Text

Lusignan Declaration with Attachments
	ML15169B121
Person / Time
Site:	Indian Point
Issue date:	06/18/2015
From:	Lusignan B; State of NY, Office of the Attorney General
To:	; Atomic Safety and Licensing Board Panel
	SECY RAS
Shared Package
ML15169B118	List: ML15169B119; ML15169B120; ML15169B121; ML15169B122;
References
	50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, RAS 27962
	Download: ML15169B121 (122)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD

x In re: Docket Nos. 50-247-LR; 50-286-LR License Renewal Application Submitted by ASLBP No. 07-858-03-LR-BD01 Entergy Nuclear Indian Point 2, LLC, DPR-26, DPR-64 Entergy Nuclear Indian Point 3, LLC, and Entergy Nuclear Operations, Inc. June 18, 2015

x DECLARATION BRIAN LUSIGNAN Pursuant to 28 U.S.C. § 1746, Brian Lusignan hereby declares as follows:

1. I serve as an Assistant Attorney General for the State of New York, counsel for petitioner-intervenor State of New York in this proceeding. I submit this declaration and accompanying attachments a part of the State of New Yorks reply to the June 4, 2015 Joint Brief of Entergy and Westinghouse Regarding Proprietary Documents (Joint Industry Brief).

2. Attached to this declaration as Attachment 1 is a true and correct copy of the publicly available, full-text copyrighted document authored by Christopher Kupper and Mark Gray, entitled License Renewal Environmental Fatigue Screening Application, Paper No.

PVP2014-29093, Proceedings of the ASME 2014 Pressure Vessels and Piping Conference, Anaheim, California (July 20-24, 2014). This paper is publicly available at http://proceedings.asmedigitalcollection.asme.org/proceeding.aspx?articleID=1937745.

3. Attached to this declaration as Attachment 2 is a true and correct copy of EPRI Report 1024995 entitled EAF Screening: Process and Technical Basis for Identifying EAF Limiting Locations. It is publicly available at http://www.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=000000000001024995.

4. Attached to this declaration as Attachment 3 is a true and correct copy of the 1

September 19, 2013 NRC Staffs License Renewal Team Inspection Report 05000247/2013010 (ML13263A020). It is publicly available at http://pbadupws.nrc.gov/docs/ML1326/ML13263A020.pdf. NRC Staff disclosed this document in September 2013 as NRC document 56-0012. Entergy disclosed this document in October 2013 as Entergy document 1652.

5. On behalf of the State of New York, I attended the February 19, 2015 public meeting convened by NRC Staff and industry representatives at Three White Flint North at the NRC Headquarters in Rockville, Maryland, to discuss reactor pressure vessel issues. To my knowledge, no transcript was made of the meeting. The meeting was attended by many industry representatives and NRC Staff members. Individuals attempted to listen to the meeting via telephone, and at certain times those individuals also attempted to ask questions or make comments. A variety of technical issues appeared to substantially impair the publics ability to participate by telephone.

6. During the February 19, 2015, public meeting, NRC Staff presented a PowerPoint presentation entitled Part II: Assessment of Impact on Plants Using BTP 5-3 to Estimate RTNDT(u). This slideshow is publicly available on ADAMS at Accession No. ML15061A075.

The slideshow evaluated the potential impact of BTP 5-3 non-conservatisms on 19 pressurized water reactors (PWRs), and in particular on the Pressurized Thermal Shock (PTS) screening criteria. Slide 2 indicated that adjustment of the PTS evaluation to account for BTP 5-3 non-conservatism may cause one of these plants to exceed the PTS screen criteria during the license renewal (LR) period. In response to this presentation, a person who had been listening over the telephone asked NRC Staff to identify which plant would exceed the PTS screening limit. NRC Staff refused to identify that plant, although a staff member did state that the plant was not in Michigan. A person on the telephone also asked NRC Staff to identify the 19 PWRs that had

been evaluated as part of this study. Again, NRC Staff refused to identify the plants that had been studied.

7. Attached to this declaration as Attachment 4 is a summary of the February 19, 2015 public meeting in Rockville, Maryland to discuss reactor pressure vessel issues, including a list of attendees (April 6, 2015) (ML15096A128).

8. Attached to this declaration as Attachment 5 is an e-mail I received from Hearingdocket@nrc.gov on June 4, 2015, notifying me that Entergy and Westinghouse had filed their Joint Brief using the non-public Electronic Information Exchange (EIE).

9. Attached to this declaration as Attachment 6 is an e-mail I sent to attorneys for Westinghouse, Entergy and NRC Staff on June 15, 2015 regarding the apparent failure of Westinghouse and Entergy to properly file the Joint Brief publicly.

10. I declare under penalty of perjury that the foregoing is true and correct.

Executed June 18, 2015 Signed (electronically) by Brian Lusignan Assistant Attorney General Office of the Attorney General of the State of New York The Capitol Albany, New York 12224 (518) 776-2399 Brian.Lusignan@ag.ny.gov

Attachment 1 License Renewal Environmental Fatigue Screening Application Authored by Christopher Kupper and Mark Gray Paper No. PVP2014-29093 Proceedings of the ASME 2014 Pressure Vessels and Piping Conference Anaheim, California (July 20-24, 2014)

Publicly Available http://proceedings.asmedigitalcollection.asme.org/proceeding.aspx?articleID=1937745 Full-Text Copyrighted

Contains Full-Text Copyrighted Material Proceedings of the ASME 2014 Pressure Vessels & Piping Conference PVP2014 July 20-24, 2014, Anaheim, California, USA PVP2014-29093 LICENSE RENEWAL ENVIRONMENTAL FATIGUE SCREENING APPLICATION Christopher T. Kupper Mark A. Gray Westinghouse Electric Company Westinghouse Electric Company Cranberry Township, PA, USA Cranberry Township, PA, USA ABSTRACT components for 60-years of plant operation. The scope of GSI-In NUREG-1801 (GALL) Revision 0 and Revision 1, the 190 included design-basis fatigue transients, studying the US Nuclear Regulatory Commission (NRC) defined the probability of fatigue failure of selected metal components for locations evaluated in NUREG/CR-6260 as a minimum 60-year plant life. Studies related to the resolution of GSI-190 acceptable set for evaluation of environmentally assisted showed that some components have cumulative probabilities of fatigue (EAF), in addressing license renewal for nuclear plant crack initiation and through-wall growth that approach unity components. Within GALL Revision 2, the NRC revised the within the 40- to 60-year period. Therefore, it was concluded expectation, so that plants also investigate the possibility of that environmentally assisted fatigue degradation should be other locations being more limiting. To address GALL addressed in aging management programs developed for license Revision 2 and NUREG-1800 Revision 2, an EAF screening renewal as stated in the Standard Review Plan for Review of methodology was developed that considers all Safety Class 1 License Renewal Applications for Nuclear Power Plants reactor coolant pressure boundary components in major (NUREG-1800) [2].

equipment and piping systems that are susceptible to EAF, Fatigue evaluations for a sample set of components in the including those locations listed in NUREG/CR-6260. While reactor coolant pressure boundary were performed in the overall screening process steps are similar to those NUREG/CR-6260 [3], including the effects of the reactor water published by EPRI, elements of the detailed application of environment. The sample set consisted of components from some steps were performed using alternative techniques. The facilities designed by each of the four United States nuclear screening process utilized the comprehensive database of plant steam supply system vendors. For each facility, six locations component fatigue qualifications available in NSSS vendor were studied. Revisions 0 and 1 of the Generic Aging Lessons documentation, and yielded a comprehensive list of lead Learned (GALL) Report (NUREG-1801) [4] defined the indicator locations for EAF consideration. This paper describes sample set of components from NUREG/CR-6260 as the the overall process and alternate methods in the context of a minimum acceptable set for evaluation of environmentally specific plant license renewal application. assisted fatigue. However, within GALL Rev. 2, the NRC revised the expectation so that operating plants address the INTRODUCTION concern of other locations that may be more limiting than the Nuclear power plants were originally given a license to sample set. GALL Rev. 2 explicitly states in Section: X.M1 operate for 40 years. To extend the license to allow for FATIGUE MONITORING, under the Evaluation and continued operation through 60 years, evaluations are required Technical Basis item 1. Scope of Program:

to ensure that the 40-year current licensing basis remains applicable. The evaluations to justify operation through the For purposes of monitoring and tracking, applicants license renewal period include the evaluation of Time Limited should include, for a set of sample reactor coolant system Aging Analyses (TLAA). TLAAs are plant-specific safety components, fatigue usage calculations that consider the analyses that are based on explicitly assumed 40-year plant life. effects of the reactor water environment. This sample set The evaluation of metal fatigue is included as a TLAA. Metal should include the locations identified in NUREG/CR-6260 fatigue is the loss of structural integrity due to fluctuating and additional plant-specific component locations in the stresses. NRC Generic Safety Issue (GSI) 190 [1] was reactor coolant pressure boundary if they may be more established to address residual concerns regarding the limiting than those considered in NUREG/CR-6260.

NUREG-1800 Rev. 2 also states in Section 4.3.2.1.3 A methodology was also proposed in EPRI Report Environmental Fatigue Calculations for Code Class 1 1024995 to perform an EAF screening evaluation. The overall Components: method is similar to that described here. The only fundamental difference between the proposed EPRI method and the method Applicants should consider adding additional component described in this paper is in the application of the step to locations if they are considered to be more limiting than compare component fatigue usage on a common basis with those considered in NUREG/CR-6260. respect to the stress analysis methods used for qualification.

The approach described herein utilizes a large database of To determine if there were any possible locations that may component fatigue evaluations and related experience to be more limiting than the NUREG/CR-6260 locations, an establish the analysis method basis of comparison, as opposed environmental fatigue screening evaluation was performed. to a more rudimentary approach utilizing a combination of new The EAF screening evaluation is a detailed review of the analysis and approximations.

current licensing basis (CLB) fatigue evaluations for all Safety The process elements of the overall screening method are Class 1 reactor coolant pressure boundary components in major summarized below:

equipment and piping systems, including the NUREG/CR-6260 locations, to determine the lead indicator locations in terms of 1. Data Collection EAF. A report was published by EPRI (1024995) [5] to

a. All of the pertinent inputs must be collected from the document a methodology associated with performing a plant-utility. This includes all of the Safety Class 1 reactor specific EAF screening evaluation. The overall methodology coolant pressure boundary component drawings, described in this paper is similar to the EPRI report but differs materials, and current licensing basis fatigue in that it utilizes a comprehensive database of plant component evaluations.

fatigue qualifications available in NSSS vendor documentation.

2. Transient Section Considerations NOMENCLATURE

a. Determine the transient sections for all piping systems NRC - Nuclear Regulatory Commission and major equipment included in the screening GALL - Generic Aging Lessons Learned evaluation. Components within a common transient EAF - Environmentally Assisted Fatigue section are evaluated initially as a group before they EPRI - Electric Power Research Institute are compared against components in other transient NSSS - Nuclear Steam Supply System sections within the same system/equipment.

TLAA - Time Limited Aging Analysis GSI - Generic Safety Issue 3. Screening Fen Application Fen - Environmental Fatigue Correction Factor

a. Determine the maximum environmental fatigue AOR - Analysis of Record correction factor (Fen) for each transient section based CUF - Cumulative Usage Factor on material (Material Fen) and calculate a screening CUFen - CUF with Environmental Effects cumulative usage with EAF for each component P&ID - Piping and Instrumentation Diagram (Material CUFen). Components with a Material CUFen PWR - Pressurized Water Reactor

< 1.0 can be eliminated from consideration at this DO - Dissolved Oxygen point.

CLB - Current Licensing Basis FE - Finite Element b. As needed, calculate refined estimated Fen factors for T* - Transformed Temperature [6-8] each component in the transient section based on O* - Transformed Dissolved Oxygen Content [6-8] maximum section temperature (Temperature Fen) to

- Transformed Strain Rate [6-8] reduce the CUFen (Temperature CUFen) to a value S* - Transformed Sulfur Content [6-8] below 1.0. Components with a Temperature CUFen <

1.0 can be eliminated from consideration at this point.

METHOD OVERIEW

4. Stress Basis Comparison To perform the EAF screening process for a particular plant, all of the Safety Class 1 reactor coolant pressure a. Determine the level of technical rigor and qualification boundary components that are susceptible to EAF were criteria used in the stress and fatigue evaluation for reviewed and categorized into common groups for the purpose each component within the transient section.

of identifying leading locations for EAF consideration. These

b. Using a consistent stress analysis method ranking leading locations supplement those identified in NUREG/CR-basis for comparison, assign a rank to each limiting 6260, resulting in a comprehensive list of plant-specific lead component within a transient section based on the indicator locations for EAF consideration. The process information developed in step 4a.

developed is outlined below, and then described in more detail in the subsequent sections.

5. Leading Location Identification the methodology described in this paper, NUREG/CR-5704 is used for austenitic stainless steels, NUREG/CR-6583 is used

a. Those components with a Material (step 3a) or for carbon and low alloy steels, and NUREG/CR-6909 is used Temperature (step 3b) screening CUFen less than 1.0 for nickel alloy steels.

are removed from the final leading location list.

For screening, the maximum Material Fen is first calculated

b. Identify the location(s) with the maximum screening using bounding assumptions regarding the parameters used in CUFen and least conservative method of stress analysis the Fen equations from the NUREG reports (temperature, in each transient section. dissolved oxygen, sulfur content, strain rate). Generally, parameters are chosen such that the maximum Fen penalty

c. Compare components of different transient sections factor for the material is calculated, with the exception of within common systems/equipment. This may require dissolved oxygen content. A value of 0.005 ppm is used for the additional stress basis comparisons to determine one dissolved oxygen (DO) content, which is typical of the PWR or two leading locations per system/equipment. environment. For PWRs, the DO content is generally well

d. Compare candidate leading locations against any below 0.05 ppm, except for short periods during NUREG/CR-6260 locations within the system. Those heatup/cooldown operations. However, elevated DO content components with a screening CUFen less than the usually only occurs when reactor coolant temperature is low.

NUREG/CR-6260 location and with the same or lower During these periods of operation, fluid temperatures are in the analysis method ranking are removed from the final range where T* = 0 and the T*O* term in the applicable Fen set of leading locations. equation still reduces to zero.

The Material Fen is applied to the current licensing basis Each of these process elements is expanded in the (CLB) CUF to calculate a screening cumulative usage with subsequent sections of this paper. EAF for each component (Material CUFen). If required, CLB fatigue curve adjustments are considered. Components with a DATA COLLECTION Material CUFen < 1.0 are eliminated from consideration at this All of the required data necessary to perform an EAF point. The basis for elimination is that a detailed EAF screening evaluation must be obtained. This information may evaluation of these components would result in a lower CUFen include the following: than that obtained using the bounding maximum material

1. Component geometries penalty, and therefore would remain below 1.0. Eliminating

2. Component material properties such locations allows the screening comparisons to focus on the

3. Component fatigue analysis of record (AOR) remaining locations that are more limiting.

4. Plant transient characteristics and/or For the locations still remaining in the transient section, specifications refined estimated Fen penalty factors are calculated for each

5. Drawings: component within the transient section based on temperature in

a. P & IDs an effort to reduce the CUFen to a value below 1.0

b. Isometrics (Temperature Fen and Temperature CUFen). The maximum

c. Detailed component drawings Temperature Fen is calculated using the same assumptions as

6. Plant water chemistry requirements the maximum Material Fen, except that the maximum temperature for each transient section is input to the applicable TRANSIENT SECTIONS NUREG equation. Components with a Temperature CUFen <

A transient section is defined as a group of sub- 1.0 are eliminated from consideration at this point. The basis components/locations that experience the same transients. for elimination is that a detailed EAF evaluation of these Components that reside in the same transient section can be components would result in a lower CUFen than that obtained compared with each other to determine the most limiting using the bounding maximum temperature penalty, and component (or leading location). The differences in stresses therefore would remain below 1.0. Eliminating such locations experienced by each component in a transient section are allows the screening comparisons to focus on the remaining generally the result of the material and geometry differences, locations that are more limiting. If the Temperature CUFen is and possibly differing stress analysis methods. greater than 1.0, this location is retained for the next step. The Temperature CUFen is the final screening CUFen.

SCREENING FEN APPLICATION The applicable NUREG Fen penalty factor equations, as When performing an EAF evaluation, GALL Revision 2 well as the assumptions used for screening, are shown below recommends either guidance from NUREG/CR-5704 [6] for for each material.

austenitic stainless steels, NUREG/CR-6583 [7] for carbon and low alloy steels, and NUREG/CR-6909 [8] for nickel alloy Austenitic Stainless Steels (NUREG/CR-5704):

steels, or guidance from NUREG/CR-6909 for all materials.

Note that if NUREG/CR-6909 is used, its corresponding fatigue Max Fen = 15.348* when:

curves must be considered in the EAF screening process. For Service Temperature, T = 200°C (392°F) or higher*

Dissolved Oxygen, DO = 0.005 ppm When performing such an assessment, the following stress Strain Rate, = 0.0004%/sec or lower analysis characteristics are considered in determining the

below 392°F the Max Fen = 2.547 limiting locations within a given transient section:

Carbon Steels (NUREG/CR-6583): 1. Qualification Criteria (NB-3200, NB-3600, etc.)

2. Stress Analysis Method Max Fen = 1.740 for Carbon when: a. Interaction Analysis Sulfur Content, S = 0.015 weight percent or higher b. Simplified or One-Dimensional Analysis (e.g., NB-Dissolved Oxygen, DO = 0.005 ppm 3600 formula, etc.)

Service Temperature, T = 350°C (662°F) or higher* c. Finite Element Analysis Strain Rate, = 0.001%/sec or lower* i. Thermal ii. Mechanical

since O* is 0.0 for PWR conditions, S*, *, & T* are

d. Elastic/Plastic Analysis irrelevant for Carbon and LAS To perform these stress basis comparisons, a hierarchy of Low Alloy Steels (NUREG/CR-6583): stress analysis methods was developed based on fatigue analysis experience to define the relative complexity of the Max Fen = 2.455 for LAS when: various methods. In general, fatigue analysis performed to NB-Sulfur Content, S = 0.015 weight percent or higher 3200 criteria are regarded as more complex than those Dissolved Oxygen, DO = 0.005 ppm performed to NB-3600 criteria. The hierarchy used is presented Service Temperature, T = 350°C (662°F) or higher* below, ordered from the least complex to the most complex Strain Rate, = 0.001%/sec or lower* methods within typical NB-3200 and NB-3600 analyses. Note

since O* is 0.0 for PWR conditions, S*, *, & T* are that combinations of the various methods are assessed on a case irrelevant for Carbon and LAS by case basis.

Ni-Cr-Fe- Alloys (NUREG/CR-6909): 1) Standard NB-3600 analysis

2) NB-3600 with mechanical FE stress quantities substituted in stress formulas Max Fen = 4.524* when:

3) NB-3600 with thermal FE stress quantities substituted in Service Temperature, T = 325°C (617°F) or higher*

stress formulas Dissolved Oxygen, DO = 0.005 ppm

4) Combination of 2) and 3)

Strain Rate, = 0.0004%/sec or lower

5) NB-3200 Fatigue Analysis:

Below 617°F, T* is a function, and maximum Fen is a. NB-3200 with interaction analysis calculated as a function of the maximum temperature of the b. NB-3200 with elastic FE analysis transient section. c. NB-3228 Plastic analysis STRESS BASIS COMPARISON Note that this hierarchical list is used primarily for the A major consideration in the comparison process for a stress-based comparison of the piping components. Since the plant-wide screening of this nature is the fact that different majority of the components associated with the equipment are stress analysis techniques may have been used for each performed using an NB-3200 analysis, the equipment component CLB usage factor calculation. For example, components are ranked using an independent ranking system, presume there is a piping component that was analyzed using similar to category 5) above, within each piece of equipment, NB-3600 analysis methods and yielded a usage factor of 0.9. based on the amount of conservatism in the analysis.

Also, presume there is another component in the same transient In executing the stress basis comparison, elimination of the section that had a usage factor of 0.9, but was qualified using location with the lower final screening CUFen value and NB-3200 plastic analysis methods. Although both of these analysis method ranking is justified, since, if it were analyzed locations have the same usage factor, the amount of technical with the same rigor as the retained location, its CUF would be rigor that was applied to the second component far exceeds that even lower, and result in an even lower CUFen. For cases of the first component. Therefore, the screening method must where a clear qualitative assessment cannot be completed, it consider the various stress analysis methods and techniques that may be required to perform additional analyses to determine were used in the usage factor evaluation. which component is actually the most limiting for a given The limiting locations for each of the transient sections transient section or system. However, it is not expected that within a system/equipment are initially compared against each many of these cases will exist during the screening process.

other only and are not yet compared against other components Generally, within a transient section there are only one or of different piping systems or equipment. This reduces the two components at most where more advanced stress analysis need to perform a plant-wide stress basis comparison of all methods were used relative to the other components in the same components at the outset.

transient section. Therefore, determining the most limiting stress analysis characteristics outlined previously.

location within a transient section is usually not a difficult task, since the component where the most technical rigor was applied Stage 3: Comparison of System/Equipment Level Leading is often the limiting location within a transient section. Note Locations with NUREG/CR-6260 locations that components in a common transient section will have Stage 3 (Step 5d from Method Overview Section) involves similar screening Fen correction factors applied since they comparing the leading locations to the NUREG/CR-6260 experience the same transients. locations. For those systems where the NUREG/CR-6260 However, comparisons made between components in the locations have the highest screening CUFen, no additional same system/equipment, but different transient sections, must locations need to be considered for EAF. For those systems consider the potential differences in the Fen correction factors. where the NUREG/CR-6260 locations do not have the highest These comparisons must also consider the previously discussed screening CUFen or a NUREG/CR-6260 location does not exist, stress basis comparison characteristics in determining the the locations within that system that have the highest analysis ultimate leading location for each system/equipment. ranking and screening CUFen should be considered as leading locations. In cases where multiple locations could be the LEADING LOCATION IDENTIFICATION leading location, additional analyses may need to be performed (as stated in Stages 1 and 2) to determine the most limiting Stage 1: Determine Transient Section Leading Locations location, or each of the locations should be considered for Stage 1 (Steps 5a and 5b from Method Overview Section) further EAF management.

has two aspects: Screening Fen application and transient section The screening process yields a comprehensive and plant-stress basis comparisons. The ultimate goal of Stage 1 is to specific list of locations that supplement those evaluated in determine the leading locations for each transient section. NUREG/CR-6260 as lead indicator locations for EAF Once the appropriate screening Fen factors are calculated consideration. The final set of leading locations should then be for the components in each transient section, they can be included in the plants fatigue management program to ensure applied to those components. Those components with a that the CUFen for each component remains below 1.0 for the screening CUFen of less than 1.0 can be removed from the list duration of plant operation.

as explained previously.

A stress basis comparison must be performed on remaining APPLICATION OF METHODOLOGY transient section components. Typically, a stress basis The EAF screening methodology described above was comparison will yield one or two distinct leading locations. If a applied to a plant with major equipment and piping designed to further reduction in the number of leading locations is required, ASME Code Section III. An example of the results of applying then a more detailed quantitative analysis can be performed. the process for one piping system is provided below. The The remaining components represent the leading locations for example system is the cold leg safety injection/accumulator EAF consideration for each transient section. piping in a Westinghouse PWR, as shown in Figure 1.

Upon gathering the CLB data, the system transient sections Stage 2: Determine System/Equipment Level Leading were defined consistent with the CLB transient definitions and Locations component fatigue evaluations.

Stage 2 (Step 5c from Method Overview Section) involves components that reside in different transient sections, but are within a common system or piece of major equipment. These can also be compared to determine one or two leading locations to represent their respective system/equipment. Often, it is the transients themselves that control which components have the highest usage factors in a given system. So within a particular system, those transient sections with the most severe system transients will usually have components with the highest usage factors. However, the comparison of components in different transient sections must be done after the appropriate Fen correction factor is applied to the component usage factor, and stress basis ranking is applied. This is because the Fen correction factor is dependent on temperature and strain rate and therefore can vary for each transient section.

The ultimate goal is to compare those components within a FIGURE 1: SAFETY INJECTION/ACCUMULATOR -

common system against each other to determine the leading TRANSIENT SECTIONS locations on a system/equipment level. Similarly to Stage 1, when comparing components of different transient sections, but of common systems/equipment, it is necessary to consider the 5 Copyright © 2014 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 05/07/2015 Terms of Use: http://asme.org/terms

Tables 1 through 5 show the components in each section and illustrate the results of applying Steps 3 through 5b from the Method Overview Section. The Analysis Ranking values TABLE 3: SAFETY INJECTION/ACCUMULATOR EAF correspond to the hierarchical list as described in the Stress SCREENING - TRANSIENT SECTION 3 Material Basis Comparison Section. (1=SS, Max Final EAF Temp >

Design 2=CS, Material Material Temp Temp Screening Analysis Section Component 392°F CUF 3=LAS, Fen CUFen Fen CUFen Ranking TABLE 1: SAFETY INJECTION/ACCUMULATOR EAF (1=yes, CUFen 4=Ni 2=no)

SCREENING - TRANSIENT SECTION 1 Alloy)

Material Elbow 0.002 1 15.35 0.03 Max 10" x 3/4" (1=SS, Final EAF 0.086 1 15.35 1.33 2 2.547 0.22 Temp > Branch Design 2=CS, Material Material Temp Temp Screening Analysis 3 Section Component 392°F Valve Butt CUF 3=LAS, Fen CUFen Fen CUFen Ranking 0.001 1 15.35 0.01 (1=yes, CUFen Weld 4=Ni 2=no) Valve 0.19 1 15.35 2.92 2 2.547 0.48 Alloy)

RCL 0.95 1 15.35 14.58 1 15.35 14.58 14.58 4 Nozzle TABLE 4: SAFETY INJECTION/ACCUMULATOR EAF Elbow 0.0875 1 15.35 1.34 1 15.35 1.34 1.34 1 Butt Weld 0.099 1 15.35 1.52 1 15.35 1.52 1.52 1 SCREENING - TRANSIENT SECTION 4 1 Small 0.099 1 15.35 1.52 1 15.35 1.52 1.52 1 Material Branch/Plug Max (1=SS, Final EAF Valve Butt Temp >

0.5368 1 15.35 8.24 1 15.35 8.24 8.24 3 Design 2=CS, Material Material Temp Temp Screening Analysis Weld Section Component 392°F CUF 3=LAS, Fen CUFen Fen CUFen Ranking Valve 0.48 1 15.35 7.37 1 15.35 7.37 7.37 1 (1=yes, CUFen 4=Ni 2=no)

Alloy)

As shown in Table 1, all components have a screening Elbow 0.007 1 15.35 0.11 6" x 2" CUFen > 1.0 after the maximum Material and Temperature Branch 0.243 1 15.35 3.73 2 2.547 0.62 4

specific Fen penalty factors were applied. All components were Valve Butt Weld 0.013 1 15.35 0.2 analyzed using NB-3600 equations (analysis ranking = 1) Valve 0.29 1 15.35 4.45 2 2.547 0.74 except for the RCL Nozzle and the valve butt weld. The section 1 RCL nozzle was qualified to NB-3600 but using TABLE 5: SAFETY INJECTION/ACCUMULATOR EAF Finite Element Analysis for thermal and mechanical stress SCREENING - TRANSIENT SECTION 5 Material quantities (analysis ranking = 4). The section 1 valve butt weld (1=SS, Max Temp > Final EAF was qualified to NB-3600 but using Finite Element Analysis Section Component Design 2=CS, Material Material 392°F Temp Temp Screening Analysis CUF 3=LAS, Fen CUFen Fen CUFen Ranking only for thermal stress quantities (analysis ranking = 3). The 4=Ni (1=yes, CUFen 2=no) final screening CUFen for the section 1 valve butt weld is less Alloy) than that of the section 1 RCL Nozzle, and the section 1 valve Socket Weld 0.1 1 15.35 1.54 2 2.547 0.25 5

butt weld was qualified using a less rigorous analysis Pipe 0.016 1 15.35 0.25 Elbow 0.1 1 15.35 1.54 2 2.547 0.25 methodology than the section 1 RCL Nozzle. Therefore, it is concluded that the section 1 RCL Nozzle is more limiting than the section 1 valve butt weld. The RCL Nozzle is chosen as the As shown in Tables 3, 4, and 5, all components in transient limiting location from section 1. sections 3, 4, and 5 have a screening CUFen < 1.0 after maximum material and temperature specific Fen penalty factors TABLE 2: SAFETY INJECTION/ACCUMULATOR EAF were applied. Therefore, all components in transient sections 3, SCREENING - TRANSIENT SECTION 2 4, and 5 can be eliminated from consideration. The basis for Material Max elimination is that a detailed EAF evaluation of these (1=SS, Design 2=CS, Material Material Temp > Temp Temp Final EAF components would result in a lower CUFen than that obtained Section Component 392°F Screening Analysis CUF 3=LAS, Fen CUFen (1=yes, Fen CUFen CUFen Ranking using the bounding maximum material penalty, and therefore 4=Ni Alloy) 2=no) would remain below 1.0.

Valve Butt 0.091 1 15.35 1.4 2 2.547 0.23 Applying Step 5c for the five transient sections in the cold Weld 2

Elbow 0.061 1 15.35 0.94 leg safety injection accumulator system, only two potential 10" x 10" x 6" Tee 0.093 1 15.35 1.43 2 2.547 0.24 leading locations remain. The final screening CUFen for the Valve 0.48 1 15.35 7.37 2 2.547 1.22 1.22 1 section 2 valve is less than that of the section 1 RCL Nozzle, and the section 2 valve was qualified using a less rigorous In Table 2 for section 2, with the exception of the valve, all analysis methodology (lower analysis ranking) than the section components have a screening CUFen < 1.0 after maximum 1 RCL Nozzle. Therefore, it is concluded that the section 1 Material and Temperature specific Fen penalty factors were RCL Nozzle is more limiting than the section 2 valve. The applied. The valve was qualified to NB-3545 (analysis ranking RCL Nozzle is chosen as the limiting location for the system.

= 1). Because it is the only location remaining, the valve is The final step (5d) applies for this system, since the RCL chosen as the limiting location from section 2. Nozzle is a NUREG/CR-6260 location. The step is trivial in this example since the nozzle is the final leading location for the system.

Similar methodology would be used for performing an comparisons are based on common data and experience from EAF screening evaluation with piping not designed to ASME plant component fatigue evaluation methods and results. The Code Section III. The difference is in the data collection step. approach can also be supplemented by the NSSS vendor fatigue Because the current licensing basis for the Safety Class 1 evaluation database for similar components if required. The list piping does not include an ASME Code Section III fatigue of leading locations can be further reduced without detailed evaluation, plant-specific usage factors are not available for analysis by removing conservatism from the fatigue analysis of each piping component. Instead, the comprehensive database record (including application of 60-year projected cycles) or of plant component fatigue qualifications available in NSSS further refinement of the Fen penalty factors. The final list of vendor documentation is utilized. Comparisons are made leading locations should be included in the aging management between the plant-specific piping components and those which program for the plant.

are available in the database to justify the applicability and make relative comparisons for screening purposes. Such REFERENCES comparisons include materials, geometry, and transients. [1] Thadani, A. C., 1999, GENERIC SAFETY ISSUE 190, The final result of the process applied to all plant Class 1 FATIGUE EVALUATION OF METAL COMPONENTS systems and components is a list of locations requiring further FOR 60-YEAR PLANT LIFE, GSI-190, NRC Office of evaluation and/or inclusion in the fatigue management Nuclear Regulatory Research, Washington, DC.

program. This list includes the NUREG/CR-6260 locations [2] U.S. Nuclear Regulatory Commission, 2010, plus any additional screened-in locations. STANDARD REVIEW PLAN FOR REVIEW OF LICENSE RENEWAL APPLICATIONS FOR NUCLEAR PHASE 2 EAF SCREENING POWER PLANTS, NUREG-1800, Division of License The EAF screening process described above can be Renewal, Office of Nuclear Reactor Regulation, considered a first-pass evaluation (Phase 1) designed to identify Washington, DC.

the potential locations of concern for further EAF analysis. The [3] Ware, A. G., Morton, D. K., and Nitzel, M. E., 1995, next step of the EAF screening process (Phase 2) is to perform APPLICATION OF NUREG/CR-5999 INTERIM a more refined EAF evaluation for the identified leading FATIGUE CURVES TO SELECTED NUCLEAR POWER locations if required. PLANT COMPONENTS, NUREG/CR/6260, Idaho The goal of Phase 2 is to minimize the number of detailed National Engineering Laboratory, Idaho Falls.

analyses (e.g., finite element or integrated CUFen calculations) [4] U.S. Nuclear Regulatory Commission, GENERIC required for the identified leading locations. For this step, the AGING LESSONS LEARNED (GALL) REPORT, screening CUFen values can potentially be reduced below 1.0 NUREG-1801 All Revisions, Division of License Renewal, by using one (or a combination) of the following methods: Office of Nuclear Reactor Regulation, Washington, DC.

[5] Electric Power Research Institute, 2012,

1. Application of projected transient cycles for 60 years ENVIRONMENTALLY ASSISTED FATIGUE

2. Reduction of conservatism in existing stress SCREENING PROCESS AND TECHNICAL BASIS FOR calculations (without performing any new detailed IDENTIFYING EAF LIMITING LOCATIONS, EPRI analyses) Report No. 1024995, Structural Integrity Associates, Inc,

3. Comparison with similar component detailed analyses San Jose.

Fen refinement [6] Chopra, O. K., 1999, EFFECTS OF LWR COOLANT ENVIRONMENTS ON FATIGUE DESIGN CURVES OF CONCLUSIONS AUSTENITIC STAINLESS STEELS, NUREG/CR-5704, Revision 2 of the NRC GALL report required plants Argonne National Laboratory, Argonne.

applying for license renewal to consider the effects of reactor [7] Chopra, O. K. and Shack, W. J., 1998, EFFECTS OF water environment on fatigue for the sample set of components LWR COOLANT ENVIRONMENTS ON FATIGUE defined in NUREG/CR-6260 as well as any other reactor DESIGN CURVES OF CARBON AND LOW-ALLOY coolant pressure boundary component(s) that may be more STEELS, NUREG/CR-6583, Argonne National limiting. To determine if there are any locations more limiting Laboratory, Argonne.

than the NUREG/CR-6260 sample set, a plant-wide [8] Chopra, O. K. and Shack, W. J., 2007, EFFECT OF LWR environmental fatigue screening evaluation has been described. COOLANT ENVIRONMENTS ON THE FATIGUE LIFE The ultimate goal of the EAF screening evaluation was to OF REACTOR MATERIALS, NUREG/CR-6909, compare all Safety Class 1 reactor coolant pressure boundary Argonne National Laboratory, Argonne.

components within common systems to determine the leading locations with respect to EAF. Components within each system were compared on the bases of common transients and common stress analysis methods to determine the most limiting locations. A systematic method for comparison of components and stress analysis methods has been described. The 7 Copyright © 2014 by ASME Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 05/07/2015 Terms of Use: http://asme.org/terms

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Environmentally Assisted Fatigue Screening: Process and Technical Basis for Identifying EAF Limiting Locations.

EPRI, Palo Alto, CA: 2012.

1024995.

iii

Abstract This report provides the technical basis and process for a screening evaluation of a nuclear power plant. This screening will identify appropriate limiting locations for systematic monitoring of the environmentally assisted fatigue (EAF) effects in a Class 1 reactor on the reactor coolant pressure boundary components that are wetted with primary coolant. Use of this process will ensure that the most limiting locations for EAF are determined on a consistent basis.

The process developed in this report provides guidance for the evaluation and relative ranking of estimated Uen values for locations in components and systems where EAF is a concern in order to minimize the possibility of the need for a formal fatigue evaluation.

The estimated values are compared to Uen values for locations in each given system/component that have been specifically identified in regulatory guidance as being of concern for EAF. Locations from previous guidance are to be monitored, and if estimated values for other locations are higher or as high, the other locations should also be monitored. For components or systems where there are no locations included in previous regulatory guidelines, the recommendation is to monitor up to three locations with the highest estimated Uen.

This report is a public document available for reference by license renewal applicants.

Keywords Design basis Environmentally assisted fatigue Fatigue monitoring Fatigue usage License renewal v

Acronyms The following acronyms are used in this report.

ASME American Society of Mechanical Engineers BWR Boiling Water Reactor CS Carbon Steel CVCS Chemical and Volume Control System CUF Cumulative Usage Factor DSR Design Stress Report DO Dissolved Oxygen EAF Environmentally Assisted Fatigue EPRI Electric Power Research Institute Fen Environmental Fatigue Correction Factor FMP Fatigue Management Program FSRF Fatigue Strength Reduction Factor FSAR Final Safety Analysis Report HWC Hydrogen Water Chemistry ID Inside Diameter LRA License Renewal Application LWR Light Water Reactor LAS Low-Alloy Steel NRC Nuclear Regulatory Commission NWC Normal Water Chemistry OD Outside Diameter P&ID Piping and Instrumentation Drawing PWR Pressurized Water Reactor RCS Reactor Coolant System RHX Regenerative Heat Exchanger RAI Request for Additional Information SS Stainless Steel Sm Design Stress Intensity Value SCF Stress Concentration Factor U ASME Code Cumulative Usage Factor Uen EAF Cumulative Usage Factor vii

Table of Contents Section 1: Introduction ............................................1-1 Background ..................................................................... 1-1 Purpose of this Report ....................................................... 1-1 Criteria for Process ........................................................... 1-1 Benefits to Plant Owners and NRC ..................................... 1-2 License Renewal ......................................................... 1-2 Section 2: Fatigue and EAF Basics ............................2-1 Definition of Terms ........................................................... 2-1 Basics of Fatigue Analysis ................................................. 2-2 Background on EAF Calculations ....................................... 2-3 Fen Formulations .......................................................... 2-4 Maximum Fen Values ................................................... 2-6 Requirements for License Renewal ...................................... 2-7 Section 3: Process Outline and Technical Basis .........3-1 Outline of Process Steps .................................................... 3-1 Process Development Assumptions and Characteristics ......... 3-3 Determining Thermal Zones ............................................... 3-4 Technical Basis for Fen Estimation Evaluation ........................ 3-5 Technical Basis for Common Basis Stress Evaluation ............. 3-7 Rationale for the Procedure .......................................... 3-7 Formulas and Equations .............................................. 3-9 Comparison of Common Basis Stress Evaluation Screening Rules to NB-3600 Evaluation Process ........... 3-11 Limitations and Assumptions of the Process .................. 3-13 Section 4: The Screening Process .............................4-1 Detailed Screening Procedure............................................ 4-1 Gather Required Inputs for all Systems Containing Class 1 Reactor Coolant Pressure Boundary Components ............................................................... 4-1 Determine Thermal Zones for Each System ..................... 4-2 Identify Materials and Candidate Locations, .................. 4-4 Calculate Uen* for Each Candidate Location .................. 4-4 Uen Estimation Evaluation Procedure ................................... 4-5 Fen Formulations .......................................................... 4-5 ix

Determine Input Values ................................................ 4-6 Fen Estimation Evaluation Procedure .................................... 4-6 Final Step for Uen Estimation Evaluation Procedure .......... 4-6 Common Basis Stress Evaluation Procedure ......................... 4-6 Determine Input Values ................................................ 4-7 Perform Stress Evaluation ............................................. 4-7 Perform Fatigue Evaluation........................................... 4-8 Evaluate Next Candidate Location ................................ 4-9 Guidelines for Reducing Number of Sentinel Locations ................................................................. 4-10 Section 5: Pilot Plant Application .............................5-1 Section 6: Concluding Remarks................................6-1 Section 7: References ..............................................7-1 Appendix A: Summary of Fen Formulations as Accepted by the NRC .............................. A-1 Fen Formulations for Ferritic Materials ................................. A-1 NUREG/CR-6583 (Old Rules) ..................................... A-1 NUREG/CR-6909 (New Rules) ................................... A-2 Fen Formulations for Austenitic Stainless Steel Materials ........ A-3 NUREG/CR-5704 (Old Rules) ..................................... A-3 NUREG/CR-6909 (New Rules) ................................... A-3 Fen Formulations for Nickel Alloy Materials ......................... A-4 NUREG/CR-6909 (New Rules) ................................... A-4 Dissolved Oxygen (DO) .............................................. A-4 Strain Rate ( )........................................................... A-4 Appendix B: Rules for Evaluation of Class 1 Piping .....................................................B-1 Rules for Evaluation of Class 1 Piping ................................. B-1 x

List of Figures Figure 3-1 Determination of Transient Stresses for Ramp Transients ...................................................................... 3-11 Figure 4-1 Screening Flow Chart ........................................... 4-13 Figure 4-2 Estimated Stress / CUF Evaluation ......................... 4-14 xi

List of Tables Table 2-1 Fen Equation Parameters ........................................... 2-6 Table 3-1 Strain Rate Categories ............................................. 3-6 Table 5-1 Results from Pilot PWR Plant Evaluation Using Fen Estimation Evaluation Procedure......................................... 5-3 Table 5-2 Final Sentinel Locations for PWR Pilot Plant ................ 5-7 Table 5-3 Sample Results from Pilot PWR Plant Evaluation Using Common Basis Stress Evaluation Procedure (Properties) ...................................................................... 5-8 Table 5-4 Sample Results from Pilot PWR Plant Evaluation Using Common Basis Stress Evaluation Procedure (Transient Details) ............................................................. 5-9 Table 5-5 Sample Results from Pilot PWR Plant Evaluation Using Common Basis Stress Evaluation Procedure (Computed Values) ......................................................... 5-10 xiii

Section 1: Introduction

Background

The NRC requires license renewal applicants to assess the fatigue usage effects from a reactor water environment and demonstrate acceptable fatigue cumulative usage factors (CUF) with the effects of a reactor water environment considered for Class 1 components for the entire period of extended operation. This is commonly referred to as environmentally assisted fatigue (EAF).

This demonstration requires an EAF evaluation and screening of Class 1 components, some of which lack a CUF calculation. Evaluating Class 1 components with a CUF calculation is straightforward and may require a relatively simple evaluation process to estimate and apply environmental fatigue correction factors (Fen) to the existing CUF values. Class 1 components without CUF calculations may require a more extensive evaluation process to evaluate plant locations on a similar stress basis and apply Fen factors.

Purpose of this Report The purpose of this report is to describe the technical basis of and define a process that may be used for EAF screening and ranking of components in nuclear power plant Class 1 systems. This example process is documented to allow for consistent application, but it is not the only acceptable way to identify limiting locations. This process must be effective for PWRs and BWRs, both with ASME Section III [1] / B31.7 [2] piping and B31.1 [3] piping. This process can be used to screen plant locations in order to rank them on the basis of EAF values. These ranked locations can then be compared to the NUREG/CR-6260 [4] sample locations and may augment a plants Fatigue Management Program (FMP).

The desired outcome of this process is to determine plant locations which can be demonstrated to bound other locations of like materials and can serve as limiting EAF locations for the plant.

Criteria for Process The procedure developed for this report has the following properties:

No need for new formal stress or fatigue analysis 1-1

Includes procedures that are practical to use, with readily available design input Provides appropriate relative EAF rankings of components Allows the use of either NUREG/CR-5704 [5] (stainless steel)/

NUREG/CR-6583 [6] (carbon and low alloy steel)/ NUREG/CR-6909 [7]

(Ni-Cr-Fe) or just NUREG/CR-6909 for all materials Benefits to Plant Owners and NRC License Renewal This process will enable plant owners to demonstrate knowledge of the locations in their plant that can serve as limiting locations for EAF evaluations as the plants enter the period of extended operation. This process provides the rationale for selecting these bounding locations. Plant owners will minimize costs by avoiding the necessity of formal fatigue analyses, while meeting the regulatory requirements of determining the bounding EAF locations in the plant.

The process will provide the NRC with a uniform approach to determination of limiting locations for EAF evaluations in license renewal applications.

1-2

Section 2: Fatigue and EAF Basics This section provides a definition of terms used in this report and a background on fatigue analysis and license renewal requirements.

Current plants are qualified via a design process defined in ASME Code,Section III [1] or B31.7 [2] or B31.1 [3]. For components subject to cyclic loadings, this generally includes a fatigue analysis. The ASME Code,Section III and B31.7 design codes provide rules for the explicit determination of CUF. The B31.1 rules provide rules for evaluation of cyclic loads using a cyclic reduction factor method that addresses sustained stresses and cyclic thermal moment stresses, but does not produce CUF values.

Evaluating components with a CUF calculation is straightforward and may require a relatively simple evaluation process to estimate and apply environmental fatigue correction factors (Fen) to the existing CUF values. If necessary, components without CUF calculations may require a more extensive evaluation process to evaluate plant locations on a similar stress basis and apply Fen factors.

Definition of Terms U: Design CUF (documented in Design Stress Reports (DSR) and design analyses)

Fatigue Table: The compilation of incremental CUF values (Uincr) determined from load pairs in a fatigue analysis (found in DSRs and design analyses)

Load Pair: A row in a fatigue table representing a local maximum and minimum value of stress.

U*: Estimated CUF produced on a common basis (produced in the Common Basis Stress Evaluation Procedure)

U6260*: Estimated CUF produced on a common basis for NUREG/CR-6260 location (produced in the Common Basis Stress Evaluation Procedure)

Uincr*: Estimated incremental CUF for a load pair (produced in the Common Basis Stress Evaluation Procedure) 2-1

Fen*: Estimated Fen produced by an evaluation that does not evaluate the Fen value for each Load Pair in a fatigue analysis (produced in the Fen Estimation Procedure)

Uen*: Estimated Uen produced from Estimated CUF and Estimated Fen Uen incr*: Estimated incremental Uen* for a load pair produced from Uincr* and Fen*

values for each transient pair (produced in the Common Basis Stress Evaluation Procedure as Uen incr* = Fen* x Uincr*)

Uen 6260*: Estimated Uen produced from Estimated CUF for a NUREG/CR-6260 location (produced in the Common Basis Stress Evaluation Procedure as Uen 6260*

= Fen* x U6260*)

Uen max*: The maximum value of Uen* in a set of locations Uen max-1*: The second highest value of Uen* in a set of locations Uen max-2*: The third highest value of Uen* in a set of locations Leading Transient: A thermal or pressure transient that contributes significantly to CUF of a component. Prime examples are temperature shocks from starting and stopping flow at nozzles.

Bundled Transients: Enveloping of multiple plant transients by one conservative plant transient.

Common Basis Model: A model in which components are assessed on a common basis for ranking comparison purposes by evaluating load pairs with unbundled transients on a linear elastic basis.

Basics of Fatigue Analysis According to the ASME Code [1], the CUF is a value computed as the summation of incremental fatigue contributions arising from thermal and mechanical stress fluctuations in a metal component. The CUF is compared to a maximum value of 1.0 to demonstrate acceptable design behavior.

As specified in the ASME Code, the fatigue analysis procedure consists of the following steps:

Determine the stress tensor (composed of six stress components) for all normal service conditions. Select one or more controlling component locations for evaluation. For piping components, the analysis is typically conducted for all welds and fitting locations.

Determine the stress differences (i.e. stress intensity ranges) for all pair-wise combinations of service conditions.

Determine the alternating stress amplitude (Salt) for each transient pair. This must include stress concentration effects, if present. Also any additional 2-2

strain due to plasticity must be accounted for. For newer Code editions, one way to do this is with a local plastic strain "penalty" factor Ke, which is calculated in accordance with Section III, NB-3228.5 if the primary plus secondary stress range exceeds three times the design stress intensity (3Sm).

Determine the allowable cycles (Na) for each transient pair, using a fatigue curve. Prior to entering the fatigue curve (graph showing allowable cycles at various Salt values) to determine the allowable number of cycles, the stress amplitude must be modified by multiplying it by the ratio of the modulus of elasticity on the fatigue curve, divided by that used in the stress analysis.

Determine fatigue usage contribution for all transient pairs where the modified stress amplitude exceeds the endurance limit of the fatigue curve.

The fatigue endurance limit is the value of Salt corresponding to an infinite number of allowable cycles Na (taken as either 106 or 1011 cycles depending upon material and edition of the ASME Code).

The fatigue usage factor, u, is determined by the summation:

M ni U = Equation 2-1 i =1 Ni where:

ni = number of stress cycles for transient pair i Ni = number of allowable cycles for the transient pair i M = total number of sets of transient pairs In the fatigue analysis process, a component meets the design criterion if the maximum CUF for all analysis locations is less than or equal to 1.0.

Background on EAF Calculations Over the past two decades, concerns about EAF have arisen because the rules for design of Class 1 components in nuclear power plants do not explicitly address the effects of light water reactor (LWR) coolant environments, and these deleterious effects may be significant. Recent laboratory work has identified the influence of key parameters on fatigue crack initiation and has established the effects of these key parameters on the fatigue life of selected carbon and low alloy steels, austenitic stainless steels and nickel alloy steels used in nuclear plants [7].

With respect to these parameters, an environmental fatigue correction factor (Fen) approach has been developed to evaluate the effects of LWR environments into ASME Section III fatigue evaluations.

NRC report NUREG-1801, Revision 2, [8] the Generic Aging Lessons Learned (GALL) Report, identifies acceptable aging management programs for fatigue and cyclic operation for the period of extended operation. It describes a process for assessing the impact of the reactor coolant environment on a set of 2-3

sample critical components for the plant, examples of which are identified in NUREG/CR-6260 [4].

NRC guidance in NUREG/CR-5704 [5], NUREG/CR-6583 [6] and NUREG/CR-6909 [7] defines the Fen correction factor as the ratio of the number of stress cycles in air at room temperature (Nair,RT ) to that in reactor water at the service temperature (Nwater ). This ratio is used because fatigue usage for one stress cycle is the inverse of its allowable number of cycles (1/N), thus:

Fen = N air , RT / N water Equation 2-2 Environmental effects are incorporated into ASME NB-3200 [1] fatigue analyses by multiplying the partial ASME usage factors for each stress cycle by the Fen correction factor computed for that stress cycle. For example, given n different stress cycle pairs, the cumulative environmental fatigue usage is:

U en = U 1 Fen ,1 + U 2 Fen , 2 + U 3 Fen ,3 + U i Fen ,i ... + U n Fen ,n

[7, Eq. A.20] Equation 2-3 where:

Ui = computed fatigue usage using the air fatigue curve for the ith stress cycle Fen,i = computed Fen for the ith stress cycle An effective Fen factor is computed as follows:

Feneffective = U en / U Equation 2-4 Where U is:

U = Un Equation 2-5 n

Fen Formulations Per Reference [8], plants have the option of computing Uen in accordance with guidance from EITHER NUREG/CR-5704 [5] (for austenitic stainless steels),

NUREG/CR-6583 [6] (for carbon and low alloy steels) and NUREG/CR-6909

[7] (for nickel alloy steels), OR NUREG/CR-6909 (for all materials).

In the case of NUREG/CR-5704 and NUREG/CR-6583 for austenitic stainless steel and carbon/low-alloy steel, respectively, the ASME Code fatigue curve is used, and the Fen factors are applied to the ASME Code fatigue usage values.

When using NUREG/CR-6909 rules, the special fatigue curve provided in Appendix A, page A.5 of Reference [7] must be used for austenitic stainless steel 2-4

materials, and optionally for ferritic materials (the ASME curve is more conservative for ferritic materials).

A generic equation for computing Fen factors is:

Fen = exp( A ( B x S

x T

x O

x * )) Equation 2-6 Parameters used in these equations are presented in Table 2-1 and further discussed in Appendix A.

A and B - Numeric constants derived from statistical analysis; the specific values depend on which material is involved and which NUREG is being used; see Table 2-1.

S* - a factor based on Sulfur Content (this term is only used for some materials in all other cases, use S* = 1).

T* - a factor based on Service Temperature.

O* - a factor based on Dissolved Oxygen.

- a factor based on Strain Rate.

2-5

Table 2-1 Fen Equation Parameters NUREG/

Material CR-A B S* Fatigue Curve Carbon Steel 6583 0.554(1) 0.101 In equation ASME Low-Alloy Steel 6583 0.898(1) 0.101 In equation ASME Austenitic 5704 0.935 1 Not in equation ASME Stainless Steel Ni-Cr-Fe Alloy 5704 0.935 1 Not in equation ASME (2)

Carbon Steel 6909 0.632 0.101 In equation Fig. A.1 (6909)

Low-Alloy Steel 6909 0.702 0.101 In equation Fig. A.2 (6909)

Austenitic 6909 0.734 1 Not in equation Fig. A.3 (6909)

Stainless Steel Ni-Cr-Fe Alloy 6909 0 1 Not in equation Fig. A.3 (6909) (3)

(1) Equation evaluated at a Test Temperature of room temperature (25°C) [6].

(2) Guidance prior to GALL, Revision 2 [8] allowed use of austenitic stainless steel equation and ASME fatigue curve for Ni-Cr-Fe steels.

(3) GALL, Revision 2 [8] cites only this option for Ni-Cr-Fe steels.

Service Temperature - temperature of the metal in contact with the primary fluid environment. Per equation rules, the temperature is in °C. For all materials, higher temperature = higher Fen (up to a maximum value).

Test Temperature - temperature during the tests (25°C).

Dissolved Oxygen - level of dissolved oxygen (DO) in the primary fluid.

Strain Rate - rate of strain in the metal during the increasingly tensile time periods. This is the parameter that has the largest effect on the value of Fen for all materials.

Sulfur Content-the weight percent of sulfur in the material.

Strain Amplitude Threshold - a minimum strain amplitude below which LWR environments have no effect on the fatigue life of these steels.

Maximum Fen Values In cases where parameters are uncertain, it is conservative to use the following worst-case values for the following parameters:

NUREG/CR-6909 [7]

For carbon steel and low-alloy steel:

2-6

S= 0.015 weight percent or higher DO = 0.5 ppm or higher T= 350°C or higher

= 0.001%/sec or lower For austenitic stainless steels and Ni-Cr-Fe alloys:

T= 325°C or higher

= 0.0004%/sec or lower NUREG/CR-5704 [5]

For austenitic stainless steel:

T= 200°C or higher DO = 0.05 ppm or higher

= 0.0004%/sec or lower NUREG/CR-6583 [6]

For carbon steel and low-alloy steel:

S= 0.015 weight percent or higher DO = 0.5 ppm or higher T= 350°C or higher

= 0.001%/sec or lower Requirements for License Renewal Part 54 to Title 10 of the U.S. Code of Federal Regulations (10CFR54) specifies the Requirements for Renewal of Operating Licenses for Nuclear Power Plants.

NUREG/CR-6260 [4], published in 1995, established a set of six locations (by plant model and vintage) that were expected to be representative of components that had higher CUFs and/or were important from a risk perspective. It made an explicit assumption that if those sample locations could be shown to have acceptable EAF values, then it would be possible to demonstrate the same for other similar locations in the plant.

2-7

The NRC has recently questioned whether the NUREG/CR-6260 locations effectively cover all locations in the plant (see Requests for Additional Information (RAIs) on recent license renewal applications and GALL Revision 2

[8]). This concern has led to requests for plants to demonstrate the validity of the NUREG/CR-6260 locations, or else augment the list with additional locations to cover any outliers.

Determination of this list is not as easy as multiplying each design CUF value by a factor or factors. Examples of the complicating factors are:

Not all CUF values represent the same degree of analytical rigor.

- Analysis of design severity plant transients produces different CUF values for a component than analysis of actual severity plant transients.

- Analysis using bundled transients yield significantly higher CUF values than analyses of the same component with un-bundled transients For a given plant transient, Fen factors often will trend counter to the computed CUF values, thus potentially complicating the ranking of the Uen values for a component.

- Faster rise times for a thermal transient will tend to produce lower Fen factors, but larger CUF values. Since Uen = Fen x U, the product of the two is not known a priori without further analysis.

Analysis of design numbers of plant transients can yield different rankings of CUF and Uen values than analyses of projected numbers of plant transients.

- The two different mixes of plant transients, each with their unique transient characteristics, can cause the weighted Fen factors and Uen values to vary significantly.

Different materials of construction exhibit different EAF characteristics, even in the same component.

- The same plant transients applied to one component will produce different Uen values for different material of construction.

- DO content affects materials of construction differently:

o Higher DO gives lower Fen values for austenitic stainless steels [5]

o Constant effect for all DO content for austenitic stainless steels [7]

o Higher DO gives higher Fen values for carbon and low-alloy steels

[6,7]

o Constant effect for all DO content for Ni-Cr-Fe steels for PWRs and HWC BWR water and constant effect for all DO content for Ni-Cr-Fe steels for NWC BWR water [7]

Further factors that influence the evaluations are:

Use of the alternate rules of NUREG/CR-5704 [5] (stainless steel) and NUREG/CR-6583 [6] (carbon and low alloy steel) will produce somewhat different values of Fen than the newer rules of NUREG/CR-6909 [7] for those materials.

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Components in similar plants will likely have similar estimated EAF characteristics, although some may have computed CUF values and others may not. This conclusion is based on an EPRI review of piping fatigue [9]

where it was determined that:

- Although ANSI B31.1 and ASME Code,Section III, Class 1 piping rules are fundamentally different, experience in operating plants has shown that piping systems designed to B31.1 are adequate.

- The operation of B31.1 plants is also not different from that of plants designed to ASME Code,Section III.

Providing a robust solution without resorting to a complete reanalysis requires a new approach. EPRI has developed a process for screening the primary coolant-wetted Class 1 reactor coolant pressure boundary fatigue-sensitive components in a plant by ranking them in terms of Uen and then determining a set of Sentinel Locations such that every plant component is covered by one or more Sentinel Locations.

1. A Sentinel Location is a specific location in a piping system or component that serves as a leading indicator for EAF damage accumulation. It serves by itself or as one of a small group of locations that bound other locations in a Thermal Zone for a given material of construction. These Sentinel Locations are expected to remain bounding as plant transients occur in plant life. Thus, monitoring of the Sentinel Locations would maintain assurance that the system or component remains bounded throughout its operating life and can be used to trigger any necessary actions with sufficient time to provide appropriate remedies for the system or component. Sentinel Locations should be periodically reevaluated as plant transients accumulate to ensure that they continue to serve the sentinel function.

2. Thermal Zone is defined and discussed in Section 3.

EPRI Technical Report 1022873, Improved Basis and Requirements for Break Location Postulation, October, 2011 [10] concludes that fatigue usage calculated using ASME fatigue curves, including environmental effects, does not directly correlate with the probability of failure. In addition, for the range of components evaluated, a Uen of less than 1.0 has an insignificant safety impact, since the estimated core damage frequency was in all cases below 1E-6. Thus, the consideration of special EAF limit for the purposes of screening or ranking is not considered in this report.

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Section 3: Process Outline and Technical Basis This section provides an outline and technical basis of the screening process for reviewing plant components susceptible to fatigue, categorizing them into groups, and identifying one or more Sentinel Locations for each group that can be analyzed and monitored for EAF usage. In this context, a Sentinel Location is a location in a plant system that is expected to accumulate more EAF usage than other locations in that system.

The idea of Sentinel Locations extends the basic approach that was used in NUREG/CR-6260 [4]. It retains the core concept of analyzing a few challenging locations to represent the entire plant, but it adds a semi-quantitative ranking system to demonstrate that each plant component is represented by at least one Sentinel Location.

It is necessary to evaluate components and/or locations in a component on a uniform common basis to accomplish valid ranking and identification of Sentinel Locations in each Thermal Zone. Plants with fatigue design bases can have:

Sets of components evaluated to a reduced, bundled set of plant transients and/or a mixture of bundled and unbundled transients.

Components or locations in components evaluated to additional refined analyses while other components or locations are not.

To assure uniform determination of relative fatigue accumulation, these differences must be accounted for or eliminated. The screening processes described in this report are designed to make this common basis determination.

The reader is reminded that this report is NOT provided as a Quality Assured document. Application of the processes described will require appropriate review and quality dedication on a site-specific basis.

Outline of Process Steps This screening process consists of four stages: data collection, determination of Thermal Zones, evaluation of locations, and ranking and identification of Sentinel Locations. Each of these stages is explained in the sections that follow.

1. Data Collection 3-1

Data collection is necessary to equip the user to perform screening evaluations. Input data such as component geometry and material properties, plant transient characteristics and projections of plant transients for the licensed operating period are required to compute relative stress, CUF and Uen values for evaluated components.

2. Determination of Thermal Zones A Thermal Zone is defined as a collection of piping and/or vessel components which undergo essentially the same group of thermal and pressure transients during plant operations. Thermal Zones are determined on the basis of common plant transients during plant operation. Components are assigned to appropriate Thermal Zones and evaluated as a group. This allows definitive rankings to be determined.

3. Evaluation of Locations Locations in plant components are evaluated to establish relative stress, CUF and EAF values. In keeping with the principle that no new detailed stress and fatigue analyses are required to perform this screening, locations in each Thermal Zone will be evaluated with a common basis approach. This common basis approach mitigates the skewing effects of refined analyses (such as elastic-plastic analysis) for selected components. The purpose of ranking on a common basis is to assure that the most highly stressed and cycled locations in each Thermal Zone are identified as leading indicators of fatigue damage for the Thermal Zone.

Two analytical evaluation procedures are developed to aid in the evaluation process; one to perform Common Basis Stress Evaluations and the other to perform Fen Estimation Evaluations. A brief description of the analytical flow follows. Detailed descriptions of the procedures are provided in Section 4.

The evaluation incorporated in the Common Basis Stress Evaluation Procedure is based on the rules of ASME NB-3600 modified to address a screening evaluation for relative ranking of locations. Rationales for this approach are that:

- The majority of the components in the screening population are piping components for which the rules of NB-3600 are appropriate.

- The NB-3600 equations are explicitly defined and require minimal analyst interpretation so that they can be easily included in a spreadsheet.

- The NB-3600 rules are representative of the more general rules of ASME NB-3200 design by analysis, which are appropriate for all plant components.

For cases where a component or location has no explicit design fatigue analysis available or where is it desired to put components with an analysis on a common basis, the user may need to estimate a common basis CUF (i.e., an estimated CUF value that is determined on the same transient basis with all other locations in the system). The Common Basis Stress Evaluation is used to 3-2

perform the following stress computations to determine the common basis CUF:

- Through-wall transient thermal stresses are computed for leading transients. Transients with thermal shocks are found to be the leading fatigue usage contributor in component stress analyses.

- Piping moment range stresses and pressure stresses are extracted from the plant piping Class 1 stress report. Use of actual piping results avoids the use of piping umbrella loads and helps differentiate moment loadings for locations within a piping system.

- Peak Stresses at discontinuities are accounted for using SCF/FSRFs taken from the ASME Code.

The Fen Estimation Evaluation Procedure is developed to estimate Fen for locations in plant components on the basis of the relevant parameters -

Dissolved Oxygen (DO), maximum temperature and estimated tensile strain rate - of the leading transient. This procedure can be used as a source of estimated Fen values for plants both with and without fatigue design analyses.

The Procedures are developed to:

- Use for plants with and without explicit fatigue design analyses available.

- Use with design transients or actual transients (as long as they are consistently applied).

- Use with design numbers of transients or licensed operating period (e.g.,

60-year) projected numbers of transients.

4. Ranking and Identification of Sentinel Locations An estimated Uen (Uen*)is determined by multiplying the common basis CUF by the estimated Fen. Those locations within each group with the highest estimated Uen are reviewed to determine one or more Sentinel Locations.

These leading locations for environmental fatigue accumulation from ongoing plant transients should be managed by the plant FMP to assure adequate margin for fatigue considering EAF.

The end result of this screening process is a listing of fatigue-sensitive plant components, organized into groups, ranked by Uen* severity, with at least one Sentinel Location identified for each group of components. Thus, a Sentinel Location may represent a number of other locations. This information can be used to augment the existing plant Fatigue Management Program to assure that bounding EAF locations in each Thermal Zone will be monitored and serve as early warning beacons and action triggers for components which might approach Uen = 1.0.

Process Development Assumptions and Characteristics Several assumptions are inherent in the process developed in this report, as provided in the following list:

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Estimated Fen method is sufficient for a screening process; it is not intended to qualify components using this process.

Several characteristics of the process are important.

Common stress evaluation basis, consistent S-N curves should be used.

Linear elastic stress analysis and superposition of stress contributions are used.

The Fen factor is applied only for increasingly tensile portions of transients, based on the guidance of MRP-47, Revision 1 [11].

The Ke factor is included in both the determination of strain range and in the estimated strain rate determination. This approach is recommended in proposed ASME-Code Case N-792-1.

Thermal Zones are employed to provide consistency in development of estimated Fen values and common basis stress approximations.

Common analytical basis (un-bundled transients) is used to put analyses in a Thermal Zone on the same transient basis.

Calculated plant piping loads and stresses are used instead of piping attachment point umbrella loads.

Design severity transients (can use actual severity, if available and consistently applied) are used.

Geometric factors are applied to stress terms.

Materials of construction are evaluated together as a group in each Thermal Zone.

Determining Thermal Zones For the purpose of this process, a Thermal Zone is defined as a collection of piping and/or vessel components which undergoes essentially the same group of thermal and pressure transients during plant operations. The idea of Thermal Zones is similar to the way that design specifications separate components into groups with common transient definitions. Within a Thermal Zone, thermal shocks and thermal bending stresses vary depending only on the materials, geometry, and location of the component in the system. Therefore it is possible to rank the locations in a Thermal Zone for fatigue independent of the transients that actually occur.

An important step in this process is dividing plant systems into Thermal Zones.

For each system, one or more Thermal Zones must be determined on the basis of similar thermal and pressure transients. Operating procedures, design specifications and piping isometric drawings are used to determine which components undergo essentially the same set of thermal and pressure transients in terms of the transient variation in temperature and pressure. Components in the same flow path or in the same sector of a vessel would be included in the same Thermal Zone. When performing this step, it is important to make the 3-4

Thermal Zones as inclusive as possible, to capture the largest number of components in the ranking. Some components may be considered to be part of two adjacent Thermal Zones.

For instance, the Class 1 portions of the Charging and Volume Control System (CVCS) in a PWR is comprised of piping that connects the Regenerative Heat Exchanger (RHX) to the cold leg charging nozzle(s) and the pressurizer spray system and from the letdown nozzle on a cold leg to the RHX. Considering one charging flow path from the RHX to one of the cold legs, components in that charging flow path experience essentially the same transients during operation, with only minor variations depending upon location in the flow path. This characteristic establishes these components as a Thermal Zone (see Section 4.2 for more detail of the process).

Technical Basis for Fen Estimation Evaluation This section describes the technical basis and procedure developed to compute estimated Fen (Fen*) and estimated Uen (Uen*) values for locations in individual plant components where there is a DSR from which CUF values can be extracted.

For the purpose of this screening, the rules for calculating Fen values may either be taken from (1) NUREG/CR-5704 [5] for stainless steel material, NUREG/CR-6583 [6] for carbon/low alloy steel material and NUREG/CR-6909 [7] for Ni-Cr-Fe material, or (2) from NUREG/CR-6909 [7] for all materials. It is noted that the substitution of the rules of NUREG/CR-6909 for stainless steel and carbon/low alloy steels causes the values of Uen to differ from the values using NUREG/CR-5704 and NUREG/CR-6583.

These rules allow calculation of Fen factors based on the material at the postulated failure location (SS, CS, LAS and Ni-Cr-Fe) and the following environmental parameters:

Estimated strain rate ( ) during the transients, in [%/sec].

Concentration of dissolved oxygen (DO) in the water, in [ppm].

Maximum fluid/metal temperature (T) during the transients, in [°C].

(Note: sulfur content of the metal (S) is also a factor for CS and LAS. However, this procedure will conservatively assume all CS/LAS components have the worst possible sulfur content.)

Since the procedure is developed as an aid to a screening evaluation for the purpose of relative ranking, the exact values of these parameters will not be calculated from qualified design input. Instead, estimated values will be determined based on knowledge of the operation of the various plant systems and components during both normal operation and the transient conditions as defined in the plant design specifications. Specifically:

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Any components which have no exposure to the environment (i.e., heated primary/secondary coolant water) will be assigned an Fen value of 1.0. This includes components such as bolts and studs, and components where the critical location with respect to fatigue is on the outside surface (i.e., exposed to ambient air). These components will not rise to the Uen necessary to be considered as a limiting location, and are not further evaluated in this report.

Material identification for the component may be obtained from available drawings, such as flow diagrams, piping isometrics and material specifications.

A qualitative estimate of the strain rate for the controlling fatigue transient(s) whose resultant stresses become increasingly tensile during the course of the transient will be determined, based on knowledge of the corresponding plant system. Each component will be identified with one of eight possible categories shown in Table 3-1. Transient pairs composed of seismic loadings will be assigned Fen = 1.0.

The effect of Ke should be accounted for in the estimation of strain rate.

This method is justified by additional study documented in the Background for Revision to CC N-792 that determined that it is acceptable to include the stress due to Ke in the strain rate calculation.

Table 3-1 Strain Rate Categories Strain Rate Category Estimated [%/sec]

Extreme 5.0 V.High ~ 1.3 High ~ 0.33 Mid-High ~ 0.087 Medium ~ 0.023 Low-Mid ~ 0.0059 Slow ~ 0.0015 V.Slow 0.0004 Note: any components which have no exposure to the environment (i.e., heated primary coolant water) will be assigned an Fen value of 1.0. This includes exterior locations and vessel head and manway studs, for example.

An estimated DO value of Low ( 0.04 ppm) will be applied for all components exposed to reactor water for PWRs. This determination is based on the observation that for the entire history of most PWRs, the concentration of dissolved oxygen is maintained below 0.04 ppm at all times when water temperature is 150°C (302°F) (with rare exceptions). (Note:

when water temperature is below 150°C, DO is no longer a factor in the value of Fen for any of the materials considered in this procedure.) For BWRs, the DO values must be determined based on the procedural policies of the plant for water chemistry control. If plant reactor water or metal 3-6

surface DO data is known, the actual DO values may be used. For NUREG/CR-6909 applications, a consistent O is used for austenitic stainless steel materials, which is invariant to DO level.

An estimated upper-bound T value will be determined based on the collected design transients for the respective plant systems. (It will be converted to °C in the Fen procedure as necessary.) If plant reactor water temperature data is known, the actual temperature values may be used.

For each component, this evaluation computes two hypothetical Fen values, one using the estimated parameter values described above, and the second using the same estimated values for DO and T, but using the worst possible (i.e. most conservative) value for strain rate, . These two computed values are averaged to produce an expected Fen for each component. This two-part expected Fen is based on experience with performing detailed Fen analyses; in general, the estimated Fen from a detailed analysis is close to the Fen value computed for just the controlling transient pairs, but slightly higher due to contributions from the less-significant fatigue pairs. A simple average is judged to magnify the contributions of the less-significant transient pairs to yield a reasonably conservative value suitable for ranking without performing a detailed analysis.

This method is considered to be a reasonable approach, but other methods for assigning Fen values may be used, as determined by the user.

Technical Basis for Common Basis Stress Evaluation The Common Basis Stress Evaluation Procedure is used to compute an estimated CUF (U*), estimated Fen (Fen*) and estimated Uen (Uen*) values in components where:

There is no Design Stress Report (DSR) (or the DSR does not include an explicit CUF calculation).

There is a DSR with CUF values available, but there is reason to put the components on a common stress basis for comparison purposes.

Rationale for the Procedure Taking guidance from the EPRI Fatigue Management Handbook [12], formulas have been developed to compute stresses arising from maximum transient through-wall temperature distributions, axial temperature differences, thermal and mechanical bending stresses and geometric characteristics for piping and vessel components. These formulas ensure a common level of analysis so that the computed stresses are directly comparable between locations.

These formulas assume that stresses are linear elastic, and so may be combined using linear superposition. Non-linear plasticity effects are accounted for using elastic-plastic penalty factors (Ke) in accordance with ASME Code Subarticles NB-3200 and NB-3600 [1]. Use of linear elastic rules for computing CUF retains technical parity among the components in a Thermal Zone. By contrast, using elastic-plastic non-linear techniques in a fatigue analysis may significantly 3-7

reduce the computed CUF for that component, which would give it a much lower CUF than other locations with comparable fatigue duty.

The linear elastic stress state for a location may be computed as the linear summation of the individual stresses caused by various types of loads. Most pressure vessels and piping system components include stresses due to internal pressure, thermal (due to temperature distribution in the component), and boundary interface loads, such as forces and moments caused by thermal expansion, thermal stratification, anchor displacement, seismic movement, etc.

Deadweight and residual stresses may be ignored, because they do not vary with time and therefore do not impact the computed stress range.

For a linear elastic stress analysis, stress contributions may be classified as one of two types:

1. Stresses due to loads, such as pressure, piping thermal expansion, etc. that are directly scalable to pertinent parameters (pressure, temperature, etc.), and

2. Time-dependent thermal stresses, which depend on the axial and radial temperature distributions in the component rather than any single instantaneous parameter.

Stress contributions of the second type depend on the temperature history and are typically calculated by a time integration of the product of a predetermined Greens function, or influence function, and the transient temperature data.

Performing this integration is more complex than is desired for this screening process. Instead, an estimate of the maximum stress range during each significant thermal transient is computed, as described below. This estimate applies a uniform level of conservatism, and is sufficiently precise to determine a relative ranking among the components in a Thermal Zone.

The stress computation combines stresses from the following terms:

Through-wall transient thermal stresses are computed using the graph shown in Figure 3-1. For each transient, two non-dimensional factors (k/hL) and (kt0/cpL2) are computed as entry into the curve for the determination of the normalized thermal peak stress.

Piping moment range and pressure stresses are extracted from the plant piping Class 1 stress report. Umbrella loads (conservative loads assigned to the system to facilitate design of adjoining systems) are not recommended, as they dont inform the relative severity at different locations.

Thermal stratification moment stresses are assumed to be negligible or included in the computed piping moment stress range.

Seismic stresses.

Peak Stresses at discontinuities are accounted for using appropriate SCFs.

Actual values of these stresses may be used, if applied consistently within a Thermal Zone evaluation.

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The Common Basis Stress Evaluation Procedure is used to determine approximate stress ranges arising from pairs of selected significant transients, compute alternating stress values including simplified elastic-plastic (Ke) effects, and produce estimated incremental CUF (Uincr*) for input numbers of plant transients (either design numbers or projected numbers). These estimated incremental CUF (Uincr*) values if summed would produce the common basis CUF (U*). Estimated Fen values (Fen*) are computed (using either the older or newer EAF rules), and multiplied by Uincr* to produce an estimated incremental Uen (Uen incr*) for each transient pair. These Uen incr* values are summed over the significant transients to yield an estimated Uen* for that location.

Formulas and Equations The evaluation addresses the following equation.

Speak = Spress + Smom + Strans + Sts, Equation 3-1 (valid for cylindrical or flat plate components)

Where:

Speak = total peak stress range (including Ke factor as appropriate)

Spress = peak stress range from pressure Smom = peak stress range from moments (includes seismic loads and stratification loads)

Strans = peak stress range (axial) from gross structural or material discontinuity Sts = peak stress range from through-wall thermal gradient In this process, stresses, moments and thermal transients are input and several thermal-hydraulic characteristics are computed. Stresses are addressed as follows:

For Spress and Smom: extract values for transients from the Class 1 stress reports For Strans: ignore for screening purposes or extract from the Class 1 stress reports For Sts: evaluate by determining the maximum through-wall thermal stress using the graphical relationship in Figure 3-1.

Inputs for this determination for a piping component are:

Geometric: Piping Diameter (D), Wall thickness (t), Moment SCF (K2) and Peak SCF (K3).

Materials: Sm ,Thermal Conductivity (k), Elastic Modulus (E), Coefficient of Thermal Expansion (), Density (), Heat Capacity (Cp).

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Thermal-hydraulic: Flow Rate (Q), Maximum Temperature Tmax, Transient Temperature Change (T), Transient Time (t0), Transient Pressure Change (P).

Computed values are Heat Transfer Coefficient (h), k/ht, kt0/cpt2, normalized thermal stress (max / [(ET)/(1-)]) and Transformed Transient Strain Rate ( * ). The equations in Figure 3-1 use the terminology of L for wall thickness, instead of t.

The heat transfer coefficient, h, must be computed for each of the plant transients (up and down portions), as follows:

h = 0.023 * (VD/) 0.3 Pr 0.4 (k/D) (valid for turbulent flow conditions - Re

> 2000)

Rearranging to define in terms of Q and D and accounting for units:

h = 12.92874 * (Pr 0.4 k/ 0.8 ) * (Q 0.8 / D 1.8) h = * (Q 0.8 / D 1.8) has been curve fit for use in spreadsheets as (valid for temperatures from 0°C (32°F) to 315.6°C (600°F)):

= 56.45 + 1.270

T - 2.927 x10-3

T2 + 3.952 x 10-6

T3 - 2.654 x 10-9

T4 where:

h = heat transfer coefficient (Btu/hr-ft2-°F)

= density (lbf/ft3)

V = velocity (ft/sec)

D = inside diameter (inches)

= kinematic viscosity (ft2/hr)

Pr = Prandtl Number (dimensionless) cp = specific heat (lbf/ft3-°F) k = thermal conductivity (Btu/hr-ft-°F)

Sts is computed by computing two non-dimensional factors (k/ht and kt0/cpt2) and using those factors to determine the value of maximum thermal stress as a percentage of the maximum thermal stress from a steep temperature step with infinite heat transfer from the graph on Figure 3-1.

1. Spress is computed or extracted from the plants Class 1 stress reports.

2. Smom is developed by extracting thermal and mechanical moments for transients from the plants Class 1 stress reports.

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1.00 0.90 0.80 0.70 max/[(ET0)/(1-)]

0.1 0.60 0.50 0.40 0.30 0.20 2.0 0.10 0.00 0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90 1.00 k/hL Figure 3-1 Determination of Transient Stresses for Ramp Transients Computations and Resulting Outputs from Common Basis Stress Evaluation Procedure For each selected component location, the procedure is used to:

Define the leading transients for evaluation.

Compute stresses for each of the leading transients, accounting for FSRFs.

Pair tensile and compressive stress pairs of leading transients.

Compute transient pair Sa from Sp-p, Sm and Ke.

Compute transient pair Uincr* from Nallow and nevents. Note: nevents is based on the design basis number of events; it may also be of interest to evaluate based on the projected number of events (and such a comparison is shown in Table 5-3), however, the recommended approach for ranking is to use the design basis number of events.

Compute transient pair Uen incr* from Uincr*and Fen*

Sum all transient pair Uen incr* into component Uen*.

Comparison of Common Basis Stress Evaluation Screening Rules to NB-3600 Evaluation Process An evaluation of the analysis procedure presented in NB-3600 for the evaluation of piping components is made to compare the elements of the screening process to the ASME rules.

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A comparison of the Common Basis Stress Evaluation process is made to the basis equations of NB-3600 (provided in Appendix B as quoted in Section 2 of

[4]). For the express purpose of screening, certain judgments are made about the equations in NB-3653.

For the purpose of this fatigue ranking, it is assumed that the basic primary stress limit is assumed to have been addressed in the component design and will not be evaluated for screening.

NB-3653.1 Equation 10 [1] is accommodated in the screening by manual input of a PPSMAX value for each component, based on previous analysis and judgment of the maximum ratio of Primary + Secondary / Primary + Secondary +

Peak Stresses (defined as PPSMAX). The peak stress range will be multiplied by the PPSMAX value to supply the secondary stress range used to compute the elastic-plastic penalty factor, Ke in NB-3653.1 Equation 14 [1].

Judgment is used to determine the comparison to the terms in the NB-3653.1 fatigue Equation 11. The prevailing fatigue-driving loading in nuclear plant Class 1 piping is thermal shock, where cold water enters a hot pipe in a step change or short time duration. Pressure and moment loadings play an important secondary role and must also be considered. Common Basis Stress Evaluation Procedure addresses the following terms of Equation 11 [Eq. 2-7 of Ref. 4]:

PD D 1 1 S =KC o o +K C o M + K C E x a T a T + ( ) K E a T + E a T p 1 1 2t 2 2 2I i 3 3 ab a a b b 21 v 3 1 1 v 2 Equation 3-2 The first three terms of Equation 11 are represented by the first three terms of the equation below. These terms generally do not provide the primary loadings for components undergoing thermal shock and experiencing relatively high CUF accumulations. The fourth term of Equation 11 is represented by the last term of the equation below. The fifth term of Equation 11 is generally a small contributor to CUF compared to the fourth term and requires more detailed analysis to define. For the purposes of a screening evaluation where thermal shock plays a large role, it can be ignored.

Speak = Spress + Smom + Strans + Sts Equation 3-3 Stresses are addressed in Common Basis Stress Evaluation Procedure as follows:

For Spress and Smom: extract values for transients from the Class 1 stress reports For Strans: ignore for screening purposes For Sts: evaluate 3-12

NB-3653.1 Equations 12 and 13 [1] are not evaluated in the screening process because it is assumed that they met their limits in the piping component design and will not change in the course of screening.

NB-3653.1 Equation 14 [1] is used to compute CUF in the screening process, A thermal ratcheting evaluation is not performed in the screening process because it is assumed that limit was satisfied in the piping component design and will not change in the course of screening.

Limitations and Assumptions of the Process Stresses caused by complex loading, such as thermal stratification, are not used in the Common Basis Stress Evaluation process. It is typically not practical to compute stratification stresses using a basic methodology. However, for components subjected to this type of loading, fatigue calculations are expected to have been performed already. Such is the case, for example, with PWR surge lines.

Likewise, axial thermal gradient stresses produced by geometry or material transitions are also not considered in this process. Branch nozzles without thermal sleeves are commonly subject to stresses caused by axial thermal gradients. Such loading may be attributed to the injection of colder fluid into a hot header, giving rise to significant thermal stresses of a steady state nature near the nozzle corner. Sophisticated fatigue analyses are typically employed to disposition these types of components, and many of them, such as the charging and safety injection nozzles, are the NUREG/CR-6260 locations (the GALL report requires evaluation of the 6260 locations at a minimum).

The Common Basis Stress Evaluation Procedure is strictly valid only for cylindrical or flat plate components where the simplified methodology is most appropriately applied.

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Section 4: The Screening Process This section describes a screening process with step-by-step instructions for evaluating the Class 1 systems in a plant to produce a list of Sentinel Locations to include in the Fatigue Management Program. This process is depicted in Figure 4-1 and Figure 4-2. Thermal Zones are established in each Class 1 system.

Candidate Sentinel Locations are determined in each Thermal Zone. Each of the materials of construction of the candidate Sentinel Locations are evaluated as a group using one of two procedures to produce an estimated Uen ranking for comparison with estimated Uen values determined for any NUREG/CR-6260 locations in the Thermal Zone. The NUREG/CR-6260 locations will be retained as Sentinel Locations in this process. Those candidate Sentinel Locations of all materials in all Thermal Zones in each system which meet certain grouping criteria are selected as Sentinel Locations.

The use of readily available data is encouraged for a screening evaluation, however a set of consistent design input is necessary. For example, within a given Thermal Zone screening evaluation, one of the methods should be used exclusively (design results with design numbers of cycles, design results with projected numbers of cycles, monitoring system results with design numbers of cycles, or monitoring system results with projected numbers of cycles).

Detailed Screening Procedure Refer to Figure 4-1 for this procedure.

Gather Required Inputs for all Systems Containing Class 1 Reactor Coolant Pressure Boundary Components Determine the following data for the components in each Class 1 system:

Materials (austenitic stainless steel (SS), carbon steel (CS), low-alloy steel (LAS) or Inconel (Ni-Cr-Fe))

Layout (connectivity and flow paths)

Geometry (ID, OD, material/geometric discontinuities)

Fatigue Strength Reduction Factors: K2 and K3 CUFs (for those components with fatigue analyses)

DO history (for the contained fluid) 4-1

List of thermal and pressure transients For the leading transients (available from stress report or transient monitoring program)

- Tmax and Tavg

- Material properties (Sm, k, E, , , cp) at average temperature

- Thermal-hydraulic characteristics (Q (or V), Tmax, Tmin, t0, P)

- Estimated strain rate

- Moment ranges for thermal and seismic loadings (Mthermal and Mseismic)

- Pressure stress range

- (optional) projection of number of cycles (nprojected)

This input data can be obtained from:

- Vessel design drawings or P&ID and piping isometric drawings:

o Vessel and pipe ID, OD and geometric factors o Connectivity and flow paths

- Flow diagrams:

o Connectivity and flow paths

- Plant design specifications or design stress report (DSR):

o Material properties at average temperature o Properties of each leading transient (see above)

- Either DSR or FSAR or transient monitoring system:

o Moments, stresses and loadings (see above) o Design CUF (U)

- Plant operating logs and procedures:

o DO history Determine Thermal Zones for Each System For each system, one or more Thermal Zones must be determined on the basis of similar thermal and pressure transients. Operating procedures, design specifications and piping isometric drawings are used to determine which components undergo essentially the same set of thermal and pressure transients in terms of the transient variation in temperature and pressure. Components in the same flow path or in the same sector of a vessel would be included in the same Thermal Zone.

Starting with a location at the boundary of the candidate component, identify the thermal and pressure transients at that location, then expand the selection to any neighboring components which are subject to the same set of transients. At the point where the transients change, such as at a branch, pump, or temperature source, another boundary is established. When there are no additional locations left to evaluate, the Thermal Zone consists of all of the components within the boundaries.

4-2

For example, consider the Charging and Volume Control System (CVCS) in a PWR. It is comprised of piping that connects the Regenerative Heat Exchanger (RHX) to the cold leg charging nozzle(s) and the auxiliary spray tee on the pressurizer spray system and from the cold leg letdown nozzle to the RHX.

There may be one or two charging nozzles in these systems (this example assumes two charging nozzles where flow is alternated between them every operating cycle).

The major transients in the CVCS are Loss of Letdown and Loss of Charging with prompt and delayed returns and charging flow adjustments. The transients are significantly different between the charging and auxiliary spray flow paths through the RHX tube side and the letdown flow path through the RHX shell side. Thus, the first division of potential Thermal Zones is between the charging flow path and the letdown flow path. Downstream components in the RHX tube side flow path will experience very similar thermal and pressure transients, only differing by which flow paths are active (which of the charging nozzles is in operation and whether or not auxiliary spray operations are active). The flow path for the active charging nozzle will experience essentially the same severity and duration of the transients from that charging nozzle to the RHX while the other charging nozzle flow path is stagnant and experiencing minimal transient behavior. The components in the active charging path comprise a single Thermal Zone. Since the currently inactive charging path will alternately be the active flow path, this flow path will also experience essentially the same transients from the charging nozzle to the RHX and would also qualify as a second Thermal Zone.

The flow path through the auxiliary spray piping will also experience unique transient depending upon the activity of that line and qualifies as a third Thermal Zone. The letdown piping is a single run of piping conducting the same letdown flow from the cold leg nozzle to the RHX and qualifies as the fourth Thermal Zone.

A determination must be made about the components on the boundaries of Thermal Zones. Locations on a boundary will typically experience transients from both Thermal Zones. A boundary component should generally be assigned to one of the two Thermal Zones. Focusing on the purpose of the screening process - to identify the leading fatigue accumulation locations - the important factor is whether the boundary location experiences all of the transients of the Thermal Zone. If it does, then it should be included in the Thermal Zone. In some cases, it may belong to both Thermal Zones, or it may become a Thermal Zone of its own.

For instance, the cold leg charging nozzles are at the boundary of the CVCS and the RCS cold leg piping. These nozzles could belong either to a unique Thermal Zone, the charging piping Thermal Zone or the cold leg Thermal Zone, or in several of them. The decision must be made about where they best fit. The charging nozzles experience all of the transients common to the rest of the charging line piping, but also exhibit additional transients caused by cold leg fluid reflood when charging flow is terminated. Due to the presence of additional 4-3

transients at the charging nozzle compared to the remainder of the charging piping, the charging nozzles should be included in the Thermal Zone with the associated charging line piping, instead of in a unique Thermal Zone. The charging nozzles do experience all of the global transients in the RCS cold leg piping, but review of the design fatigue analysis shows the primary fatigue duty for them are CVCS transients rather than RCS transients. This fact leads to the decision not to include the charging nozzles in the RCS cold leg Thermal Zone.

This is expected to be the usual assignment of boundary components on the RCS. The RCS piping nozzles connecting to adjoining systems, such as the CVCS, Safety Injection, Residual Heat Removal or Shutdown Cooling, etc., will be expected to be included in the Thermal Zones of those adjoining systems.

Continuing the evaluation would lead to the identification of the following four Thermal Zones for the CVCS:

1. One charging nozzle connected to a cold leg and piping back to the RHX

2. The other charging nozzle connected to a cold leg and piping back to the RHX

3. Auxiliary spray piping from the main spray line back to the charging line

4. Letdown nozzle on cross-over leg and piping between RHX and cross-over leg Identify Materials and Candidate Locations, For Each Thermal Zone in the system, Select candidate locations by material Identify thermal, pressure and seismic transients Determine candidate locations by examining vessel nozzles and wall thickness transitions and piping nozzles, elbows, tees and wall transition changes. Look for the largest changes in wall transitions, nozzles connected to piping or components with largest temperature differences, etc. The NUREG/CR-6260 locations will be included as candidate Sentinel Locations, evaluated on a common stress basis, and retained as Sentinel Locations in this process.

Identify the major thermal and pressure transients. Among these will be transients with a large change of value over a short time duration (e.g., step temperature shocks or rapid pressure drops).

Calculate Uen* for Each Candidate Location For each candidate location in the Thermal Zone Determine if a fatigue table exists from design stress reports

- If YES o Do CUFs fit Common Basis model?

4-4

o If YES, Go to Uen Estimation Evaluation Procedure (A) o If NO, Go to Common Basis Stress Evaluation Procedure (B)

- If NO o Go to Common Basis Stress Evaluation Procedure (B)

The common basis concept is needed so that candidate Sentinel Locations are not disproportionately promoted or demoted due to assumptions of the analysis.

This happens because many calculations use transient lumping and other strategies to reduce computational complexity. It is important to apply the same level of rigor in transient definition across all locations to assure proper relative ranking of fatigue results.

For CUF locations, a common basis determination is made by examining the fatigue table for each component. Examine the load set pairs with the largest Uincr that account for at least 75% of the total CUF. If those load set pairs do not include any lumped transients, then the CUF satisfies the common basis and Part A below can be used. Otherwise, a common basis CUF needs to be calculated as part of the evaluation (Part B should be used).

For locations without a CUF, follow the procedure in Part B.

Uen Estimation Evaluation Procedure The Fen Estimation Evaluation Procedure (included below) is used to calculate an estimated Fen (Fen*) and a final step is added to calculate an estimated Uen (Uen*)

for each evaluated location.

The Uen Estimation Evaluation Procedure is comprised of the following activities:

1. For each location, the user:

- Inputs the material type, DO, estimated strain rate and Tmax for the leading transient (from stress report or transient monitoring system)

- Enters design (or computed) CUF value for the location

2. For each location, the procedure is used to compute:

- Estimated Fen value (Fen*)

- Estimated Uen value (Uen*)

Fen Formulations The Fen Estimation Evaluation Procedure allows the user to select from the Fen formulations in NUREG/CR-5704 [5] and NUREG/CR-6583 [6], or NUREG/CR-6909 [7]. The details of these formulations are given in Appendix A. The user must select whether to use the older guidance [5 and 6] or the newer guidance [7]. When the old guidances are used, Fen formulations for austenitic stainless steel materials are conservatively used for nickel alloy materials.

4-5

Determine Input Values For each plant location, determine the following five variables:

CUF: The common basis CUF Material Type: Either SS, LAS, CS, or Ni-Cr-Fe DO: Either Low, Med, or High Tmax: The maximum fluid temp to which the component is exposed, in

[°F]it will be converted to °C in the Fen calculation. The value of the average temperature of each load pair (Tave) is used instead of Tmax if the rules in [7] are used.

Strain Rate: A qualitative estimate of the strain rate ( ) for the typical or controlling fatigue transient(s), based on knowledge of the corresponding plant system. The strain rate must be specified as one of eight possible categories listed in Table 3-1.

Fen Estimation Evaluation Procedure Note: When this procedure is applied to Part A, the Strain Rate defined above is used and only one Fen* is calculated for the location. When this procedure is applied to Part B, it is performed separately for each transient pair and each transient pair has a different Fen*. This is illustrated in Table 5-3.

1. Compute an Expected Fen for each location as Fen = exp ( A ( B x S * )

T *O* * ) Equation 4-1 where the constants (A, B) and the transformed parameters (T*, O*, etc.) are defined according to the specific NUREG guidance document; see Appendix A for details.

2. Compute Maximum Fen for each location. Calculate as above, but use the most conservative (lower bound) strain rate for the Material Type instead of the Strain Rate value determined above.

3. Compute Estimated Fen as: Fen* = (Expected Fen + Maximum Fen)/2.

Final Step for Uen Estimation Evaluation Procedure Compute Estimated Uen* for each location as: Uen* = (Fen*) x (CUF).

Common Basis Stress Evaluation Procedure Refer to Figure 4-2 for this procedure.

The Common Basis Stress Evaluation Procedure is used to calculate an estimated CUF (U*), an estimated Fen (Fen*) and an estimated Uen (Uen*) for each evaluated location.

4-6

The Common Basis Stress Evaluation Procedure consists of the following activities:

1. For each location, the user:

- Inputs the material type and geometric properties, thermal-hydraulic characteristics, DO, estimated strain rate and Tmax for the leading transient (judgment and/or evaluation of stress report)

2. For each location, the procedure is used to compute:

- Estimated U value (U*)

- Estimated Fen value (Fen*)

- Estimated Uen value (Uen*)

Determine Input Values

1. For each plant location, determine the following variables:

- Material Type: Either SS, LAS, CS, or Ni-Cr-Fe

- DOmax,

- Mi, Z, (Ta-Tb), Mstrat: the moment ranges, moment of inertia, axial temperature difference, and moment stress from stratification (for piping locations only), taken from the DSR.

- Geometric values [material type, ID, OD, C1, C2, K2, K3]

- Material properties at average transient temperature [k, E, , cp, , , Sm]

- Major transients in terms of fluid velocity (V), Tinit, Tfinal, to, Pmin, Pmax, nprojected

2. Calculate several dependent parameters:

h = convective heat transfer coefficient (use the formulas in Section 3) t = wall thickness = (OD - ID)/2 To = temperature change = (Tinit - Tfinal)

= estimated strain rate = To / to (determined as shown in Figure 3-1)

R = (OD + ID) / 2 Perform Stress Evaluation

1. Create load set pairs from the leading transients (or up-down pairs from single transients) using:

- Extract Spress as the maximum pressure stress range for the transient pair (i.e., C1(Po Do/2t)) from the components DSR (Po defined as the range of pressure for the transient pair and Do is OD)

- Determine Smom as the maximum thermal and mechanical moment stress for the transient pair - ( i.e., C2(Do/2I)Mi) from the components DSR (Mi is the range of moments for the transient pair and Do is OD).

- Determine Sts for each paired transient. Compute the two non-dimensional factors (k/ht and kt0/cpt2), then using Figure 3-1 to determine a value for max/(E T0/(1-)). Note: the equations in Figure 3-1 use the letter L for wall thickness, instead of t. Multiply that value by E T0/(1-) to get max = Sts for the transient. Note that Sts for step-up transients will be negative (less than zero), while Sts for step-down transients will be positive.

4-7

2. Compute Sp-p for each transient pair as: Sp-p = K1Spress + K2Smom + K3(Sts,max -

Sts,min)

3. Determine a PPSMAX (based on experience with fatigue analyses of the component)

4. Compute Ke for each transient pair using the rules in NB-3653.6, with Sn =

PPSMAX x Sp-p as: Ke = 1.0 + [(1-n)/n(m-1)] (Sn / 3Sm-1) within the bounds of 3Sm < Sn < 3mSm

5. Compute the alternating stress amplitude for each pair as: Sa = Sp-p/2 x Ke.

6. Compute Nallow for each pair using the appropriate fatigue curve from either NUREG/CR-6909 or the ASME Code.

Perform Fatigue Evaluation

1. Compute Uincr* (nevents / Nallow) for each transient pair. The total common-basis CUF (U*) is equal to the sum of these values for all transient pairs.

Note: nevents is based on the design basis number of events, it may also be of interest to evaluate based on the projected number of events (and such a comparison is shown in Table 5-3), however, the recommended approach for ranking is to use the design basis number of events

2. Determine EAF inputs for appropriate material and estimate Fen for each transient pair. Use Fen Estimation Evaluation Procedure described in Part A using the input gathered above to compute Fen* for each transient pair.

3. Compute Uen incr* for each transient pair as: U en incr* = Fen* x U incr*

4. Sum all Uen incr* to determine Uen* for each component Repeat the process for the next candidate location of that material in the Thermal Zone in the system. When Uen* has been determined for all candidate locations of that material in the Thermal Zone in the system, go to the next step of ranking the candidate locations.

RANKING and SENTINEL LOCATION IDENTIFICATION for Each Material in Each Thermal Zone

1. Identify all NUREG/CR-6260 locations as official Sentinel Locations for that material in that Thermal Zone and denote their Uen* values as Uen 6260*.

Sort all remaining locations by Uen* from highest to lowest value.

2. Identify candidate Sentinel Locations for final consideration.

- Is Uen max* 1.0 o If NO, evaluate candidate Sentinel Location Uen max* in step 3, remove all other locations from further consideration.

o If YES, Is Uen max* > 2 x Uen max-1*?

If YES, evaluate candidate Sentinel Location Uen max* in step 3, remove all other locations from further consideration.

4-8

If NO, are the top 3 (Uen max* , Uen max-1* and Uen max-2* ) within 25%

(i.e., (Uen max* - Uen max-2*) / Uen max* < 0.25)?

If NO, evaluate 2 candidate Sentinel Locations (Uen max* and Uen max-1*) in step 3, remove all other locations from further consideration.

If YES, evaluate 3 candidate Sentinel Locations (Uen max*, Uen max-1* and Uen max-2*) in step 3, remove all other locations from further consideration.

3. Consolidate candidate Sentinel Locations with any resident NUREG/CR-6260 Uen* (Uen 6260*) location(s)

- Does the Thermal Zone include one or more NUREG/CR-6260 locations?

o If YES, Is candidate Sentinel Location Uen* 0.5 x lowest Uen 6260*

location?

If YES, candidate Sentinel Location is promoted to an official Sentinel Location If NO, remove from candidate Sentinel Location list o If NO, candidate Sentinel Location is promoted to an official Sentinel Location Evaluate Next Candidate Location Repeat the process for the next candidate Sentinel Location of that material in the Thermal Zone in the system. When the ranking of all candidate Sentinel Locations of that material in the Thermal Zone have been determined, move to the next material in Thermal Zone.

Evaluate Next Thermal Zone When all candidate Sentinel Locations for all materials in the Thermal Zone in the system are evaluated and ranked, move to the next Thermal Zone and repeat the process.

Evaluate Next System When all candidate Sentinel Locations for all materials in all Thermal Zones in a system are evaluated and ranked, move to the next system and repeat the process.

Compile Final List of Sentinel Locations When all candidate Sentinel Locations for all materials in all Thermal Zones in all systems are evaluated and ranked, compile the list of Sentinel Locations for inclusion in the Fatigue Management Program.

4-9

Guidelines for Reducing Number of Sentinel Locations This screening and ranking process can produce a fairly large number of Sentinel Locations, with at least one Sentinel Location assigned for each material in each Thermal Zone in each system. With specific engineering justification, a specific Sentinel Location may be justified to bound one or more other Sentinel Locations and allow the bounded Sentinel Location(s) to be removed from the final list. Possible criteria that could be used to make these judgments and guidelines for their evaluation are included below:

Possible Criteria for Determination of Sentinel Location Boundedness:

One Thermal Zone can bound another Thermal Zone in a System This circumstance could be achieved if within the same system, both the CUF and Fen values for one Sentinel Location in one Thermal Zone are each higher than the CUF and Fen values for the Sentinel Locations in other Thermal Zones. It is expected that the Uen for the former Sentinel Location would be more than double the Uen values of the other Sentinel Locations.

The determination that this highest Sentinel Location of one material could bound the other locations could be justified on this basis.

One material in a Thermal Zone can bound other materials in the same Thermal Zone This circumstance could be achieved if within the same Thermal Zone, both the CUF and Fen values for one Sentinel Location composed of one material are each higher than the CUF and Fen values for the Sentinel Locations composed for all other materials. It is expected that the Uen for the former Sentinel Location would be more than double the Uen values of the other Sentinel Locations. The determination that this highest Sentinel Location of one material could bound the other locations could be justified on this basis.

One material in a Thermal Zone can bound other materials in another Thermal Zone This circumstance combines the guidelines of the two listed above and must satisfy both criteria listed.

A non-NUREG/CR-6260 location can bound and replace a NUREG/CR-6260 location This is not allowed within the guidelines for NRC review of GALL, Revision 2 [8], which states in paragraph 4.3.3.1.3: If an applicant has chosen to assess the impact of the reactor coolant environment on a sample of critical components, the reviewer verifies the following: 1. The critical components include a sample of high-fatigue usage locations. This sample is to include the locations identified in NUREG/CR-6260, as a minimum (emphasis added), and proposed additional locations based on plant specific considerations 4-10

A location with Uen* < 0.8 may be removed from the Sentinel Location list This judgment relies upon the relative value of the Uen* value of the Sentinel Location. Generally, if the Sentinel Location Uen* for the projected number of design cycles is low (e.g., Uen* < 0.8), that Sentinel Location may be removed from the final list due to the small likelihood that it will be the leading Sentinel Location in a system. If, however, the Sentinel Location Uen* for the projected number of design cycles is fairly high (e.g., Uen* 0.8), the possibility exists that it could remain the Sentinel Location for its group and should be included in the monitoring program that ensures that it does not exceed a value of 1.0.

Evaluation of the guidelines listed above leads to the possibility that a Sentinel Location is not required for every Thermal Zone and that a Sentinel Location is not required for every system.

4-11

Gather Required Inputs for all Class 1 Systems For each System - Determine Thermal Zones For each Thermal Zone in the System - Identify Material Used, Identify Candidate Locations, Identify Transients For each Material in the Thermal Zone in the System For each Candidate Location in the Material in the Thermal Zone in the System Fatigue on Fen Evaluation Fatigue Table? Yes Common Yes using CUF Basis?

No No To/from Define Transients Perform Estimated Figure 4-2 To/from Stress / CUF Evaluation To/from Calculate Uen*

Evaluate Next Candidate Location in Next Yes the Material Candidate?

No Identify NUREG/CR-6260 Locations, compute Uen 6260* values, move them to List of Sentinel Locations Sort all Remaining Locations by Uen*

Uen max* > 2 x Top 3 Uen*

Uen max* 1? Yes No Uen max-1*? within 25%?

Yes No Top 3 =

Yes No Yes 3 Candidate Locations, 1 Candidate Location, Remove Top 2 = 2 Candidate Locations, Remove Yes Remainder Remove Remainder Remainder For all Candidate Locations Yes 6260 location in Uen candidate* 0.5 Remove from Yes No Thermal Zone? x lowest Uen 6260*? candidate list Yes No Add to Sentinel Location List for that Material Next Material?

No Next Thermal Zone?

No Next System?

No Compile Complete List of Sentinel Locations Figure 4-1 Screening Flow Chart 4-13

Figure 4-2 Estimated Stress / CUF Evaluation 4-14

Section 5: Pilot Plant Application Table 5-1 provides the list of candidate Sentinel Locations and their evaluation as initial and final Sentinel Locations using the Uen Estimation Evaluation Procedure process of screening and the guidelines for reducing the number of Sentinel Locations. The Uen Estimation Evaluation procedure employed calculated CUF values; the pilot plant was designed to ASME Section III requirements and plant components had these data available. Components to be evaluated are selected to include locations constructed of each material residing in each Thermal Zone.

For each row in the table, the Component is evaluated in this process to determine whether it will be a Sentinel Location for its Thermal Zone and material. For instance, the top line shows that RPV Outlet Nozzle is the location evaluated to be the Sentinel Location for the LAS RPV Nozzle Thermal Zone. Some of the entries represent multiple Sentinel Locations in a Thermal Zone, arising from the rules permitting establishment of two and three Sentinel Locations depending upon the relative Uen* values of the top three locations.

Details of this evaluation presented in Table 5-1 follow:

In the Input section: The Design Basis CUF, material of construction, expected DO, maximum temperature for the leading transients and strain-rate are tabulated in individual columns. Based on knowledge of fatigue analyses of many plant components, a strain-rate category was selected to represent each component. This selection was based on identifying the transients that would govern the CUF at the given component, and ranking them with respect to how quickly the maximum and minimum stress states are established. For instance, if the majority of CUF for this component would generally derived from very large temperature step changes, the Extreme strain rate category was used.

Conversely, components governed by hours-long ramps of temperature and pressure would be assigned to the V.Slow category.

In the Fen Computation section: The parameters T*, O*, S*, * , and the three values of Fen, are calculated as follows:

The Expected Fen (Fen Exp.) is calculated from the values of the T*, O*, S* and

parameters using the the rules of NUREG/CR-5704 [5] for stainless steel material, NUREG/CR-6583 [6] carbon and low-alloy steel materials and NUREG/CR-6909 [7] for Ni-Cr-Fe steels material.

5-1

The Maximum Fen (Fen Max.) is calculated using the same formulas as above, except for using a bounding value for *, i.e., * = ln( 0.001 ) for CS, LAS and SS, or *= ln( 0.0004/5.0 ) for Ni-Cr-Fe.

The Fen Average (Fen*) is computed as [(Expected Fen) + (Maximum Fen)] / 2.

In the Comparison section: The Estimated Uen design (Uen*) is computed as the product of the Fen Average (Fen*) and the Design Basis CUF.

In the Identification of Sentinel Locations section:

Evaluations were made to determine whether each Evaluated Component qualified as an Initial Sentinel Location. This determination was made using the rules of the detailed screening process in Section 4. Descriptive reasons for the determination are provided in the column entitled Criterion for Initial Sentinel Locations. These descriptive reasons are related to the Ranking and Sentinel Location Identification logic presented in paragraph 5 of the Detailed Screening Procedure (page 4-8).

The Initial Sentinel Location Count column indexes those Evaluated Components that are established as Initial Sentinel Locations.

The last three columns present the results of using the Guidelines for Reducing the Number of Sentinel Locations (pages 4-9 and 4-10). The Criterion for Final Sentinel Location column provides a descriptive reason about why each Initial Sentinel Location is or is not a Final Sentinel Location. The Final Sentinel Location Count column indexes those Evaluated Components that are established as Final Sentinel Locations.

Table 5-2 provides a summary list of the results of Table 5-1. The last column shows the components that are identified as the final list of Sentinel Locations.

These components represent the Sentinel Locations assigned for each material in each Thermal Zone of each system or vessel evaluated.

The sample results of an evaluation of the pilot PWR plant using the Common Basis Stress Evaluation Procedure are provided in Tables 5-3 through 5-5.

Geometric and material properties for evaluated components are listed in Table 5-3. The input and computed thermal-hydraulic and stress characteristics of two transients are listed in Table 5-4. Table 5-5 shows the resulting combined stresses, Fen*computation and Uen* values. Results of this procedure demonstrate concurrence with the conclusion of the Uen Estimation Evaluation Procedure using design CUF input. Both methods indicate that the charging nozzle will bound the charging piping, although the Uen* values are not exactly the same.

This table also includes a comparison of Uen* calculated with the design number of transients and Uen* calculated with the projected number of transients based on actual operating experience.

5-2

Table 5-1 Results from Pilot PWR Plant Evaluation Using Fen Estimation Evaluation Procedure Compariso Input Fen Computation Identification of Sentinel Locations n

Design Worst Fen Criterion for Final System or Vessel Thermal Zone Component Basis Material D0 Temp Strain Rate T* O* S*

Fen Exp.

Worst O*

Fen Max.

Average Uen design (Uen*)

Initial Sentinel Initial Sentinel Location Count Criterion for Final Sentinel Location Sentinel Count CUF (Fen*) Locations Location RPV Outlet Nozzle 0.1078 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.265 NUREG/CR-6260 Y 1 NUREG/CR-6260 Y 1 RPV Nozzle RPV Inlet Nozzle 0.0795 LAS Low 558 Slow 14.36 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.195 NUREG/CR-6260 Y 1 NUREG/CR-6260 Y 1 RPV Core Exit Thermocouple Nozzle Assembly Upper 0.37 SS Low 619 Slow 1 0.26 1 -5.586 10.885 0.26 -6.908 15.348 13.117 4.853 Uen > 1.0 Y 1 Highest SS Uen in T.Z. Y 1 Nozzle Housing RPV Core Exit Thermocouple

< 50% of top Nozzle Assembly Head Port 0.123 SS Low 619 Slow 1 0.26 1 -5.586 10.885 0.26 -6.908 15.348 13.117 1.613 N 0 N 0 RPV Upper location Adapter Head Uen < 1.0; sole Reactor RPV Vessel Flange 0.196 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.481 Y 1 Uen < 0.8 N 0 location Pressure Vessel CRDM Housing 0.1093 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.268 N 0 N 0 RPV Head Flange 0.0155 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.038 N 0 N 0 Uen < 1.0; sole RPV Vessel Wall Transition 0.0105 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.026 Y 1 Uen < 0.8 N 0 location RPV Bottom Head-to-Shell 0.007 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.017 NUREG/CR-6260 Y 1 NUREG/CR-6260 Y 1 RPV Bottom Juncture Head RPV Bottom Head Instrument Highest Ni-Cr-Fe Uen in 0.3184 Ni-Cr-Fe Low 619 Slow 1 0.16 1 -8.112 3.662 0.16 -9.433 4.524 4.093 1.303 Uen > 1.0 Y 1 Y 1 Tubes (pos. 2) T.Z.

RPV Bottom Head Instrument < 50% of top 0.0014 Ni-Cr-Fe Low 619 Slow 1 0.16 1 -8.112 3.662 0.16 -9.433 4.524 4.093 0.006 N 0 N 0 Tubes (pos. 1) location Pressurizer Heater 0.562 SS Low 653 Slow 1 0.26 1 -5.586 10.885 0.26 -6.908 15.348 13.117 7.372 Uen > 1.0 Y 1 Highest SS Uen in T.Z. Y 1 Penetration

< 50% of top Pressurizer Immersion Heater 0.123 SS Low 653 Slow 1 0.26 1 -5.586 10.885 0.26 -6.908 15.348 13.117 1.613 N 0 N 0 location Low- < 50% of top Pressurizer Heater Well 0.128 SS Low 653 1 0.26 1 -4.217 7.624 0.26 -6.908 15.348 11.486 1.470 N 0 N 0 Mid location Pressurizer Thermowells 0 SS Low 653 Slow 1 0.26 1 -5.586 10.885 0.26 -6.908 15.348 13.117 0.000 N 0 N 0 Pressurizer Shell at Support Low-0.992 LAS Low 653 19.7 0 0.015 -5.133 2.455 0 -6.908 2.455 2.455 2.435 Uen > 1.0 Y 1 Highest LAS Uen in T.Z. Y 1 Pressurizer Lug Mid Lower Head > 50% of top Pressurizer Surge Nozzle 0.963 LAS Low 653 V.Slow 19.7 0 0.015 -6.908 2.455 0 -6.908 2.455 2.455 2.364 Y 1 > 50% of top location Y 1 location Pressurizer Lower Low- Within 25% of top 0.734 LAS Low 653 19.7 0 0.015 -5.133 2.455 0 -6.908 2.455 2.455 1.802 Y 1 Within 25% of top location Y 1 Head/Support Skirt Mid location Low- Not in top 3 Pressurizer Instrument Nozzle 0.236 LAS Low 653 19.7 0 0.015 -5.133 2.455 0 -6.908 2.455 2.455 0.579 N 0 N 0 Mid locations Pressurizer Manway Pad 0.141 LAS Low 653 Slow 19.7 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.346 N 0 N 0 Low-Pressurizer Lower Head 0.112 LAS Low 653 19.7 0 0.015 -5.133 2.455 0 -6.908 2.455 2.455 0.275 N 0 N 0 Mid Pressurizer Pressurizer Spray Pressurizer Spray Nozzle 0.411 SS Low 653 High 1 0.26 1 -0.192 2.678 0.26 -6.908 15.348 9.013 3.704 Uen > 1.0 Y 1 Highest SS Uen in T.Z. Y 1 Nozzle 6-inch and 3-inch Pressurizer Low-0.975 SS Low 653 1 0.26 1 -4.217 7.624 0.26 -6.908 15.348 11.486 11.199 Uen > 1.0 Y 1 Highest SS Uen in T.Z. Y 1 Safety and Relief Valve Piping Mid Pressurizer 3-inch x 6-inch Low- > 50% of top Power Operated Relief Valve 0.68 SS Low 653 1 0.26 1 -4.217 7.624 0.26 -6.908 15.348 11.486 7.811 Y 1 > 50% of top location Y 1 Mid location Solenoid Pressurizer Pressurizer Safety/Relief Low- Not within 25% of SRV/PORV 0.169 SS Low 653 1 0.26 1 -4.217 7.624 0.26 -6.908 15.348 11.486 1.941 N 0 N 0 Valve Nozzle Mid top value Pressurizer 3-inch x 6-inch Low- Not in top 3 0.139 SS Low 653 1 0.26 1 -4.217 7.624 0.26 -6.908 15.348 11.486 1.597 N 0 N 0 Power Operated Relief Valve Mid locations Pressurizer 6-inch Pressurizer Low-0.018 SS Low 653 1 0.26 1 -4.217 7.624 0.26 -6.908 15.348 11.486 0.207 N 0 N 0 Safety Valve Mid The SS 6-inch and 3-inch Pressurizer Safety and Low-Pressurizer Instrument Nozzle 0.236 SS Low 653 1 0.26 1 -4.217 7.624 0.26 -6.908 15.348 11.486 2.711 Uen > 1.0 Y 1 Relief Valve Piping bound N 0 Pressurizer Mid the SS Pressurizer Upper Head Instrument Nozzle

< 50% of top Pressurizer Thermowells 0 SS Low 653 Slow 1 0.26 1 -5.586 10.885 0.26 -6.908 15.348 13.117 0.000 N 0 N 0 location 5-3

Table 5-1 (continued)

Results from Pilot PWR Plant Evaluation Using Fen Estimation Evaluation Procedure Compariso Input Fen Computation Identification of Sentinel Locations n

Design Worst Fen Criterion for Final System or Vessel Thermal Zone Component Basis Material D0 Temp Strain Rate T* O* S*

Fen Exp.

Worst O*

Fen Max.

Average Uen design (Uen*)

Initial Sentinel Initial Sentinel Location Count Criterion for Final Sentinel Location Sentinel Count CUF (Fen*) Locations Location The SS 6-inch and 3-inch Pressurizer Safety and Pressurizer Upper Low-0.928 LAS Low 653 19.7 0 0.015 -5.133 2.455 0 -6.908 2.455 2.455 2.278 Uen > 1.0 Y 1 Relief Valve Piping bound N 0 Head/Upper Shell Mid the LAS Pressurizer Upper Head/Upper Shell.

Low-14-inch Hot Leg Surge Nozzle 0.3 SS Low 653 1 0.26 1 -4.217 7.624 0.26 -6.908 15.348 11.486 3.446 NUREG/CR-6260 Y 1 NUREG/CR-6260 Y 1 Surge Surge Line Mid Piping Piping 14-inch Pressurizer Surge < 50% of top 0.099 SS Low 653 Slow 1 0.26 1 -5.586 10.885 0.26 -6.908 15.348 13.117 1.299 N 0 N 0 Line (includes Thermowell) location SS Pressurizer Spray 4-inch Spray Piping at Mid-0.84 SS Low 558 0.813 0.26 1 -1.526 3.516 0.26 -6.908 10.964 7.240 6.082 Uen > 1.0 Y 1 Nozzle bounds SS N 0 Pressurizer Spray Nozzle High Pressurizer Spray Piping Accounted for in Spray Spray Line Auxiliary Spray Piping 0.72 SS Low 500 Medium 0.629 0.26 1 -2.856 4.062 0.26 -6.908 7.877 5.970 4.298 N 0 N 0 CVCS T.Z.

Piping Piping 4-inch Pressurizer Spray Mid- Not within 25% of 0.25 SS Low 558 0.813 0.26 1 -1.526 3.516 0.26 -6.908 10.964 7.240 1.810 N 0 N 0 Piping High top value Pressurizer Spray Line Mid-0.021 SS Low 558 0.813 0.26 1 -1.526 3.516 0.26 -6.908 10.964 7.240 0.152 N 0 N 0 Thermowell High CVCS 3-inch Cold Leg Loop 1 Mid-0.9 SS Low 558 0.813 0.26 1 -1.526 3.516 0.26 -6.908 10.964 7.240 6.516 NUREG/CR-6260 Y 1 NUREG/CR-6260 Y 1 Normal Charging Nozzle High The CVCS 3-inch Cold Charging Leg Loop 1 Normal Loop 1 CVCS 3-inch Normal > 50% of top 0.93 SS Low 500 Medium 0.629 0.26 1 -2.856 4.062 0.26 -6.908 7.877 5.970 5.552 Y 1 Charging Nozzle bounds N 0 Charging Loop 1 Piping location the CVCS 3-inch Normal Charging Loop 1 Piping.

CVCS 3-inch Cold Leg Loop 4 Mid-0.9 SS Low 558 0.813 0.26 1 -1.526 3.516 0.26 -6.908 10.964 7.240 6.516 NUREG/CR-6260 Y 1 NUREG/CR-6260 Y 1 Alternate Charging Nozzle High The CVCS 3-inch Cold Charging Leg Loop 4 Normal Loop 4 CVCS 3-inch Alternate > 50% of top 0.93 SS Low 500 Medium 0.629 0.26 1 -2.856 4.062 0.26 -6.908 7.877 5.970 5.552 Y 1 Charging Nozzle bounds N 0 Charging Loop 4 Piping location the CVCS 3-inch Normal Charging Loop 4 Piping.

The SS Charging Nozzle bounds the SS Letdown CVCS 3-inch Normal Letdown 0.95 SS Low 558 Medium 0.813 0.26 1 -2.856 4.657 0.26 -6.908 10.964 7.811 7.420 Uen > 1.0 Y 1 Nozzle on a Common N 0 Piping, Crossover Loop 3 Letdown Basis Stress Evaluation Loop 3 basis.

CVCS 3-inch Crossover Leg

< 50% of top Loop 3 Normal Letdown 0.1 SS Low 558 Medium 0.813 0.26 1 -2.856 4.657 0.26 -6.908 10.964 7.811 0.781 N 0 N 0 location Nozzle CVCS The SS Charging Nozzle CVCS 2-inch Crossover Leg bounds the SS Letdown Loop 4 Excess Letdown 0.804 SS Low 558 Medium 0.813 0.26 1 -2.856 4.657 0.26 -6.908 10.964 7.811 6.280 Uen > 1.0 Y 1 Nozzle on a Common N 0 Letdown Nozzle Basis Stress Evaluation Loop 4 basis.

CVCS 2-inch Excess Letdown < 50% of top 0.099 SS Low 558 Medium 0.813 0.26 1 -2.856 4.657 0.26 -6.908 10.964 7.811 0.773 N 0 N 0 Piping, Crossover Loop 4 location Auxiliary Auxiliary Spray Piping 0.72 SS Low 500 Medium 0.629 0.26 1 -2.856 4.062 0.26 -6.908 7.877 5.970 4.298 Uen > 1.0 Y 1 Highest SS Uen in T.Z. Y 1 Spray The SS Charging Nozzle bounds the SS Drain CVCS Drain Line, Loop 2 0.95 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 9.147 Uen > 1.0 Y 1 Nozzle on the basis of N 0 Common Basis Stress Evaluation.

The SS Charging Nozzle bounds the SS Drain Drain > 50% of top CVCS Drain Line, Loop 3 0.95 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 9.147 Y 1 Nozzle on a Common N 0 location Basis Stress Evaluation basis.

Not within 25% of CVCS Drain Line, Loop 1 0.09 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 0.867 N 0 N 0 top value Not within 25% of CVCS Drain Line, Loop 4 0.09 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 0.867 N 0 N 0 top value 5-4

Table 5-1 (continued)

Results from Pilot PWR Plant Evaluation Using Fen Estimation Evaluation Procedure Compariso Input Fen Computation Identification of Sentinel Locations n

Design Worst Fen Criterion for Final System or Vessel Thermal Zone Component Basis Material D0 Temp Strain Rate T* O* S*

Fen Exp.

Worst O*

Fen Max.

Average Uen design (Uen*)

Initial Sentinel Initial Sentinel Location Count Criterion for Final Sentinel Location Sentinel Count CUF (Fen*) Locations Location The SS Charging Nozzle, CVCS 1-1/2-inch, 2-inch Seal Letdown Nozzle and Drain Water Injection Loops 3 0.114 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 1.098 Uen > 1.0 Y 1 N 0 Line bound the SS Seal Piping Water Piping.

CVCS 1-1/2-inch, 2-inch Seal Seal Water > 50% of top Water Injection Loops 4 0.067 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 0.645 Y 1 Uen < 0.8 N 0 location Piping CVCS 1-1/2-inch, 2-inch Seal Not within 25% of Water Injection Loops 1, 2 0.066 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 0.635 N 0 N 0 top value Piping RCP Casing/Discharge 0.915 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 8.810 Uen > 1.0 Y 1 Highest SS Uen in T.Z. Y 1 Nozzle Junction The SS Charging Nozzle bounds the crossover leg RCS 2-inch Crossover Leg > 50% of top 0.7 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 6.740 Y 1 components on a N 0 Loops 1, 2 Drain Nozzles location Common Basis Stress RCS Cold Evaluation basis.

Leg RCS Crossover Leg Loops 1, Not within 25% of 0.5 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 4.814 N 0 N 0 2, 3, 4 top value RCS Cold Leg Loops 1, 2, 3, Not in top 3 RCS 0.37 SS Low 558 Slow 0.813 0.26 1 -5.586 8.292 0.26 -6.908 10.964 9.628 3.562 N 0 N 0 4 locations Uen < 1.0; sole RCS Cold Leg Thermowells 0.025 Ni-Cr-Fe Low 558 Slow 0.899 0.16 1 -8.112 3.212 0.16 -9.433 3.885 3.549 0.089 Y 1 Uen < 0.8 N 0 location The Hot Leg Surge Nozzle bounds the hot leg RCS Hot Leg Loops 1, 2, 3, 4 0.95 SS Low 619 Slow 1 0.26 1 -5.586 10.885 0.26 -6.908 15.348 13.117 12.461 Uen > 1.0 Y 1 components on a N 0 RCS Hot Leg Common Basis Stress Evaluation basis Uen < 1.0; sole RCS Hot Leg Thermowells 0.017 Ni-Cr-Fe Low 619 Slow 1 0.16 1 -8.112 3.662 0.16 -9.433 4.524 4.093 0.070 Y 1 Uen < 0.8 N 0 location RHR 12-inch Hot Leg Loops 0.81 SS Low 619 Medium 1 0.26 1 -2.856 5.352 0.26 -6.908 15.348 10.350 8.384 Uen > 1.0 Y 1 NUREG/CR-6260 Y 1 1, 4 RHR Nozzles RHR 12-inch RHR Loops 1, 4 Low- < 50% of top 0.661 SS Low 390 0.279 0.26 1 -4.217 3.460 0.26 -6.908 4.207 3.833 2.534 N 0 N 0 (Hot Leg SI portion) Mid location RHR 12-inch RHR Pump Low- < 50% of top RHR Inlet 0.64 SS Low 390 0.279 0.26 1 -4.217 3.460 0.26 -6.908 4.207 3.833 2.453 N 0 N 0 Suction Isolation Valves Mid location RHR 12-inch RCS Hot Leg to Low- < 50% of top 0.64 SS Low 390 0.279 0.26 1 -4.217 3.460 0.26 -6.908 4.207 3.833 2.453 N 0 N 0 RHR Pump Isolation Valves Mid location RHR RHR 12-inch RHR Suction , Low- < 50% of top 0.296 SS Low 390 0.279 0.26 1 -4.217 3.460 0.26 -6.908 4.207 3.833 1.135 N 0 N 0 Loops 1, 4 Mid location RHR 6-inch SI & RHR System Low- Uen < 1.0; sole Loop 2 & 3 Recirculation 0.17 SS Low 390 0.279 0.26 1 -4.217 3.460 0.26 -6.908 4.207 3.833 0.652 Y 1 Uen < 0.8 N 0 Mid location Supply Header Check Valves RHR Outlet RHR 6-inch SI/RHR Isolation Low- Uen < 1.0; sole 0.17 SS Low 390 0.279 0.26 1 -4.217 3.460 0.26 -6.908 4.207 3.833 0.652 Y 1 Uen < 0.8 N 0 Valves Mid location RHR 6-inch RHR Pumps to Low-0.165 SS Low 390 0.279 0.26 1 -4.217 3.460 0.26 -6.908 4.207 3.833 0.633 N 0 N 0 RCS Cold Leg Check Valves Mid SI 3-inch Cold Leg (All Loops) 0.999 SS Low 558 Medium 0.813 0.26 1 -2.856 4.657 0.26 -6.908 10.964 7.811 7.803 NUREG/CR-6260 Y 1 NUREG/CR-6260 Y 1 Boron Injection Nozzle SI 1.5-inch BIT Line Loops 1, < 50% of top BIT 0.93 SS Low 160 Medium 0 0.26 1 -2.856 2.547 0.26 -6.908 2.547 2.547 2.369 N 0 N 0 2, 3, 4 location SI 1.5-inch BIT Line Common < 50% of top 0.773 SS Low 160 Medium 0 0.26 1 -2.856 2.547 0.26 -6.908 2.547 2.547 1.969 N 0 N 0 Header location SI 10-inch Cold Leg (All 0.95 SS Low 558 Medium 0.813 0.26 1 -2.856 4.657 0.26 -6.908 10.964 7.811 7.420 Uen > 1.0 Y 1 Highest SS Uen in T.Z. Y 1 Loops) Accumulator Nozzle SI SI 10-inch Accumulator Line < 50% of top 0.98 SS Low 160 Medium 0 0.26 1 -2.856 2.547 0.26 -6.908 2.547 2.547 2.496 N 0 N 0 Loops 1, 2, 3, 4 Piping location SI 10-inch Cold Leg (All Accumulator Loops) Accumulator Nozzle 0.26 SS Low 160 Medium 0 0.26 1 -2.856 2.547 0.26 -6.908 2.547 2.547 0.662 N 0 N 0 Check Valves SI 10-inch Accumulator Line Loops 1, 2, 3, 4 Outlet 0.26 SS Low 160 Medium 0 0.26 1 -2.856 2.547 0.26 -6.908 2.547 2.547 0.662 N 0 N 0 Upstream Check Valves 5-5

Table 5-1 (continued)

Results from Pilot PWR Plant Evaluation Using Fen Estimation Evaluation Procedure Compariso Input Fen Computation Identification of Sentinel Locations n

Design Worst Fen Criterion for Final System or Vessel Thermal Zone Component Basis Material D0 Temp Strain Rate T* O* S*

Fen Exp.

Worst O*

Fen Max.

Average Uen design (Uen*)

Initial Sentinel Initial Sentinel Location Count Criterion for Final Sentinel Location Sentinel Count CUF (Fen*) Locations Location SI 6-inch Hot Leg Loops 2, 3 0.1 SS Low 619 Medium 1 0.26 1 -2.856 5.352 0.26 -6.908 15.348 10.350 1.035 Uen > 1.0 Y 1 Highest SS Uen in T.Z. Y 1 SIS Nozzle SIS SI 6-inch Hot Leg Loops 2,3 < 50% of top 0.09 SS Low 160 Medium 0 0.26 1 -2.856 2.547 0.26 -6.908 2.547 2.547 0.229 N 0 N 0 SIS Piping location The LAS RSG Tubesheet RSG Primary Manway Drain Uen < 1.0; sole (Continuous Region) 0.391 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.960 Y 1 N 0 Tube location bounds the LAS Primary Head Locations.

Primary RSG Primary Manway Cover 0.349 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.857 N 0 N 0 Head RSG Primary Nozzle Drain 0.316 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.776 N 0 N 0 Tube RSG Primary Inlet Nozzle 0.009 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.022 N 0 N 0 Steam RSG Primary Outlet Nozzle 0.007 LAS Low 558 Slow 14.36 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.017 N 0 N 0 Generator RSG Tubesheet (Continuous 0.428 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 1.051 Uen > 1.0 Y 1 Highest LAS Uen in T.Z. Y 1 Region)

RSG Tubesheet (Perforated < 50% of top Tubesheet 0.122 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.299 N 0 N 0 Region) location RSG Tube-to-Tubesheet Uen < 1.0; sole 0.068 Ni-Cr-Fe Low 619 Slow 1 0.16 1 -8.112 3.662 0.16 -9.433 4.524 4.093 0.278 Y 1 Uen < 0.8 N 0 Connection location RSG Channel Head at Uen < 1.0; sole 0.059 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.145 Y 1 Uen < 0.8 N 0 Channel Primary Manway location Head RSG Channel Head near 0.004 LAS Low 619 Slow 17.79 0 0.015 -6.502 2.455 0 -6.908 2.455 2.455 0.010 N 0 N 0 Tubesheet 44 22 5-6

Table 5-2 Final Sentinel Locations for PWR Pilot Plant System or Sentinel Locations Thermal Zone Material Component Vessel Initial Final Reactor RPV Outlet Nozzle LAS 1 1 RPV Nozzle Pressure RPV Inlet Nozzle LAS 1 1 Vessel RPV Core Exit Thermocouple Nozzle Assembly Upper RPV Upper SS 1 1 Nozzle Housing Head RPV Vessel Flange LAS 1 0 RPV Vessel Wall Transition LAS 1 0 RPV Bottom RPV Bottom Head-to-Shell Juncture LAS 1 1 Head RPV Bottom Head Instrument Tubes (pos. 2) Ni-Cr-Fe 1 1 Pressurizer Heater Penetration SS 1 1 Pressurizer Pressurizer Shell at Support Lug LAS 1 1 Lower Head Pressurizer Surge Nozzle LAS 1 1 Pressurizer Lower Head/Support Skirt LAS 1 1 Pressurizer Pressurizer Spray Nozzle SS 1 1 Pressurizer Spray Nozzle 6-inch and 3-inch Pressurizer Safety and Relief Valve Piping SS 1 1 Pressurizer Pressurizer 3-inch x 6-inch Power Operated Relief Valve SRV/PORV SS 1 1 Solenoid Pressurizer Pressurizer Instrument Nozzle SS 1 0 Upper Head Pressurizer Upper Head/Upper Shell LAS 1 0 Surge Surge Line 14-inch Hot Leg Surge Nozzle SS 1 1 Piping Piping Spray Spray Line 4-inch Spray Piping at Pressurizer Spray Nozzle SS 1 0 Piping Piping CVCS 3-inch Cold Leg Loop 1 Normal Charging Nozzle SS 1 1 Charging Loop 1 CVCS 3-inch Normal Charging Loop 1 Piping SS 1 0 CVCS 3-inch Cold Leg Loop 4 Alternate Charging Nozzle SS 1 1 Charging Loop 4 CVCS 3-inch Alternate Charging Loop 4 Piping SS 1 0 Letdown Loop 3 CVCS 3-inch Normal Letdown Piping, Crossover Loop 3 SS 1 0 CVCS Letdown Loop 4 CVCS 2-inch Crossover Leg Loop 4 Excess Letdown Nozzle SS 1 0 Auxiliary Spray Auxiliary Spray Piping SS 1 1 CVCS Drain Line, Loop 2 SS 1 0 Drain CVCS Drain Line, Loop 3 SS 1 0 CVCS 1-1/2-inch, 2-inch Seal Water Injection Loops 3 Piping SS 1 0 Seal Water CVCS 1-1/2-inch, 2-inch Seal Water Injection Loops 4 Piping SS 1 0 RCP Casing/Discharge Nozzle Junction SS 1 1 RCS Cold Leg RCS 2-inch Crossover Leg Loops 1, 2 Drain Nozzles SS 1 0 RCS RCS Cold Leg Thermowells Ni-Cr-Fe 1 0 RCS Hot Leg Loops 1, 2, 3, 4 SS 1 0 RCS Hot Leg RCS Hot Leg Thermowells Ni-Cr-Fe 1 0 RHR Inlet RHR 12-inch Hot Leg Loops 1, 4 RHR Nozzles SS 1 1 RHR 6-inch SI & RHR System Loop 2 & 3 Recirculation RHR SS 1 0 RHR Outlet Supply Header Check Valves RHR 6-inch SI/RHR Isolation Valves SS 1 0 BIT SI 3-inch Cold Leg (All Loops) Boron Injection Nozzle SS 1 1 SI Accumulator SI 10-inch Cold Leg (All Loops) Accumulator Nozzle SS 1 1 SIS SI 6-inch Hot Leg Loops 2, 3 SIS Nozzle SS 1 1 Primary Head RSG Primary Manway Drain Tube LAS 1 0 Steam RSG Tubesheet (Continuous Region) LAS 1 1 Tubesheet Generator RSG Tube-to-Tubesheet Connection Ni-Cr-Fe 1 0 Channel Head RSG Channel Head at Primary Manway LAS 1 0 44 22 5-7

Table 5-3 Sample Results from Pilot PWR Plant Evaluation Using Common Basis Stress Evaluation Procedure (Properties)

System Component Type Material OD t ID k E (ksi) K2 K3 Sm S1e6 1.00E-SS 3.5 0.437 2.626 10 2.67E+04 1.1 1.1 16.7 28.3 05 Charging Nozzle CVCS 1.00E-SS 3.5 0.437 2.626 10 2.67E+04 1.1 1.1 16.7 28.3 05 Charging Piping 5-8

Table 5-4 Sample Results from Pilot PWR Plant Evaluation Using Common Basis Stress Evaluation Procedure (Transient Details)

Transient 1 System Component Detailed Type Name V Q (gpm) Tmean Tmax h k/ht t0 cp kt0/cpt^2 T ET/(1-) (max)/(ET/(1-)) Sts (detailed) Smom Sp (detailed)

LOC/LOL Trip and Return Down 5 84.24901871 330 560 267.35 1632.30 0.168 10 100 0.21 460 175 84 55 139 0.48 LOC Delayed Down 5 84.24901871 330 560 267.35 1632.30 0.168 0.01 100 0.00 460 175 88 55 143 0.50 Charging LOL Delayed Down 1 3.8 64.02925422 315 530 263.46 1291.48 0.213 100 100 2.09 430 164 25 48 72 Nozzle 0.15 LOL Delayed Down 2 3.8 64.02925422 335 570 268.57 1316.52 0.209 10 100 0.21 470 179 81 55 136 CVCS 0.45 LOL Prompt Down 5 84.24901871 300 500 259.23 1582.71 0.173 100 100 2.09 400 153 23 44 67 0.15 LOC/LOL Trip and Return Down 55 300 500 259.23 1125.22 0.244 100 100 2.09 400 153 21 41 63 0.14 LOC Delayed Down 55 330 560 267.35 1160.48 0.237 100 100 2.09 460 175 25 48 72 Charging 0.14 Piping LOL Delayed Down 1 55 300 500 259.23 1125.22 0.244 100 100 2.09 400 153 21 41 63 0.14 LOL Prompt Down 55 300 500 259.23 1125.22 0.244 100 100 2.09 400 153 21 41 63 0.14 PPSMAX 0.66 Transient 2 System Component Detailed Type Name V gpm Tmean Tmax h k/ht t0 cp kt0/cpt^2 T ET/(1-) (max)/(ET/(1-)) Sts (detailed) Smom Sp (detailed)

LOC/LOL Trip and Return Up 5 84.24901871 300 500 259.23 1582.71 0.173 90 100 1.885 400 153 23 44 67 0.15 LOC Delayed Up 5 84.24901871 300 500 259.23 1582.71 0.173 90 100 1.885 400 153 23 44 67 0.15 Charging LOL Delayed Up 1 3.8 64.02925422 335 570 268.57 1316.52 0.209 0.01 100 0.000 470 179 86 55 141 Nozzle 0.48 LOL Delayed Up 2 3.8 64.02925422 315 530 263.46 1291.48 0.213 100 100 2.095 430 164 25 48 72 0.15 CVCS LOL Prompt Up 5 84.24901871 300 500 259.23 1582.71 0.173 100 100 2.09 400 153 23 44 67 0.15 LOC/LOL Trip and Return Up 55 300 500 259.23 1125.22 0.244 100 100 2.095 400 153 23 44 67 0.15 LOC Delayed Up 55 330 560 267.35 1160.48 0.237 100 100 2.095 460 175 26 51 77 Charging 0.15 Piping LOL Delayed Up 1 55 300 500 259.23 1125.22 0.244 100 100 2.095 400 153 23 44 67 0.15 LOC/LOL Trip and Return Up 55 300 500 259.23 1125.22 0.244 100 100 2.095 400 153 23 44 67 0.15 PPSMAX 0.66 5-9

Table 5-5 Sample Results from Pilot PWR Plant Evaluation Using Common Basis Stress Evaluation Procedure (Computed Values)

Combined Design Projected System Component Fen factor Transient Pair Detailed Tensile Compressive Tensile Compressive Type Speak Ke Sa Nallow n design Uincr Uincr Uincr n proj Uincr proj Uincr Uincr T* O* used

Worst

Fen expect Fen worst Fen average LOC/LOL Trip and Return Up/Down 207 2.370 245 305 60 0.1967 0.1326 0.0641 1 0.0033 0.0022 0.0011 0.82 0.26 0.0521 0.0521 -2.037491708 -6.907755279 3.93 9.07 6.50 LOC Delayed Up/Down 210 3.190 335 86 20 0.2326 0.1580 0.0745 1 0.0116 0.0079 0.0037 0.82 0.26 53.4564 0.4000 0 -6.907755279 2.55 9.07 5.81 Charging Nozzle LOL Delayed Up 1/Down 1 213 3.190 340 86 20 0.2326 0.0789 0.1537 1 0.0116 0.0039 0.0077 0.85 0.26 0.0027 0.0027 -4.994484904 -6.907755279 7.69 9.60 8.65 CVCS LOL Delayed Up 2/Down 2 208 2.430 253 270 20 0.0741 0.0483 0.0258 1 0.0037 0.0024 0.0013 0.85 0.26 0.0508 0.0508 -2.063301759 -6.907755279 4.02 9.60 6.81 LOL Prompt Up/Down 135 2.054 138 567 200 0.3527 0.1764 0.1764 1 0.0018 0.0009 0.0009 0.63 0.26 0.0025 0.0025 -5.066805566 -6.907755279 5.83 6.44 6.14 LOC/LOL Trip and Return Up/Down 130 2.054 134 567 60 0.1058 0.0511 0.0547 1 0.0018 0.0009 0.0009 0.63 0.26 0.0024 0.0024 -5.135798437 -6.907755279 5.90 6.44 6.17 Charging LOC Delayed Up/Down 150 2.180 163 436 20 0.0459 0.0221 0.0237 1 0.0023 0.0011 0.0012 0.82 0.26 0.0027 0.0027 -4.996036495 -6.907755279 7.38 9.07 8.23 Piping LOL Delayed Up 1/Down 1 130 2.054 134 567 20 0.0353 0.0170 0.0182 1 0.0018 0.0009 0.0009 0.63 0.26 0.0024 0.0024 -5.135798437 -6.907755279 5.90 6.44 6.17 LOC/LOL Trip and Return Up/Down 130 2.054 134 567 200 0.3527 0.1703 0.1824 1 0.0018 0.0009 0.0009 0.63 0.26 0.0024 0.0024 -5.135798437 -6.907755279 5.90 6.44 6.17 Design Uen (Uen*) Projected Uen (Uen*)

System Component Transient Pair Tensile Compressive Total Sum Tensile Compressive Total Sum Rank Rank Type Uen incr Uen incr Uen total Uen Uen incr Uen incr Uen total Uen LOC/LOL Trip and Return 0.8620 0.0641 0.9262 4.3676 1 0.0069 0.0011 0.0080 0.1239 1 Up/Down LOC Delayed Up/Down 0.9180 0.0745 0.9925 0.0216 0.0037 0.0254 Charging Nozzle LOL Delayed Up 1/Down 1 0.6817 0.1537 0.8355 0.0664 0.0077 0.0741 LOL Delayed Up 2/Down 2 0.3291 0.0258 0.3548 0.0088 0.0013 0.0101 CVCS LOL Prompt Up/Down 1.0823 0.1764 1.2586 0.0054 0.0009 0.0063 LOC/LOL Trip and Return 0.3152 0.0547 0.3699 1.9321 2 0.0056 0.0009 0.0065 0.0306 2 Up/Down Charging LOC Delayed Up/Down 0.1822 0.0237 0.2059 0.0098 0.0012 0.0109 Piping LOL Delayed Up 1/Down 1 0.1051 0.0182 0.1233 0.0056 0.0009 0.0065 LOC/LOL Trip and Return 1.0506 0.1824 1.2330 0.0056 0.0009 0.0065 Up/Down 5-10

Section 6: Concluding Remarks This report provides the technical basis of a screening process that can be used to evaluate a plant to determine EAF limiting locations for fatigue monitoring.

Procedures for this example screening evaluation are described and applied to a pilot PWR plant.

A primary reason for developing this process is to equip license renewal applicants with a consistent method to identify EAF limiting locations additional to the sample locations evaluated in NUREG/CR-6260 for their reactor type and vintage.

Guiding principles for the process included:

1. Consistent technical basis.

2. Analytical method using readily available design input from P&IDs, piping isometric drawings and piping stress reports.

3. Only basic stress or fatigue analysis required.

The following are the basic areas of new technology developed by this project:

1. Procedure for Estimating Fen Factors.

2. Procedure for Estimating Uen.

Each of these areas is discussed in detail. An example of the process is shown in Section 5.

The process developed in this report provides guidance for the evaluation and relative ranking of estimated Uen values for locations in components and systems where EAF is a concern to minimize the possibility of the need for a formal fatigue evaluation. The estimated values for a number of Sentinel Locations are compared to Uen values for locations in each given system/component that have been specifically identified in regulatory guidance as of concern for EAF.

Locations from previous guidance are to be managed and if estimated values for other locations are higher or as high, these other locations should also be managed. For components/systems where there are no locations included in previous regulatory guidelines, the recommendation is to manage up to the three locations with highest estimated Uen values.

Further analysis beyond the basic screening steps may be applied to reduce the number of locations.

6-1

Section 7: References

1. ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Power Plant Components," American Society of Mechanical Engineers.

2. ANSI/ASME B31.7-1969, Nuclear Power Piping, American National Standards Institute.

3. ANSI/ASME B31.1, Power Piping, American National Standards Institute.

4. NUREG/CR-6260 (INEL-95/0045), Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, March 1995.

5. NUREG/CR-5704 (ANL-98/31), Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels, 1999.

6. NUREG/CR-6583 (ANL-97/18), Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, 1998.

7. NUREG/CR-6909 (ANL-06/08), Effects of LWR Coolant Environments on the Fatigue Life of Reactor Materials Final Report, February 2007.

8. NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report, U.

S. Nuclear Regulatory Commission, December 2010.

9. EPRI Report "Fatigue Comparison of Piping Designed to ANSI B31.1 and ASME Section III, Class 1 Rules," TR-102901, EPRI, Palo Alto, CA, December 1993.

10. EPRI Technical Report, Improved Basis and Requirements for Break Location Postulation, EPRI, Palo Alto, CA: 2011. 1022873.

11. Materials Reliability Program: Guidelines for Addressing Fatigue Environmental Effects in a License Renewal Application, MRP-47, Revision 1, September 2005.

12. Materials Reliability Program: Fatigue Management Handbook, Revision 1 (MRP-235 w/ corrections), EPRI, Palo Alto, CA: 2009. 1015010.

13. EPRI/BWRVIP Memo. No. 2005-271, Potential Error in Existing Fatigue Reactor Water Environmental Effects Analyses, July 1, 2005.

14. Regulatory Guide 1.207, Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components due to the Effects of Light-Water Reactor Environment for New Reactors, March 2007.

7-1

Appendix A: Summary of Fen Formulations as Accepted by the NRC Fen Formulations for Ferritic Materials NUREG/CR-6583 (Old Rules)

The following are the appropriate Fen relationships from NUREG/CR-6583 [6]

for carbon and low alloy steels. These expressions are:

For Carbon Steel [6, p. 69]:

Fen = exp(0.585 0.00124T '0.101S *T *O * * ) = exp(0.554 0.101S *T *O * * )

Equation A-1 For Low Alloy Steel [6, p. 69]:

Fen = exp(0.929 0.00124T '0.101S *T *O * * ) = exp(0.898 0.101S *T *O * * )

Equation A-2 Note that the above expressions have been corrected as summarized in Reference [13].

where:

Fen = fatigue life correction factor T' = 25°C (NUREG/CR-6583, Section 6, Fen relative to room temperature air)

S* = S for 0 < S 0.015 wt. %

= 0.015 for S > 0.015 wt. %

S = weight percent sulfur of steel T* = 0 for T < 150°C

= (T - 150) for 150 T 350°C T = service temperature (°C)

O* = 0 for DO < 0.05 parts per million (ppm)

= ln(DO/0.04) for 0.05 ppm DO 0.5 ppm

= ln(12.5) for DO > 0.5 ppm A-1

DO = dissolved oxygen

= 0 for > 1%/sec

= ln( ) for 0.001 1%/sec

= ln(0.001) for < 0.001%/sec

= strain rate, %/sec NUREG/CR-6909 (New Rules)

For Carbon Steel (CS) [7, p. A.1]:

Fen = exp(0.632 0.101S *T *O * * ) Equation A-3 For Low Alloy Steel (LAS) [7, p. A.1]:

Fen = exp(0.702 0.101S *T *O * * ) Equation A-4 Where S * , T * , O * , and

are the transformed sulfur content, service temperature, dissolved oxygen (DO), and strain rate, respectively, which are defined as follows [7, p. A.1 and A.2]:

Fen = fatigue life correction factor S = 0.001 for S 0.001 wt.%

= S for S 0.015 wt. %

= 0.015 for S > 0.015 wt. %

T* = 0 for T < 150°C

= (T - 150) for 150 T 350°C T = service temperature (°C)

O = 0 for dissolved oxygen, DO 0.04 parts per million (ppm)

= ln(DO/0.04) for 0.04 ppm < DO 0.5 ppm

= ln(12.5) for DO > 0.5 ppm

= 0 for strain rate, > 1%/sec

= ln( ) for 0.001 1%/sec

= ln(0.001) for < 0.001%/sec For both carbon and low-alloy steels, a threshold value of 0.07% for strain amplitude (one-half the strain range for the cycle) is defined, below which environmental effects on the fatigue life of these steels do not occur. This strain threshold corresponds to 21 ksi (145 MPa) alternating stress intensity from the fatigue analysis. That is, if alt 21 ksi then Fen = 1.0.

Fen = 1 for strain amplitude, a 0.07% or Salt (Ec)(0.07%)/(100%) = 21 ksi (145 MPa)

A-2

Fen Formulations for Austenitic Stainless Steel Materials NUREG/CR-5704 (Old Rules)

For Types 304 and 316 Stainless Steel [5]:

Fen = exp(0.935 T * *O * ) Equation A-5 where:

Fen = fatigue life correction factor T = service temperature of transient, °C T = 0 for T < 200°C

= 1 for T 200°C

= 0 for strain rate, > 0.4%/sec

= ln( /0.4) for 0.0004 0.4%/sec

= ln(0.0004/0.4) for < 0.0004%/sec O* = 0.260 for dissolved oxygen, DO < 0.05 parts per million (ppm)

= 0.172 for DO 0.05 ppm NUREG/CR-6909 (New Rules)

For wrought and cast austenitic stainless steels (SS) [7, p. A.2]:

Fen = exp(0.734 T ' O' ' ) Equation A-6 where:

Fen = fatigue life correction factor T' = 0 for T < 150°C

= (T - 150)/175 for 150 T < 325°C

= 1 for T 325°C T = service temperature (°C)

' = 0 for strain rate, > 0.4%/sec

= ln( /0.4) for 0.0004 0.4%/sec

= ln(0.0004/0.4) for < 0.0004%/sec O' = 0.281 for all dissolved oxygen levels For wrought and cast austenitic stainless steels, a threshold value of 0.10% for strain amplitude (one-half the strain range for the cycle) is defined, below which environmental effects on the fatigue life of these steels do not occur. This strain threshold corresponds to 28.3 ksi (195 MPa) alternating stress intensity from the fatigue analysis. That is, if Salt 28.3 ksi then Fen = 1.0. Thus, Fen = 1 for strain amplitude, a 0.10% or Salt (Ec)(0.10%)/(100%) = 28.3 ksi (195 MPa)

A-3

Fen Formulations for Nickel Alloy Materials NUREG/CR-6909 (New Rules)

New rules are required for Nickel Alloy materials per both the GALL report (license renewal) [8] and Reg. Guide 1.207 [14] (new plants).

For Ni-Cr-Fe alloys [7, p. A.2]:

Fen = exp(T ' ' O' ) Equation A-7 where:

Fen = fatigue life correction factor T' = T/325 for T < 325°C

= 1 for T 325°C T = service temperature (°C)

' = 0 for strain rate, > 5.0%/sec

= ln( /5.0) for 0.0004 5.0%/sec

= ln(0.0004/5.0) for < 0.0004%/sec O' = 0.09 for NWC BWR water

= 0.16 for PWR or HWC BWR water For Ni-Cr-Fe alloys, a threshold value of 0.10% for strain amplitude (one-half the strain range for the cycle) is defined, below which environmental effects on the fatigue life of these steels do not occur. This strain threshold corresponds to 28.3 ksi (195 MPa) alternating stress intensity from the fatigue analysis. That is, if Salt 28.3 ksi then Fen = 1.0. Thus, Fen = 1 for strain amplitude, a 0.10% or Salt (Ec)(0.10%)/(100%) = 28.3 ksi (195 MPa)

Dissolved Oxygen (DO)

This is a value determined by the evaluator. For PWRs, the value has typically been established as below 50 ppm for plant conditions in which EAF is active (temperature above 150°C). For BWRs, this value is determined based on several factors, such as use of Hydrogen Water Chemistry program or not, and others. This value is typically obtained from individuals responsible for maintaining and monitoring plant water chemistry.

Strain Rate ( )

This is determined as the strain difference over the increasingly tensile strain change divided by the time duration of the strain change.

A-4

Appendix B: Rules for Evaluation of Class 1 Piping Rules for Evaluation of Class 1 Piping This section is extracted in part from References [4] The ASME Section III rules for evaluation of Class 1 piping components have many similarities to the design by analysis" rules of NB-3200. NB-3650 is based on the maximum shear stress theory, and primary, secondary and peak stress categories are evaluated. The allowable stress limits for the different stress categories are the same as for NB-3200, and if the limitations on primary plus secondary stresses are exceeded, simplified elastic-plastic analysis with consideration of thermal stress ratcheting is allowed. The major difference between NB-3200 and NB-3650 is that the latter takes a "design by formula" approach, with the design being considered acceptable if it passes a series of equations for the various loadings to which the component is exposed.

After primary stress limits are satisfied, Equation 10 must be satisfied for all pairs of load sets:

Po Do D S n = C1 + C 2 o M i + C3Eab aT a - abT b 3 Sm 2t 2I (Section III, Cl. 1, Eq. 10) [Eq. 2-6 of Ref. [4)) Equation B-1 Where from [4]:

C1, C2, C3 = secondary stress indices for the specific component under investigation Do = outside diameter of pipe, in. (mm) t = nominal wall thickness of product, in. (mm)

I = moment of inertia, in4 (mm4)

Sm = allowable design stress intensity, psi (MPa)Po

= range of service pressure, psi (MPa)

Mi = resultant range of moment which occurs when the system goes from one service load set to another, in.-lb. (N-mm)

B-1

Eab = average modulus of elasticity of the two sides of a material or structural discontinuity at room temperature, psi (MPa) a,b = coefficient of thermal expansion on side a and side b of a structural or material discontinuity, in/in°F (mm/mm/°C)

Ta, Tb = range of average temperature on side a and side b of a structural discontinuity, when the system goes from one service load to another, °F (°C)

The fatigue resistance of each piping component is assessed by evaluating the range of peak stress. For every pair of load sets, S, values are calculated using Equation 11:

Po Do D 1 1 S =KC +K C o M +K C E x a T a T + K E a T + E a T p 1 1 2t 2 2 2I i 3 3 ab a a b b 2(1 v )

3 1 1 v 2 (Section III, Cl. 1, Eq. 11) [Eq. 2-7 of Ref. [4)) Equation B-2 where:

K 1, K2, K3 = local stress indices for the specific component under investigation E = modulus of elasticity (E) times the mean coefficient of thermal expansion (), both at room temperature, psi /°F (MPa/°C)

T 1 = absolute value of range of the temperature difference for each load set pair between the temperature of the outside surface T0 and the temperature of the inside surface Ti of the piping product assuming a moment generating equivalent linear temperature distribution, °F (°C)

T2 = absolute value of range for that portion of the nonlinear thermal gradient through the wall thickness not included in T1 , °F (°C)

A load set pair is defined as two loading sets or cases, which are used to compute a stress range.

If Equation 10 cannot be satisfied for all load set pairs, the alternative analysis described below may still permit qualifying the component. Only those load set pairs which do not satisfy Equation 10 need to be considered.

1. Equation 12 shall be met:

Do S e = C2 M i 3S m 2I Section III, C1. 1, Eq. 12) [Eq. 2-8 of Ref. [4)) Equation B-3 where:

Se = nominal value of expansion stress, psi (MPa)

B-2

Mi* = same as Mi in Equation 10, except that it includes only moments due to thermal expansion and thermal anchor movements, in-lb (N-mm)

2. The primary plus secondary membrane plus bending stress intensity, excluding thermal bending and thermal expansion stresses, shall be <3Sm.

This requirement is satisfied by meeting Equation 13:

Po Do DM C + C o i + CE T T 3S 1 2t 2 2I 3 ab a a b b m (Section III, Cl. 1, Eq. 13) [Eq. 2-9 of Ref. [4)) Equation B-4 where:

C3 = stress index

3. If these conditions are met, the value of Salt (also called Sa later in the Code and in this document) shall be calculated by Equation 14:

Sp S alt = K e 2

(Section III, C1. 1, Eq. 14) [Eq. 2-10 of Ref. [4)) Equation B-5 Where from [4]:

Salt = alternating stress intensity, psi (MPa)

Sp = peak stress intensity value calculated by Equation 11, psi (MPa)

Ke = 1.0 for Sn 3Sm [Eq. 2-11 of Ref. [4))

= 1.0 + [(1-n)/n(m-1)](Sn/3Sm -1),

for 3Sm < Sn < 3mSm [Eq. 2-12 of Ref. [4))

= 1/n, for Sn 3mSm [Eq. 2-13 of Ref. [4))

Sn = primary plus secondary stress intensity value calculated in Equation 10, psi (MPa) m,n = material parameters The alternating stress for all load set pairs is then computed as one-half of the peak stress intensity values adjusted by Ke. The fatigue analysis is then performed using the applicable Code fatigue curve and the number of design cycles for each load case from the Design Specification.

B-3

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Attachment 3 NRC Staffs License Renewal Team Inspection Report 05000247/2013010 September 19, 2013 (ML13263A020)

Publicly available at:

http://pbadupws.nrc.gov/docs/ML1326/ML13263A020.pdf

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 2100 RENAISSANCE BOULEVARD, SUITE 100 KING OF PRUSSIA, PENNSYLVANIA 19406-2713 September 19, 2013 Mr. John Ventosa Site Vice President Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT 2 - NRC LICENSE RENEWAL TEAM INSPECTION REPORT 05000247/2013010

Dear Mr. Ventosa:

On September 12, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Indian Point Nuclear Generating Unit 2. The enclosed inspection report documents the results of our review of your completed actions for the remaining 10 commitments, which were discussed on September 12, 2013, with you and members of your staff.

The inspectors examined activities conducted by your staff to complete commitments Entergy made as part of your application for a renewed facility operating license. The inspectors also reviewed selected procedures and records, observed activities, and interviewed personnel.

This inspection was conducted to follow-up on several commitments that were determined to merit additional inspection during a previous NRC License Renewal Team Inspection.

No findings were identified during this inspection. The NRC determined that the commitments reviewed associated with the license renewal application had been appropriately implemented.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs Agencywide Document Access and Management System (ADAMS). ADAMS is accessible from the NRC website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

James M. Trapp, Chief Engineering Branch 1 Division of Reactor Safety Docket No. 50-247 License No. DPR-26

ML13263A020 Non-Sensitive Publicly Available SUNSI Review Sensitive Non-Publicly Available OFFICE RI/DRS RI/DRP RI/DRS NAME GMeyer/MM ABurrit JTrapp DATE 9/18/13 9/19/13 9/19/13 J. Ventosa 2

Enclosure:

Inspection Report 05000247/2013010 w/

Attachment:

Supplementary Information cc w/encl: Distribution via ListServ

J. Ventosa 3 Distribution w/encl: (via E-mail)

W. Dean, RA (R1ORAMAIL RESOURCE)

D. Lew, DRA (R1ORAMAIL RESOURCE)

D. Roberts, DRP (R1DRPMAIL RESOURCE)

M. Scott, DRP (R1DRPMAIL RESOURCE)

R. Lorson, DRS (R1DRSMAIL RESOURCE)

J. Rogge, DRS (R1DRSMAIL RESOURCE)

A. Burritt, DRP T. Setzer, DRP S. McCarver, DRP L. McKown, DRP S. Stewart, DRP, SRI K. Dunham, DRP, RI Ami Patel, DRP, RI D. Hochmuth, DRP, AA D. Rich, RI OEDO RidsNrrPMIndianPoint Resource RidsNrrDorlLpl1-1 Resource ROPReports Resources

U.S. NUCLEAR REGULATORY COMMISSION REGION I Docket No.: 50-247 License No.: DPR-26 Report No.: 05000247/2013010 Applicant: Entergy Nuclear Northeast (Entergy)

Facility: Indian Point Energy Center Unit 2 Location: 450 Broadway Buchanan, NY 10511-0249 Dates: September 9-12, 2013 Inspectors: G. Meyer, Senior Reactor Inspector M. Modes, Senior Reactor Inspector Approved By: James M. Trapp, Chief Engineering Branch 1 Division of Reactor Safety i Enclosure

SUMMARY

OF FINDINGS IR 05000247/2013010; 09/09/2013 - 09/12/2013; Indian Point Nuclear Generating Unit 2; License Renewal Inspection.

This report covers an announced one week inspection, using the guidance provided in NRC inspection procedure Temporary Instruction 2516/001, Review of License Renewal Activities, of activities conducted by Entergy to complete commitments made to the NRC as a part of the Indian Point Energy Center, Unit 2, application for a renewed operating license. The commitments reviewed during this inspection are recorded in Supplement 1 to NUREG-1930, Safety Evaluation Report Related to the License Renewal of Indian Point Generating Units Numbers 2 and 3, Attachment 1, dated August 2011, and in other related correspondence.

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity No findings were identified. The inspectors concluded Entergy had made sufficient progress to complete our review of 44 commitments.

ii Enclosure

REPORT DETAILS

4. OTHER ACTIVITIES 4OA5 Review of License Renewal Activities (TI 2516/001)

.1 Background The expiration date of the operating license for Indian Point Unit 2 is midnight on September 28, 2013. Indian Point Unit 2 meets the criteria in Title 10 of the Code of Federal Regulations (10 CFR) 2.109(b), Effect of timely renewal application, and will likely operate beyond the current operating license expiration date. Due to the Commissions decision to revise the Waste Confidence Decision and Rule and because of the ongoing Atomic Safety and Licensing Board hearings, the Commission is not expected to issue a renewed license for Indian Point Unit 2 before the expiration date of the original license. Therefore, Indian Point would continue operations under the timely renewal provisions of 10 CFR 2.109(b).

The team used NRC Inspection Manual Temporary Instruction 2516/001 to conduct this inspection. The Temporary Instruction was written specifically for plants like Indian Point Unit 2, where the holders of an operating license meet the criteria of 10 CFR 2.109, for timely renewal, but a final decision by the NRC on the license renewal application is not expected prior to the period of extended operation. The inspection objectives and requirements of the Temporary Instruction are to report the status of license renewal commitment implementation, the status of aging management program implementation, and to verify the description of programs and activities for managing the effects of aging are consistent with the Updated Final Safety Analysis Report.

The NRC has conducted three separate license renewal inspections that have reviewed a total of 44 license renewal commitments. Our first license renewal inspection conducted during a refueling outage, reviewed four commitments as documented in NRC Inspection Report 05000247/2012008 (ML12110A315). On May 23, 2013, the NRC completed a License Renewal Commitment Team Inspection, as documented in NRC Inspection Report 05000247/2013009 (ML13186A179). The team inspection concluded that Entergy had made sufficient progress to complete review of 30 commitments and identified 11 commitments that merited further assessment during a planned follow-up inspection. During the current planned follow-up inspection, the inspectors completed our review of ten of the 11 commitments identified by the team as requiring additional review. Commitment 47, one of the 11 commitments previously identified for additional review, was not assessed during this inspection because Entergy revised the completion date for this commitment to March 1, 2015, in a letter to the NRC (ML13142A202).

Enclosure

2

.2 Commitment Reviews

.2.1 Commitment 6: Enhance the Fatigue Monitoring Program to monitor steady state cycles and feedwater cycles or perform an evaluation to determine monitoring is not required.

Review the number of allowed events and resolve discrepancies between reference documents and monitoring procedures.

a. Inspection Scope In the prior inspection, the inspectors noted that Entergy had awarded contracts to perform calculations to determine whether monitoring of steady state cycles and feedwater cycles was required. The inspectors also noted that Entergy planned to revise procedure 2-PT-2Y015, Thermal Cycle Monitoring Program, if the calculations demonstrated that a change in the number of allowable steady state cycles and feedwater cycles was identified.

During this inspection, the inspectors reviewed Westinghouse calculation IPP-13-20, Revision 1, dated August 14, 2013, reporting the Steady State Fluctuations and Feedwater Cycling Transient Disposition, presenting LTR-PAFM-13-87, Revision 2.

This result determined that both transients do not significantly affect the fatigue of the primary system and can be removed from the transient cycle counting program. It was determined the feedwater cycle transient must still be tracked for the secondary side of the steam generator and feedwater piping, with a 25,000 cycle limit. The inspectors verified that procedure 2-PT-2Y015, Thermal Cycle Monitoring Program, Revision 4, tracked the feedwater cycle transients.

Findings and Observations No findings were identified.

.2.2 Commitment 13: Enhance the Metal-Enclosed Bus (MEB) Inspection Program to add a 480 volt bus, visually inspect the external surface of MEB enclosure assemblies, include acceptance criteria, inspect bolted connections, and remove reference to re-torquing connections from the applicable site procedure.

a. Inspection Scope In the prior inspection, the inspectors identified that Entergy had not included all accessible portions of the MEB within the scope of the maintenance inspection program.

The inspectors noted that sections of the emergency diesel generator 480 volt MEB in the electrical tunnel had not been visually inspected and were not included in the scope of the maintenance procedure which performed the MEB inspections and tests. As a result of the NRCs observations, Entergy initiated a Condition Report (CR-IP2-2013-01786) to revise site procedures and conduct visual inspections of those additional sections of the bus ducts prior to the period of extended operations.

During this inspection the inspectors reviewed the revised inspection and test procedures for the portions of the MEB previously considered to be inaccessible. The Enclosure

3 inspectors reviewed the work orders under which the inspections and tests were completed for the three MEB sections and corrective action documents which resolved any identified conditions.

Findings and Observations No findings were identified.

.2.3 Commitment 19: Implement the One-Time Inspection Program as described in License Renewal Application, Section B.1.27. This new program will be implemented consistent with the corresponding program described in NUREG-1801, Section XIM32, One-Time Inspection.

a. Inspection Scope In the prior inspection, the inspectors noted the One-Time Inspection Program involved over 400 inspections. Because Entergy had only completed approximately half of the planned inspections, the inspectors determined that additional NRC review was merited to assess the remaining inspection results; any further actions needed, and program conclusions.

During this inspection the inspectors reviewed the One-Time Inspection Summary Report, the completed inspection tracking matrix, and 20 additional inspection reports.

The inspectors determined that all inspection results were acceptable and there was no need for additional action. Further, Entergy concluded that the inspection results demonstrated that the existing aging management programs for water chemistry and the diesel fuel monitoring and oil analysis had been effective in managing aging. The test program results also indicated that additional inspections on components specified in the License Renewal Application had not identified any unacceptable degradation.

Findings and Observations No findings were identified.

.2.4 Commitment 23: Implement the Selective Leaching Program as described in License Renewal Application, Section B.1.33. This new program will be implemented consistent with the corresponding program described in NUREG-1801, Section XIM33 Selective Leaching of Materials.

a. Inspection Scope In the prior inspection, the inspectors noted Entergy was in the progress of completing selective leaching inspections. The inspectors determined that additional NRC inspection was merited to review the results of the remaining inspections, any further actions needed, and program conclusions.

During this inspection, the inspectors reviewed Selective Leaching Summary Report, including the final sampling plan, the destructive evaluation results of seven Enclosure

4 components, and associated corrective action documents. Entergy determined that the visual inspections of 22 gray cast iron components and 17 copper alloy components showed an absence of selective leaching. However, destructive evaluations demonstrated that significant selective leaching (i.e., graphitization, had occurred in gray cast iron components) was occurring. While the commitment actions were complete, Entergy concluded that they would refine the process and documented a corrective action to perform engineering evaluations to develop a continuing monitoring program to manage and evaluate selective leaching.

Findings and Observations No findings were identified

.2.5 Commitment 26: Implement the Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program as described in License Renewal Application, Section B.1.37. This new program will be implemented consistent with the corresponding program described in NUREG 1801,Section XI.M12, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Program.

a. Inspection Scope In the prior inspection, the inspectors noted that Entergy was planning to screen all cast austenitic stainless steel components to determine which were potentially susceptible to a loss of fracture toughness. These components were to be further evaluated, using a refined analytical technique, to determine the components susceptibility to reduction in fracture toughness. Entergy chose to perform a unique fracture mechanics analysis of these components, which was in progress at the time of the prior inspection. The inspector noted that there may be a question regarding the submittal of the analysis under ASME requirements, as stipulated in 10 CFR 50.55a.

Subsequently the inspector, in consultation with the Division of License Renewal, determined submittal of the analysis was not required. During this inspection the inspectors reviewed the completed evaluation of the screened components susceptibility to reduction in fracture toughness: Report 1300066.403, Revision 0, Aging Management of CASS Piping at Indian Point 2, Flaw Tolerance Evaluation of CASS Piping at Indian Point, dated August 2013. The methodology of probabilistic fracture mechanics determined a postulated starting reference flaw of one-quarter thickness would remain below the maximum allowable flaw size during a 60 year life for all components.

Findings and Observations No findings were identified.

.2.6 Commitment 27: Implement the Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program as described in License Renewal Application, Section B.1.38. This new program will be implemented consistent with the Enclosure

5 corresponding program described in NUREG-1801,Section XI.M13, Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel program.

a. Inspection Scope In the prior inspection, the inspectors noted a site document had not been developed that defined and implemented the screening criteria of Electric Power Research Institute (EPRI) Technical Report 1013234, Materials Reliability Program: Screening Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design, (MRP-191), listed in Table 3-5, as applied to the Indian Point components listed in Table 5-1 of EPRI Report 1022863, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, (MRP-227-A).

During this inspection the inspectors reviewed Entergy Nuclear Engineering Report:

Indian Point Energy Center: Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel Aging Management Program, IP-RPT-13-00049, Revision 0, dated August 14, 2013. This report implemented the above screening criteria and ranking methodology for Indian Point.

Findings and Observations No findings were identified.

.2.7 Commitment 33: For the locations identified in License Renewal Application, Table 4.3-13 (IP2), update the fatigue usage calculations using refined fatigue analyses to determine if Cumulative Usage Factors (CUF) remain less than 1.0 when accounting for the effects of reactor water environment, using valid Fen factors.

a. Inspection Scope In the prior inspection, the inspectors noted that calculations documented in calculation WCAP 17149-P, Evaluation of Pressurizer Insurge/Outsurge Transients for Indian Point Unit 2, and WCAP 17199-P, Environmental Fatigue Evaluation for Indian Point Unit 2, concluded that cumulative fatigue usage factors, including reactor water environment effects, were below the American Society of Mechanical Engineers (ASME) Code allowable value of 1.0 for transients postulated for 60 years of operation. The inspectors noted Entergys action plan included revising the Thermal Cycle Monitoring Program procedure to reflect changes in the number of projected cycles used in WCAP 17199-P.

During this inspection the inspectors reviewed Indian Point Programs and Components Engineering Procedure 2-PT-2Y015, Revision 4, Thermal Cycle Monitoring Program.

The inspectors noted the procedure was updated to reflect the number of allowable cycles derived from the above analysis. The procedure referenced WCAP 17199-P and included a Table (Attachment 1) that included actual plant cycles rather than the number of cycles used in the original design calculations. The procedure also was revised to better reflect operational assumptions used and the bases for the revised calculations of the WCAP.

Enclosure

6 Findings and Observations No findings were identified.

.2.8 Commitment 40: Evaluate plant specific and appropriate industry operating experience and incorporate lessons learned in establishing appropriate monitoring and inspection frequencies to assess aging effects for the new aging management programs.

Documentation of the operating experience evaluated for each new program will be available on site for NRC review prior to the period of extended operation.

a. Inspection Scope During the prior inspection, there was insufficient material to review this commitment.

Subsequently, Entergy completed an operational review for the following new aging management programs embodied in separate commitments:

3, Buried Piping and Tanks Inspection Program

14, Non-EQ Bolted Cable Connections Program

15, Non-EQ Inaccessible Medium-Voltage

16, Non-EQ Instrumentation circuits Test Review Program

17, Non-EQ Insulated Cables and Connections Program

19, One-Time Inspection Program

20, One-Time Small Bore Piping Program

23, Selective Leaching Program

26, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Program

27, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program.

During this inspection the inspectors reviewed a selected sample of the reviews and noted each of the programs was subject to an operational review which included Indian Point specific experience. For example, the Buried Piping Program for Unit 2 was assessed for a period of six years, 2007 through 2013. This period of time enveloped the operational experience included in the guidance documents, such as Generic Aging Lessons Learn (NUREG-1801), Revision 2. External operational experience considered included license event reports, NRC generic letters, NRC information notices, and Institute of Nuclear Power Operations (INPO) documents.

Findings and Observations No findings were identified.

.2.9 Commitment 43: Indian Point Energy Center will review design basis ASME Code Class 1 fatigue evaluations to determine whether the NUREG/CR-6260 locations that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting locations for the Unit 2 and Unit 3 configurations. If more limiting locations are identified, the most limiting location will be evaluated for the effects of the Enclosure

7 reactor coolant environment on fatigue usage. Indian Point Energy Center will use the NUREG/CR-6909 methodology in the evaluation of the limiting locations consisting of nickel alloy, if any.

a. Inspection Scope In the prior inspection, the inspectors noted that Entergy had awarded contracts to perform calculations to support closure of this commitment. Because the results of the calculations were not available at the time, the inspectors deferred inspection of this commitment.

During this inspection the inspectors reviewed calculation CN-PAFM-13-32, Revision 0, Indian Point Unit 2 (IP2) and Unit 3 (IP3) Refined EAF Analyses and EAF Screening Evaluations. This calculation was the evaluation of locations previously screened by calculation CN-PAFM-12-35, Revision 1, Indian Point Unit 2 and Unit 3 EAF Screening Evaluations, that could be more limiting than the locations identified in NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plants Components. The inspectors noted the CUFen result for the pressurizer nozzle was 0.999 at 60 years. Entergy was aware that accumulation of cycles at a rate greater than assumed in the calculation would require a more refined analysis, application of a non-destructive monitoring technique, or replacement of the pressurizer nozzle.

Findings and Observations No findings were identified.

.2.10 Commitment 48: Entergy will visually inspect in-scope underground piping prior to the period of extended operation and then on a frequency of at least once every 2 years during the period of extended operation. Visual inspections will be supplemented with surface or volumetric non-destructive testing if indications of significant loss of material are observed. Adverse indications will be entered into the plant corrective action program for evaluation of extent of condition and for determination of appropriate corrective actions (e.g., increased inspection frequency, repair, or replacement).

a. Inspection Scope In the prior inspection the inspectors noted that work orders had been issued to perform the inspections prior to the period of extended operation. The inspectors determined that additional inspection was merited regarding review of the results of the underground piping inspections.

The inspectors reviewed the results of selected inspections performed subsequent to the May 23, 2013, NRC inspection. EN-EP-S-002-MULTI, Attachment 7.2 Pipe/Tank Coating Visual Inspection Checklist was reviewed for the following lines:

Enclosure

8 21-EDGE-2/EDG FOST 3-inch equalizing line 21-EDGE-2/EDG FOST 4-inch fill line 21-EDGE-2/EDG FOST 4-inch vent line 22-EDGE-2/EDG FOST 3-inch equalizing line 22-EDGE-2/EDG FOST 4-inch fill line 22-EDGE-2/EDG FOST 4-inch vent line 23-EDGE-2/EDG FOST 3-inch equalizing line 23-EDGE-2/EDG FOST 4-inch fill line 23-EDGE-2/EDG FOST 4-inch vent line The check list included the inspection of the piping for aging affects such as mechanical damage, coating breaks (referred to as a holiday), and blistering.

Findings and Observations No findings were identified.

4OA6 Meetings, Including Exit On September 12, 2013, the inspectors presented the inspection results to Mr. John Ventosa, Site Vice President, and other members of the Entergy staff. The inspectors verified that no proprietary information was retained by the inspectors or documented in this report.

ATTACHMENT: SUPPLEMENTARY INFORMATION Enclosure

A-1 SUPPLEMENTARY INFORMATION KEY POINTS OF CONTACT Entergy Personnel J. Ventosa, Site Vice President N. Azevedo, Code Programs Supervisor C. Caputo, License Renewal Team J. Curry, Senior Project Manager G. Dahl, Licensing Engineer P. Guglielmino, Implementation Team Manager L. Lubrano, Component Electrical Engineer R. Sporbert, One-Time Inspection Coordinator LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED Closed The inspectors determined that the following 10 commitments had been appropriately implemented:

6 Fatigue cycles analysis 13 Metal enclosed bus inspection 19 One-time inspection 23 Selective leaching inspection 26 Embrittlement of CASS analysis 27 Embrittlement of CASS analysis 33 Fatigue monitoring 40 Operating experience for new programs 43 Fatigue monitoring analysis 48 Underground piping inspection LIST OF ACRONYMS ADAMS Agencywide Documents Access Management System ASME American Society of Mechanical Engineers CASS Cast Austenitic Stainless Steel CFR Code of Federal Regulations CUF Cumulative Usage Factor ENTERGY Entergy Nuclear Northeast EPRI Electric Power Research Institute INPO Institute of Nuclear Power Operations IPEC Indian Point Energy Center LRA License Renewal Application MEB Metal-Enclosed Bus Attachment

A-2 MRP Materials Reliability Project NRC Nuclear Regulatory Commission UFSAR Updated Final Safety Evaluation Report LIST OF DOCUMENTS REVIEWED Commitment 6, 33, and 43 2-PT-2Y015, Revision 3, Thermal Cycle Monitoring Program IP-RPT-11-LRD13, Revision 0, Review of the Fatigue Monitoring Aging Management Program for License Renewal Implementation EN-LI-100, Revision 13, Process Applicability Determination EN-AD-101, Revision 16, Procedure Process WCAP-17199-P, July 2010, Environmental Fatigue Evaluation for Indian Point Unit 2 WCAP-17149-P, July 2010, Evaluation of Pressurizer Insurge/Outsurge Transients for Indian Point Unit 2 Entergy Letter dated August 9, 2010 (NL-10-82), License Renewal Application - Completion of Commitment #33 Regarding the Fatigue Monitoring Program Indian Point Nuclear Generating Unit Nos. 2 and 3 [ML102300504]

Commitment 13 2-ELC-016-BUS, Inspection, Cleaning and Testing of 480V Buses, Revision 2 CR-IP2-2009-03029, Water dripping at the bend in the electric tunnel CR-IP2-2012-01903, Bus 5A surface rust noted on the interior divider panel Work Order 52293872, Inspection of Bus 5A Switchgear and Station Service Transformer Work Order 52294517, Inspection of Bus 5A (480 V Switchgear to EDG) 2-ELC-016-BUS, Revision 4, Inspection, Cleaning and Testing of 480V Buses 2-ELC-403-BUS, Revision 7, Inspection and Cleaning of 480 Volt Bus Duct Work Order 351381, 21 EDG Bus Visual, Cleaning, Bolted Checks in Electrical Tunnel, completed on July 18, 2013 Work Order 351382, 22 EDG Bus Visual, Cleaning, Bolted Checks in Electrical Tunnel, completed on June 12, 2013 Work Order 351448, 23 EDG Bus Visual, Cleaning, Bolted Checks in Electrical Tunnel, completed on August 12, 2013 CR-IP2-2013-01738, MEB acceptance criteria CR-IP2-2013-01748, Leaks in Unit 2 electrical tunnel CR-IP2-2013-02375, 22 EDG bus inspection CR-IP2-2013-02923, 21 EDG bus inspection CR-IP2-2013-03330, 23 EDG bus inspection CR-IP2-2013-02912, Thermography procedures CR-IP2-2013-02913, Water intrusion into Unit 2 electrical tunnel Attachment

A-3 Commitment 19 IP-RPT-11-LRD28, Revision 0, Review of the One-Time Inspection Program EN-FAP-LR-024, Revision 1, One-Time Inspection NL-13-046, Amendment 13 to LRA for One-Time Inspection and Selective Leaching Programs, March 18, 2013 IPEC Unit 2 One-Time Inspection Tracking Matrix, May 3, 2013 and May 21, 2013 20 Inspection reports for one-time inspections IP-RPT-13-LRD03, Revision 0, Unit 2 License Renewal One-Time Inspection Summary Report IPEC Unit 2 One-Time Inspection Tracking Matrix, August 28, 2013 20 Additional inspection reports for one-time inspections Commitment 23 IP-RPT-11-LRD34, Revision 0, Review of the Selective Leaching Program EN-FAP-LR-02, Revision 3, Selective Leaching Inspection NL-13-046, Amendment 13 to LRA for One-Time Inspection and Selective Leaching Programs, March 18, 2013 IPEC Unit 2 Selective Leaching Inspection Tracking Matrix, May 20, 2013 10 Inspection reports for copper-alloy selective leaching inspections 12 Inspection reports for gray cast iron selective leaching inspections WO 00326036-01 WO 00326216-01 IP-RPT-13-LRD07, Revision 0, License Renewal Selective Leaching Inspection Summary Report Altran 13-0313-TR-001, Laboratory Analysis of Several Valves for Selective Leaching, Revision 0 Altran 13-0313-TR-001, Laboratory Analysis of Several Valves for Selective Leaching, Revision 1 CR-IP2-2013-03037, Selective leaching of gray cast iron components CR-IP2-2013-03360, Selective leaching of copper alloy components Commitment 26 Entergy Letter, NL-09-018, Reply to Request for Additional Information - Miscellaneous Items, January 27, 2009 Entergy Letter NL-11-101, Clarification for Additional Information (RAI) Aging Management Programs, August 22, 2011 LR# 173, LR Request, Confirm each AMP will be implemented with ten elements IP-RPT-11-LRD38, Review the Thermal Aging Embrittlement of CASS Aging Management Program for License Renewal Implementation, 1/2/2013 NRC Letter, License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components, May 19, 2000 WCAP-10977, Supplement 1, Additional Information in Support of the Technical Justification for Eliminating Large Primary LOOP Pipe Rupture as the Structural Design Basis for Indian Point Unit 2, January 1989 Attachment

A-4 Commitment 27 IP-RPT-11-LRD39, Revision), Review of the Thermal Aging & Neutron Embrittlement of CASS Aging Management Program for License Renewal Implementation, ED41109, 1/23/2013 EPRI 1013234 Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191), November 2006 EPRI 1022863 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), December 2011 NRC Letter Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Nos. 2 and 3, License Renewal Application, May 15, 2012 Entergy Letter, NL-11-101, Clarification for Request for Additional Information (RAI) Aging Management Programs, 8/22/2011 Entergy Letter NL-13-052, Reply to Request for Additional Information Regarding the License Renewal Application, 5/7/2011 LR Request #173 Confirm new programs will be implemented consistent with 10 elements of NUREG-1801.

IP-RPT-11-LRD39 Review of the Thermal Aging and Neutron Irradiation Embrittlement of CASS Aging Management Program for License Renewal Implementation, EC41109, 1/2/13 Commitment 48 Work Order PMRQ 00349816-01, 23 Fuel Oil Storage Tank Underground Piping Inspections Work Order PMRQ 00349802-01, 21 Fuel Oil Storage Tank Underground Piping Inspections Work Order PMRQ 00349814-01, 22 Fuel Oil Storage Tank Underground Piping Inspections Work Order 00342492-01, Inspect Underground Piping by the IP2 EDG Building DF-2 Area Work Order 00342493-01, Inspect Underground Piping by the IP2 EDG Building DF-2-1 Area Work Order 00342494-01, Inspect Underground Piping by the IP2 EDG Building DF-2-2 Area Attachment

Attachment 4 Summary of the February 19, 2015 public meeting in Rockville, Maryland to discuss reactor pressure vessel issues including a list of attendees April 6, 2015 (ML15096A128)

April 6, 2015 MEMORANDUM TO: John W. Lubinski Division of Engineering Office of Nuclear Reactor Regulation FROM: Robert O. Hardies, Sr. Level Advisor /RA by E-mail//

Division of Engineering Office of Nuclear Reactor Regulation

SUBJECT:

SUMMARY

OF FEBRUARY 19, 2015, PUBLIC MEETING TO DISCUSS REACTOR PRESSURE VESSEL ISSUES On February 19, 2015, a Category 2 public meeting was held between the U.S. Nuclear Regulatory Commission (NRC) staff and representatives of industry to allow an exchange of information about reactor pressure vessel integrity issues. A portion of the meeting related to the staffs recent evaluation of potential non-conservatisms in NRC Branch Technical Position (BTP) 5-3, Fracture Toughness Requirements. Other topics addressed included reactor vessel surveillance programs, Title 10, Code of Federal Regulations, Part 50, (10 CFR 50) Appendices G and H evaluations, status of the NRC Regulatory Guide for Alternate Pressurized Thermal Shock (PTS) Rule Implementation, a discussion of ASME Codes requirements for pressure testing while the reactor vessel is critical, and a status of NRC work activities on the Reactor Embrittlement Archive Project (REAP) Database.

The NRC staff made a two-part presentation on the recent NRC activities performed regarding the potential non-conservatism of BTP 5-3. The first part of the presentation provided a review of definitions and estimates of unirradiated reference temperature (RTNDT) and unirradiated upper shelf energy (USE) addressed in BTP 5-3, summarized the background of the recent questions concerning the potential non-conservatism of BTP 5-3, and summarized the objectives of the staffs analysis of BTP 5-3. The second part of the staffs presentation provided an assessment of the impact of the potential non-conservatisms in BTP 5-3 on U.S. plants with regard to PTS and pressure temperature (P-T) limit evaluations based on recent docketed information. The staff concluded that one PWR PTS evaluation is potentially affected, some PWRs and BWRs P-T limits are potentially affected, and that there are no immediate safety concerns. The staff provided a tentative schedule for completion of a report that documents the staffs findings.

The industry followed with three presentations on their recent activities performed regarding the potential non-conservatism of BTP 5-3. The first presentation provided background on the industry focus groups that have been formed to address the BTP 5-3 issues, the membership of the focus groups, and the focus groups activities that are currently underway. The second presentation provided a summary of the Materials Reliability Project (MRP)/Boiling Water Reactor Vessel and Internals Project (BWRVIP) focus group activities and results to-date.

The objectives of the focus group activities include conducting a survey regarding use of BTP 5-3 in the PWR and BWR fleets, evaluating the BTP 5-3 procedures which had previously been identified as potentially non-conservative, determining if application of BTP 5-3 for defining reactor vessel P-T limits provides adequate margins against failure through the 60-year license period (EOLE), and, if needed, recommending alternative procedures to ensure that adequate margins against failure are maintained through EOLE. The focus group

J. Lubinski has completed a draft report that is under review with a target completion in June 2015. The focus group is considering similar work for a procedure developed by GE-Hitachi Nuclear Energy that is similar to BTP 5-3. The third presentation provided a summary of the Pressurized Water Reactor Owners Group (PWROG) activities regarding their Material Orientation Toughness Assessment (MOTA) for the purpose of mitigating BTP 5-3 uncertainties. The objective of the MOTA is to explore existing deterministic margin that may be potentially available in ASME Code Section XI, Appendix G and other NRC-approved sources to address potential non-conservatisms in BTP 5-3. The results of the industrys investigation to-date demonstrate that current methods for developing P-T limits are acceptable in light of the identified BTP 5-3 estimation uncertainties. The PWROG intends to document the MOTA in a final report later this year.

The NRC staff provided two presentations on the status of their 10 CFR 50 Appendix G research activities to-date. The first presentation provided a summary of overall activities to-date, and summarized the new version of the FAVOR computer code to address software errors previously identified by the industry. A release of FAVOR v15.1 that remedies the software errors is anticipated in Spring 2015. The second presentation summarized the staffs efforts regarding an advanced residual stress model under investigation for implementation into FAVOR. Further efforts are underway to explore other fracture mechanics models and to assess the ability to adopt an appropriate model into the FAVOR code at a later date.

The PWROG discussed an assessment of the margins associated with American Society of Mechanical Engineers (ASME)Section XI Appendix G pressure-temperature (P-T) limits for pressurized water reactor (PWR) nozzles. The assessment is intended to generically address 10 CFR 50 Appendix G requirements recently clarified by the NRC in Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components. This RIS clarifies that 10 CFR 50 Appendix G requires that P-T limits sufficiently address all ferritic materials of the reactor vessel, including the impact of structural discontinuities such as nozzles. The purpose of the industrys activity is to develop a basis for generically addressing nozzles supporting P-T limit submittals and to justify the use of the reactor vessel shell region with the highest embrittlement as the limiting region to be used for P-T limits.

The NRC staff presented their plans to begin rulemaking for revising 10 CFR 50 Appendix H based on recent commission approval to do so. The planned revisions will include changes to permit the use of updated and modern revisions of ASTM standards and, in response to industry requests, to allow an increase in the required time to submit surveillance program test results. The longer reporting time is intended to mitigate scheduler complications that modern integrated surveillance programs have experienced related to the logistics of withdrawal, shipping, testing and reporting results of testing of surveillance capsules.

The NRC staff presented a status and the latest tentative schedule for issuing a draft Regulatory Guide (RG) describing guidance for implementation of the Alternate Pressurized Thermal Shock Rule, 10 CFR 50.61a. The RG was published in the Federal Register for a 60-day public comment period on March 13, 2015.

The NRC staff presented an ASME Code item related to pressure testing with the reactor vessel core critical. Although NRC regulations in 10 CFR 50 specifically prevent pressure

J. Lubinski testing of the pressure vessel while the core is critical, ASME Section XI repair/replacement activities allow pressure testing of other non-RPV Class 1 components while the reactor vessel core is critical. The staff is considering additional actions to clarify the NRC position on this item.

The NRC staff discussed the Reactor Embrittlement Archive Project. This project catalogues historical records of surveillance data and research irradiation information.

Action items captured during the meeting were as follows:

1. NRC and industry will explore opportunities to share longitudinal and transverse Charpy data sources to facilitate data verification.

2. The NRC will evaluate the need to request and review the EPRI sponsored evaluation of BTP 5-3 conservatism.

3. The NRC will evaluate the need to request and review the PWROG report on MOTA.

4. The NRC and industry will evaluate further communication opportunities at ASME Code meetings.

A list of attendees is enclosed. The slide presentations presented by the NRC staff and the industry representatives can be found in the Agencywide Documents Access and Management System (ADAMS) at Accession Number ML15061A072.

Enclosure:

List of Attendees

Package: ML15061A072 Meeting Summary: ML15096A128

By E-mail OFFICE NRR/DE

NAME RHardies DATE 04/06/2015 List of Attendees Public Meeting with U.S. Nuclear Regulatory Commission (NRC) Staff to Discuss Reactor Pressure Vessel Issues February 19, 2015

List of Attendees (concluded)

Public Meeting with U.S. Nuclear Regulatory Commission (NRC) Staff to Discuss Reactor Pressure Vessel Issues February 19, 2015 ENCLOSURE

Attachment 5 E-mail from Hearingdocket@nrc.gov to Brian Lusignan Notification of the Non-public Filing of the Joint Brief of Entergy and Westinghouse Re Proprietary Documents June 4, 2015

From: Hearingdocket@nrc.gov To: pbessette@morganlewis.com; wglew@entergy.com; brian.harris@nrc.gov; hearingdocket@nrc.gov; rkuyler@morganlewis.com; Lisa S. Kwong; Brian Lusignan; lawrence.mcdade@nrc.gov; OCAAMAIL@nrc.gov; martin.oneill@morganlewis.com; drepka@winston.com; John J. Sipos; ksutton@morganlewis.com; sherwin.turk@nrc.gov; richard.wardwell@nrc.gov; alana.wase@nrc.gov; edward.williamson@nrc.gov

Subject:

Re: NRC Proceeding "Indian Point 50-247-LR and 50-286-LR" Date: Thursday, June 04, 2015 3:52:15 PM MESSAGE FROM THE OFFICE OF THE SECRETARY, NUCLEAR REGULATORY COMMISSION REGARDING LIMITED ACCESS PROTECTED DOCUMENT Re: NRC Proceeding "Indian Point 50-247-LR and 50-286-LR" The Office of the Secretary has received a limited access protected document entitled "Joint Brief of Entergy and Westinghouse Re Proprietary Documents" submitted by David A Repka who is affiliated with Winston and Strawn LLP.

It is intended for inclusion in the referenced proceeding and was submitted through the NRC Electronic Information Exchange (EIE) system. It arrived on 06/04/15 at 15:52 EDT.

The submitter has designated you as a hearing participant who is eligible to view this/these document(s). You are entitled to view and/or retrieve/save this/these document(s) by visiting the following web link(s):

Joint Brief of Entergy and Westinghouse re Proprietary Documents -

https://eieprod.nrc.gov/EIE25L3/downloadAttachment.do?submissionID=50749&docID=1172 (1998 KB)

This/these document(s) is/are subject to special handling and limited distribution pursuant to an adjudicatory directive. The document(s) will remain available to you through the link(s) above for 14 days after which it/they will be removed from the E-Filing system. After 3 business days, you may also access the document(s), along with any other protected document(s) to which you have been authorized access, through the "Access Authorized Protected Documents" selection on the following page: https://eieprod.nrc.gov/EIE25/portal.do or by going directly to the Electronic Hearing Docket at: https://adams.nrc.gov/ehd Receipt of this message constitutes completion of service of this filing.

Attachment 6 E-mail from Brian Lusignan to attorneys for Westinghouse, Entergy and NRC Staff regarding the apparent failure of Westinghouse and Entergy to properly file the Joint Brief publicly June 15, 2015

From: Brian Lusignan To: "drepka@winston.com"; "Sutton, Kathryn M."; "Bessette, Paul M."

Cc: John J. Sipos; Lisa S. Kwong; Turk, Sherwin

Subject:

Nonpublic Filing of Joint Brief Date: Monday, June 15, 2015 11:48:39 AM Counsel:

The State is in the process of preparing its response to the joint brief of Westinghouse and Entergy regarding the proprietary designation of certain documents in the Indian Point relicensing proceeding. Recently, we noticed that although the joint brief has not been designated as containing proprietary information pursuant to 10 C.F.R. section 2.390 (b) and the ASLB's September 4, 2009 Protective Order, paragraphs A and K, it was served via the NRC's non-public electronic information exchange (EIE) on a limited number of hearing participants. The Certificate of Service that accompanied the joint brief does not reference the non-public EIE exchange; instead, the Certificate of Service states that the brief was served on those on the EIE Service List for the captioned proceeding, indicating that the joint brief should have been served on all parties and interested governmental entities in the proceeding. The State respectfully requests that you re-file the joint brief on the public EIE exchange, so that all hearing participants may access it.

Very truly yours, Brian Lusignan Assistant Attorney General NYS Office of the Attorney General Environmental Protection Bureau The Capitol Albany NY 12224-0341 (518) 776-2399 Please note the new phone number.

ML15169B121

Text

Background

SUBJECT:

Dear Mr. Ventosa:

Enclosure:

Attachment:

SUMMARY

SUBJECT:

SUMMARY

Enclosure:

Subject:

Subject:

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