ML12151A405

10 CFR 50.55a Request ISI-07 and ISI-08: Proposed Alternative Reactor Vessel Closure Head Penetration Nozzle Examinations for the Fifth Interval Inservice Inspection (ISI) Program

ML12151A405
Person / Time
Site:	Ginna
Issue date:	05/24/2012
From:	Mogren T Constellation Energy Group, EDF Group, Ginna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML12151A405 (23)

Text

Thomas Mogren R.E. Ginna Nuclear Power Plant, LLC Manager - Engineering Services 1503 Lake Road Ontario, New York 14519-9364 585.771.5208 585.771.3392 Fax CENG a joint venture of Thomas. Mogren@cengllc.com May 24, 2012 U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 10 CFR 50.55a Request ISI-07 and ISI-08: Proposed Alternative Reactor Vessel Closure Head Penetration Nozzle Examinations for the Fifth Interval Inservice Inspection (ISI) Program Pursuant to 10 CFR 50.55a(a)(3)(i), R.E. Ginna Nuclear Power Plant, LLC requests relief from ASME Code Case N-729-1, Examination Requirements of the Reactor Vessel Replacement Head Penetration Nozzles.

Relief Requests ISI-07 (Enclosure 1) and ISI-08 (Enclosure 2) are being submitted for the fifth 10-year ISI interval to propose alternatives to complying with the code examination requirements specified in 10 CFR 50.55a(g)(6)(ii)(D). The enclosures describe how the proposed alternatives provide an acceptable level of quality and safety. Approval is requested by October 5, 2012 in support of our 2012 Refueling Outage.

There are no new regulatory commitments identified in this correspondence. Should you have any questions regarding this request, please contact Mr. Thomas Harding at (585) 771-5219 or Thomas.HardinpqJr(ccengllc.com.

Very truly yours, Thomas Mogren I )ýý A0(4-7 (AftlLAJeC -- jo~?65~

Document Control Desk May 24, 2012 Page 2 Enclosure (1): Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - "Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i), Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations" Enclosure (2): Relief Request Number ISI-08 R.E. Ginna Nuclear Power Plant - "Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i), "Reactor Vessel Replacement Head Penetration Nozzle Weep Channel Examination" cc: W.M. Dean, NRC M.C. Thadani, NRC Ginna Resident Inspector, NRC

ENCLOSURE 1 Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations

1. ASME CODE COMPONENT(S) AFFECTED:

Code Class: I

Reference:

ASME Code Case N-729-1/10 CFR 50.55a(g)(6)(ii)(D)

Item Number: B4.40

==

Description:==

SB-167 UNS N06690 Nozzles and UNS N06052 Partial-penetration welds of PWSCC-resistant materials in head Drawing: 083NE001 Revision 2

2. APPLICABLE CODE EDITION AND ADDENDA:

The current code of record for the R.E. Ginna Nuclear Power Plant is the ASME Section Xl Code, 2004 Edition with no addenda, as augmented by ASME Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1," (Reference 1) as amended and noticed in the Federal Register (73 FR 52730, September 10, 2008 and 76 FR 36232, June 21, 2011).

3. APPLICABLE CODE REQUIREMENT 10 CFR 50.55a(g)(6)(ii)(D)(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)(D)(2) through (6).

For examination requirements, Paragraph - 2500 of Code Case N-729-1 states, in part:

"If obstructions or limitations prevent examination of the volume or surface requiredby Fig. 2 for one or more nozzles, the analysis procedure of Appendix I shall be used to demonstrate the adequacy of the examination volume or surface for each such nozzle. If Appendix I is used, the evaluation shall be submitted to the regulatory authorityhaving jurisdictionat the plant site" .

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 (0) < 30'; 1 inch for 8 > 30'; or to the end of the tube, whichever is less. These coverage requirements are applicable to Ginna Station reactor vessel head penetrations as follows:

Page 1 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Penetration Nozzle Incidence Angle, 8 Required Coverage, "a" Number (degrees) (inches) 1, 6-13, 18-21 < 30 1.5 14-17, 22-35, 37 >30 1.0 Note: Penetration nozzle numbers 2-5 and 36 do not exist

4. REASON FOR REQUEST Reactor vessel upper head control rod drive mechanism (CRDM) penetration nozzles at Ginna Station have two different nozzle bottom configurations. Penetration nozzle numbers 1, 6 through 33 inclusive, are essentially a smooth wall cylinder with a 0.188" corner radius on both the outer diameter and inner diameter at the bottom of the penetration nozzle (Reference 2).

The volumetric examination coverage for these penetration nozzles is therefore limited to the region between the toe of the J-groove weld and near the tangent point of the inside diameter corner radius at the bottom of the penetration nozzle. Penetration numbers 34, 35 and 37 have a 0.188" corner radius on the outer diameter and an internal taper 0.25" x 0.433" on the inside diameter at the bottom of the penetration nozzle (Reference 2). The volumetric examination coverage for these three outermost penetration nozzles is limited to the region between the toe of the J-groove weld and near the tangent point at the top of the internal taper at the bottom of the penetration nozzles. Due to the physical configuration and limitations of the examination equipment associated with a majority of the reactor vessel head penetration nozzles at Ginna Station, the full examination volume required by ASME Code Case N-729-1 Table 1 for Item No. B4.40 cannot be achieved. For most of the penetration nozzles, the geometric limitation reduces the volumetric examination coverage distance below the toe of the J-groove weld to less than the required coverage dimension "a" shown in Figure 2 of Code Case N-729-1.

R.E. Ginna obtained the achievable examination coverage data on all 32 CRDM penetration nozzles in the reactor vessel replacement head during the pre-service inspection (Reference 2).

The distance from the toe of the J-groove weld, identified as "a" in Figure 2 of Code Case N-729-1, varies based on location of the penetration nozzle in the reactor vessel head. This distance is in general longer for the penetration nozzle rows near the centerline of the reactor vessel head and shorter for the outermost penetration nozzle rows. The dimensional configuration at all the nozzles, with the exception of one nozzle, is such that the achievable examination coverage zone below the toe of the J-groove weld is less than that required in Figure 2 of Code Case N-729-1.

Table 1 lists the extent of the achievable examination coverage for the reactor vessel head penetration nozzles at Ginna Station. The achievable examination coverage below the toe of the J-groove weld on the downhill side, which is the limiting side of each penetration nozzle, corresponds to the proposed alternative examination coverage being requested. For Page 2 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations comparison, the examination coverage required by Figure 2 of ASME Code Case N-729-1 is also shown in Table 1. Based on the achievable examination coverage tabulated in Table 1, the volumetric and surface examination coverage requirements of ASME Code Case N-729-1 for Item B4.40 cannot be met for all the penetration nozzles with the exception of penetration nozzle number 24 as shaded in Table 1.

Table 1 Ginna Station Achievable Examination Coverage (Alternative Coverage Requested) Below the J-groove Weld for CRDM Penetration Nozzles and Code Case N-729-1 Requirements Page 3 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Penetration Code Case N-729-1 Achievable Examination Penetration Nozzle Required Coverage (Alternative Nozzle Number Incidence Angle Examination Coverage Requested)

(G) Coverage Below J- Below J-groove weld

(______ ___)___ groove Weld Below__ -grooveweld 29 37.00 1.0" 0.76" 30 37.00 1.0" 0.64" 31 37.00 1.0" 0.80" 32 37.00 1.0" 0.60" 33 37.00 1.0" 0.56" 34 43.50 1.0" 0.52" 35 43.50 1.0" 0.20" 37 43.50 1.0" 0.40" Note: Penetration nozzle numbers 2-5 and 36 do not exist

5. PROPOSED ALTERNATIVE AND BASIS FOR USE 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, R.E. Ginna proposes the 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. Paragraph 1-1000 of ASME Code Case N-729-1 requires that for alternative examination zones that eliminate portions of the Figure 2 examination zone below the J-groove weld, 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. In support of this relief request, the techniques of both Paragraph 1-2000 and Method 1 of Paragraph 1-3200 were used as appropriate.

5.1 ASME Code Case N-729-1 Paragraph 1-2000 Stress Analysis Paragraph 1-2000 of ASME Code Case N-729-1 requires that plant specific analysis be performed to demonstrate that the hoop and axial stresses remain below 20 ksi over the entire region outside the alternative examination coverage zone but within the required examination zone defined in Figure 2 of the Code Case. Finite element analyses were performed for five different CRDM penetration nozzle geometries, including the center penetration nozzle (incidence angle of 00), penetration nozzle rows with incidence angle of 13.7', 31.9', 37.00 and Page 4 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations 43.50. The results for penetration nozzle row with incidence angle of 31.90 were used to bound the results for those penetration nozzles with incidence angles of 28.20 and 30.00 since the differences in incidence angles are small and the limiting hoop stresses are in general more governing for the outermost penetration nozzles. The penetration nozzle numbers that are bounded by the analyzed penetration nozzles are shown in Table 2. The hoop stress distribution (Reference 3) for the analyzed penetration nozzles obtained from finite element stress analysis are provided in Figures 1 through 5 of this relief request. In addition, the distance below the toe of the J-groove weld where the steady state hoop stress is less than 20 ksi is also shown in Figures 1 through 5. It should be noted that the steady state hoop stresses are the most limiting stress components and therefore axial stresses are not considered and only hoop stress distributions need to be considered. The steady state hoop stresses are due to the residual stress resulting from the J-groove weld fabrication process as well as the normal operating temperature and pressure condition loads.

Table 2 Applicable Hoop Stress Distributions for Ginna Head Penetration Nozzles Analyzed Penetration Nozzle Applicable Figure No.

Penetration Nozzle Numbers Bounded by the for Hoop Stress Incidence Angle (0) Analyzed Nozzle Distribution 00 1 1 13.70 6-9 2 31.90 10-13, 14-17, 18-21, 22-25 3 37.00 26-33 4 43.50 34, 35, 37 5 Note: Penetration nozzle numbers 2-5 and 36 do not exist 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 is shown in Table 3 as well as in Figures 1 through 5.

With the exception of the outermost penetration nozzle row 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 Page 5 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Figure 2 of Code Case N-729-1. Penetration nozzle numbers 34, 35, and 37 are the outermost penetration nozzles with an incidence angle of 43.50. 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.

Table 3 Comparison Between Minimum Achievable Examination Coverage and the 1-2000 Required Examination Coverage Analyzed Range of Achievable Minimum Achievable 1-2000 Required Penetration Examination Examination Coverage Examination Coverage Nozzle Incidence Angle Coverage (inches) Below J-groove weld Below J-groove Weld Ie From Table 1 (inches) From Table 1 (inches) From Figures 1-5 00 1.00 1.00 0.52 13.70 0.88- 1.12 0.88 0.28 31.90 0.48- 1.12 0.48 0.31 37.00 0.56 - 0.88 0.56 0.31 43.50 0.20 - 0.52 0.20 0.33 Note: There are only three penetration nozzles (penetration numbers 34, 35 and 37) with incidence angle of 43.5*. The achievable examination coverage for penetration nozzle numbers 34, 35 and 37 from Table 1 are 0.52", 0.20" and 0.40" respectively.

5.2 ASME Code Case N-729-1 Paragraph 1-3200, Method I Deterministic Fracture Mechanics Analysis A deterministic fracture mechanics analysis was performed for penetration nozzle number 35 and documented in Reference 3. The purpose of the deterministic fracture mechanics analyses is to demonstrate that a potential axial crack in the unexamined zone below the J-groove weld will not grow to the toe of the J-groove weld prior to the next scheduled examination. The crack growth analysis was based on the stress results from the finite element stress analysis performed to address Paragraph 1-2000 of the Code Case and considered an improvement factor of 100 over Alloy 600 for Alloy 690 material in accordance with MRP-258 (Reference 4)

Page 6 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations and MRP-309 (Reference 5). NUREG/CR-7103 (Reference 6) also indicated similar results except for high levels of cold work during one-dimensional rolling but such level and orientation of cold work is not anticipated in light water reactor service components. An axial through-wall flaw is postulated below the J-groove weld with its upper extremity to be initially located at the bottom edge of the achievable examination coverage zone tabulated in Table 1 and the lower extremity to be located where either the inside or the outside surface hoop stress becomes compressive. For penetration nozzle number 35, an axial through-wall flaw is postulated with its upper extremity located at 0.20 inch below the J-groove weld. The result of the analysis using an improvement factor of 100 over Alloy 600 is shown as a crack growth curve in Figure 6 and demonstrated that a postulated through-wall axial flaw at the bottom edge of the achievable examination coverage zone for penetration nozzle number 35 would take more than 100 effective full power years to reach the toe of the J-groove weld. Therefore any flaws initiated below the J-groove weld, in the region of the penetration nozzle number 35 not being inspected, would not reach the toe of the J-groove weld before the next inspection period at Ginna Station.

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 allow 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. Penetration nozzle number 35 is the most limiting nozzle for Ginna Station with respect to the achievable examination coverage. The remaining penetration nozzles are not as limiting since there are margins available between the minimum achievable examination coverage, which is the proposed alternative examination coverage, and the minimum examination coverage required by Code Case N-729-1 Appendix 1-2000 as shown in Table 3 and Figures 1- 5.

In summary, the proposed alternative examination coverage tabulated in Table 1 for all the Ginna Station penetration nozzles have been demonstrated to be acceptable in accordance with the arialysis procedure for alternative examination area or volume definition in Mandatory Appendix I of Code Case N-729-1. As a result, the probability is extremely low for PWSCC to occur in the unexamined region of the Ginna Station reactor vessel upper penetration nozzles and that any potential undetected PWSCC cracks would not lead to a safety concern before the next inspection period. Relief Request ISI-07 applies to the penetration nozzles listed in Table 1 with the exception of penetration nozzle number 24, which meets Code Case N-729-1 examination coverage requirements.

Page 7 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations

6. DURATION OF PROPOSED ALTERNATIVE The duration of the proposed alternative is for the remainder of the Ginna Station 5th Interval ISI Program.

7. PRECEDENT Precedent for relief from the requirements of examination coverage exist since Beaver Valley, Unit 2; Arkansas Nuclear One, Unit 2; San Onofre, Unit 2; Indian Point, Unit 3; Braidwood, Units 1 and 2; and Byron, Units 1 and 2 have all been granted relief for the same configuration limitations.

The corresponding relief request and NRC acceptance letter for each plant are listed as follows:

Beaver Valley Unit 2:

1. Proposed Alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-729-1 Examination Requirements, December 30, 2008 (ML090020385)

2. Beaver Valley Power Station, Unit No. 2 - Relief Request No. 2-TYP-3-RV-02 Regarding the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-729-1 Examination Requirements, September 28, 2009 (ML092640111)

Arkansas Nuclear One, Unit 2:

1. Request for Alternative to 10 CFR 50.55a(G)(6)(ii)(D) Examination Requirements, Arkansas Nuclear One, Unit 2, February 9, 2009 (ML090400962)

2. Arkansas Nuclear One, Unit 2 - Request for Alternative ANO2-ISI-002 for the Remainder of the Current (Third) 10-Year ISI Interval and the Fourth ISI Interval Until the Reactor Vessel Head Is Replaced, August 27, 2009 (ML092300551)

San Onofre Unit 2:

1. Third Ten-Year Inservice Inspection (ISI) Interval Relief Request ISI-3-29, Reactor Vessel Head Inspection San Onofre Nuclear Generating Station, Units 2 and 3, February 27, 2009 (ML090620358)

2. Revision 1 to Third Ten-Year Inservice Inspection (ISI Interval Relief Request ISI-3-29, Inspection of Reactor Vessel Head Control Element Drive Mechanism Nozzles San Onofre Nuclear Generating Station, Units 2 and 3, October 2, 2009 (ML092790153)

3. San Onofre Nuclear Generating Station, Units 2 and 3 - Relief Request ISI-3-29, Request for Relief from Inspection Requirements of ASME Code Case N-729-1 for Control Element Drive Mechanism Penetrations, December 22, 2009 (ML093441035)

Indian Point Unit 3:

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Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations

1. Requests for Relief 3-45, 3-46, 3-47(l) and 3-48 to Support the Unit 3 Refuel Outage 15 Inservice Inspection Program, January 22, 2009 (ML090420062)

2. Indian Point Nuclear Generating Unit No. 3 - Relief Requests RR-3-45 and RR-3-46 for Reactor Vessel Head Penetrations Examination, July 8, 2009 (ML091880161)

Braidwood and Byron Units I and 2:

1. Request for Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds, March 12, 2010 (ML100710764)

2. Response to Request for Additional Information Regarding Request for Relief from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds, December 6, 2010 (ML103410394)

3. Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 - Relief Request from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds, March 3,2011 (ML110590921)

8. REFERENCES

1. ASME Code 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," Dated March 28, 2006.

2. WesDyne Report WDI-PJF-1306902-TCR-001 Rev.1, "R.E. Ginna Reactor Vessel Head Penetration Code Case Pre-Service Data Analysis Review." (Westinghouse proprietary)

3. LTR-PAFM-12-62-NP, Rev. 0, "Technical Basis For Alternative Examination Volume Below The J-Groove Weld For Ginna Replacement Reactor Vessel Head Penetration Nozzles."

4. Materials Reliability Program: Resistance to Primary Water Stress Corrosion Cracking of Alloy 690 in Pressurized Water Reactors (MRP-258). EPRI, Palo Alto, CA: 2009. 1019086. (Proprietary)

5. Materials Reliability Program: Primary Water Stress Corrosion Testing of Alloys 690 and Weld Metals-An Update (MRP-309). EPRI, Palo Alto, CA: 2011. 1022854.

(Proprietary)

6. NUREG/CR-7103 Volume 2, "Pacific Northwest National Laboratory Investigation of Stress Corrosion Cracking in Nickel-Base Alloys," April 2012 (ML12114A011)

7. 083NE001, Revision 2, General arrangement (Babcock & Wilcox Canada Proprietary)

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Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Figure 1 Downhill Side Hoop Stress Distribution Below J-groove Weld Toe for Center Penetration Nozzle with Incidence Angle of 0*

80,000 70,000 60,000 50,000 40,000 30,000 C 20,000 0.

0

= 10,000 0

-10,000

-20,000

-30,000

.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 Distance from Bottom of Weld (in)

I -insided Page 10 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Figure 2 Downhill Side Hoop Stress Distribution Below J-groove Weld Toe for Penetration Nozzles with Incidence Angle of 13.70 80,000 I 1 I 70,000 Minimum Achievable Examination --- --

Coverage (0.88 inch) Ij

. N-729. 1 Required Examination _ _

40,000 C-"overage (1.5 Inch) . - -_-_ - . - -

3-30,000 20,000 . .. . ... .. . .

0 0,0 0 'Para....gr..... .- - -_1-200.....

1 C10,000 Pargrah -20 Required IIIt ..

j Examination Coverage (0.28 I I "

-2000 Inch

-30,000 , n 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 Distance from Bottom of Weld (in)

I -Inside -U-Outside Page 11 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Figure 3 Downhill Side Hoop Stress Distribution Below J-groove Weld Toe for Penetration Nozzles with Incidence Angle of 31.9*

80,000 70,000 60,000 50,000 40,000 30,000 M

CL 20,000 0

0 0

-10,000

-20,000

-30,000 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 Distance from Bottom of Weld (in)

I -insideOutside Note: For Penetration numbers 10-13, 18-21 (incidence angles of 28.20 and 30') which are bounded by hoop stress distribution shown in this figure, the N-729-1 required examination coverage is 1.5".

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Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Figure 4 Downhill Side Hoop Stress Distribution Below J-groove Weld Toe for Penetration Nozzles with Incidence Angle of 37.00 80,000 70,000 60,000 50,000 40,000 M

30,000 u) 0 0, 20,000 0

10,000 0

-10,000

-20,000

-30,000 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 Distance from Bottom of Weld (in)

Ide-inside Page 13 of 15

Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Figure 5 Downhill Side Hoop Stress Distribution Below J-groove Weld Toe for Penetration Nozzles with Incidence Angle of 43,50 80,000 70,000 60,000 50,000

'=- 40,000 30,000 0 20,000 0

z 10,000 0

-10,000

-20,000

-30,000 +-

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 Distance from Bottom of Weld (in)

I In Outside Note: There are only three penetration nozzles (penetration numbers 34, 35 and 37) with incidence angle of 43.50. 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.

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Relief Request Number ISI-07 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzles Code Case N-729-1 Examination Volume Limitations Figure 6 Crack Growth Curve for Postulated Axial Through-wall Flaw Below the J-groove Weld for Penetration Nozzle Number 35 (Alloy 690 Improvement Factor of 100 over Alloy 600) 1.4 1.2

- 1.0 0

0 E 0.8

-- '-- ..... CRDM Nozzle Weld Region -

9 0.6 0.4

• 0.4

(.)

-0.2

-0.0 I

-0.2 . . I  ! 1 1

. I I T Service Life - 124.0 Effective Full Power Years -1

-0.4 . . . . I I

. I

. . 1 . . I . I I 1 1 1 1 1 1 M l I l I l I I 0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 Time (Effective Full Power Years)

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ENCLOSURE 2 Relief Request Number ISI-08 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzle Weep Channel Examination

Relief Request Number ISI-08 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzle Weep Channel Examination Alternative provides an acceptable level of quality and safety

1. ASME CODE COMPONENT(S) AFFECTED:

Code Class 1: 1

Reference:

ASME Code Case N-729-1/10 CFR 50.55a(g)(6)(ii)(D)

Item Number: B4.30, B4.40

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Description:==

Head with nozzles and partial penetration welds with PWSCC resistant materials.

SB-167 UNS N06690 Nozzles and UNS N06052 Partial-penetration welds of PWSCC-resistant materials in head.

Drawing number: 083NE001 Revision 2

2. APPLICABLE CODE EDITION AND ADDENDA The current code of record for the R. E. Ginna Nuclear Power Plant is the ASME Section XI Code, 2004 Edition, No Addenda, as augmented by ASME Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1," (Reference 1) as amended and noticed in the Federal Register (73 FR 52730, September 10, 2008 and 76 FR 36232, June 21, 2011).

3. APPLICABLE CODE REQUIREMENT 10 CFR 50.55a(g)(6)(ii)(D) (1) requires that examinations of the reactor vessel head be performed in accordance with ASME Code Case N-729-1 subject to conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6). The applicable requirement is the ASME Code Case N-729-1 condition in 10CFR50.55a(g)(6)(ii)(D)(3) "A demonstrated volumetric or surface leak path assessment through all J-groove welds shall be performed".

Additionally, Code case N-729-1 Table 1, Item B4.30 requires visual inspection at a frequency of "Every third refueling outage or 5 calendar years whichever is less".

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Relief Request Number ISI-08 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzle Weep Channel Examination Alternative provides an acceptable level of quality and safety

4. REASON FOR REQUEST Relief Request Number ISI-08 is being submitted to provide an alternative to the ASME Code Case N-729-1 condition in 10CFR50.55a(g)(6)(ii)(D)(3) to perform a volumetric or surface leak path assessment through all J-groove welds. The reason for this request is a design feature of the replacement Ginna Reactor Vessel Head that incorporated a weep channel in each control rod drive mechanism (CRDM) nozzle bore through the interference zone (reference 6). The weep channel is machined axially into the low alloy low alloy steel through the interference zone between the CRDM nozzle and penetration. This design feature provides a tell tale path for visual leak detection on the vessel at the CRDM nozzle to low alloy steel reactor vessel head bore annulus. This feature is unique to only a few replacement Reactor Vessel Heads in the US.

The ultrasonic leak path assessment is performed by monitoring the increase or decrease in sound amplitude reflections from the CRDM backwall, primarily within the CRDM interference zone annulus area (reference 4, reference 5). Sound is normally transmitted from the CRDM nozzle through the annulus to the low alloy steel bore base material in the interference zone due to the intimate interference fit tolerances associated with the CRDM tube and the low alloy steel bore base material. If an inservice leak path exists as a result of degradation from the CRDM or CRDM J weld, scenarios can be established that would modify these sound transmissions through the annulus in the following options. A compaction of boric acid or a wetted region in the annulus would result in an increase of sound transmission through the CRDM nozzle to low alloy steel base metal annulus. Conversely, ifthe leak conditions corroded and or eroded the low alloy steel base material this would result in a decrease of sound transmission through the CRDM to low alloy steel base metal annulus. The leak path assessment monitors for an increase or decrease of sound amplitude and distinct leak path or "riverbed" patterns in the ultrasonic c-scan images.

The weep channel that is designed into the R. E. Ginna reactor vessel head bore will not allow for the demonstrated ultrasonic leak path monitoring in the specific area of the weep channel.

The weep channel design has distinct machining characteristics when viewed ultrasonically versus distinct leak path riverbed C-scan geometries. Any potential reactor coolant system leakage through the CRDM nozzle or CRDM J groove weld that passes through the annulus would be free to leak through the machined weep channel to the exterior of the reactor vessel head. For R. E. Ginna, based upon the designed in leak path bypass there would be a smaller probability of a leak causing an increase or decrease of ultrasonic sound transmission of the CRDM nozzle to reactor vessel head low alloy steel bore annulus in the interference fit region.

But, there is an increased probability of detecting a leak visually due to the same design considerations.

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Relief Request Number ISI-08 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzle Weep Channel Examination Alternative provides an acceptable level of quality and safety Industry experience with removed CRDM housing and non destructive examination as well as destructive analyses has not detected 100% of leak paths with non destructive testing (reference 4), which states "the ultrasonic leakage path measurements were partially successful". Industry field experience with degraded CRDM nozzles have not always resulted in detectable ultrasonic leak path detection or in visual leakage detection on the exterior of the reactor vessel head.

With the advantage of the designed in weep channel in the R.E. Ginna reactor vessel head an improved leak path monitoring can be performed through bare metal visual examination.

5. PROPOSED ALTERNATIVE AND BASIS FOR USE R.E. Ginna Nuclear Power Plant proposes to perform the bare metal visual examination of the CRDM nozzle to reactor vessel head bore annulus area in lieu of a demonstrated volumetric or surface leak path by virtue of the weep channel designed into the Ginna replacement Reactor Vessel Head. The weep channel that is machined axially through the region of interference fit connects the larger tolerance (non interference fit) low alloy steel nozzle bore above and below the interference fit region. Any CRDM J weld pressure boundary leakage would be free to flow and promote detection.

R.E. Ginna has an improved reactor pressure vessel head design which makes the N-729-1 Table 1, Item B4.30 visual examination for leak path more effective than the conventional reactor vessel head designs using ultrasonic leak path detection.

Industry experience (reference 4) has shown that visual examination in conjunction with ultrasonic leak path examination is optimum for detecting reactor coolant leakage. These examinations are further improved upon with a visual examination of the R.E. Ginna Reactor head with the weep channel design.

6. DURATION OF PROPOSED ALTERNATIVE The duration of this relief request is the remainder of the Fifth Interval ISI Program. The proposed alternative provides an acceptable level of quality and safety as a result of the design feature of the replacement Reactor Vessel Head nozzle bore having a weep channel through each bore permitting any leakage to easily bypass the full axial distance of interference fit.

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Relief Request Number ISI-08 R.E. Ginna Nuclear Power Plant - Fifth Interval Inservice Inspection Program Proposed Alternative in Accordance with 10 CFR 50.55a (a)(3)(i)

Reactor Vessel Replacement Head Penetration Nozzle Weep Channel Examination Alternative provides an acceptable level of quality and safety

7. PRECEDENT Precedent for relief from the requirements of ultrasonic examination leak path assessment exists in NUREG/CR-6996 PNNL-18372 "Nondestructive and Destructive Examination Studies on Removed from Service Control Rod Drive Mechanism Penetrations". The integrated results from Non Destructive Examinations performed on North Anna removed nozzles, with focus on results from nozzle 31 and nozzle 54, clarified the capabilities of visual and ultrasonic examination.

8. REFERENCES

1. 10 CFR 50.55a, Industry Codes and Standards; Amended Requirements; Final Rule

2. ASME Code Case N-729-1, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure - Retaining Partial - Penetration WeldsSection XI, Division I

3. ASME Section Xl Code, 2004 Edition with no Addenda

4. NUREG/CR-6996 PNNL-18372 Nondestructive and Destructive Examination Studies on Removed from Service Control Rod Drive Mechanism Penetrations

5. WDI-TJ-006-03-P, Revision 4 Ultrasonic Testing of Interference Fit Samples for Leak Path Detection (Westinghouse proprietary)

6. 083ND1 12, Revision 2 Closure Head Initial Machining (Babcock & Wilcox Canada proprietary)

7. 083NE001, Revision 2 General Arrangement (Babcock & Wilcox Canada proprietary)

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