ML16358A444

Relief from the Requirements of the ASME Code Regarding Alternate IP3-RR-10 to the Full Circumferential Inspection Requirement of Code Case N-513-3

ML16358A444
Person / Time
Site:	Indian Point
Issue date:	01/11/2017
From:	Stephen Koenick Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Pickett D, NRR/DORL/LPL1, 415-1364
References
CAC MF8792
Download: ML16358A444 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 11, 2017 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 - REQUEST FOR RELIEF FROM THE REQUIREMENTS OF THE ASME CODE REGARDING IP3-RR-10 ALTERNATIVE TO THE FULL CIRCUMFERENTIAL INSPECTION REQUIREMENT OF CODE CASE N-513-3 (CAC NO. MF8792)

Dear Sir/Madam:

By letter dated November 7, 2016, Entergy Nuclear Operations Inc. (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, IWD-3100, for the repair of degraded service water system piping at Indian Point Nuclear Generating Unit No. 3 (Indian Point 3).

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee submitted Relief Request No. IP3-ISl-RR-10 on the basis that compliance with the specified ASME Code requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The relief request is for an alternate repair of a leaking 3-inch diameter service water system pipe associated with the number 31 fan cooling unit at Indian Point 3.

On November 7, 2016, the U.S. Nuclear Regulatory Commission (NRC) staff verbally authorized the use of Relief Request No. IP3-ISl-RR-10 for the remainder of the fourth 10-year interval until the 3R19 refueling outage, which is scheduled for March 2017, or until the flaw size exceeds the acceptance criteria. The enclosed safety evaluation documents the technical basis for the NRC's verbal authorization.

The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the subject piping and that complying with the ASME Code requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Accordingly, the staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the staff authorizes the use of Relief Request No. IP3-ISl-RR-1 O for the remainder of the fourth 10-year interval until the 3R19 refueling outage, which is scheduled for March 2017, or until the flaw size exceeds the acceptance criteria at Indian Point 3.

All other requirements of the ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.

If you have any questions, please contact Douglas V. Pickett, at 301-415-1364 or by e-mail at Douglas.Pickett@nrc.gov.

Sincerely, Stephen S. Koenick, Acting Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NO. IP3-ISl-RR-10 ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

1.0 INTRODUCTION

By letter dated November 7, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16314E601}, Entergy Nuclear Operations Inc. (the licensee),

requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code}, Section XI, IWD-3100, for the repair of degraded service water system piping at Indian Point Nuclear Generating Unit No. 3 (Indian Point 3).

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), the licensee submitted Relief Request No. IP3-ISl-RR-10 on the basis that compliance with the specified ASME Code requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. The relief request is for an alternate repair of a leaking 3-inch diameter service water system pipe associated with the number 31 fan cooling unit at Indian Point 3.

On November 7, 2016 (ADAMS Accession No. ML16313A012}, the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff verbally authorized the use of Relief Request No. IP3-ISl-RR-10 for Indian Point 3 for the remainder of the fourth 10-year interval until the 3R19 refueling outage, which is scheduled for March 2017, or the flaw size exceeds the acceptance criteria at Indian Point 3. This safety evaluation documents the technical basis for the staff's verbal authorization.

2.0 REGULATORY EVALUATION

Section 50.55a(g)(4) of 10 CFR states, in part, that ASME Code Class 1, 2, and 3, components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code, Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components."

Enclosure

Section 50.55a(z) of 10 CFR states, in part, that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if the licensee demonstrates that

( 1) the proposed alternatives would provide an acceptable level of quality and safety or (2) compliance with the specified requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the NRC staff to grant, the relief requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 ASME Code Component Affected The affected component is ASME Code Class 3, butt weld 8297, on the 3-inch diameter service water system return piping associated with the number 31 fan cooler unit. The nominal wall thickness for service water piping is 0.216 inches. The maximum design pressure and temperature are 150 pounds per square inch gauge (psig) and 169 degrees Fahrenheit (°F),

respectively. The material of construction of the subject pipe is Haynes Alloy 20 (Unified Numbering System (UNS) N08320) or 904L stainless steel (UNS N08904). The weld filler metal is ASME SFA 5.11 ENiCrMo-1 or ASME SFA 5.14 ERNiCrMo-1. The pipe has no internal lining.

3.2 Applicable Code Edition and Addenda The code of record for the inservice inspection fourth interval is the ASME Code, Section XI, 2001 Edition through 2003 Addenda.

3.3 Applicable Code Requirement The ASME Code, Section XI, Articles IWD-3120 and IWD-3130, require that flaws exceeding the acceptance criteria be corrected by repair or replacement activities or be evaluated and accepted by analytical evaluation. Subarticle IWD-3120(b) requires that components exceeding the acceptance standards of IWD-3400 be subject to supplemental examination or to a repair or replacement activity.

ASME Section XI, Code Case N-513-3 (N-513-3), "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping, Section XI, Division 1," paragraph 2(a), requires the full circumference of the pipe to be inspected at the flaw location to characterize the length and depth of all flaws in the pipe section. The NRC has conditionally accepted N-513-3 for use in Regulatory Guide (RG) 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1" (ADAMS Accession No. ML13339A689). This code case provides alternatives to the requirements of the ASME Code, Section XI, IWD-3120(b), which require that components exceeding the acceptance standards of IWD-3400 be subject to supplemental examination or to a repair or replacement activity. The RG 1.147 condition on the code case requires that the repair or replacement activity temporarily deferred shall be performed during the next scheduled outage (i.e., refueling outage 3R 19).

3.4 Reason for Request On November 3, 2016, the licensee detected a pinhole through-wall leak in the 3-inch diameter 31 fan cooler unit service water return line located inside the vapor containment building. The leak is located on weld 8297, line number 128. The licensee initially identified the leak based on an increase in the sump pump rate, indicating a leak rate of about 0.16 gallons per minute (gpm).

The licensee ultrasonically inspected approximately 70 percent of the circumference of the weld.

However, the licensee could not inspect the remaining 30 percent of the circumference of the weld (approximately 3 inches) because of space constraints between the weld and the adjacent fan cooler unit plenum wall. Field measurements indicate the distance between the two components is 0.2 inches at the narrowest point. Therefore, the licensee was not able to satisfy the requirement of inspecting the full pipe circumference in accordance with paragraph 2(b) of N-513-3 due to interference.

The normal service water operating pressure is 90 psig, but it will be lower at the location of the leak because the fan cooler unit is located in one of the highest points of the system, and the downhill discharge causes a pressure reduction of approximately 7 to 10 psig. The operating temperature will be based on the river water temperature whose design value is 95 °F. The licensee noted that the river is currently below 70 °F and will remain lower during the period until the next refueling outage. There is some small increase in the outlet temperature due to containment cooling.

The licensee stated that the service water piping to the 31 fan cooler unit must be isolated when allowable leakage is exceeded. The Technical Specification (TS) 5.5.15, "Containment Leakage Rate Testing Program," allowable leakage is s 0.36 gpm in order to assure containment leakage is limited to an acceptable level. The maximum leakage to assure that the post-accident containment leakage remains within allowable limits is 0.023 gpm. This limit is based upon an evaluation to calculate the amount of service water that can leak through this pinhole under normal system operating conditions to ensure that the containment leakage limits of 10 CFR Part 50, Appendix J, are not exceeded under any mode of operation, including accident conditions. The licensee noted that the leak does not impinge upon any safety-related equipment. As a result, no damage from direct leakage is expected to occur. The licensee stated that approval of the relief request will allow it to reestablish service water to the fan cooler unit to verify that leakage limits are met using a clamp over the pinhole leak. The licensee further stated that it will inspect the leakage daily in accordance with Code Case N-513-3.

The licensee explained that the basis for hardship is the requirements of TS 3.6.6, "Containment Spray System and Containment Fan Cooler System," in which Action C states that an inoperable containment fan cooler unit must be restored to operable status within 7 days or the plant must be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The licensee recognized that relief is required in order to apply N-513-3 to evaluate the leak for operability. The licensee's evaluation demonstrates that there is no decrease in quality and safety because the acceptance criteria of the code case are met. The licensee contended that an online repair was not considered practical because of the amount of time required (preparation of a modification package and partial disassembly of the fan cooler unit to allow welding) and the potential for

excessive sump filtration loading due to the amount of equipment involved in the ASME Code repair, both of which would increase the risk level.

3.5 Proposed Alternative and Basis for Use In lieu of repairing or replacing the subject degraded pipe in accordance with the ASME Code, Section XI, the licensee proposed to use N-513-3, meeting all the requirements of the code case and the RG 1.147 condition on the code case, except for the requirement to inspect the full pipe circumference at the flaw location as required by paragraph 2(a) of the code case. As an alternative to N-513-3, the licensee proposes to inspect only the accessible portion of the pipe circumference at the flaw location.

The licensee measured the pipe wall thickness immediately around the pinhole using ultrasonic testing (UT). Based on the UT data, the licensee characterized the flaw at the leak location as a single flaw with respect to the proximity of other thinned regions. In addition, the licensee noted that the degradation could be characterized as a nonplanar flaw with a through-wall pinhole leak, rather than a leak originating from general area thinning, meaning that the flaw will not propagate from its location because wall thinning is not a contributor to the flaw. UT readings at the site of the leak and surrounding areas indicate acceptable wall thicknesses.

Basis for Use The licensee evaluated the pipe adjacent to weld B297 for structural integrity for general wall thinning. The licensee determined that the minimum required thickness is 0.073 inches, which is below the minimum measured thickness of 0.117 inches. The licensee stated that any reading below 0.073 inches would be considered to be a through-wall flaw. The licensee calculated that the maximum allowable axial flaw size (in length) is 4.11 inches, and the maximum allowable circumferential flaw size is 3.65 inches. In its calculation, the licensee characterized the existing flaw to be contained within an approximate area of 0.5 inches by 0.5 inches, which is bound by the axial and circumferential limits. The licensee noted that this is a conservative assumption because the actual flaw is a small pinhole, and the lowest reading on the pipe wall thickness was obtained when the UT probe was placed at the leak and recorded the 0.117 inch reading. The licensee stated that approximately 3 inches of the pipe circumference is inaccessible for UT measurements. The licensee further stated that while it could be inferred that this portion is sound based on the consistency of UT results, consideration of this approximate 3-inch portion of the pipe as a flaw would still be within the allowable limits. The licensee indicated that the pinhole leak is opposite the inaccessible area, so the flaw areas would be independent and not additive in size.

Following the extent of condition inspection requirement of paragraph 5(a) of N-513-3, the licensee examined five weld locations: weld numbers B298, B299, B300, B301, B302, all located on the same service water return line. These additional inspections confirm the integrity of the piping where all UT data measurements are above 87.5 percent of the pipe nominal wall thickness.

The licensee stated that it will meet the remaining N-513-3 requirements as follows:

(a) Perform a permanent repair or replacement in accordance with the ASME Code, Section XI, IWA-4000, prior to the end of the next scheduled outage.

(b) Conduct daily walkdowns to confirm that the analysis conditions used in the evaluation remain valid.

(c) Conduct periodic ultrasonic inspections at no more than 90-day intervals to verify the flaw growth analysis predictions.

The licensee noted that the weld location qualifies as a straight pipe because it connects the straight pipe to the elbow.

3.6 Duration of Proposed Alternative The licensee requested relief for the fourth 10-year interval and second period until shutdown for refueling outage 3R 19 (March 2017) or until the flaw size exceeds the acceptance criteria.

The licensee stated that if the flaw size exceeds the maximum allowable axial flaw size of 4.11 inches, the maximum allowable circumferential flaw size of 3.65 inches, or the flaw grows into the elbow beyond the applicability limits provided in paragraph 1(c) of N-513-3, it will repair the defect or request additional relief.

3. 7 NRC Staff Evaluation The NRC staff evaluated the proposed alternative based on the provisions of N-513-3, which are divided into the following areas of interest: scope, procedure, flaw evaluation, acceptance criteria, and augment examination.

Scope The NRC staff finds that the affected weld satisfies the requirement of paragraph 1(a) of N-513-3 because the weld location qualifies as a straight pipe as it connects the straight pipe to an elbow.

The NRC staff finds that the criteria of paragraph 1(b) of N-513-3 are applicable to the affected weld because it is an ASME Class 3 component and the normal service water operating pressure of 90 psig and the operating temperature of 95 °F are within the criteria for moderate energy piping.

Paragraph 1(c) of N-513-3 is related to flaw evaluation involving fittings and flanges, which are not applicable to the subject relief request.

With regard to paragraph 1(d) of N-513-3, the TS 5.5.15 allowable leakage is~ 0.36 gpm, and the maximum leakage to assure that the post-accident containment leakage remains within allowable limits is 0.023 gpm. Any leakage that exceeds these TS limits requires the licensee to enter the limiting condition of operation and take corrective actions. In addition, the licensee

noted that the leak does not impinge upon any safety-related equipment. As a result, no damage from direct leakage is expected to occur. As stated above, the licensee will install a clamp on the affected weld to limit the leakage. Based on the above information, the NRC staff finds that the licensee has satisfactorily addressed the leakage requirement of paragraph 1(d) of N-513-3.

The NRC staff finds that the affected weld satisfies the requirement of paragraph 1(e) of N-513-3 because the licensee stated that it will perform a permanent repair or replacement of the subject pipe in accordance with the ASME Code, Section XI, IWA-4000, before the end of the next scheduled outage.

Procedure Paragraphs 2(a), (b), and (c) of N-513-3 require flaw characterization using volumetric inspection methods or physical measurement. The NRC staff finds that the licensee has performed extensive UT of the degraded area of the weld and has characterized the flaw in accordance with paragraphs 2(b) and 2(c). However, the licensee was not able to satisfy paragraph 2(a), which requires that the full pipe circumference at the flaw location be inspected to characterize the length and depth of all flaws in the degraded pipe area. The staff recognizes that a limited region of the degraded pipe area is inaccessible for inspection. The staff finds that the licensee has used flaw evaluation to demonstrate that even if the inaccessible area is degraded and the licensee was not able to satisfy the requirement of paragraph 2(a), the pipe would still perform its intended function until the next refueling outage.

The NRC staff finds that the licensee has satisfied the requirement of paragraph 2(d) of N-513-3 because the licensee has performed an acceptable flaw evaluation as discussed further in this safety evaluation.

The NRC staff finds that the licensee has satisfied the requirement of paragraphs 2(e) and 2(f) of N-513-3 because the licensee will conduct daily walkdowns to confirm that the analysis conditions used in the evaluation remain valid. The licensee will also conduct periodic ultrasonic inspections at no more than 90-day intervals to verify the flaw growth analysis predictions.

The NRC staff finds that that the licensee has satisfied the requirements of paragraphs 2(g) and 2 (h) of N-513-3 because the licensee stated that if the flaw exceeds the maximum allowable axial flaw size of 4.11 inches, the maximum allowable circumferential flaw size of 3.65 inches, or the flaw grows into the elbow beyond the applicability limits provided in paragraph 1(c) of ASME Code Case N-513-3, it will repair the defect or request additional relief.

Flaw Evaluation The NRC staff finds that the licensee satisfactorily followed the methodology in Section 3 of N-513-3 to derive the allowable flaw size in the circumferential (3.65 inches) and axial (4.11 inches) direction. The detected flaw is a pinhole leak, but the licensee conservatively assumed the flaw size is 0.5 inch by 0.5 inch in the circumferential and axial direction. With this conservative assumption, the existing flaw is still within the allowable size.

The NRC staff notes that the minimum required thickness is 0.073 inches. The minimum measured thickness was found to be 0.117 inches. This demonstrates that the degraded area of the pipe has a margin in terms of wall thickness because measured wall thickness is greater than the minimum required wall thickness. The staff notes that the licensee used the wear rate from the plant-specific operating experience such as from the flow-accelerated corrosion program to demonstrate that the flaw will still be within the allowable size until the next refueling outage. The staff further notes that the licensee will periodically inspect the degraded area in accordance with N-513-3 to monitor the corrosion of the subject pipe and verify the flaw growth rate used in its flaw evaluation.

Based on the licensee's flaw evaluation, the NRC staff finds that the licensee has satisfied Section 3 of N-513-3.

Acceptance Criteria The NRC staff finds that the licensee used the structural factors in the ASME Code, Section XI, Appendix C, C-2621 and C-2622, in calculating the allowable flaw size as required by Section 3 of N-513-3. The staff determined that the licensee has satisfied the requirement of the Section 4, "Acceptance Criteria," of N-513-3.

Augment Examination The NRC staff finds that the licensee has satisfied the requirement of Section 5, "Augment Examination," of N-513-3 because the licensee has inspected five weld locations that are located on the same service water return line. These five welds have satisfactory wall thickness.

Hardship Justification The basis for hardship is the requirement of TS 3.6.6, Action C, which states that an inoperable fan cooler unit must be restored to operable status within 7 days or the plant must be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. As previously discussed, approximately 30 percent of the circumference of the degraded weld is inaccessible for inspection due to interferences, and the licensee was not able to satisfy the requirement of inspecting the full pipe circumference in accordance with paragraph 2(b) of N-513-3. Performing an online repair in accordance with the ASME Code requirements would require physical modifications to the piping. Therefore, the NRC staff concludes that performing an ASME Code repair would result in hardship and unusual difficulty, without a compensating increase in the level of quality and safety.

4.0 CONCLUSION

The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the subject piping and that complying with the ASME Code requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Accordingly, the staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the staff authorizes the use of Relief Request No. IP3-ISl-RR-10 for the remainder of the fourth 10-year

interval until the 3R 19 refueling outage, which is scheduled for March 2017 or until the flaw size exceeds the acceptance criteria at Indian Point 3.

All other requirements of the ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including third party review by the Authorized Nuclear lnservice Inspector.

Principal Contributor: J. Tsao Date: January 11, 2017

If you have any questions, please contact Douglas V. Pickett, at 301-415-1364 or by e-mail at Douglas.Pickett@nrc.gov.

Sincerely, IRA/

Stephen S. Koenick, Acting Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:

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NAME DPickett LRonewicz DAiiey SKoenick DATE 1/04/2017 12/27/2016 12/19/2016 1/11/2017 OFFICIAL RECORD COPY