ML12296A108

Relief Request lSl-07 and lSl-08 Proposed Alternative Reactor Vessel Head Penetration Nozzle Examinations for Fifth Interval of Inservice Inspection Program

ML12296A108
Person / Time
Site:	Ginna
Issue date:	10/23/2012
From:	George Wilson Plant Licensing Branch 1
To:	Joseph Pacher Ginna
Thadani M
References
TAC ME8800, TAC ME8801
Download: ML12296A108 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 23, 2012 Mr. Joseph E. Pacher Vice President R. E. Ginna Nuclear Power Plant R. E. Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519

SUBJECT:

R. E. GINNA NUCLEAR POWER PLANT - RE: RELIEF REQUESTS ISI-07 AND ISI-08 PROPOSED ALTERNATIVE REACTOR VESSEL HEAD PENETRATION NOZZLE EXAMINATIONS FOR FIFTH INTERVAL OF INSERVICE INSPECTION PROGRAM (TAC NOS. ME8800 AND ME8801)

Dear Mr. Pacher:

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated May 24,2012 as supplement by letter dated July 27,2012, R. E Ginna Nuclear Power Plant, LLC, licensee for R. E. Ginna Nuclear Power Plant (Ginna), submitted relief requests Nos. ISI-07 and ISI-08 for relief from certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(6)(ii)(D) for the fifth ten-year inservice inspection interval (lSI) of Ginna.

In relief request ISI-07, the licensee requests authorization of a proposed alternative from the inspection requirements of the American Society of lVIechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for In-service Inspection of Nuclear Power Plant Components," Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Pressure Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division I, "subject to the conditions specified in paragraph 10 CFR 50.55a(g)(ii)(D)(2) throl1gh(6), for the examination of reactor pressure vessel upper head penetrations for the control rod drive mechanisms (CRDMs). Specifically, pursuant to 10 CFR 50.55a(a)(3)(0),as stated in the licensee's July 27, 2012 response to the NRC staff's request for additional information, the licensee requested to use the alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty.

In relief request ISI-08, the licensee requests authorization to the proposed alternative to the ASME Code Case N-729-1 condition specified in paragraph 10 CFR 50.55a(g)(6)(ii)(D)(3),for performing a volumetric or surface leak path assessment through the J-groove welds. The reason for the licensee's request is a design feature of the replacement reactor vessel head that incorporates a weep change in each CRDM nozzle bore through the interface zone. The weep change is modified axially into the alloy steel through the interference zone between the CRDM nozzle to low alloy steel reactor vessel head bore annulus. This feature is unique to only a few replacement reactor pressure vessel heads in the U.S.

The NRC staff has reviewed the licensee's requests as outlined in the safety evaluation for request ISI-07 in Enclosure 1, and for relief request ISI-08 in Enclosure 2, and concludes that

J. Pacher -2 the proposed alternatives provide reasonable assurance of structural integrity of the specified components.

Based on the information provided for relief request ISI-07, the NRC staff concludes that the licensee's proposed alternative demonstrates that compliance with the specified requirements of 10 CFR 50.55a(g)(6)(ii)(D) would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes the licensee's proposed alternative to define an alternate examination zone below the J-groove weld at Ginna for the fifth ten-year inservice inspection interval, scheduled to end in December 2019.

Similarly, based on the information provided by the licensee for relief request ISI-08, the NRC staff concludes that the licensee has adequately addressed all of the Code of Federal Regulations requirements set forth in 10 CFR 50.55a(a)(3)(i), and the proposed alternative is in compliance with those requirements. Therefore, in accordance with 10 CFR 50.55a(a)(3)(i) the NRC staff authorizes the licensee's proposed alternative, ISI-08, at Ginna for the remainder of the fifth ten-year inservice inspection interval, scheduled to end in December 2019.

All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

If you have any questions, please contact the Ginna Project Manager, Mohan Thadani, at (301) 415-1476.

Sincerely, George Wilson, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

1) Safety Evaluation for ISI-07

2) Safety Evaluation for ISI-08 cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR RELIEF REQUEST NO. ISI-07 FOR RELIEF FROM CODE CASE INSPECTION REQUIREMENTS FOR EXAMNATION OF REACTOR VESSEL HEAD PENETRATION NOZZLES HAVING PRESSURE-RETAINING PARTIAL PENETRATION WELDS R. E. GINNA NUCLEAR POWER PLANT, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. SO-244

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated May 24, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 121S1A40S) as supplement by letter dated July 27,2012 (ADAMS Accession No. ML 1221SA318), R. E Ginna Nuclear Power Plant, LLC (the licensee) submitted a relief request No. ISI-07 from certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) SO.55a(g)(6)(ii)(D) for the fifth ten-year inservice inspection (lSI) interval of R. E.

Ginna Nuclear Power Plant (Ginna).

The licensee requests authorization of a proposed alternative from the inspection requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section XI, "Rules for In-service Inspection of Nuclear Power Plant Components," Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Pressure Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division I,"

for the examination of reactor pressure vessel (RPV) upper head penetrations for the control rod drive mechanisms (CRDMs). Specifically, pursuant to 10 CFR 50.SSa(a)(3)(ii), as stated in the licensee's July 27,2012 response to the NRC staff's request for additional information (RAI), the licensee requested to use an alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty.

2.0 REGULATORY EVALUATION

The ASME Code,Section XI, states that lSI of Class 1, 2 and 3 components shall be performed in accordance with the requirements of Section XI of the ASME Code and applicable editions and addenda as required by 10 CFR 50.5Sa(g), except where specific written relief has been granted by the Commission. Pursuant to 10 CFR 50.55a(g)(4), throughout the service-life of a pressurized water reactor (PWR), components which are classified as ASME Code Class 1, 2, and 3 must meet the requirements, except design and access provisions and pre-service Enclosure 1

-2 examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry and materials of construction of the components.

Further, regulations under 10 CFR 50.55a(g)(4)(i) require that 151 of components and system pressure tests conducted during the first ten-year 151 interval and subsequent intervals shall comply with the requirements in the latest edition and addenda of the ASME Code,Section XI, incorporated by reference in paragraph (b) of 10 CFR 50.55a on the date 12 months prior to the start of the 120-month 151 interval subject to the limitations and modifications listed therein. The Section XI, ASME Code of Record for the fifth ten-year 151 interval at Ginna is the 2004 Edition with no addenda.

In 10 CFR 50.55a(g)(6)(ii), it states that the Commission may require the licensee to follow an augmented 151 program for systems and components for which the Commission deems that added assurance of structural reliability is necessary. Regulations under 10 CFR 50.55a(g)(6)(ii)(O) require augmented 151 of RPV head penetration nozzles of PWRs in accordance with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (2) through (6) of 10 CFR 50.55a(g)(6)(ii)(O).

Pursuant to 10 CFR 50.55a(a)(3), proposed alternatives to the requirements of 10 CFR 50.55a(g) may be used when authorized by the Commission if: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

3.0 TECHNICAL EVALUATION

3.1 Affected Systems and Components The CROM penetration nozzles deSignated as Item No. 84.40, "Nozzles and Partial-Penetration Welds of PWSCC [primary water stress corrosion cracking]-Resistant Materials in Head," in Table 1 of Code Case N-729-1, for Ginna are as follows:

Penetration nozzle Nos. 1,6 to 35, and 37.

3.2 Applicable Code Requirement (as stated by the licensee) 10 CFR 50.55a(g)(6)(ii)(O)(1) requires that examinations of the reactor vessel head be performed in accordance with ASME Code Case N-729-1 subject to the conditions specified in paragraphs 10 CFR 50.55a(g)(6)(ii)(O)(2) through (6).

Figure 2 in Code Case N-729-1, as referenced by Paragraph -2500, requires that the volumetric or surface examination coverage distance below the toe of the J-groove weld (dimension "a") be 1.5 inches for incidence angle (8) < 30°; 1 inch for 8> 30°; or to the end of the tube, whichever is less.

3.3 Proposed Alternative (as stated by the licensee)

As an alternative to the volumetric and surface examination coverage requirements shown as dimension "a" in Figure 2 of ASME Code Case N-729-1, Ginna proposes the

-3 use of the achievable examination coverage as the alternative examination coverage for the penetration nozzles listed in Table 1 with the exception of penetration nozzle number 24, which meets the Code Case examination coverage requirements.

Appendix I of ASME Code Case N-729-1 provides the analysis procedure for alternative examination area or volume definition to that specified in Figure 2 of the Code Case if impediments prevent the examination of the complete zone ... the analyses shall be performed using at least the stress analysis method (Paragraph 1-2000) or the deterministic fracture mechanics analysis method (Paragraph 1-3000) to demonstrate that the applicable criteria are satisfied.

Table 1: Ginna CRDM Nozzle Alternative (Achievable) Inspection Coverage Proposed Inspection Proposed Inspection Penetration Coverage Below the J-Groove Penetration Coverage Below the J-Groove No. Weld Toe on the Downhill No. Weld Toe on the Downhill Side (in) Side (in) I 1 1.00 21 0.64 6 1.00 22 0.48 i 7 1.12 23 0.88 8 0.88 24 1.00 9 1.08 25 0.76 10 0.72 26 0.84 11 1.00 27 0.88 12 1.12 28 0.72 13 1.00 29 0.76 14 0.96 30 0.64 15 0.72 31 0.80 16 0.80 32 0.60 17 0.76 33 0.56 18 0.88 34 0.52 19 1.04 35 0.20 20 0.88 37 0.40 3.4 Licensee's Basis for Request (as stated by the licensee)

The minimum distance below the J-groove weld that needs to be examined is determined by the location below the J-groove weld where the steady state hoop stress distribution is less than 20 ksi. Figures 1 through 5 show the minimum required inspection coverage distance below the J-groove weld such that the stresses over the entire region outside the minimum examination coverage zone are below 20 ksi but within the examination zone defined in Figure 2 of the ASME Code Case N-729-1. A comparison of the minimum required examination coverage determined in accordance with the 1-2000 procedure and the minimum achievable examination coverage for each of the analyzed upper head penetration nozzle[s] is shown in Table 3 as well as in Figures 1 through 5.

- 4 With the exception of the outermost penetration nozzle row [34, 35, and 37] with incidence angle of 43.50", the hoop stress remains below 20 ksi for the entire region outside the minimum achievable examination coverage zone below the J-groove weld but within the examination zone defined in Figure 2 of Code Case N-729-1. The achievable examination coverage for penetration nozzle numbers 34, 35 and 37 from Table 1 are 0.52",0.20" and 0.40" respectively. Only the achievable examination coverage for penetration nozzle number 35 is below the Paragraph 1-2000 required minimum examination coverage. Therefore, in accordance with Mandatory Appendix I analysis procedure for alternative examination area or volume definition, the achievable examination coverage for all the penetration nozzles, with the exception of penetration nozzle number 35, has been shown to be acceptable as the alternative examination coverage in accordance with Paragraph 1-2000. The achievable examination coverage at penetration nozzle number 35 was evaluated further using the deterministic fracture mechanics analysis approach outlined in Paragraph 1-3200 of Code Case N-729-1.

The result of the deterministic fracture mechanics analysis has demonstrated that an assumed hypothetical through-wall axial flaw in the unexamined region of penetration nozzle number 35 would take more than 100 effective full power years to reach the toe of the J-groove weld. For penetration nozzle number 35, the adequacy of using the achievable examination coverage tabulated in Table 1 as the alternative examination coverage has been demonstrated and thus allows Ginna Station to continue to operate prior to the hypothetical postulated flaw reaching the toe of the J-groove weld before the next scheduled examination.

3.5 NRC Staffs Evaluation The NRC staff's review of this request was based on 10 CFR 50.55a(a)(3)(;;) which states that:

Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The specific regulatory requirements for which relief is requested are defined in 10 CFR 50.55a(g)(6)(ii)(D)(3), which states in part:

Instead of the specified 'examination method' requirements for volumetric and surface examinations in Note 6 of Table 1 of Code Case N-729-1, the licensee shall perform volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Figure 2 of ASME Code Case N-729-1.

As stated in Section 3.3 of this safety evaluation (SE), the licensee proposed to perform ultrasonic (UT) inspections for the CRDM penetration nozzles in accordance with the UT achievable coverage indicated in Table 1. For penetration nozzles of which the achievable UT coverage is less than the required coverage determined by Paragraph 1-2000, "Stress Analysis,"

a deterministic fracture mechanics analysis is used to justify the alternative examination coverage. Section 3.4 of this SE indicated that, with the exception of Nozzle No.35, which is one of the three outermost penetration nozzles with incidence angle of 43.50", the minimum achievable examination coverage zone below the J-groove weld of each nozzle bounds the examination coverage defined by the 20 ksi hoop stress limit per Paragraph 1-2000 of Code

-S-Case N-729-1. The hoop stresses are due to the residual stress resulting from the J-groove weld fabrication process and normal operating temperature and pressure condition loads. For the three outermost nozzles, the stress of 20 ksi defines the minimum required examination coverage of 0.33 inch for them, which Nozzle No. 3S can only cover 0.20 inch. This is the basis for the licensee to claim that all penetration nozzles, with the exception of Nozzle No. 3S, are acceptable to have the alternative examination coverage in accordance with Paragraph 1-2000.

Since the stress analysis in ASME Code Case N-729-1 calls for consideration of both axial and hoop stresses, the NRC staff issued RAI-1 requesting information on axial stresses. The licensee's response to RAI-1 dated July 27,2012, provides stress plots for nozzles of all five incident angles, demonstrating that hoop stresses are bounding. Therefore RAI-1 is resolved.

For Nozzle No. 3S the licensee used a deterministic fracture mechanics analysis to justify the alternative examination coverage. The NRC staff is not certain that the proposed coverage of 0.2 inch is sufficient to tolerate the uncertainties in the calculated residual stresses and the determination of the final crack length based on the proposed crack growth rate for Alloy 690 CRDM material. Hence, the NRC staff issued RAI-2 requesting the licensee provide the basis for not performing a surface examination of the lower portion of each penetration nozzle necessary to meet the inspection requirements of 10 CFR SO.SSa(g}(6}(ii)(D}. The licensee's July 27,2012, response to RAI-2 states, 'The Reactor Pressure Vessel Head outside surface NDE [non-destructive evaluation] Penetrant Testing (PT) will expose workers to between SOO 700 mRem of dose per penetration. There also exists the potential for additional dose for additional preparation, cleaning or repeat PT exams in resolution of geometric indications due to material surface conditions that potentially could double personnel exposure. The Ginna Station replacement Reactor Vessel Head was designed and fabricated to the ASME Section III 1995 Edition, 1996 Addenda and to be compatible with the original vessel design which did not identify any requirements to extend nozzle material configurations to achieve Code Case N-721[9]-1 specified coverage above and below the J welds. To obtain the current material volume requirements, the Reactor Vessel Head would need to be redesigned/modified to accommodate an increased inspection zone below the J-groove weld. 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 licensee's response to RAI-2, the NRC staff determined that the licensee has shown a physical and radiological hardship which would be incurred in order to be within compliance with the specified requirements. The NRC staff finds that a physical hardship exists due to the inability of UT inspection to effectively scan the ASME Code Case N-729-1 required examination coverage of each CRDM penetration. While dye penetrant inspection would be a viable option for the licensee, the inspection would require manual application in a high radiation area. Furthermore, additional setup work would require additional accumulation of dose for each nozzle. Therefore, the NRC staff finds that to perform the additional inspection would be a significant radiological hardship for the limited additional inspection coverage, and the licensee's revised relief request basis via 10 CFR SO.SSa(a)(3)(ii), instead of 10 CFR SO.SSa{a){3){i), is acceptable. RAI-2 is resolved.

In the licensee's fracture mechanics analysis, PWSCC was identified as the major degradation mechanism for CRDM nozzle penetrations. PWSCC typically initiates in the areas of the highest tensile stress in susceptible materials, such as alloy 600 materials, and propagates in response to time, environment (i.e., temperature) and stress intensity. In the current application, the licensee used a crack growth rate with an improvement factor of 100 over Alloy 600 for Alloy 690

- 6 materials in its fracture mechanics analyses. The NRC staff found that the test results reported in NUREG/CR-71 03, Volume 2, "Pacific Northwest National Laboratory Investigation of Stress Corrosion Cracking in Nickel-Base Alloys," revealed that Alloy 690 with 30 percent cold work has a similar crack growth rate as Alloy 600. To ensure that the Alloy 690 material for the Ginna CRDM penetrations did not receive significant cold work, so that using the improvement factor is justified, RAI-3 requests the licensee provide an estimate of the cold work for the Ginna CRDM penetrations based on the fabrication and installation record to support use of the much reduced crack growth rate for Alloy 690.

The July 27,2012, response to RAI-3 states that, "The CRDM tube material received no cold rolling operations. With regard to any cold work surface layers produced by final machining processes, Babcock and Wilcox performed an empirical and analytical evaluation where X-ray diffraction was used to determine the cold work layer depth that resulted from inside diameter and outside diameter machining and grinding processes. These dimensions were used as a basis for establishing electro-polishing parameters to result in removal of the entire cold worked layer that resulted from machining and grinding on the wetted portions of the nozzle and J-groove weld."

Regarding the estimate of the cold work, it further states, "This calculation shows that the R.E.

Ginna CRDM nozzles originally had 2% [percent] of their cross sectional area in the cold-worked condition but by electro-polishing each CRDM nozzle, all the cold work induced by machining was effectively removed prior to nozzle installation. Therefore, prior to welding the CRDM nozzles were considered to have very small amount << 2%) of cold work present."

Since the CRDM tube material received no cold rolling operations and the estimated cold work of less than 2 percent was obtained through an empirical and analytical evaluation of cold work surface layers produced by final machining processes, the estimate is credible. The estimated cold work of less than 2 percent for the Ginna CRDM penetration nozzles is far less than the 30 percent cold work which would make the improvement factor of the PWSCC rate for Alloy 690 vanish. Hence, the NRC staff determined that a PWSCC rate accounting for minimum cold work in the Alloy 690 CRDM tube could be used in the proposed fracture mechanics analysis.

RAI-3 is resolved. However, instead of the proposed improvement factor of 100, the staff determined that a factor of 50 is more appropriate considering the recent information presented in the August 2012, ASME Code Meeting, "scc Growth Rates for Alloy 690 and Associated Welds."

The licensee's fracture mechanics analysis showed that a conservative through-wall axial flaw located in the uninspected region of the nozzle would not grow to the toe of the J-groove weld in 124 effective full power years (EFPYs). The NRC staffs assessment of the licensee's fracture mechanics analysis conclusions is based on the stress analysis results and the proposed PWSCC rate for Alloy 690 with negligible cold work as discussed above. The calculated EFPYs are large and can tolerate an improvement factor of 50. Therefore, the licensee's stress and fracture mechanics analysis demonstrated that compliance with the specified requirements stated in the beginning of this section would not result in a compensating increase in the level of quality and safety.

Since the May 24, 2012, submittal did not specify the examination frequency, RAI-4 requests the licensee provide EFPYs operated so far for the Ginna CRDM penetration nozzles and specify the time for the next examination. The July 27,2012, response to RAI-4 states that, "In

-7 compliance with code case N-729-1 B4.40 examination frequency for PWSCC resistant materials, Ginna is performing the first in-service head inspection during the fall 2012 refueling outage. Subsequent examinations to the fall of 2012 refueling outage will also follow the examination frequency of ASME code case N-729-1. Ginna will have 8.52 [EFPYs] of operation on the replaced reactor head at the time of the fall 2012 examination." This response provides the information requested by the NRC staff regarding EFPYs operated to date for the Ginna CRDM penetration nozzles and the time for the next examination. Therefore, RAI-4 is resolved.

Given the above, the NRC staff finds sufficient technical basis, such that the re-inspection frequencies above will provide reasonable assurance of structural integrity of each nozzle, with the reduced area of inspection coverage, as defined in Table 1.

Regulations under 10 CFR 55a(g}(6)(ii)(D} calls for ASME Code Case N-729-1 to address reactor vessel head inspections. The NRC staff's evaluation confirmed that the licensee's proposed alternative inspection meets the ASME Code Case N-729-1 analysis procedure for alternative examination of CRDM penetration nozzles at Ginna. The licensee has noted that while surface examination could be performed to increase the inspection coverage for the nozzle, these additional inspections would be of limited value and require extensive work in very high radiation fields. The NRC staff finds that performing these additional surface examinations would result in hardship through significant radiation exposure without a compensating increase in the level of quality or safety.

4.0 CONCLUSION

As set forth above, the NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the subject components. Based on the information provided in the licensee's submittals and given the maximum time between volumetric re-inspection in accordance with the requirements of Item No. B4.40 of Table 1 of ASME Code Case N-729-1 for Ginna, the NRC staff has determined that the licensee's proposed alternative demonstrates that compliance with the specified requirements of 10 CFR 50.55a(g}(6}(ii)(D) would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Therefore, pursuant to 10 CFR 50.55a(a}(3)(ii), the use of the proposed alternative to define an alternate examination zone below the J-groove weld is authorized for Ginna for the fifth ten-year lSI interval.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Principal Contributors: S. Sheng C. Sydnor Date: October 23,2012

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR RELIEF REQUEST NO. ISI-08 FOR RELIEF FROM ASME CODE CASE INSPECTION REQUIREMENTF FOR DEMONSTRATED VOLUMETRIC LEAK PATH EXAMINATION OF REACTOR PRESSURE VESSEL UPPER HEAD PENETRATION NOZZLES R. E. GINNA NUCLEAR POWER PLANT, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

1.0 INTRODUCTION

By letter dated May 24, 2012, (Agencywide Document Access and Management System (ADAMS) accession number ML 121S1A40S), enclosure 2, and supplemented by letter dated October 5, 2012 (ADAMS accession number ML 1228SA3S9). R. E. Ginna Nuclear Power Plant LLC, the licensee, requested the U. S, Nuclear Regulatory Commission (NRC, Commission) for authorization of an alternative to Title 10 of the Code of Federal Regulations (10 CFR)

SO.SSa{g){6){ii){D){3) for R. E. Ginna Nuclear Power Plant (Ginna). The alternative would allow the licensee to perform a volumetric leak path inspection of each of the reactor pressure vessel (RPV) upper head penetrations without a specific demonstration to include the effects of a grooved channel in the low alloy steel upper head. The licensee installed the channel in each penetration to enhance the capability of a bare metal visual examination to identify leakage from the top of the head.

2.0 REGULATORY EVALUATION

The inservice inspection (lSI) of American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components is to be performed in accordance with Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," of the AS ME Code and applicable editions and addenda as required by Title 10, Code of Federal Regulations (CFR) , except where specific written relief has been granted by the Commission.

Regulations under 10 CFR SO.SSa{g)(6){ii) states that the Commission may require the licensee to follow an augmented lSI program for systems and components for which the Commission deems that added assurance of structural reliability is necessary. Regulations under 10 CFR SO.SSa(g){6)(ii)(D) requires, in part, augmented inservice inspection of RPV upper heads, associated penetration nozzles and J-goove welds of pressurized water reactors in accordance with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (2) through (6) of 10 CFR SO.SSa(g)(6)(ii)(D).

Enclosure 2

- 2 Alternatives to requirements under 10 CFR 50.55a(g) may be authorized by NRC pursuant to 10 CFR 50.S5a(a)(3)(i) or 10 CFR SO.55a(a)(3)(ii). In proposing alternatives or requests for relief, the licensee must demonstrate 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.

By letter dated May 24, 2012, the licensee proposed an alternative, RR-ISI-OB, in accordance with 10 CFR 50.5Sa(a)(3)(i) for relief from the requirements of 10 CFR 50.55a(g)(6)(ii)(0)(3) for all RPV upper head penetration nozzles. 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 Commission to authorize the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 Licensee Relief Request 3.1.1 Component Identification ASME Code Case N-729-1 Item Number B4.40 penetration nozzles and associated partial penetration J-groove welds in the RPV upper head at Ginna.

3.1.2 Code Requirements for Which Relief is Requested Regulations under 10 CFR 50.55a(g)(6)(ii)(0)(3) requires, in part, that a demonstrated volumetric leak path assessment be performed for each RPV upper head penetration.

3.1.3 Licensee's Proposed Alternative The licensee proposes to perform a volumetric examination for leak path assessment augmented by a bare metal visual examination in accordance with ASME Code Case N-729-1.

3.1.4 Licensee's Ouration of Relief Request The licensee requests relief from 10 CFR 50.55a(g)(6)(ii)(0)(3) for the remainder of the fifth ten-year inservice inspection interval which is scheduled to be complete by Oecember 2019.

3.1.5 Licensee's Basis for Relief The licensee submitted this relief request due to a design feature of the new replacement RPV upper head at Ginna. The design incorporates a weep channel in each control rod drive mechanism (CROM) nozzle bore through the interference fit region. The axial weep channel is a small portion of the CROM interference fit region, less than O.S percent of the total area. The weep channel was included to ensure leakage from the annulus region would be able to go through the interference fit region. The licensee believes this design will improve the ability of a bare metal visual examination to detect even small amounts of leakage into the annulus region.

In the licensee's May 25,2012 letter, the licensee initially stated that performance of the bare metal visual examination alone would be equivalent to a demonstrated volumetric or surface

-3 leak path assessment, due to the weep channel. In response to an NRC request for additional information, the licensee, by letter dated October 5,2012, restated its proposed alternative to include a volumetric leak path assessment for each penetration as well as the bare metal visual examination. However, the licensee stated that it was not able to meet the full requirements of 10 CFR 50.55a(g)(6)(ii)(0)(3) as it was not able to demonstrate the volumetric leak path assessment on a penetration with an axial weep channel as used at Ginna.

While the licensee stated that it would not be able to demonstrate the volumetric leak path assessment for the unique configuration at Ginna, it did note that the same technique has been demonstrated on other reactor head designs successfully. The licensee notes that industry experience has shown that visual examination in conjunction with ultrasonic leak path examination is optimum for detecting reactor coolant leakage from the RPV upper head penetration nozzles and associated J-groove welds. Further the licensee explained that the proposed alternative would allow identification of leakage through scenarios 1,2, and 3 of the recently published NRC NUREG/CR-7142, "Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation," (ADAMS accession number ML12241A160). The licensee states that delectability using the proposed alternative would be consistent with other head designs as stated in a Westinghouse technical justification document, WOI-T J-006-03-P, Revision 4, "Ultrasonic Testing of Interference Fit Samples for Leak Path Detection (Westinghouse Proprietary)."

The licensee also states that it has performed baseline examinations of the interference fit regions using an ultrasonic leak path assessment technique. This baseline will be used to compare to the inservice inspection data, and it will provide an excellent comparison of ultrasonic response to facilitate the examination.

3.2 NRC STAFF'S EVALUATION The staff reviewed the licensee's proposed alternative under the requirements of 10 CFR 50.55a(a)(3)(i), such that; "The proposed alternative would provide an acceptable level of quality and safety."

The current requirements of 10 CFR 50.55a(g)(6)Oi)(O)(3) would require a demonstrated volumetric leak path assessment for each RPV upper head penetration during the scheduled fall 2012 refueling outage at Ginna. This inspection is required once every ten years, so this examination would be the only volumetric examination of the RPV upper head at Ginna in its fifth ten-year inservice inspection interval.

The volumetric leak path assessment uses an ultrasonic probe running along the inside diameter of the penetration nozzle to volumetrically look for indications of leakage between the penetration nozzle outside diameter and the low alloy steel wall of the head penetration. The indications of leakage can be a leak path signal, showing where leakage is flowing through the area, deposits, either dried boric acid or corrosion products, or degradation of the low alloy steel wall of the penetration itself. There are three general areas along the axial length of the outside diameter of the penetration nozzle and the wall of the head penetration. Going from the triple point, where the J-groove weld, low alloy steel head and the penetration nozzle are joined, to the

- 4 top of the RPV head, these areas are the annulus region, the interference fit region and the counterbore region. Indications of leakage, a leak path; deposits; or corrosion of the low alloy steel head, can be found in each of these regions but are most noticeable in the annulus and interference fit regions. In a cold condition, the interference fit region provides metal to metal contact and therefore good sound transition for the ultrasonic signal. If any deposits or corrosion of the low alloy steel wall are within this region, the ultrasonic signal will change, indicating a potential leak path. The lower annulus region is limited in area but has a general air gap. If filled with deposits, the ultrasonic signal differs from a general air gap signal, indicating a potential leak.

The licensee's proposed alternative will be to conduct the same examinations, volumetric leak path assessment and bare metal visual examinations, as performed at all other US pressurized water reactors. The only difference between the two exams is in the design of the Ginna RPV upper head. It has axial weep channels machined into the low alloy steel CRDM interference fit region of each penetration. These weep channels, however, are very small, less than 0.5 percent of the circumferential area of the interference fit region.

The staff reviewed the technical basis and supporting research for the volumetric leak path assessment to determine the impact of the axial weep channels at Ginna on the effectiveness of the assessment. Due to the small size of the axial weep channels and because during operation, the interference fit does not exist due to the thermal expansion differences between the nozzle material and the low alloy steel material of the head penetration, the staff does not expect significantly more flow through the weep channel than through the interference fit region.

As such, the amount of deposits and potential corrosion material in the annulus region is not expected to significantly change. The staff bases this conclusion, in part, on reviews of the experimental testing conducted in Task 4 of the EPRI Boric Acid Corrosion Program. A summary of the program's findings is available in a slide presentation made at an NRC public meeting on February 29,2012, entitled, "Boric Acid Corrosion Testing Program Overview."

(ADAMS accession number ML120690174) Of note, slides 14 through 20 and 27 through 29 show experimental examples of the amount of deposits in an annular region during simulated leakage of primary coolant. In addition test 12a, on slide 20, shows the potential effect of an axial weep channel in the interference fit region. In this test a significant axial channel was machined into the test block representing the low alloy steel head. Deposits and degradation of wall material around the edges of the axial channel were still clearly evident, even with this much larger channel effect than the one machined into the Ginna RPV upper head penetrations.

Therefore the staff finds minimal impact on the accumUlation of boric acid or corrosion products, or the evidence of a leak path in the annulus and interference fit regions of a penetration at Ginna due to the axial weep channels.

Given that the indications of leakage would remain within the inspection area of the volumetric leak path assessment, the staff reviewed the potential for any limitation in the ability for the technique to volumetrically detect these indications of leakage. The licensee's proposed*

alternative will use a volumetric leak path technique which has been previously demonstrated on penetration nozzles of a similar size, with similar configurations, as those found at Ginna, with the single exception of the axial weep channel. NRC NUREG/CR-7142, provides a strong basis for even small boric acid or corrosion product deposits to be detected using a volumetric leak path assessment. Further, as the licensee has a baseline examination to compare results of inservice examinations, even small changes in the data output can be identified and analyzed to ensure effective implementation of the technique. Given that boric acid and corrosion product

- 5 deposits will remain in the annulus and interference fit regions, with or without an axial weep channel, the staff finds no significant difference in the reliability of a volumetric leak path technique demonstrated on a penetration without an axial weep channel to be used on a RPV upper head penetration with an axial weep channel of the size found at Ginna. As such, the staff finds an equivalent and acceptable level of quality and safety in the licensee performing its volumetric leak path assessment technique on each of the Ginna RPV upper head penetrations.

4.0 CONCLUSION

S As set forth above, the NRC staff concludes that the licensee provided sufficient technical basis to demonstrate that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i), and therefore the alternative is in compliance with the Code of Federal Regulation's requirements. As such, in accordance with 10 CFR 50.55a(a)(3)(i) the NRC staff authorizes the licensee's proposed alternative, RR-ISI-08, as supplemented by letter dated October 5, 2012, at Ginna for the remainder of the fifth ten-year inservice inspection interval, scheduled to end in December 2019.

All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: J. Collins Date: October 23, 2012

J. Pacher -2 the proposed alternatives provide reasonable assurance of structural integrity of the specified components.

Based on the information provided for relief request ISI-07, the NRC staff concludes that the licensee's proposed alternative demonstrates that compliance with the specified requirements of 10 CFR 50.55a(g)(6)(ii)(D) would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes the licensee's proposed alternative to define an alternate examination zone below the J-groove weld at Ginna for the fifth ten-year inservice inspection interval, scheduled to end in December 2019.

Similarly, based on the information provided by the licensee for relief request ISI-08, the NRC staff concludes that the licensee has adequately addressed all of the Code of Federal Regulations requirements set forth in 10 CFR 50.55a(a)(3)(i), and the proposed alternative is in compliance with those requirements. Therefore, in accordance with 10 CFR 50.55a(a)(3)(i) the NRC staff authorizes the licensee's proposed alternative, ISI-08, at Ginna for the remainder of the fifth ten-year inservice inspection interval, scheduled to end in December 2019.

All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.

If you have any questions, please contact the Ginna Project Manager, Mohan Thadani, at (301) 415-1476.

Sincerely, Ira!

George Wilson, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

1) Safety Evaluation for ISI-07

2) Safety Evaluation for ISI-08 cc w/encls: Distribution via Listserv DISTRIBUTION:

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