Text

Inservice Inspection Program Fifth Ten-Year Interval 10 CFR 50.55a Relief Requests
	ML17241A048
Person / Time
Site:	Cooper
Issue date:	08/17/2017
From:	Dent J; Nebraska Public Power District (NPPD)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NLS2017071
	Download: ML17241A048 (132)
	v • d • e

{{#Wiki_filter:H Nebraska Public Power District Alwa)S there when )OU need us 10 CFR 50.55a NLS2017071 August 17, 2017 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

10 CFR 50.55a Relief Requests for Fifth Ten-Year Inservice Inspection Interval Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District (NPPD) to request that the Nuclear Regulatory Commission (NRC) grant relief from, and authorize alternatives to, certain inservice inspection code requirements for the Cooper Nuclear Station (CNS) pursuant to 10 CFR 50.55a. The 10 CFR 50.55a requests pertain to inservice examination test requirements in Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Relief Request RI5-02, Revision 1, is included which provides an updated inspection history to include the Fall 2016 refueling outage. Updates are designated by change bars in the right-hand column. The applicable ASME Code for the fifth ten-year interval is the 2007 Edition, 2008 Addenda. These requests are applicable to the fifth ten-year inservice inspection interval, which began on April 1, 2016. In order to support planning for Refueling Outage 30, NPPD requests approval of these requests by August 1, 2018. The attachment contains the relief requests for the fifth ten-year interval. Enclosures 1 and 2 are included to support Relief Requests RR5-02 and RR5-03 . Formal licensee commitments are being made in this submittal and are included immediately following this cover letter. Should you have any questions concerning this matter, please contact Jim Shaw, Licensing Manager, at (402) 825-2788. Sincerely, KH v'\v1 0, L / ,., (Tu khO+.i J- ~ John Dent, Jr. Vice President - Nuclear and Chief Nuclear Officer /dv COOPER NUCLEAR STATION PO Box 98 / Brownville, NE 68327-0098 Telephone: (402) 825-3877 / Fax: (402) 825-5277 www.nppd .com

NLS2017071 Page 2 of2

Attachment:

Cooper Nuclear Station Inservice Inspection Program Fifth Ten-Year Interval 10 CFR 50.55a Relief Requests

Enclosures:

1. RR5-02, Cooper Nuclear Station; Structural Integrity Associates, Inc.

2. RR5-03, Cooper Nuclear Station; Structural Integrity Associates, Inc.

cc: Regional Administrator w/ attachment and enclosures USNRC - Region IV Senior Project Manager w/ attachment and enclosures USNRC - NRR Plant Licensing Branch IV Senior Resident Inspector w/ attachment and enclosures USNRC-CNS NPG Distribution w/o attachment or enclosures CNS Records w/ attachment and enclosures

 .T.
  ~Entergy NUCLEAR MANAGEMENT QUALITY RELATED        O-EN-Ll-106            REV. 13C3 MANUAL                   INFORMATIONAL USE                PAGE 28 OF 46 Regulatory Correspondence ATTACHMENT    9.3                                                   LICENSEE - IDENTIFIED COMMITMENTS TABLE Sheet 1of1 This table identifies actions discussed in this letter which Nebraska Public Power District commits to perform . Any other actions discussed in this submittal are described for the NRC's information and are not regulatory commitments.

TYPE (Check one) SCHEDULED ONE-TIME CONTINUING COMPLETION DATE COMMITMENT/COMMITMENT NO. ACTION COMPLIANCE Commitment# - NLS2017071-01 x Within 60 days of For a leaking flaw, the allowable leakage rate NRC approval of will be determined by dividing the critical Relief Request RR5-leakage rate by a safety factor of four (4). The 03 critical leakage rate is determ ined as the maximum leakage rate that can be tolerated . The critical leakage rate may be based on the allowable loss of inventory or the maximum that can be tolerated relative to room flooding , among others. Commitment# - NLS2017071-02 x Within 60 days of The number of augmented examination NRC approval of locations shall be increased from 5 (current Relief Request RR5-number in Code Case N-513-4) to 10. 03 Commitment# - NLS2017071-03 x Within 60 days of For leakage rates greater than 5 gpm , the NRC approval of leakage shall be stopped throughout the Relief Request RR5-temporary acceptance period by the use of 03 engineered mechanical clamping designed by CNS . The engineered mechanical clamping design shall be sufficient to withstand the maximum operating pressure and removable such that frequent periodic inspections defined in paragraph 2(e) of N-513-4 may be performed .

NLS2017071 Attachment Page 1of68 Attachment Cooper Nuclear Station Inservice Inspection Program Fifth Ten-Year Interval 10 CFR 50.SSa Relief Requests Page # Acronyms 2 RI5-01 Implementation of BWRVIP-05 (GL 98-05) 3 RI5-02, Rev. 1 Implementation ofBWRVIP Documents in Lieu ofB-N-1 and B-N-2 9 RI5-03 Implementation of Code Case N-702 47 RR5-02 Use Code Case N-513-4 58 RR5-03 Use Code Case N-513-4 at a Higher System Operating Pressure 63

NLS201707 1 Attachment Page 2of68 Acronyms ADAMS Agencywide Documents Access and NRC Nuclear Regulatory Commission Management System AHC Access Hole Cover NRI No Recordable Indications ART Adjusted Reference Temperature NV Nozzle-to-Vessel ASME American Society of Mechanical PEO Period of Extended Operation Engineers BWR Boiling Water Reactor PoF Probability of Failure BWRVIP Boiling Water Reactor Vessel PSIG Pounds per Square Inch Internals Project CE Combustion Engineering PT Liquid Penetrant Examination CF Chemistry Factor Pff Pressure and Temperature CFR Code of Federal Regulations PTLR Pressure and Temperature Limits Report CNS Cooper Nuclear Station OE Operating Experience CRDRL Control Rod Drive Return Line RAMA Radiation Analysis Modeling Application cs Core Spray RE xx Refuel Outage EFPY Effective Full Power Year RG Re!!Ulatory Guide EPRI Electric Power Research Institute RHR Residual Heat Removal EVT-1 Enhanced Visual (VT-1) Testing RHRSWB Residual Heat Removal Service Water Booster FEM Finite Element Method RIC SIL Rapid Information Communication Service Information Letter FF Fluence Factor RPV Reactor Pressure Vessel FW Feed water RTNDT Reference Temperature, Nil Ductility Temperature GE SIL Generic Electric Service Information sec Stress Corrosion Cracking Letter GL Generic Letter SER Safety Evaluation Report GPM Gallons Per Minute SIL Service Information Letter HAZ Heat Affected Zone SLC Standby Liquid Control ID Inner Diameter SRM Source Range Monitor IEB Inspection and Enforcement Bulletin SW Service Water IGSCC Intergranular Stress Corrosion TG Top Guide Cracking IR Inner Radius TLAA Time Limited Aging Analyses IRM Intermediate Range Monitor USNRC United States Nuclear Regulatory Commission ISI Inservice Inspection USAR Updated Safety Analysis Report JP Jet Pump UT Ultrasonic Testing LRA License Renewal Application VT Visual Inspection LTOP Low Temperature Overpressure VT-1 Detailed Inspection MVT-1 Modified Visual Inspection VT-2 Leakage Inspection NPPD Nebraska Public Power District VT-3 General Condition Inspection

NLS2017071 Attachment Page 3 of68 10 CFR 50.55a Request No. RI5-01 Implementation of BWRVIP-05 (GL 98-05) Proposed Alternative in Accordance with 10 CFR 50.55a(z)(l) Acceptable Level of Quality and Safety ASME Code Component(s) Affected Code Class: ASME Section XI Code Class l Component Numbers: RPV Circumferential Shell Welds (VCB-BB-1 , VCB-BA-2, VCB-BB-3, VCB-BB-4) Code

References:

ASME Section XI, 2007 Edition with 2008 Addenda Examination Category: B-A Item Number(s): Bl.11 Unit/Inspection Interval: Cooper/Fifth 10-year interval April 1, 2016 - February 28, 2026

Applicable Code Edition and Addenda

ASME Section XI, 2007 Edition through the 2008 Addenda Applicable ASME Code Requirements Table IWB-2500-1 , Examination Category B-A, Item No. B 1.11 , requires a volumetric examination of the circumferential shell welds each interval.

Reason for Request

NPPD is requesting an alternative in accordance with 10 CFR 50.55a(z)(l) on the basis that this alternative provides an acceptable level of quality and safety. This request for alternative would provide relief from circumferential weld examinations required by the ASME Section XI Code for the extended period of operation. CNS was previously granted this relief for the remainder of the original 40-year license term (Reference 6). During the staffs review of the CNS LRA (Reference 1), the staff concluded that CNS had demonstrated, in accordance with 10 CFR 54.21(c)(l)(ii), that for RPV circumferential weld examination relief, the analysis had been projected to the end of the period of extended operation. The staff also concluded that the USAR Supplement contained an appropriate summary description of the TLAA evaluation in accordance with 10 CFR 54.21 (d) and therefore, was acceptable (Reference 2).

NLS2017071 Attachment Page 4of68 10 CFR 50.55a Request No. RI5-01 (continued) Implementation of BWRVIP-05 (GL 98-05) Proposed Alternative and Basis for Use Proposed Alternative CNS requests the use ofBWRVIP-05 with supporting information described herein as the bases for excluding the RPV shell circumferential welds from the examinations required by ASME Section XI, Examination Category B-A, Item No. B 1.11 for the extended license period ending on January 18, 2034. The axial weld seams (Examination Category B-A, Item No. Bl.12) and their intersection with the associated circumferential weld seams will be examined in accordance with ASME Section XI except where specific relief is granted when essentially 100% (>90%) coverage cannot be obtained. Basis for Use The technical basis supporting the requested alternative is provided by BWRVIP-05, (EPRI TR-105697) "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations" as accepted in the staffs final safety evaluation report enclosed in a July 28, 1998, letter (Reference 4). In this letter, the staff concluded that because the failure frequency for circumferential welds in BWR plants is significantly below the criterion specified in RG 1.154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," and below the core damage frequency of any BWR plant, continued inspection would result in a negligible decrease in an already acceptably low RPV failure probability. Therefore elimination of the ISI requirements for RPV circumferential welds is justified. The staffs letter indicated that BWR applicants may request relief from ASME Code Section XI requirements for volumetric examination of circumferential RPV welds by demonstrating that (1) the failure frequency for circumferential welds in BWR plants must be significantly below the criterion specified in RG 1.154 and below the core damage frequency of any BWR plant, therefore, the failure frequency for RPV circumferential welds; and (2) the applicants must implement operator training and operating procedures that limit the frequency of cold over-pressure events to the amount specified in the July 28, 1998, SER for the BWRVIP-05 report. The letter also indicated that the requirements for inspection of RPV circumferential welds during an additional 20-year license renewal period would need plant-specific reassessment as part of any BWR LRA. The applicant also must request relief from the ASME Code Section XI requirements for volumetric examination of circumferential welds for the extended license term in accordance with 10 CFR 50.55a(z). LRA Section 4.2.5 provided a comparison of the plant-specific information with the generic analysis information in BWRVIP-05 SER to support the conclusion that the CNS RPV beltline circumferential weld parameters at 54 EFPY remained within the bounding parameters for CE RPVs at 64 EFPY from the BWRVIP-05 SER. Since the 54 EFPY mean ART value for CNS is

NLS2017071 Attachment Page 5of68 10 CFR 50.55a Request No. RI5-01 (continued) Implementation of BWRVIP-05 (GL 98-05) less than the 64 EFPY value from the BWRVIP-05 SER, the staff review concluded that the RPV conditional failure probability for CNS at 54 EFPY was bounded by the staffs generic analysis in the BWRVIP-05 SER. Therefore, the staff determined that CNS' s RPV circumferential welds satisfy the limiting conditional failure probability for circumferential welds at the end of the period of extended operation (the first condition established in the BWRVIP-05 SER). Table 1 compares the CNS reactor vessel limiting circumferential weld parameters to those used in the NRC analysis. The data in the second column is from Table 2.6-5 of the NRC SER for BWRVIP-05. The data in the third column is from Table 4.2-6 of the LRA (Reference 1). The data in the last column is the projected 54 EFPY data for CNS and has been updated to include the changes contained in the PTLR Revision 1 (Reference 3). Consistent with previous submittals, this table uses surface fluence rather than 1/4t fluence and no margin for RT NDT to be comparable with NRC assessment data, hence the reported ART is lower than that reported in the PTLR (Reference 3). Table 1 CNS Circumferential Weld Evaluation for 54 EFPY Parameter Description CE(VIPfl 64 EFPY CNS 54 EFPY Beltline CNS 54 EFPY Beltline Bounding Parameters Circumferential Weld Circumferential Weld (LRA Values) [Ref l] (Updated PIT Values) [Ref 3] Initial reference 0 -50 -50 temperature (RTNDT), °F Neutron fluence at the end of the requested 4.0E+ l8 1.48E+ l8 l.75E+ l8 relief period, n/cm2 Weld copper content, % 0.13 0.183 0.183 Weld nickel content, % 0.71 0.704 0.704 Weld chemistry factor 151.7 172.22 172.22 (CF) Increase in reference temperature (.6.RT NDT ) , 113.2 86.1 92.6 Of Mean adjusted reference temperature (ART), °F 113.2 36. l 42.6 (Initial RT NDT +

.6.RTNDT)

(1) Based on chemistry report by BWRVIP For the second condition, the staff review of the original request for the 4th Interval, concluded that the CNS implementation of operator training and establishment of procedures, limiting the frequency of code over-pressure events to the frequency specified in BWRVIP-05 SER for the remaining initial licensed period of operation described in the letter dated February 6, 2008, was

NLS2017071 Attachment Page 6 of68 10 CFR 50.55a Request No. RI5-01 (continued) Implementation of BWRVIP-05 (GL 98-05) acceptable. In LRA Section 4.2.5, CNS stated that the same procedures and training will be used for the period of extended operation. Based on this the staff determined that continued implementation of operator training and establishment of procedures limiting the frequency of cold over-pressure events would be satisfied during the period of extended operation (the second criterion established in the BWRVIP-05 SER). In addition to the above criterion, in the BWRVIP-05 SER (Reference 4), the staff concludes that the failure probability of the RPV circumferential shell welds is substantially less than that of the RPV axial shell welds. In the LRA Table 4.2-7, CNS summarized the effects of irradiation on the limiting axial weld at CNS and compared its properties to the NRC limiting plant-specific data used in the July 28, 1998, SER for BWRVIP-05. The higher copper content and chemistry factor for the CNS weld is offset by the CNS weld's lower initial RT NOT. Consequently, the CNS axial welds are less susceptible to irradiation damage than the NRC limiting plant-specific case. During the staffs review of the LRA, a comparison of the mean ART values of CNS weld data in Table 4.2-6 and Table 4.2-7 of the LRA and concluded that the mean ART for the axial welds at CNS is higher than the mean ART for the circumferential weld, indicating that the axial welds are more susceptible to radiation embrittlement than the circumferential welds. Table 2 compares the CNS reactor vessel limiting axial weld parameters to those used by the NRC analysis in BWRVIP-05. The data in the second column is from Table 2.6-5 of the NRC SER for BWRVIP-05 (Reference 4) and Table 1 of the NRC SER for BWRVIP-74 (Reference 5). The data in the third column is from Table 4.2-7 of the LRA (Reference 1). The data in the last column is the projected 54 EFPY data for CNS and has been adjusted to include the changes contained in the PTLR (Reference 3). Consistent with previous submittals, this table uses surface fluence rather than l /4t fluence and no margin for RTNoTto be comparable with NRC assessment data hence the reported ART is lower than that reported in the PTLR (Reference 3). Table 2 Effects of irradiation on CNS RPV Axial Weld Properties Parameter Description NRC Limiting CNS Data for Weld CNS Data for Weld Plant-Specific Data 2-233-B [Ref. I] 2-233-B [Ref. 3] (LRA Values) (Updated P/T Values) EFPY 64 54 54 Initial (unirradiated) reference temperature -2* -50 -50 (RTNor), °F Neutron fluence, n/cm' 0.40E+ l9** l.46E+ l8 1.72E+ l8 Fluence factor (FF) NIA 0.497 0.534 (calculated per RG 1.99) Weld copper content, % 0.219 0.27 0.27 Weld nickel content, % 0.996 1.035 1.035

NLS2017071 Attachment Page 7of68 10 CFR 50.55a Request No. RI5-01 (continued) Implementation of BWRVIP-05 (GL 98-05) Weld chemistry factor (CF) (calculated per RG 231.l ** 254.43 254.4 1.99) Increase in reference temperature (~RT NOT), 116.0 126.5 135.8

°F (FF x CF)

Mean adjusted reference temperature (ART), °F 114.0* 76.5 85 .8 (RTNDT + ~RTNoT)

NRC SER to BWRVIP-74[Ref. 5]

     ** NRC SER to BWRVIP-05[Ref 4]

To summarize, the additional analysis described in Section 4.2.5 of the LRA as modified by the changes contained in the PTLR (Reference 3) shows that the parameters projected to 54 EFPY for the CNS RPV are bounded by the staff's (64 EFPY) bounding parameters for a CE vessel in the BWRVIP-05 SER. Additionally, the CNS RPV axial welds are less susceptible to irradiation damage than the NRC limiting plant-specific case, but are more susceptible than the CNS circumferential welds. Therefore, continuation of ISi for axial welds provides additional assurance that the structural integrity of the circumferential welds is adequate. The procedures and training used to limit low temperature over-pressure events will be the same as those in use when CNS requested approval of the BWRVIP-05 technical alternative for the initial license term. The TLAA associated with reactor vessel circumferential weld inspection relief has been projected to the end of the period of extended operation in accordance with 10 CFR 54.2l(c)(l)(ii). Based on the information presented in this request, the referenced LRA with the corresponding NRC SER, and the information provided in the PTLR, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staff's July 28, 1998, safety evaluation. Duration of Proposed Alternative The duration of this request is for the extended license period ending January 18, 2034. Precedents "Peach Bottom Atomic Power Station, Units 2 and 3 - Requests for Reliefl4R-51 and 15R-52," dated January 24, 2012 (ADAMS Accession Number MLl 12770217). "Browns Ferry Nuclear Plant, Units 2 and 3 - Request for ASME Code, Section XI, Alternatives 2-ISI-30 and 3-ISI-27 for the Periods of Extended Operation Regarding Reactor Pressure Vessel Circumferential Shell Weld Examinations," dated March 14, 2017 (ADAMS Accession Number MLl 7045A 772).

NLS2017071 Attachment Page 8of68 10 CFR 50.55a Request No. RI5-01 (continued) Implementation of BWRVIP-05 (GL 98-05) References

1. NPPD letter to the U.S. NRC, "Cooper Nuclear Station, License Renewal Application, Preface through Chapter 4, References," dated September 24, 2008 (ADAMS Accession No. ML083030239).

2. US NRC letter to NPPD, "Safety Evaluation Report Related to the License Renewal of Cooper Nuclear Station," dated September 1, 2010 (ADAMS Accession No. ML102000270).

3. NPPD letter to the U.S. NRC, "Pressure and Temperature Limits Report, Revision l ," dated January 9, 2017 (ADAMS Accession No. MLl 7018Al51 and MLl 7018Al52).

4. J. Strosnider (NRC), to C. Terry (BWRVIP Chairman), "Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report {TAC NO. M93925)," letter dated July 28, 1998.

5. C. Grimes (NRC), to C. Terry (BWRVIP Chairman), "Acceptance for Referencing ofEPRI Proprietary Report TR-113596, 'BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74)' and Appendix A,

  'Demonstration of Compliance with the Technical Information Requirements of the License Renewal Rule (10 CFR 54.21)' "dated October 18, 2001 (ADAMS Accession No. MLO 12920549).

6. US NRC letter to NPPD, "Cooper Nuclear Station - Request for Relief No. RI-29 for Fourth 10-Year Inservice Inspection Interval Regarding Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds (TAC No. MD5260)," dated February 6, 2008 (ADAMS Accession No. ML080230288).

NLS2017071 Attachment Page 9 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(l) Acceptable Level of Quality and Safety ASME Code Component(s) Affected Code Class: ASME Section XI Code Class 1 Examination Category: B-N-1, B-N-2 Item Number(s): Bl3.10, Bl3.20, Bl3.30, and Bl3.40 Component Numbers: Various

Applicable Code Edition and Addenda

ASME Section XI, 2007 Edition through the 2008 Addenda Applicable ASME Code Requirements Table IWB-2500-1, Examination Categories "B-N-2, Welded Core Support Structures and Interior Attachments to Reactor Vessels," "B-N-3, Removable Core Support Structures" requires examinations based on the following Item Numbers: B 13 .10 Examine accessible areas of the reactor vessel interior (B-N-1) each period by the VT-3, visual examination method; includes only those spaces above and below the core made accessible by removal of components during normal refueling outages Bl3.20 Examine accessible interior welded attachments within the beltline region each interval by the VT-1, visual examination method (B-N-2) Bl3 .30 Examine accessible interior welded attachments beyond the beltline region each interval by the VT-3 , visual examination method (B-N-2) B 13 .40 Examine the accessible surfaces of welded core support structures each interval by the VT-3 , visual examination method (B-N-2) These examinations are performed to assess the structural integrity of the reactor vessel interior, its welded attachments, and the welded core support structure within the boiling water reactor pressure vessel.

Reason for Request

In accordance with 10 CFR 50.55a(z)(l ), NPPD is requesting NRC approval of a proposed alternative to the Code requirements provided above on the basis that the use of the BWRVIP guidelines discussed below provide an acceptable level of quality and safety.

NLS2017071 Attachment Page 10of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2 The BWRVIP Inspection and Evaluation Guidelines recommend specific inspection by BWR owners to identify material degradation with BWR components. A wealth of inspection data has been gathered during these inspections across the BWR industry. The BWRVIP Inspection and Evaluation Guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying known or potential degradation mechanisms, and require re-examination at appropriate intervals. The scope of the BWRVIP Inspection and Evaluation Guidelines exceed that of ASME Section XI and in most instances include components that are not part of the ASME Section XI jurisdiction. Use of this proposed alternative will maintain an adequate level of quality and safety and avoid duplicate or unnecessary inspections, while conserving radiological dose. Revision 1 updates the BWRVIP-18 reference to Revision 2-A and provides an updated inspection history to include the Fall 2016 (RE29) refueling outage. Revision 0 of this Relief was approved by the NRC on February 17, 2016 (Reference 14). Proposed Alternative and Basis for Use Proposed Alternative NPPD requests authorization to utilize the alternative requirements of the BWRVIP Guidelines in lieu of the requirements of ASME Code Section XI. NPPD will satisfy the Examination Category B-N-1 and B-N-2 requirements as described in Table 1 on page 15 in accordance with BWRVIP guideline requirements. This relief request proposes to utilize the identified BWRVIP guidelines in lieu of the associated Code requirements, including examination method, examination volume, frequency, training, successive and additional examinations, flaw evaluations, and reporting. Not all of the components addressed by these guidelines are Code components. The proposed alternative includes: For Examination Category B-N-1: As an alternative to meeting ASME Section XI and performing a VT-3 examination of the RPV interior above and below the core made accessible by a normal refuel outage, NPPD will implement the BWRVIP Guidelines listed below and as outlined in Table 1 on page 15. By this request for alternative the BWRVIP Guidelines will be used as an alternative to the requirements of ASME Section XI.

BWRVIP-03, Revision 17, "Reactor Pressure Vessel and Internals Examination Guidelines"

NLS2017071 Attachment Page 11of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu ofB-N-1 and B-N-2

BWRVIP-18, Revision 2-A, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines"
BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines"
BWRVIP-26-A, "BWR Top Guide Inspection and Flaw Evaluation Guidelines"
BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate ~p Inspection and Flaw Evaluation Guidelines"
BWRVIP-41, Revision 3, "BWR Jet Pump Assembly Inspection and Evaluation Guidelines"
BWRVIP-47-A, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines"
BWRVIP-138 Revision 1-A, "Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines" For Examination Category B-N-2:

As an alternative to meeting ASME Section XI and performing a VT-1 or VT-3, as required by ASME Section XI, examination of the RPV welded attachments and welded core support structures, NPPD will implement the BWRVIP Guidelines listed below and as outlined in Table 1 on page 15. By this request for alternative the BWRVIP Guidelines will be used as an alternative to the requirements of ASME Section XL

BWRVIP-03, Revision 17, "Reactor Pressure Vessel and Internals Examination Guidelines"
BWRVIP-38, "BWR Shroud Support Inspection and Flaw Evaluation Guidelines"
BWRVIP-48-A, "Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines"
BWRVIP-76, Revision 1-A, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines"
BWRVIP-100-A, "Updated Assessment of the Fracture Toughness oflrradiated Stainless Steel for BWR Core Shrouds" Note: If flaw evaluations are required for BWRVIP-76, Revision 1-A, examinations, the fracture toughness values ofBWRVIP-100-A will be utilized.

When a BWRVIP Guideline refers to ASME Section XI, the technical requirements of ASME Section XI as described by the BWRVIP Guideline will be met, but the examination is under the auspices of the BWRVIP program as defined by BWRVIP-94NP, Revision 2, "BWRVIP Vessel and Internals Project Program Implementation Guide." The NPPD reactor vessel internals inspection programs have been developed and implemented to satisfy the requirements ofBWRVIP-94NP, Revision 2. It is recognized that the BWRVIP executive committee periodically revises the BWRVIP guidelines to address industry operating experience, include enhancements to inspection techniques, and add or adjust flaw evaluation methodologies. BWRVIP-94NP, Revision 2, states that where guidance in existing BWRVIP documents has been supplemented or revised by subsequent correspondence approved by the

NLS2017071 Attachment Page 12of68 10 CFR 50.55a Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 BWRVIP Executive Committee, the vessel and internals program shall be modified to reflect the new requirements and implement the guidance within two refueling outages, unless a different schedule is specified by the BWRVIP. However, if new guidance approved by the Executive Committee includes changes to NRC approved BWRVIP guidance that are less conservative than those approved by the NRC, the less conservative guidance shall be implemented only after NRC approves the changes, which generally means publication of a "-A" document or equivalent. Therefore, where the revised version of a BWRVIP inspection guideline continues to also meet the requirements of the version of the BWRVIP inspection guideline approved by the NRC, it may be implemented. Otherwise, the revised guidelines will only be implemented after NRC approval of the revised BWRVIP guidelines or a plant-specific request for alternative has been approved. Table 1 below only represents the most current comparison. Any deviations from the referenced BWRVIP Guidelines for the duration of the proposed alternative will be appropriately documented and communicated to the NRC, per the BWRVIP Deviation Disposition Process. Note that other regulatory commitments (i.e., NUREG-0619) are still being implemented separately from the ASME Section XI Program or this request for alternative. In the event that conditions are identified that require repair or replacement and the component is within the jurisdiction of ASME Section XI (welded attachments to the RPV or Core Support Structure), the repair or replacement activities will be performed in accordance with ASME Section XI, Article IW A-4000. Subsequent examinations will be in accordance with the applicable BWRVIP Guideline. Basis for Use As part of the BWRVIP initiative, the BWR reactor internals and attachments were subjected to a safety assessment to identify those components that provide a safety function and to determine if long-term actions were necessary to ensure continued safe operation. The safety functions considered are those associated with (1) maintaining a coolable geometry, (2) maintaining control rod insertion times, (3) maintaining reactivity control, (4) assuring core cooling and (5) assuring instrumentation availability. The results of the safety assessment are documented in BWRVIP-06, Revision 1-A, "BWR Vessel and Internals Project Safety Assessment ofBWR Reactor Internals" which has been approved by the NRC. As a result of BWRVIP-06, Revision 1-A, component specific BWRVIP guidelines were developed providing appropriate examination and evaluation requirements to address the specific component safety function and potential degradation mechanism.

NLS2017071 Attachment Page 13 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Along with the component specific guidelines, the BWRVIP has established a reporting protocol for examination results and deviations. The NRC has agreed with the BWRVIP approach in principal and has issued Safety Evaluations for many of these guidelines (see References). As additional justification, page 17, "Comparison of ASME Code Section XI Examination Requirements to BWRVIP Examination Requirements," provides specific examples which compare the inspection requirements of ASME Code Section XI Table IWB-2500-1 , Item Numbers Bl3 .10, Bl3 .20, Bl3.30 and Bl3.40 to the inspection requirements in the BWRVIP documents. Specific BWRVIP documents are provided as examples. This comparison also includes a discussion of the inspection methods. Therefore, use of the BWRVIP guidelines as an alternative to ASME Section XI, as shown by the comparison provides an acceptable level of quality and safety. Duration of Proposed Alternative This proposed alternative will be used for the Fifth Ten-Year Interval of the Inservice Inspection Program for CNS. Precedents Similar request for alternatives has been previously approved for the following other licensees.

1. US NRC letter to Entergy Operations, "Grand Gulf Nuclear Station, Unit 1 - Request for Relief GG-ISI-017, Alternative to Use Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (TAC No. MF2357)", dated June 30, 2014 (ML14148A262).

2. US NRC letter to Entergy Operations, "River Bend Station, Unit 1 - Request for Relief No.

RBS-ISI-019, Alternative to Use Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of ASME Code, Section XI Requirements for the Fourth 10-Year Inservice Inspection Interval (TAC No. MF1867), dated May 30, 2014 (ML14127A327).

3. US NRC letter to Exelon Generation Company, LLC, "Dresden Nuclear Power Station, Units 2 and 3 - Safety Evaluation in Support of Request for Relief Associated With the Fifth l 0-y ear Inservice Inspection Interval Program (TAC Nos. ME9682 , ME9683 , ME9684, ME9685 , ME9686, ME9687, ME9688, ME9689, ME9690, ME9691 , ME9692, ME9693 ,

ME9694, ME9695 , ME9696, and ME9697), dated September 30, 2013 (ML13260A585).

4. US NRC letter to Exelon Generation Company, LLC, "Quad Cities Nuclear Power Station Units 1 and 2 - Safety Evaluation in Support of Request for Relief Associated With the Fifth 10 Year Interval Inservice Inspection Program (TAC Nos. ME9668 , ME9669, ME9670,

NLS2017071 Attachment Page 14of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 l\.1E9671 , l\.1E9672, l\.1E9674, l\.1E9675, l\.1E9676, l\.1E9677, l\.1E9678, l\.1E9679, l\.1E9680, l\.1E968 l ), dated September 30, 2013 (l\.1L 13267 A097).

5. US NRC letter to Exelon Nuclear, "Oyster Creek Nuclear Generating Station - Relief From the Requirements of the AS1\.1E Code, Relief Request No. I5R-Ol (TAC No. l\.1E9490), dated August 5, 2013 (l\.1Ll3169A062).

References

1. US NRC letter to BWRVIP, "Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report, BWRVIP-06-A: BWR (Boiling Water Reactor) Vessel and Internals Project (BWRVIP), Safety Assessment ofBWR Reactor Internals, Revised Section 4.0:

Consideration of Loose Parts" (TAC No. l\.1C7448) dated July 29, 2008 (l\.1L082030758).

2. US NRC letter to BWRVIP, "Final Proprietary Safety Evaluation for Electric Power Research Institute Topical Report, "BWRVIP-18, Revision 2: Boiling Water Reactor Core Spray Internals Inspection and Flaw Evaluation Guidelines (TAC No. l\.1F8809}," dated February 22, 2016 (l\.1Ll6011Al99).

3. US NRC letter to BWRVIP, "Final Safety Evaluation ofBWRVIP Vessel and Internals Project, BWR Vessel and Internals Project, BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25)," EPRI Report TR-107284, December 1996 (TAC No. l\.197802),"

dated December 19, 1999.

4. US NRC letter to BWRVIP, "NRC Approval Letter ofBWRVIP-26-A, BWR Vessel and Internals Project Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines," dated August 29, 2005 (l\.1L052490550).

5. US NRC letter to BWRVIP, "Non-Proprietary Version ofNRC Staff Review of BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate ~p Inspection and Flaw Evaluation Guidelines," dated June 9, 2004 (l\.1L04 l 700446).

6. US NRC letter to BWRVIP, "Final Safety Evaluation of the "BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38),"

EPRI Report TR-108823 (TAC No. l\.199638)," dated July 24, 2000 (l\.1L003735498).

7. US NRC letter to BWRVIP, "Final Safety Evaluation of the "BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines (BWRVIP-41)," (TAC No. 1\.199870)," dated February 4, 2001 (l\.1L0104601 l l).

NLS2017071 Attachment Page 15of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2

8. US NRC letter to BWRVIP, "NRC Approval Letter of BWRVIP-47-A, "BWR Vessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines," dated September 1, 2005 (ML052490537).

9. US NRC letter to BWRVIP, "NRC Approval Letter of BWRVIP-48-A, "BWR Vessel and Internals Project Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines,"

dated July 25, 2005 (ML052130284).

10. US NRC letter to BWRVIP, "Final Safety Evaluations of the Boiling Water Reactor Vessel and Internals Project 76, Rev. 1 Topical Report, "Boiling Water Reactor Core Shroud Inspection and Flaw Evaluation Guidelines" (TAC No. ME8317)," dated November 12, 2014.

11. Letter from Chairman, BWR Vessel and Internals Project to NRC, "Project No. 704 -

BWRVIP Program Implementation Guide (BWRVIP-94NP, Revision 2)," dated September 22, 2011 (ML11271A058).

12. US NRC letter to BWRVIP, "NRC Approval Letter with Comment for BWRVIP-100-A, BWR Vessel and Internals Project, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds," dated November 1, 2007 (ML073050135).

13. US NRC letter to BWRVIP, "Electric Power Research Institute Final Safety Evaluation for Technical Report 1016574 "BWRVIP-138, Revision 1: BWR [Boiling Water Reactor]

Vessel and Internals Project 'Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines' (TAC No. ME2191)," dated May 14, 2012 (ML1208A139).

14. US NRC letter to NPPD, "Cooper Nuclear Station - Request for ReliefRI5-02, Alternative to Use Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (CAC No. MF6336)," dated February 17, 2016 (ML16034A479).

NLS2017071 Attachment Page 16of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu ofB-N-1 and B-N-2 Table 1 Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements (Note 1) ASMEitem ASME Applicable BWRVIP ASMEExam ASME BWRVIPExam BWRVIP No. Table Component Exam BWRVIP Exam Scope Frequency Scope Frequency IWB-2500-1 Type Document Type B13.10 Reactor Vessel Interior Accessible Areas VT-3 Each Period BWRVIP-18, Overview examinations of components during (Non-specific) 25, 26, 38, BWRVIP examinations are performed to satisfy 41, 47, 48, Code VT-3 inspection requirements. 76, 138 B13.20 Interior Attachments Accessible VT-1 Each 10-year BWRVIP-48 Riser Brace EVT-1 100% in first 12 within Beltline - Riser Welds Interval Table 3-2 Attachment years, 25% during Braces each subsequent 6 years Lower Surveillance BWRVIP-48, Bracket VT-1 Each l 0-Year Specimen Holder Table 3-2 Attachment Interval Brackets B13.30 Interior Attachments Accessible VT-3 Each 10-year BWRVIP-48, Bracket VT-3 Each l 0-Year beyond Beltline - Steam Welds interval Table 3-2 Attachment Interval Dryer Hold-down Brackets Guide Rod Brackets BWRVIP-48, Bracket VT-3 Each 10-Year Table 3-2 Attachment Interval Steam Dryer Support BWRVIP-48, Bracket EVT-1 Each 10-Year Brackets Table 3-2 Attachment Interval Feedwater Sparger BWRVIP-48, Bracket EVT-1 Each 10-Year Brackets Table 3-2 Attachment Interval Core Spray Piping BWRVIP-48, Bracket EVT-1 Every 4 Refueling Brackets Table 3-2 Attachment Cycles Upper Surveillance BWRVIP-48, Bracket VT-3 Each 10-Year Specimen Holder Table 3-2 Attachment Interval Brackets

NLS2017071 Attachment Page 17of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2 Table 1 Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements (Note 1) ASME Item ASME Applicable BWRVIP ASMEExam ASME BWRVIPExam BWRVIP No. Table Component Exam BWRVIP Exam Scope Frequency Scope Frequency IWB-2500-1 Type Document Type Shroud Support (Weld BWRVIP-38, WeldH-9 EVT-1 or Maximum of6 H9) including gussets 3.1.3.2, including gussets UT years for EVT-1 , Figures 3-2 Maximum of 10 and 3-5 years for UT B13.40 Integrally Welded Core Accessible VT-3 Each 10-year BWRVIP-38, Shroud support EVT-1 or Based on as-found Support Structure Surfaces interval 3.1.3.2, welds H8 and H9 UT conditions, to a Figures 3-2 including gussets maximum 6 years and 3-5 for one side EVT-1, 10 years for UT where accessible Shroud Horizontal BWRVIP-76, Welds Hl-H7 as UT or Based on as-found Welds 2.2 applicable EVT-1 conditions, to a maximum of 10 years for UT when inspected from both sides of the welds Shroud Vertical Welds BWRVIP-76, Vertical Welds as EVT-1 or Maximum 10 2.3 applicable UT years for UT based on inspection of horizontal welds Note:

1. This Table provides only an overview of the requirements. For more details, refer to ASME Section XI, Table IWB-2500-1 and the appropriate BWRVIP document.

NLS2017071 Attachment Page 18of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Comparison of ASME Code Section XI Examination Requirements to BWRVIP Examination Requirements The following provides a comparison of the examination requirements provided in ASME Code Section XI Table IWB-2500-1 , Examination Category B-N-1 and B-N-2, Item Numbers B13.10, B13.20, 813.30, and 813.40, to the examination requirements in the BWRVIP Guidelines. Specific BWR VIP Guidelines are provided as examples for comparisons. This comparison also includes a discussion of the examination methods. Code Requirement - 813.10 - Reactor Vessel Interior Accessible Areas (B-N-1) The ASME Section XI Code requires a VT-3 examination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during normal refueling outages. The frequency of these examinations is specified as the first refueling outage, and at intervals of approximately 3 years during the first inspection interval, and each period during each successive l 0-year Inspection Interval. Typically, these examinations are performed every other refueling outage of the Inspection Interval. This examination requirement is a non-specific requirement that is a departure from the traditional Section XI examinations of welds and surfaces. As such, this requirement has been interpreted and satisfied differently across the licensees, and vendors of this inspection service. Based on the acceptance criteria specified in IWB-3520.2, the examination is to identify relevant conditions such as distortion or displacement of parts, loose, missing, or fractured fasteners, foreign material, corrosion, erosion, or accumulation of corrosion products, wear, and structural degradation. Portions of the various examinations required by the applicable BWRVIP Guidelines require access to accessible areas of the reactor vessel during each refueling outage. Examination of Core Spray Piping and Spargers (BWRVIP-18-R2-A), Top Guide (BWRVIP-26-A), Jet Pump Welds and Components (BWRVIP-41-R3), Interior Attachments (BWRVIP-48-A), Core Shroud Welds (BWRVIP-76-Rl-A), Shroud Support (BWRVIP-38), and Lower Plenum Components (BWRVIP-47-A) provides such access. Locating and examining specific welds and components within the reactor vessel areas above, below (if accessible), and surrounding the core (annulus area) entails access by remote camera systems that essentially perform equivalent VT-3 examination of these areas or spaces as the specific weld or component examinations are performed. This provides an equivalent method of visual examination on a more frequent basis than that required by the ASME Section XI Code. Evidence of wear, structural degradation, loose, missing, or displaced parts, foreign materials, and corrosion product buildup can be, and has been observed during the course of implementing these BWRVIP examination requirements. Therefore, the requirements specified by the BWRVIP Guidelines meet or exceed the subject Code requirements for examination method and frequency of the interior of the reactor vessel. Accordingly, these BWRVIP examination requirements provide an acceptable level of quality and safety as compared to the subject Code requirements.

NLS2017071 Attachment Page 19 of68 10 CFR 50.55a Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Code Requirement - B13.20 - Interior Attachments within the Beltline (B-N-2) The ASME Section XI Code requires a VT-1 examination of accessible reactor interior surface attachment welds within the beltline each 10-year interval. In the BWR, this includes the Jet Pump Riser Brace Weld-to-Vessel Wall and the Lower Surveillance Specimen Support Bracket Welds-to-Vessel Wall. In comparison, the BWRVIP requires the same examination method and frequency for the Lower Surveillance Specimen Support Bracket Welds, and requires an EVT-1 examination on the remaining attachment welds in the beltline region in the first 12 years, and then 25% during each subsequent 6 years. The Jet Pump Riser Brace examination requirements are provided below to show a comparison between the Code and the BWRVIP examination requirements. Comparison to BWRVIP Requirements - Jet Pump Riser Braces (BWRVIP-41-R3 and BWRVIP-48-A)

The ASME Code requires a 100% VT-1 examination of the Jet Pump Riser Brace-to-Reactor Vessel Wall Pad welds each 10-year Interval.
The BWRVIP requires an EVT-1 baseline examination of 100% of the Jet Pump Riser Brace-to-Reactor Vessel Wall Pad welds in the first 12 years with at least 50% being inspected in the first 6 years. Reinspection consists of 25% during each subsequent 6 year period.
BWRVIP-48-A specifically defines the susceptible regions of the attachment that are to be examined.

The Code VT-1 examination is conducted to detect discontinuities and imperfections on the surfaces of components, including such conditions as cracks, wear, corrosion, or erosion. The BWRVIP EVT-1 is conducted to detect discontinuities and imperfections on the surface of components and is additionally specified to detect potentially very tight cracks characteristic of fatigue and IGSCC, the relevant degradation mechanisms for these components. General wear, corrosion, or erosion although generally not a concern for inherently tough, corrosion resistant stainless steel material, would also be detected during the process of performing a BWRVIP EVT-1 examination. The ASME Code visual examination method requires (depending on applicable ASME Edition) that a letter character with a height of 0.044 inches can be read. The BWRVIP EVT-1 visual examination method requires the same 0.044 inch resolution on the examination surface and additionally the performance of a cleaning assessment and cleaning as necessary. While the Jet Pump Riser Brace configuration varies depending on the vessel manufacturer, BWRVIP-48-A includes diagrams for each configuration and prescribes examination for each configuration. I_

NLS2017071 Attachment Page 20 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 The calibration standards used for BWRVIP EVT-1 examinations utilize the same Code characters, thus assuring at least equivalent resolution compared to the Code. Although the BWRVIP examination may be less frequent, it is a more comprehensive method. Therefore, the BWRVIP guidance provides an acceptable level of quality and safety to that provided by the ASMECode. Code Requirement - 813.30 - Interior Attachment Beyond the Beltline Region (B-N-2) The ASME Section XI Code requires a VT-3 examination of accessible Reactor Interior Surface Attachment Welds beyond the beltline each 10-year Interval. In the Boiling Water Reactor, this includes the Core Spray Piping Primary, the Upper Surveillance Specimen Support Bracket Welds-to-Vessel Wall, the Feedwater Sparger Support Bracket Welds-to-Reactor Vessel Wall, the Steam Dryer Support and Hold-Down Bracket Welds-to-Reactor Vessel Wall, the Guide Rod Support Bracket Weld-to-Reactor Vessel Wall, the Shroud Support Plate-to-Vessel Welds, and Shroud Support Gussets. BWRVIP-48-A requires as a minimum the same VT-3 examination method as the Code for some of the interior attachment welds beyond the beltline region, and in some cases specifies an enhanced visual examination technique EVT-1 for these welds. For those interior attachment welds that have the same VT-3 method of examination, the same scope of examination (accessible welds), the same examination frequency (each 10 year interval) and ASME Section XI flaw evaluation criteria, the level of quality and safety provided by the BWRVIP requirements are equivalent to that provided by the ASME Code. The Core Spray Piping Bracket-to-Vessel Attachment Weld is used as an example for comparison between the Code and BWRVIP examination requirements as discussed below: Comparison to BWRVIP Requirements - Core Spray Piping Bracket Welds relative to BWRVIP-48-A

The Code examination requirement is a VT-3 examination of each weld every 10 years.
The BWRVIP examination requirement is an EVT-1 for the Core Spray Piping Bracket Attachment Welds with each weld examined every four cycles (8 years for units with a 2 year fuel cycle)

The BWRVIP examination method EVT-1 has superior flaw detection and sizing capability than the Code VT-3 , the examination frequency is greater than the Code requirements, and the same flaw evaluation criteria are used. The Code VT-3 examination is conducted to detect component structural integrity by ensuring the component's general condition is acceptable. An enhanced EVT-1 is conducted to detect discontinuities and imperfections on the examination surfaces, including such conditions as tight cracks caused by IGSCC or fatigue, the relevant degradation mechanisms for BWR internal attachments. Additionally, BWRVIP-48-A guidance requires indications detected by an EVT-1

NLS2017071 Attachment Page 21 of 68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2 to be examined by ultrasonics to determine if the indication has propagated into the reactor vessel base material. Therefore, with the EVT-1 examination method, the same examination scope (accessible welds), an increased examination frequency (8 years instead of 10 years) in some cases, and the same flaw evaluation criteria (ASME Code Section XI), the level of quality and safety provided by the BWRVIP criteria is superior to that provided by the ASME Code. Code Requirement - B13.40 - Integrally Welded Core Support Structures (B-N-2) The ASME Code requires a VT-3 examination of accessible surfaces of the welded core support structure each 10-year interval. In the boiling water reactor, the welded core support structure has primarily been considered the shroud support structure, including the shroud support plate (annulus floor), the shroud support ring, the shroud support welds, and the shroud support gussets. In later designs, the shroud itself is considered part of the welded core support structure. Historically, this requirement has been interpreted and satisfied differently across the industry. The proposed alternate examination replaces this ASME requirement with specific BWRVIP guidelines that examine susceptible locations for known relevant degradation mechanisms.

The Code requires a VT-3 of accessible surfaces each 10-year interval.
The BWRVIP requires as a minimum the same examination method (VT-3) as the Code for integrally welded Core Support Structures, and for specific areas, requires either an enhanced visual examination technique (EVT-1) or volumetric examination (UT).

BWRVIP recommended examinations of integrally welded core support structures are focused on the known susceptible areas of this structure, including the welds and associated weld heat affected zones. As a minimum, the same or superior visual examination technique is required for examination at the same frequency as the Code examination requirements. In many locations, the BWRVIP guidelines require a volumetric examination of the susceptible welds at a frequency identical to the Code requirement. For other integrally welded core support structure components, the BWRVIP requires an EVT-1 or UT of core support structures. The core shroud is used as an example for comparison between the Code and BWRVIP examination requirements as shown below. Comparison to BWRVIP Requirements - BWR Core Shroud Examination and Flaw Evaluation Guideline (BWRVIP-76)

The Code requires a VT-3 examination of accessible surfaces every 10 years.
The BWRVIP requires an EVT-1 examination from the inside and outside surface where accessible or ultrasonic examination of each core shroud circumferential weld that has not been structurally replaced with a shroud repair at a calculated "end of interval" that will vary depending upon the amount of flaws present, but not to exceed ten years.

NLS2017071 Attachment Page 22of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 The BWRVIP recommended examinations specify locations that are known to be vulnerable to BWR relevant degradation mechanisms rather than "all surfaces." The BWRVIP examination methods (EVT-1 or UT) are superior to the Code required VT-3 for flaw detection and characterization. The BWRVIP examination frequency is equivalent to or more frequent than the examination frequency required by the Code. The superior flaw detection and characterization capability, with an equivalent or more frequent examination frequency and the comparable flaw evaluation criteria, results in the BWRVIP criteria providing a level of quality and safety equivalent to or superior to that provided by the Code requirements.

NLS2017071 Attachment Page 23of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Reactor Internals Inspection History Plant: Cooper Nuclear Station Dated: Nov 08 , 20 16 (RE29 outage end) Component in BWRVlP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Core Shroud Fall 1995 UT Baseline UT performed on welds H 1 through H7 (RE 15) per BWRVIP guidelines. Indications identified in 4 circumferential welds. No examinations on vertical welds. No repair required. Spring 2005 UT UT examinations were performed on welds H- 1 (RE22) through H-4 including a portion of vertical weld Vl6. Examination of welds H5-H7 was deferred to fa ll 2006. Single sided UT examinations were performed on welds H-1 through H-3 with welds H-4 and vertical weld (V- 16) receiving dual sided examinations. Percentage of welds examined: H 1 (54.9%), H2 (55 .7%), H3 (63 .9%), H4 (58.4%). The previously identified eight (8) flaws in HI showed a net decrease in length. No new flaws in H2 were identified. The eight (8) flaws in H3 were reexamined with one (1) new flaw identified for a total increased change in flaw length relative to total weld length of7.5 %. Two (2) new minor flaws were discovered in the HAZ ofH4. In addition, a total of eleven ( 11) minor indications were identified in the base metal adjacent to H4. Six (6) of the indications exhibited characteristics associated with Stress Corrosion Cracking (SCC) in areas subjected to cold working during the shroud fabrication/installation process. The remaining five (5) indications did not exhibit characteristics of sec but appeared to exhibit characteristics commonly observed from localized attachment removal sites. The indications were determined to be acceptable by analysis. No indications were observed in the vertical weld. Fall 2006 UT UT examinations were performed on welds H5 , (RE23) H6a, H6b, and H7 using phased array. Two- (2) sided examinations were performed on all welds except H7 that received a one-sided UT examination. Coverage was estimated at greater than 72% for welds H5 , H6a, and H6b. H7 received greater than 53% coverage. A previously identified indication in H5 was re-examined with no apparent change. A previously identified indication in H6a was re-examined with no

NLS2017071 Attachment Page 24 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection apparent change. A new minor indication was discovered in weld H6b in an area previously scanned in RE 16 (1995). Two (2) new minor indications were discovered in weld H7 , one in a previously scanned location and the other in an area not previously scanned. VT-3 VT-3 examination of shroud per ASME Section XI, B-N-2 requirements. Discovered an indication approximately ten (10) inches long behind JP-19. Analyzed as acceptable. Spring 2008 VT-3 Performed first ASME B-N-2 VT-3 successive (RE24) examination of flaw discovered in base metal behind JP-19. No changes in the indication. Spring 2011 VT-3 Performed second ASME B-N-2 VT-3 successive (RE26) examination of flaw discovered in base metal behind JP-19. No changes in the indication. Fall2014 UT UT exams were performed on the HI thru H7 (RE28) welds along with the V 16 and base material flaw behind JP-19. The previously identified indications showed no apparent changes in growth and none were through wall. VT-3 examination of shroud per ASME Section XI, B-N-2 requirements. Also performed third ASME B-N-2 VT-3 successive examination of flaw discovered in base metal behind JP-19. No changes in the indication. Shroud Support/ Access 1993-1995 VT-3 and VT-3 examinations of welds on 50% of core plate Hole Covers UT each outage. No indications. UT of access hole covers (AHC) in 1993 . No indications. Spring 1997 VT-3 VT-3 examinations on 50% of the core shroud (RE17) support plate. No indications. VT-1 VT-I examinations of AHC in accordance with GE SIL 462. No indications. Fall 1998 VT-3 VT-3 examinations on 50% of the core shroud (RE18) support plate. No indications. VT-1 VT-1 of AHC's in accordance with GE SIL 462 . No indications. VT-1 of gusset plate welds between 0-180° to B-N-2.

NLS201 7071 Attachment Page 25of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Spring 2000 VT-3 VT-3 examinations on 50% of the core shroud (RE19) support plate. No indications. VT-1 VT-1 examinations of AHCs in accordance with GE SIL 462. No indications. Fall 2001 EVT-1 EVT-1 examinations on 17% of the H8 and H9 (RE20) welds. EVT-1 examinations on 6 gusset welds and AHCs. No indications. UT UT examination of AHCs. No indications. Spring 2003 EVT-1 EVT-1 examinations on four (4) gusset welds. No (RE21) indications. Spring 2005 UT UT examinations on 11. 7% of the H9 weld length. (RE22) No indications. Fall 2006 EVT-1 EVT-1 examinations performed on approximately (RE23) 16% ofH8 weld length with no relevant indications. EVT-1 examinations of ARC per SIL 462. No indications. Spring 2008 EVT-1 EVT-1 examinations performed on accessible (RE24) lengths of welds on seven (7) gussets. No indications. Fall 2012 EVT-1 EVT- 1 examinations on 16.7% of the H8 weld. (RE27) EVT- 1 examinations on accessible lengths of welds on two (2) gussets @ 195° and 315°. No indications. Fall2014 UT/EVT- 1 UT performed on H9 with 13.4% coverage. EVT-(RE28) 1 performed on both ABC's. No indications. Core Spray Piping 1980' s to 1995 VT-l NT-3 IEB 80-13 examinations of piping and welds in annulus. Three(3) indications identified in Fall 1995 outage by EVT-1 . No repair required. Spring 1997 UT UT examinations of CS P8a and P8b welds . (REI 7) Indications on one P8a and one P8b weld (first discovery). Evaluated as acceptable. EVT-1 EVT-1 examinations on balance of piping.

NLS2017071 Attachment Page 26 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Fall 1998 UT UT examinations on the P8a and P8b indications (RE18) were re-examined. EVT-1 Balance examined by EVT-1. No visual indications. Spring 2000 UT UT examinations on P8a and P8b welds with (RE19) indications. No repair required. EVT-1 EVT-1 of P3 , P4, PS , P6, and P7 welds. No visual indications. Fall 2001 UT UT examinations on P3's, three (3) P4 's, PS's, (RE20) P6' s, P7's, P8a's and P8b's. EVT examinations of thirty-one of the CS piping welds. EVT-1 EVT-1 examinations on fifteen (15) welds. Indications re-examined on P8a weld and P8b welds. Spring 2003 UT UT examinations on all P8a and P8b welds. (RE21) Identified three (3) flaw indications on one P8b weld and one (1) flaw indication on one P8a weld. No change in length. EVT-1 EVT-1 examinations on both junction box covers and accessible portions of both Pl ' s, 2 - P2 's, 4 - P3 ' s, l-P4a , l-P4b, l-P4c, l-P4d. EVT-1 all P8a and P8b welds. No indications. Spring 2005 EVT-1 EVT-1 examinations of both Pl ' s . The (RE22) examination revealed that the P 1 weld is not a creviced weld based on the presence of an external weld on the tee box near the nozzle thermal sleeve. EVT-1 examinations were performed on both P2 welds, the four (4) P3 welds, the 4a - 4d welds at 190°, the PS 's, P6's, and P7's, the four (4) P8a's, and four (4) P8b 's. Fall 2006 UT UT examinations of P8b welds. Previous (RE23) indications showed no change in size. EVT-1 EVT-1 examinations of piping welds and bracket attachment welds. No new relevant indications observed. Spring 2008 EVT-1 EVT-1 of indication near P 1 at 90°. No change. (RE24) EVT-1 of Pl at 270°. EVT-1 ofP2 ' s and P3 's at 90° and 270°. EVT- 1 of P4a, -b, -c, and-d at 170° J

NLS2017071 Attachment Page 27of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection EVT-1 ofPS's, P6 's, and P7' s at 10°, 170°, 190°, and 3S0°. Fall 2009 EVT-1 EVT- 1 examinations near P 1welds at 90° and (RE2S) 270°. No change with the indication near the Pl at 90° (Loop A). EVT-1 examinations of the four (4) P3 , PS, P6 and P7 welds, EVT-1 examinations of downcomer welds P4a, P4b, P4c, and P4d at 10°. EVT- 1 examinations of four (4) P8a and P8b welds. No change with visual indication of P8b at 10°. UT UT performed on all four (4) P8a and P8b welds. Previously identified indications on the P8a at 190° (Loop B) and the P8b at 10° (Loop A) did not show any change. Spring 2011 EVT-1 EVT- 1 of area and indication adjacent to Pl weld (RE26) at 90° (Loop A). No change to the indication. EVT-1 of area adjacent to P 1 weld at 270°. No indications. EVT-1 of the P2 welds at 90° and 270°. EVT-1 of the four (4) P3, PS, P6, and P7 welds. EVT-1 of downcomer welds P4a, P4b, P4c, and P4d at 190°. No indications. Fall2012 EVT-1 EVT-1 of area and indication adjacent to P 1weld (RE27) at 90° on A Loop. No change to the indication. EVT-1 of the P2, P3a, & P3b @ 90°. EVT-1 of PS, P6, & P7 @ 10° & 170°. No indications. EVT- 1 of area adjacent to P 1 weld at 270° on B Loop. No indications. EVT- 1 of the P2, P3a, & P3b welds at 270°. EVT-1 of PS , P6, and P7 welds @ 190° & 3S0°. EVT-1 of downcomer welds P4a, P4b, P4c, and P4d @ 3S0°. No indications. Loops A & B, EVT-1 of the bracket attachment welds PB @ 30°, 1S0°, 210°, and 330°. No indications. UT UT performed on all four (4) P8a and P8b welds. Previously identified indications on the P8a at

NLS201 707 1 Attachment Page 28 of 68 10 CFR 50.55a Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection 190° (Loop B) and the P8b at 10° (Loop A) did not show any change. Fa112014 EVT-1 Loop A (RE28) EVT-1 of area and indication adjacent to P 1weld at 90° on A Loop. No change to the indication. EVT-1 of the P2, P3a & P3b welds@ 90°. EVT-1 of P5, P6, & P7 @ 10° & 170°. EVT-1 of downcomer welds P4a, P4b, P4c, and P4d @ 170°. No indications observed. LoopB EVT-1 of area adjacent to Pl weld at 270°. EVT-l of the P2, P3a, & P3b welds at 270°. EVT-l of P5, P6, and P7 welds @ 190° & 350°. No indications observed. Fall 2016 EVT-1 Loop A (RE29) EVT- l of area and indication adjacent to Pl weld at 90° on A Loop. No change to the indication. EVT- l of downcomer welds P4a, P4b, P4c, and P4d at 10°. No indications. EVT-1 of two (2) P8a and two (2) P8b welds at l 0° and 170°. No change with visual indication of P8b at 10°. LoopB EVT-1 of two (2) P8a and two (2) P8b welds at 190° and 350°. No indications. Core Spray Sparger 1980's to 1995 VT-1 /UT IEB 80-13 of welds on sparger. No indications. Spring 1997 EVT-1 EVT-1 examinations of sparger welds and brackets (REI 7) per BWRVIP-18 . Debris (wire) in C-sparger Nozzle I 5C identified. No other indications. Fall 1998 EVT-1 EVT- 1 examinations of sparger welds and brackets (RE18) inspected in accordance with BWRVIP-18. Debris (wire) in C-sparger Nozzle 15C was reconfirmed. No other indications. Spring 2000 EVT-1 EVT- 1 examinations of sparger and brackets. (RE19) Five (5) indications evaluated as acceptable.

NLS2017071 Attachment Page 29of68 10 CFR 50.55a Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Fall 2001 VT- 1 VT-I of25% ofS3a, S3b, and S3c welds. No (RE20) indications. EVT-1 EVT- 1 examinations of all S l , S2, and S4 welds examined with no indications. Spring 2003 VT-1 VT-I of25% ofS3a & S3b' s and all bracket (RE2 l) welds. No indications. EVT- 1 EVT-1 examinations of two S l 's, two S2 ' ,s, both XTRW welds near t-boxes, and four (4) S4 welds. No indications. Spring 2005 NI A Sparger examinations deferred to fall 2006 (RE22) (RE23) . Fall 2006 VT-1 VT-1 on 50% of the S3a, S3b, and S3c welds and (RE23) l 00% on sparger brackets. No indications. EVT-1 EVT- 1 on 100% of SI 's and S2 ' s and S4 's. No indications. Spring 2008 VT- I VT-I on 25% of the S3a, S3b, and S3c welds. (RE24) VT-I ofSB ' s at 90°, 92°, 119°, 149°, 210°, 24I 0 and 268°. EVT-I EVT- 1 examinations of S l 's and S2 ' s at 170° and 190°. EVT-1 examinations ofS3a, S3b at 92° to 269°. EVT- I examinations ofS3c at 99°. EVT-1 examinations of S4' s at 91 ° and 269°. Fall 2009 VT- 1 VT-1on25% of the S3a, S3b, and S3c welds. (RE25) VT-1 ofSB 's at 272°, 299°, 30°, 329°, 61 °, 88° and 270°. EVT-1 EVT-1 examinations of S l and S2 and at l 0° and 350°. EVT-1 examinations of two (2) additional welds near the 350° tee-box S2 welds. Spring 2011 VT-I VT-I on 25% of the S3a, S3b, and S3c welds. VT-(RE26) l ofsparger brackets at 90°, 92°, 119°, 149.5°, 210.5°, 241 ° and 268°. No indications. EVT- 1 EVT- 1 on C Sparger, SI @ 170°, S2 @ 168° & 172°, S4 @ 9I 0 & 269°. No indications. EVT-1 on D Sparger, Sl @ 190°, S2 @ I88 ° & 192°, S4 @ 91 ° & 269°. No indications.

NLS2017071 Attachment Page 30 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Fall 2012 EVT-1 EVT-1 on A sparger, SI @ 10, S2 @ 8° & 12°, (RE27) and S4 @ 89° & 271 °. No indications. EVT-1 on B sparger, SI @ 350°, S2 @ 348° & 352°, XTRW welds near T-box @ 346° & 354°, and S4 @ 89° & 271 °. No indications. VT- I VT-1 on the B sparger, S3a & S3b @ 27 1°-89° and S3c @ 279°. No indications. VT- I of the Sparger brackets at 30.5°, 61°, 88°, 270°, 272°, 299°, and 329.5°. No indications. Fa112014 EVT-1 EVT- 1 on C Sparger, Sl @ 170°, S2 @ 168° & (RE28) 172°, S4 @ 91 ° & 269°. No indications observed. EVT- 1 on D Sparger, Sl @ 190°, S2 @ 188° & 192°, S4's @ 91 ° & 269°. No indications observed. VT-1 on 25% of the S3a, S3b, and S3c welds. VT-VT- I I ofsparger brackets (SB) at 90°, 92°, 119°, 149.5°, 210.5°, 241 ° and 268°. No indications observed. Top Guide (Rim, etc.) 1991-1995 VT VT of top guide beams of fifty (50) cells was performed in 1991 per RICSIL 059. No indications. VT exams of the members in the load path between the top guide and core shroud in 1995 per SIL 588. One (1) indication on the 90° aligner pin keeper was observed and evaluated as acceptable (indication not on load bearing portion of assembly). Spring 1997 VT-I VT-I re-examination of Top Guide Aligner Pin (REI 7) located at 90° in accordance with SIL 588, R 1. Indication on aligner pin keeper did not appear to change in size. Spring 2000 VT-1 VT-1 of two (2) hold down assemblies. No (REl9) indications. Fall2001 VT-1 VT-1 of two (2) horizontal aligner pins with no (RE20) new indications. VT-1 of four (4) hold down assemblies. EVT-1 EVT- 1 examinations of accessible areas of the Rim weld.

NLS2017071 Attachment Page 31 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Fall 2006 VT-1 VT-1 on two (2) hold down assemblies and aligner (RE23) pin assemblies at 90° and 270°. A previous indication identified on the non-load bearing keeper of the aligner pin assembly at the 90° location was observed with no apparent change. However, two (2) new but similar type indications were also observed on the same keeper. Three (3) new indications were observed on the non-load bearing aligner pin keeper at the 270° location. Indications were evaluated as acceptable. VT-3 VT-3 examinations performed on accessible areas of top guide per B-N-2 . No indications. Spring 2008 VT-I VT-1 examinations performed on hold down and (RE24) aligner assemblies at 0 and 180°. One (1) new indication identified on non-structural keeper at 180°. Similar to indications in keepers seen at 90° and 270°. Evaluated as acceptable. EVT- 1 EVT-1 examinations of accessible areas of rim weld. VT-3 VT-3 examinations performed of accessible top guide hold down assemblies, rim pins per B-N-2 . Fall 2009 VT-I VT-1 examinations performed on hold down and (RE25) aligner assemblies at 90°. No change in the indication at the 90° aligner pin keeper. EVT-1 EVT-1 examinations of 10% or fourteen (14) of top guide grid beams per BWRVIP-183 . No indications. However, only eight (8) were credited as quality examinations. VT-3 VT-3 examinations of accessible areas of top guide per B-N-2. Spring 2011 VT-3 VT-3 of accessible areas of Top Guide per B-N-2 . (RE26) No indications. VT- 1 VT-1 for BWRVIP-26 credit was performed on the Hold Down assemblies and Aligner Pin assemblies at 270°. An indication not previously reported was observed adjacent to the attachment weld adjoining the Aligner Block to the Top Guide. Indication appears to be a manufacturing remnant that was not completely removed during

NLS2017071 Attachment Page 32 of 68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVlP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection construction. Previously identified indications were also observed with no changes. Scope was expanded to include the remaining other three (3) Aligner Pin assemblies located at 0°, 90°, and 180°. VT-3 for Sect XI B-N-2 and VT- 1 for BWRVIP-26 credit was performed. Aligner Pin assembly at 0° was found to have seven (7) previously unidentified indications, with four (4) identified in the Aligner Pin Keeper and three (3) identified in the Aligner Block. Review of previous inspection video showed faint presence of indications. Evaluated as acceptable. Aligner Pin assembly at 90° was found to have one (1) previously unidentified indication located on the Aligner Pin Keeper. Review of previous inspection video showed a faint presence of the indication. Three (3) previously identified indications were also observed with no changes. Evaluated as acceptable. Aligner Pin assembly at 180° was fo und to have two (2) previously unidentified indications located on the Aligner Pin Keeper. Review of previous inspection video shows presence of indications. Three (3) previously identified indications were also observed with no changes. Evaluated as acceptable. EVT-1 EVT-1 examinations of accessible areas of the Rim weld. EVT- 1 of two (2) top guide cell locations per BWRVIP- 183 . No indications. Fall 2012 VT- 1 VT- I examinations performed on hold down (RE27) assembly at 180°. No indications. VT-I VT-1 of the aligner pin assembly at 0° was performed to confirm seven (7) flaws identified in RE26 . Four (4) of the flaws on the keeper were confirmed and verified to have no changes. One (!)flaw on the aligner pin block was confirmed and verified to have no changes. The two (2) other previously identified flaws on the block were determined to be non-relevant surface scratches.

NLS2017071 Attachment Page 33 of 68 10 CFR 50.55a Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection VT-1 of the aligner pin assembly at 90° was performed to confirm seven (7) flaws identified on the keeper in RE26. 4 of the flaws were confirmed and verified to have no changes. One ( 1) additional flaw on the keeper was also reported. This flaw is similar to flaws seen on the other aligner pin keepers, but could not be verified in previous video due to camera positioning. VT-! of the aligner pin assembly at 180° was performed to confirm three (3) flaws identified in RE26. All of the flaws on the keeper were confirmed and verified to have no changes. VT-1 of the aligner pin assembly at 270° was performed to confirm four (4) flaws identified in RE26. Three (3) of the flaws on the keeper were confirmed and verified to have no changes. One ( l) previously reported flaw adjacent the aligner block to top guide weld was examined using an improved camera and delivery mechanism and determined to be a non-relevant surface scratch. Fa112014 EVT-1 EVT- 1 examinations of accessible areas of Rim (RE28) weld. No indications. VT-I VT-1 examinations performed on accessible top portion of the top guide hold down assembly at 0°. No indications. VT-1 of the aligner pin assembly at 0° was performed to confirm five (5) previously identified flaws. Four (4) flaws on the keeper were confirmed to have no changes. The identified flaw on the aligner pin block showed slight increase in length. VT- I of the aligner pin assembly at 90° was performed to confirm five (5) previously identified flaws. The 5 flaws on the keeper were confirmed to have no changes. Five (5) unreported flaws on the aligner block were detected. VT- I of the aligner pin assembly at 180° was performed to confirm three (3) previously identified flaws. The three flaws on the Keeper were confirmed to have no change. Five (5)

NLS20 17071 Attachment Page 34of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection umeported flaws on the aligner block and two (2) on the top guide were detected. VT-1 of the aligner pin assembly at 270° was performed to confirm three (3) previously identified flaws . The three (3) flaws on the keeper were confirmed and verified to have no changes. One (1) umeported flaw on the aligner block and one (I) on the top guide were detected. EVT- 1 of beams near impact site of dropped control rod blade. (Ref OE 3 13327). No crack indications identified. Fa112016 EVT-1 EVT-1 of five (5) of top guide grid beams per (RE29) BWRVIP-183 . No indications. EVT- 1 of beams near impact site of dropped control rod blade. (Ref OE 3 13327). No indications identified. VT-I VT- I of accessible top portion of the TG hold down assembly at 90°. No indications. VT- 1/EVT-l VT-1 and EVT-1 of aligner pin assembly at 0° confirmed five (5) previously identified flaws. All flaws showed no changes, except for Flaw 6 on the keeper which appeared longer compared to previous examination but difference was attributed to improved tooling and lighting. VT-I and EVT-1 of aligner pin assembly at 90° confirmed ten (10) previously identified flaws on the keeper and block. All flaws showed no changes except for, Flaw 5 (keeper) and 10 (block) which appeared longer compared to previous exam with difference attributed to improved tooling an lighting. VT- I and EVT-1 of aligner pin assembly at 180° confirmed nine (9) previously identified flaws . All flaws showed no changes, except for Flaws 2 and 3 on keeper which appear longer compared to previous exam with difference attributed to improved tooling and lighting. One(!) new flaw was reported on the block to slider weld that was later confirmed to be present in the previous exam. Change attributed to improved tooling and lighting.

NLS2017071 Attachment Page 35of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection VT-1 and EVT- 1 of aligner pin assembly at 270° confirmed five (5) previously identified flaws that showed no changes. One (I) new flaw reported on keeper to washer weld that was later confirmed be present in the previous exam. Core Plate (Rim, etc.) Fall 1995 VT-3 VT-3 examinations of hold down bolts examined in 1995 per SIL 588 . No indications. Spring 2000 VT-3* VT-3 examinations of 48 bolts examined from top (RE 19) side.

                                                    *(Bolts are not accessible for EVT- 1)

Fall 2001 to VT-3 VT-3 examinations performed on accessible areas Fall 2009 (RE20 per B-N-2. No indications.

                        -RE26)

Fall 2012 VT-3 VT-3 examination of three (3) hold down bolt (RE27) locations (70, 71, and 72) from the top side. No indications. Fall2014 VT-3 VT-3 exam of36 (50%) hold down bolt locations from the top side. No indications. SLC 1986-2001 VT-2 VT-2 examinations ofSLC penetration during Class l RPV pressure test each outage. Spring 2003 EVT-2 Enhanced VT-2 examinations during Class 1 (RE2 1) pressure test. No indications. Spring 2005 EVT-2/UT Enhanced VT-2 performed of safe-end and (RE22) penetration in conjunction with ASME Section XI Class I pressure test. Manual UT to Appendix VIII performed on nozzle to safe-end weld. No indications. Fall 2006 EVT-2 Enhanced VT-2 examinations of safe-end and (RE23) penetration performed in conjunction with ASME Section XI Class I system leakage test. No indications. Spring 2008 EVT-2 Enhanced VT-2 examinations performed of safe-(RE24) end and penetration in conjunction with ASME Section XI Class I system leakage test. No indications.

NLS2017071 Attachment Page 36of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in B WRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Fall 2009 EVT-2 Enhanced VT-2 examinations of safe-end and (RE25) penetration performed in conjunction with ASME Section XI Class I system leakage test. No indications. Spring 2011 EVT-2 Enhanced VT-2 examinations of safe-end and (RE26) penetration performed in conjunction with ASME Section XI Class I system leakage test. No indications. UT examination ofNIO SLC nozzle to safe-end Fall 2012 UT per Risk-Informed ISI Program and Appendix (RE27) VIII. No indications. Enhanced VT-2 examinations of safe-end and EVT-2 penetration performed in conjunction with ASME Section XI Class I system leakage test. No indications. Enhanced VT-2 examinations of safe-end and Fall2014 EVT-2 penetration performed in conjunction with ASME (RE28) Section XI Class I system leakage test. No indications. Jet Pump Assembly 1986-1995 VT-l N T-3/ VT examinations on ten (10) Jet Pumps each UT outage. Exam includes applicable GE SILS. Jet pump beams replaced in 1985. Jet pump beam UT first performed in 1993. Spring 1997 VT-INT-3 Ten (10) jet pumps VT examined. Exam includes (REI7) applicable GE SILs. No indications. Fall 1998 VT - lNT-3 Ten (10) jet pumps VT examined. Exam includes (RE18) applicable GE SILs. No indications. Spring 2000 NI A Examinations deferred to Fall 2001. (REl9) Fall2001 VT-3 VT-3 examinations on all 20 jet pump nozzle (RE20) inlets per SIL 465 . No indications. VT-I VT- I examinations on all WD-1 's. No indications. EVT-1 EVT- 1 examinations on BB-I and BB-2 on JP 's 1-

10. EVT-1 on MX-2's on JP ' s I - 10. EVT-1 on RB-1 ' sand RB-2 ' s on JP ' s 1/2, 3/4, and 5/6. No indications. EVT- 1 on RS-1 ' s, RS-2' s, and RS-3 ' s on JP 's 1 - 10. EVT-1 on RS-6 ' s on JP's 1, 3, and

NLS2017071 Attachment Page 37 of 68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection

5. EVT-1 on RS-7's on JP ' s 2, 4, and 6. EVT-1 on RS-S's and RS-9 ' s on JP ' s 1/2, 3/4, and 5/6.

No indications. Spring 2003 VT-3 VT-3 examinations on the JP nozzle inlet mixers (RE21) on JP's 11 - 20 per SIL 465. VT-3 examinations of set screws, gaps, and tack welds on JP 's I - 20 per SIL 574. No indications. EVT-1 EVT- 1 examinations on the IN-4 on JP ' s 5, 6, 11, 12, 13, and 14. EVT-1 examinations on the MX-2 on JP's 11 , 12, 13, and 14. EVT-1 examinations on the RB-1 ' sand RB-2's, on JP's 11/12 and 13/14. EVT-1 examinations on RS-1 and RS-2 on JP ' s 11 / 12, 13/14, 15/ 16, and 17/ lS; RS-6 on JP ' s 11and13; RS-7's on JP's 12 and 14; RS-S 's and RS-9's on JP's 11/12 13/ 14. No indications. UT UT examinations on the BB-1 ' sand BB-2's for JP ' s I - 20. No indications. Spring 2005 VT-3 VT-3 on the JP nozzle inlet mixers on JP s 1 - 10 (RE22) per SIL 465 . No indications. VT-1 VT-1 examinations on JP set screws, gaps and tack welds on JP ' s 1, 2, 15, and 16 per SIL 574. No indications. EVT-1 EVT-1 examinations on RS- I, RS-2 , and RS-3 welds on JP ' s 1 and 2 and the IN-4 welds on JP ' s 7, S, 9, and 10. No indications. Fall 2006 VT-1 VT-1 per SIL 574 of adjustment screw and gap (RE23) and tack welds on JPs 9 10. VT-1 ofWD-1 at JP's 9 I 0. No indications. EVT-1 EVT-1 ofRS-1 and RS-2 on JP ' s 15/1 6 and 19/20. Spring 200S EVT-1 EVT- 1 examinations ofIN-4's at JP's 19 and 20. (RE24) EVT-1 examinations ofRB-la' s, -lb' s, -!e's, and

                                               -Id' s between JP ' s 9/ 10 and 19/20. EVT-1 examinations ofRB-2a's, -2b's, -2c' s, and -2d' s between JP ' s 9/10 and 19/20. EVT-1 examination ofRS-3 between JP's 19/20. EVT-1 examinations ofRS-6 at JP's 9 and 19. EVT-1 examination of RS-7 at JP ' s 10 and 20. EVT-1 examinations of RS-Sand RS-9 at JP ' s 19/20 and 9/ 10. No indications.

NLS2017071 Attachment Page 38of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection UT UT of BB-I , -2 and -3 on all 20 JP beams. No indications. UT ofMX-2 (and AD-I , AD-2 , DF-1 , DF-2, DF-3 note in Diffuser Section) on all 20 jet pumps. Fall 2009 VT-3 VT-3 of JP nozzle inlets per SIL 465 on JP ' s 15, (RE25) 16, 17 and 18. No indications. VT-I VT-1 per SIL 574 of adjustment screw and gap andtackweldsonJPs 10, 15, 16, 19, and20. VT-1 ofWD-1 at JP 's 17 and 18. No indications. EVT- 1 EVT-1examinations ofIN-4onJP's 15, 16, 17, and 18. EVT-1 examinations ofRB-1 ' sand RB-2' sonJP' s7/8, 15/16, and 17/1 8. EVT-1 examinations on RS-1 's and RS-2 's on JP ' s 1111 2 and 1711 8. EVT-1 examinations on RS-3 ' s on JP 's 11112, 15/16, and 17/ 18. EVT-1 examinations on RS-6 ' s on JP ' s 7, 15, and 17 and RS-7's on JP ' s 8, 16, and 18. EVT-1 examinations on RS-8 ' s and RS-9 ' s on JP ' s 7/8, 15/ 16, and 17/1 8. No indications. Spring 2011 VT-3 VT-3 of JP nozzle inlets on JP 9 and 10. No (RE26) indications. VT-1 VT-1 of the JP Restrainer Wedge (WD-1) at JP- I thru JP-20 . No indications of movement or wear observed. EVT-1 EVT-1 ofRS-8 and RS-9 on JP-1 thru JP-14, JP-19, and JP-20. No indications. EVT-1 of JP-9 and JP-lO's IN-4, RB-la, RB-lb, RB-le, RB-ld, RB-2a, RB-2b, RB-2c, RB-2d, RS-3, RS-1 RS-2. EVT-1 of JP-9's RS-6 and JP-lO's RS-7. No indications. EVT-1 of JP-7 and JP-8's RS-3 . No indications. Fall2012 EVT-1 EVT-1 ofJP-13 and JP-14's IN-4, RB-la, RB-lb, (RE27) RB- le, RB- Id, RB-2a, RB-2b, RB-2c, RB-2d, RS-I, RS-2 , & RS-3. EVT-1 ofRS-6 on JP-13 and RS-7 on JP-14. No indications. EVT-1 ofJP-7 and JP-8's RS-3. No indications. EVT-1 ofJP-15 and 16's RS-3. No indications.

NLS2017071 Attachment Page 39 of 68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection VT-1 VT-1 of the JP Restrainer Wedge (WD-1) at JP-1 , 2, 9, 10, 13, 14, 15, 16, 19, & 20. No indications of movement or wear observed. VT- 1 of the JP-15 set screw gaps and slip joint. Previously identified shroud side gap was found to have an increase of0.003" with no signs of movement. Vessel side set screw confirmed to have partial contact. No indications on slip joint. VT-1 of the JP-20 set screw gaps and slip joint. Previously identified shroud side gap was found to have an increase of .004" with no signs of movement. Newly reported vessel side set screw gap measured to be .013" . No indications on slip joint. VT-3 VT-3 of JP nozzle inlets on JP-13 and 14. No indications. Fall 2014 EVT-1 EVT-1 ofMX-2 on Jet Pumps 1, 2, 8, 9, & 10. No (RE28) Indications. EVT-1 of JP-1 and JP-2's IN-4, RB-1 a, RB-1 b, RB-le, RB-ld, RB-2a, RB-2b, RB-2c, RB-2d, RS-1, RS-2 , RS-3 , RS-6 on JP-1 and RS-7 on JP-2. No indications. VT-1 VT-1 of the JP Restrainer Wedge (WD-1) at JP-1 thru JP-20 . No indications of movement or wear observed. VT-1 of the set screws and auxiliary wedges on Jet Pump 1, 2, 9, &10. JPlO had a previously identified shroud side gap that was found to be 0.011" with no signs of movement. JPs 1, 2, and 9 set screws and aux wedges were found to be in full contact with no signs of wear. VT-3 VT-3 of JP nozzle inlets on JP-I and 2. No indications. Fall 2016 EVT-1 EVT- 1 ofMX-2 and IN-4 on JP-6. No (RE29) indications. EVT-1 ofRS-1 , RS-2, and RS-3 on JP-3&JP-4 and RS-1 and RS-2 on JP-7&JP-8 . No indications. EVT-1 ofJP-5&JP-6's RB-la, RB-lb, RB-le, RB-ld, RB-2a, RB-2b, RB-2c, RB-2d, RS-1 , RS-2,

NLS2017071 Attachment Page 40of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection RS-3 , RS-6, RS-7, RS-8 , and RS-9 . No indications. VT-! VT-1 of set screw gaps and slip joints on JP-15 and JP-20 . No indications. VT-1 of the JP Restrainer Wedge (WD-1) at JP-5 and JP-6. No indications. VT-3 VT-3 of JP nozzle inlets on JP-6. No indications. Jet Pump Diffuser 1986-1998 VT-3 10 Jet Pumps VT-3 examined each outage. No indications. No indications. Spring 1997 VT-lNT-3 Ten jet pumps VT examined. Exam includes (REI 7) applicable GE SILs. No indications. Fall 1998 VT-lNT-3 VT examinations on ten (10) jet pumps. Exam (RE18) includes applicable GE SILs. No indications. Spring 2000 NIA Exams deferred to Fall 2001 . (RE19) Fall 2001 EVT-1 EVT-1 examinations on ten (10) jet pumps (5 (RE20) assemblies). Identified an indication thought to be a broken jet pump sensing line upper bracket retaining weld. Evaluated as acceptable. Spring 2003 VT-3 VT-3 on JP sensing lines for all jet pumps per SIL (RE21) 420. No indications. VT-1 VT-1 on sensing line brackets for all jet pumps per SIL 420. Previously reported cracked bracket weld was determined not to be cracked. No indications. EVT-1 EVT-1 examinations of AD- I, AD-2 , AD-3a, AD-3b welds on JP ' s 11 through 20. No indications. Spring 2005 VT-3 VT-3 on JP sensing lines for JP 's 1 - 11 and 14 per (RE22) SIL 420. No indications. VT-1 VT-1 on JP sensing line brackets for JP ' s 1- 11 and 14. No indications. Fall 2006 EVT-1 EVT- 1 on AD-1 on JP ' s l, 2, and 5. EVT-1 (RE23) examinations on AD-2, AD-3a, AD-3b, DF-1 on

NLS2017071 Attachment Page 41of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection JP-15 , 16, 17, 18, 19, and20andDF-2onJP 's 15, 16, 19, and 20. No indications. Spring 2008 UT UT on AD-1 , AD-2 , DF-1 , DF-2, and DF-3 (and (RE24) MX-2). One (1) indication on DF-1 at JP-14. EVT-1 EVT-1 examinations on DF- 1 at JP-14 in addition to UT. Appeared to be a defect from original construction. Fall 2009 EVT-1 EVT- 1 examinations of indication to DF-1 on JP-(RE25) 14 identified during the previous outage. No change. Spring 2011 EVT-1 EVT- 1 re-examination of indication located on the (RE26) inside surface of JP-14 at the DF-1 weld. Indication was found to have no changes. Fall2012 EVT-1 EVT- 1 re-examination of indication located on the (RE27) inside surface of JP-14 at the DF-1 weld. Indication did not change. Fall2014 EVT-1 EVT- 1 of AD-1 , AD-2 , AD-3a, AD-3b, DF-1 , DF-(RE28) 2 on Jet Pumps 1, 2, 8, 9, & 10. No Indications. EVT- 1 re-examination of indication located on the inside surface of JP-14 at the DF-1 weld. Indication did not change. Fall 2016 EVT-1 EVT- 1 of AD-I , AD-2, AD-3a, AD-3b, DF-2 on (RE29) Jet Pumps 3, 4, 5, 6, & 7. No indications. EVT-1 ofDF-1 (outside diameter) on Jet Pumps 6 and 14. No indications. EVT-1 ofDF-1 (inside diameter) for re-examination of indication on the inside surface of JP-14. Indication showed no change. CRD Guide Tube Fall 1995 VT-3 VT-3 exams of accessible guide tubes. No indications. Spring 1997 VT-3 VT-3 exams of accessible guide tubes. No (RE17) indications. Fall 1998 VT-3 VT-3 exams of accessible guide tubes. No (RE I8) indications.

NLS2017071 Attachment Page 42 of68 10 CFR 50.55a Request No. RIS-02, Revision 1 (continued) Implementation ofBWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Spring 2000 VT-3 VT-3 examinations of eighteen {18) anti-rotation (RE19) pins and eleven ( 11) CRGT-1 welds. No indications. EVT-1 EVT-1 examinations of four (4) CRGT-2 and CRGT-3 welds. No indications. Fall 2001 VT-3 VT-3 examinations of thirteen {13) anti-rotation {RE20) pins and thirteen (13) CRGT-1 welds. No indications. EVT-1 EVT-1 examinations of five (5) CRGT-2 and CRGT-3 welds. No indications. Spring 2005 EVT-1 EVT- 1 examinations ofone (1) CRGT-2 weld and (RE22) one (1) CRGT-3 weld. No indications. Fall 2006 EVT-1 EVT-1 examinations ofone (1) CRGT-2 weld and (RE23) one (1) CRGT-3 weld. No indications. Spring 2008 EVT-1 EVT-1 examinations of two (2) CRGT-2 welds (RE24) and three (3) CRGT-3 welds. No indications. Fall 2009 EVT-1 EVT-1 examinations on one (1) CRGT-2 weld and (RE25) two (2) CRGT-3 welds. No indications. CRD Stub Tube NIA NI A No record of examination. In-core Housing NA NA No record of examination back to 1996. Dry Tube 1989-1991 VT VT exam in 1989, 1990,and 1991 perSIL409Rl. All dry tubes replaced in 1993. Spring 2005 VT Replaced one {l) dry tube. (RE22) Fall 2012 VT-I VT- I was performed on dry tube locations at 12-(RE27) 09 and 28-25 . No indications. Fall 2014 VT-I VT-1 performed on IRM dry tube locations at 20-(RE28) 25 and 36-41. No indication observed. Replaced IRM dry tube at 12-41 . Fall 2016 VT-1 VT-I on Dry Tube Locations. IRM 28-33 & 36-(RE29) 09 and SRM 12-33, 20-17, & 36-25 . No indications.

NLS2017071 Attachment Page 43of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Instrument Penetrations 1986-2000 VT-2 VT-2 examination performed during RPV system leakage test each outage for all six (6) instrument nozzle penetrations. No indications. Spring 2000 PT PT examination ofN16A instrument penetration (RE19) nozzle to safe-end weld. Fall2001 VT-2 VT-2 examination performed during RPV system (RE20) leakage test. No indications. Spring 2003 VT-2 VT-2 examination performed during RPV system (RE21) leakage test. No indications. Spring 2005 VT-2 VT-2 examination performed during RPV system (RE22) leakage test. No indications. UT UT examination ofN16B nozzle to safe-end per Risk-Informed ISI Program and Appendix VIII. No indications. Fall 2006 VT-2 VT-2 examination performed during RPV system (RE23) leakage test. No indications. Spring 2008 VT-2 VT-2 examination performed during RPV system (RE24) leakage test. No indications. Fall 2009 VT-2 VT-2 examination performed during RPV system (RE25) leakage test. No indications. Spring 2011 VT-2 VT-2 examination performed during RPV system (RE26) leakage test. No indications. Fall 2012 VT-2 VT-2 examination performed during RPV system (RE27) leakage test. No indications. Fall2014 UT UT examination ofN16B nozzle to safe-end per (RE28) Risk-Informed ISI Program and Appendix VIII. No indications. VT-2 VT-2 examination performed during RPV system leakage test. No indications. Fall 2016 VT-2 VT-2 examination performed during RPV system (RE29) leakage test. No indications. Vessel ID Brackets 1986-1995 VT-INT-3 ASME XI VT-3 (non-beltline) and VT-1 (beltline examinations) of jet pump riser brace, dryer, FW Sparger, Core Spray, guide rod, and surveillance

NLS2017071 Attachment Page 44 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu ofB-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection capsule holder brackets performed once per interval. No indications. Spring 1997 VT-l N T-3 VT- l N T-3 ASME Section XI examinations on (RE17) five (5) jet pump riser brackets, FW brackets and welds examined. No indications. Fall 1998 VT-l N T-3 VT- l NT-3 ASME Section XI examinations on (RE 18) five (5) jet pump riser brackets, FW brackets and welds examined. No indications. EVT-1 EVT- 1 examinations on fo ur (4) CS bracket attachment welds. No indications. Spring 2000 VT-3 VT-3 examinations of guide rod attachment welds. (RE19) No indications. VT- 1 VT- 1 on FW sparger brackets. No indications. EVT-1 EVT- 1 examinations on CS bracket attachment welds. No indications. Fall2001 EVT-1 EVT-1 examinations on all FW sparger bracket (RE20) attachment welds and all dryer support attachment welds. No indications. Spring 2003 EVT-1 EVT-1 examination of on JP riser brace pad (RE2 1) attachment weld at 150°. No indications. Spring 2005 VT-3 VT-3 examination of steam dryer hold down (RE22) brackets. Fall 2006 EVT-1 EVT- 1 of eight (8) FW sparger brackets and four (RE23) (4) CS piping bracket attachment welds. No indications. Spring 2008 VT-3 VT-3 of guide rod attachment welds. No (RE24) indications. EVT-1 EVT- 1 examinations of JP riser brace pad attachment welds at 30°, 150°, 210°, 270°, and 330°. EVT-1 examinations of steam dryer support bracket attachment welds at 2 15° and 325°. No indications. Fall 2009 EVT-1 EVT-1 examinations of JP riser brace pad (RE25) attachment welds at 60°, 90°, and 120°. No indications.

NLS2017071 Attachment Page 45 of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu ofB-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Spring 2011 EVT-1 EVT-1 of the JP riser brace pad attachment welds, (RE26) JP-RBPAD-ATTWLDS @ 30°. No Indications. Fall 2012 EVT-1 EVT-I of four (4) CS piping bracket attachment (RE27) welds at 30°, I50°, 210°, and 330°. No indications. EVT- I NT- I of JP riser brace pad attachment welds at 270°. No indications. EVT-INT-3 of steam dryer support bracket attachment welds at 2 I 5° and 325°. No indications. VT-I VT- I of surveillance capsule holder brackets at 300°. No indications. VT-3 VT-3 (direct) examination of steam dryer hold down brackets@ 35°, I45°, 215°, and 325°. No indications. Fall 20I4 EVT-I EVT-1 of Riser Brace attachment welds on JP-1 & (RE28) 2 at I50°. No indications. EVT-INT-3 EVT- l N T-3 of steam dryer support bracket attachment welds at 145° and 35°. No indications. VT-I VT- I of surveillance capsule holder brackets at 30° & I20°. No indications. VT-3 VT-3 (direct) examination of steam dryer hold down brackets@35°, 145°, 215°, and 325°. No indications. Fal120 I6 EVT-1 EVT-1 of the JP riser brace pad attachment welds, (RE29) JP-RBPAD-ATTWLDS Lal 90°. No indications. LPCI Coupling NIA NIA Not applicable to this plant. Steam Dryer Fall 200I VT-1 VT- I of twenty four (24) drain channel welds per (RE20) SIL 474. Spring 2003 EVT-I EVT- 1 of twenty four (24) drain channel welds per (RE2 1) SIL 474. Spring 2005 VT-1 VT- 1 of leveling screws per OE 16110. (RE22)

NLS2017071 Attachment Page 46of68 10 CFR 50.SSa Request No. RIS-02, Revision 1 (continued) Implementation of BWRVIP Documents in Lieu of B-N-1 and B-N-2 Component in BWRVIP Date of Inspection Summarize the Following Information: Inspection Scope Frequency of Method Used Results, Repairs, Replacements, Re-inspections Inspection Fall 2006 VT-I Performed baseline VT- I examinations to (RE23) w/Character BWRVIP-I39 and SIL 644, Rev 2. Re-examined Card five (5) minor indications previously identified per SIL 474 adjacent to several drain channels. Two (2) new indications were observed in a weld adjacent to a drain channel and both tack welds on one (I) lifting lug were observed. The indications were evaluated as acceptable. Fall 2009 VT-I(89) VT- I examinations on seven (7) previously (RE25) identified indications on dryer. With additional cleaning, six ( 6) of the indications disappeared with only one (1) remaining (i.e. , the cracked tack welds on one (1) lifting lug - no change in the lifting lug). Fall 20I6 VT-I(89) Performed VT-I examinations to BWRVIP-I39-(RE29) A. Re-examined one (I) previous indication of cracked tack welds on lifting lug @ 145° with no changes. The other three (3) lifting lugs were observed to have similar cracked tacks that were not previously reported. Five (5) Tie Rods were reported to be slightly bent with no broken welds. One (I) additional indication was observed on the lower guide bracket weld at I 80° that extended into the skirt. The indication was arrested with a 5/8th inch Stop-drill hole. No other dryer indications reported. Dissimilar metal welds Spring 2008 UT Automated UT performed on four (4) CAT A (RE24) welds per Appendix VIII. Manual UT performed on two (2) CAT A welds. All welds included in Risk-Informed ISI Program. No indications. Spring 2011 UT Manual UT inspection performed on one (1) CAT (RE26) D nozzle to cap weld (CRD Return) per Appendix VIII and Risk-Informed ISI Program. No indications. Fall2014 UT Manual UT inspection performed on three (3). (RE28) CAT A welds per Appendix VIII. All welds included in Risk-Informed ISI Program. No indications. Fall 20I6 UT Automated Phased Array UT performed on one (I) (RE29) CAT D weld and manual UT performed on one (I) CAT A weld. Both welds included in Risk-Informed ISi Program. No indications.

NLS2017071 Attachment Page 47of68 10 CFR 50.55a Request No. RIS-03 Implementation of Code Case N-702 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(l) Acceptable Level of Quality and Safety ASME Code Component(s) Affected Code Class: ASME Section XI Code Class 1 Component Numbers: Various (see Table 1 for detailed list of components) Code

References:

ASME Section XI, 2007 Edition with 2008 Addenda Code Case N-702 Examination Category: 8-D Item Number(s): 83 .90 and 83 .100

Applicable Code Edition and Addenda

ASME Section XI, 2007 Edition through the 2008 Addenda Applicable ASME Code Requirements Table IWB-2500-1 , Examination Category 8 -D, "Full Penetration Welded Nozzles in Vessels" requires a volumetric examination of all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles each 10-year interval. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented, as required and conditioned by 10 CFR 50.55a(b )(2)(xv). The RPV nozzle-to-vessel welds and inner radii subject to this request are listed below in Table 1: TABLE 1 Identification Description Total Number Minimum Number Number to be examined Nl Recirculation Outlet 2 1 N2 Recirculation Inlet 10 3 N3 Main Steam Outlet 4 1 NS Core Spray 2 1 N6 Head Instrument 2 1 N7 Head Vent 1 1 N8 Jet Pump Instrumentation 2 1

NLS2017071 Attachment Page 48of68 10 CFR 50.SSa Request No. RIS-03 (continued) Implementation of Code Case N-702

Reason for Request

NRC Regulatory Guide 1.147 Rev. 17 conditionally accepts the use of Code Case N-702. This code case provides an alternative to performing examination of 100% of the Nozzle-to-Vessel Welds and Inner Radii for Examination Category B-D nozzles with the exception of the Feedwater and CRDRL Nozzles. The alternative is to perform examination of a minimum of 25% of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size, excluding the Feedwater and CRDRL Nozzles. Proposed Alternative and Basis for Use Proposed Alternative Pursuant to 10 CFR 50.55a(z)(l), NPPD requests approval to implement the alternative of Code Case N-702 in lieu of the code required 100% examination of all nozzles identified in Table 1. As an alternative, for the nozzle-to-shell welds and inner radii identified in Table 1, NPPD proposes to examine a minimum of25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702. Basis for Use BWRVIP has issued two topical reports:

BWRVIP-108, "Technical Basis for the Reduction oflnspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1003557, dated October 2002 (ML023330203).
BWRVIP-241, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1021005, dated October 2010 (ML11119A041).

The BWRVIP-108 report contains the technical basis supporting ASME Code Case N-702 "Alternative Requirements for BWR Nozzle Inner Radius and Nozzle-to-Shell Welds" for reducing the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radius areas from 100% to 25% of the nozzles for each nozzle type during each 10-year interval. BWRVIP-241 provides supplemental analyses for BWR RPV recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii. BWRVIP-241 was submitted to address the limitations and conditions specified in the December 19, 2007, safety evaluation for the BWRVIP-108 report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of

NLS2017071 Attachment Page 49 of68 10 CFR 50.55a Request No. Rl5-03 (continued) Implementation of Code Case N-702 Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii." Based on the two evaluations (BWRVIP-241 and BWRVIP-108), the failure probabilities due to a LTOP event at the nozzle blend radius region and the nozzle-to-vessel shell welds for CNS recirculation inlet and outlet nozzles are very low and meet the NRC safety goal. RG 1.14 7, Revision 17 conditionally accepts the use of Code Case N-702 with the following condition "The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 ofNRC Safety Evaluation regarding BWRVIP-108 dated December 18, 2007 (ML073600374) or Section 5.0 ofNRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (ML13071A240) are met." Section 5.0 of the NRC Safety Evaluation for BWRVIP-241 states:

   "Licensees who plan to request relief from the ASME Code, Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-241 report as the technical basis for use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by demonstrating all of the following:"

(1) The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour CNS Technical Specifications Surveillance Requirement 3.4.9.1, Reactor Coolant System heatup and cooldown rates are limited to a maximum of 100°F when averaged over any one hour period and thus meets the requirement of Condition 1. Note: Inputs used in Conditions 2 through 5 representing the CNS configuration are in bold text. Recirculation inlet nozzles (N2) (2) (pr/t)/CRPv _:: : 1.15 p = RPV normal operating pressure (psi) (1020 psig per CNS Technical Specifications 3.4.10 for Reactor Steam Dome Pressure) r = RPV inner radius (inch) (110.375) t = RPV wall thickness (inch) (6.875) CRPY= 19332 (based on the BWRVIP-108 recirculation inlet nozzle/RPV FEM model);

NLS2017071 Attachment Page 50of68 10 CFR 50.55a Request No. RI5-03 (continued) Implementation of Code Case N-702 CNS specific calculations for Condition 2 above: (1020 x 110.375)/6.875)/19332 = 0.85 ~ 1.15 The CNS result is 0.85 and thus meets the requirement of Condition 2 to be~ 1.15. P = RPV normal operating pressure (psi) (1020) r0 = nozzle outer radius (inch) (10.219) ri = nozzle inner radius (inch), (6.188) C OZZLE = 1637 based on the BWRVIP-108 recirculation inlet nozzle/RPV FEM model); CNS specific calculations for Condition 3 above: [1020(10.219 2 + 6.188 2)/(10.219 2 - 6.188 2))/1637 = 1.34 ~ 1.47 The CNS result is 1.34 and thus meets the requirement of Condition 3 to be~ 1.47 Recirculation outlet nozzles (Nl) (4) (pr/t)/CRPv :S 1.15 p = RPV normal operating pressure (psi) (1020) r = RPV inner radius (inch) (110.375) t = RPV wall thickness (inch) (6.875) CRPv = 16171 (based on the BWRVIP-108 recirculation outlet nozzle/RPV FEM model); CNS specific calculations for Condition 4 above: (1020 x 110.375)/6.875)/16171 = 1.013 ~ 1.15 The CNS result is 1.013 and thus meets the requirements of Condition 4 to be~ 1.15 P = RPV normal operating pressure (psi) (1020) r0 = nozzle outer radius (inch) (21.656) ri = nozzle inner radius (inch) (12.875) CNoZZLE = 1977 (based on the BWRVIP-108 recirculation outlet nozzle/RPV FEM model).

NLS2017071 Attachment Page 51 of 68 10 CFR 50.55a Request No. RI5-03 (continued) Implementation of Code Case N-702 CNS specific calculation for Condition 5 above: [1020(21.656 + 12.875 2)/(21.656 2 - 12.875 2))/1977 = 1.08 ~ 1.59 2 The CNS result is 1.08 and thus meets the requirements of Condition 5 to be ~ 1.59 The analyses for the N2 nozzles in BWRVIP-108 and BWRVIP-241 are based on the assumption that fluence at the nozzles is negligible because the analysis is for the initial 40 years of plant operation and do not address the extended operating period. Based on analysis performed in support of license renewal for CNS, the beltline was re-evaluated for 60 years based on the axial flux profile and the active fuel and nozzle elevations. In 4.2. l "Reactor Vessel Fluence" of the License Renewal Application, it is documented that fluence at the recirculation inlet nozzles (the closest ferritic nozzles to the beltline) will not exceed IE+ 17 n/cm2 during the period of extended operation. Since the LRA, CNS has updated the fluence values using the NRC approved RAMA fluence method in support of the current Pressure Temperature Limits Report. As part of that evaluation, the predicted fluence at 54 EFPY was also determined for the N2 nozzles which support the conclusion of the LRA that the fluence at the recirculation inlet nozzles will not exceed IE+ 17 n/cm2 . The plates and welds in the beltline remain the limiting materials for the period of extended operation. Therefore, the fluence assumptions used in BWRVIP-108 and BWRVIP-241 remain valid and are applicable to CNS. The analyses in BWRVIP-108 and BWRVIP-241 were based on predicted fatigue crack growth over the initial licensed operating period and assumed additional fatigue cycles in evaluating fatigue crack growth. CNS is projected to exceed the total number of thermal cycles used in the BWRVIP analysis during the extended operating period. However, the usage factor for the N2 nozzles is expected to remain below 1.0. Previous BWRVIP documents have demonstrated that sec crack growth represents the majority of the crack growth and that crack growth due to additional mechanical/thermal fatigue cycles introduced by the extended operation time is insignificant compared to hypothetical SCC growth. Thus, the amount of thermal cycle driven fatigue crack growth due to the extended operation to 60 years is not a controlling factor in the probability of failure of the BWR reactor vessel nozzles. CNS performed a plant specific probabilistic fracture mechanics to supplement the criteria of Code Case N-702 and BWRVIP-241 in order to demonstrate that the PoF remains acceptable over the period of extended operation. Conservatively assuming zero inspection for the initial 40 years of operation and examination of25% for PEO, the evaluation concluded the average PoF for a LTOP event is 2.92x 10-11 per year (Reference I) for the nozzle inner radius, and <8.33xl0-13 per year (Reference 1) for the nozzle-to-shell weld, both of which are less than the NRC safety goal of 5x I o-6 per year. The examination history for the subject nozzles is included in Table 2.

NLS2017071 Attachment Page 52of68 10 CFR 50.SSa Request No. RIS-03 (continued) Implementation of Code Case N-702 Duration of Proposed Alternative The duration of this request is for the extended license period ending January 18, 2034. Precedents US NRC letter to Entergy, "Pilgrim Nuclear Power Station - Relief Request PRR-50, Use of Alternatives, Implementation of Code Case N-702," dated January 5, 2016 (ADAMS Accession Number ML15338A309). US NRC letter to NPPD, "Cooper Nuclear Station - Request for Relief No. RI-04 for the Fourth 10-Year Inservice Inspection Interval Regarding Inspection of Reactor Vessel Nozzle-to-Vessel Shell Welds," dated October 8, 2010 (ADAMS Accession Number ML102220449). US NRC letter to NPPD, "Cooper Nuclear Station - Relief Request No. RI-08, Revision 0 Applicable to Fourth 10-Year Inservice Inspection Interval," dated May 20, 2015 (ADAMS Accession Number ML15134A242). Reference

1. "ER 2017-027, Review of Structural Integrity Calculations 1400334.301 & 1400334.302 for Code Case N-702 Relief Request, Revision O," (Proprietary) dated June 7, 2017.

NLS20I707I Attachment Page 53of68 10 CFR 50.SSa Request No. RIS-03 (continued) Implementation of Code Case N-702 Table 2 - Nozzle-to-Shell Welds and Nozzle Blend Radii Component Nozzle ID Nominal PiQe Last AQQendix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size (Inches} Examination Inner Radius (IR) Recirculation Yes NIA (NV) B-D B3 .90 28 I0/20I6 NRI (Outlet) Recirculation VT- I NIA (IR) B-D B3.100 28 10/20I6 NRI (Outlet) Recirculation Yes NIB (NV) B-D B3 .90 28 OI /2005 NRI (Outlet) Recirculation MVT-I NIB (IR) B-D B3.100 28 OI /2005 NRI (Outlet) Recirculation Yes N2A (NV) B-D B3.90 I2 OI /2005 NRI (Inlet) Recirculation Yes N2A (IR) B-D B3 .100 I2 04/2011 NRI (Inlet) Recirculation Yes N2B (NV) B-D B3 .90 I2 01 /2005 NRI (Inlet) Recirculation Yes N2B (IR) B-D B3.100 I2 04/20I I NRI (Inlet) Recirculation Yes N2C (NV) B-D B3.90 I2 04/201 I NRI (Inlet) Recirculation Yes N2C (IR) B-D B3.100 I2 04/201 I NRI (Inlet) Recirculation Yes N2D (NV) B-D B3.90 I2 OI /2005 NRI (Inlet)

NLS2017071 Attachment Page 54of68 10 CFR 50.SSa Request No. RIS-03 (continued) Implementation of Code Case N-702 Nozzle ID Nominal PiQe Last AQQendix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size (Inches) Examination Inner Radius (IR) Recirculation Yes N2D (IR) B-D B3.100 12 04/2011 NRI (Inlet) Recirculation Yes N2E (NV) B-D B3 .90 12 10/2016 NRI (Inlet) Recirculation Yes N2E (IR) B-D B3.100 12 10/2016 NRI (Inlet) Recirculation Yes N2F (NV) B-D B3.90 12 04/2011 NRI (Inlet) Recirculation Yes N2F (IR) B-D B3.100 12 04/2011 NRI (Inlet) Recirculation Yes N2G (NV) B-D B3.90 12 04/2011 NRI (Inlet) Recirculation Yes N2G (IR) B-D B3.100 12 04/2011 NRI (Inlet) Recirculation Yes N2H (NV) B-D B3 .90 12 1012016 NRI (Inlet) Recirculation Yes N2H (IR) B-D B3 .100 12 10/2016 NRI (Inlet) Recirculation Yes N2J (NV) B-D B3.90 12 04/2011 NRI (Inlet) Recirculation Yes N2J (IR) B-D B3 . l 00 12 04/2011 NRI (Inlet) Recirculation Yes N2K (NV) B-D B3.90 12 10/2016 NRI (Inlet)

NLS2017071 Attachment Page 55of68 10 CFR 50.55a Request No. RIS-03 (continued) Implementation of Code Case N-702 Nozzle ID Nominal Pipe Last Appendix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size (Inches) Examination Inner Radius (IR) Recirculation Yes N2K (IR) B-D B3. 100 12 10/2016 NRI (Inlet) B-D B3 .90 Main Steam 24 10/2016 NRI Yes N3A (NV) B-D B3.100 Main Steam 24 09/2016 VT-1 N3A (IR) NRI B-D B3.90 Main Steam 24 03/2000 NoC3l N3B (NV) NRI B-D B3.100 Main Steam 24 0312000 NoC3l N3B (IR) NRI B-D B3.90 Main Steam 24 01 12005 Yes N3C (NV) NRI B-D B3.100 Main Steam 24 0212005 MVT-1 N3C (IR) NRI B-D B3.90 Main Steam 24 03/2000 NoC3l N3D (NV) NRI B-D B3 .100 Main Steam 24 03/2000 NoC3l N3D (IR) NRI N4A(NVi 1) B-D B3.90 Feedwater 12 10/2014 NRI Yes Feedwater 12 Yes N4A (IRi 1l B-D B3.100 10/2014 NRI B-D B3.90 Feedwater 12 NRI Yes N4B CNVil) 10/2014 Feedwater 12 Yes N4B (1Ri 1l B-D B3.100 10/2014 NRI B-D B3.90 Feedwater 12 NRI Yes N4C CNVil) 10/2014 B3.100 Feedwater 12 Yes N4C (IRi 1> B-D 10/2014 NRI

NLS2017071 Attachment Page 56of68 10 CFR 50.SSa Request No. RIS-03 (continued) Implementation of Code Case N-702 Nozzle ID Nominal Pipe Last Appendix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size (Inches} Examination Inner Radius (IR) Feedwater 12 Yes N4D (NVi 1> 8-D 83.90 10/2014 NRI Feedwater 12 Yes N4D (IR)< 1> 8-D 83.100 10/2014 NRI 8-D 83.90 Core Spray 10 10/2016 NRI Yes N5A (NV) 8-D 83 .100 Core Spray 10 10/2016 Yes N5A (IR) NRI 8-D 83.90 Core Spray 10 01 /2005 Yes N58 (NV) NRI 8-D 83.100 Core Spray 10 NRI Yes N58 (IR) 03/2011 Head Yes N6A (NV) 8-D 83.90 6 02/2005 NRI Instrumentation Head 8-D 83 .100 Instrumentation 02/2005 NRI MVT-1 N6A (IR) 6 Head 8-D 83 .90 Instrumentation 03/2011 NRI Yes N68 (NV) 6 Head 8-D 83.100 Instrumentation 03/2011 NRI MVT-1 N68 (IR) 6 8-D 8 3.90 Head Vent 4 03/2011 NRI Yes N7 (NV) 8 -D 83 .100 Head Vent 4 03/2011 NRI MVT-1 N7 (IR) Jet Pump No 0 > N8A (NV) 8-D 83.90 4 03/2000 NRI Instrumentation

NLS2017071 Attachment Page 57of68 10 CFR 50.55a Request No. RIS-03 (continued) Implementation of Code Case N-702 Nozzle ID Nominal Pi2e Last A22endix Nozzle-to-Vessel (NV) Category Number Item Number System Results VIII Exam Size {Inches) Examination Inner Radius (IR) Jet Pump No(3) N8A (IR) B-D B3 .100 4 03/2000 NRI Instrumentation Jet Pump Yes N8B (NV) B-D B3.90 4 10/2014 NRI Instrumentation Jet Pump Yes N8B (IR) B-D B3 .100 4 10/2014 NRI Instrumentation Yes N9 <NVi2) B-D B3.90 CRD Return Line 5 03/2011 NRI NRI Yes N9 (IRi2) B-D B3.100 CRD Return Line 5 03/2011 Notes:

1) Code Case N-702 excludes these nozzles.

2) CNS CRD Line is capped and as such is no longer the return for CRD. Because N9 is a unique nozzle, it will be examined and an alternative is not being sought for this nozzle.

3) Exams performed prior to adoption of Appendix VIII.

NLS2017071 Attachment Page 58of68 10CFRSO.SSa Request No. RRS-02 Cooper Nuclear Station Request to Use Code Case N-513-4 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Hardship Without a Compensating Increase in Quality and Safety ASME Code Component(s) Affected All ASME, Section XI, Class 2 and 3 piping components that meet the operational and configuration limitations of Code Case N-513-4, paragraphs l(a), l(b), l(c) , and l(d) at CNS .

Applicable Code Edition and Addenda

CNS applicable Code for the fifth 10-year ISi interval and the ISI program is the 2007 Edition of Section XI with the 2008 Addenda. CNS fifth interval started April 1, 2016 and ends February 28, 2026. Applicable ASME Code Requirements ASME Code, Section XI, IWC-3120 and IWC-3130 require that flaws exceeding the defined acceptance criteria be corrected by repair/replacement activities or evaluated and accepted by analytical evaluation. ASME Code, Section XI, IWD-3120(b) requires that components exceeding the acceptance standards of IWD-3400 be subject to supplemental examination, or to a repair/replacement activity.

Reason for Request

In accordance with 10 CFR 50.55a(z)(2), NPPD is requesting a proposed alternative from the requirement to perform repair/replacement activities for degraded Class 2 and 3 piping whose maximum operating temperature does not exceed 200° and whose maximum operating pressure does not exceed 275 psig. Moderately degraded piping could require a plant shutdown within the required action statement timeframes to repair observed degradation. Plant shutdown activities result in additional dose and plant risk that would be inappropriate when a degraded condition is demonstrated to retain adequate margin to complete the component's function. The use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition will allow NPPD to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long term repair actions if necessary. Actions to remove degraded piping from service could have a detrimental overall risk impact by requiring a plant shutdown, thus requiring use of a system that is in standby during normal operation. Accordingly, compliance with the current code requirements results in a hardship without a compensating increase in the level of quality and safety. ASME Code Case N-513-3 does not allow evaluation of flaws located away from attaching circumferential piping welds that are in elbows, bent pipe, reducers, expanders, and branch tees. ASME Code Case N-513-3 also does not allow evaluation of flaws located in heat exchanger

NLS2017071 Attachment Page 59of68 10CFR50.55a Request No. RRS-02 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 external tubing or piping. ASME Code Case N-513-4 provides guidance for evaluation of flaws in these locations. Proposed Alternative and Basis for Use NPPD is requesting approval to apply ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l ," to the SW System piping that meets the operational and configuration limitations of Code Case N-513-4, paragraphs l(a), l(b), l(c), and l(d). Application of the Code Case will avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with the cited Code requirements. The NRC issued GL 90-05 (Reference 1), "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptability of limited degradation in moderate energy piping. The generic letter defines conditions that would be acceptable to utilize temporary non-code repairs with NRC approval. The ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed Code Case N-513. NRC approval of Code Case N-513 versions in RG 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1 (Reference 3)," allows temporary acceptance of partial through-wall or through-wall flaws provided all conditions of the Code Case and NRC conditions are met. The temporary acceptance period has historically been the time to the next scheduled refueling outage. The Code Case also requires the Owner to demonstrate system operability due to leakage. The ASME recognized that the limitations in Code Case N-513-3 were preventing needed use in piping components such as elbows, bent pipe, reducers, expanders, and branch tees and external tubing or piping attached to heat exchangers. Code Case N-513-4 was approved by the ASME to expand use on these locations and to revise several other areas of the Code Case. , Attachment 1 provides a marked-up N-513-3 version of the Code Case to highlight the changes compared to the NRC approved N-513 -3 version. The following provides a high level overview of the Code Case N-513-4 changes:

Revised the maximum allowed time of use from no longer than 26 months to the next scheduled refueling outage.
Added applicability to piping elbows, bent pipe, reducers, expanders, and branch tees where the flaw is located more than (Rot) 112 from the centerline of the attaching circumferential piping weld.
Expanded use to external tubing or piping attached to heat exchangers.
Revised to limit the use to liquid systems.
Revised to clarify treatment of Service Level load combinations.
Revised to address treatment of flaws in austenitic pipe flux welds.

_J

NLS2017071 Attachment Page 60of68 10CFRSO.SSa Request No. RRS-02 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4

Revised to require minimum wall thickness acceptance criteria to consider longitudinal stress in addition to hoop stress.
Other minor editorial changes to improve the clarity of the Code Case.

The technical basis for changes in Code Case N-513-4 when compared to NRC approved Code Case N-513-3 is provided in Enclosure 1, Attachment 2. Enclosure 1, Attachment 3 provides additional technical justification for the use of Code Case N-513-4 at CNS. The design basis is considered for each leak and evaluated using the NPPD Operability Evaluation process. The evaluation process must consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgment. As required by the Code Case, the evaluation process considers but is not limited to system make-up capacity, containment integrity with the leak not isolated, effects on adjacent equipment, and the potential for room flooding. Leakage rate is not typically a good indicator of overall structural stability in moderate energy systems, where the allowable through-wall flaw sizes are often on the order of inches. The periodic inspection interval defined using paragraph 2(e) of Code Case N-513-4 provides evidence that a leaking flaw continues to meet the flaw acceptance criteria and that the flaw growth rate is such that the flaw will not grow to an unacceptable size. The effects ofleakage may impact the operability determination or the plant flooding analyses specified in paragraph l(f). For a leaking flaw, the allowable leakage rate will be determined by dividing the critical leakage rate by a safety factor of four (4). The critical leakage rate is determined as the lowest leakage rate that can be tolerated and may be based on the allowable loss of inventory or the maximum leakage that can be tolerated relative to room flooding, among others. The safety factor of four (4) on leakage is based upon Code Case N-705 (Reference 2), which is accepted without condition in RG 1.147, Revision 17. Paragraph 2.2( e) of N-705 requires a safety factor of two (2) on flaw size when estimating the flaw size from the leakage rate. This corresponds to a safety factor of four (4) on leakage for nonplanar flaws. Although the use of a safety factor for determination of an unknown flaw is considered conservative when the actual flaw size is known, this approach is deemed acceptable based upon the precedent of Code Case N-705. Note that the alternative herein does not propose to use any portion of Code Case N-705 and that citation of N-705 is intended only to provide technical basis for the safety factor on leakage. During the temporary acceptance period, leaking flaws will be monitored daily as required by paragraph 2(f) of Code Case N-513-4 to confirm the analysis conditions used in the evaluation remain valid. Significant change in the leakage rate is reason to question that the analysis conditions remain valid, and would require re-inspection per paragraph 2(f) of the Code Case. Any re-inspection must be performed in accordance with paragraph 2(a) of the Code Case.

NLS2017071 Attachment Page 61of68 10CFR50.55a Request No. RRS-02 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 The leakage limit provides quantitative measurable limits which ensure the operability of the system and early identification of issues that could erode defense-in-depth and lead to adverse consequences. CNS will follow all requirements of Code Case N-513-4. With regard to augmented examination process as described in Section 5 of the Code Case, a sample size of at least five of the most susceptible and accessible locations shall be examined within 30 days of detecting the original flaw. The intent of this requirement is to identify the extent of condition that exists within similar system piping that could also be susceptible to similar flaws. Specific to the CNS SW Class 3 piping, if a single flaw is identified within 30 days of scheduled volumetric examinations, CNS may take credit for any previous examination if performed prior to identification of the flaw, as a part of the same set of inspections. Credit will be taken for these examinations to meet the requirements of Code Case N-513-4 paragraph 5(a), provided that the examination meets the inspection method requirements of the Code Case and was performed on SW Class 3 piping components, the inspected segments are of the same design and operation, and the inspected segments are considered to be of the most susceptible and accessible locations determined from the engineering evaluation. This is consistent with the augmented examination approach in GL 90-05 as well as previous NRC approved versions ofN-513 . In summary, NPPD will apply ASME Code Case N-513-4 in its entirety along with RG 1.147, Revision 17 (or later NRC defined revision as applicable) for evaluation of Class 2 and 3 piping flaws at CNS if Code repairs cannot reasonably be completed within the Technical Specifications required time limit. Code Case N-513-4 utilizes technical evaluation approaches that are based on principals that are accepted in other Code documents already acceptable to the NRC. The application of this Code Case will maintain acceptable structural and leakage integrity while minimizing plant risk and personnel exposure by minimizing the number of plant transients that could be incurred if degradation is required to be repaired based on ASME Section XI acceptance criteria only. Duration of Proposed Alternative The proposed alternative is for use of Code Case N-513-4 for Class 2 and 3 piping and components within the scope of the Code Case and the request herein. A Section XI compliant repair/replacement will be completed prior to exceeding the next scheduled refueling outage or allowable flaw size, whichever comes first. This relief request will be applied for the duration of the fifth 10-year inservice inspection interval. If a flaw is evaluated near the end of the interval and the next refueling outage is in the subsequent interval, the flaw may remain in service under this relief request until the next refueling outage. Precedent US NRC letter to Exelon Generation Company Nuclear Fleet - "... Proposed Alternative to Use ASME Code Case N-513-4," NRC Safety Evaluation dated September 6, 2016 (ML16230A237).

NLS2017071 Attachment Page 62of68 10CFR50.55a Request No. RRS-02 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 References

1. NRC Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," dated June 15, 1990.

2. ASME Boiler and Pressure Vessel Code, Code Case N-705 , "Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks Section XI, Division l ," dated October 12, 2006.

3. NRC Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 17," dated August 2014.

NLS2017071 Attachment Page 63 of68 10CFR50.55a Request No. RRS-03 Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Hardship Without a Compensating Increase in Quality and Safety ASME Code Component(s) Affected All ASME, Section XI, Class 3 RHRSWB system piping with a maximum operating pressure less than or equal to 490 psig and a maximum operating temperature less than 200°F at the Cooper Nuclear Station (CNS). The RHR and RHRSWB Systems are designed such that RHRSWB operates at a higher pressure than RHR. The RHR and RHRSWB Systems are standby systems that typically operate during testing or plant shutdown. Under this design, if there is an internal leak within a RHR heat exchanger, RHRSWB water, which is raw water from the Missouri River, will leak into the RHR System. The safety objective of the RHRSWB System is to provide cooling to the RHR System without an uncontrolled release of radioactive material to the environment. The RHRSWB System is designed to provide an adequate supply of cooling water to the RHR heat exchangers during postulated accident and transient conditions to remove the design RHR System heat load and at adequate pressure to prevent uncontrolled release of fission products to the environment due to a RHR heat exchanger tube failure. RHRSWB System at CNS has exhibited a history of degradation similar to raw fresh water systems throughout the nuclear industry. Degradation requiring immediate action to address leakage or observed thinning in the system is generally due to localized corrosion mechanisms.

Applicable Code Edition and Addenda

CNS' applicable Code for the fifth 10-year ISi interval and the ISi program is the 2007 Edition of Section XI with the 2008 Addenda. CNS' fifth interval started April 1, 2016 and ends February 28, 2026. Applicable ASME Code Requirements ASME Code, Section XI, IWD-3 l 20(b) requires that components exceeding the acceptance standards of IWD-3400 be subject to supplemental examination, or to a repair/replacement activity.

Reason for Request

In accordance with 10 CFR 50.55a(z)(2), NPPD is requesting a proposed alternative from the requirement to perform repair/replacement activities for degraded RHRSWB piping which has a

NLS2017071 Attachment Page 64 of68 10CFR50.55a Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure maximum operating pressure in excess of275 psig. Moderately degraded piping could require a plant shutdown within the required action statement timeframes to repair observed degradation. Plant shutdown activities result in additional dose and plant risk that would be inappropriate when a degraded condition is demonstrated to retain adequate margin to complete the component's function. The use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition will allow NPPD to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long term repair actions if necessary. Actions to remove degraded piping from service could have a detrimental overall risk impact by requiring a plant shutdown, thus requiring use of a system that is in standby during normal operation. Accordingly, compliance with the current code requirements results in a hardship without a compensating increase in the level of quality and safety. ASME Code Case N-513-3 does not allow evaluation of flaws located away from attaching circumferential piping welds that are in elbows, bent pipe, reducers, expanders, and branch tees. ASME Code Case N-513-3 also does not allow evaluation of flaws located in heat exchanger external tubing or piping. ASME Code Case N-513-4 provides guidance for evaluation of flaws in these locations. Proposed Alternative and Basis for Use NPPD is requesting approval to apply ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l ," to the RHRSWB System piping having a maximum operating pressure of 490 psig. The operational and configuration limitations of Code Case N-513-4, paragraphs l(a), l(b), and l (d), shall apply. The maximum operating temperature of 200°F in paragraph l(c) shall also apply. Application of the Code Case will avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with the cited Code requirements. The NRC issued Generic Letter 90-05 (Reference 1), "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptability of limited degradation in moderate energy piping. The generic letter defines conditions that would be acceptable to utilize temporary non-code repairs with NRC approval. The ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed Code Case N-513. NRC approval of Code Case N-513 versions in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 (Reference 3)," allows temporary acceptance of partial through-wall or through-wall flaws provided all conditions of the Code Case and NRC conditions are met. The temporary acceptance period has historically been the time to the next scheduled refueling outage. The Code Case also requires the Owner to demonstrate system operability due to leakage.

NLS2017071 Attachment Page 65 of68 10CFRSO.SSa Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure The ASME recognized that the limitations in Code Case N-513-3 were preventing needed use in piping components such as elbows, bent pipe, reducers, expanders, and branch tees and external tubing or piping attached to heat exchangers. Code Case N-513-4 was approved by the ASME to expand use on these locations and to revise several other areas of the Code Case. , Attachment 1 provides a marked-up N-513-3 version of the Code Case to highlight the changes compared to the NRC approved N-513-3 version. The following provides a high level overview of the Code Case N-513-4 changes:

Revised the maximum allowed time of use from no longer than 26 months to the next scheduled refueling outage.
Added applicability to piping elbows, bent pipe, reducers, expanders, and branch tees where the flaw is located more than (Rot) 112 from the centerline of the attaching circumferential piping weld.
Expanded use to external tubing or piping attached to heat exchangers.
Revised to limit the use to liquid systems.
Revised to clarify treatment of Service Level load combinations.
Revised to address treatment of flaws in austenitic pipe flux welds.
Revised to require minimum wall thickness acceptance criteria to consider longitudinal stress in addition to hoop stress.
Other minor editorial changes to improve the clarity of the Code Case.

The technical basis for changes in Code Case N-513-4 when compared to NRC approved Code Case N-513-3 is provided in Enclosure 2, Attachment 2. The design basis is considered for each leak and evaluated using the NPPD Operability Evaluation process. The evaluation process must consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgment. As required by the Code Case, the evaluation process considers but is not limited to system make-up capacity, containment integrity with the leak not isolated, effects on adjacent equipment, and the potential for room flooding. Leakage rate is not typically a good indicator of overall structural stability, where the allowable through-wall flaw sizes are often on the order of inches. The periodic inspection interval defined using paragraph 2( e) of Code Case N-513-4 provides evidence that a leaking flaw continues to meet the flaw acceptance criteria and that the flaw growth rate is such that the flaw will not grow to an unacceptable size. The effects ofleakage may impact the operability determination or the plant flooding analyses specified in paragraph l(f). For a leaking flaw, the allowable leakage rate will be determined by dividing the critical leakage rate by a safety factor of four (4). The critical leakage rate is

NLS2017071 Attachment Page 66of68 10CFR50.55a Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure determined as the lowest leakage rate that can be tolerated and may be based on the allowable loss of inventory or the maximum leakage that can be tolerated relative to room flooding, among others. The safety factor of four (4) on leakage is based upon Code Case N-705 (Reference 2), which is accepted without condition in RG 1.147, Revision 17. Paragraph 2.2(e) ofN-705 requires a safety factor of two (2) on flaw size when estimating the flaw size from the leakage rate. This corresponds to a safety factor of four (4) on leakage for nonplanar flaws. Although the use of a safety factor for determination of an unknown flaw is considered conservative when the actual flaw size is known, this approach is deemed acceptable based upon the precedent of Code Case N-705. Note that the alternative herein does not propose to use any portion of Code Case N-705 and that citation of N-705 is intended only to provide technical basis for the safety factor on leakage. During the temporary acceptance period, leaking flaws will be monitored daily as required by paragraph 2(f) of Code Case N-513-4 to confirm the analysis conditions used in the evaluation remain valid. Significant change in the leakage rate is reason to question that the analysis conditions remain valid, and would require re-inspection per paragraph 2(f) of the Code Case. Any re-inspection must be performed in accordance with paragraph 2(a) of the Code Case. The maximum operating pressure of the Class 3 RHRSWB system is 490 psig. The background, history, and effects of using Code Case N-513-4 at a pressure of 490 psig in lieu of the current 275 psig limitation provided in the Code Case are contained in Enclosure 2, . A review of previous NRC submittals identified that the NRC has previously granted specific relief for leaks on high energy systems (see Enclosure 2, Attachment 3). NPPD is seeking relief for general application for limited degradation in the RHRSWB System raw water piping for a maximum operating pressure of 490 psig. Raw water piping degradation is a well understood phenomenon and the evaluation methods in Code Case N-513-4 are widely applied by the industry in raw water piping systems that operate at a pressure less than or equal to 275 psig without incident. The structural aspects of raising the allowable operating pressure to 490 psig were evaluated as discussed in Enclosure 2, Attachment 3. It was determined that Code Case allowable flaw sizes by both the Linear Elastic Fracture Mechanics and branch reinforcement methods used in Code Case N-513-4 were smaller as would be expected. The effects of jet thrust force were evaluated and it was determined there was little difference in force for a 0.50" diameter flaw size at 275 psig versus 490 psig. The study also determined that jet thrust force increases with increasing leakage rate and that it is appropriate to limit the application of this relief request to 490 psig. , Attachment 3 provides:

1) A review of relevant NRC approved relief requests

2) A structural integrity evaluation that includes:

Design minimum wall thickness comparison

NLS2017071 Attachment Page 67 of68 10CFR50.55a Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure

Code Case N-5 13-4 allowable flaw s ize comparison
Code Case N-51 3-4 cover thickness requirement comparison

3) A jet thrust force evaluation CNS will follow all requirements of Code Case N-513-4. With regard to augmented examination process as described in Section 5 of the Code Case, a sample size of at least five of the most susceptible and accessible locations shall be examined within 30 days of detecting the original flaw. The intent of this requirement is to identify the extent of condition that exists within similar system piping that could also be susceptible to similar flaws. If a single flaw is identified during scheduled volumetric examinations, CNS may take credit for any previous examination if performed within 30 days prior to identification of the flaw, as a part of the same set of inspections. ~redit will be taken for these examinations to meet the requirements of Code Case N-513-4 paragraph 5(a), provided that the examination meets the inspection method requirements of the Code Case and was performed on Class 3 RHRSWB piping, the inspected segments are of the same design and operation, and the inspected segments are considered to be of the most susceptible and accessible locations determined from the engineering evaluation.

However, CNS will increase the number of augmented examinations from 5 to 10 as the RHRSWB system is considered high energy. This is consistent with the augmented examination approach in GL 90-05. This approach adequately meets the intent of the Code Case with respect to augmented examinations. If the Code Case is applied to a leaking flaw in the RHRSWB system, for leakage greater than 5 gpm, the leakage shall be stopped throughout the temporary acceptance period by the use of engineered mechanical clamping designed by NPPD. The engineered mechanical clamping shall be sufficient to withstand the maximum operating pressure and removable such that the frequent periodic inspections defined in paragraph 2( e) of N-513-4 may be performed. In summary, NPPD will apply ASME Code Case N-513-4 and RG 1.147, Revision 17 (or later NRC defined revision as applicable) for evaluation ofRHRSWB piping flaws at CNS if Code repairs cannot reasonably be completed within the Technical Specifications required time limit. NPPD will apply a 490 psig maximum operating pressure in lieu of the 275 psig maximum operating pressure defined in paragraph l(c) of the Code Case. Code Case N-513-4 utilizes technical evaluation approaches that are based on principals that are accepted in other Code documents already acceptable to the NRC. In order to bolster defense-in-depth and avoid adverse consequences as a result of increasing the maximum operating pressure to 490 psig, NPPD is making three additional commitments:

Safety factor of 4 applied to leakage limit
Augmented inspections increased from 5 to 10
Leakage shall be stopped by the use of engineered mechanical clamp

NLS2017071 Attachment Page 68 of68 10CFR50.55a Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure The application of this Code Case, along with the additional commitments above, will maintain acceptable structural and leakage integrity while minimizing plant risk and personnel exposure by minimizing the number of plant transients that could be incurred if degradation is required to be repaired based on ASME Section XI acceptance criteria only. Duration Of Proposed Alternative The proposed alternative is for use of Code Case N-513-4 for Class 3 RHRSWB piping and components within the scope of the Code Case and the request herein. A Section XI compliant repair/replacement will be completed prior to exceeding the next scheduled refueling outage or allowable flaw size, whichever comes first. This relief request will be applied for the duration of the fifth 10-year ISi interval. If a flaw is evaluated near the end of the interval and the next refueling outage is in the subsequent interval, the flaw may remain in service under this relief request until the next refueling outage. Precedents US NRC letter to Exelon Generation Company Nuclear Fleet - ".. .Proposed Alternative to Use ASME Code Case N-513-4," dated September 6, 2016 (ML16230A237). US NRC letter to "Peach Bottom Atomic Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units l and 2 - Relief from the Requirements of the ASME Code," dated March 19, 2015 (MLl 5043A496). References

1. NRC Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," dated June 15, 1990.

2. ASME Boiler and Pressure Vessel Code, Code Case N-705 , "Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks Section XI, Division 1," dated October 12, 2006.

3. NRC Regulatory Guide 1.14 7, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 17," dated August 2014.

NLS2017071 Page 1 of 27 Enclosure 1 RR5-02, Cooper Nuclear Station Structural Integrity Associates, Inc. Final Code Case N-513-3 Markup for Revision 4 Technical Basis for Proposed Fourth Revision to ASME Code Case N-513 Cooper Nuclear Station Technical Basis for Proposed Alternative to Use ASME Code Case N-513-4 Note: Enclosed report contains 26 pages not counting this cover page.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 1 of 26 Attachment 1 Final Code Case N-513-3 Markup for Revision 4 to 1700405.401.RO 1-1 e Structural Integrity Associates, Inc."'

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 2 of 26 Record # 12-8-t I CASE CASES O F ASME BOILER AND PRESSURE VESSEL CODE N-513- 4~ Approva l Date: Ja nua ry 26, 2009 Code Cases will remain available for use until anm1/led by the applicable Standards Committee. Case -513-:!J 1c l he l:>![llng ,Jes1gn <'.pc.le shall be .1sed u1 Evaluation Criteria for Tern porary Acceptance of J.:ter1111nmg th< stress m<l1ces B. and f3 <md stre" Flaws in Moderate Energy C lass 2 or J Piping llltl!ns1f!Ci.\tlVll foctor I for flaw I!\ .du.RtWl1 fo(irnVtn!.!, Section XI, Division 1 \ 'odt> anplicabil1tv 1!1111~ m tenns ~)f ~Pmnone-nt gl:ometrv such ns D,._1um1 ratio 11 the plPlfll..!. Jes1gn Inquiry: What requirements may be used for f'ode doe' not pro\lde stress md1ccs. ~ccuon Ill 2004 temporary acceptance of flaws, including through-wall h.!ttlllJl or lakr Edrnons "d .\dden<la 111m h,* used 10 flaws, in moderate energy Class 2 or 3 piping mcludmµ define R and B _ db...)\.\'S. bent nipe. reducer:;;. C'xpandcrs. ~md hrnnch t..:"t!s. (i;i} The provisions of this Case demonstrate the without performing a repair/replacement activ ity? integrity of the item and not the consequences of leakage. It is the responsibility of the Owner to Reply: It is the opinion of the Com mittee that the ol*m 'A .tFttl* ) ol*P* "~"AEHiit 1 consider"'!S effects of following requirements may be used to accept flaws, leakage m demonstratm 0 S\stem oncrnhtl1l\ aml including through-wall flaws. in moderate energy Class nl"rfonn mg nlant lloodm~ anal\"')C'<.;. 2 or 3 piping mduc.lmg e "'"'" h,'nt p1 pc reduca,_ (t} Th* o\"Hl>l8EIPA r*netl. 1'~ l.l lA* "I ""llBAUI expander>. ,mt! hrnnch tel's, without performing a l*ffi* frr <<h1eh the temrernr<< 0ee*1*tnAee ofle<Fta 11r,* repair/replacement activ ity for a lim ited time, not 1Bl!il1'd llm APl e*ce**d*A! 2e ffidAlA.I fn "' tke IAllH.! exceeding the eYal~et1eR pu1ea a:, d*fu<<a tA tilt> d1.;ee *01'" cftl" ee11e!1£;e11.

         ~t1111e to the next scheJubl reful'lmg ,,utaq.c.

2 PROCEDURE SCOPE (a) The flaw geometry sha ll be characterized by (a) These requirements apply to the ASNIE Section volum etric inspection methods or by physical III. SI B3 I. l , and ANSI 831.7 piping, classified by measurement. The full pipe circum fere nce at the fl aw the Owner as Class 2 or 3 that is ,1ccesS1ble for location sha U be inspected to characterize the length and mspectwn. The provisions of this Case do not apply to depth of all flaws in the pipe section. the fo llow ing: (b) Flaw shall be classified as planar or nonplanar. (1) pumps, valves. expansion joints, and heat (cl When multiple fla ws, including irregular exchangers c:xc~pt as pronded m rhJ; (compound) shape flaws, are detected the interaction (2) weld metal of socket welded joints; and combined area loss of flaws in a given pipe section (3) leakage through a flange joint; shall be accounted for in the flaw evaluation. (4) threaded connections employing (d) A flaw evaluation shall be perform ed to nonstructural seal welds for leakage protection. determine the conditions for fl aw acceptance. Section 3 {IJJ I his r1s,* 111.1' be <mplted l*> heai exchanger provides accepted methods for conducting the required

         ~xtem~I tubm" or ptpmn. pnw1ded tli<* flaw is                         analysis.

charact.:nZl<d m accordance with '.:lat and kakage is (ei Frequent periodic inspections of no more than morutorcd 30 day intervals shall be used to detem1ine if flaws are (<;.~)The provisions of this Case apply to Class 2 or growing and to establish the tim e~___._at which the 3 piping 111 liyu1d svskms whose maximum operating detected flaw will reach the allowab le size. temperature does not exceed 200°F (93°C) and whose Alternatively, a flaw growth evaluation may be maximum operating pressure does not exceed 275 psig performed to predict the tim e___,;........, at which the (1.9MPa). detected flaw wiU grow to the a llowable size. The flaw (s_it;) The followi ng flaw evaluation criteria are growth analysis shall consider the relevant growth perm ined for pipe and tube mcludmg dhow'. bent rme. mechanisms such as genera l corrosion or wastage. reducers. e:-~panders. ,ui...i branch tees. The strmeht pipe fatigue. or stress corrosion cracking. When a fl aw flaw evaluation criteria are permined for adjoining growth analysis is used to establish the allowable time fittings and flanges to a distance of (Rot).,, from the weld for temporary operation, periodic examinations of no centerline . more than 90 day intervals shall be conducted to verify the flaw growth ana lysis predictions. Draft 15 (05/05/14) to 1700405.401.RO 1-2 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RRS-02, Cooper Nuclear Station Page 3 of 26 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-:!~ (/) For through-wall leaking flaws, leakage shall be When through-wall ax ial flaws are eva luated . the a~seffed Ii, momtored daily 'l"Bll<aa"'A< to confirm the allowable flaw length is: analysis conditions used in the evaluation remain valid. (g) If examinations reveal flaw growth rate to be unacceptable, a repair-t>F-_replacement aclt\ 1t.v shall be performed. (h) Repair--ff.._replacement acti1:itie' shall be performed no later than when the predicted flaw size from either periodic inspection or by flaw growth analysis exceeds the acceptance criteria of 4, or duruuz where the next scheduled ~outage, whichever occurs p = pressure for the loading condition tl2!.I:: Z has l>een first. Repair~rep lacement act1v1lies shall be in Do = pipe outside diameter added to equation accordance with IWA-4000 er [J\",\ ,l!(I(,), re*1ieetl'.'el). Uj = flow stress (! i. IA J";flit1PA' **HI ,\d*leflRR ~nar IA ll*e 1001 .\adi!flaa; QllB, iA tile ! 0 0! o\dflel'BR aAd later iA Rfl1FQ811<8 ""llH

                                                                                 .s;. =Code specified y ield strength
                                                                                 ~ = Code specified ultimate tensile strength and P' '" lll'lO.                                                          SFm= structural factor on primary membrane stress (i) Evaluations and examination shall be as specified in C-2622 documented in accordance with IWA-6300. The Owner shall document the use of th is Case on the applicable z      load multjpl1ei for ductile tlaw extefbhl!l
                                                                                        !equal to Ill \\hen u*mg lurnl lc1ad cnteriaJ data report form.

Material properties at the temperature of interest 3 FLAW EVALUATION shall be used. Planar flaws to straight pipe shall be evaluated in accordance with the requirements in 3.1. Nonplanar "G' rnROOG$We<O*~ flaws m straight pipe shall be evaluated in accordance with the requirements in 3.2. Through-wall !law" in elbow-, and bent pipe ohall be evaluated in acwrdanc e with the regu1rements m 3 3 Tlu*ough-wall flaw' in reducer<. e'IOpander<i. and branch tee' *hall be evaluated in accordance with the regmrements in 3.-1 and 35. re~mectivelv F law growth evaluation shall be performed g in accordance with the requirements in 3.Q.;. onferrous (c) For planar flaws in ferritic piping, the evaluation materials shall be evaluated in accordance with the procedure of Appendix C shall be used. Flaw depths up requirements in 3 ..:+. to 100% of wall thickness may be evaluated. fllllic For all tlaw evaluaUon; all Scrx1cc Lc,*cl load depU1 <L 15 detjned to Figures C-4310-1 and C=-1310-;: combu1ations shall be evaluated to determine the mpst When through-wall circumferential flaws are evaluated li1niting allo\yable 1ltJV.' .,ize. in accordance with C-5300 or C-6300, the flaw depth to thickness ratio, alt, shall be set to unity. When applying 3.1 Planar Flaws in Straight Pipe the Appendix C screening criteria for through-wall axial (qi For planar flaws , the flaw shall be bounded by a flaws, alt shall be set lo unity , and the reference limit rectangu lar or circumferential planar area in accordance load hoop stress, cri. shall be defined as cr, J/1-1,,. When with the methods described in Appendix C. IW A-33 00 through-wall axial flaws are evaluated in accordance shall be used to determine when multiple proximate with C-5400 or ('-6400 the allowable length is defined flaws are to be evaluated as a single flaw. The geometry by eqs. (1) through (3), with U1e appropriate structural of a through-wall planar flaw is shown in Fig. I. factors from Appendix C, C-2622 When through-wall (b) For planar flaws in austenitic piping, the flaws are evaluated in accordance with C-7300 or C-evaluation procedure in Appendix C shall be used. Flaw 7400, the formu las for evaluation given in C-43 00 may depths up to 1000/o of wa ll thickness may be evaluated. be used, but with values for Fm, F0

and P applicab le to When through-wall circumferential flaws are evaluated, through-wall flaws. Relations for Fm, Fi. and F that take the f01mulas for evaluation given in C-5320 or C-63'.!0 into account flaw shape and pipe geometry (R/I rati o) a<, rnmlicable. of Appendix C may be used, with the flaw shall be used. The appendix to thi s Case provid es depth to thicknes s ratio, alt, equal to unity. equations for Fm, Fb, and P for a selected range of Rlt.

Geometry of a through-wall crack is shown in F ig. I. Draft.15 (05/05/14) 2 Attachment I to 1700405.401.RO 1-3 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RRS-02, Coope r Nuclear Station Page 4 of 26 CASE (co ntinu ed) CASES OF ASME BOU,ER AND PRESSURE VESSEL CODE N-513-4.l FIG. 2 SEPARATION REQUIRENENTS FOR ADJACENT THINNED AREAS

                                                 .r, , - ---'-""'--*-i L.*----*-""-**'

L. ... - O.HL. 1*1. / 3.2 Nonp lanar Flaws in Straight Pipe defined in Fig. 3. When the above requirement is not __fa) The evalmtion shall consider the depth and satisfied, (;,..!l.) shall be met extent of the affected area md require that the wall (:..:_~) When Lm is less than or equal to 2.65 thickness exceed t,,,,,, for a distance thal is the greater of (Rctmm"f' and tnorn is greater than 1.13 tmm , ta1oc is 2.5 ../Rtnom or U...~8 between adjacent thinned regions, determined by satisfying both of the following where R is the mean radius of the piping item based on equation s: nominal wall thickne ss and Lm,.-.g is the average of the extent of Lm below tm'" for adjacent areas (see Fig. 2). Alternatively, the adjacent thinned reg ions shall be (5) considered a sing le thinned region in the evaluation. __fb) For nonplanar flaws, the pipe is acceptable when ~itherl!l.J11l.J!illL!b)(!). or (b)(l) and !bli3l are (6) met. _____{J__I!he remaining pipe thickness (tp) is greater than or e<pJal to the minimum wall thickness tmm: When the above re<pJirements are not satisfied, (~ shall be met. t~ pDp (4) (~_..) When the requirements of (~J . (.;w, md 2(S+ 0.4 p) (J<l.!) above are not satisfied, ta1oc is determined from Curve 2 of Fig. 4. In RElelitietL :_. J*all R11.f) llte where lallenlll? OftURfietl: p = maximum operating pressure al flaw localion S = allowable stress al operating temperature 1-I TI1e remrunmg degraded pipe sectJon m eet> the longitudinal stress limits in the de,ign Code for the Illilli1&. ~""th* nemmal f'lf'* l en~11>1<1inal-&.....1~s

                 ,_.) Ati an altetJtal!ve to ibl!lh\ltem..r1.d) , an          re , ~itmg II am all 8u, iee Lt , el B I'""""' l*if'*

evaluation ol tl1e remru11ing ptpe thickness (t,) may be le..~ performed as given below. The evalualion procedure is (c) When there is through-wall leakage along a a function of the depth and the extent of the affected portion of the thinned wall, as illustrated in Fig. 5 , the area as illustrated in Fig. 3. flaw may be evaluated by the branch reinforcement (:;) When Wm is less than or equal to 0.5 method The thinned area including the ilirough-wall (Rct) 112 , where Ro is the outside radius and Wm is defined opening shall be represented by a circular penetration at in Fig. 3, the flaw can be classified as a planar fl aw and the flaw location. Only the portion of the flaw lying evaluated in accordance with 3.l(a) through 3.l(c), within t.,. need be considered as illustrated in Fig. 6. above. When the above requirement is not satisfied, When evaluating multiple flaws in accordance with (.;w shall be met. 3.2(a), only the portions of the flaws contained within C..~J When Lm~J is not greater than (Rot.. .. )112 , todi need be considered. t.1oc is detennined from Curve 1 of Fig. 4, where Lm~J is Draft 15 (05/05/14) to 1700405.401.RO 1-4 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 5 of 26 CASE (continued) CASES OF ASME BOILERAND PRESSURE VESSEL CODE N- 513 -~ FIG . 3 ILLUSTRATION OF NON PLANAR FLAW DUE TO WALL THINNING l Lmeo Transverse (clrcumferentlall Axial direction direction Editors Nc*te This Figure 3 u t*. t-e delete :I an.J replaced wdi the F1g11re 3 '" the f *l: **H'tfi<J" page Draft 15 (05105114) 4 to 1700405.401.RO 1-5 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02, Cooper Nuclear Station Page 6 of 26 CASE (continued) CASES OF AS!\llE BOILER AND PRESSURE VESSEL CODE N-513-4.} FIG. 3 ILLUSTRATION OF NONPLANAR FLAW DUE TO WALL THINNING

                                                       -     1..m(o) -

Axial direction 1 Lm!(ti

                                                                       ---\------

I Transverse (circumferential) direction FIC. 4 ALLOWABLE WALL THICKNESS AND LENCTH OF LOCALLY THINNED AREA 0 '-----'----L-----'----L----'----------'---__J 0 2 6 7 8 Draft 15 (05105114) 5 Attachment l to 1700405.401.RO 1-6 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 7 of 26 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL C ODE N-51J-4J. FIG. 5 ILLUSTRATlON OF THROUGH-WALL NONPLANAR FLAW DUE TO WALL THINNING Transverse Cclrcumferentlal) Axial direction direction ( Editor"s Note 11u' F1gur* 51s to be deleted and replaced with the Furure < on U1e fullowrng page Draft 15 (05105/ 14) to 1700405.401.RO 1-7 SJ Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 8 of 26 CASE (continu ed) CASES OF ASME BO ILER AND PRESSURE VESSEL CODE N-513-4J FIG. 5 ILLUSTRATION OF THROUGH-WALL NONPLANAR FLAW DUE TO WALL THINNING Through-wall

                  +h. . t t .

mm

                                  ~~~.
                                                         /  opening
                                                              ~~
                                             - l . . . a xi a l -

Axial direction I __.___ l f Lcirc I Transverse (circumferential) direction

                                  ~l
                                            ~

Draft 15 (05/05/14) 7 to 1700405.401.RO 1-8 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 9 of 26 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-43 The minimum wall thickness, t., .,. shall be are ad1u:"ll:d. to th.:count for the f"'C1..m1etn difference$. as determined by eq. (4). For eva luation purposes, the d~scnhL'J helL'w \ltl!maU\'l: mi::thods mm he used to adjusted wall thickness. l.,;1, is a postulated thickness as L-<1lculah.' the stre$ses U'it!<l m t!\"aluatwn shown in Fig. 6. The pipe wall thickness is defined as The lwor stress. (f, for elbow and bent pipe the thickness of the pipe in the non-degraded region as cvaluarton sh<lll be. shown in Fig. 6(a). The diameter of the opening is equal to dod; as defined by lo4J as shown in Fig. 6(a). The postulated value for t.,;1 shall be greater than t.,,. and (J . =( pD 0 )[ 2R,,.,,,,+R sin¢ ]+(~) R,M. ro 0

                                                                                     '          21       2(R,,.,., + R sin¢)          h'"       l   ....(£).

shall not exceed the pipe wall thickness. The lod; value 0 may be varied between Im., and the pipe wall thickness to determine whether there is a combination of t.,;1 and d.,;1 that satisfies the branc h reinforcement requirements. The values of lod; and d.d; of Fig. 6(b) shall satisfy : R ,.., dbc>w or bent 01rc bend r.1d1us

                                                                                          ¢          c1rcumlerenttal angle defined m F1i:!ure 7 h            f1.:,1b1lm d1arnctenst1c
                                                                                         .\ l,       re~ultant     pnm an b..:ndmg m. .1m ent I           momt!nl vf mert1?1 hased on 1.!ValualKm \Vall The remaining ligament average thickness, I,,,,                                      thickness I over the degraded area bounded by d.,;1 shall satisfy :

f,qw1t 1on 9 is only *<[?phcablc for elbows and bent pipe where h " I (l!.9) fht! ;1xml mt!mbrane pressure strt!ss a'f, for t"lhow anJ bent ptpe eva!Uath'Tl shalJ be In ad<l1tl<.m the pr12c secthm mdu<lmg the eyUivalt:nt hole representa11011 shall meet the long!ludmal >tress lllmts 111 the design t 'o<le f('r the Q!J.lli1£... __If a flaw growth analysis is performed, the growth where B. 1s n prnnan su-es.<:; 1ndt""\. a~ defined m St~ct10n in flaw dimensions shall consider the degradation [fl for the piping item B. shall he egu.11to11 5 t<'r mechanisms as relevant to the application. The flaw is dbm.\"'i and bent Plpe acceptable when there is sufficient thickness in the Th.: a"\.ial hendin~ :-ar~ss u. fi.1r t*lb<~w and bi::nt degraded area to provide the required area pipe ernluat1on shall be reinforcement. (d) A lternative ly, if there is a through-wall ope ning along a portion of the thinned wall as illustrated in F ig. --

(] I I 5 the flaw may be eva luated as two independent planar through-wall fl aws, one oriented in the axial direction and the other oriented in the circum ferentia l di rect ion. where /3 1s a prtmarY :;tress mdex as ddmed m Sectton The minimum wall thickness Im.,, shal l be determined lfl for th\! 111pulg ttem. by eq. ( 4). The allowable through-wall lengths in the Th~ th(!nnnl t!Xnanshm stress_ rJ for dbo\\' and axial and circumferential directions shall be determined hent Dtl'e ernluauon shall he by vary ing lo4J shown in Fig. 5 from lnom to tm.,. The allowable through-wall flaw lengths based on t.,;1 shall be greater than or equa l to the corresponding L""' and -- *

                                                                                                         <J   =i(D,M   '21
                                                                                                                             ,) -------~~

1: L"n (see Fig. 5) as determined from 3. i (a) and 3. l(b) or

3. l (c), as appropriate . The remaining ligament average thickness, t,,,8 , over the degraded area bounded by L.,,.,

and Lm, sha ll satisfy eq. @'). strt!ss Lrll~ns1ficatton fact.0r as defi.nc;<l m tht! di:s1*~n C'<'dl! for the! rmmg item 3.3 Through-wall Flaws in Elhmn and Bent Pipo Through-wall flaws m dbows ,mu bent pipe Jtla\ be .\/ resultant rl1ermal exb'ans10n mom~nt (!Valuated lL';tng the strmnht pme Prr..-icc..:dures 11n*t:n m 3 1 or 3 2!d'l JX1.lvidt!d the stresses used m tht! evalunt1on Draft 15 (05/05/14) to 1700405.401.RO 1-9 l} Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 10 of26 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513- 4~ 3.4 Through-\\ull fla"' in Rrd1Jr1'r' and I"ht! axial bending s1rl!ss " ;111J thl*rnrnl l'XJ'fms1on E'l'.pandrr; s1rl!s:-- rJr for brnnch tel! eraluauon shall he dl'tl.!mHnl.!d Through-wall tlaws m rt! ...h1cers :md expander..;; mav from l!Q 15'1 and ~9 1 161 re~m~clJVl.!h be ev~luated us1rw* the stramht pme nrocedurc~ g1v~n in 3 l or 3 2(d) provided the stresses used m the 3 .~ Flaw Growth Evaluation

          .;valuat10n are ad1u~tcd t0 accvunt for th.; geomctn                              If a flaw growth analysis is perfonned. the growth
          ,ltfferen<;cs '" dcscnbed below Alternative mcth,,J.,                      ana lysis shall consider both corrosion and crack-growth mav be used h' calculate the stress<" lL'cd 111 c\*allli1t1,m              mechanisms as relevant to the application.

l'i!! ~ 1lluslrates the reducer anJ exp,mder z.mes In perfonni ng a fl aw growth analysis, the discussed below Evaluatton or ilaws m the snrnll end procedures in C-3000 may be used as guidance. trnns1t1on z0nl! ts ou1s1JI! the scope l*I this "asl.' Relevant growth rate mechanisms shall be considered. The hoop str\!ss. (J,., and axial mt!mbran~ pressure When stress corrosion cracking (SCC) is active, the stress."". fc.~r n:duc.;r or expand.;r i:vnluatton shall bl! following growth rate equation shall be used : ___ er,=( P~. )------~1~l~-11 ( 114) where da/dt is flaw growth rate in inches/hour, Kmax is er = B ( pD,

                                     ,,,     I   '21 l) ______i_,_1-l~\             the maxim um stress intensity factor under long-tenn steady state conditions in ksi in.0 *' , ST is a temperature where D. ts the small end OJ) for !laws m the small end                    correction factor, and C and n are material constants.

and the lame end OD for all .. .thcr llaws For intergranular SCC in austenitic steels, where T ~ The axrnl hendtn>! stress ll. and thennal cxpanston stress, ll,, for reducer or expander ernluathm shall bt* 200°F (93°C). c = 1.79 " 10-* _ _ _ er.= B,( D~* )------'"(~!~'\ Sr n

                                                                                                   =1
                                                                                                   = 2.16 1

_ _ _ er, = { 1_* )-------'-1~!=0*

                                             -D-;M-                                  For transgranu lar SCC in austenitic steels, where T             ~

200°F (93°C). where/ 1s based on the de!Crackd ,,,ctwn C = l.79x 10- 1 3.5 Through -wall Haw~ in Branch fr" Sr = 3.7I x l O" (JO(o.01942 T- 12.2'1] Bmnch reinforceme nt rt!qwn.*ments shall he: met in n = 2.161 accordance with the design Cude I l the dest>!n <:ode did rhX rcyuu~ reinforcement for cv8luauon num(">Scs a The temperature, T, is the metal temperature in remforc~menl tl.!g1on i:-; defim:<l as a region of radius D degrees FahrenheiL The flaw growth rate curves for the of the brnnch pipe from the center of the branch above SCC growth mechanism s are shown in Figs. 2-7 connect11.m Throue*h-waJl tlaws in branch k"e'> ouL"tdt! and .!.<_~. Other growth rate parameters in eq. (J ]u) may of the rC"mforcement region mJv b1..* t!valuatcd usmg the be used. prov ided they are supported by appropriate slra11!ht nm~ nr0cedtlft!$ QJ\*en m 3 1 1.)f 1 ;i(d prond.:<l1 data . t.h~ stresses ust!d tn tht' e\*aluatton are dd1usted IG a~count for the (,.~0mern: differences as dl!s~nhcd 3.]4 Nonferrous Materials bdc>w .\ltemJllve methods mav be uud r..;. calculate the For nonferrous materials, nonplanar and planar

         ' tresses used m ,*rnluatton. l\*aluanon or llaws m the                    fl aws may be evaluated fo llowing the general approach re2mn <'f hran...:h remforcernc.;:nt ls 0uL'>u.lc th~ scop\.! 1.>f         of 3. l through 3._<9. For planar flaws in ductile thts \L,C                                                                  materials, the approach given 111 ' l f'P) ElRH 11fnr The hoon '>lre:".':i a .. , anJ ~tx:1at m~mbranc pre-ssurt:         :mstcnil1c mpe may be used; otherwise, the approach slrt:ss_ fi . fi..1r hraru.:h lt!<! eralua110n shall l)I;;! Jl!t.enrnnc:d  given n ' I 'a' t1AJ 1 1 for fernt1c pm~ should be from eg ( 13) and 1..*q (14), respccll\'dv The outside                     applied. Structural factors provided in 4 sha ll be used. It dmmcta for each ,>f these cquatrnns shall be for the                       is the responsibility of the eva luator to establish branch 0r nm ptpe, 1..l..:pending vn tht! flmv locanon                     conservative estimates of strength and fracture toughness for the piping materia l.

Dra ft 15 (05/0511 4) 9 Attachment l to 1700405.401.RO l-10 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 11 of26 CASE (continued) CASES O F ASME BOU.ER AND PRESSURE VESSEL COD E N-5 JJ-4J FIG. 6 ILLUSTRATION OF ADJUSTED WALL THICKNESS AND EQUIVALENT HOLE DIAMETERS T a...>o...:U..~"'-4------+--'""'~...,._,..,.f" l*I ~Justed W II Thlcic.iie O I

                                                                            .I I

Sltift figure (b) to the

: rigltt so that dadj r,,,., : ! width lines up with

                                   -.--{                 :                  :            figure (a).
                                   ~~-!-----------~ ~~~:e--d~~--.~
                                                      ~d.i1-l 1

lb) Equtval nt Hole Representation FIG. 7 CIRCUMFERENTIAL ANGLE DEFINED

                                                                    ~
                                                              ----~-~~~~~~

Orafi 15 (05/05/1 4) 10 to 1700405.401.RO 1-11 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02, Cooper Nuclear Station Page 12 of 26 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4J. FIG. 8 ZONES OF A REDUCER OR EXPANDER large end transition zon e tran sition zone GENERAL NOTE: Transit io~ zones &><tend from the point on the ends where the diameter beg ins to change to the pomt on the central cone where the cone ongle is constant . Draft 15 (05/05/ 14) 11 to 1700405.401.RO 1-12 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RRS-02 , Cooper Nuclear Station Page 13 of26 CASE (continued) CASES O F ASME BOILER AND PRESSURE VESSEL CODE N-513-4.J FIG . ~ FLAW GROWTH RA TE FOR IGSCC IN AUSTENITIC PIPING l.(E.02

                                                  ~ A"'1onHl:i Plpl~

T s 211C*' I/

                                   ,t;
                                   ~                                                                        ,,

i /

i~ 1.c:E.ai 0

tJ 1£E,48

                                                                                /   "

I/

                                                         /
                                                  ,, /

IJ:E-Oll I 10 100 sar ... tntw1'1tv FiKIDr,Jr 11<51 ln."'l Dra!l 15 (05/0511 4) 12 to 1700405.401.RO 1-13 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 14 of 26 CA~ (continued) CASES OF AS/all! BOILER AND PRl!SSURI! VESSl!L CODI! N-513-.U FIG . ..1.Q8 FLAW GROWTH RATE FOR TG SC C IN AUST ENIT IC PIPIN G UE..O 2 Fl AurllW'fUo ,..pl"'iJ T< ZD' I.; 1.1£>4 1.CE.44 v I/

                                                    =       T*llJ IF
                                                                /                              /        I T* IOIF L,*

1.CE.47 1.CE..01 v 1.l'E,.011 I to 100 S tr11n runsnv F.Jaor, tbl 1n.u 1 Draft 15 (05/05/l 4) 13 to 1700405.401.RO 1-14 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 15 of 26 CASE (continued ) CASES O F ASME BOILER AND PRESSURE VESSEL CODE N-513-4J 4 ACCEPTANCE CRITERIA nondirnensional stress intensity factor for through-wall circumferential !law under Piping containing a circumferential planar flaw is membrane stress acceptable for temporary service when flaw evaluation mom~nt of mcrtrn hA.s~-..1 ('.Jl l.!\:Rluatirm provides a margin using the strucrural factors in th.tckne" 1 Appendix C, C-2621. For ax.ial planar flaws, the max.imum stress intensity factor under structural factors for temporary acceptance are as long term steady state conditions specified in Appendix C, C-2622. Stra1\.\ht *-.J2ipi<.<Ag L maximum extent of a local thinned area containing a nonp lanar part:::-through-wall flaw is with t < !,10m acceptable for te mporary service if the remo1nme. "'I"' length of idealized through-wa ll planar secl1on meel.s the long1tudmal stress lunw; in the Jesil!n fl aw opening in the axial direction of the Cc>de for the p1p111g and Ip 2: laio" where 4 1vc is pipe, as illustrated in Fig . 5 determined from 3.2(b) . Strn1!!ht I * !'£~ conta ining a length of idea lized through-wa ll planar nonplanar through-wall flaw is acceptable for temporary flaw opening in the circun1 ferential service when the flaw conditions of 3.2(c) or 3.2(d) are direction of the pipe. as il lustrated in F ig. 5 satisfied . . \n dhow or bent ptpe contamtn11 a nonr lanar maximum extent of a local thinned area lhr<lLlh-wall !law IS :1cceptable for km p,irnrv* service 1f with I < l.. m the (Jaw comllllons of 1 1 *Ire sa11sfieu \ retlutw or axial extent of wall thinning below t,.m l!XPandcr l .mtamm1! n ncmnlanar through-\wll flnw is circumferential extent of wall thinning acce ptable f<>r tcm porarv "'rv1ce 1f the flaw cond!l1nn' belowt.,.in Jf 1 4 are ~ausfted \branch U!e contmnmg ~i n. .mplanar average of the extent of L.. below thr<'UZ.h-wnll llaw 1s ncc~pi...1ble ft-,r t~m p\}faf'v serv1~l! 1f Imm for adjacent thinned areas the tJaw cond1t10ns qf 1 'i are satisfied L..., maximum extent of thinned area. i M2 bulging factor for axial Oaw 5 AUGMENTED [;'{ MINATION .\./ r~su ltant prunap. hcnJinu mc'ltl~nt Ji rcsu [t,rnt th(!nnal c.!.\.panshm moment An augmented volumetri c examination or physical R mean pipe radius measure ment to assess degradation of the affected fi ...,,_ __c~*l~b~o~w_o=r~b~e=n=l*p~m~e~c~*e=n~k=*r~h~n"~*~h=e1~1d~n=1J=11=L' system shall be performed as follows : R0 outside pipe radius (a) From the engineering evaluation, the most S allowable stress at operating temperature susceptible locations shall be identified A sample size SFm strucrural factor on primary membrane of at least five of the most susceptible and accessible stress locations, or. if fewer than five , all susceptible and Sr coefficient for tem perature dependence in accessible locations shall be examined within 30 days of the crack growth relationship detecting the flaw . Code-specified ultiniate tensile strength (b} When a flaw is detected, an additional sample Code-specified yield strength of the same size as defined in S(a) shall be exam ined. metal temperature (c) This process shall be repeated within 15 days llHL'<..im wn extent of a local thinned area for each successive sample, until no significant flaw is perpendicular to L,., with I < 1,.., detected or until I 00% of suscepti ble and access ible x,, mini mum distance between thinned areas i locations have been examined. andj load mult1pher for du(;lJk 11,1w c,;tens1on 6 NO MENCLATURE a fl aw depth c half crack length fj. 8, - Sect ion I II nnmarv $tr~s$ 111d1ct"s daldI flaw growth rate for stress corrosion c coefficient in the crack growth re lationship cracking i inside p1rt! dmmt:tcr diameter equivalent circular hole at laa1 Do outside pipe diameter diameter of equiva lent circular hole at F nondimensionaJ stress intensity factor for Imm through-wall ax.ial flaw under hoop stress ll tlex1b1htY charncterht1c nondirnensional stress intensity factor for :st.res::> mtt"ns1ficat10n factor through-wall circum ferential flaw unde r £ total crack length = 2c pipe bending stress £ ,11 allowable axial through-wall flaw length n exponent in the crack growth relationship Draft 15 (05/05/ 14) 14 to 1700405.401.RO 1-15 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RRS-02 , Cooper Nuclear Station Page 16 of26 CASE (continued ) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4.J p maximum operating pressure at flaw location evaluat1<>n wall thickness. sum*und1m! the de,,radcd area to<1, adjusted wall thickness which is varied for evaluation purposes in the evaluation of a through-wall nonplanar flaw t.ioc allowable local thickness for a nonplanar flaw t""'E average remaining wall thickness coveri ng degraded area with through-wall leak bounded by dad; lmu* minim um wall thickness required for pressure loading tnom nom inal wall thickness tp minimum re maining wall thickness (.t maxunum c0nc ane.k at the ~!!nk1 of ,1 n:duca A. nondimensional half crack length for through-wall axial flaw

               ¢            ctrcum fcrenual ,mgle from elht>w c-r bend
                            .l1illk g              .\:srnl bcndtng srrt.:ss fur pnmarv    k'atlm~

a ax1.il thermal expans11._:.n slrcs~

             <If            material fl ow stress m,             pipe hoop stress due to pressure and hc.!ndmg mom~nt (ft,r elbows md hcnL

_.r.!J2£l ff_ A JffltA1ll lE'Re1tudrAAl h 'R ln~g ;tr**, lPr flFllPRi:: lae8iRg w1tl1@ett *~ra ~ tnta1 , 1ti aet1a1\ la2tar

             <TL            reference limit load hoop stress rr,            axwl nrt:'\sure ';lres.s
             <I7            material yield strength at temperature, as defined in C-4300
            '"~             Hiii<  I"""'    f. r the det,eted 11,,,,  = "

ta th' 1tlJO oAi'i< flA" >lo.t' IJ~I All t'Hee:~EitAg ~( Rh nrh_; trJm the IAtflAI

                            ,iuee\ ary aft!" eeAElit1eF1 B              ha lf crack angle for through-wa ll circum ferential flaw 7             APPLICABILITY TJ.i. C 1

l' Affh if.I* {nm 1"1* 19~' 1'aH1"R " 11.R ti.* u*mror 1*1~5 \88oR<la, rhrs*ig,h !Re '.)Vl7 b8n1rn I' 1th lRo JO>' '.l.l*Rd* Reference to Appendix C in J,d1tor s '\ore [*or .\ppltcabiltt\ lndc.,

this Case shall apply to Appendix C of the 2004 Edition 'lpnl1c,1hilit\ I"> fr\,_,m I l19h _\JJ!!n<l.1 Ul ir later ed1uons or addenda. For editions ttf>J..<ir addenda ) I_~ .* . . ltthm prior to the 2004 Edition, Class 1 pipe flaw -;;aluation procedures may be used for other piping classes. As a matter of defin ition, the current term "structural factor" is equiva lent to the term "safety factor," which is used in earlier editions and addenda. Draft 1- (05/05/ 14) 15 to 1700405.401.RO 1-16 ~Structural Integrity Associates, Inc."'

Attachments 1 - 3 RR5-02, Cooper Nuclear Station Page 17 of 26 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4.J MANDATORY APPENDIX I RELATIONS FOR Fm, Fb, AND F FOR THROUGH-WALL FLAWS 1-1 DEFINITIONS Ao = - 3.'.l6543 + 1.52784 (Rft) - 0.072698 (Rlt )'

                                                                                  + 0.00 160 11 (Rlt'j' For through-wall flaws, the crack depth._ ...La,* will       Bb = 11.36322 - 3.914 12 (Rlt) + 0.186 19 (Ric )'

be replaced with half crack length..._- c.4 in the stress - 0.004099 (R!t) 3 intensity factor equations in C-7300 and C-7400 of Cb = - 3.18609 + 3.84763 (Rlt) - 0. 18304 (Rli )1 Section XI, Appendix C. Also. Q will be set equal to + 0.00403 (R/t) 3 unity in C-7400. Equations for Fm and Fo are accurate for R/t between 5 1-2 CIRCUMFERENTIAL FLAWS and 20 and become increasingly conservative for R!t greater than 20. Alternative solutions for Fm and Fb may For a range of R/t between 5 and 20, the following be used when R/t is greater than 20. equations for Fm and Fb may be used: l-3 AXIAL FLAWS Fm = I + Am (Bltr)l.l + Bm(BltrY..' + Cm (B' tr)3 -' Fb = I + Ab (8/tr)u + Bb (8/tr)2*1 + Cb ((J;r)u For internal pressure loading. the following equallon for F may be used: where F = I + 0 072449). + 0.6485&' - 0.2327 ;l.3 e = half crack angle = c!R + O.Q38 I 54 :i.* - 0.0023487 J.' R = mean pipe radius t =~wall thickness where and Am = -2.029 17 + 1.67763 (Rft) - 0.07987 (Rlt )' c = hal f crack length

                     + 0.00 176 (R/t'f                                    ). = cl(Rt)112 Bm = 7.09987 - 4.42394 (R!t) + 0.21 036 (Rlt )'          The equation for F is accurate for :i. between 0 and 5
                     - 0.00463 (RI t)'                                Alternative solutions for F may be used when ). is Cm = 7.7966 1 + 5. 16676 (R/t) - 0.24577 (IM)2           greater than 5.
                    + 0.0054 1 (R!t)'

Drarr 15 (05/0511 4) 16 to 1700405.401.RO 1-17 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02, Cooper Nuclear Station Page 18 of26 Attachment 2 Technical Basis for Proposed Fourth Revision to ASME Code Case N-513 to 1700405.401 .RO 2-1 e Structural Integrity Associates, Inc.lg

Attachments 1 - 3 RRS-02, Cooper Nuclear Station Page 19 of 26 Proceedings of the ASME 2014 Pressure Vessels & Piping Conference PVP2014 July 20-24, 2014, Anaheim, California, USA PVP2014-28355 TECHNICAL BASIS FOR PROPOSED FOURTH REVISION TO ASME CODE CASE N-51 3 Robert 0 . McGill Guy DeBoo Structural Integrity Associates Exelon Generation Company San Jose, CA Warrenvil le, IL rmcg11l@struct1 nt com guv deboo@exeloncorp.com Russell C. Cipolla Eric J. Hou ston lntertek AIM structural Integrity Associates Sunnyvale, CA Sa n Jose, CA russell c1oolla@lntertek com ehouston@struct1nt com ABSTRACT Resultant !hernial expansion moment Elbow or bent pipe bend radius Code Case -513 provides evaluation rules and criteria Outside radius for temporary acceptance of flaws. including through-wall Flexib ility characteristic flaw s, in moderate energy piping. The application of the Code Stress intensification factor as defined in the Code of Case is restricted to moderate energy, Class 2 and 3 systems, Record for the pip ing item so that safety issues regarding short-term, degraded system Maximum operating pressure at flaw location operation are minimized. The first version of the Code Case Evaluation thickness was published in 1997. Since then, there have been three Axial bending stress revisions to augment and clarify the evaluation requirements Thermal expansion ;tress and acceptance criteria of the Code Case that have been Hoop s1ress published by ASME. The technical bases for the original Axial membrane slress version of the Code Case and the three re visions have been Circumferential angle previously published. There is currently work underway to incorporate INTRODUCTION additional changes to the Code Case and this paper provides Background the technical basis for the changes proposed in a fourth revision. These changes include addressing !he current Code Case N-513-3 [I] (currently approved revision) condition on !he Code Case acceptance by the US N uclear provides evaluation rules and criteria for the temporary Regulatory Comm iss ion (NRC), c larification of !he Code Case acceptance of llaws, inc luding through-wal l llaws, in applicability lin1 its and expansion of Code Case scope to moderate energy piping. The provisions of this Code Case are additional piping components. New llaw evaluation focused on preventing gross fai lure of the affected pipe for a procedures are g iven for through-wall flaws in elbows, bent temporary per iod. However, it also requires the piping system pipe, reducers, expanders and branch tees. Th~e procedures and adjacent equipment functionality be demonstrated for lost evaluate llaws in the piping components as if in straight pipe lluid inventory, spray mg and flooding caused by the leakage. by adjusting hoop and axial stresses to account for the The Code Case provides rules for !he evaluation of degraded geometry differences. These changes and their technical bases pipe and tube for a short operating period, with inspection and are described in this paper. monitoring requirements of the degraded condition as part of the overall integrity assessment. The application of the Code NOMENCLATURE Case is restricted to moderate energy Class 2 and Class 3 systems, so !hat the safety issues regarding short-term system B,, B2 Primary stress index deftned in ASME Section III for operation are minimized. Moderate energy piping is deftned the piping item as those piping systems where the maximum operating Do Outside diameter pressure and te mperature do not exceed 275 psig ( 1.9 tv!Pa) l Moment of inert ia based on I (degraded section) and 200°F (93°C), respectively. Mb Resultant primary bendi ng moment Copyrigh t 0 2014 by AS:NIE to 1700405 .40 l.RO 2-2 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 20 of 26 Currently. the scope of the Code Case is lim ited to to accept flaws . .. without perfonn ing a repa ir/replacement straight pipe with some prov ision for flaw evaluation into activ ity for a limited tim e. not exceeding the tim e to the next fittings for a short distance from the weld attachment to scheduled refueling outage." N ote that the word "'refueling straight pipe. There have been many instances where through- has been added for further clarification rega rding the nature wa ll leaks have been observed in pipe components ( i.e., and duration of the plant outage. elbows. reducers, expanders or branch tees) or bent pipe outside of the current Code Case scope . As a resull utilities This language is consistent with the introduction of NRC are forced into taking alternate actions (e.g. , repa ir, Generic Letter 90-05 [3] which the original N-513 was based. replacement, or making requests for NRC relief) that can be The second paragraph of 90-05 states : "Temporary non-code significantly more burdensome without any measurable repairs are applicable until the next scheduled outage increase in plant safety. The fourth revision of the Code Case exceeding 30 days, but no later than the next scheduled prov ides evaluation rules and criteria for the temporary refueling outage."' acceptance of flaws for these instances. In addition. other enhancements are included with the fourth rev is ion as Flaw Evaluation Criteria for Piping Components a nd Bent summ arized below. ~ Evaluation and acceptance criteria have been added to - Summarv of Code Case N-513-4 C hanges 513-4 for flaws in elbows, bent pipe, reducers, expanders and The list below summari zes the proposed changes included branch tees. A sim plified approach has been adopted based on in Revision 4 of Code Case N-5 13. ote that a brief reason the evaluations and results from the Second International for each change is included in parenthesis. Piping lntegrity Research Group (IPIRG-2) program reported in Reference [4]. The flaw evaluation for the piping Temporary acceptance period redefined (addresses component is conducted as if in straight pipe by scaling hoop NRC condition given in Regulatory Guide I 147 ['.!D and axial stresses using ASME piping design code stress Flaw evaluation criteria included for elbows, bent indices and stress intensificati on factors to account for the pipe. reducers, expanders and branch tees (scope stress variations caused by the geometric differenc es. In expansion) Refe rence [4], this approach was determ ined to be very Allow flaw evaluati on of heat exchanger tubing in conservative by comparing the failure moments predicted specific instances (scope expansion) using this approach to the measured fail ure moments from the Daily walkdown requirement for through-wal l leaks elbow tests for through-wall circumferential flaws. Details of provides additiona l flex ibility for user the simplified approach are given in the following sections. implementation (scope expansion) Limit scope to only Ii uid systems (scope Flaw Evaluation in E lbows and lknt Pipe. Through-clarification) wall flaws in elbows and bent pipe may be evaluated using the Treatment of Service Level load combinations (scope straight pipe procedures given in N-513-4 provided the clarification) stresses used in the evaluation are adjusted to account for geometry differences. The hoop stress for elbow and bent Treatment of flaws in austenitic pipe flw< welds pipe evaluation shall be (Equation 9 of N-513-4): (scope clarification) Minim um wall thickness acceptance criteria to consider longitudinal stresses in addition to hoop a = ( pD 0 ) [ 2R.... + R0 sin¢ ] + (~) R)vl* (9 ) 213 stress (scope clarification) ' 21 '.!(R..., + R, sin ¢) h I In addition, several editorial changes to improve the Equation 9 is only applicable for elbows and bent pipe clarity of the Code Case are included. where h 2: 0. I . CODE CASE N-513-4 CHANGES AND TECHNICAL The axia l membrane pressure stress for elbow and bent BASIS pipe evaluation shall be (Equation 10 ofN-513- 4): The fo llowing subsections provide details regarding each change and their technical basis. ( 10) Temporarv Acceptance Period

            -513-3 specifies a temporary acceptance period that               8 1 shall be equal to 0. 5 for e lbows and bent pipe .

could extent out to 26 months. The NRC did not endorse this maximum period length and in the latest re vision ofNRC The axial bending stress for elbow and bent pipe Regulatory Guide 1. 147, placed a condition on N-5 13-3 evaluation shall be (Equation 11 of N -5 13-4): stating, ** ... repair or replacement activity temporarily deferred under the provisions of this Code Case shall be perfonned during the next scheduled outage ."' -5 13-4 rr, = B, ( -R I-0 M, J (1 1) addresses this condition by remov ing the maximum duration limit and stating, " ... the follow ing requirements may be used Copyright 0 20 14 by AS.ME to 1700405.401.RO 2-3 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 21of26 Note that if the piping design Code of Record does not provide a 8 2. it may be taken from the ASNIE Section ill, 2004 Edition or later Editions and Addenda . CY

                                                                                                         = i(D.M 21
                                                                                                                 ,)                       ( 16)

The thermal expansion stress for elbow and bent pipe evaluation shall be (Equation 12 of -513-4) : Equations 13-1 6 are cons istent with the piping design by rule approach in ASME Section ill The reducer and expander zones shown in Figure 8 of 1 -5 13-4 are consistent with those

                              "* = {  R.~f,)                         ( 12)     shown in Code Case N-597-2 [8]. As with elbows and bent pipe, -513-4 allows for alternate methods to be used to calculate stresses used in the flaw evaluation.

The first term of Equation 9 accounts for the hoop membrane stress variation around the circum l~rence of the .Flaw Evaluatio n in :Branch Tees. B ranch reinforcement elbow (higher at the intrados, lower at the extrados assum ing a require ments shall be met in accordance wi th the Construction uni fonn wa ll thickness). The second tenn conservatively Code . [f the Construction Code did not req ui re reinforce ment, accounts for the through-wall bending in the hoop direction a reinforcement region is defined as a region of radius D, of that results fro m elbow ovalization due to in-plane or out-of- the branch pipe from the center of the branch connection for plane bendi ng and was taken from Reference [ 5). Equations evaluation purposes. Through-wall flaws in branch tees I 0- 12 are consistent with the piping design by rule approach outside of the reinforcement region may be evaluated using in ASME Section ill, NC/ND-3600 (6). the straight pipe procedures given in N-513-4 provided the stresses used in the evaluation are adjusted as described below It is recognized in Reference [4) that the simplified flaw to account for the geometry differences. Evaluation of flaws evaluation approach may be overly conservative in some in the region of branch reinforcement is outside the scope of instances. Thus, N-513-4 allows for alternate methods to be this Case. used to calculate stresses used in the flaw evaluation. For example, the French have developed a more comprehensive The hoop stress and axial membrane pressure stress for approach in analyzing flaws m cracked elbows. They have branch tee evaluation shall be determined from N-513-4 implemented finite element analysis based correlations to Equation 13 and 14, respectively. The outside diameter for determine location specific stresses for flaw evaluation. each of these equations shall be for either the branch or run Reference [7) provides additional details. pipe depending on the flaw location. The axial bending stress and thenn al expansion stress for branch tee evaluation shall be Flaw Evaluation in R educe rs and Expa nders. Through- determined from N-5 13-4 Equation 15 and 16. respective ly. wall flaws in reducers and expanders may be evaluated using the straight pipe procedures given in N-513-4 prov ided the As discussed previously, Equations 13-16 are consistent stresses used in the evaluation are adjusted to account for the wi th the piping design by rule approach in AS IE Section III. geometry differences. Figure 8 of N-5 13-4 illustrates the The limitation regarding flaw evaluation within the branch reducer and expander zones discussed below. Evaluation of reinforcement region is consistent with guidance given in flaws in the small end transition zone is outside the scope of Code Case N-597-2. As with elbows and bent pipe, N-513-4

       -513-4. The hoop stress and axial membrane pressure stress              allows for alternate methods to be used to calculate stresses for reducer and expander eva luation shall be :                            used in the flaw evaluation.

Flaw Evalua t ion of Heat Exc hanger T ubing (l3) N -513-4 allows for flaw evaluation in heat exchanger tubing provided the flaw can be characterized and the leakage monitored. The technica l basis is that knowledge of flaw ( 14) geometry is needed to evaluate the structural integrity of the tubing contai ning the flaw . An example application from plant operating experience would be for leaks in nonfe rrous Note that 0 0 is either the small end OD for the small end tubing of air cooling coils. The heat exchangers being of or is the large end OD for the large end. For the large end plate- fin design could have tubing accessible for volumetric transition zone and central conica l section. 0 0 is the large end inspection. OD Also. if the piping design Code of Record does not provide a 8 1, it may be taken from the ASNIE Section ill, Flexi bilih

in D aily W alkdown Requ irement 2004 Edition or later Editions and Addenda.

N-513-4 changes the daily walkdown requirement for The axia l bending stress and thermal expansion stress for leaking flaw s from ". leakage shall be observed by daily reducer and expander evaluation shall be : walkdowns .. .to** ... leakage shall be monitored daily . This change allows for other techniques to be employed in meeting the intent of the requirement instead of only physical _B' (D21M,) 0 ( 15) walkdowns. Such techniques could include remote visual Copyright 0 20 14 by ASNIE to 1700405.401.RO 2-4 e Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 22 of 26 equipment or leakage detection systems to dete rmine if and branch tees. In addition, severa l other scope expansions leakage rates are changing. and clarifications are included as well as editorial improvements. With these changes, it is believed the Code Scope L imited to .Liquid Svstem s Case wi ll be of even greater use to utilities in avoiding unscheduled plant shutdowns without negative ly impacting Provisions ofN-513-4 are now specifica lly lim ited to only plant safety. liquid systems. The methods ofN-513-4 should not be applied to the piping of air or other compressible fluid ACKNOWLEDGEMENTS systems. The deve lopment of Code Case N-5 13-4 was through the T reatm ent of Service Level Load Combinations efforts of the Working Gro up on Pipe Flaw Evaluation. The authors wish to recognize the efforts of all Working Group N-513 -4 now specifica lly requires all Serv ice Level load members especially the Working Group Chair, Dr. Doug combinations to be considered in flaw evaluations to Scarth. determine the most limiting. N-5 13-3 was previous ly silent as to what Service Leve l loading to consider. However, this REFERENCES requirement was im plied in prev ious Code Case revisions when it required the use of pipe flaw evaluation procedures ASME Code Case N -513-3 , "Evaluation Criteria for specified in the refere nced, non-mandatory appendices of Temporary Acceptance of F laws in Moderate Energy ASME Section XI. Appendix C in the 2002 Addenda and Class 2 or 3 P iping Section XI, Division 1," Cases of later editions and adde nda, and Appendix H for ferritic the ASME Boiler and Pressure Vessel Code, January materials in earlier editions, required all operating conditions 26, 2009. or Service Levels be evaluated. 2. Regulatory Guide 1. 147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," T reatment of Flaws in A ustenitic Pipe Flux Welds Revision 16. October 2010.

3. NRC Generic Letter 90-05, "Guidance for A reference to ASME Section XI, Appendix C, C-6320 Perform ing Temporary Non-Code Repair of ASME has been added to paragraph 3.1 (b) to address the instance Code Class I. 2, and 3 Piping," (June 15, 1990).

where a flaw in an austenitic pipe flux weld requires

4. Ki linski, I , Mohan, R. , Rudl and, D., Flem ing, M .,

eva luation. Such a flaw would require the use of Elastic O lson, R. , Scott, P., Brust, F, Ghadia li, N., Plastic Fracture Mechanics (EPFlvl) criteria instead of Limit Witkowski, G., and Hopper, A , "Fracture Behav ior Load criteria. Equation 1 of N-5 13-4 now includes a Z factor of C ircum ferentially Surface-Cracked E lbows," to acco unt for this specific application. NUREG/CR-6444, BMI-2192, Decem ber 1996.

5. Moore, S.E. , and Rodabaugh, E.C., "Background for When the original version ofN-5 13 was developed, Changes in the 198 1 Edition of the ASME N uclear Appendix C of AS.tvJE Section XI only included flaw Power Plant Components Code for Controlling evaluation criteria for austenitic piping. The change from N -

Prim ary Loads in Piping System s, " Journal of 5 13- 1 to -2 included the reference changes to the new Pressure Vessel Technology, Volum e 104, pp. 351 - Appendix C that combined austenitic and ferritic flaw 361, November 1982. evaluation criteria . The need to reference EPFM criteria for

6. ASME Boiler and Pressure Vessel Code, Section III, an austenitic tl ux we ld flaw eva luation was overlooked.

NC/ND-3600, 2004 Edition.

7. Marie, S., Chapuliot, S., Kayser, Y., Lac ire, M.H.,

M inimum W all T hickness Acce pta nce Criteria Drubay, B., Barthe let, B ., Le-De ll iou, P , Rougier, In estab lishing a minimum wall thickness acceptance V , Naudin, C., Gilles, P ., and Triay, M , "French criteria for paragraphs ' .2(b) and 3.2 (c) ofN-5 13-4. the RSE-M and RCC-MR Code Appendices for Flaw require ment to consider longirudinal stresses in addition to Analysis: Presentation of the Fracture Paran1eters hoop stress is added. While it is unlikely that a longirudinal Calculation - Part IV: Cracked E lbow," International stress based minim um wall thickness would be Limiting Journal of Pressure Vessels and Piping 84, pp. 659-compared to the hoop stress based minimum wall thickness, 686, May 2007. the user should be aware of this possibility and confim1 an 8. ASME Code Case N-597-2, "Requirements for appropriate Im ,. as an acceptance criterion. Analytical Evaluation of Pipe Wall Thinning, Section XI, Div ision l ," Cases of the AS.tvJE Boiler and CONCLUSIONS Pressure Vessel Code, November 18, 2003 . This docum ent serves as the technical basis for the fourth revision of Code Case -513 . This Code Case provides eva luation rules and criteria for temporary acceptance of tlaws, including throug h-wall flaws, in moderate energy C lass 2 or Class 3 piping. Currently, the scope of the Code Case is lim ited to straight pipe with some provision for flaw evaluation into fittings for a short distance from the we ld attachment to straight pipe . N -513-4 provides flaw evaluation rules and criteria for elbows, bent pipe, reducers, expanders Copyright 0 20 14 by ASME to 1700405.401.RO 2-5 l} Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02, Cooper Nuclear Station Page 23 of 26 Attachment 3 Cooper Nuclear Station Technical Basis for Proposed Alternative to Use ASME Code Case N-513-4 to 1700405.401 .RO 3-1 SJ structural Integrity Associates, Inc."'

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 24 of 26 BACKGROUND Code Case N-513-4 [1] provides evaluation rules and criteria for the temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. Moderate energy piping is defined as those piping systems where the maximum operating pressure and temperature do not exceed 275 psig and 200°F, respectively. The provisions of the Code Case are focused on preventing gross failure of the affected pipe for a temporary period while permitting leakage within the plant' s Technical Specification. The Code Case provides rules for the evaluation of degraded pipe and tube for a short operating period, with inspection and monitoring requirements of the degraded condition as part of the overall integrity assessment. The application of the Code Case is restricted to moderate energy Class 2 and Class 3 systems, so that the safety issues regarding short-term system operation are minimized. The previous Code Case revision, N-513-3 [2], is conditionally approved by the NRC in Regulatory Guide 1.14 7 [3]. The single condition deals with a requirement to perform the repair or replacement activity, temporarily deferred, during the next scheduled outage. Since the introduction of this Code Case, many utilities have used it as a basis for continued operation of degraded piping in moderate energy systems and that has resulted in significantly fewer relief requests to the NRC. Consequently, the industry has benefited from substantial cost savings while maintaining safety. To date, there have been no known instances where the use of the Code Case has resulted in any safety issues at the plants. The scope of Code Case N-513-3 is limited to straight pipe with some provision for flaw evaluation into fittings for a short distance from the weld attachment to straight pipe. There have been many instances where through-wall leaks have been observed in pipe components (i.e., elbows, reducers, expanders or branch tees) or bent pipe outside of the current Code Case scope. As a result, utilities are forced into taking alternate actions (e.g., repair, replacement or making requests for NRC relief) that can be significantly more burdensome without any measurable increase in plant safety. Code Case N-513-4 provides evaluation rules and criteria for the temporary acceptance of flaws for these instances. In addition, other enhancements are included with the fourth revision as summarized below.

SUMMARY

OF CODE CASE N-513-4 CHANGES The list below summarizes the major differences between the NRC-approved Code Case N-513-3 and Code Case N-513-4. Note that a briefreason for each change is included in parenthesis.

Temporary acceptance period redefined from no longer than 26 months to the next refueling outage (addresses NRC condition given in Regulatory Guide 1.14 7)
Flaw evaluation criteria included for elbows, bent pipe, reducers, expanders and branch tees (scope expansion)
Allow flaw evaluation of heat exchanger tubing in specific instances (scope expansion)
Daily walkdown requirement for through-wall leaks provides additional flexibility for user implementation (scope expansion)
Limit scope to only liquid systems (scope clarification)
Treatment of Service Level load combinations (scope clarification) to 1700405.401.RO 3-2 e Structural Integrity Associates, Inc.

l Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 25 of 26

Treatment of flaws in austenitic pipe flux welds (scope clarification)
Minimum wall thickness acceptance criteria to consider longitudinal stresses in addition to hoop stress (scope clarification)

In addition, several editorial changes to improve the clarity of the Code Case are included. A detailed markup of Code Case N-513-3 showing all of the changes made for Revision 4 is provided in Attachment 1. TECHNICAL BASIS The technical basis for the changes listed above is given in an ASME Pressure Vessel and Piping (PVP) conference paper [4] and is provided in Attachment 2. PRECEDENTS There have been several submittals approved for N-513-4 use in specific applications and for generic use by Exelon [5]. The table below lists several Safety Evaluation Reports as precedents for use of Code Case N-513-4. Table 1: Code Case N-513-4 Precedents SER Accession No. Plant Application Additional Requirements ML16230A237 Exelon Fleet Generic N-513-4 Critical leakage determination MLl 5070A428 ANO Leaking sweepolet 5 gpm leakage limit ML14316Al67 Fort Calhoun Leaking elbow None ML14335A551 Peach Bottom Leaking elbow 5 gpm leakage limit ADDITIONAL REQUIREMENT Consistent with the previous NRC N-513-4 safety evaluation for the Exelon fleet Relief Request [5] , the following limitation on leakage is included. For a leaking flaw, the allowable leakage rate will be determined by dividing the critical leakage rate by a safety factor of four (4). The critical leakage rate is determined as the maximum leakage rate that can be tolerated. The critical leakage rate may be based on the allowable loss of inventory or the maximum leakage that can be tolerated relative to room flooding, among others. The safety factor of four to the critical leakage rate is believed adequate to ensure system operability and allow for early identification of conditions that may lead to adverse consequences. to 1700405.401.RO 3-3 SJ Structural Integrity Associates, Inc.

Attachments 1 - 3 RR5-02 , Cooper Nuclear Station Page 26 of 26 CONCLUSIONS This attachment provides a technical basis for proposed alternative use of ASME Section XI Code Case N-513-4. The discussion provided herein demonstrates that the use of Code Case N-513-4 will reduce plant burden without any adverse effect on safety. Consistent with a previous NRC safety evaluation of a generic Code Case N-513-4 request, an additional leakage limitation requirement has been included. REFERENCES

1. ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division I ," Cases of the ASME Boiler and Pressure Vessel Code, May 7, 2014.

2. ASME Code Case N-513-3 , "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l ," Cases of the ASME Boiler and Pressure Vessel Code, January 26, 2009.

3. Regulatory Guide 1.14 7, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division I ," Revision 17, August 2014.

4. McGill, R.O., Cipolla, R.C ., DeBoo, G. and Houston, E.J., "Technical Basis for Proposed Fourth Revision to ASME Code Case N-513," Proceedings of the ASME 2014 Pressure Vessels & Piping Conference, PVP2014-28355 .

5. NRC Relief Request Approval and Safety Evaluation Report for Exelon Generation Company Fleet Request for Proposed Alternative to Use ASME Code Case N-513-4, ADAMS Accession No. ML16230A237. to 1700405.401.RO 3-4 e Structural Integrity Associates, Inc.

NLS2017071 Page 1of34 Enclosure 2 RR5-03 , Cooper Nuclear Station Structural Integrity Associates, Inc. Final Code Case N-513-3 Markup for Revision 4 Technical Basis for Proposed Fourth Revision to ASME Code Case N-513 Cooper Nuclear Station Technical Basis for Proposed Alternative to Use ASME Code Case N-513-4 and Scope Expansion to a Higher Pressure Limit Note: Enclosed report contains 33 pages not counting this cover page.

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 1 of 33 Attachment 1 Final Code Case N-513-3 Markup for Revision 4

Atta chments 1 - 3 RR5-03 Cooper Nuclear Station Page 2 of 33 Record# 12-841 CASE CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4J Approval Date: January 26, 2009 Code Cases will remain available for use until annulled by rhe applicable Standards Committee. Case -513-9 C'l The p1pmg <.b1gn <..ode sha ll be used m Evaluation C riteria fo r Tem porary Acceptance of determinmg the stress md1ces B and B and stre,;,, Flaws in Moderate Energy C lass 2 or 3 Pipi ng mtens1f1callon factor / for Oaw evaluation following Section XI, Division 1 Code anplicab1ht1 lim its m term s of component

                                                                  <teomelf\ such as D.,J,,.,m ra tio ff the piping design Inquiry: What requirements may be used for                  Code does not prov ide stress indices, Secllon lH . 2004 temporary acceptance of flaw s, including through-wall             Ed1t10n or later Ed1llons and Addenda may be used to flaws, in moderate energy Class 2 or 3 piping mc\udmg              defme B and B elbows. bent pipe, reducers. expanders and branch tees,                  ([i i) The provisions of this Case demonstrate the without performing a repair/replacement activity?                  integrity of the item and not the consequences of leakage. It is the responsibility of the Owner to Reply: It is the opinion of the Comm ittee that the          101" aRs!Fal* s;**l*i:* e~ l Fee1li1y consider.,;g effects of fo llowing requirements may be used to accept fl aws,              leakage in demonstrating svstem operabl11\) and including through-wall flaws, in moderate energy Class             perform ing plant llooding analvses.

2 or 3 piping mc\udmg dbows. bent pipe reducer; (ti Hil "' ahiet1eR ~en ea 1 ' - i.; !he Sfl<ret1e11al expanders. and branch tees, without performing a time feF ,;h1 eh !he lefl1~eF0f)* eee l~laRe o enlrna eFo repair/replacement activity for a lim ited tim e, not ;atl!;liea bill Ret e'teeetl1Rg 20 mtlR!hs frem !ho Mdl!sl exceeding the l'iahiallaR ~uteEI as EleflfllEI 111 th~; d1see 1*ef)* e fthl ee11ame11. Atsel!me to the next scheduled refuelm g outa ~e. 2 PROCEDURE SCOPE (a) The flaw geometry shall be characterized by (a) These requirements apply to the ASME Section volumetric inspection methods or by physical III, ANSI B3 l . I, and ANSI B3 l .7 piping, classified by measurement. The full pipe circumference at the flaw the Owner as Class 2 or 3 that 1s access ible for location sha ll be inspected to characterize the length and 111:>p ecllon. The prov is ions of this Case do not apply to depth of all flaws in the pipe section. the following : (b) Flaw shall be c lassified as planar or nonplanar. (1) pumps, valves, expansion joints, and heat (c) When multiple flaws, including irregular exchangers except as pron ded m lb! ; (compound) shape flaws, are detected, the interaction (2) weld metal of socket welded joints; and combined area loss of flaws in a given pipe section (3) leakage through a flange joint; shall be accounted for in the flaw evaluation. (4) threaded connections employing (d) A !law evaluation shall be perform ed to nonstructural seal welds for leakage protection. determine the conditions for flaw acceptance. Section 3 lb! This Case ma\ be applied to heal exchanger provides accepted methods for conducting the required external tub mg or pir mg provided the fl aw 1s analysis. charactenzed m accordance wtth 2ca*1 and leakage 1s (e} Frequent periodic inspections of no more than momtoml 30 day intervals shall be used to determ ine if fl aws are (f.b) The provisions of this Case apply to Class 2 or growing and to establish the tim e~_,..._a l which the 3 piping m liyu1d svstems whose maximum operating detected fl aw will reach the allowable size. temperature does not exceed 200"F (93°C) and whose Alternatively, a fl aw growth evaluation may be maximum operating pressure does not exceed 275 psig performed to predict the time,----;......., at which th.e (1.9 MPa) detected fl aw will grow to the a llowable size. The flaw (!if!;) The following flaw evaluation cnteria are growth analysis shall consider the relevant growth perm itted for pipe and tube mcludmg elbows. bent pipe. mechanisms such as general corrosion or wastage, reducers. expanders. and branch tees. The straight pire fatigue, or stress corrosion cracking. When a flaw flaw evaluation criteria are permitted for adjoining growth analysis is used to establish the allowable time fi ttings and flanges to a distance of (R0 t)" from the weld for temporary operation, periodic examinations of no centerline . more than 90 day intervals shall be conducted to verify the flaw growth analysis predictions. Draft 15 (05105114)

Attachments 1 - 3 RRS-03 Cooper Nuclear Station Page 3 of 33 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-:!;l (/) For through-wall leaking flaws, leakage shall be When through-wall axial flaws are eva luated, the eesep; ea e:*monilored daily '>"B lliB0WRS to confinn the allowable flaw length is: analysis conditions used in the evaluation remain valid. {g) If examinations reveal flaw growth rate to be unacceptable, a repair~replacement acti> 1tv shall be performed. td,=u&./R![((sF~)zaJ-T (I) (h) Repair-eF-Lreplacement activities shall be performed no later than when the predicted flaw size a , = pD0 /2t ~ (2) from either periodic inspection or by flaw growth analysis exceeds the acceptance criteria of 4, or during a, = (s, +s,) 12 ~ (3) where the next scheduled rrliJ..cling_outage, whichever occurs p = pressure for the loading condition ~: Z has been firsL Repair~replacement activities shall be in added to equation Do = pipe outside diameter accordance with IWA-4000 er l"'A 7009. res1ieet1.,* e1~* . (! ). a1 = flow stress iA Bditiens aml AdaeflEla ~ rier ta ll<l e 199 1 AdEl...,Ela.

                                                                    .s;. =Code specified yield strength anEI. 1R !Re 1991 \aEleA Ela anEI later. iR aeee*Elatiee "' i~i S, =Code specified ultimate tensile strength and pv' IQOO.

SFm= structural factor on primary membrane stress (i) Evaluations and examination shall be as specified in C-2622 documented in accordance with IWA-6300. The Owner shall document the use of this Case on the applicable Z = load multiplier for ductile flaw mensioo

                                                                           !equal to l. Owhen using limit load criteria )

data report form. Material properties at the temperature of interest 3 FLAW EVALUATION shall be used. Planar flaws tn strai2ht pipe shall be evaluated in accordance with the requirements in 3.1. Nonplanar flaws in straight pipe shall be evaluated in accordance with the requirements in 3.2. Through-wall llaws in elbows and bent pipe shall be evaluated in accordance with the requirements in 3.3. Through-wall llaws in reducer>. expander., and bran ch tees shall be evalualed ' --' in accordance with the requirements in 3.-l and 3.5 respectively . Flaw growth evaluation shall be perfo1med in accordance with the requirements in 3.§.>. Nonfe1rous materials shall be evaluated in accordance with the

                                                                                        -"---~\

(c) For planar flaws in ferritic piping, the evaluation procedure of Appendix C shall be used. Flaw depths up requirements in 3.2+. to l 00% of wall thickness may be evaluated. flID:\' For all !law evaluauon. all Service Level load deoth a 1* defined in figures C'-4 31 0-1 and C-43 10-2 combinations shall be evaluated to determine the most When through-wall circumferential flaws are evaluated limiting all owable flaw size. in accordance with C-5300 or C-6300, the flaw depth to thickness ratio, alt, shall be set to unity. When applying 3.1 Planar Flaws in Str aight Pip e the Appendix C screening criteria for through-wall axial (~ For planar flaws, the flaw shall be bounded by a flaws, alt shall be set to unity , and the reference limit rectangu lar or circumferential planar area in accordance load hoop stress, 0 1, shall be defined as o/ Jvl,.. When with the methods described in Appendix C. IW A-3300 through-wall axial flaws are evaluated in accordance shall be used to determine when multiple proximate with C-5400 or C'-MOO the allowable length is defined flaws arc to be evaluated as a single flaw. The geometry by eqs. ( L) through (3), with the appropriate slructural of a through-wall planar flaw is shown in Fig. I. factoro from Appendix C, C-2622 When through-wall {b) For planar flaws in austenitic piping, the flaws are evaluated in accordance with C-7300 or C-evaluation procedure in Appendix C shall be used. Flaw 7400, the fonnulas for evaluation given in C-4300 may depths up to LOOOA> of wall thickness may be evaluated. be used, but with va lues for Fm, Fo, and F applicable to When through-wall circumferentia l flaws are evaluated, through-wall flaws. Relations for Fm, Fi. and F that take the formulas for evaluation given in C-5320 or C-6320. into account flaw shape and pipe geometry (Rlt ratio) as applicable of Appendix C may be used, with the flaw shall be used. The appendix to this Case prov ides depth to thicknes s ratio, alt, equal to unity. equations for Fm. Po, and F for a selected range of Rft. Geometry of a through-wall crack is shown in Fig. I. Draft 15 (05/05/14) 2

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 4 of 33 CASE (co ntinued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513--'J FIG. 2 SEPARATION REQUIREIVENTS FOR ADJACENT THINNED AREAS 3.2 Nonphuur Flaws in Straight Pipe defmed in Fig. 3. When the above requirement is not __fa) The evalmtion shall consider the depth and satisfied, (~ shall be met extent of the affected area md require thit the wall (;Ji~) When L., is less than or equal to 2.65 thickness exceed tmm for a distance th!t is the greater of (Rot..,,,)112 and tnom is greater than 1.13 tmm , ta1oc is 2.5 "Rt..,., or U,,,,,.,. between adjacent thinned regions, determined by satisfying both of the following where R is the mean radius of the piping item based on equations: nominal wall thickness and Lm.a-.g is the average of the extent of Lm below t.,,;" for adjacent a-eas (see Fig. 2). Alternatively, the adjacent thinned regions shall be ~t ~ ~1 1.5 " " [ - ....!!!!!!.. t ] + l .O (5) considered a single thinned region in the evaluation. t_ L t_ __jb) For nonplanar flaws, the pipe is acceptab le when eithorr !b)(l) and (bH 2l. or (b)(2) and !bl(31 are t."" ~ 0.353L,,, (6) met. --1.llJ:. e remaining pipe thickness (tp) is greater t..., JRot,,,. than or e<pal to the minimum wall thickness t..,": When the above re<pirements a-e not satisfied, (41_D shall be met t _ PL\, (4) (il.>.4) When the requirements of (.,<.,j , (~. md m 2(S +0.4p ) (~.ill above are not sltisfied, ta1oc is determined from Curve 2 of Fig. 4. [n eaaille1t . f - Jlmll :!81i~* tfte where fella .tng eftuetien . p = mrocimum operating pressure at flaw locition S = allowable stress at operating temperlture 12/ The remaining degraded pipe section meets the longitudinal stress limits in the design Code for the P.iPi.!!& ~~is the naminal )ll~e leng1lueli11al eElleling sire"' Iii As an alternative to 1b)(ll,\lte1'HBlt*el) , an evaluition of the remaining pipe thickness It,,) may be

                                                              "'"'llin,,. ft em all Se ... iee Le, el B
                                                              ~
                                                                                                                  " "II!'  !'ii'*

performed as given below. The evaluit ion procedure is (c) When there is through-wall leakage along a a function of the depth and the extent of the affected portion of the thinned wall, as illustrated in Fig. 5, the area as illustrated in Fig. 3. flaw may be evalu!ted by the branch reinforcement {L-/.) When Wm is less than or equal to 0.5 method The thinned area including the through-wal l (Rol)112 , where Ro is the outside radius and Wm is defmed opening shall be represented by a circular penetration at in Fig. 3, the flaw can be classified as a planar flaw and the flaw location. Only the portion of the flaw lying evaluited in accordance with 3.l(a) through 3.l(c), within ta1; need be considered as illustrated in Fig. 6. above. When the above requirement is not satisfied, When evaluating mu ltiple flaws in accordance with (-rji) shall be met. 3.2(a), on ly the portions of the flaws contained within L_..,) When Lm~) is not greater than (Rolmm )Ill, tad/ need be considered. taJoc is determined from Curve 1 of Fig. 4, where Lm~) is Draft 15 (05/05/14)

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 5 of 33 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL C ODE N-513-~ FIG. 3 ILLUSTRATION OF NONPLANA R FLAW DUE TO WA LL THIN NING i Lm1n Transve rse lclrcumferentlall Ax ial direction direction Editor's N ote. This Figure 3 ts to be delete d and replaced with the Figure 3 on the foll owing page Draft 15 (05105114) 4

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 6 of 33 CASE (continued) CASES O F ASl\llE BOILER AND PRESSURE VESSEL CODE N-513-4J FIG. 3 ILLUSTRATION OF NONPLANAR FLAW DUE TO WALL THINNING t lmm i

                                    -     L,,,1*1-Transverse (circumferential) direction FIG. 4 ALLOWABL E WALL THICKN ESS AND LENGTH OF LOCALLY THINNED AR EA

0.6

_E ) 0.4 0.2 f----+-- 0 '----~--~----"-----"'-----------'-----" 0 6 7 8 Draft 15 (05/05/1 4)

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 7 of 33 CASE (continued) CASES OF ASME BOil..ER AND PRESSURE VESSEL CODE N-5134.! FIG. 5 ILLUSTRATION OF THROUGH-WAL L NONPLANAR FLAW DUE TO WALL THINNING Through-wall

                                    /   opening r

l...,1rc Tra111sverse (circumferential) Axial di rection { Editor's Note Tlus Figure 5 is to be deleted and replaced with lhe Figure 5 on lhe fullowmg page Draft 15 (05/05/14)

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 8 of 33 CASE (continued) CASES OF ASME BOILER AND PRESS URE VESSEL CODE N-513-4.J FIG. 5 ILLUSTRATION OF THROUGH-WALL NONPLANAR FLAW DUE TO WALL THINNING Throug h-wall tm .in / opening t l

 *r::_~          ~~.                     ~
                            -    Laxia1-Axial direction I                                   Lcirc Transverse (circumferential) direction Draft 15 (05/05/ 14)

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 9 of 33 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4J The minim um wall thickness, t,. ,,., shall be are ad1usted to account for the geomdrv differences as determ ined by eq. (4). For evaluation purposes, the Jescnbed belo\\ Alternative metho<ls ma' be used to adjusted wall thickness, f ad;, is a postulated thickness as calculate the stresses used m evaluauon shown in Fig. 6. The pipe wall thickness is defined as The hoop stress a for dbov. and bent pme the thickness of the pipe in the non-degraded region as evaluation shall be. shown in Fig. 6(a). The diameter of the opening is equal to d.d1 as defmed by l ad; as shown in Fig. 6(a). The postulated value for t ad; shall be greater than t,. 0 , and a

                                                                       =( pD*)[

2t 2R...,+Rsin¢ ]+(~) 0 2(R,.., + R sin¢) h'" R.M, 9 l --1.21 shall not exceed the pipe wall thickness. The f ad; value 0 may be varied between Imm and the pipe wall thickness to determine whether there is a combination of laJ; and d.d1 that satisfies the branch reinforcement requirements. The values of l ad; and d,,d; of Fig. 6(b) shall satisfy : R ...0 elbow or bent pipe bend radtu' rp C!fcumferentml anl!lc defined m figure 7 h flex1bilit\ charactensuc Cl~) M,. resultant pnmary ben<lmg moment I moment of menta based on evaluation wall thickness f The remaining ligament average thickness, 1""'"' over the degraded area bounded by d.d; shall satisfy : Euuat10n 9 ts only apphcabk for elbov.s and bent pipe where h - 0 I (]9) The ""'"I membrane pres.,ure stress a, for elbow and bent pipe evaluatton shall be In addition the pipe sect10n mcludmg the equ11*aknt hole representation shall meet the lon1utud mal stress limits m the design Code for the ll!l2!!l&. __If a flaw growth analysis is performed, the growth v.here B. is a pnm arv stress mdex as defined m Secuon in flaw dimensions shall consider the degradation lll for the pmml! item B sha ll be equal to 11 5 for mechanism s as relevant to the application. The flaw is elbows and bent pipe acceptable when there is sufficient thickness in the The axial bending stress, a ., for elbo" and bent degraded area to provide the required area pipe evaluation shall be reinforcement. (d) A lternatively, if there is a through-wall opening along a portion of the thinned wall as illustrated in Fig. --

a = B,- ( D.M 21

                                                                                                      ' )
                                                                                                          ------~~

II ' 5 the flaw may be evaluated as two independent planar through-wall flaws, one oriented in the axial direction and the other oriented in the circumferential direction. where B is a prun ary stress index as defined m Section The minimum wall thickness lmm, shall be determ ined lll for the pmmg item by eq. (4). The allowable through-wall lengths in the The themial expansion stress a for elbow am! a'Cial and circum ferential directions shall be determined bent pme evaluation shall be by varying lad; shown in Fig. 5 from !,,.,. to lm m. The allowable through-wall flaw lengths based on tad; shall be greater than or equal to the corresponding L.,,.1 and - - a* = i( D.Alf,)------~~'-= 21 ~ L"" (see Fig. 5) as determined from 3. l(a) and 3. l(b) or

3. l(c), as appropriate . The remaining ligan1ent average thickness, le.avg, over the degraded area bounded by L.,,.i and L"" shall satisfy eq. @!)).

stress mtens1tical!on factor as d~fmcd in th~ 3.3 Th rough -wa ll Flaws in Elbow' an d Bent Pipe <les1im \o<le for the pmmg item Through-wall flaws m elbows and bent pme ma' be .If. resultant !hernial exoans1on moment evaluated usmg the straight pme procedures given m 3 I or 3.2(d) pro,*1ded the stres.~es usecl m the evaluation Draft 15 (05/05/1 4) 8

Attachments 1 - 3 RRS-03 Cooper Nuclear Station Page 10 of 33 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4.1 3.4 T hrough-wa ll Flaws in R educers a nd [he axial bending stress u, and !hernial expansion Expanders stress. u, for branch tee evaluation shall be tletem11ned Throu!!h-wall llaws m reducers and expanders ma' from eg (1 'land eu (16). respecuveh be evaluated usmg the stra1!!ht pme procedures given in 3 l or 3.2(d) provided the stresses used m the 3.@ Flaw G rowth Evaluation evaluallon are adjus1etl to account for the e.eometr- If"7i flaw growth analysis is performed, the growth diffrrences as descnbed below Alternative methods analysis shall consider both corrosion and crack-growth mav be used to calculate the stresses used m evaluation mechanisms as relevant to the application. Fig 8 ti Justrates the reducer and expander zones In performing a tlaw growth analysis, the discussed below. Evaluation of llaws in the small end procedures in C-3000 may be used as guidance. translllon zone is outside the scope of this Case. Relevant growth rate mechanisms shall be considered. The hoop stress u,, and axial membrane pressure When stress corrosion cracking (SCC) is active, the stress u,,. for reduce r or expander evaluat10n shall be fo llowing growth rate equation shall be used : a-

            - * = u -~

(PD.) (131 where da/dt is flaw growth rate in inches/hour, Km ox is ___ a-~= B ~ )-----~l-1~ 1( the maxim um stress intensi ty factor under long-term steady state conditions in ksi in. 0 ~, S r is a temperature where D, is the small end OD for flaws m the small end correction factor, and C and n are mater ial constants. and the lame end OD for all other flaws The axia l bendtnl! stress. u.. and thermal expansion For intergranular SCC in austenitic steels, where T S stress, u. for reducer or expander evaluatton shall be: 200°F (93°C). C = 1.79 x 10-s

            ---      a-b = B2 ( D*M*)

21 ----~'-"L,_ ,_' Sr n

                                                                         =I
                                                                         = 2. 161

_ __ a-, ={ D;~* ) -------'-("'"'lb'-"-) For transgranular SCC in austenitic steels, where T S 200°F (93°C). where I ts based on the de11.raded sect10n C = l.79x 10- 1 3.5 Through-wall F laws in & *anch Tees Sr = 3.7 ! x I O" [I O(o.oi842 r- i2.2')] Branch reinforcement requirement' shall he met m n = 2. 161 accordance with the design Code If the design Code d id not reqmre remforcement for eva luation purposes. a The tem perature, T, is the metal temperature in n:mforcement re11.ion is defined as a regmn of radius D degrees Fahrenheit The flaw growth rate curves for the o[ the branch pme from the center of the branch above SCC growth mechanisms are shown in Figs. 2.+ connecllon Through-wall tlaws m branch tees outskle and l!_>>:. Other growth rate para meters in eq. ( l li.l) may of the reinforcement region may be ernluated using the be used, prov ided they are supported by appropriate strai11.ht pme procedures given m 3 1 or 3 2(dl prov ided data. the stresses used m th~ evaluat10n are ad1usted to account for the geometrv differences as described 3.24 Nonferrous Materials below. Altemauve methods mav be used to calculate the For nonferrous materials, nonplanar and planar stresses used in evalua11on. Evaluatton of tlaws in the flaw s may be evaluated fo llowing the general approach re>!iOn of branch reinforcement IS OUtslCk the scope of of 3.1 through 3 . ~ For planar flaws in duc tile this Case materials, the approach given IA J I (b 1 RRd 3 Jfor The hoop s'tress U;, and axial membrane pressure austeruttc pipe may be used; otherwise, the approach stress, u~ for branch tee evaluation shall 11<'. determined given "" J l f81 aRil J Jfor femtic pine should be from eq. (13) and eg (1-1) respecuveh The outside applied. Structural factors provided in 4 shall be used. It thameter for each of these equat10ns shall be for the is the responsibility of the evaluator to establish branch or run pipe, dependmg on the flaw location conservative estimates of strength and fracture touglmess for the piping material. Draft 15 (05/0511 4) 9

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 11 of33 CASE (co ntinued) CASES OF ASME BO ILER A:'ID PRESS URE VESS EL COD E N-513-..i.l FIG. 6 ILLUSTRATION OF ADJUSTED WALL THICKNESS AND EQUIVALENT HOLE DIAMETERS Through-wall T Pipe CU-"'-"-"'-"..::..J..+-------+....L:..-"'-"'-"'-"'~lwall lal ~djusted Wall Thlck~ess

                                                                   '                  Shift fi gure (b ) lo the
                                              ,                                       right so that dadj TItm,,.

width lines up with

                         +m-i----------a r:.~::e-d--.~

t figure (a). f - - - dod/ --l (bl Equivalent Hole Representation FIG. 7 CIRCUMFERENTIAL ANGLE DEFINED Dra ft 15 (05/05114) 10

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 12 of 33 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513--lJ. FIG. SZONES OF A REDUCER OR EXPANDER l arge e nd transition zone transition zo ne GENE RAL NOTE: Transit io~ zones exten d from the poim o n the ends where th e diameter begins to c hange to the pomt on the cen tra l cone where the cone ang le is co nstan t. Draft 15 (05/05/ 14) 11

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Pag e 13 of 33 CASE (continued) CASES OF ASME BOILER AND PRES.SURE VESSEL CODE N-513- 4~ FIG. ~ FLAW GROWTH RATE FOR IGSCC IN AUSTENITIC PIPING H Auslontti: Plplr>~ T <.lOCf lr.E-43

1.CE-04

v

                                                           /
                                                      /
                                              /

1.CE.-07

                                        /
                                 /
                               /

1£E .Ql3 I 10 ICO Draft 15 (05/05/1 4) 12

                                                                                                                              --1 Atta chments 1 - 3 RR5-03 Cooper Nuclear Stati on Page 14 of 33 CASE (continued)

N- 513- ~ CASES OF A SME BORER AND PRESSURE VESSEL CODE F I G . ~~ FLAW GROWTH RA TE FOR T GSCC IN A USTEN ITIC PIPING U E-02 Fl AIKIW"IU,:; flt pln;J T ::.: .2D~ / UE.JJ3

                                                                              /
                                                                           /

l.CE44 v ii

                     ~                                          V'                     v
                     ~ 1.CE-'lS
                      ..           ====1    T* l>ltf
                                                        /                /

J

[ T

IOCf 1.CE.06

                                               /                      /

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                                     /
                                                    /

1.CE-08 v

1.CE-00 I 10 Dru t 15 (05/05/14) 13

Attach ments 1 - 3 RR5-03 Cooper Nuclear Station Page 15 of 33 CASE (continued ) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4.1 4 ACCEPTANCE CRITERIA nondim ensional stress intensity factor for through-wall circum ferential fl aw under Piping containing a circumferential planar flaw is membrane stress acceptable for temporary serv ice when flaw evaluation moment of mertia based on evaluauon provides a margin using the structural factors in thickness I Appendix C, C-262 1. For axial planar flaws, the maximum stress intensity factor wider structural factors for tem porary acceptance are as long term steady state conditions specified in Appendix C, C-2622. Straight rPi~<Rg L maximum extent of a local thinned area containing a nonplanar part.::-through-wall flaw is with t < l.om acceptable for temporary service if the remammg pme L axial length of idealized through-wa II planar section meets the long1tu<lmal stress limits m the design flaw opening in the axial direction of the Code for the p1pmg and lp !! t.J"" where laJoc is pipe, as illustrated in F ig. 5 detennined from 3.2(b) . Straight oPiJ)£<Rg containing a length of idealized through-wall planar nonplanar through-wall flaw is acceptable for temporary flaw opening in the circumferential service when the flaw conditions of 3.2(c) or 3.2(d) are direction of the pipe, as illustrated in Fig. 5 satisfied. An elbow or bent pipe contamml! a nonplanar maximum extent of a local thinned area through-wall flaw "acceptable for temporar' service 11 with t < lmin the llaw comht1ons of 3 3 are sallsfied A reducer or l.. (o) axial extent of wall thinning below lmm expander contammg a nonplanar through-wall fla\\ ts L..111 circumferential extent of wall thinning acceptable for temporao sen*1ce 1f the flaw condi!tons below lmm of 3 .J are sallsfied A branch tee contammg a nonplanar L.. .~. average of the extent of L.. below through-wall llaw is acceptable for temporary service if lmm for adjacent thinned areas the flaw conditions of 3 5 are satisfied. maximum extent of thinned area, i bulging fac tor for axial flaw 5 AUGMENTED LXAMINATION resultant [lrtmarv bending moment

                                                            .\I     resultant thcnnal expansion mom~nt An augm ented volumetric examination or physical        R       mean pipe radius measurem ent to assess degradation of the affected           ii_     elbow or bent pipe centerlme bend radius system shall be performed as follows :                       R,      outside pipe radius (a) From the engineering evaluation, the most           s       allowable stress at operating temperature susceptible locations shall be identified A sample size      SFm     structural factor on prim ary membrane of at least five of the most susceptible and accessible              stress locations, or, if fewer than five , all susceptible and              coefficient for temperature dependence in accessible locations shall be examined w ithin 30 days of            the crack growth relationshi p detecting the flaw .                                                 Code-specified ultimate tensile strength (b) When a flaw is detected, an additional sample               Code-specified yield strength of the sam e size as defined in 5(a) shall be exam ined.             metal temperature (c) This process shall be repeated within 15 days               maximum extent of a local thinned area for each successive sample, until no significant flaw is             perpendicular to Lmwith t < Imm detected or until I 00% of susceptible and accessible                minimum distance between thinned areas i locations have been exam ined.                                       and}

L load mult1pl1er for ducllk fla\\ extension 6 NOMENCLATURE a flaw depth c half crack length B.,B Secuon 111 rnmarv stre" m<lices daldt flaw growth rate for stress corrosion c coefficient in the crack growth relationship cracking D 1n.s1de pipe diameter diameter equivalent circular hole at 1,.1 D, outside pipe diameter diameter of equivalent circular hole at F nondim ensional stress intensity factor for lm m through-wall axial flaw under hoop stress It flex1b1htv charactenstK nondimensional stress intensity factor for stress mten.~ificat1on factor through-wall circumferential flaw under total crack length = 2c pipe bending stress allowable axial through-wall flaw length exponent in the crack grow th relationship Draft 15 (05/05/14) 14

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 16 of 33 CASE (continued) CASES OF ASME BOILER AND PRES.SURE VES.SEL CODE N-513-4~ p maximum operating pressure at flaw location eva luation wall thickness surrounc.hng the de2raded area tad1 adjusted wall thickness which is varied for evaluation purposes in the evaluation of a through-wall nonplanar flaw t.,"" allowable local thickness for a nonplanar flaw t,.w8 average remaining wall thickness covering degraded area with through-wall leak bounded by d.d} tm~ minimum wall thickness required for pressure loading tnom nominal wall thickness tp minimum remaining wall thickness a maximum cone anl!l e at the center of a reducer A. nondimensional half crack length for through-wall axial flaw If! circum ferenual angle from dbow or bend llimk (J,, axrnl bendmg stress for pnm mv loading rI axial thermal expansion stress rIJ material flow stress rih pipe hoop stress due to pressure and bending moment (for elbows and bent

6, ReniRal leR51t1<8!Ral 13sRd* g .-1ra:s !er 13 riffi e ~

lsaelffig w itRsHt S~Fe1 0t; I F1h!R ~ l~I ea!ieR fe ater a1 reference lim it load hoop stress a:r; ax.ml prc!ssure stress riy material yie ld strength at te mperature, as defined in C-4300

 .._          !lllle rettttHetl fe r!ke tleteetetl Rs .,* to g re ..

le tka a ll e .. ee le Raw .nu. li ttt Rel a <eeea1A3 ~0 '" e Atks tfe1i1 ;!;a 1m1ial Elwee*1eFy ef tfie eeREllHeR B half crack angle for through-wall circumferential flaw 7 APPLICABILITY Hu* Gao

ts Bf"phaael* freR1 ti>* 19gJ J:le1tteR ,,. ,11>

IA* ll mt*r 19g? ." EleaR<la. tli:s1<gfi Elle 21J' l7 e dmeR

qtfi tfi* 201W ,\.JEleR,la. Reference to Appendix C in hdttor*s '!ote. for Apphcabdtt' Index.

this Case shall apply to Appendix C of the 2004 Edition appl1cabili\) is lrom 1996 Addenda to or later ed1t1ons or addenda. For editions !!R<l-or addenda

                                                                         ~IJ13 Ed1uon prior to !h!;_2004 Edition, C lass I pipe flaw evaluation procedures may be used for other piping classes. As a matter of definition, the current term "structural factor" is equivalent to the term "safety factor," which is used in earlier editions and addenda.

Draft 15 (05/05/1 4) 15

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 17 of 33 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4~ MANDATORY APPENDIX I RELATIONS FOR Fm, Fb, AND FFOR THROUGH-WALL FLAWS 1-1 DEFINITIO S Ab = - 3.26543 + l .52784 (Rft) - 0.072698 (R/t f

                                                                    + 0.001601 l (RJt)'

For through-wall Oaws, the crack depth,_..!a.; will Bb = 11.36322 - 3.91412 (Rft) + 0. 186 19 (R/t ) 2 be replaced with half crack length,_--tcJ in the stress - 0.004099 (Rft)' intensity factor equations in C-7300 and C- 7400 of Cb = - 3.18609 + 3.84763 (Rlt) - 0.18304 (R/t f Section XI, Appendix C. Also, Q will be set equal to + 0.00403 (Rf 1)3 unity in C- 7400. Equations for Fm and Fb are accurate for Rlt between 5 1-2 CIRCUMFERENTIAL FLAWS and 20 and become increasingly conservative for Rft greater than 20. Alternative solutions for F~ and Fb may For a range of Rlt between 5 and 20, the following be used when Rft is greater than 20. equations for Fm and Fb may be used: 1-3 AXIAL FLAWS Fm= 1 + Am (8!7r)'-' + Bm(8/7rf' +Cm (t'J1i)B Fb - l + Ab (8/7r)u + Bb (817rf*' +Cb (fJKj'-' For internal pressure loading, the following equation for F may be used : where F = I + 0.0724491. + 0.648561.2 - 0.2327 A.3 e = half crack angle = c!R + 0.038154 A.4 - 0.0023487 A.' R = mean pipe radius t = ~wall thickness where and Am = -2.02917 + l .67763 (Rlt) -0.07987 (R/t f c = half crack length

           + 0.00176 (Rlt)'                                 A. = c/(Rt)vi Bm = 7.09987 - 4.42394 (Rlt) + 0.2 1036 (R/t ) 2    The equation for F is accurate for A. between 0 and 5.
           - 0.00463 (Rlt)'                             Alternative solutions for F may be used when A. is Cm - 7.79661 + 5.16676 (Rlt) - 0.24577 (R/t) 2      greater than 5.
           + 0.0054 l (Rft)'

Draft 15 (05/05/ 14) 16 _J

Attachments 1 - 3 RRS-03 Cooper Nuclear Station Page 18 of 33 Attachment 2 Technical Basis for Proposed Fourth Revision to ASME Code Case N-513

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 19 of 33 Proceedings of the ASME 2014 Pressure Vessel s & Piping Conference PVP201 4 July 20-24, 2014, Anaheim, California, USA PVP2014-28355 TECHNICAL BASIS FOR PROPOSED FOURTH REVISION TO ASME CODE CASE N-513 Roberto. McGill GuyOeBoo Structural Integrity Associates Exelon Generation Company San Jose, CA Warrenville , IL rmcgjl l@structint.com guv.deboo@exeloncorp.com Russell C. Cipolla Eric J. Houston lntertek AIM Structural Integrity Associates Sunnyvale, CA San Jose, CA russell .cioolla@intertek.com ehouston@structint.com ABSTRACT Resultant therm al expansion moment Elbow or bent pipe bend radius Code Case -5 13 provides evaluation rules and criteria Outside radius for temporary acceptance of flaws, including through-wall Flexibility characteristic flaw s, in moderate energy piping. The application of the Code Stress intensification factor as defined in the Code of Case is restricted to moderate energy, Class 2 and 3 systems, Record for the piping item so that safety issues regarding short-term, degraded system Ma'<imum operating pressure at flaw location operation are minimi zed The first version of the Code Case Evaluation thickness was published in 1997. Since then, there have been three Axial bending stress rev isions to augment and clarify the evaluation requirements Thermal expansion stress and acceptance criteria of the Code Case that have been Hoop stress published by AS:tvJE. The technical bases for the original Axial membrane stress version of the Code Case and the three revisions have been Circumferential angle previously published. There is currently work underway to incorporate INTRODUCTION additional changes to the Code Case and this paper prov ides Background the technica l basis for the changes proposed in a fourth revision. These changes include addressing the current Code Case N-513-3 [!] (currently approved rev ision) condition on the Code Case acceptance by the US Nuclear prov ides evaluation rules and criteria for the temporary Regulatory Comm ission (NRC), clarification of the Code Case acceptance of flaw s, including through-waLI flaws, in applicability lin1 its and expansion of Code Case scope to moderate energy piping. The prov is ions of this Code Case are additional piping corn ponents. New flaw evaluation focused on preventing gross fai lure of the affected pipe for a procedures are given for through-wall flaws in elbows, bent temporary period. However, it also requires the piping system pipe, reducers, expanders and branch tees. These procedures and adj acent equipment functionality be demonstrated for lost evaluate flaws in the piping components as if in straight pipe fluid inventory, spraying and flooding caused by the leakage. by adjusting hoop and axial stresses to account for the The Code Case provides rules for the evaluation of degraded geometry differences. These changes and their technical bases pipe and tube for a short operating period, with inspection and are described in this paper. monitoring requirements of the degraded condition as part of the overall integrity assessment. The application of the Code NOMENCLATURE Case is restricted to moderate energy Class 2 and Class 3 systems, so that the safety issues regarding short-term system B,, 82 Primary stress index defined in ASME Section III for operation are minimized. Moderate energy piping is defined the piping item as those piping systems where the maxim um operating Do Outside diameter pressure and temperature do not exceed 275 psig (1.9 MPa) I Moment of inertia based on t (degraded section) and 200°F (93°C), respectively. Mb Resultant primary bending moment Copyright 0 20 14 by ASME

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 20 of 33 Currently. the scope of the Code Case is limited to to accept flaws ... without performing a repa ir/replacement straight pipe with some prov ision for flaw evaluation into activity for a lin1ited time. not exceeding the time to the next fittings for a short distance from the weld attachment to scheduled refueling outage." Note that the word ..refueling'* straight pipe. There have been many instances where through- has been added for further clarification regarding the nature wa ll leaks have been observed in pipe components (i.e., and duration of the plant outage . elbows, reducers, expanders or branch tees) or bent pipe outside of the current Code Case scope . As a resull utilities This language is consistent with the introduction ofNRC are forced into taking alternate actions (e.g. , repair, Generic Letter 90-05 [3] which the original N-5 13 was based. replacement, or making requests for NRC relie!) that can be The second paragraph of90-05 states : "Temporary non-code significantly more burdensome without any measurable repairs are applicable until the next scheduled outage increase in plant safety. The fourth revision of the Code Case exceeding 30 days, but no later than the next scheduled provides evaluation rules and criteria for the temporary refueling outage.., acceptance of flaws for these instances. In addition. other enhancements are included with the fourth revision as Flaw Evaluation Criteria for Piping Components a nd Bent summarized below. !'.!.!!! Evaluation and acceptance criteria have been added to N-Summarv of Code Case N-513-4 C hanges S13-4 for flaws in elbows, bent pipe, reducers, expanders and The list below summarizes the proposed changes included branch tees. A sin1 plified approach has been adopted based on in Revision 4 of Code Case N-513. Note that a brief reason the ev aluations and results from the Second International for each change is included in parenthesis. Piping Integrity Research Group (IPIRG-2) program reported in Re ference [4]. The tlaw evaluation for the piping Temporary acceptance penod redefi ned (addresses component is conducted as if in straight pipe by scaling hoop NRC condition given in Regulatory Guide l .147 ['.!]) and axial stresses using ASME piping design code stress Flaw evaluation criteria included for elbows, bent indices and stress intensification factors to account for the pipe, reducers, expanders and branch tees (scope stress variations caused by the geometric diffe rences. In expansion) Reference [4], this approach was determ ined to be very Allow flaw evaluation of heat exchanger tubing in conservative by comparing the failure moments predicted specific instances (scope expansion) using this approach to the measured fa ilure moments from the Daily walkdown requirement for through-wall leaks elbow tests for through-wall circumferential flaws. Details of provi des additional flexibil ity for user the simplified approach are given in the following sections. imp lementation (scope expansion) Limit scope to only liquid systems (scope Flaw Evaluation in E lbows and Bent P ipe. Through-clarificat ion) wall flaws in elbows and bent pipe may be evaluated using the Treatment of Service Level load combinations (scope straight pipe procedures given in N-5 13-4 provided the clarification) stresses used in the eva luation are adjusted to account for geometry differences. The hoop stress for elbow and bent Treatment of flaws in au.stenitic pipe tlm<welds pipe evaluation shall be (Equation 9 of N-513-4) (scope clarification) Minimum wall thickness acceptance criteria to consider longitudinal stresses in addition to hoop stress (scope clarification) a

                                                                               =( pD 2t 0 ) [   2R,,.,.,+ R sin¢ ] + (~) RJvl, 0
                                                                                             '.!(R.,,., + R sin ¢)

0 h"' I (9) [n addition, severa l editorial changes to improve the Equation 9 ts only applicable for elbows and bent pipe clarity of the Code Case are included. where h 2'. 0. I . CODE CASE N-513-4 CHANGES AND TECHNICAL The axial membrane pressure stress for el bow and bent BASIS pipe evaluation sha ll be (Equation I 0 ofN-513-4): The follow ing subsections prov ide details regardmg each change and their technical basis. (10) Temporarv Acceptance Period N-513-3 specifies a temporary acceptance period that 8 1 shall be equal to 0.5 for e lbows and bent pipe . could extent out to 26 months. The NRC did not endorse this maximwn period length and in the latest revision of NRC The ax[al bending stress for elbow and bent pipe Regulatory Gu ide l . l 47, placed a condition on N-513-3 evaluation shall be (Equation l 1 of N-513-4) : stating, *' repair or replacement activity temporarily (RM,) deferred under the provisions of this Code Case shall be perfom1ed during the next scheduled outage ." N-513-4 " * = B, - !- 0 (l I ) addresses this condition by removing the maximum duration limit and stating, ". the follow ing requirements may be used Copyright <O 20 l 4 by ASME

Attachments 1 - 3 RRS-03 Cooper Nuclear Station Page 21 of 33 Note that if the piping design Code of Record does not provide a Bi, it may be taken from the ASlvffi Section ill, 2004 Edition or later Editions and Addenda. a ; ;( D.M,) 2/ ( 16) The thermal expansion stress for elbow and bent pipe evaluation shall be (Equation 12 of N -513-4) : Equations 13-1 6 are consistent with the piping design by rule approach in ASME Section ill . The reducer and expander zones shown in Figure 8 of -5 13-4 are conststent with those (j

                                *(RIM,)
                            = 1 --

0 ( 12) shown in Code Case -597-2 [8]. As with e lbows and bent pipe, N -513 -4 allows for alternate methods to be used to calculate stresses used in the flaw evaluation. The first term of Equation 9 accounts for the hoop membrane stress variation around the circumference of the Flaw Evaluation in Branch Tees. Branch reinforcement elbow (higher at the intrados, lower at the extrados assuming a requirements shall be met in accordance with the Construction uniform wa ll thickness). The second term conservati vely Code If the Construction Code did not require reinforce ment, accounts for the through-wall bending in the hoop direction a reinforcement region is defined as a regio n of radius D, of that results from elbow ovalization due to in-plane or out-of- the branch pipe from the center of the branch connection for plane bendi ng and was taken from Reference [5]. Equations eva luation purposes. Through-wall flaws in branch tees I 0-1 2 are consistent with the piping design by rule approach outside of the reinforcement region may be evaluated using in ASME Section ill, NC/ND-3600 [6]. the straight pipe procedures given in N -513-4 provided the stresses used in the evaluation are adj usted as described below It is recognized m Reference [4] that the simpli fied flaw to account for the geometry differences. Evaluation of flaws evaluation ap proach may be overly conservative in some in the region of branch reinforcement is outside the scope of instances. Thus. N-513-4 allows for alternate methods to be this Case. used to calculate stresses used in the flaw evaluation. For example, the French have developed a more comprehensive The hoop stress and axial membrane pressure stress for approach in analyzing tlaws in cracked elbows. They have branch tee eva luation shall be determined from -513-4 implemented finite element analysis based correlations to Equation 13 and 14, respectively The outside diameter for determine location speci fic stresses for flaw evaluation. each of these equations shall be for e ither the branch or run Reference [7] provides additional details. pipe depending on the flaw location. The axial bending stress and thermal expansion stress for branch tee evaluation shall be Flaw Evaluation in Hcduccrs and Expanders. Through- determined from N-51 3-4 Equation 15 and 16. respectively. wall flaws in reducers and expanders may be eva luated usmg the straight pipe procedures given in N-513-4 prov ided the As discussed previous ly, Equations 13-1 6 are consistent stresses used in the evaluation are adjusted to account for the wi th the piping design by rule approach in SwlE Section III. geometry differences. Figure 8 ofN-513-4 illustrates the The limitation regarding flaw evaluation within the branch reducer and expander zones d iscussed below. Eva luation of reinforcement region is consistent with guidance given in flaws in the small end trarisition zone is outside the scope of Code Case 1 -597-:. . As with elbow s and bent pipe, N-5 13-4

  -513-4. The hoop stress and axial membrane pressure stress         allows for alternate methods to be used to calculate stresses for reducer and expander evaluation shall be:                        used in the flaw ev aluation.

Flaw Evaluation of Heat Exchanger Tubing ( 13)

                                                                             -513-4 allows for flaw evaluation in heat e"changer tubing provided the flaw can be characteri zed and the leakage a,. ; 8 1 pD.)

(21 ( 14) monitored. The technical basis is that knowledge of tlaw geometry is needed to eva luate the structural mtegrity of the tubing containing the fl aw. An e"ample appli cation from plant operating experience wou ld be for leaks in nonferrous ote that D, is either the smal l end OD for the small end tubing of air coo ling coils. The heat exchangers being of or 1s the large end OD for the large end. For the large end plate-fin design could have tubing accessible for vol umetric transition zone and central conica l section. D, is the large end inspection. OD Also. ifthe piping design Code of Record does not provide a 8 1, it may be take n from the ASME Section ill, Flexibilitv in Dailv Walkdown R equirement 2004 Edition or later Editions and Addenda . N-513-4 changes the daily walkdown requirem ent for The axia l bending stress and therm al expansion stress for leaking flaws from **... leakage shall be observed by daily reducer and e"pander evaluation shall be : walkdowns . .. ** to *' ... leakage shall be monitored daily . This change allows for other techniques lo be employed in a, -

                          - B (D,M21
                                        ,)                  (15) meeting the intent of the requirement instead of only physical walkdowns. Such techniques could include remote visual Copyright 0 20 14 by ASME

Attachments 1 - 3 RR 5-03 Cooper Nuclear Station Page 22 of 33 equipment or leakage detection systems to determine if and branch tees. In addition, severa l other scope expansions leakage rates are changing. and clarifications are included as well as editorial improvements. With these changes, it is be lieved the Code Scope Li mited to Liquid Svstems Case wi ll be of even greater use to utilities in avoiding unscheduled plant shutdowns without negatively impacting Provisions of N-513-4 are now specifically lim ited to only plant safety. liquid systems. The methods ofN-513-4 should not be applied to the piping of air or other compressible fluid ACKNOWLEDGEMENTS system s. The deve lopment of Code Case N-513-4 was through the Treatment of Service Leve l Load Combinations efforts of the Working Group on Pipe Flaw Evaluation. The authors wish to recognize the efforts of all Working Group N-513-4 now specifically requires all Service Level load members especially the Working Group Chair, Dr. Doug combinations to be considered in flaw evaluations to Scanh. detennine the most limiting. N-5 13-3 was previous ly silent as to what Service Leve l loadi ng to consider. However, this REFERENCES requirement was im plied in previous Code Case revisions when it requi red the use of pipe fl aw evaluation procedures I. ASME Code Case N-5 13-3, "Evaluation Criteria for specified in the referenced, non-m andatory appendices of Temporary Acceptance of Flaws in Moderate Energy ASlvlE Section XI. Appendix C in the 2002 Addenda and Class 2 or 3 P iping Section XI, Division 1," Cases of later editions and addenda, and Appendix H for ferritic the ASME Boiler and Pressure Vessel Code, January materials in earlier editions, required all operating conditions 26, 2009. or Service Levels be evaluated. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division l ," Treatment of Flaws in Auste nitic Pioe Flux We lds Revision 16. October 20 10.

3. NRC Generic Letter 90-05. "Guidance for A reference to SlvlE Section XI, Appendix C, C-6320 Perfonning Temporary Non-Code Repair of ASME has been added to paragraph 3. l (b) to address the instance Code Class 1. 2, and 3 Piping, " (June 15, 1990).

where a flaw in an austenitic pipe flux weld requires

4. Ki linski. T., Mohan, R. , Rudland, D. , Fleming, M. ,

evaluation. Such a flaw would require the use of Elastic O lson, R., Scott, P., Brust, F., Ghadiali, N., Plastic Fracture Mechanics (EPFM) criteria instead of Lim it Witkowski, G., and Hopper, A. , "Fracture Behavior Load criteria. Equation I of -5 13-4 now includes a Z factor of Circum ferentially Surface-Cracked E lbows," to account for this specific application. NUREG/CR-6444, BMI-219'.!, December 1996.

5. Moore, SE. , and Roda baugh, E.C., "Backgro und for When the original version of N -513 was deve loped, Changes in the 198 1 Edition of the ASME Nuclear Appendix C of ASlvlE Section XI only included flaw Power Plant Components Code for Controlling evaluation criteria for austenitic piping. The change from Primary Loads in Piping Systems," Journal of 513-1 to -2 included the reference changes to the new Pressure Vessel Technology. Volum e 104, pp. 351 -

Appendix C that combined austenitic and ferritic flaw 361, ovember 1982. evaluation criteria . The need to reference EPFM critena for

6. ASME Boiler and Pressure Vessel Code. Section ill, an austenitic tlux weld flaw evaluation was over looked.

NC/ND-3600, 2004 Edition_

7. Marie, S.. Chapuliot, S.. Kayser, Y., Lacire, M.H.,

Mi nimum Wall T hickness Acceptance Criteria Drubay, B., Barthe let, B., Le-De lliou, P., Rougier, In establishing a minimum wall thickness acceptance V , Naudin, C., Gilles, P., and Triay, M, "French cr iteria for paragraphs ' .2(b) and 3.2 (c) ofN-5 13-4. the RSE-M and RCC-MR. Code Appendices for Flaw requirement to consider longitudinal stresses in addition to Ana lysis : Presentation of the Fracture Parameters hoop stress is added. While it is unlikely that a longitudinal Ca lculation - Part IV: Cracked Elbow," International stress based minimum wall thickness would be Limiting Journal of Pressure Vessels and Piping 84, pp. 659-compared to the hoop stress based minimum wall thic kness, 686, May 2007. the user should be aware of this possibility and confinn an 8. ASlvlE Code Case N-597-2. "Requirements for appropriate t~,. as an acceptance crite rion. Analytical Evaluation of Pipe Wall Thinning, Section XI, Division l ."Cases of the ASME Boiler and CONCLUSIONS Pressure Vessel Code, November 18, 2003 . This document serves as the technical basis for the fourth revision of Code Case N-513 This Code Case provides evaluation rules and cntena for tem porary acceptance of flaws, including through-wall flaws, in moderate energy Class 2 or Class 3 piping. Currently, the scope of the Code Case is limited to stra ight pipe with some prov ision for flaw evaluation into fittings for a short distance from the weld attachm ent to straight pipe . N- -13_4 provides flaw evaluation rules and criteria for elbows, bent pipe, reducers. expanders Copyright 0 2014 by ASME

Attachments 1 - 3 RRS-03 Cooper Nuclear Station Page 23 of 33 Attachment 3 Cooper Nuclear Station Technical Basis for Proposed Alternative to Use ASME Code Case N-513-4 and Scope Expansion to a Higher Pressure Limit

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 24 of 33 INTRODUCTION Background of Code Case N-513-4 Code Case N-513-4 [1] provides evaluation rules and criteria for the temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. Moderate energy piping is defined as those piping systems where the maximum operating pressure and temperature do not exceed 275 psig and 200°F, respectively. The provisions of the Code Case are focused on preventing gross failure of the affected pipe for a temporary period while permitting leakage within the plant's Technical Specification. The Code Case provides rules for the evaluation of degraded pipe and tube for a short operating period, with inspection and monitoring requirements of the degraded condition as part of the overall integrity assessment. The previous Code Case revision, N-513-3 [2], is conditionally approved by the NRC in Regulatory Guide 1.147 [3] . The single condition deals with a requirement to perform the repair or replacement activity, temporarily deferred, during the next scheduled outage. Since the introduction of this Code Case, many utilities have used it as a basis for continued operation of degraded piping in moderate energy systems and that has resulted in significantly fewer relief requests to the NRC . Consequently, the industry has benefited from substantial cost savings while maintaining safety. To date, there have been no known instances where the use of the Code Case has resulted in any safety issues in the industry. The scope of Code Case N-513-3 is limited to straight pipe with some provision for flaw evaluation into fittings for a short distance from the weld attachment to straight pipe. There have been many instances where through-wall leaks have been observed in pipe components (i.e., elbows, reducers, expanders or branch tees) or bent pipe outside of the current Code Case scope. As a result, utilities are forced into taking alternate actions (e.g., repair, replacement or making requests for NRC relief) that can be significantly more burdensome without any measurable increase in plant safety. Code Case N-513-4 provides evaluation rules and criteria for the temporary acceptance of flaws for these instances. In addition, other enhancements are included with the fourth revision as described herein. Background of Moderate Energy Definition The genesis of Code Case N-513 was NRC Generic Letter (GL) 90-05 [4]. Prior to Code Case N-513, this GL was the only available guidance for plants regarding operational leakage in moderate energy piping. The definition of moderate energy piping in Code Case N-513 is consistent with GL 90-05. The scope of GL 90-05 is limited to Class 3 piping, but does address moderate and high energy systems. While non-code repairs are allowed by GL 90-05 (with NRC review) for the temporary period of operation prior to Code compliant repair/replacement, an additional requirement for the repair having load-bearing capability is necessary for high energy pipe applications. Both GL 90-05 and the latest approved revision of Code Case N-513 are identified as methods available to evaluate the structural integrity of piping with discovered leakage in the NRC Inspection Manual [5]. The definition of moderate energy (the 200°F and 275 psig limits) was first introduced in a NRC letter from A. Giambusso to licensees in 1972 to address postulated piping breaks in fluid

Attachments 1 - 3 RRS-03 Cooper Nuclear Station Page 25 of 33 systems outside containment (discussed in the Branch Technical Position 3-3 of the current Standard Review Plan [6]) . The importance of the 200°F temperature limit being below the boiling point of water at atmospheric pressure is clear in the definition of moderate energy piping. However, the basis for the pressure limit of 275 psig is less evident. In conversations with the NRC, there appears to be a link between the 275 psig limit and the recommended working pressure limit for Class 150 pipe flanges and flanged fittings given in the ASME Bl6.5 standard [7]. This basis is not known to be documented. A technical approach and basis are presented herein to justify the scope expansion of Code Case N-513-4 to a higher pressure limit. The Class 3 RHRSW system at Cooper has a design pressure of 490 psig [8]. It is desired to expand the scope of N-513-4 such that it may be generically applied to this specific system as it has experienced part-wall degradation in the past. TECHNICAL BASIS FOR ALTERNATE USE OF ASME CODE CASE N-513-4 Summary of Code Case N-513-4 Changes The list below summarizes the major differences between the NRC-approved Code Case N-513-3 and Code Case N-513-4. Note that a brief reason for each change is included in parenthesis.

Temporary acceptance period redefined from no longer than 26 months to the next refueling outage (addresses NRC condition given in Regulatory Guide 1.14 7)
Flaw evaluation criteria included for elbows, bent pipe, reducers, expanders and branch tees (scope expansion)
Allow flaw evaluation of heat exchanger tubing in specific instances (scope expansion)
Daily walkdown requirement for through-wall leaks provides additional flexibility for user implementation (scope expansion)
Limit scope to only liquid systems (scope clarification)
Treatment of Service Level load combinations (scope clarification)
Treatment of flaws in austenitic pipe flux welds (scope clarification)
Minimum wall thickness acceptance criteria to consider longitudinal stresses in addition to hoop stress (scope clarification)

In addition, several editorial changes to improve the clarity of the Code Case are included. A detailed markup of Code Case N-513-3 showing all of the changes made for Revision 4 is provided in Attachment l. Technical Basis The technical basis for the changes listed above is given in an ASME Pressure Vessel and Piping (PVP) conference paper [9] and is provided in Attachment 2.

Attachments 1 - 3 RRS-03 Cooper Nuclear Station Page 26 of 33 Precedents There have been several submittals approved for N-513-4 use in specific applications and for generic use by Exelon. The table below lists several Safety Evaluation Reports as precedents for use of Code Case N-513-4. Table 1: Code Case N-513-4 Precedents SER Accession No. Plant Application Additional Requirements ML16230A237 Exelon Fleet Generic N-513-4 Critical leakage determination ML15070A428 ANO Leaking sweepolet 5 gpm leakage limit ML14316Al67 Fort Calhoun Leaking elbow None ML14335A55 l Peach Bottom Leaking elbow 5 gpm leakage limit TECHNICAL BASIS FOR CODE CASE N-514-4 SCOPE EXPANSION TO IDGHER PRESSURE LIMIT The technical approach includes several elements that support the following objectives: (i) show NRC precedent in approving relief requests for through-wall leakage in piping or components operating at high energy, (ii) show that structural integrity is not overly impacted when comparing flaw evaluations between piping systems operating at 275 and 490 psig, and (iii) show that possible jet thrust forces resulting from leaking flaws are not of concern. An outline of the technical basis detailing the elements used to support these objectives follows :

Review of Relevant NRC Approved Relief Requests
Structural Integrity Evaluation o Design minimum wall thickness comparison o Code Case N-513-4 allowable flaw size comparison o Code Case N-513-4 cover thickness requirement comparison
Jet Thrust Force Evaluation Note that the evaluation rules and acceptance criteria of Code Case N-513-4 are not limited by pressure and remain valid for the higher limit of 490 psig.

Review of Relevant Relief Requests A detailed search of the NRC ADAMS database was conducted to identify any relief requests for continued operation of degraded high energy piping or components. The search results are summarized in Table 2.

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 27 of 33 Table 2: Summary of Relevant Relief Requests ADAMS Plant I Operating Accession Description Status Condition Number McGuire, Ul I Application for continued ML072140851 Approved 2,500 psig operation of leaking valve San Onofre U2, Generic application for continued ML101440381 U3 I< 275 psig operation of high temperature Approved (275 ° F) through-wall leaking pipe Exelon Plants I Generic request to use Code Case ML15043A496 Approved* 375 psig N-513-3 at a higher pressure limit

In add1t1on to the N-513-3 requirements, the NRC imposed an allowable leak rate limit of 5 gpm .

Structural Integrity Evaluation Design minimum wall thickness comparison ASME B31. l [ 1O] defines the design minimum wall thickness required for pressure loading, tm, as (without a corrosion allowance): t = p0o (1) m 2(S+0.4p) where: p Design pressure Do Outside pipe diameter s Material allowable stress. Substituting maximum operating pressure for design pressure in Equation 1, the minimum required wall thickness increases 77% with an increase in pressure from 275 to 490 psig. While this percentage increase appears significant, the actual change in minimum required wall thickness is relatively low as the hoop stress at these pressures is small. Table 3 shows a comparison of minimum wall thicknesses for various pipe sizes. Table 3: Minimum Required Wall Thickness Comparison Nominal Pipe Size tmin for 275 psig* (in) tmin for 490 psig* (in) 6-inch 0.053 0.094 12-inch 0.102 0.181 18-inch 0.144 0.255

A material allowable stress of 17 .1 ksi is assumed.

Attachments 1 - 3 RRS-03 Cooper Nuclear Station Page 28 of 33 Code Case N-513-4 allowable flaw size comparison Code Case N-513-4 allows for nonplanar, through-wall flaws to be evaluated as two independent planar through-wall flaws, one orientated in the axial direction and one orientated in the circumferential direction (i.e. , a planar flaw characterization approach). The Code Case acceptance criteria require the flaw region be bounded by the area defined by the allowable axial and circumferential flaw sizes. Several example N-513-4 calculations were conducted illustrating the influence of higher pressure on allowable flaw size. Following the Code Case guidance, a linear elastic fracture mechanics (LEFM) evaluation was performed for various carbon steel pipe sizes to determine the maximum allowable flaw sizes at maximum operating pressures of 275 and 490 psig. Table 4 summarizes the results and several notes are provided giving more details regarding the analysis inputs. The influence of the higher pressure is clearly seen and a greater impact is observed in the axial direction as expected since pressure hoop stress is twice the axial membrane stress due to pressure. While the higher pressure does decrease the allowable flaws sizes, the effect does not impact the functionality or validity of the Code Case approach. Table 4: Allowable Axial and Circumferential Flaw Size Comparison Nominal Axial Direction (in) Circumferential Direction (in) Pipe Size 275 psig 490 psig %~ 275 psig 490 psig %~ 6-inch Mb = 40 3.1 1.9 39% 2.8 2.1 25% in-kips 12-inch Mb = 170 3.0 1.5 50% 3.3 2.1 36% in-kips 18-inch Mb = 260 2.2 0.9 59% 3.3 1.8 44% in-kips Notes: - Piping material assumed Al06 Grade B; standard schedule thickness.

                 - Applied bending moment for each pipe size results in a stress ratio of about 0.25 at 275 psig.
                 - Allowable flaw sizes based on Service Level B structural factors.
                 - Analysis based on a conservative lower shelf toughness value of 45 in-lb/in 2

Code Case N-513-4 cover thickness requirement comparison Code Case N-513-4 provides a branch reinforcement method to evaluate nonplanar, through-wall flaws. As part of the branch reinforcement approach, an opening is modeled such that its diameter fully bounds the leaking flaw. In practice, there could exist a remaining wall ligament within the modeled opening. Equation 9 of Code Case N-513 -4 provides assurance against pressure blowout (i.e., wall ligament failure) by requiring an average cover thickness, tc,avg, within the modeled opening:

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 29 of 33 tc,avg 2::: 0.353dadjJ% (2) where: dactj Modeled opening diameter p Maximum operating pressure s Material allowable stress. The average cover thickness requirement increases about 33% with an increase in maximum operating pressure from 275 to 490 psig. Figure 1 illustrates the average cover thickness required as a function of adjusted diameter for 275 and 490 psig. Note that a material allowable stress of 17.1 ksi is assumed. Typically, modeled openings have a diameter < 1 inch and the change in the required cover thickness is small(< 15 mils). 0.10 0.09 0.08

.§. 0.07 - 275 psig Q)

     ~     0.06                                                       -    490 psig u
     ..c f-(jj  0.05 0

u

     " 0.04
      ~

i fil' a:::

0.03 0.02 0.01 0.00 0 0.25 0.5 0.75 1 Adjusted Diameter (in) Figure l : Required Cover Thickness vs. Adjusted Diameter for 275 and 490 psig

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 30 of 33 Jet Thrust Force Evaluation Part 3.6.2 of the NRC Standard Review Plan [11] provides a simplified dynamic analysis model to quantify the jet thrust force, T, of water from a pipe break. The following equation is given: T=KpA (3) where: K Thrust coefficient (2.0 for subcooled, nonflashing water) p System pressure prior to pipe break A Pipe break area. Figure 2 shows a comparison of jet thrust force for pressures of 275 and 490 psig over a range of through-wall opening diameters. For small through-wall opening diameters(< 0.5 in), the difference in the jet thrust force is small. It is unlikely larger openings would be tolerated by the plant as this would result in excessive leakage and possibly impact system operability or room flooding abatement. 800 700 600

                                                                -  275 psig 500                                                  -  490 psig
   .0 QI u     400 0

u.. QI 300 200 100 0 0 0 .25 0 .5 0 .75 1 Through-Wall Hole Diameter (in) Figure 2: Jet Force vs. Through-wall Hole Diameter for 275 and 490 psig

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 31 of 33 ADDITIONAL REQUIREMENTS Implementation of Code Case N-513-4 requires additional actions be satisfied by the plant including observing leakage daily to confirm analysis conditions used in the evaluation remain valid, frequent periodic inspections to track flaw growth and augmented examinations to assess degradation of the affected system. Furthermore, the following additional requirements are included as part of the Relief Request in order to bolster defense-in-depth and avoid adverse consequences:

1. For a leaking flaw, the allowable leakage rate will be determined by dividing the critical leakage rate by a safety factor of four (4). The critical leakage rate is determined as the maximum leakage rate that can be tolerated. The critical leakage rate may be based on the allowable loss of inventory or the maximum leakage that can be tolerated relative to room flooding, among others. The safety factor of four to the critical leakage rate is believed adequate to ensure system operability and allow for early identification of conditions that may lead to adverse consequences. This requirement is consistent with a previous NRC Relief Request Safety Evaluation for the Exelon fleet [12] to generically use Code Case N-513-4.

2. The number of augmented examination locations shall be increased from 5 (current number in Code Case N-513-4) to 10. This is consistent with the NRC Generic Letter 90-05 [4] requirement for high energy Class 3 applications.

3. For leakage rates greater than 5 gpm, the leakage shall be stopped throughout the temporary acceptance period by the use of engineered mechanical clamping designed by Cooper. The engineered mechanical clamping design shall be sufficient to withstand the maximum operating pressure and removable such that the frequent periodic inspections defined in paragraph 2(e) ofN-513-4 may be performed.

CONCLUSIONS This attachment provides a technical basis for proposed alternative use of ASME Section XI Code Case N-513-4 and its scope expansion to a higher pressure limit from 275 to 490 psig for the Cooper Nuclear Station, specific to the RHRSW system. The discussion provided herein demonstrates that the use of Code Case N-513-4 with the higher pressure limit will reduce plant burden without any adverse effect on safety. The technical basis is comprised of four primary elements. First, Attachment 2 provides the technical basis for the new content and evaluation methods in Code Case N-513-4 including those for flaw analysis in piping components such as elbows, reducers and branch tees. Second, relevant NRC relief requests were reviewed. They demonstrate that there is precedent for the temporary acceptance of leaking flaws in Class 2/3 piping and components at pressures higher than 275 psig. Third, a structural integrity evaluation was performed to determine the impact of the increased pressure on design minimum wall thickness, N-513-4 allowable flaw sizes and the N-513-4

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 32 of 33 cover thickness requirement. While the influence of the higher pressure was observed in the evaluation, structural integrity can still be demonstrated. The functionality and validity of the Code Case methods at higher pressure were confirmed. Finally, jet thrust forces were estimated for a leaking pipe at 275 and 490 psig. A significant change in jet thrust forces was only seen with large opening areas that would result in high leak rates, i.e., rates that would challenge system functionality or local spray and/or compartment flooding requirements. As part of the generic Relief Request, several additional requirements are introduced in order to bolster defense-in-depth. They include expanding the number of augmented inspection locations from 5 to 10, introducing a limit on discovered leakage and requiring the leakage be stopped during the temporary acceptance period. REFERENCES

1. ASME Code Case N-513 -4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division I," Cases of the ASME Boiler and Pressure Vessel Code, May 7, 2014.

2. ASME Code Case N-513-3 , "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division I," Cases of the ASME Boiler and Pressure Vessel Code, January 26, 2009 .

3. Regulatory Guide 1. 147, " Inservice Inspection Code Case Acceptability, ASME Section XI, Division I," Revision 17, August 2014.

4. NRC Generic Letter 90-05 , "Guidance for Performing Temporary Non-Code Repair of ASME Code Class I, 2, and 3 Piping," (June 15, 1990).

5. NRC Inspection Manual, Chapter 0326, "Operability Determinations & Functionality Assessments for Conditions Adverse to Quality or Safety," Issue Date: 12/03 / 15 .

6. NRC Standard Review Plan, NUREG-0800, Branch Technical Position 3-3 , "Protection Against Postulated Piping Fai lures in Fluid Systems Outside Containment," Revision 3, March 2007 .

7. American National Standard, " Pipe Flanges and Flanged Fittings," ASME B 16 .5-2003 .

8. Nebraska Public Power District Design Calculation No . NEDC 02-051, Revision 1,

          " Design Temperature Determination for SW, CW, WW, & CS Piping; Design Pressure Determination for SW-I piping," September 17, 2002, SI Fi le No. 1700405.20 I.

9. McGill, R.O., Cipolla, R.C., DeBoo, G . and Houston, E.J., "Technical Basis for Proposed Fourth Revision to ASME Code Case N-513 ," Proceedings of the ASME 2014 Pressure Vessels & Piping Conference, PVP2014-28355 .

10. USA Standard, USAS 831.1.0, " Power Piping," 1967 Edition.

11. NRC Standard Review Plan, NUREG-0800, Part 3.6.2, " Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping,"

Revision 2, March 2007.

Attachments 1 - 3 RR5-03 Cooper Nuclear Station Page 33 of 33 12 . NRC Relief Request Approval and Safety Evaluation Report for Exelon Generation Company Fleet Request for Proposed Alternative to Use ASME Code Case N-513-4, ADAMS Accession No. ML16230A237.}}

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