RS-15-155, Response to Preliminary RAI Regarding Braidwood and Byron Stations Relief Request for Alternative Requirements for the Repair of Reactor Vessel Head Penetrations
| ML15149A424 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 05/29/2015 |
| From: | Gullott D Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15152A254 | List: |
| References | |
| RS-15-155 | |
| Download: ML15149A424 (45) | |
Text
Exelon Generation 4300 W1nf1eld Road Warrenville. IL 60555 www.exeloncorp.com 10 CFR 50. 55a RS-15-155 May 29, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455
Subject:
Response to Preliminary RAI Regarding Braidwood and Byron Stations Relief Request for Alternative Requirements for the Repair of Reactor Vessel Head Penetrations
References:
(1) Letter from David M. Gullatt, (EGC) to U.S. NRC, "Revision to the Third 10-Year lnservice Inspection Interval Requests for Relief for Alternative Requirements for the Repair of Reactor Vessel Head Penetrations," Dated September 8, 2014, ADAMS Accession No. ML14251A536 (2) Email from Joel Wiebe, (U . S. NRC) to Jessica Krejcie, (EGC),
"Preliminary RAI Regarding Braidwood and Byron Stations Relief Request for Alternative Requirements for the Repair of Reactor Vessel Head Penetrations," Dated March 23, 2015 In accordance with 10 CFR 50 .55a, "Codes and standards," paragraph (a)(3)(i) , in a letter dated September 8, 2014, (Reference 1), Exelon Generation Company, LLC (EGC) submitted a revision to the Third 10-Year lnservice Inspection Interval Requests for Relief for Alternative Requirements for the Repair of Reactor Vessel Head Penetrations at Braidwood Station , Units 1 and 2 and Byron Station, Units 1 and 2. The submitted letter requested inspection frequency relief for the Reactor Vessel Head Penetrations repair weld surface examinations (i.e., dye penetrant (PT)) .
In Reference 2, the NRC requested additional information related to its review of Reference 1. The additional information was discussed in a teleconference with the NRC on April 24, 2015. Additionally, during this teleconference it was identified that EGC would
May 29, 2015 U. S. Nuclear Regulatory Commission Page 2 submit a revision to the relief request that would reduce the frequency of PT examinations in lieu of elimination of the PT examinations. includes the requested additional information. Attachment 2 provides the revised relief request. Attachment 3 provides the revised technical basis supporting the Attachment 2 relief request. Attachment 4 provides the Non-Proprietary version of Attachment 3. contains proprietary information as defined by 10 CFR 2.390, "Public inspection, exemption, requests for withholding." Westinghouse Electric Company, LLC, (Westinghouse),
as the owner of the proprietary information has executed the enclosed affidavit, which identifies that the enclosed proprietary information has been handled and classified as proprietary, is customari ly held in confidence, and has been withheld from public disclosure. The proprietary information was provided to EGC by Westinghouse as referenced by the affidavit. The proprietary information has been faithfully reproduced in the attached information such that the affidavit remains applicable. Westinghouse hereby requests that the attached proprietary information be withheld, in its entirety, from public disclosure in accordance with the provisions of 10 CFR 2.390 and 10 CFR 9.17. The affidavit supporting the proprietary nature of the information is contained in Attachment 5.
EGC requests approval of this proposed relief request by September 4, 2015, prior to the beginning of the Byron Station refueling outage in Fall 2015 (81 R20).
There are no regulatory commitments contained in this submittal.
If you have any questions regarding this matter, please contact Jessica Krejcie at (630) 657-2816.
Respectfully, w~~..-----
David M. Gullatt Manager - Licensing Exelon Generation Company, LLC Attachment 1: Response to Request for Additional Information Related to Alternative Requirements for the Repair of Reactor Vessel Head Penetrations In Accordance with 10 CFR 50.55a(z)( 1) (Non-Proprietary)
Attachment 2: 10 CFR 50.55a Relief Requests 13R-09 and 13R-20, Revision 2 Alternative Requirements for the Repair of Reactor Vessel Head Penetrations In Accordance with 10 CFR 50.55a(z)( 1) (Non-Proprietary)
Attachment 3: Westinghouse Report LTR-PSDR-T AM-14-005, Revision 3, May 2015 (PROPRIETARY)
STN 50-454 Attachment 4: Westinghouse Report LTR-PSDR-TAM-14-005, Revision 3, May 2015 (Non-Proprietary)
Attachment 5: Westinghouse Electric Company, LLC Affidavit for Report LTR-PSDR-TAM 005 Revision 3, May 2015
Attachment 1 Response to Request for Additional Information Related to Alternative Requirements for the Repair of Reactor Vessel Head Penetrations In Accordance with 10 CFR 50.55a(z)(1)
ATTACHMENT 1 Response to Request for Additional Information By letter dated September 8, 2014, (Agencywide Documents Access and Management System (ADAMS) Accession Number ML14251A536), Exelon Generation Company, LLC (EGG) submitted a revision to the Third 10-Year lnservice Inspection Interval Requests for Relief for Alternative Requirements for the Repair of Reactor Vessel Head Penetrations at Braidwood Station, Units 1 and 2 and Byron Station, Units 1 and 2. The submitted letter requested inspection frequency relief for the Reactor Vessel Head Penetrations repair weld surface examinations (i.e., dye penetrant (PT. During its review, the NRG requested additional information related to its review of Reference 1. The additional information was discussed in a teleconference with the NRG on April 24, 2015. Additionally, during this teleconference it was identified that EGG would submit a revision to the relief request that would reduce the frequency of PT examinations in lieu of elimination of the PT examinations. A response to the NRG question as provided in Reference 1 below: NRC RAI: The NRG staff needs to make a finding that the proposed alternative provides an acceptable level of quality and safety in order to authorize the alternative. The NRG authorization of the embedded flaw repair technique by letter dated March 29, 2012, (ADAMS Accession No. ML120790647) was a temporary repair technique with known issues that require reinspection each refueling outage to verify a continuing acceptable level of quality and safety. The NRG staff has reviewed the technical justification provided in your submittal. However, additional operating experience appears necessary to make a finding that the level of quality and safety is acceptable. Provide operating experience with repaired penetrations that have been in service for at least 4 cycles (4 sets of examinations) that demonstrate that the current specified PT examinations are not required to maintain an acceptable level of quality and safety. EGC Response to NRC RAI: Between Braidwood Station and Byron Station, seven penetrations have been repaired using the Embedded Flaw Repair (EFR) Technique. Of those seven penetrations, one repair has been in service at least four cycles (i.e., four sets of examinations). Byron Station Unit 2, penetration 68 was repaired with the EFR during the refueling outage in Spring 2007 (B2R13) . The detailed history of this penetration's PT results are included in Table 2 of Attachment 3 to this submittal. After installation, five sets of subsequent examinations verified successful PT examination results . Page 1 of 1
Attachment 2 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20, Revision 2 Alternative Requirements for the Repair of Reactor Vessel Head Penetrations In Accordance with 10 CFR 50.55a(z)(1)
/SI Program Plan Units 1 and 2, Third Interval 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 1 of 10) Request for Relief Alternative Requirements for the Repair of Reactor Vessel Head Penetrations In Accordance with 10 CFR 50.55a(z)(1) 1.0 ASME CODE COMPONENT(S) AFFECTED Component Numbers Braidwood and Byron Station, Units 1 and 2, Reactor Vessels 1RC01 R (Unit 1) and 2RC01 R (Unit 2)
Description:
Alternative Requirements for the Repair of Reactor Vessel Head Penetrations (VHPs) and J-groove Welds Code Class: Class 1 Examination Category: ASME Code Case N-729-1 Code Item: B4.20 Identification: Byron Units 1 and 2, VHP Numbers 1 through 78, (P-1 through P-78) Previous repairs (13R-14): Unit 2, P-68 (13R-19) : Unit 1, P-31, P-43, P-64, and P-76 1 (13R-20): Unit 2, P-6 1 Braidwood Units 1 and 2, VHP Numbers 1 through 78, (P-1 through P-78) Previous repairs (13R-09): Unit 1, P-69 1 Drawing Numbers: Various 2.0 APPLICABLE CODE EDITION AND ADDENDA lnservice Inspection and Repair/Replacement Programs: American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) , Section XI, 2001 Edition, through 2003 Addenda. Examinations of the VHPs are performed in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-1, with conditions. Code of Construction [Reactor Pressure Vessel (RPV)] : ASME Section Ill , 1971 Edition through Summer 1973 Addenda. 1 This relief request includes lnservice Inspection (ISi) examination requirements for repairs previously completed in accordance with 13R-14, 13R-19, 13R-09 and 13R-20.
ISi Program Plan Units 1 and 2, Third Interval 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 2 of 10) 3.0 APPLICABLE CODE REQUIREMENT IWA-4000 of ASME Section XI contains requirements for the removal of defects from and welded repairs performed on ASME components. The specific Code requ irements for which use of the proposed alternative is being requested are as follows: ASME Section XI, IWA-4421 states: Defects shall be removed or mitigated in accordance with the following requirements: (a) Defect removal by mechanical processing shall be in accordance with IWA-4462. (b) Defect removal by thermal methods shall be in accordance with IWA-4461. (c) Defect removal or mitigation by welding or brazing shall be in accordance with IWA-4411 . (d) Defect removal or mitigation by modification shall be in accordance with IWA-4340. Note that use of the "Mitigation of Defects by Modification" provisions of IWA-4340 is prohibited per 10 CFR 50.55a(b)(2)(xxv) . For the removal or mitigation of defects by welding, ASME Section XI, IWA-4411 states, in part, the following. Welding, brazing, and installation shall be performed in accordance with the Owner's Requirements and ... in accordance with the Construction Code of the item ... The applicable requirements of the Construction Code required by IWA-4411 for the removal or mitigation of defects by welding from which relief is requested are as follows . Base Material Defect Repairs: For defects in base material, ASME Section 111, NB-4131 requires that the defects are eliminated, repaired, and examined in accordance with the requirements of NB-2500. These requ irements include the removal of defects via grinding or machining per NB-2538. Defect removal must be verified by a Magnetic Particle (MT) or Liquid Penetrant (PT) examination in accordance with NB-2545 or NB-2546, and if necessary to satisfy the design thickness requirement of NB-3000, repair welding in accordance with NB-2539. ASME Section Ill , NB-2539.1 addresses removal of defects and requires defects to be removed or reduced to an acceptable size by suitable mechanical or thermal methods. ASME Section Ill, NB-2539.4 provides the rules for examination of the base material repair welds and specifies they shall be examined by the MT or PT methods in accordance with NB-2545 or NB-2546. Additionally, if the depth of the repair cavity exceeds the lesser of 3/8-inch or 10% of the section thickness, the repair weld shall be examined by the radiographic method in accordance with NB-5110 using the acceptance standards of NB-5320.
IS/ Program Plan Units 1 and 2, Third Interval 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 3of10) Weld Metal Defect Repairs (This applies to the CROM penetration J-Groove weld.) ASME Section Ill, NB-4450 addresses repair of weld metal defects. ASME Section Ill, NB-4451 states; that unacceptable defects in weld metal shall be eliminated and , when necessary, repaired in accordance with NB-4452 and NB-4453. ASME Section Ill, NB-4452 addresses elimination of weld metal surface defects without subsequent welding and specifies defects are to be removed by grinding or machining. ASME Section Ill, NB-4453.1 addresses removal of defects in welds by mechanical means or thermal gouging processes and requires the defect removal to be verified with MT or PT examinations in accordance with NB-5340 or NB-5350 and weld repairing the excavated cavity. In the case of partial penetration welds where the entire thickness of the weld is removed, only a visual examination is required to determine suitability for re-welding. As an alternative to the requirements above, repairs will be conducted in accordance with the appropriate edition/addenda of ASME Section Ill and the alternative requirements, based on WCAP-15987-P, Revision 2-P-A, "Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations," December 2003, (Refer to Reference 1, hereafter known as WCAP-15987-P). 4.0 REASON FOR THE REQUEST Exelon Generation Company, LLC (EGC) will conduct examinations of the reactor Vessel Head Penetrations (VHPs) in accordance with Code Case N-729-1, as amended by 10 CFR 50.55a. Flaw indications that require repair may be found on the VHP tube material and/or the J-groove attachment weld(s) on the underside of the reactor vessel head. Relief is requested from the requirements of ASME Section XI , IWA-4411 to perform permanent repair of future defects that may be identified on the VHP's and/or J-groove attachment weld(s) in accordance with the rules of the ASME Section Ill Construction Code as described in thi s relief request. Specifically, relief is requested from:
- The requirements of ASME Section Ill, NB-4131, NB-2538, and NB-2539 to eliminate and repair defects in materials.
- The requirements of ASME Section Ill, NB-4450 to repair defects in weld metal.
/SI Program Plan Units 1 and 2, Third Interval 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 4 of 10) 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE 5.1 Proposed Alternative EGC proposes to use the less intrusive embedded flaw process (Reference 1) for the repair of VHP(s) as approved by the NRC (Reference 2) as an alternative to the defect removal requirements of ASME Section XI and Section Ill. 5.1 .1 The criteria for flaw evaluation established in 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-1, will be used in lieu of the "Flaw Evaluation Guidelines" specified by the NRC Safety Evaluation for WCAP-15987-P (Refer to Reference 5). 5.1 .2 Consistent with WCAP-15987-P, Revision 2-P-A methodology, the following repair requirements will be performed.
- 1. Inside Diameter (ID) VHP Repair Methodology
- a. An unacceptable axial flaw will be first excavated (or partially excavated) to a maximum depth of 0.125 inches. Although this depth differs from that specified in WCAP-15987-P, the cavity depth is not a critical parameter in the implementation of a repair on the ID surface of the VHP. The goal is to isolate the susceptible material from the primary water (PW) environment. The purpose of the excavation is to accommodate the application of at least two (2) weld layers of Alloy 52 or 52M, which is resistant to Primary Water Stress Corrosion Cracking (PWSCC), to meet that requirement. The depth specified in WCAP-15987-P is a nominal dimension and the depth needed to accommodate three weld layers while still maintaining the tube ID dimension. Since two (2) weld layers will be applied, less excavation is required and only 0.125 inches of excavation is necessary. The shallower excavated cavity for 2 weld layers would mean a slightly thinner weld, which would produce less residual stress.
The excavation will be performed using an Electrical Discharge Machining (EDM) process to minimize VHP tube distortion. After the excavation is complete, either an ultrasonic test (UT) or surface examination will be performed to ensure that the entire flaw length is captured. Then a minimum of 2 layers of Alloy 52 or 52M weld material will be applied to fill the excavation . The expected chemistry of the weld surface is that of typical Alloy 52 or 52M weldment with no significant dilution. The finished weld will be conditioned to restore the inside diameter and then examined by UT and surface examination to ensure acceptability.
- b. If required , unacceptable ID circumferential flaw will be either repaired in accordance with existing code requirements; or will be partially excavated to reduce the flaw to an acceptable size, examined by UT or surface examination ,
inlaid with Alloy 52 or 52M, and examined by UT and surface examination as described above.
- 2. Outside Diameter (OD) VHP and J-groove Weld Repair Methodology
/SI Program Plan Units 1 and 2, Third Interval 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 5 of 10)
- a. An unacceptable axial or circumferential flaw in a tube below a J-groove attachment weld will be sealed off with an Alloy 52 or 52M weldment. Excavation or partial excavation of such flaws is not necessary. The embedded flaw repair technique may be applied to OD axial or circumferential cracks below the J-groove weld because they are located away from the pressure boundary, and the proposed repair of sealing the crack with Alloy 690 weld material would isolate the crack from the environment as stated in Section 3.6.1 of the NRC Safety Evaluation for WCAP-15987 -P.
- b. Unacceptable radial flaws in the J-groove attachment weld will be sealed off with a 360 degree seal weld of Alloy 52 or 52M covering the entire weld. Excavation or partial excavation of such flaws is not necessary.
- c. If EGC determines an excavation is desired (e.g., boat sample) , then
- The excavation will be filled with Alloy 52 or 52M material.
- It is expected that a portion of the indication may remain after the boat sample excavation; however, a surface examination will be performed on the excavation to assess the pre-repair condition.
- Depend ing on the extent and/or location of the excavation, the repair procedure requires the Alloy 52 or 52M weld material to extend at least one half inch outboard of the Alloy 82/182 to stainless steel clad interface.
- d. Unacceptable axial flaws in the VHP tube extending into the J-groove weld will be sealed with Alloy 52 or 52M as discussed in Item 5.1.2.2.a above. In addition, the entire J-groove weld will be sealed with Alloy 52 or 52M to embed the axial flaw. The seal weld will extend onto and encompass the portion of the flaw on the outside diameter of the VHP tube.
- e. For seal welds performed on the J-groove weld, the interface boundary between the J-groove weld and stainless steel cladding will be located to positively identify the weld clad interface to ensure that all of the Alloy 82/182 material of the J-groove weld is seal welded during the repair.
- f. The seal weld that will be used to repair an OD flaw in the nozzles and the J-groove weld will conform to the following .
- Prior to the application of the Alloy 52 or 52M seal weld repair on the RPV clad surface, at least three beads (one layer) of ER309L stainless steel buffer will be installed 360° around the interface of the clad and the J-groove weld metal.
- The J-groove weld will be completely covered by at least three (3) layers of Alloy 52 or 52M deposited 360° around the nozzle and over the ER309L stainless steel buffer. Additionally, the seal weld will extend onto and encompass the outside diameter of the penetration tube Alloy-600 material by at least one half inch.
ISi Program Plan Units 1 and 2, Third Interval 10 CFR 50.SSa RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 6 of 10)
- The VHP tube will have at least two (2) layers of Alloy 52 or 52M deposited over the flaw on the VHP tube, extending out at least one half inch beyond th e flaw, or to the maximum extent allowed by the nozzle geometry (e.g.,
limited length of the VHP tube).
- g. Nondestructive examinations of the finished seal weld repair (i. e., Repair NOE) and during subsequent outages (i.e., ISi NOE) are summarized in the table below.
Repair Location in Flaw Orientation Repair Repair NOE ISi NOE Original in Original Method Note (2) Note (2) Component Component Axial or UT and VHP Nozzle/Tube ID Seal weld UT or Surface Circumferential Surface VHP Nozzle/Tube Axial or OD above J-groove Note (1) Note (1) Note (1) Circumferential weld VHP Nozzle/Tube Axi al or OD below J-groove Seal weld UT or Surface UT or Surface Circumferential weld UT and UT and Surface, J-groove weld Axial Seal weld Surface, Notes (3) and (4) Note (3) UT and UT and Surface, J-groove weld Circumferential Seal weld Surface, Notes (3) and (4) Note (3) Notes: (1) Repair method to be approved separately by NRC. (2) Preservice and lnservi ce Inspection to be consistent with 10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of Code Case N-729-1 with conditions; or NRG-approved alternatives to these specified conditions. (3) UT personnel and procedures qualified in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of Code Case N-729- 1 with conditions. Examine the accessible portion of the J-groove repai red region. The UT plus surface exam ination coverage equals to 100%. (4) Surface examination of the embedded flaw repair (EFR) shall be performed to ensure the repair satisfies ASME Section Ill, NB-5350 acceptance standards. The frequency of examination shall be as fo llows:
- a. Perform surface examination during the refueling outage after installation or repair of the EFR.
- b. When the examination results in 4.a above verify acceptable results then re-inspection of the EFR will be continued at a frequency of every other outage. If these examinations identify unacceptable
ISi Program Plan Units 1 and 2, Third Interval 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 7 of 10) results that require flaw removal, flaw reduction to acceptable dimensions or welded repair the requirements of 4.a above shall be applied during the next refueling outage. 5.1.3 J-Groove Weld ISi NOE Requirements Note 4 permits a reinspection frequency of every other cycle when the surface examination results of the EFR are verified to be acceptable. Westinghouse Report LTR-PSDR-TAM-14-005, Revision 3 (Reference 13; provided in Attachment 3) provides the technical bases for reducing surface exam ination requirements for J-groove weld repairs. This technical justification includes a detailed review of PT examination history, review of potential causes of PT indications in EFRs, and the use of crack resistant alloys in the EFR. The EFR is a robust design that is resistant to PWSCC. EFR installation, examination, and operational history indicate that the EFR performs acceptably. Examination and removed sample history indicate that the flaws identified shortly after installation of EFR weld material were due to embedded weld discontinuities and not due to service induced degradation. With inspection of the EFR every other cycle of operation, the nozzles are adequately monitored for degradation by ultrasonic examination methods similar to the nozzles without EFR repairs. EGC projects that the reduction of the PT examination of nozzles would result in a dose savings of approximately 0.4 to 0.7 REM per nozzle examination. The historical radiation dose associated with these examinations is presented in Reference 13, Table 2. The proposed changes to the inservice examination requirements assure that the EFR repaired nozzles are adequately monitored through a combination of volumetric and surface examinations throughout the life of the installation at a frequency approved by the NRC, thus ensuring the EFR repaired nozzles will continue to perform their required function. 5.1.4 Reporting Requirements and Conditions on Use EGC will notify NRC of the Division of Component Integrity or its successor of changes in indication(s) or findings of new indication(s) in the penetration nozzle or J-groove weld beneath a seal weld repair, or new linear indications in the seal weld repair, prior to commencing repair activities in subsequent outages.
ISi Program Plan Units 1 and 2, Third Interval 10 CFR 50.SSa RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 8 of 10) 5.2 Technical Basis for Proposed Alternative As discussed in WCAP-15987-P, the embedded flaw repair technique is considered a permanent repair. As long as a PWSCC flaw remains isolated from the Primary Water (PW) environment, it cannot propagate. Since an Alloy 52 or 52M weldment is considered highly resistant to PWSCC, a new PWSCC flaw should not in itiate and grow through the Alloy 52 or 52M seal weld to reconnect the PW environment with the embedded flaw. Structural integrity of the affected J-groove weld and/or nozzle will be maintained by the remaining unflawed portion of the weld and/or the VHP. Alloy 690 and Alloy 52/52M are highly resistant to stress corrosion cracking, as demonstrated by multiple laboratory tests, as well as over twenty years of service experience in replacement steam generators. The residual stresses produced by the embedded flaw technique have been measured and found to be relatively low because of the small seal weld thickness. This implies that no new flaws will initiate and grow in the area adjacent to the repair weld . There are no other known mechanisms for significant flaw propagation in the reactor vessel closure head and penetration tube region since cyclic loading is negligible, as described in WCAP-15987-P. Therefore, fatigue driven crack growth should not be a mechanism for further crack growth after the embedded flaw repair process is implemented . The thermal expansion properties of Alloy 52 or 52M weld metal are not specified in the ASME Code. In this case the properties of the equivalent base metal (Alloy 690) should be used. For Alloy 690, the thermal expansion coefficient at 600 degrees Fis 8.2E-6 in/in/degree Fas found in Section II part D. The Alloy 600 base metal has a coefficient of thermal expansion of 7.8E-6 in/in/deg ree F, a difference of about 5 percent. The effect of this small difference in thermal expansion is that the weld metal will contract more than the base metal when it cools, thus producing a compressive stress on the Alloy 600 tube or J-groove weld. This beneficial effect has already been accounted for in the residual stress measurements reported in the technical basis for the embedded flaw repair, as noted in the WCAP-15987 -P. WCAP-16401-P, Revision O (Reference 3) provides the plant-specific analysis performed for Byron and Braidwood Stations using the same methodology as WCAP-15987-P. This analysis provides the means to evaluate a broad range of postulated repair scenarios to the reactor vessel head penetrations and J-groove welds relative to ASME Code requirements for allowable size and service life. The above proposed embedded flaw repair process is supported by applicable generic and plant specific technical bases , and is therefore considered to be an alternative to Code requirements that provides an acceptable level of quality and safety, as requ ired by 10 CFR 50.55a(z)(1).
/SI Program Plan Units 1 and 2, Third Interval 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 9 of 10) 6.0 DURATION OF THE PROPOSED ALTERNATIVE The duration of the proposed alternative is for the remainder of the Byron Station Units 1 and 2, Third lnservice Inspection Interval currently scheduled to end in July 15, 2016. The duration of the proposed alternative is for the remainder of the Braidwood Units 1 and 2, Third lnservice Inspection Interval currently scheduled to end in July 28, 2018, and October 16, 2018, respectively. 7.0 PRECEDENTS In Reference 2, the NRC generically approved the embedded flaw repair process described in Reference 1. Requests to use the embedded flaw technique to repair cracks on the OD of VHPs as well as to repair flaws in the J-groove attachment welds of VHPs have been previously approved by the NRC on a plant specific basis. The NRC approved a similar repair for Byron Station Unit 2 in Reference 9. On March 28, 2011, Byron Station Unit 1 received verbal authorization for use of the seal weld repairs methodology on P-64 and P-76, and again on April 10, 2011, for P-31 and P-43 (References 10 and 11 ). This alternative incorporates lessons that are learned regarding the significant radiation dose incurred for seal weld repair surface examinations at Beaver Valley, Unit 2, during the fall 2009 outage repair activities, which were discussed in the previously approved 10 CFR 50.55a request for Beaver Valley, Unit 2 (Reference 8) . As such, this alternative requests provisions that permit original construction code acceptance criteria for the post weld overlay surface examination, and a barrier layer of ER309L filler material, prior to the application of three Alloy 52M repair weld layers on the clad surface, at the periphery of the weld overlay (at the repair-to-clad interface).
8.0 REFERENCES
- 1. Westinghouse WCAP-15987-P, Revision 2-P-A, "Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations," December 2003
- 2. Letter from H. N. Berkow (U. S. NRC) to H. A. Sepp (Westinghouse Electric Company), "Acceptance for Referencing -Topical Report WCAP-15987-P, Revision 2, 'Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations,' (TAC NO. MB8997)," dated July 3, 2003
- 3. Westinghouse WCAP-16401-P , Revision 0, "Technical Basis for Repair Options for Reactor Vessel Head Penetration Nozzles and Attachment Welds: Byron and Braidwood Units 1 and 2," March 2005
- 4. Letter LTR-NRC-03-61 from J. S. Galembush (Westinghouse Electric Company) to Terence Chan (U. S. NRC) and Bryan Benney (U.S. NRC), "Inspection of Embedded Flaw Repair of a J-groove Weld ," dated October 1, 2003
ISi Program Plan Units 1 and 2, Third Interval 10 CFR 50.55a RELIEF REQUESTS 13R-09 and 13R-20 Revision 2 (Page 10of10)
- 5. Letter from R. J. Barrett (U. S. NRC) letter to A. Marion (Nuclear Energy Institute),
"Flaw Evaluation Guidelines," dated April 11, 2003
- 6. Byron Station, Unit No. 2 - Relief Request 13R-14 for the Evaluation of Proposed Alternatives for lnservice Inspection Examination Requirements (TAC NO. MD5230)
- 7. American Society of Mechanical Engineers Boiler and Pressure Vessel Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds Section XI, Division 1"
- 8. Letter from N. L. Salgado (U. S. NRC) to P. A. Harden (FirstEnergy), "Beaver Valley Power Station, Unit No. 2 - Relief Request Regarding an Alternative Weld Repair Method for Reactor Vessel head Penetrations J-Groove Welds (TAC No. ME4176),"
Request 2-TYP-3-RV-03, February 25, 2011 (ADAMS Accession No. ML110470557)
- 9. Letter from R. Gibbs (U. S. NRC) to C. M. Crane (EGC) , "Byron Station, Unit No. 2 -
Relief Request 13R-14 for the Evaluation of Proposed Alternatives for lnservice Inspection Examination Requirements (TAC No. MD5230)," dated May 23, 2007
- 10. NRC Memorandum, "Byron Station, Unit No. 1 - Verbal Authorization of Relief Request 13R-19 -Alternative Requirements for Repair of Reactor Vessel Head Penetrations 64 and 76 (TAC No. ME5877)," dated March 29, 2011 11 . NRC Memorandum, "Byron Station Unit No. 1 - Verbal Authorization of Relief Request 13R-19 -Alternative Requirements for Repair of Reactor Vessel Head Penetrations Nos. 31and43 (TAC No. ME5948)," dated April 13, 2011
- 12. Letter from Jacob Zimmerman, (U.S . NRC) to M. J. Pacilio, (EGC), "Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 - Relief Requests 13R09 and 13R-20 Regarding Alternative Requirements for Repair of Reactor Vessel Head Penetrations (TAC Nos. ME6071, ME6073, and ME6074)," dated March 29, 2012, ADAMS Accession No. ML120790647
- 13. Westinghouse Report LTR-PSDR-TAM-14-005, Revision 3, "Techn ical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair," dated May 2015
- 14. Letter from J. Zimmerman (U .S. NRC) to M. Pacilio (EGC), "Byron Station, Unit No.
1 - lnservice Inspection Relief Request 13R-19: Alternative Requirements for the Repair of Reactor Vessel Head Penetrations (TAC Nos. ME5877 and ME5948) ," dated February 1, 2012
Attachment 4 Westinghouse Report LTR-PSDR-TAM-14-005, Revision 3 May 2015 Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair (Non-Proprietary)
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005-NP Revision 3 TECHNICAL BASIS FOR OPTIMIZATION OR ELIMINATION OF LIQUID PENETRANT EXAMS FOR THE EMBEDDED FLAW REPAIR May 2015 Author: W. H. Bamford* , Consu lting Engineer, Primary Systems Design & Repair Verifier: A. Udyawar*, Piping Analysis and Fracture Mechanics Approved: J. R. Stukus*, Manager, Structural Design and Analysis Engineering
- Electro11ically approved records are a11the11ticated i11 the Electro11ic Doc11111e11t Ma11agement System.
© 2015 Westinghouse Electric Company LLC All Rights Reserved 8 Westinghouse
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005-NP FOREWORD TI:iis document contains Westinghouse Electric Company LLC proprietary infonnation and data which has been identified by brackets. Coding <a,c.cJ associated with the brackets sets fo1th the basis on which the infonnation is considered proprietary. These codes are listed with their meanings in WCAP-7211 Revision 6 (March 2015), "Proprietary Infonnation and Intellectual Property Management Policies and Procedures." The proprietary infonnation and data contained in this report were obtained at considerable Westinghouse expense and its release could seriously affect our competitive position. This infonnation is to be withheld from public disclosure in accordance with the Rules of Practice 10CFR2.390 and the infonnation presented herein is to be safeguarded in accordance with LOCFR2.903. Withholding of this infomrntion does not adversely affect the public interest. This information has been provided for your internal use only and should not be released to persons or organizations outside the Directorate of Regulation and the ACRS without the express written approval of Westinghouse Electric Company LLC. Should it become necessary to release this infomrntion to such persons as prut of the review procedure, please contact Westinghouse Electric Company LLC, which will make the necessary arrangements required to protect the Company's proprietary interests. The proprietary information in the brackets has been deleted in this report, the deleted information is provided in the proprietary version of this report (LTR-PSDR-TAM-14-005-P Revision 3). Revision I: Provided more detail of examinations of embedded flaw repairs, for both Beaver Valley and Byron/Braidwood. Revision 2: Created proprietary and non-proprietary versions. Revision 3: Corrected an entry in Table 3 and editorial changes in Tables 2 and 3. Page 2 of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005-NP Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair ABSTRACT The embedded flaw repair for reactor vessel head penetrations was first applied at the DC Cook plant in 1996, and since that time it has been implemented on over 50 reactor vessel head penetrations world-wide. The technical basis for the approach has been accepted by the Nuclear Regulatory Commission (NRC) tJu*ough their published Safety Evaluation of WCAP- 15987-P Rev 2-P-A [1 ], December 2003. As an additional requirement of the safety evaluation, a Liquid penetrant examination was imposed on the embedded flaw repair weld each refueling outage after its implementation. While this requirement may have been a reasonable one when the process was new, the dose required for implementation of this examination is no longer justified. This rep01t provides a technical justification for eliminating the liquid penetrant examination. Some of the key points to be covered in this report are listed below:
- The original technical basis for these repairs has stood the test of time. Multiple layers have resulted in an impervious ba1Tier of highly resistant material, which continues to fulfill its intended purpose.
- The service history of these repairs has been excellent, with no failures of the repair to protect the head penetrations from Primaiy Water Stress Corrosion Cracking (PWSCC).
- AJloy 52/Alloy 152 applied weld material is highly resistant to PWSCC, with no crack initiations in over 21 years[5].
- The initial and post installation PT's perfom1ed on the embedded flaw repair weld established the integrity of the repair. Degradation with service leading to PWSCC is highly unlikely. No PWSCC of these repairs has been experienced to date.
- Follow-up PT exatns have not revealed any service-induced cracking or structural degradation. A boat satnple removed and examined at San Onofre proved that the penetrant indication found in the Embedded Flaw Repair (EFR) weld during service was not PWSCC[3].
- Continued UT examination of the repaired nozzle is sufficient to find degradation of the EFR and nozzle material.
- The UT leak-path examination technique is demonstrnted per LOCFR50 and is the current standard for inspection of nozzles which have not been repaired. This approach has become accepted since the embedded flaw repair was first licensed and will provide the same level of confidence/safety for the embedded flaw repaired nozzles. Therefore, additional exai11ination of these welds by PT adds little to any value.
- There is significant radiation dose associated with the penetrant examinations. This extensive dose is not justifiable in light of the extensive positive service experience for this repair teclmique.
- The zinc addition implemented at Byron and Braidwood adds further benefit towards mitigating PWSCC of the already resistant Alloy 52 weld material.
Page 3 of2 I
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14 -005-NP
- 1. BACKGROUND AND PURPOSE The embedded flaw repair technique was developed by Westinghouse in 1994, and involves the deposition of at least two layers of A11oy 52 weld metal to isolate existing flaws and susceptible material from the primary water environment (three layers on welds, to avoid chromium dilution).
The embedded flaw repair technique is considered a pennanent repair because as long as the degraded region remains isolated from the primary water (PW) environment, flaws cannot initiate or propagate due to PWSCC. Since Alloy 52 weld metal is highly resistant to PWSCC, a new PWSCC crack will not initiate and grow through the Alloy 52 overlay pem1itting the PWR environment to contact the susceptible material. In fact, these repairs have been in service for nearly 20 years now, with the longest single repair in place for over 10 years, and have perfonned well. The resistance of Alloy 690 and its associated weld metal, Alloys 52 and 152, has been demonsh*ated by laborat01y testing in which no cracking has been observed in sim ulated PWR enviromnents, and by approximately 24 years of operational service in steam generator hibes, where no PWSCC has occurred. The crack growth resistance of this material has been docmDented in reference [2], as well as numerous other papers. When the embedded flaw repair process was developed, as part of the regulatory review process, liquid penetrant examinations were requ ired every outage, to ensure the reliability of the process. Although these examinations may have been wan-anted initially, the extensive service experience to be discussed here will demonstrate that they are no longer needed.
- 2. EMBEDDED FLAW REPAIR: OVERVIEW J-weld flaws have a complex, three dimensional geometry, which does not lend itself to conventional repairs. Since conventional repair was considered impractical and ineffective, Westinghouse developed a unique repair alternative called the embedded flaw repair method.
This method enables the installation of a non-structural Alloy 52/52M weld barrier that effectively isolates the Control Rod Drive Mechanism (CROM) rube and J-weld from Reactor Coolant System (RCS) water. This Westinghouse design was developed by over lO years of direct interface with nuclear Owners (both US and international) and regulatmy authorities, culminating in a fomrnl Westinghouse report [l] recognized and accepted by the US NRC. Regulatory endorsement of this Westinghouse methodology has become routinely accepted by plant owners and the NRC as an acceptable method to prevent further PWSCC degradation to reactor vessel head penetrations. This repair method has also been licensed and implemented in both Japan and Korea, as well as two countries in Europe. Several factors make J-weld repair welding difficult to address. The J-welds themselves are located on the inside of the RPV bead. The result is that each nozzle is unique, in that its location relative to the outer edges of the head results in a constantly varying degree of curvature in the J-weld itself. Welding challenges associated with this cm*vatm*e are further complicated by the fact that the CRDM tubes are all vertically oriented. This means that each J-weld, as it extends armmd the circumference of the CROM tube, has significant variations in height. Also, the J-weld itself, due to the fact that it welds a vertical CROM h1be to a sloped vessel ID surface, is oval in shape. In light of the high radiation levels (typically 2 to 3 REM per hour) underneath the reactor vessel head, any attempt at manual repair of the J-welds will result in unacceptably high radiation exposure levels; levels that are both undesirable, costly, and contrmy to Utility AIARA Page 4 of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005-NP objectives. These exposure levels make remote, machine welding a necessity. They also make liquid penetrant examinations of the completed welds a costly endeavor. To address these and other concerns, PCI designed and manufactured a welding system specifically tailored for installation of a J-weld overlay. This four-axis, custom manufactured weld head is unique in the industry, in that it can deposit a high quality weld in this challenging environment[ I]. PCI began this equipment development effort in 2000. Over the following years, PCI has made major upgrades, enhancements, and improvements in this welding equipment. The result is a proven, robust, and reliable equipment set capable of consistent, schedule-effective delivery of quality welds. Specifically, the repair involves at least two (offset) layers of weld metal over the head penetration base metal, and at least three layers over the J-groove weld. The multiple layers are to ensure that no path exists to allow the PWR water to contact the susceptible materials. The additional layer over the weld region is to minimize dilution of the chromium.
- 3. EMBEDDED FLAW REPAIR EXPERIENCE Embedded flaw repairs serve to mitigate reactor vessel head penetration (RV.HP) flaws in the following locations: Tube-to-vessel J-groove weld; Tube OD, and; Tube ID. For purposes of this docwnent, the review of EFR perfonnance history will focus only on OD repairs, i.e., repairs that deposit weld metal on the surface of the Tube-to-vessel J-groove weld and the OD surface of the tube. This constitutes the majority of the repairs which have been implemented.
Over fifty EFR OD repairs have been installed in at least 12 separate nuclear power plants. Thirty six are currently in service; others have been removed from service by head replacement or other reasons not related to EFR effectiveness. These repairs are summarized in Table 1. uclear plants where EFR's have been removed from service include North Anna 2, Arkansas I, Beaver Valley 1, Ohi 3 (Japan), San Onofre 3, Hanbit 3, and DC Cook 2. Nuclear plants with in-service EFR's include Beaver Valley 2, Byron 1, Byron 2, Beznau, VC Summer, Braidwood 1, and Hanbit 4 (Korea). Service exposure duration for individual EFR varies among units, with 9 years constituting the longest period of service exposure to date for an OD repair. (An ID embedded flaw repair implemented at the DC Cook plant was in service for ten years, without any evidence of degradation, before the head was replaced, in 2006. Details are provided in Appendix A.)
- 4. PT EXAMINATION DETAILED HISTORY Every EFR is presently required to be PT examined every outage, as a result of an NRC condition imposed on the generic relief request, which was approved in July 2003[1 ]. To date, no PT examination has shown evidence of PWSCC in EFR deposits; however, PT examinations have periodically identified fabrication flaws and/or discontinuities in installed EFR deposits. The following paragraphs summarize PT examination results for in-service EFR welds.
Of the over 50 EFR 's of the OD that were placed in service (see Table I), 13 were install ed in 2013 and 2014 (5 at VC Summer, 5 at Hanbit 4, 2 at Beaver Valley 2, and I at Byron 2). These newly installed E FRs have not yet completed their first cycle of operating service, and therefore have not been PT examined subsequent to being placed in service. PT examinations have been perfo1111ed on each of the remaining repairs dw*ing subsequent refueling outages, and these PT examinations have been repeated for the entirety of the service life of each EFR. TI1is means that, for these installed EFR's, a large number of outage-related PT examinations have been Page5of21
Westinghouse Non-Proprietaty Class 3 LTR-PSDR-TAM-14-005-NP performed. The sites with the longest service of these repairs are Beaver Valley, Byron, and Braidwood. The history of these repairs will be reviewed in detail below, and the history of San Onofre will be reviewed, since this was the site of the most in-depth evaluation of an indication found during the required liquid penetrant (PT) exams.
- 5. BOAT SAMPLE REMOVAL AND TESTING FROM SAN ONOFRE UNIT 3 [3]
The first example of PT indications observed on an embedded flaw repair surface was at San Onofre Unit 3 in October of 2008. As a result of this finding, the NRC was concerned that the observed cracking in the repair weld might be PWSCC, and so a boat sample was removed and examined futiher. The repair had been in service since the EFR was completed in 2004, and had undergone an acceptable PT examination in 2006, after one cycle of operation. The boat sample contained a rejectable rounded indication, in the EFR for penetration #64. The examinations included visual inspections, stereo-visual inspections, X-ray radiography, high resolution replication, extended dwell fluorescent PT, scaiming electron microscopy (SEM), energy dispersive spectroscopy, and optical metallography. The primary purpose was to identify the most likely cause of the delayed appearance of the PT indication. Thin fragments of material (0.0005" or less) were identified surrounding the entrance to the rejected void. These Iigaments were found to likely explain the delayed presentation of the void, because the underlying cavity had been protected from the surface by a very thin layer of Alloy
- 52. The exact cause of the failure of this layer is not known, but it is likely that the protective ligaments failed due to operational stresses or cleaning efforts following the 2006 penetrant examination. The evidence obtained did not support PWSCC, thus demonstrating that the Alloy 52 weld material continued to protect the Alloy 600/82/182 from the PWR water.
- 6. POTENTIAL CAUSES OF PT INDICATIONS IN EFR WELD REPAIRS Several factors can affect PT examination accuracy and results. These factors can contribute to differing PT results on a given weld from one examination to the next, even when the examination is perfom1ed under ideal conditions. These factors include:
Minor Changes in Weld Surface: EFR weld surfaces are PT examined upon initial installation, and rejectable (ASME Section ill criteria) indications are removed. Upon initial introduction to service conditions, EFR weld surfaces are ASME Section III compliant. Themlal expansion, vessel head dilation due to pressurization, and water flow can have an effect on the EFR surface. For example, a small welding flaw in an EFR surface may be present, but the flaw face (i.e., the portion of the flaw exposed to the EFR weld surface) may be compressed too tightly to permit penetrant absorption. The face of this small welding fl.aw, when exposed to service conditions, may be slightly changed, widening the gap at the weld surface. Upon subsequent PT examination, this indication surface may now have sufficient width to enable penetrant entry, causing it to appeai* as rejectable during subsequent examinations. Such indications do not indicate failure of the EFR; rather, they are indicative of minor change in surface configuration Page 6 of2l
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM 005-NP resu lting from service exposure. A similar situation may occur in areas where separate weld passes/beads join together. These areas often constitute geometric discontinuities, and (as explained previously) can cause penetrant to become entrapped. Operating conditions may cause minor additional changes in the surface geometry in these locations, and these slight changes can increase the risk of penetrant entrnpment. An example is shown in Figures l and 2. These changes may then preclude effective cleaning during subsequent PT examinations. When these situations occur, metal removal is typically required to smooth the area, and restore a configuration suitable for PT acceptance. Again, these minor configmation changes do not constitute weld failure of the EFR; they constitute only a condition wammting smface conditioning to facilitate the PT process, thus confinning the integrity of the weld . This situation is the most common somce of indications which are found in subsequent PT inspections. Since the OD of the head penetrations near the J-groove weld is located in a very high dose region, subsequent surface preparation by grinding has a significant in1pact on personnel radiation exposure, and is only perfonned when absolutely necessary to faci litate the PT process . It is also possible that EFR welds contain welding flaws that, at the time of introduction to service, are not open to the weld surface, and are not, therefore, detectable by PT. Service conditions can cause welding flaws of this nature to become surface exposed, and any such exposed flaws would likely appear as PT indications during subsequent PT examinations. Indications of this type may be rounded (i.e., porosity) or linear (i.e., lack of fosion between beads, such as localized flaws at tie-in areas). When flaws of this nature are detected, they are repaired by excavation and, where necessaiy, localized weld repair (typically manual GTAW). Indications of this nature have occuned in limited situations, and repairs have been successfully performed, as confomed by subsequent PT inspections. The EFR specifically employs multiple weld layers to ensure weld integrity, and flaws of this nature are typically limited in size and depth to less than one weld layer. As demonstrated by ongoing EFR inservice experience, these types of welding flaws do not compromise the integrity or acceptability of the EFR weld to perform its intended purpose. As-Welded Surfaces: Embedded flaw repair welds require that the final PT be perfonned on an as-welded surface, which requires careful and thorough cleaning to adequately remove penetrant prior to developer application. Due to inherent differences in cleaning from one examination to the next, some variation in indication size is an unavoidable aspect of the PT examination process. It is possible, therefore, that the original surface (with or without a preexisting welding flaw) has not changed, but that the more recent PT examination produced different results due to this variability between PT examinations. This is not indicative of an unacceptable examination technique; rather, it is a degree of examination variabil ity inherent in the PT process. The ASME Code recognizes the liquid penetrant test method to be inherently susceptible to indications of this nature, and cites ' surface conditions' (Ref. NB-535l(a)) as a common contributing factor. ASME Section ITl , NB-535 l (2010 Edition) states: Page 7 of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM- 14-005 -NP NB-5350 LIQUID PENETRANT ACCEPTANCE STANDARDS NB-5351 Evaluation of Indications (a) Me chanical discontin uities at the s urface are re-vealed by bleeding out of t he penetrant; however, localized su rface discontin uities, such as may occur from machi ning marks, surface cond itions, or an incomplete bo nd betwee n base metal and daddin g, may produce simila r indications which are non releva nt. (b) Any ind ication which is believed to be nonre levant shall be reexami ned to verify whether or not actual defects are present. Su rface conditi oning may precede the reex-amination. Nonreleva nt indications and broad areas of pigmentation which would mask defects are unacceptable. (c) Releva nt ind ications are ind ications which res ult fr om impe rfections. Linear ind ications are indications in which the length is more than t hree ti mes the wid th. Rounded indications are indica tions which are circular or ell ipti cal with the length equal to or less than three times the wi dth. As confim1ed by ASME, ' surface conditions' can produce PT indications which may initially appear rejectable. ASME addresses this issue by specifically pennitting minor metal removal to address indications of this nature. Accessibility for Exan1ination: As described in the preceding paragraphs, as-welded surfaces pose tmique challenges to the PT process. Careful removal of penetrant from valleys between weld beads is essential for accurate EFR PT, and this process can be challenging. EFR repair welds are located in the high radiation environment within the Reactor Pressure Vessel head, and radiation controls require strict limitations on entry times. Work near the surface of the RPV head, which is an inherent aspect of EFR PT's, increases radiation exposure levels. The welds themselves are located overhead, requiring the PT examiner to reach upward and apply considerable pressure to the surface of the weld, in order to achieve effective penetrant cleaning. Accessibility to these welds may be lim ited by themrnl sleeves, guide funnels, and other obstructions, any of which increase the difficulty of the exan1ination. These factors combine to increase the difficulty of precise implementation of the PT process, and introduce an inherent level of variabil ity in the examination process. This inherent variability may result in differences in examination results from one examination to the next, even when no changes have occurred in the weld surface. Variability in the PT examination process is unavoidable, and may lead to differences in PT examination results. 1t is weU known that, although these differences exist, they do not affect the functional integrity of the embedded flaw repair, or any other pressure boundary structure. If they did, the ASME code would not allow this flexibil ity.
- 7. BRAIDWOOD PENETRATION 69 RESULTS:
In 2012, Westinghouse imp lemented an Embedded Flaw Repair (EFR) to mitigate fl aws in RPV Head Penetration P69 at Braidwood Unit I. Th is repair wel d was success fu lly install ed, accepted by liquid penetrant (PT) testing, and the unit wa s rehrrned to servi ce. After one cycle, PT examination was re-perfo rmed on the P69 EFR as required. This examination identified 27 PT indications. Page 8 of2l
Westinghouse Non-Proprieta1y Class 3 LTR-PSDR-TAM-1 4-005-NP All indications were rounded; none were linear. Of the 27 indications, 5 were non-relevant (i.e., <I /16" diameter), and 9 were acceptable (i.e., not rejectable in accordance with ASME Section III PT acceptance criteria). The remaining 13 rounded indications were deemed rejectable and required remediation. Minor grinding completely removed two of these welding defects, leaving 11 that required additional remediation. Continued grinding removed the remaining welding defects, and localized manual welding was perf01111ed where it was necessmy to restore EFR thickness. Final PT accepted these localized weld repairs. As explained elsewhere in this rep01t, a number of factors can contribute to detection of PT indications in previously accepted EFR weld surfaces, and any of these factors may have played a role in this application. An exact explmrntion regarding the nature/cause of these PT indications is not possible. Lacking other exculpato1y evidence, it must be acknowledged that there were 11 EFR weld defects in Penetration 69 requiring excavation, and application of three layers of weld metal. The nature and extent of these welding defects is evidenced by the extent of the repair effott required to mitigate each. Thi s repair effort was localized and limited in extent, demonstrating that none of these 11 welding defects adversely affected the EFR's continued suitability for service. Each flaw has subsequently been removed (with removal verified by PT examination) and repaired. No degradation of the original Alloy 600/821182 material was identified during any of these repairs.
- 8. SERVICE HISTORY OF EMBEDDED FLAW REPAIRS AT BYRON AND BRAIDWOOD PLANTS Table 2 provides a summary of the repair hfatmy of the head penetrations at Byron and Braidwood units. No indications have been found in Braidwood 2, but the other three have had indications, and therefore embedded flaw repairs have been performed.
In reviewing Table 1, it becomes clear that there is a cost to the PTs that have been catTied out as an additional requirement of the repair methodology. Also, when indications are discovered and grinding or welding is required, the man-rem dose increases significantly. This dose needs to be compared with the benefits obtained by the process. Since no PWSCC flaws have been found, the cost, once welding flaws have been eliminated, is not justified. In light of the service experience of Alloy 52 welds, which will be discussed below, PWSCC is not likely to occur in the remaining operating life of the plant.
- 9. SERVICE HISTORY OF EMBEDDED FLAW REPAIRS AT BEAVER VALLEY PLANT Table 3 provides a summmy of the repair history of the head penetrations at Beaver Valley units I at1d 2. 11Je head at Unit l was replaced in 2006, so the histo1y is rather short, but that has not been the case fo r unit 2 .
Review of Table 3 further supp01ts the conclusion that the implementation of penetrant exams for the embedded flaw repair has been costly, in terms of radiation exposure. Although there were several cases where grinding was required to eliminate the PT exam indication, there is no evidence that the grinding extended beyond the repair, into the original base metal. As with Byron Page 9 of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005-NP and Braidwood above, the dose associated with grinding and/or welding of these already completed repairs is significant, and provides no added value.
- 10. RESISTANCE TO CRACKING OF ALLOY 690 AND ITS ASSOCIATED WELDS Alloy 690 is known to be much more highly resistant to PWSCC than Alloy 600. This conclusion is also true for its associated welds, Alloy 52 and 152, and results from increased resistance to crack initiation, as well as crack growth.
From the standpoint of crack initiation, there are no known incidents of PWSCC for Alloy 690 materials in service. While this does not prove that the material will not initiate cracks, it does argue strongly that the material is much more resistant than Alloy 600. One of the most challenging locations for either Alloy 600 or Alloy 690 is the steam generator, where the use of Alloy 600 has been prevalent since the earliest steam generators. Leaks were observed in Alloy 600 steam generator tubing in the first few fuel cycles. The leakage and detected SCC cracks led utilities to replace steam generators when the percentage of plugged and degraded tubes rose to levels that caused operational and economic difficulties. As a result, some Alloy 600 steam generators were replaced within 7 years service and as little as 3.6 EFPY of operation. Conversely, Alloy 690 steam generator tubes have been in service since 1989, a period now exceeding 25 years, with no observed stress conosion cracking. Looking at this susceptibility another way, one could compile the 'effective degradation years ' (EDY) when cracking occurred for the two materials. An effective degradation year has been defmed as one effective full power year at 600°F (3 l6°C) [1]. Looking at the service experience from this point of view, we see that the earliest steam generator replacement occuned in Alloy 600 mill annealed after 4 EDY due to extensive SCC problems, while the longest service for an Alloy 690 steam generator has been over 30 EDY, with no cracking. In the US fleet of PWRs, 59 PWRs started operation with mill annealed Alloy 600 steam generator tubes. As of the end of 2009, 52 of those 59 PWRs have replaced their steam generators, principally due to stress corrosion cracking issues. Forty-one of those utilities have replaced their steam generators with new steam generators tubes with thennally treated Alloy 690 tubes. Considering that there are typically 8000 to 20000 steam generator tubes in a PWR and multiple high stress locations in a single tube, the Alloy 690 operating experience is impressive. The remaining replacement steam generators and some 10 original steam generators were tubed with thermally treated Alloy 600 that had improved microstructural features and lower residual stresses. The lead plants with this tubing had operated about 35 EDY when the first SCC related plugging operations occurred. The operation of thennally treated Alloy 690 tubing would be expected to be significantly improved over the Alloy 600 TT. Based on the above, there is at least a 25 calendar year lead time wllich would provide advanced warning of any potential trouble for Alloy 690. Similar service experience has been recorded for Alloy 152/52 welds, except that the first weld went into service in contact with a primary water environment in 1994, about five years later. Therefore, operating history provides evidence of at least 2 I years of crack-free service for Alloy 152/52 welds. From the standpoint of crack growth, it is clear that PWSCC has been observed in laboratory tests Pagel0of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005-NP for Alloy 690 in the standard PWR environment. Experimentally, it is difficult to get a crack to initiate, and also very difficult to keep the crack growth going long enough to obtain a measurement. The only successful attempts have employed a fatigue pre-sharpened crack, and the growth rates that have been obtained are about two orders of magnitude slower than that of Alloy 600 [2]. Therefore, we can conclude that these materials are very highly resistant to PWSCC. Although cracking can be measured, crack initiation has not been seen in service or in the lab, and if a flaw does not initiate, it will ce1tainly not propagate.
- 11. DISTORTION EFFECTS OF THE EMBEDDED FLAW REPAIR PROCESS One factor which can enhance the PWSCC growth rate of both Alloy 690 and Alloy 52/152 welds is cold work [2]. Although cold work has not been purposely applied, analytical work was recently completed to quantify the amount of distortion introduced by the EFR process.
The embedded flaw repair was developed, and continues to fw1ction as an effective SCC-resistant baITier between the RCS water and the SeC-susceptible J-weld and eRDM tube. The repair is not intended to be structural, and therefore does not need to meet ASME Section III requirements. It is also important to mention that there are no Section III design requirements for residual stresses, since they do not affect potential failure (even though they do affect crack growth). That said, however, great care was taken in the design of the repair technique, and especially the weld thickness of the EFR, to minimize any distortion of the tube as a result of the welding process, and the repair has been effectively applied to many bead penetrations over the past 20 years. Since it is well known that cold work can degrade the sec resistance of Alloy 52/152 welds, it is important to examine the amount of distortion in1posed by the embedded flaw repaiJ". To investigate this issue, a detailed analysis was recently completed for the repair method, as applied to an operating plant [4]. The maximum distortion of the tube ID was estimated to be 0.0 I inch from this analysis. Measurements were made of the actual distortion of the tube after the embedded flaw repair was actually completed, and the maximLm1 distortion of the tube was detem1ined to be less than 0.1 mm, or 0.0039 inch[4] . Therefore it can be concluded that the entire repair causes little or no distortion or cold work of the tube. A small area of welding to refill the area removed by grinding would cause even less distortion, and would therefore be of no concern.
- 12.
SUMMARY
AND CONCLUSIONS The embedded flaw repair (EFR) process has been described in detail, and it was pointed out that the requirement for PT of the OD of the repair surface was not pait of the original proposed inspections, but was added as part of the Safety Evaluation Rep01t for the process. At the time of the original implementation of the generic relief request, the UT leak path examination was not accepted, but now it has replaced the need to even examine the J-groove welds. The summary provided in this report indicates excellent service performance of Westinghouse Embedded Flaw Repairs. The penetrations with PT indications represent a relatively small portion of installed EFR welds, and all of the embedded flaw repairs continued to perform their Page 11 of2 l
Westinghouse Non-Proprietaiy Class 3 LTR-PSDR-TAM-14-005 -NP function, which is to protect the susceptible Alloy 600 tubes from the water environment. Each of these penetrations which were found to have indications has subsequently been repaired and PT-accepted. The nature of each of these repairs confinned no PT indications compromised the integrity or perfonnance of the associated EFR. In each case, each EFR was restored to a fully acceptable condition by minor buffing/grinding and, where needed, limited manual repair welding. In no case was PWSCC identified in these welds. Frnthennore, the absence of PT indications for the remaining PT examinations demonstrates that PT indications are not typical in these repair welds. As can be seen in these results, the quantity of PT indications detected on the Braidwood P69 EFR differs significantly from all prior PT examination results, making P69 results atypical. Significant radiation dose is incuITed as the result of each liquid penetrant exam, and this dose is increased by an order of magnih1de if grinding or welding is needed to clear the PT indications. Counting the man-Rem dose for just the Beaver Valley and Byron/Braidwood plants, the dose is above 85 man-Rem. This dose is not justified in light of the negligible advantage gained with respect to the integrity of the head penetration. The EFR is a complex repair developed for implementation using entirely remote teclmiques, so as to minimize man-rem dose. This is counteracted by the requirement for PT every outage. UT inspection of the entire volume of the head penetrations is already required, but this can be accomplished remotely, with minimal exposure. Any evidence of further growth on indications in the susceptible tube material can be obtained from this remote examination. The resistance to PWSCC of Alloy 52/l 52 and Alloy 690 continues to be confinned by testing and operating experience. CuITently, at least 21 years of experience exists in the welds with no initiation, regardless of application. Contrast that with only a few years of experience before PWSCC occun-ed with Alloy 600 and its welds, and the EFR has proven to be a robust process. Efforts are continuing to forther improve its reliability. The fast embedded flaw repair to have PT indications (San Onofre Unit 3) has been sampled, and carefully examined, and two key findings emerged. First, it is likely that the protective ligaments failed due to operational stresses or cleaning efforts following the 2006 penetrant examination. The protective boundary provided by the EFR was still intact. Secondly, the evidence obtained confinned that PWSCC was not present. Since no PWSCC tlaws have been found, and there has been no evidence of new or continued degradation of the original Alloy 600/82/182 material protected by the EFR, the cost of PT exams every outage does not seem to be justified. In light of the service experience of Alloy 52 welds, it does not seem likely that any PWSCC is going to occur in the future, at least for the next 21+ years. Future perfonnance of the EFR can be monitored using the same UT examinations currently perfonned for the other reactor vessel head penetrations. Page 12 of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005 -NP
- 13. REFERENCES
- l. Bamford, W.H. "Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations" WCAP 15987-P, Rev. 2-P-A, December 2003.
- 2. Bamford, W.H., Alloy 690 and Weld Metal Reference SCC Growth Rate Models for ASME Section XI" (with Guy DeBoo) in Proceedings Sixteenth International Conference on Environmental Degradation of Materials in Nuclear Power Systems, NACE, 2013.
- 3. Hyres, J. W., "Filial Report: Laboratory analysis of a Boat sample Removed from CEDM #64 at San Onofre", B&W Repo11 S-1456-001, May 2009.
- 4. Ng, Chris K., "Finite element Residual Stress Analysis Results for the Yonggwang Unit 3 Outermost Reactor Vessel Control element Drive Mechanism Penetration Nozzle Before and After the Embedded Flaw Repair Process", Westinghouse Letter report LTR-PAFM-13-43, Rev. O,April 2013.
- 5. Gonrnm, J., "Stunmaiy of Available Laborato1y PWSCC Initiation Data for Alloys 690, 52, and I 52, and Identification of Desired Additional Initiation Data", presented to Alloy 6901521152 Collaborative meeting, Tampa, FL, November 2013 .
Page 13 of2 I
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM- 14-005-NP Table 1: History of OD Embedded Flaw Repairs a,c,e Page 14 of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM- 14-005-NP Table 2: Summary oflnspection and Repair History of Embedded Flaw Repairs at Byron and Braidwood Plants New Repair PT Results Buffing Welding PT and Repair Required Required Exposure Byron B1R1 7 4 Repairs Acceptable NIA NIA NIA P31143164176 after Repairs completed B IR18 P3 1 - 1 Yes No 5. 701 Rem indication rejectable P43 - 1 indication rejectable BJR19 P3 1 - 1 No No 1.439 Rem indication acceptable P43 - l indication acceptable B2Rl3 1RepairP68 NIA B2Rl4 P68 No No 697 mrem A cceptable B2R l 5 P68 No No 345 mrem Acceptable B2R16 P68 No No 422 mrem Acceptable B2Rl7 P68 No No 428 mrem Acceptable B2Rl8 I Repair P6 P68 Yes Yes* 15 .694 ~em Acceptable " -, Note, the repair of P6 Braidwood resulted in approximately AlRl6 1RepairP69 Acceptable NIA NIA NIA 15 Rem . The remainder after Repairs of the exposure indicated here is from the PT of completed P68 . AlR17 P69 - 13 Yes YES 10 10.333 Rem indications indications rt:iectable required welding AlR18 P69 - l Yes No 1.028 Rem indication rejectable Total: 36.087 Rem
- Weld ing associated with repair only Page 15 of 2 1
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM- 14-005-NP Table 3: Summary oflnspection and Repair History of Embedded Flaw Repairs at Beaver Valley Plants a,c,e Page 16 of2 l
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005 -NP .F igure 1: Example of an indication at the boundary of the vertical weld passes on the penetration tube, and the elliptical weld passes on the head Page 17 of 21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM- 14-005-NP Figure 2: Closer View of the indication of Figure l Page 18 of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM-14-005-NP Appendix A DC Cook Unit 2 Embedded Flaw Repair Experience One of the best examples of service experience of an Alloy 52 weld repair is provided by the experience of the D.C. Cook Unit 2 embedded flaw repair, because of the length of service provided by this repair. Penetration number 75 at this plant was found to have an inside surface flaw with a depth of approximately 40 percent of the tube wall thickness. This penetration was repaired with the embedded flaw repair process in 1996, and the repair was re-inspected in January of 2002. The inspection of January 2002 was catTied out with both dye penetrant and eddy cunent testing. The penetrant examination showed no indications, as did the eddy cuJTent testing. The eddy cunent results are more quantitative, and will be discussed here in some detail The method was demonstrated and qualified under a progrmn in response to the NRC Generic Letter 91-01. The process uses an eddy current coil with high-resolution gray scale imaging, with a magenta response at 50 percent of the amplitude of the calibration notch (0.004 inch long and 0.040 inch deep). This was shown empirically to conespond to the response to actual PWSCC. An example of such a response is shown in Figure 1, which shows actual clustered axial flaws in a penetration tube. The coil design is optimized for high spatial resolution, in order to distinguish individual responses among clusters of cracks, such as those shown in Figure l . This eddy cunent testing and display process was applied to the D.C. Cook penetration 75 in January 2002, and the results are shown in Figure 2. The results show no evidence of cracking after six years of service. Further inspections conducted up until the head was replaced in 2006 also showed no evidence of cracking. Therefore, the ID embedded flaw repair was confim1ed to last at least 10 years with no deterioration. Page 19 of21
Westinghouse Non-Proprietary Class 3 LTR-PSDR-TAM- 14-005-NP Figm*e 1 ECT View of Craze Cracking ~t-d(ly <;111n *1 11 /\11,ily ~i~ DCC :Jo _. / I _02 - Lockr*c.J *"""~ Fi1e F.!:_equencv Probe C - Scan LissaJ_ous Too1s S~ttings File : 0 C_2_7 4 _02 x*m Oat*: 01/26/2002 Tim*: 09:31 - 1 0:43 F re u*nc~: 1 Pro b e: 1 Probe T ype: Driver P ick up G i n: 3 2.0 dB Drive: 1.2 .0 V Fr e q: 4 00 Hz R o*t
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Westinghouse Non-Proprietary Class 3 LTR-PS DR-TAM- 14-005 -NP Figure 2 ECT View of 1996 Repair, Taken in 2002 1Mlysb PCC2_/~,_Ul Luckl*cl .. ~- .a!dgj~ E_i.le F!:equen c y ~robe ~-S een Lissa.Joos !ools S.!_l.ti.ns s !!e lp fi le: DCC2_7!5.._ i ( ><*"' Oat.*: 01127/ 20 2 Tin*: 0 9 : 3 2 - 09:33 r r equ.nc y: 1 Probe: 1 Probe Tu pe: Driver ,.1ckYp Ga in: 32 .0 dB Ori v*: 12 .0 V Fr*q : 4 001< H:z: Rot.at.ion : 98 d-e O~r*t-or: Unknown COl'lp onent: ()pein ou*ina CROM Ser No : 0 Sc*n Ad*: Clre Tro111: 0.000 To: 370.000 Intv: 1.000 Off"* 1. : 0.0000 lnde)( Ad*: Axi*l F"°": O.~O To : 11.090 Intv: 0.040 Of"f'*e t : 0.0000 0 0 9.0 80* 60-V*rt (ECU ) 8.0 20 7.0 6.0
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Attachment 5 Westinghouse Electric Company, LLC Affidavit for Report LTR-PSDR-TAM-14-005, Revision 3 May 2015
@Westinghouse Westinghouse Electric Company Engineering, Equipment and Major Projects 1000 Westinghouse Drive, Building 3 Cranberry Township, Pennsylvania 16066 USA U.S . Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8560 11 SSS Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 208S2 Proj letter: CAE-15-35/CCE- I 5-35 CAW-lS-4199 May 28, 201S APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
LTR-PSDR-TAM-14-005-P, Revision 3, "Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair" (Proprietary) The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CA W-15-4199 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations. Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Exelon Corp., LLC. Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse Affidavit should reference CA W-15-4199, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066. Very truly yours, i~+J:;;:nager Regulatory Compliance Enclosures
CAW-15-4199 May 28, 2015 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA: SS COUNTY OF BUTLER: I, James A. Gresham, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.
/J..James A. Gresham, Manager Regulatory Compliance
2 CAW-15-4199 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse. (2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit. (3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld. (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse. (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for. determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of
3 CAW-15-4199 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies. (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability. (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product. (d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse. (f) It contains patentable ideas, for which patent protection may be desirable. (iii) There are sound policy reasons behind the Westinghouse system which include the following: (a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position. (b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
4 CAW-15-4199 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries. (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage. (iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission. (v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief. (vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-PSDR-TAM-14-005-P, Revision 3, "Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair" (Proprietary) for submittal to the Commission, being transmitted by Exelon Corp., LLC letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the technical basis for optimization or elimination of liquid penetrant exams for the embedded flaw repair, and may be used only for that purpose.
5 CAW-15-4199 (a) This information is part of that which will enable Westinghouse to: (i) Provide guidance on optimization or elimination of liquid penetrant exams for the embedded flaw repair (ii) Provide guidance on performing and examining embedded flaw repair (b) Further this information has substantial commercial value as follows: (i) Westinghouse plans to sell the capability to perform embedded flaw repairs. (ii) The information requested to be withheld reveals plant specific information that was used by Westinghouse. Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar environmental fatigue screening and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money. In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended. Further the deponent sayeth not.
PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of documents f umished to the NRC in connection with technical basis for optimization or elimination of liquid penetrant exams for the embedded flaw repair, and may be used only for that purpose. In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l). COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.}}