L-MT-13-108, Request to Utilize an Alternative to the Requirements of 10 CFR 50.55a(g) for Implementation of a Risk-Informed, Safety-Based Lnservice Inspection Program Based on ASME Code Case N-716

Request to Utilize an Alternative to the Requirements of 10 CFR 50.55a(g) for Implementation of a Risk-Informed, Safety-Based Lnservice Inspection Program Based on ASME Code Case N-716

ML13308A390
Person / Time
Site:	Monticello
Issue date:	10/31/2013
From:	Fili K Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-13-108
Download: ML13308A390 (58)

Text

Monticello Nuclear Generating Plant Xcel Energy@ 2807 W County Road 75 Monticello, MN 55362 October 31, 2013 L-MT-13-1 08 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22

Subject:

Request to Utilize an Alternative to the Requirements of 10 CFR 50.55a(g) for Implementation of a Risk-Informed, Safety-Based lnservice Inspection Program Based on ASME Code Case N-716 In accordance with 10 CFR 50.55a(a)(3)(i), Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy requests authorization to implement a risk-informed, safety-based inservice inspection (lSI) program based on American Society of Mechanical Engineers (ASME) Code Case N-716, "Alternative Piping Classification and Examination Requirements,Section XI, Division 1 ," as documented in the enclosed 10 CFR 50.55a Request RR-003. The information provided in the enclosed request demonstrates that the proposed alternative provides an acceptable level of quality and safety.

NSPM plans to implement the proposed alternative during the fifth ten-year lSI interval and requests NRC approval by November 1, 2014.

If you have any questions or require additional information, please contact Mr. Randy Rippy at 612-330-6911.

Summary of Commitments T~'O~ commitments and no revisions to existing commitments.

Karen D. Fili Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company- Minnesota Enclosure

Document Conrol Desk Page 2 cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce

ENCLOSURE MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 REQUEST FOR AUTHORIZATION OF RISK-INFORMED/SAFETY BASED INSERVICE INSPECTION.ALTERNATIVE FOR CLASS 1 AND 2 PIPING

MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Plant Site - Unit: Monticello Nuclear Generating Plant (MNGP) t Interval- Dates: Fifth lnservice Inspection (lSI) Interval- September 1, 2012 to May 31, 2022 Requested Date for Authorization is requested by November 1, 2014.

Authorization:

ASME Code All Class 1 and 2 piping welds- Examination Categories 8-F, 8-J, C-F-1, and Components Affected: C-F-2.

The applicable Code edition and addenda is American Society of Mechanical Applicable Engineers (ASME) Boiler & Pressure Vessel Code,Section XI, Rules for Code Edition and Addenda: lnservice Inspection of Nuclear Power Plant Components, 2007 Edition through the 2008 addenda.

For the Fifth Interval lSI program Code of Record, IWB-2200, IWB-2420, Applicable IWB-2430, and IWB-2500 provide the examination requirements for Category Code 8-F and Category 8-J welds. Similarly, IWC-2200, IWC-2420, IWC-2430, and Requirements: IWC-2500 provide the examination requirements for Category C-F-1 and C-F-2 welds.

Reason for The objective of this submittal is to request the use of a risk-informed/safety Request: based (RIS_B) lSI process for the inservice inspection of Class 1 and 2 piping.

Proposed In lieu of the ASME Code requirements, Northern States Power Company-Alternative and Minnesota (NSPM) proposes to use a RIS_B process at MNGP as an Basis for Use: alternative to an ASME Section XI lSI program for Class 1 and 2 piping. The RIS_B process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI, Division 1.

Code Case N-716 is founded, in large part, on the risk-informed lSI (RI-ISI) process described in Electric Power Research Institute (EPRI) Topical Report

{TR) 112657 Rev. 8-A, Revised Risk-Informed lnservice Inspection Evaluation Procedure, December 1999 (ADAMS Accession No. ML013470102) which was previously reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC).

In general, a risk-informed program replaces the number and locations of nondestructive examination (NDE) inspections based on ASME Code,Section XI requirements with the number and locations of these inspections based on the risk-informed guidelines. These processes result in a program consistent with the concept that, by focusing inspections on the most safety-significant welds, the number of inspections can be reduced while at the same time maintaining protection of public health and safety.

NRC approved EPRI TR 112657, Rev. 8-A includes steps which, when successfully applied, satisfy the guidance provided in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment In Risk-Informed 1 of 31

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Decisions On Plant-Specific Changes to the Licensing Basis and RG 1.178, An Approach For Plant-Specific Risk-Informed Decision Making for lnservice Inspection of Piping. These steps are:

Scope definition Consequence evaluation Degradation mechanism evaluation Piping segment definition Risk categorization lnspection/NDE selection Risk impact assessment Implementation monitoring and feedback These same steps were also applied to this RIS_B process and it is concluded that this RIS_B process alternative also meets the intent and principles of Regulatory Guides 1.174 and 1.178.

In general, the methodology in Code Case N-716 replaces a detailed evaluation of the safety significance of each pipe segment required by EPRI TR 112657, Rev. B-A with a generic population of high safety-significant segments, supplemented with a rigorous flooding analysis to identify any plant-specific high safety-significant segments (Class 1, 2, 3, or Non-Class).

The flooding analysis was performed in accordance with Regulatory Guide 1.200 and ASME RA-Sb-2009, Standard for Probabilistic Risk Assessment for Nuclear Plant Applications.

By using risk-insights to focus examinations on more important locations, while meeting the intent and principles of Regulatory Guides 1.174 and 1.178, this proposed RIS_B program will continue to maintain an acceptable level of quality and safety. Additionally, all piping components, regardless of risk classification, will continue to receive ASME Code-required pressure testing, as part of the current ASME Code,Section XI program. Therefore, authorization of this alternative to the requirements of IWB-2200, IWB-2420, IWB-2430, and IWB-2500 (Examination Categories B-F and B-J) and IWC-2200, IWC-2420, IWC-2430, and IWC-2500 (Examination Categories C-F-1 and C-F-2) is requested in accordance with 10 CFR 50.55a(a)(3)(i). A Monticello specific template is attached that mirrors previous RIS_B submittals to the NRC.

All other ASME Code,Section XI requirements for which alternatives were not specifically requested in this 10 CFR 50.55a Request remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.

Duration of The proposed alternative will be used for the Fifth Interval of the MNGP Proposed lnservice Inspection Program that began on September 1, 2012 and is Alternative: scheduled to end May 31, 2022.

Precedents: Similar alternatives have been authorized for Vogtle Electric Generating Plant, Donald C. Cook 1 and 2, Grand Gulf Nuclear Station, Waterford-3 and North Anna 1 & 2.

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Vogtle Electric Generating Plant Safety Evaluation - See ADAMS Accession No. ML100610470.

D. C. Cook Safety Evaluation - See ADAMS Accession No. ML072620553.

References:

Grand Gulf Nuclear Station Safety Evaluation - See ADAMS Accession No. ML072430005.

Waterford-3 Safety Evaluation -See ADAMS Accession No. ML080980120.

Status: Awaiting NRC authorization.

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TEMPLATE SUBMITTAL APPLICATION OF ASME CODE CASE N-716 RISK-INFORMED/SAFETY-BASED (RIS_B)

INSERVICE INSPECTION PROGRAM PLAN 4 of 31

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Technical Acronyms/Definitions Used in the Template AC Alternating Current AS Accident Sequence Analysis ASEP Accident Sequence Evaluation Program ASME American Society of Mechanical Engineers BER Break Exclusion Region CAFTA Computer-Aided Fault Tree Analysis cc PRA abbreviation for Capacity Category cc Crevice Corrosion CCDP Conditional Core Damage Probability CCF Common Cause Failure ccw Component Cooling Water CDF Core Damage Frequency CIV Containment Isolation Valve Class 2 LSS Class 2 Pipe Break in LSS Piping CLERP Conditional Large Early Release Probability cs Core Spray DA Data analysis DC Direct Current*

DM Degradation Mechanism E-C Erosion-Corrosion ECSCC External Chloride Stress Corrosion Cracking EOOS Equipment Out of Service FAC Flow-Accelerated Corrosion F&O Facts and Observations FT Fault tree FW Feedwater HELB High Energy Line Break (synonymous with BER)

HEP Human Error Probability HFE Human Failure Event HR Human Reliability HRA Human Reliability Analysis HSS High Safety-Significant IE Initiating Events Analysis IF Internal Flooding IF IV Inside First Isolation Valve IGSSC lntergranular Stress Corrosion Cracking ILOCA lsolable Loss of Coolant Accident IPE Individual Plant Evaluation LE LERF Analysis LERF Large Early Release Frequency LOCA Loss of Coolant Accident 5 of 31

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Technical Acronyms/Definitions Used in the Template (Continued)

LOSP Loss of Off-Site Power LSS Low Safety-Significant MAAP Modular Accident Analysis Program MIC Microbiologically-Influenced Corrosion MOV Motor Operated Valve MS Main Steam MU Model Update NDE Nondestructive Examination NNS Non-Nuclear Safety NPS Nominal Pipe Size PBF Pressure Boundary Failure PIT Pitting PLOCA Potential Loss of Coolant Accident POD Probability of Detection PRA Probabilistic Risk Assessment PSA Probabilistic Safety Assessment PWSCC Primary Water SCC QU Quantification RCPB Reactor Coolant Pressure Boundary RG Regulatory Guide RHR Residual Heat Removal RI-BER Risk-Informed Break Exclusion Region RI-ISI Risk-Informed lnservice Inspection RIS_B Risk-Informed/Safety Based lnservice Inspection RM Risk Management RPV Reactor Pressure Vessel SBO Station Blackout sc Success Criteria soc Shutdown Cooling SLB Steam Line Break sse Systems, Structures, and Components SR Supporting Requirements SURF Surface SXI Section XI SY Systems Analysis TASCS Thermal Stratification, Cycling, and Striping TGSCC Transgranular Stress Corrosion Cracking TR Technical Report TT Thermal Transients VOL Volumetric 6 of 31

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Table of Contents

1. Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 PRA Quality

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

3. Risk-Informed/Safety-Based lSI Process 3.1 Safety Significance Determination 3.2 Failure Potential Assessment 3.3 Element and NDE Selection 3.3.1 Current Examinations 3.3.2 Successive Examinations 3.3.3 Scope Expansion 3.3.4 Program Relief Requests 3.4 Risk Impact Assessment 3.4.1 Quantitative Analysis 3.4.2 Defense-in-Depth 3.5 Implementation 3.6 Feedback (Monitoring)

4. Proposed lSI Plan Change

5. References/Documentation Attachment A- Monticello PRA Quality Review 7 of 31

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1. INTRODUCTION Monticello Nuclear Generating Plant (Monticello) is currently in Period 1 of the Fifth lnservice Inspection (lSI) Interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section XI Code. Monticello plans to implement a risk-informed/safety-based inservice inspection (RIS_B) program for the Fifth lSI Interval. The Fifth lSI Interval began September 1, 2012.

The ASME Section XI Code of record for the Fifth lSI Interval is the 2007 Edition through the 2008 Addenda for Examination Category B-F, B-J, C-F-1, and C-F-2 Class 1 and 2 piping components.

The RIS_B process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1, which is founded in large part on the RI-ISI process as described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk-Informed lnservice 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 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis, and Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decision making lnservice Inspection ofPiping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.

1.2 Probabilistic Safety Assessment (PSA) Quality The methodology in Code Case N-716 provides for examination of a generic population of high safety significant (HSS) segments, supplemented with a rigorous flooding analysis to identify if any plant-specific HSS segments need to be added.

Satisfying the requirement for the plant-specific analysis requires confidence that the flooding PRA is capable of successfully identifying any significant flooding contributors that are not identified in the generic population.

The Monticello PRA is based on a detailed model of the plant that was originally developed from the Individual Plant Examination (IPE) and Individual Plant

. Examination for External Events (IPEEE) projects. The original model was reviewed by the NRC and underwent Boiling Water Reactor Owner's Group (BWROG) certification in 1997. NRC reviews of the IPE and IPEEE are documented in the NRC Staff Evaluations on IPE in May 1994 (TAC No M74435) and IPEEE dated April2000 (TAC No M83644). The NRC concluded that the Monticello process is capable of identifying the most likely severe accidents and no significant impacts on the PRA were identified.

The Monticello PRA has since been upgraded. It is a Level 2, at-power model. A major upgrade of the internal events model to meet the guidance of RG 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," as well as the American Society of 8 of 31

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Mechanical Engineers and American National Standard (ASME/ANS) PRA Standard RA-Sa-2009 was completed in 2013. A formal, 8WROG-sponsored industry peer review of the upgraded internal events model was completed in April 2013. The peer review utilized the process described in Nuclear Energy Institute document NEI 05-04, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard," January 2005, and the ASME/ANS PRA Standard. This review to ASME Capability Category II requirements confirmed that the PRA model met the requirements of RG 1.200, Revision 1, and ASME/ANS RA-Sa-2009. There were twenty two Finding Level F&Os identified by the peer review team. Attachment A contains a summary of these findings and their resolution. To date, all of the Peer Review Findings have been resolved and inserted into the PRA model or dispositioned for this submittal.

The original RI-ISI evaluation concluded external events are not likely to impact the consequence ranking. This position is further supported by Section 2 of EPRI Report 1021467, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs" which

. concludes that quantification of these events will not change the conclusions derived from the RI-ISI process. As a result, there is no need to further consider these events.

The PRA model used for development of the RIS_8 evaluation accounts for conditions applicable to the Extended Power Uprate which Monticello intends to implement following NRC approval.

2. PROPOSED ALTERNATIVE TO CURRENT lSI PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories 8-F, 8-J, C-F-1, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components.

The alternative RIS_8 Program for piping is described in Code Case N-716. The RIS_8 Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories 8-F, 8-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety.

Other non-related portions of the ASME Section XI Code will be unaffected.

2.2 Augmented Programs The impact of the RIS_8 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_8 application scope (e.g., Class 1 and 2 piping).

• The plant augmented inspection program for high-energy line breaks outside containment has not been revised by this application. A separate evaluation and program in accordance with the risk-informed break exclusion region methodology (RI-8ER) described in EPRI Report 1006937, Extension of EPRI Risk Informed IS/

Methodology to Break Exclusion Region Programs has not yet been established.

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• The augmented inspection program for intergranular stress corrosion cracking (IGSCC) as addressed in NRC Generic Letter 88-01, NUREG-0313, Rev. 2 and BWRVIP-75A, have been resolved by Monticello's pip_e replacement program wherein all susceptible material was replaced with resistant material. All welds are therefore classified as IGSCC Category "A". In accordance with EPRI TR-112657, piping welds identified as Category "A" are considered resistant to IGSCC, and as such are assigned a low failure potential provided no other damage mechanisms are present. Examination criteria for these welds will be in accordance with the RIS_B process.

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

3. RISK-INFORMED/SAFETY-BASED lSI 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 (see Section 3.1)

• Failure Potential Assessment (see Section 3.2)

• Element and NDE Selection (see Section 3.3)

• Risk Impact Assessment (see Section 3.4)

• Implementation Program (see Section 3.5)

• Feedback Loop (see Section 3.6)

Each of these six steps is discussed below:

3.1 Safety Significance Determination The systems assessed in the RIS_B Program are provided in Table 3.1. The piping and instrumentation diagrams and additional plant information, including the existing plant lSI Program were used to define the piping system boundaries. Per Code Case N-716 requirements, piping welds are assigned safety-significance categories, which are then used to determine the examination treatment requirements. High safety-significant (HSS) welds are determined in accordance with the requirements below.

Low safety-significant (LSS) welds include all other Class 2, 3, or Non-Class welds.

(1) Class 1 portions of the reactor coolant pressure boundary (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:

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(a) As part of the RCPB from the reactor pressure vessel (RPV) to the second isolatbn 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 [greater than 4 inch nominal pipe size (NPS)] of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve. This is not applicable to Monticello.

(4) Piping within the break exclusion region (BER) greater than 4 inch NPS for high-energy piping systems as defined by the Owner. Per Code Case N-716, this may include Class 3 or Non-Class piping.

(5) Any piping segment whose contribution to Core Damage Frequency (CDF) is greater than 1E-06 [and per NRC feedback on the Grand Gulf and D. C. Cook RIS_B applications 1E-07 for Large Early Release Frequency (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. No piping segments with a contribution to CDF greater than 1E-06 (1 E-07 for LERF) were identified.

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 NRC approved 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.

As previously approved for Monticello during last Interval, a deviation to the EPRI RIS_B methodology has been implemented in the failure potential assessment. Table 3-16 of EPRI TR-112657 contains the following 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 11 of 31

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4. The potential exists for two phase (steam/water) flow; or

5. The potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow; AND AND Richardson Number> 4 (this value predicts the potential buoyancy of a stratified flow)

These criteria, based on meeting a high cycle fatigue endurance limit with the actual L1 T assumed equal to the greatest potential L1T 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 is a swirling vertical flow structure in a branch line induced by high velocity flow in the connected piping. It 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 L1Ts can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASCS is considered for this configuration.

For upward sloping branch lines connected to the hot fluid source that turn horizontal or in horizontal branch lines, natural convective effects combined with effects of turbulent penetration will tend to keep the line filled with hot water. If there is in-leakage of cold water, a cold stratified layer of water may be formed, and significant top-to-bottom L1Ts may occur in the horizontal portion of the branch line. Interaction with the swirling motion from turbulent penetration may cause a periodic axial motion of the cold layer. Therefore, TASCS is considered for these configurations.

For similar upward sloping branch lines, if there is no potential for in-leakage, this will result in a well-mixed fluid condition where significant top-to-bottom L1Ts will not occur. Therefore, TASCS is not considered for these in-leakage 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.

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• 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. Consideration of cycle severity was used in previous NRC approved RIS_B program submittals for D. C. Cook, Grand Gulf Nuclear Station, Waterford-3, and the Vogtle Electric Generating Plant as well as Monticello during the past Interval. The methodology used in the Monticello RIS_B application for assessing TASCS potential conforms to these updated criteria.

3.3 Element and NDE Selection Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RIS_B applications provided 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.

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(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 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 RPV) and the RPV.

(4) A minimum of 10% of the welds in that portion of the RCPB that lies outside containment shall be selected.

(5) A minimum of 10% of the welds within the break exclusion region (BER) shall be selected. Currently, there are 42 welds at Monticello in the BER program. These BER welds consist of Class 1 welds in the core spray, HPCI, main steam, RCIC RHR and RWCU systems. A RI-BER program has not been implemented for these welds; if is implemented in the future, it will also require more than 10% of the population to be examined.

In contrast to a number of traditional 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% of the HSS welds be chosen. A brief summary of the number of welds and the number selected 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. Only those RIS_B inspection locations that receive a volumetric examination are included.

Class 1 Welds( 1J Class 2 Welds( 2J All Piping Welds( 3J Total Selected Total Selected Total Selected 811 87 887 0 1698 87 Notes:

(1) Includes all Category B-F and B-J locations. All Class 1 piping weld locations are HSS.

(2) Includes all Category C-F-1 and C-F-2 locations. Of the Class 2 piping weld locations, none were identified as HSS.

(3) Regardless of safety significance, Class 1, 2, and 3 ASME Section XI in-scope piping components will continue to be pressure tested as required by the ASME Section XI Program. VT-2 visual examinations are scheduled in accordance with the pressure test program that remains unaffected by the RIS_B Program.

3.3.1 Current Examinations The Fourth Interval program concluded on August 31, 2012, and implemented the previous NRC authorized alternative RI-ISI methodology using EPRI-TR 112657, Rev. B-A for Class 1 and 2 piping welds (see ADAMS Accession No. ML021490050). Monticello is currently in Period 1 of the Fifth lSI Interval, which began on September 1, 2012, and has not yet implemented 14 of 31

MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) risk-informed examinations for the Fifth Interval. Monticello intends to implement the RIS_B program for Class 1 and 2 piping welds for the Fifth Interval upon receiving NRC authorization of the proposed alternative.

3.3.2 Successive Examinations If indications are detected during RIS_B ultrasonic examinations, they will be evaluated per IWB-3514 (Class 1) or IWC-3514 (Class 2) to determine their acceptability. Any unacceptable flaw will be evaluated per the requirements of ASME Code Section XI, IWB-3600 or IWC-3600, as appropriate. As part of this evaluation, the degradation mechanism that is responsible for the flaw will be determined and accounted for in the evaluation. If the flaw is acceptable for continued service, successive examinations will be scheduled per Section 6 of Code Case N-716. If the flaw is found unacceptable for continued operation, it will be repaired in accordance with IWA-4000, applicable ASME Section XI Code Cases, or NRC approved alternatives. The IWB-3600 analytical evaluation will be submitted to the NRC. Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section XI.

3.3.3 Scope Expansion If the nature and type of the flaw is service-induced, then welds subject to the same type of postulated degradation mechanism will be selected and examined per Section 6 of Code Case N-716. 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. The need for extensive root cause analysis beyond that required for the IWB-3600 analytical evaluation will be dependent on practical considerations (i.e., the practicality of performing additional NDE or removing the flaw for further evaluation during the outage).

3.3.4 Program Relief Requests Consistent with previously approved RIS_B submittals, Monticello 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). Monticello selected welds which are expected to provide the required examination volume coverage, but actual coverage will not be determined until the examination is performed. As such, the effect on risk, if any, will not be known until the examinations are performed. Relief requests for those cases where greater 15 of 31

MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) than 90% coverage is not obtained will be submitted per the requirements of 10 CFR 50.55a(g)(5)(iv).

No Monticello relief requests are being withdrawn due to the RIS_B application.

3.4 ,Risk Impact Assessment The RIS_B Program development has been conducted in accordance with Regulatory Guide 1.174 and the requirements of Code Case N-716, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.

This evaluation categorized segments as high safety significant or low safety significant in accordance with Code Case N-716, and then determined what inspection changes were proposed for each system. The changes included changing the number and location of inspections, and in many cases improving the effectiveness of the inspection to account for the findings of the RIS_B degradation mechanism assessment. For example, examinations of locations subject to thermal fatigue will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the inspection process.

3.4.1 Quantitative Analysis Code Case N-716 has adopted the NRC approved EPRI TR-112657 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 Regulatory Guides 1.174 and 1.178. Section 3.7.2 of EPRI TR-112657 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, Conditional Core Damage Probability (CCDP)/Conditional Large Early Release Probability (CLERP) values of 1E-4/1 E-5 were conservatively used. The rationale for using these values is that the change-in-risk evaluation process of Code Case N-716 is similar to that of the EPRI risk-informed lSI (RI-ISI) methodology. As such, the goal is to determine CCDPs/CLERPs threshold values. For example, the threshold values between High and Medium consequence categories is 1E-4 (CCDP)/1 E-5 (CLERP) and between Medium and Low consequence categories are 1E-6 (CCDP)/1 E-7 (CLERP) from the EPRI RI-ISI Risk Matrix. Using these threshold values streamlines the change-in-risk evaluation as well as stabilizes the update process. For example, if a CCDP changes from 1E-5 to 3E-5 due to an update, it will remain below the 1E-4 threshold value; the change-in-risk evaluation would not require updating.

The updated internal flooding PRA was also reviewed to ensure that there is no LSS Class 2 piping with a CCDP/CLERP greater than 1E-4/1 E-5. This review identified some piping in the CS, RHR, RCIC, and HPCI Torus suction lines with a CCDP *greater than 1E-4. As a result, all LSS welds in these systems are conservatively assigned CCDP/CLERP equal to 2E-3/2E-4.

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With respect to assigning failure potentials for LSS piping, the criteria are defined in Table 3 of Code Case N-716. That is, those locations identified as susceptible to FAC are assigned a high failure potential. Those locations susceptible to thermal fatigue, erosion-cavitation, corrosion, or stress corrosion cracking are assigned a medium failure potential, unless they have an identified potential for water hammer loads. In such cases, they will be assigned a high failure potential. Finally, 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 that verified that' the LSS piping was not susceptible to water hammer. LSS piping may be susceptible to FAC; however, the examination for FAC is performed per the FAC program. 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) for use in the change-in-risk assessment. Experience with previous industry RIS_B applications shows this to be conservative.

Monticello 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-112657. 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 in the table below. Consistent with the EPRI methodology, the upper bound for all break locations that fall within the high consequence rank range was based on the highest CCDP value obtained (e.g., Medium LOCA CCDP bounds the large and small LOCA CCDPs).

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 X 0 and is expected to have a value less than 1E-08. Piping locations identified as medium failure potential have a likelihood of 20x0 . These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-112657. In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to an increased POD from application of the RIS_B approach.

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Estimated Consequence Upper I Lower Bound Break Location Description of Affected Piping CCDP I CLERP Rank CCDP CLERP LOCA 4E-04 I 4E-05 (U) 4E-04 (U) 4E-05 The highest CCDP for Medium LOCA (IE_MLOCA) was used HIGH Unisolable RCPB piping of all sizes (L) IE-04 (L) IE-05 (0.1 margin can be used for CLERP)

ILOCA 2E-06 I 2E-07 (U) IE-04 (U) IE-05 Piping between 1st and 2nd normally open isolation valve Calculated based on IE_MLOCA CCDP of 4E-4 and valve fail to MEDIUM (L) IE-06 (L) IE-07 inside containment (RWCU, RCIC, HPCI, MS, FW) close probability of3E-3 (0.1 margin can be used for CLERP)

PLOCA IE-06 I IE-07 (U) IE-04 (U) IE-05 Piping between 1st and 2nd normally closed isolation valve Calculated based on IE_MLOCA CCDP of 4E-4 and check valve MEDIUM (L) IE-06 (L) IE-07 inside containment (CS, RHR, SLC) fail open probability of2E-3 (0.1 margin can be used for CLERP)

ILOCA-OC 4E-05 I 4E-05 Piping between penetration and outside containment (U) 4E-04 (U) 4E-05 Isolable LOCA outside containment CCDP based on bounding HIGH isolation valve with normally open isolation valve inside (L) IE-04 (L) IE-05 value for lEA P280 (MS break) in the PRA (CCDP = CLERP) containment (RWCU, RCIC, HPCI, MS, FW)

PLOCA-OC 4E-05 I 4E-05 Piping between penetration and outside containment (U) 4E-04 (U) 4E-05 Potential LOCA outside containment CCDP based on bounding HIGH isolation valve with normally closed isolation valve inside (L) IE-04 (L) IE-05 value for lEA P280 (MS break) in the PRA (CCDP = CLERP) containment (CS, RHR, SLC)

Class 2 LSS IE-04 I IE-05 (U) IE-04 (U) IE-05 All Class 2 system piping designated as low safety MEDIUM Estimated based on upper bound for Medium Consequence (L) IE-06 (L) IE-07 significant except for piping connected to the Torus Class 2 Torus 2E-03 I 2E-04 Class 2 piping connected to the Torus and designated as low Piping connected to the Torus (IEF_R896-TW, IEF_RHRI-TW (U) 2E-03 (U) 2E-04 HIGH safety significant (CS, RHR, RCIC, HPCI Torus suction and IEF_RHR2-TW) although low frequency and low risk has a (L) IE-04 (L) IE-05 lines).

CCDP -2E-3 and CLERP <2E-4 (0.1 margin used for CLERP) - - -

PLOCA =potential LOCA (I closed valve)

ILOCA = isolable LOCA (I open valve)

Note: The PRA does not explicitly model potential and isolable LOCA events, because such events are subsumed by the LOCA initiators in the PRA. That is, the frequency of a LOCA in this limited piping downstream of the first RCPB isolation valve times the probability that the valve fails is a small contributor to the total LOCA frequency.

The N-716 methodology must evaluate these segments individually; thus, it is necessary to estimate their contribution. This is estimated by taking the LOCA CCDP and multiplying it by the valve failure probability 18 of 31

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Table 3.4 presents a summary of the RIS_B Program versus the Third lSI Interval (1986 Edition of ASME Section XI) program requirements on a "per system" basis. The presence of FAC was adjusted for in the quantitative analysis by excluding its impact on the failure potential rank. The exclusion of the impact of FAC on the failure potential ranks, and therefore in the determination of the change-in-risk, was performed because FAC is a damage mechanism managed by a separate, independent plant augmented inspection program. The RIS_B Program credits and relies upon this plant augmented inspection program to manage this damage mechanism. The plant FAC program will continue to determine where and when examinations shall be performed. Hence, since the number of FAC examination locations remains the same "before" and "after" (the implementation of the RIS_B program) and no delta exists, there is no need to include the impact of FAC 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 RIS_B Program, and that the acceptance criteria of Regulatory Guide 1.174 and Code Case N-716 are satisfied.

With POD Credit Without POD Credit System Delta CDF DeltaLERF Delta CDF DeltaLERF CS - Core Spray 9.87£-10 9.96£-11 9.87£-10 9.96E-11 FW- Feedwater 5.04E-11 -3.60£-12 4.60£-10 4.60£-11 HPCI - High Pressure Coolant Injection 1.72£-10 1.74E-11 1.72E-10 1.74E-11 INST- Instrument Nozzles 4.00£-12 4.00E-13 4.00E-12 4.00E-13 MS - Main Steam 5.04E-11 7.20£-12 5.04£-11 7.20£-12 PCAC- SGTS/Vent l.OOE-11 l.OOE-12 l.OOE-11 l.OOE-12 RBCCW - RB Closed Loop Cooling O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RCIC - Reactor Isolation Core Cooling 2:06E-10 2.04E-11 2.06£-10 2.04£-11 RHR - Residual Heat Removal 2.89E-10 1.20E-11 6.77E-10 6.80£-11 RR - Reactor Recirculation 4.80E-11 4.80E-12 4.80£-11 4.80£-12 RWCU- Reactor Water Clean-up 1.02£-11 1.20E-12 1.02E-11 1.20E-12 SDV- Scram Discharge Volume 3.10E-10 3.10£-11 3.10£-10 3.10£-11 SLC - Standby Liquid Control -4.00£-12 -4.00£-13 -4.00£-12 -4.00£-13 Total 2.13E-09 1.91E-10 2.93E-09 2.97E-10 As shown in Table 3.4, new RIS_B locations were selected such that the RIS_B selections exceed the Section XI selections for certain categories (Delta column has a positive number). To show that the use of a conservative upper bound CCDP/CLERP does not result in an optimistic calculation with regard to meeting the acceptance criteria, a conservative sensitivity was conducted where the RIS_B selections were set equal to the Section XI selections (Delta changed from positive number to zero). The acceptance criteria are met when the number of RIS_B selections is not allowed to exceed Section XI.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3.4.2 Defense-in-Depth The intent of the inspections mandated by 10 CFR 50.55a 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 s.electing inspection locations is based upon terminal end locations, structural discontinuities, and stress analysis results. As depicted in ASME White Paper 92-01-01 Rev. 1, Evaluation of lnservice Inspection Requirements for Class 1, Category 8-J Pressure Retaining Welds, this methodology 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; that is, a determination of each location's susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients assure defense-in-depth is maintained. First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased. Secondly, 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 piping on a plant-specific basis that has a contribution to CDF of greater than 1E-06 (or 1E-07 for LERF) be included in the scope of the application. Monticello did not identify any such piping.

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.

3.5 Implementation Upon authorization of the RIS_B Program, procedures that comply with the guidelines described in Code Case N-716 will be used to implement and monitor the program.

The new program will be implemented during the Fifth lSI Interval. No changes to the Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section XI program implementing procedures will be retained and modified to address the RIS_B process, as appropriate.

3.6 Feedback (Monitoring)

The RIS_B Program is a living program that is required to be monitored continuously for changes that could impact the basis for which welds are selected for examination.

Monitoring encompasses numerous facets, including the review of changes to the plant configuration, changes to operations that could affect the degradation 20 of 31

MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) assessment, a review of NDE results, a review of site failure information from the corrective action program, and a review of industry failure information from industry operating experience (OE), and other OE such as vendor issued communications.

Also included is a review of PRA changes for their impact on the RIS_B program.

These reviews provide a feedback loop such that new relevant information is obtained that will ensure that the appropriate identification of HSS piping locations selected for examination is maintained. As a minimum, this review will be conducted on an ASME period basis. In addition, more frequent adjustment may be required as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant-specific feedback.

If an adverse condition, such as an unacceptable flaw is detected during examinations, the adverse condition will be addressed by the corrective action program and procedures. The following are appropriate actions to be taken:

A. Identify (Examination results conclude there is an unacceptable flaw).

B. Characterize (Determine if regulatory reporting is required and assess if an immediate safety or operation impact exists).

C. Evaluate (Determine the cause and extent of the condition identified and develop a corrective action plan or plans).

D. Decide (make a decision to implement the corrective action plan).

E. Implement (complete the work necessary to correct the problem and prevent recurrence).

F. Monitor (through the audit process ensure that the RIS_B program has been updated based on the completed corrective action).

G. Trend (Identify conditions that are significant based on accumulation of similar issues).

For preservice examinations, Monticello will follow the rules contained in Section 3.0 of N-716. Welds classified HSS require a preservice inspection. The examination volumes, techniques, and procedures shall be in accordance with Table 1 of N-716.

Welds classified as LSS do not require preservice inspection.

4. PROPOSED lSI PLAN CHANGE Monticello is currently in the 1st Period of the Fifth lSI Interval and, upon receiving NRC authorization, plans to implement this RIS_B application for the Fifth Interval which began September 1, 2012.

In anticipation of NRC authorization of this RIS_B submittal, Monticello selected welds for examination in the 1st Period, using the traditional ASME Section XI examination methodology, and that also meet the examination requirements of Table 1 of Code Case N-716. After authorization of the RIS_B submittal, those welds in the RIS_B scope that were examined during the 1st Period that also met Table 1 requirements may be credited toward the RIS_B requirements for the 1st Period.

A comparison between the RIS_B Program and the 1986 Edition of Section XI program requirements for in-scope piping is provided in Table 4.

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5. REFERENCES/DOCUMENTATION EPRI Report 1006937, Extension of EPRI Risk Informed IS/ Methodology to Break Exclusion Region Programs.

EPRI TR-112657, Revised Risk-Informed lnservice Inspection Evaluation Procedure, Rev.

B-A.

ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1.

Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis.

Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking lnservice Inspection of Piping.

Regulatory Guide 1.200, Rev 2 An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities.

US NRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for Alternative GG-ISI-002-Implement Risk-Informed lSI based on ASME Code Case N-716, dated September 21, 2007.

USNRC Safety Evaluation for DC Cook Nuclear Plant, Units 1 and 2, Risk-Informed Safety-Based lSI program for Class 1 and 2 Piping Welds, dated September 28, 2007.

EPRI Report 1021467 Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs.

USNRC Safety Evaluation for Monticello Nuclear Generating Plant, Risk-Informed Inspection Plan, dated July 24, 2002, ADAMS Accession No. ML021490050.

Supporting Onsite Documentation Sl Calculation No. 1000515.301, "Degradation Mechanism Evaluation for Monticello,"

Revision 1.

Sl Calculation No. 1000515.302, "N-716 Evaluation for Monticello," Revision 0.

PRA-CALC-13-002, PRA Input to RI-ISI LAR, Revision 0, October 1, 2013.

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Table 3.1 Code Case N-716 Safety Significance Determination Safety Weld N-716 Safety Significance Determination System Significance Count RCPB SDC PWR:FW BER CDF> lE-6 High Low 24 ./ ./

cs 6 ./ ./ ./

162 ./

70 ./ ./

FW 8 ./

15 ./ ./

HPCI 2 ./ ./ ./

141 ./

INST 25 ./ ./

259 ./ ./

MS 24 ./ ./ ./

PCAC 52 ./

RBCCW 15 ./

13 ./ ./

RCIC 2 ./ ./ ./

51 ./

31 ./ ./

6 ./ ./ ./ ./

RHR 79 ./ ./ ./

417 ./

RR 194 ./ ./

24 ./ ./

RWCU 2 ./ ./ ./

SDV 41 ./

SLC 35 ./ ./

690 ./ ./

36 ./ ./ ./

Summary Results all 6 ./ ./ ./ ./

Systems 79 ./ ./ ./

887 ./

Totals 1698 CS = Core Spray FW =Main Feedwater HPCI =High Pressure Coolant Injection INST =Instrument connections to RPV MS = Main Steam PCAC = Primary Containment Cooling RBCCW = Reactor Building Closed Cooling Water RCIC = Reactor Core Isolation Cooling RR = Reactor Coolant RHR =Residual Heat Removal RWCU = Reactor Water Cleanup SDV = Scram Discharge Volume SLC =Standby Liquid Control 23 of 31

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Table 3.2 Thermal Localized Flow Fati!;ue Stress Corrosion Cracking Corrosion Sensitive System( 1J TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT cc E-C FAC cs FW ../ ../ ../

HPCI INST MS PCAC RBCCW RCIC RHR ../

RR RWCU ../

SDV SLC _L___ - L__ - --- --

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 PCAC, RBCCW and SDV in their entirety, as well as portions of the CS, FW, HPCI, RCIC and RHR systems.

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Table 3.3: Code Case N716 Selections Weld Count N716 Selection Considerations System RCPB RCPB Selections HSS LSS DMs RCPB BER (IFIV) (OC) cs 14 None o/ o/ 2 cs 10 None o/ 0 cs 6 None o/ o/ o/ 1 cs 162 None 0 FW I9 TT o/ o/ 0 FW I2 TASCS o/ o/ 3 FW 2 TASCS o/ 0 FW 2 TASCS o/ o/ I TASCS, o/ o/

FW I2 2 TT TASCS, o/

FW 2 0 TT TASCS, o/ o/

FW 2 I TT FW 19 None o/ o/ 0 FW 8 None 0 HPCI 5 None o/ o/ 2 HPCI 10 None o/ 0 HPCI 2 None o/ o/ o/ 1 HPCI I41 None 0 INST 25 None o/ o/ 3 MS 24I None o/ o/ 20 MS I8 None o/ 0 MS 24 None o/ o/ o/ 9 PCAC 52 None 0 RBCCW 15 None 0 RCIC 3 None o/ o/ 2 RCIC IO None o/ 0 RCIC 2 None o/ o/ o/ I RCIC 51 None 0 RHR 34 TT o/ o/ 8 RHR I2 TT o/ 0 RHR 4 TT o/ o/ o/ 4 RHR 57 None o/ o/ 0 RHR 7 None o/ 0 RHR 2 None o/ o/ o/ 0 RHR 4I7 None 0 RR 182 None o/ o/ 14 RR I2 None o/ 6 RWCU 22 None o/ o/ 2 RWCU 2 None o/ 0 RWCU 2 None o/ o/ o/ I SDV 41 None 0 SLC 24 None o/ o/ 3 SLC 4 None o/ 0 SLC 7 None o/ o/ I 25 of 31

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Weld Count N716 Selection Considerations System RCPB RCPB Selections HSS LSS DMs RCPB BER (IFIV) (OC) 53 TT o/ o/ 8 12 TT o/ 0 4 TT o/ o/ o/ 4 12 TASCS o/ o/ 3 2 TASCS o/ 0 2 TASCS o/ o/ 1 Summary TASCS, 12 o/ o/ 2 Results TT All TASCS, 2 o/ 0 Systems TT TASCS, o/ o/

2 1 TT 592 None o/ o/ 48 73 None o/ 6 38 None o/ o/ o/ 13 7 None o/ o/ 1 Totals 811 887 87 Note: Systems are described in Table 3.1 26 of 31

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Table 3.4 Risk Impact Analysis Results Safety Break Failure Potential Inspections CDFimpact LERFimpact System Significance Location DMs Rank SXI RISB Delta w/POD w/oPOD w/POD w/oPOD cs High LOCA None Low 5 2 -3 6.00E-12 6.00E-12 6.00E-13 6.00E-13 cs High PLOCA None Low 2 0 -2 l.OOE-12 l.OOE-12 l.OOE-13 ' l.OOE-13 cs High PLOCA-OC None Low 6 I -5 l.OOE-11 l.OOE-11 l.OOE-12 l.OOE-12 cs Low Class 2 LSS Assume Medium 18 0 -18 1.80E-10 1.80E-IO 1.80E-Il 1.80E-ll cs Low Class 2 Torus Assume Medium 4 0 -4 8.00E-10 8.00E-IO 8.00E-11 8.00E-ll CS Total 9.97E-10 9.97E-10 9.97E-11 9.97E-11 FW High LOCA TT Medium 7 0 -7 1.68E-IO 2.80E-IO 1.68E-ll 2.80E-ll FW High LOCA TASCS Medium 7 3 -4 -4.80E-Il 1.60E-IO -4.80E-12 1.60E-11 FW High ILOCA TASCS Medium 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO FW High ILOCA-OC TASCS Medium I I 0 -4.80E-ll O.OOE+OO -4.80E-12 O.OOE+OO FW High LOCA TASCS, TT Medium I 2 I -1.20E-IO -4.00E-Il -1.20E-ll -4.00E-12 FW High ILOCA TASCS, TT Medium 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO FW High ILOCA-OC TASCS, TT Medium I I 0 -4.80E-Il O.OOE+OO -4.80E-12 O.OOE+OO FW High LOCA None Low 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO FW Low Class 2 LSS Assume Medium 6 0 -6 6.00E-ll 6.00E-ll 6.00E-12 6.00E-12 FWTotal -3.60E-11 4.60E-10 -3.60E-12 4.60E-11 HPCI High LOCA None Low 3 2 -1 2.00E-12 2.00E-12 2.00E-13 2.00E-13 HPCI High ILOCA None Low I 0 -1 5.00E-13 5.00E-13 5.00E-14 5.00E-14 HPCI High ILOCA-OC None Low 2 I -1 2.00E-12 2.00E-12 2.00E-13 2.00E-13 HPCI Low Class2 LSS Assume Medium 17 0 -17 1.70E-IO 1.70E-IO 1.70E-ll 1.70E-ll HPCI Low Class 2 Torus Assume Medium 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO HPCITotal 1.75E-10 1.75E-10 1.75E-11 1.75E-11 INST Total High LOCA None Low 2 0 -2 4.00E-12 4.00E-12 4.00E-13 4.00E-13 MS High LOCA None Low 35 11 -24 4.80E-11 4.80E-ll 4.80E-12 4.80E-12 MS High ILOCA/PLOCA None Low 2 0 -2 l.OOE-12 l.OOE-12 l.OOE-13 l.OOE-13 MS High ILOCA-OC None Low 21 9 -12 2.40E-ll 2.40E-11 2.40E-12 2.40E-12 MS Total 7.30E-11 7.30E-11 7.30E-12 7.30E-12 PCAC Total Low . Class2 LSS Assume Medium I 0 -1 l.OOE-11 l.OOE-11 l.OOE-12 l.OOE-12 RBCCWTotal Low Class 2 LSS Assume Medium 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RCIC High LOCA None Low 0 2 2 -4.00E-12 -4.00E-12 -4.00E-13 -4.00E-13 RCIC High ILOCA None Low 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RCIC High ILOCA-OC None Low 0 I ---

I -

-2.00E-12 -2.00E-12 -2.00E-13 -2.00E-13 27 of 31

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Safety Break Failure Potential Inspections CDFimpact LERFimpact System Significance Location DMs Rank SXI RISB Delta w/POD w/oPOD w/POD w/oPOD RCIC Low Class 2 LSS Assume Medium I 0 -I l.OOE-11 I.OOE-11 I.OOE-12 l.OOE-I2 RCIC Low Class 2 Torus Assume Medium I 0 -I 2.00E-IO 2.00E-IO 2.00E-11 2.00E-11 RCIC Total 2.04E-10 2.04E-10 2.04E-11 2.04E-11 RHR High LOCA TT Medium 7 8 I -4.08E-10 -4.00E-Il -4.08E-11 -4.00E-12 RHR High PLOCA TT Medium 4 0 -4 2.40E-11 4.00E-11 2.40E-12 4.00E-I2 RHR High PLOCA-OC TT Medium 4 4 0 -1.92E-10 O.OOE+OO -1.92E-11 O.OOE+OO RHR High LOCA None Low 13 0 -13 2.60E-11 2.60E-11 2.60E-12 2.60E-12 RHR High PLOCA None Low 4 0 -4 2.00E-12 2.00E-12 2.00E-13 2.00E-13 RHR High PLOCA-OC None Low 2 0 -2 4.00E-12 4.00E-12 4.00E-13 4.00E-13 RHR Low Class 2 LSS Assume Medium 29 0 -29 2.90E-10 2.90E-IO 2.90E-11 2.90E-11 RHR Low Class 2 Torus Assume Medium 2 0 -2 4.00E-10 4.00E-IO 4.00E-ll 4.00E-11 RHR Total 1.46E-10 7.22E-10 1.46E-11 7.22E-11 RR High LOCA None Low 38 8 -30 6.00E-11 6.00E-11 6.00E-I2 6.00E-I2 RR High PLOCA None Low 0 6 6 -3.00E-12 -3.00E-I2 -3.00E-13 -3.00E-13 RR Total 5.70E-11 5.70E-11 5.70E-12 5.70E-12 RWCU High LOCA None Low 7 2 -5 I.OOE-11 I.OOE-11 I.OOE-12 l.OOE-12 RWCU High ILOCA None Low 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RWCU High ILOCA-OC None Low 2 I -I 2.00E-I2 2.00E-12 2.00E-13 2.00E-13 RWCUTotal 1.20E-11 1.20E-11 1.20E-12 1.20E-12 SDVTotal Low Class 2 LSS Assume Medium 3I 0 -3I 3.10E-10 3.10E-10 3.10E-11 3.10E-11 SLC High LOCA None Low 0 2 2 -4.00E-I2 -4.00E-I2 -4.00E-13 -4.00E-I3 SLC High PLOCA None Low 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO SLC High PLOCA-OC None Low 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO SLC Total -4.00E-12 -4.00E-12 -4.00E-l3 -4.00E-13 Grand Total 287 67 -220 1.95E-09 3.03E-09 1.95E-10 3.03E-10 Notes

1. Systems are described in Table 3.1

2. Only those ASME Section XI Code inspection locations that received a volumetric examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-112657.

3. Only those RIS_B inspection locations that receive a volumetric examination are included in the count. Locations subjected to VT-2 only are not credited in the count for risk impact assessment.

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4. The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium", or "Low" depending upon potential susceptibly to the various types of degradation. [Note: Low Safety Significant (LSS) locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium")

5. The "LSS" designation 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).

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Table 4 Inspection Location Selections Comparison Safety Significance Break Failure Potential Code Weld Section XI Code Case N716 System Category Count High Low Location DMs Rank Vol/Surf Surf only RISB Other cs ../ LOCA None Low B-F, B-J 14 5 2 NA cs ../ PLOCA None Low B-J 10 2 0 NA cs ../ PLOCA-OC None Low B-J 6 6 I NA cs ../ LSS Assume Medium C-F-2 I62 22 0 NA FW ../ LOCA TT Medium B-J I9 7 0 NA FW ../ LOCA TASCS Medium B-J I2 7 3 NA FW ../ ILOCA TASCS Medium B-J 2 0 0 NA FW ../ ILOCA-OC TASCS Medium B-J 2 I I NA FW ../ LOCA TASCS, TT Medium B-J I2 I 2 NA FW ../ ILOCA TASCS, TT Medium B-J 2 0 0 NA FW ../ ILOCA-OC TASCS, TT Medium B-J 2 I I NA FW ../ LOCA None Low B-J I9 0 0 NA FW ../ LSS Assume Medium C-F-2 8 6 0 NA HPCI ../ LOCA None Low B-F, B-J 5 3 2 NA HPCI ../ ILOCA None Low B-J IO I 0 NA HPCI ../ ILOCA-OC None Low B-J 2 2 I NA HPCI ../ LSS Assume Medium C-F-2 I4I I7 0 NA INST ../ LOCA None Low B-F, B-J 25 2 IO 0 3 MS ../ LOCA None Low B-F, B-J 24I 35 41 11 9 MS ../ ILOCA/PLOCA None Low B-J I8 2 5 0 NA MS ../ ILOCA-OC None Low B-J 24 2I 2 9 NA PCAC ../ LSS Assume Medium C-F-2 52 I 0 NA RBCCW ../ LSS Assume Medium C-F-2 I5 0 0 NA RCIC ../ LOCA None Low B-J 3 0 2 NA RCIC ../ ILOCA None Low B-J IO 0 3 0 NA RCIC ../ ILOCA-OC None Low B-J 2 0 2 I NA RCIC ../ LSS Assume Medium C-F-2 51 2 0 NA RHR ../ LOCA TT Medium B-F, B-J 34 7 8 NA RHR ../ PLOCA TT Medium B-J 12 4 0 NA RHR ../ PLOCA-OC TT Medium B-J 4 4 4 NA RHR ../ LOCA None Low B-F, B-J 57 13 0 NA RHR ../ PLOCA None Low B-J 7 4 0 NA

--- --- _c_

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Safety Significance Break Failure Potential Code Weld Section XI Code Case N716 System High Low Location DMs Rank Category Count Voi!Surf Surf only RISB Other RHR ./ PLOCA-OC None Low B-J 2 2 0 NA RHR ./ LSS Assume Medium C-F-2 417 31 3 0 NA RR ./ LOCA None Low B-F, B-J 182 38 19 8 6 RR ./ PLOCA None Low B-J 12 0 3 6 NA RWCU ./ LOCA None Low B-F, B-J 22 7 1 2 NA RWCU ./ ILOCA None Low B-J 2 0 0 NA RWCU ./ ILOCA-OC None Low B-J 2 2 1 NA

./

C-F-1, SDV LSS Assume Medium 41 31 0 NA C-F-2 SLC ./ LOCA None Low B-F, B-J 24 0 6 2 1 SLC ./ PLOCA None Low B-J 4 0 1 0 NA SLC ./ PLOCA-OC None Low B-J 7 0 3 0 1 Totals --

1698 287 99 67 20 Notes

1. Systems are described in Table 3.1

2. 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) in a BWR to be credited toward the 10% requirement. This option is not applicable for the Monticello RIS_B application. The "Other" column has been retained in this table solely for uniformity purposes with other RIS_B application template submittals and to indicate when RIS_B selections will receive a VT-2 examination (these are not credited in risk impact assessment).

3. The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium", or "Low depending upon potential susceptibly to the various types of degradation. [Note: Low safety significant (LSS) locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").

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Attachment A to Enclosure Consideration of the Adequacy of Probabilistic Risk Assessment Model for Application of Code Case N-716 Monticello PRA Response to RG 1.200 Peer Review Findings A-1 of 24

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Change Number: MT-13-0025 {IE)

Brief Problem

Description:

ASEP is used for assessment of all pre-initiator HFEs. A detailed assessment was not used for significant HFEs.

F&O Number: 1-6 Technical Element: HR Detailed Problem

Description:

ASEP is used for assessment of all pre-initiator HFEs. A detailed assessment was not used for significant HFEs. The standard defines "significant basic event: a basic event that contributes significantly to the computed risks for a specific hazard group. For internal events, this includes any basic event that has an FV importance greater than 0.005 or a RAW importance greater than 2. (See Part 2 Requirements OA-C13, OA-01, OA-03, OA-05, OA-08, HR-02, and HR-G1.)" For examples of pre-initiator important HEPs see Table E-1 of PRA-MT-QU. (This F&O originated from SR HR-02)

Proposed Solution: Use detailed assessment for significant pre-initiator HFEs, e.g., THERP Risk Impact: Specific requirement for Cat II not met.

Actual Solution:

Significant pre-initiator HEPs, as found in the MNGP Rev 3.00 model, were re-evaluated with the THERP methodology. The following significant pre-initiator HEPs were identified in the Rev 3.00 model:

IPTP207WXZ, KCHV-SF-9Z, SPEP111AXZ, and SPEP111 BXZ. Pre-initiator HEP KCHT8089CZ was also significant but was determined to be modeled less accurately by the THERP methodology and hence the detailed ASEP methodology was retained. KCHT8089CZ is a miscalibration pre-initiator HFE.

In the THERP methodology under Execution Steps there really is no appropriate Error of Commission to choose for miscalibration errors, so Basic Error of Commission was chosen which has a high value of 1E-02. Using the THERP methodology and not applying any testing recovery credit, the HEP was 5.6E-03 which was higher and less accurate than the ASEP value. Once the testing recovery credit was applied (again maintenance done during shutdown but EOG testing done monthly) the value was reduced to 3.5E-05 which seems far too low. An error of commission item in the HRAC tool needs to be developed for miscalibration for this to be more accurately portrayed in HRAC.

The MNGP Rev 3.01 model was re-verified to identify any newly significant pre-initiator HEPs and found one significant pre-initiator HEP (NNTRAINBXZ) that was reanalyzed with the THERP methodology.

Miscalibration HEPs Q-672ABC044Z, Q-2352AB-22Z, and NSP4237HIZ were also identified as significant but have been previously discussed to be inaccurately represented by the THERP methodology. Common cause HEPs QSV672EFX22Z and HPTRCHPSX22Z were also found to be significant. Common cause failures are not well accounted for in the THERP methodology since THERP does not allow for the inclusion of the dependence between the two single failures. There was no obvious option in the HRA Calculator tool for modeling activities that impact more than one train hence the detailed ASEP methodology was retained. Other significant HEPs that were already discussed and re-examined were KCHT8089CZ, SPEP111 BXZ, and KCHV-SF-9Z.

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Change Number:* MT-13-0026 {IE)

Brief Problem

Description:

(This F&O originated from SR IE-AS) A relatively robust process for system engineering interviews and documentation of same is included in PRA-MT-WI, Rev. 3.0.

F&O Number: 3-2 Technical Element: IE Detailed Problem

Description:

(This F&O originated from SR IE-AS) A relatively robust process for system engineering interviews and documentation of same is included in PRA-MT-WI, Rev. 3.0. However, the form used to document system engineer interviews does not include a section or any criteria relating to whether failure of the system in question could cause an initiating event. Therefore, this SR is met only to CC 1.

Proposed Solution: Possible resolution is tore-interact with system engineers regarding the potential for overlooked initiating events, and document the results of that interaction.

Risk Impact: This F&O is deemed a Finding as it prevents the SR from meeting CCII.

Actual Solution:

Since the Peer Review, an Interview with Plant System Engineers and operators was performed to decide whether any initiating events were overlooked. Each system engineer was interviewed to verify that a failure of their system would not cause a plant trip, which is documented in Rev 3.1 of the Walkdown and Interview Notebook. All system engineers concluded that the initiating event fault trees modeled in the PRA were satisfactory in that no initiating events were overlooked. of the WI Rev 3.1 Notebook under the statement below documents that this work was performed: Verify with plant personnel to determine if potential initiating events have been over looked.

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Change Number: MT-13-0027 (IE)

Brief Problem

Description:

During review of the CAFTA fault tree files, some instances of credit for repair of hardware (EDGs) were noted that do not seem to be justified. An example is under gate AA131 of the AAC fault tree model.

F&O Number: 3-6 Technical Element: SY Detailed Problem

Description:

During review of the CAFTA fault tree files, some instances of credit for repair of hardware (EDGs) were noted that do not seem to be justified. An example is under gate AA 131 of the AAC fault tree model.

Under the cited gate, failure to recover a diesel'-generator is ANDed with "EDG 11 FAILURES WITH POTENTIAL TO RECOVER". Two issues were noted. First, there is not a corresponding gate for EDG 11 comprised of "failures that cannot be recovered". Second, some of the EDG 11 failures that are recovered, e.g., EDG 11 OOS for corrective maintenance, appear to be unjustified. (This F&O originated from SR SY-A24)

Proposed Solution: Search for instances of repair of hardware faults in the logic model, and for each instance identified consider whether credit for recovery/repair is justified. If not, revise the affected fault tree model. If so, document an adequate justification for the repair credit.

Risk Impact: Moving some of the EDG failures out from under the AND (i.e., recovery) gate may result in an appreciable impact on CDF.

Actual Solution:

An evaluation was conducted to investigate the basis for which components (basic events) should or should not be in the scope of EDG recovery. EDG non-recoverable failures are modeled under several gates. Certain accident sequences which require rapid response were excluded (NO-RECOVERY) from the scope of the recovery of the EDGs (for example floods that directly fail the EDG or ATWS). Failures of EDG cooling via HVAC or ESW were also excluded since it is assumed that failures of these systems would cause EDG overheating and hence be non-recoverable (for EDG 11 gate AA049 for EDG12 gate AA086).

RADS data for plants with similar EDGs were reviewed from 1999 to the present. The failures were reviewed and binned based on their time to recover (unavailability), so that an appropriate recovery probability could be calculated. These failures were categorized to determine their potential ability to be recovered. The following failure categories were determined to be recoverable:

1. Include components within EDG boundary. These components are inherently in the scope of the recovery data provided from the RADS database. This database is includes MSPI EDG data entries from all US nuclear power plants.

2. Include corrective maintenance as part of EDG since it represents unplanned unavailability (UA) data. This unplanned UA was utilized to calculate EDG recovery fraction.

3. Include preventative maintenance since the average planned maintenance is less than six hours per month with most months less than three hours. Recovery would be complete or back out of PM and place EDG back into service.

4. Events that do not directly disable the EDG, hence EDG recovery is not applicable but has a separate recovery credit (e.g. FLOODING ANDed with ALOWFUELHY) is allowed.

5. Include support system recoverable events. These events contribute to unplanned UA for EDGs which was utilized to calculate EDG recovery fraction.

6. Include recoverable Human Actions such as the failure to address low fuel oil. These events may cause temporary loss of EDG but recovery not expected to be difficult.

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Change Number: MT 0028 (IE)

Brief Problem

Description:

PRA-MT-QU Rev 3 and PRA-MT-L2-AS document the codes used and limitations that may impact applications in Section 2.4.

F&O Number: 4-2 Technical Element: LE Detailed Problem

Description:

PRA-MT-QU Rev 3 and PRA-MT-L2-AS document the codes used and limitations that may impact applications in Section 2.4. However, there is no discussion related to the impact of HEP dependencies on applications that may be requested from the site. For example, using the current'seed values for the HFE dependencies would prevent MT from being able to obtain cutsets at a truncation of 1E-12 which is required for MSPI (i.e. 7 orders of magnitude below CDF). (This F&O originated from SR QU-F5)

Proposed Solution: Discuss this limitation in the documentation for the current model. In the future, determine the appropriate method for HFE dependency which will allow truncation at an appropriate limit for support of applications.

Risk Impact: MSPI requires that models be quantified at 7 orders of magnitude below the CDF. This will require MT to truncate at a limit of at least 1E-12 which is unattainable based on discussion in Section 3.5 of PRA-MT-QU. Therefore, this item is a finding.

Actual Solution:

The MSPI Sensitivity was quantified for the MT Rev 3.0E model and truncated down to 6E-13 without any modification to our quantification process. It is believed that the improvements to the HEP dependency analysis along with other improvements have made this possible. Model revisions after Rev 3.0E have also been tested and have been successful. Since the model's risk has gone down the truncation also has to progress down which makes quantification at these lower truncations more difficult. Similarly with PI's most recent model, quantification had to be broken up by initiator to meet these lower truncations. The resulting cutsets are simply appended afterwards. This will be documented in the MSPI Basis document and is documented in the convergence study within the Rev 3.1 Quantification Notebook.

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Change Number: MT 0030 (IE)

Brief Problem

Description:

Section 8.1 of PRA-MT-L2-AS gives a table, Table 8.2, which identifies the sources of model uncertainties and assumptions that could impact the PRA model.

F&O Number: 4-6 Technical Element: LE Detailed Problem

Description:

Section 8.1 of PRA-MT-L2-AS gives a table, Table 8.2, which identifies the sources of model uncertainties and assumptions that could impact the PRA model. In this table is a description for the impact on the PRA model, but there doesn't appear to be a sound basis. For example, for the second item in the table, namely Debris Coolability in the Drywell, the 'approach taken' states that the event is set to FALSE but is provided for the purpose 'of performing sensitivity studies on this assumption;'

however, no sensitivity [study] is performed. (This F&O originated from SR LE-F3)

Proposed Solution: Perform sensitivity analysis to characterize the key model uncertainties identified.

Risk Impact: The sensitivity analysis performed should help to characterize the uncertainties identified; however, there aren't any supporting sensitivities performed for the identified uncertainties. Therefore the SR is not met.

Actual Solution:

PRA calculation PRA-CALC-13-001 (Level2 Sensitivity Studies- Model Rev 3.1) was generated to document the detailed level 2 sensitivity studies performed by Applied Reliability Engineering Inc. (ARE!)

in response to the internal events peer review finding (F&O number 4-6, concerning ASME/ANS RA-Sa Supporting Requirement LE-F3). Section 8.3 (Level 2 Sensitivity Studies) was added to revision 3.1 of the Level 2 PRA Notebook to summarize each of the studies documented in this PRA calculation.

PRA CALC-13-001 is referred to in the Level2 PRA Notebook as reference #43. PRA-CALC-13-001 has been reviewed and reviewer signed. All the above statements are accurate. The scope of the sensitivity studies was determined by agreement between the XCEL MNGP Team and ARE!.

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Change Number: MT 0031 (IE)

Brief Problem

Description:

Section 5.2 of PRA-MT-DA states that coincident maintenance at MT is not a common practice but there is no evidence that coincident maintenance activities have been reviewed for applicability of the PRA model.

F&O Number: 4-10 Technical Element: DA Detailed Problem

Description:

Section 5.2 of PRA-MT-DA states that coincident maintenance at MT is not a common practice, but there is no evidence that coincident maintenance activities have been reviewed for applicability of the PRA model. However, review of the fault tree demonstrates that RHR-A and RHR-C are taken OOS for corrective maintenance per review of event RLOOPAXXCM in the CAFTA fault tree. This is a clear example of coincident maintenance that should be examined and reviewed for impact on the PRA model. (This F&O originated from SR DA-C14)

Proposed Solution: Document/provide evidence of a review of past planned maintenance to identify any coincident maintenance activities and determine the impact on the PRA model for any items that are identified. A process for coincident maintenance review and impact on the model has been provided via email as one example that this SR can be MET.

Risk Impact: SR is NOT MET because SR requirement is not satisfied.

Actual Solution:

First, high risk combinations were identified that could occur due to online preventive maintenance coupled with random corrective maintenance. These combinations were identified by assigning probabilistic values consistent with those observed for the system combinations from actual plant experience. Second, a detailed cycle plan review was performed for the period 2009-2012 to verify that the risk significant combinations of components/trains were not coincidently taken out of service. The result of this review concluded that there are no coincident maintenance outages as a matter of practice at Monticello.

See Section 5.2 of Rev 3.1 of the Data Notebook for further discussion on this item.

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Change Number: MT 0032 (IE)

Brief Problem

Description:

During review of the RHR fault tree, an asymmetry was identified in the fault tree related to Div I and Div II Flow diversion. Specifically, gates R009-A and R010-A should be similar just as R012-A and R013-A are.

F&O Number: 4-11 Technical Element: QU Detailed Problem

Description:

During review of the RHR fault tree, an asymmetry was identified in the fault tree related to Div I and Div II Flow diversion. Specifically, gates R009-A and R010-A should be similar just as R012-A and R013-A are. The non-symmetry occurs due to including maintenance events under the fail to run gate for one pump and not the other three pumps in the RHR system (or at the very least one of the pumps in the other division). The model should be consistent or documentation should be provided describing the reason for the asymmetry. (This F&O originated from SR QU-F2)

Proposed Solution: Correct the model or documentation for the RHR system fault tree.

Risk Impact: Documentation issue Actual Solution:

In the RHR fault tree, the description of gate R009_1 was " ... common cause failure to start of 2 or 3 pumps" where the proper gate description should have been " ... common cause failure to run of 2 or 3 pumps". All basic events under this gate are run failures. It appears that the mislabeling of this gate led to the inclusion of maintenance events under the gate, as maintenance events are included under the run failure gates for all of the RHR pumps. The gate name (R009_1) description was corrected to reflect it is a failure to run gate, and the maintenance events under the gate were removed. A review of the fault tree was conducted to verify the proper inclusion of maintenance events under the appropriate (failure to start) gates for each of the four RHR pumps.

Maintenance events are similarly located for each pump under gate R050 for P-202A, gate R055 for P-202C, gate R068 for P-202D, and gate R063 for P-2028. Gates R009-A and R010-A are now symmetrical.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0033 {IE)

Brief Problem

Description:

The use of a half failure for plant-specific data is being questioned. A SSe either fails to perform its function or it doesn't.

F&O Number: 5-12 Technical Element: DA Detailed Problem

Description:

The use of a half failure for plant-specific data is being questioned. A sse either fails to perform its function or it doesn't. Examples include:

1) A valve that had minor leakage was considered a half failure. This should be determined if the valve could perform its safety function or not.

2) A pump started, stopped, and then restarted successfully. What caused the pump to restart? Was it a manual restart or automatic?

3) A valve did not meet an 1ST requirement during a surveillance test. Would the valve still be able to perform its function?

4) A valve was not able to maintain a required flow, but could maintain a specific flow. Is that specific flow acceptable for success criteria? If it is, then no failure. (This F&O originated from SR DA-e4)

Proposed Solution: Review the cases where a half failure is given and determine if those are PRA failures or not.

Risk Impact: The basis for identifying events as failures was not clearly defined and is viewed as atypical. Therefore, this is a finding.

Actual Solution:

All half failure rates as reported in Rev 3.0 of the Data Notebook were reviewed to determine if the component failed to perform its intended function. The half failures were adjusted accordingly (to be counted as 0 or 1) depending on the scenario with a confirmation from system engineers. These changes Were corrected in the basic event type code database titled "MNGP _REV_3-0I.EGS_06-05-13_IE_&_type_code_Database.mdb" and included in any future revisions.

The database file has since been split into two: one for type codes and one for initiating events. The new type code database is titled: "MNGP_ Te_REV_3-1.mdb". Paragraph description now added in Section 4.5.4 of the Data Notebook.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0034 (IE)

Brief Problem

Description:

The times that components were in their standby statuses were estimated by using a general understanding of average system operation (i.e., a pump in a 2 out of 3 pump system would have 1/3 probability in standby).

F&O Number: 5-15 Technical Element: DA Detailed Problem

Description:

The times that components were in their standby statuses were estimated by using a general understanding of average system operation (i.e., a pump in a 2 out of 3 pump system would have 1/3 probability in standby). Plant-specific operational records were not reviewed. Therefore, this SR is MET at CAT I, but NOT MET at CAT 11/111. (This F&O originated from SR DA-C8)

Proposed Solution: Review plant-specific data to capture more realistic standby probabilities.

The use of PI data to determine running data combined with the time the sse is unavailable may be used to determine the plant-specific standby time.

Risk Impact: Because plant-specific operational records were not reviewed, this SR is not met at CAT IIIII I, therefore this is a finding.

Actual Solution:

The standby data for the DC, SW, and AIR systems were collected from the SOMs, PI Systems, or Procedure 4953-PM respectively to establish more realistic probabilities for the standby flags BE of redundant components. For systems which lacked sufficient data in PI Systems and SOMs logs, system engineer interviews were conducted to determine appropriate standby fractions. The following systems lacked sufficient plant data: CRDH, HVAC, and RBCCW. The time period reviewed was 2008 to 2012, except for the AIR system which was recently upgraded in 2010. The spreadsheet titled, "MT Standby Fraction.xls" documents this work.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0035 {IE)

Brief Problem

Description:

Mean values and uncertainty intervals were provided through Bayesian updating. However, a number of important basic events in Section 7.1 of PRA-MT-DA were given point estimates without any associated uncertainty values.

F&O Number: 5-16 Technical Element: DA Detailed Problem

Description:

Mean values and uncertainty intervals were provided through Bayesian updating. However, a number of important basic events in Section 7.1 of PRA-MT-DA were given point estimates without any associated uncertainty values. More importantly, no sensitivity studies were performed to ascertain the impact of the point estimates on the model. (This F&O originated from SR DA-D3)

Proposed Solution: Assign appropriate uncertainty values to the point estimates given in Section 7.1 and perform sensitivity studies to determine the impact of these values to the model.

Risk Impact: These point estimates may have a large impact on the model, therefore this is a finding.

Actual Solution:

For events not linked to Type Codes, the uncertainty bounds for basic events are defined by the use of lognormal distribution Error Factors (EFs) and are assigned as follows:

When available, EFs or Variances are obtained based on the distribution values provided by the data source. When such information is not available, the general EF guidelines below are used.

1-Human Error Probabilities: related to Human Error are based on information in Section 7 of NUREG/CR-1278 (Handbook of HRA). The Equipment EF guidelines are based on comparison with various data sources.

2-for failure rates probabilities related to Special BE, an engineering judgment and comparisons with similar BE data was utilized.

Table 7-1a Guideline related for probabilities missing distribution attributes is listed in the Data notebook Section 7. 1.

Sensitivity studies were performed to determine significant point estimate basic events. The Sensitivity study concluded three basic events which are fairly important to LERF only. These events are JUMPERS "Hardware needed to align alternate power supply to div. 2 250V DC fails", MVR4543XXN "Hard-pipe vent rupture disk PSD-4543 fails to open", and XPP-SRV--L "SRV tailpipe rupture in the wet-well airspace". Each basic event contributes an increase in LERF risk by more than 10% in comparison to all special events included in the study.

The insights from the study indicate that our model is sensitive to the following basic events:

JUMPERS: The original probability was conservatively estimated to be 1.0E-3. Since this failure only pertains to the jumper cables themselves this failure probability is much higher than other passive components (heat exchangers or check valves). Additionally, this procedure and equipment has been functionally tested (during the 2005 refueling outage) to provide battery charging following a complete discharge.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT-13-0035 (IE) (continued)

MVR4543XXN: The only feasible ways that the rupture disk would not open when containment pressure challenges containment integrity (> 100 psi g) is if the incorrect rupture disk is installed, multiple rupture disks are installed together, or the rupture disk is manufactured incorrectly. The probability of installing multiple disks together is considered to be negligible because of the level of training and experience of maintenance workers, as well as the quality assurance program at the site. This original estimated probability of 1.0E-3 was found to be conservative since the failure rate is much higher than the failure rate to similar passive components.

XPP-SRV--L: The probability of 3.6006 E-04 is assumed to have a testing period of 2 years for three of the SRVs, and 4 years for the other five SRVs. The probability that a rupture is present (on any SRV tailpipe) is calculated by using a calculation type 5 event with a failure rate for all eight SRV tailpipes (estimated to be a total of 100 feet of piping) with an average testing period of 3.25 years. The piping structural analysis performed in PRA-MEM0-12-007 also concluded that the most limited pipe stress in the SRV tail pipe is below water line not in the airspace. Therefore, the probability is also reasonable given the calculation method and the location of the rupture.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0036 (IE)

Brief Problem

Description:

Section 4.0 discusses the use of Bayesian Updating of generic priors to provide posterior distributions. Generic information was primarily collected from NUREG/CR-6928.

F&O Number: 5-17 Technical Element: DA Detailed Problem

Description:

Section 4.0 discusses the use of Bayesian Updating of generic priors to provide posterior distributions.

Generic information was primarily collected from NUREG/CR-6928. A number of parameters did not have plant-specific information, therefore only generic information was used. (This F&O originated from SR DA-01)

Proposed Solution: An industry reference for a probability of consequential LOOP is given in NUREG/CR-6890 Vol. 1, Section 6.3. Alternately, obtain generic data and perform a Bayesian update using plant-specific data to generate a posterior value.

Risk Impact: This SR requires realistic parameter estimates for significant basic events based on relevant generic and plant-specific evidence. There is no evidence to support a realistic parameter estimate for the consequential LOOP (ALOOPXXXXL). Therefore this is finding.

Actual Solution:

The consequential LOOP basic event was reanalyzed and broken up into two different types based on the scenario. The first is the possibility that a LOCA event with the automatic start of the large ECCS pumps induces a grid or plant instability that leads to a LOOP. The second possibility was the consideration of LOOP due to transient. EPRI has reviewed NRC references (such as NUREG/CR-6890 Vol 1) and provided estimates for conditional LOOP frequencies. The EPRI guidance provided the following recommended values:

1. Consequential LOOP for initiators with LOCA was given a frequency of 2.4E-2 per year. The new basic event name for this scenario is ALOOPWLOCA.

2. The consequential LOOP caused by a Transient was given a frequency of 2.4E-3 per year. The new basic event name for this scenario is ALOOPTRANS.

Bayesian updating of data was performed on risk significant components from the 2004 Monticello PRA Model (CDF and LERF). The 2004 Monticello model was the model of record at the time of the project's commencement. When the data is updated again, the events which are given a Bayesian update will also be revisited.

The reasonableness check of the Prior versus Posterior was performed using the mean of the Posterior to the 90% confidence intervals of the prior 5th-95th percentile. This check verified that that mean of the Posterior with 90% confidence falls between the Prior's 5th and 95th percentile values. This work was performed and documented per Suggestion 5-13 (PCD MT-13-0077).

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0037 {IE)

Brief Problem

Description:

The system description for the Condensate and Feedwater system (PRA-SY-CFW) states that the oil coolers for the new condensate pumps being installed will be dependent on Service Water; however, SW isn't listed as a required support system.

F&O Number: 6-4 Technical Element: SY Detailed Problem

Description:

The system description for the Condensate and Feedwater system (PRA-SY-CFW) states that the oil coolers for the new condensate pumps being installed will be dependent on Service Water; however, SW isn't listed as a required support system. (This F&O originated from SR SY-89)

Proposed Solution: Include SW as a required support system for CFW.

Risk Impact: Documentation and model issue. This dependency may have an appreciable impact on CDF.

Actual Solution:

The CFW fault tree (CFW Rev 3.0.B.ABS_04-30-13.caf) was revised to include service water as a dependency for the condensate pumps. Specifically, gate T_SW-SW was added under gates F009, F009-S, F010 and F010-S. The revised fault tree was renamed "CFW Rev 3.0.C.TPW_05-17-13.caf' and filed appropriately. The system notebook for the CFW system was revised to reflect these changes.

A similar PCD (MT-12-0034) was already established, and will be closed out in conjunction with this PCD. Additionally, PCD MT-11-0027, which is more general but related to EPU changes to the CFW system, will be updated to note the change from this PCD.

Also, feedwater long term DC dependencies were changed to be long term instead of short term.

Specifically gate T_D-DC111-S was changed toT_D-DC111-L under the F051 gate and gate T_D-DC211-S was changed to T_D-DC211-L under gate F054.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT-13-0038 (IE)

Brief Problem

Description:

PRA-MT-IF-IE Rev 3 includes the information regarding the potential sources of flooding. How-ever, the inadvertent actuation of the fire suppression system is not discussed as a potential initiating event.

F&O Number: 6-6 Technical Element: IFSO Detailed Problem

Description:

PRA-MT-IF-IE Rev 3 includes the information regarding the potential sources of flooding. However, the inadvertent actuation of the fire suppression system is not discussed as a potential initiating event. (This F&O originated from SR IFSO-A4)

Proposed Solution: Perform the evaluation of inadvertent fire suppression actuation and include it in PRA-MT-IF-IE.

Risk Impact: This evaluation is a requirement of this SR.

Actual Solution:

The following information has been added to the Internal Flooding Initiating Events (PRA-MT-IF-IE)

Notebook, Section 2.1: CAP 1266742 evaluated the FPS with regard to effects on plant equipmenUareas if inadvertent system actuation were to occur. The conclusion of the evaluation is that inadvertent actuation would not affect any safety-related equipment's ability to perform its function. This evaluation further supports the following positions regarding inadvertent actuation of the fire suppression system in the MNGP PRA Model.

Sprinkler System The effects of an inadvertent sprinkler discharge are bounded by the effects modeled for random pipe breaks in the same area. The spray effects of a sprinkler discharge would likely be less since the sprinkler targets a specific location while a random pipe break is assumed to spray everything in the room. Additionally, all small FPS piping containing the sprinklers is included in the random FPS pipe break frequencies calculated for the various flood areas. Given that MNGP does not have a history of random sprinkler discharges leading to a plant transient and such events are so rare throughout the industry that no failure data has been developed for such events, inadvertent sprinkler discharge is considered to be within the bounds of uncertainty for the existing flood scenarios hence is not explicitly modeled.

Deluge System The effects of an inadvertent actuation of a deluge system are bounded by the effects modeled for random pipe breaks in the same area. The spray effects of a deluge discharge would likely be less since the deluge system is focused on a specific set of targets while a random pipe break is assumed to spray everything in the room. The human actions associated with isolation of a deluge discharge are considerably different than those for a random pipe break, however. The control room will receive a unique alarm for a deluge discharge indicating exactly where the operator should investigate instead of simply receiving a 'fire pump running' alarm leaving the operator to search for the location of the event.

Additionally, the isolation. of the deluge discharge can be performed at the discharge location by simply closing the deluge valve rather than having to travel to the intake structure to locally trip the fire pumps.

While explicit inclusion of this deluge system actuation would increase the initiating event frequency slightly, this increase would be more than offset by the decreased HEP associated with the isolation of the deluge discharge. Thus, inadvertent actuation of a deluge system is considered to be within the uncertainty bounds of the existing flood scenarios and is not explicitly modeled.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0039 {IE)

Brief Problem

Description:

There is no evidence of plant-specific analysis completed or research done to identify or collect repair time information.

F&O Number: 7-6 Technical Element: DA Detailed Problem

Description:

There is no evidence of plant-specific analysis completed or research done to identify or collect repair time information. The EDG recovery values in Table 1 of PRA -MT-SY-RECAC are based solely on a generic data source. This is contrary to the SR requirement to gather both generic and plant specific information. (This F&O originated from SR DA-C15)

Proposed Solution: If EDG recovery is used in the model, base the values on plant specific experience of EDG repair or show "that the plant EDGs are sufficiently similar... such that the generic data can be used to characterize the EDG repair probability" as stated in the EPRI LOOP Technical Guidelines.

Risk Impact: No plant-specific experience is given for repairs which appear to be very plant specific in nature.

Actual Solution:

Data from the NRC's Reliability and Availability Data System (RADS) was utilized to calculate an updated EDG recovery estimate. RADS was used to obtain more data to establish a larger data population than only MNGP related failures. This analysis used similar assumptions and methodology utilized by the most recent NRC diesel generator repair time estimates. The RADS Unplanned Unavailability (UUA) outage data for similarly designed EDGs from 1998-2010 was utilized to characterize diesel generator repair probabilities. Each UUA event was categorized into their respective time span. The sum of events within the time span was calculated. The percent recovered was calculated based on the total of UUA events (1 ,401 ). The result of this analysis is provided in Appendix A in the RECAC (AC Recovery) Notebook Rev. 3.1.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT-13-0040 (IE)

Brief Problem

Description:

(This F&O originated from SR QU-83) Convergence is only demonstrated when quantifying with nominal HEP values. There is no convergence demonstrated when the method described in PRA-MT-QU Rev 3 is used for quantification.

F&O Number: 7-7 Technical Element: QU Detailed Problem

Description:

(This F&O originated from SR QU-83) Convergence is only demonstrated when quantifying with nominal HEP values. There is no convergence demonstrated when the method described in PRA-MT-QU Rev 3 is used for quantification. The truncation level (1 E-10) is quite high even given the E-05 CDF value. Also little basis is given for using the 1E-1 0 truncation rather than the possible lower E-11 truncation.

Proposed Solution: Review the model to identify what is limiting the quantification truncation. The self-identified issue of dependent HEP methodology is one potential factor but others could exist.

Risk Impact: A sufficiently low converged truncation is required for many applications such as MSPI.

Actual Solution:

Convergence of the internal events only CDF, complete CDF, and LERF has been tested and proven for each model revision since Rev 3.0E. Quantification at truncation levels required by MSPI (7 orders below converged CDF value) were also successful. No pre or post quantification modification to the model was required to establish this convergence. It is believed that since the Peer Review the improvements to the HEP dependency analysis along with other improvements have made this possible.

Since the model's risk has gone down since Rev 3.0 the truncation must also progress down to meet the MSPI requirement, which makes quantification at these lower truncations more difficult. To solve at these lower truncations, quantification is performed on smaller groups of initiators with the resulting cutsets appended post quantification. This is a simple and accurate method to quantify the model at these rarely used truncation levels. The final convergence study will be documented in the Quantification Notebook.

The Rev 3.1 model was solved at a truncation of 1E-13 by splitting the initiators into four groups by flags.

Convergence of total CDF, no flood CDF, and total LERF were established at 1E-12, 1E-13, and 1E-12 truncations respectively.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0041 (IE)

Brief Problem

Description:

Truncation is only 5 orders of magnitude below CDF. This is less than general industry practice. PRA-MT-QU Rev 3 states that it is because the model quantification fails at lower truncation levels as a result of the joint HEP methodology used.

F&O Number: 7-9 Technical Element: QU Detailed Problem

Description:

Truncation is only 5 orders of magnitude below CDF. This is less than general industry practice.

PRAMT-QU Rev 3.0 states that it is because the model quantification fails at lower truncation levels as a result of the joint HEP methodology used. (This F&O originated from SR QU-82)

Proposed Solution: Identify and resolve the issues with quantification to allow quantification at truncation levels required by the standard and for applications. This may include review of the incorporation of HEP dependencies into the quantification, and review of fault tree structure to optimize quantification effectiveness among other possibilities.

Risk Impact: Events can become significant due to many lower value cutsets containing those events whose summed valu.e is important.

Actual Solution:

Convergence of the internal events only CDF, complete CDF, and LERF has been tested and proven for each model revision since Rev 3.0E. Quantification at truncation levels required by MSPI (7 orders below converged CDF value) were also successful. No pre or post quantification modification to the model was required to establish this convergence. It is believed that since the Peer Review the improvements to the HEP dependency analysis along with other improvements have made this possible.

Since the model's risk has gone down since Rev 3.0 the truncation must also progress down to meet the MSPI requirement, which makes quantification at these lower truncations more difficult. To solve at these lower truncations, quantification is performed on smaller groups of initiators with the resulting cutsets appended post quantification. This is a simple and accurate method to quantify the model at these rarely used truncation levels. The final convergence study will be documented in the Quantification Notebook.

The Rev 3.1 model was solved at a truncation of 1E-13 by splitting the initiators into four groups by flags.

Convergence of total CDF, no flood CDF, and total LERF were established at 1E-12, 1E-13, and 1 E-12 truncations respectively.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0042 (IE)

Brief Problem

Description:

The reasonableness check appears to be limited to whether the individual values are reasonable, not whether similar (timing, complexity, out of control room, etc.) actions when compared to each other have reasonable values, which is the intent of the SR.

F&O Number: 7-15 Technical Element: HR Detailed Problem

Description:

The reasonableness check appears to be limited to whether the individual values are reasonable, not whether similar (timing, complexity, out of control room, etc.) actions when compared to each other have reasonable values, which is the intent of the SR. (This F&O originated from SR HR-G6)

Proposed Solution: Perform the reasonableness check between events.

Risk Impact: Comparison between events with similarities is an important part of validating the values.

Actual Solution:

Since the Peer Review multiple reviews of the post initiator HEPs were performed using the HRA Calculator to get an accurate picture of the overall HRA analysis. A stress review report was generated and reviewed for all HEPs to ensure that the assigned stress levels are accurate. In general actions that are rarely performed, are outside the control room, or have little time to perform were assigned a higher stress level. The locations of HEPs were also reviewed to ensure that consistent naming was used for accurate use in the dependency analysis. A review of the complete HEP values against other similar actions was also performed. Any screening value HEPs that were found to be risk significant were developed further to establish a more accurate failure probability. Similar actions, such as flood isolations and manual depressurizations, were reviewed collectively to validate that actions that have more time or are located in the control room were given a lower failure probability. These reviews have been documented in Section 8.2 of the HRA Notebook, Rev 3.1.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0043 {IE)

Brief Problem

Description:

However, it appears that combinations (up to 6) HEPs appearing in cutsets are inappropriately analyzed for dependency resulting in over 8000 dependent combinations.

F&O Number: 7-16 Technical Element: HR Detailed Problem

Description:

The dependency analysis was performed using the HRA calculator which is an acceptable method to perform the dependency analysis and there is evidence that the limitation of the HRA calculator dependency module on the sequence of events times for multiple HEPs within a cutset was addressed at least in part. However, it appears that combinations (up to 6) HEPs appearing in cutsets are inappropriately analyzed for dependency resulting in over 8000 dependent combinations. This analysis does not account for the subset of HEPs where the dependency is addressed via the separate cognitive and execution events. (This F&O originated from SR HR-G7)

Proposed Solution: Redo the dependency analysis using an awareness of the mix of HEP types, the concept of minimum values, consideration of whether the large number of HEPs in a single cutset is appropriate while retaining the current correct addressing of the cutset length limitations and time sequencing of HEPs from the current analysis. Review available industry guidance on performance of HEP dependency analysis to garner the best, most efficient way to perform the analysis. Ensure the

. software limitations are explicitly addressed in the dependency analysis process.

Risk Impact: Large sets of HEPs in a cutset may be due to the mix of methods (separate cognitive and execution events with the dependencies explicitly addressed and combined events not addressing dependency). Where the dependency was addressed by the separate events method already, these should be removed from or limited in the dependency analysis.

Actual Solution:

Post model Rev 3.0, a large effort was conducted to simplify and standardize the post initiator HRA analysis. First, cognitive and execution only HEPs were combined into one single HEP to create a more logical and simplified method for accounting for dependencies between HEPs. Second, an effort was conducted to combine the remaining HEPs that were very similar such as high pressure injection with FW or HPCI/RCIC. The number of flood isolation HEPs were reduced from 57 to 12 based on their HEP value and similarity of action. Model logic that duplicated the same operator action for different timings were also removed since these actions are not actually separate (for example depressurizing with one SRV or three). The efforts previously described have reduced the number of HEPs from 160 to 79 and reduced the HEP combos from approximately 12,000 (Rev 3.0) to 1,600 in the Rev 3.1 model. The amount of HEPs and dependencies is now well within the limitations of the current software used for HRA (HRA Calculator) and quantification (CAFTA and FTREX).

See Table 22 in the PRA-MT-HR Rev 3.1 Notebook for the summary of the work performed to standardize the HRA analysis by combining cognitive and execution HEPs from the 2004 model.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0044 (IE)

Brief Problem

Description:

Part of a reasonableness check should be an evaluation as to whether very low HEP values make sense. Common industry practice is to employ a floor or lower limit value of 1E-06 or 5E-07 to any HEP calculated value below these limits.

F&O Number: 7-17 Technical Element: HR Detailed Problem

Description:

Part of a reasonableness check should be an evaluation as to whether very low HEP values make sense. Common industry practice is to employ a floor or lower limit value of 1E-06 or 5E-07 to any HEP calculated value below these limits. This applies to both independent and dependent HEP values. For example from the DAC, combination 1 has a resultant dependent HEP value of E-11. This is mainly an issue with the dependent HEP values. (This F&O originated from SR HR-G6)

Proposed Solution: Incorporate the use of floor values in the HRA analysis.

Risk Impact: Use of extremely low HEP values is not realistic from a human performance perspective and can skew the PRA results.

Actual Solution:

All post Rev 3.0 quantifications now include a floor limit of independent or dependent combinations of HEPs of 1E-7. This number is currently being debated among the industry and regulators and is consistent with Prairie Island. A floor limit of the single HEP was set to 1E-5 in the HRAC database.

This single HEP floor limit is rarely used for post initiator HEPs, XDEPHOURSY and J2NDPHRS-Y only.

These are reserved for simple control room actions which need to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Multiple pre-initiator HEPs have this screening level value.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0045 (IE)

Brief Problem

Description:

A comparison of MNGP results with other plants was documented in Section 4.3.6.

F&O Number: 8-1 Technical Element: QU Detailed Problem

Description:

A comparison of MNGP results with other plants was documented in Section 4.3.6. Loss of 125 VDC initiators are a much higher contribution to CDF than for the other three comparison plants, but no adequate explanation was offered as to why there is such a large discrepancy. (This F&O originated from SR QU-04)

Proposed Solution: State physical reasons why the 125 VDC initiators are more significant for MNGP, or else identify why these cutsets may be "artificially" high, e.g., use of HEP dependencies.

Risk Impact: The SR states that the causes for significant differences with other comparison plants needs to be identified.

Actual Solution:

A discussion comparing Monticello's results to other similarly designed BWRs will not be created until the Rev 3.1 model is frozen July 151h. This discussion will not affect the PRA model or its applications.

The importance of the complete loss of 125VDC was researched and found to be a result of having the common cause failure of all three battery chargers fails short term injection. This basic event was removed since only two of the three battery chargers are normally in service with the backup in standby.

The basic event was replaced with the common cause of two battery chargers failing and the backup also failing independently. These changes reduce the overall CDF significance for complete loss of 125VDC from -18%( Rev 3.0) to 0.0% (3.0H) at a 1E-9 truncation, which is much more consistent with other similar BWRs. The modeling for complete loss of either single train of 125VDC is about a 1%

contributor for non-flood scenarios.

Discussion was added to describe the differences between Monticello and other similarly designed plants in Section 4.3.6 of the QU Notebook. Comparisons to other similarly designed plants were made to the best of our ability based on the information provided.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT 0046 {IE)

Brief Problem

Description:

Based on a review of the CAFTA database file (PRA-MT-DA-CDF-OPT Rev 3.0.RR), it appears that Error Factors were assigned to both Beta and Gamma distributions in the Type Code table instead of using variance values.

F&O Number: 8-5 Technical Element: QU Detailed Problem

Description:

Based on a review of the CAFTA database file (PRA-MT-DA-CDF-OPT Rev 3.0.RR), it appears that Error Factors were assigned to both Beta and Gamma distributions in the Type Code table instead of using variance values. In Attachment 1 of the Data notebook (PRA-MT-DA), the same type codes listed have assigned variance values for their Beta or Gamma distribution. This appears to be a discrepancy in the way the data was utilized in estimating the CDF uncertainty intervals. (This F&O originated from SR QU-E3)

Proposed Solution: Revise the distribution parameters in the Type Code table of the CAFTA database such that they are consistent with the PRA documentation. If necessary, the parametric uncertainty analysis may need to be re-performed in order to more accurately estimate the CDF and LERF uncertainty intervals.

Risk Impact: For RG 1.200 compliance in order to meet Capability Category II, the mechanisms listed under Category Ill of the Standard are required to be qualitatively addressed.

Actual Solution:

The Uncertainty Parameters that were implemented in CAFTA were reviewed to assure the data is using the proper uncertainty parameter (Error factor or variance) according to the type of Distribution (Gamma, Beta, and lognormal). The data notebook was updated to be consistent with the CAFTA database. A list of the appropriate parameters with the distribution type is listed in the Rev 3.1 Data notebook, Section 2.

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MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST RR-003 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Change Number: MT-13-0047 (IE)

Brief Problem

Description:

It appears that the Category Ill items, such as pipe whip, humidity, condensation, and temperature were not qualitatively addressed (see NRC Resolution, which is required for Cat. II) (This F&O originated from SR IFSN-A6)

F&O Number: 8-8 Technical Element: IFSN Detailed Problem

Description:

It appears that the Category Ill items, such as pipe whip, humidity, condensation, and temperature were not qualitatively addressed (see NRC Resolution, which is required for Cat. II). (This F&O originated from SR IFSN-A6)

Proposed Solution: The mechanisms listed for Capability Category Ill should be qualitatively addressed using conservative methods and treatments to show what additional impact, if any, may be imposed on existing internal flood scenarios.

Risk Impact: Basis for Significance For RG 1.200 compliance in order to meet Capability Category II, the mechanisms listed under Category Ill of the Standard are required to be qualitatively addressed.

This will have little to no impact of CDF or LERF.

Actual Solution:

Per the resolution requested from the Peer Review, a qualitative assessment was performed for the following flood failure mechanisms to meet Capability Category II criteria:

- Submergence

-Spray

-Jet Impingement

-Pipe Whip

-Humidity

- Condensation

-Temperature Spray and submergence effects are specifically discussed and accounted for throughout the Internal Flooding Accident Sequence Notebook for those internal flooding scenarios that are impacted. All internal flood initiators account for submergence and spray.

Jet Impingement and Pipe Whip are assumed to have no effect on the internal flooding analysis unless specifically stated in the internal flooding scenario. This assumption is based on the research provided in NUREG/CR-3231 which found that these potential impacts of pipe breaks require very specific pipe spacing/orientation, rarely result in a complete (guillotine) break, and even if pipe break occurs, significant reduction in cross sectional area of the target pipe follows. The instances which all these conditions are met are few in MNGP and the probability of these types of events coincident with a flood initiator is sufficiently remote. Assumption #5 was added to the "General Assumption" Section 2.2 of the Internal Flooding Accident Sequence Notebook, PRA-MT-IF-AS Rev. 3.1.

Humidity, Condensation, and temperature effects are assumed to be encompassed in the bounding, conservative assumption that all spray floods in the MNGP flood model fail the entire room where the flood exists irrespective of room size or pipe orientation. Assumption #6 was added to the "General Assumption" Section 2.2 of the Internal Flooding Accident Sequence Notebook, PRA-MT-IF-AS Rev. 3.1.

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