Text

Response to Request for Additional Information Regarding Relief Request I4R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs
	ML12104A067
Person / Time
Site:	Quad Cities
Issue date:	04/12/2012
From:	Gullott D; Exelon Generation Co, Exelon Nuclear
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML121040124	List: RS-12-069, Response to Request for Additional Information Regarding Relief Request I4R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs;
References
	RS-12-069
	Download: ML12104A067 (13)
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Exelon Generation Company, LLC www.exeloncorp.com 4300 Winfield Road Nuclear Warrenville, IL60555 Proprietary Information - Withhold From Public Disclosure Under 10 CFR 2.390 RS-12-069 10 CFR 50.55a April 12, 2012 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Quad Cities Nuclear Power Station, Unit 2 Renewed Facility Operating License No. DPR-30 NRC Docket No. 50-265

Subject:

Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs

References:

1) Letter from P. R. Simpson (Exelon Generation Company, LLC) to NRC, "Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs,*

dated April 6, 2012

2) Email fromJoeIWiebe(NRC)toD.M.Gullott (8<elon Generation Company, LLC), "Final RAls Regarding Quad Cities, Unit 2 Relief Request 14R-19," dated April 11, 2012

3) Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs,* dated April 11,2012

4) Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Further Clarification of RAI Question 10 for Quad Cities Unit 2 Relief Request 14R-19," dated April 12, 2012 In Reference 1, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i),

Exelon Generation Company, LLC (EGC), requested NRC approval of a relief request associated with the Fourth Inservice Inspection (lSI) Interval for Quad Cities Nuclear Power Station (QCNPS), Unit 2. Note that the fourth interval of the QCNPS lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1995 Edition with addenda through 1996.

Attachments 2 and 3 Contain Proprietary Information. Withhold From Public Disclosure Under 10 CFR 2.390 When separated from Attachments 2 and 3, this document is decontrolled

U. S. Nuclear Regulatory Commission April 12, 2012 Page 2 During review of the subject relief request, the NRC concluded that additional information would be needed to complete their review. The NRC transmitted the Request for Additional Information (RAI) to EGC in Reference 2. This request contained 10 individual items; however, during a telephone conversation between J. Wiebe (NRC) and D. M. Gullott (EGG), it was agreed that RAls 1-7 would be addressed immediately; and RAls 8-10 would be addressed in a follow-up near term response. The responses to RAls 1-7 were provided in Reference 3. The responses to RAls 8-10 are provided in Attachment 1 to this letter. Note that a clarification to RAI10 was transmitted to EGC in Reference 4. , "Corrosion Evaluation of the Quad Cities Unit 2 N-11 B Reactor Vessel Nozzle Modification," contains information in response to RAI 8; and Attachment 3, "Quad Cities Unit 2 Instrument Nozzle J-Groove Weld Flaw Evaluation," contains information in response to RAI 9.

As Attachments 2 and 3 contain information proprietary to AREVA NP Inc. (AREVA), they are supported by affidavits (Le., Attachment 4) signed by AREVA, the owner of the information. The affidavits set forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is respectfully requested that the information which is proprietary to AREVA be withheld from public disclosure.

A non-proprietary version of the information provided in Attachments 2 and 3 will be provided in a near term follow-up submittal.

There are no regulatory commitments contained in this letter.

Should you have any questions concerning this letter, please contact Mr. Joseph A. Bauer at (630) 657-2804.

Respectfully, David M. Gullott Manager - licenSing Exelon Generation Company, LLC : Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs : Corrosion Evaluation of the Quad Cities Unit 2 N-11 B Reactor Vessel Nozzle Modification : Quad Cities Unit 2 Instrument Nozzle J-Groove Weld Flaw Evaluation : AREVA NP Inc. Affidavits (2)

ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs In Reference 1, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i),

Exelon Generation Company, LLC (EGC), requested NRC approval of a relief request associated with the Fourth Inservice Inspection (lSI) Interval for Quad Cities Nuclear Power Station (QCNPS), Unit 2. Note that the fourth interval of the QCNPS lSI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1995 Edition with addenda through 1996.

As noted in Reference 1, during a recent pressure test of the reactor pressure vessel, one instrument penetration, N-11 B, was found to have approximately 60 drops per minute of leakage. Currently, there is not a qualified or demonstrated technique to perform volumetric non-destructive examination (NDE) of the partial penetration weld in this configuration that can be used to accurately characterize the location, orientation, or size of a flaw in the weld; therefore, as an alternative to performing the NDE required to characterize the flaw in penetration N-11 B, EGC proposed analyzing a maximum postulated flaw that bounds the range of flaw sizes that could exist in the J-groove weld and nozzle.

During review of the subject relief request, the NRC concluded that additional information would be needed to complete their review. The NRC transmitted the Request for Additional Information (RAI) to EGC in Reference 2. This request contained 10 individual items; however, during a telephone conversation between J. Wiebe (NRC) and D. M. Gullott (EGC), it was agreed that RAls 1-7 would be addressed immediately; and RAls 8-10 would be addressed in a follow-up near term response. The responses to RAls 1-7 were provided in Reference 3. The responses to RAls 8-10 are provided below. Note that a clarification to RAI10 was transmitted to EGC in Reference 4.

8. The proposed repair will result in a gap between the original and new nozzles inside the bore of the reactor vessel shell. The gap region may be susceptible to crevice or general corrosion affecting the reactor vessel shell and the new nozzle. Provide analysis to demonstrate that the corrosion of the shell will not affect the structural integrity of the new nozzle and the reactor vessel shell.

Response

The analysis which demonstrates that corrosion will not affect the structural integrity of the new nozzle and the reactor vessel shell is provided in Attachment 2.

9. Application Section 5, Basis, (e) states that a flaw evaluation will be performed for one cycle of operation. The licensee stated that qualification beyond one cycle will be performed based on detailed residual stress analysis and fatigue crack growth analysis.

a. Submit the one-cycle flaw evaluation,

Response

The one-cycle flaw evaluation is provided in Attachment 3.

Page 1 of 4

ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs

b. Deleted

c. Discuss how the licensee can demonstrate the acceptability of the repair or the one cycle flaw evaluation without an examination, after one cycle of operation after the repair,

Response

The repair for this nozzle establishes the pressure boundary at the outside of the vessel wall and abandons the original pressure boundary at the J-groove weld on the inside of the vessel wall. The repair will be examined as defined in response to RAI 7 (see Reference 3). Future examinations of the repair will be performed as required by the QCNPS ASME Section XI Inservice Inspection Program.

Although the actual leak path through the abandoned J-groove weld and nozzle remnant has not been determined, an initial flaw was assumed which encompassed all of the weld cross section subject to intergranular stress corrosion cracking (IGSCC), plus a small distance into the low alloy steel material. This was determined to be the largest flaw which could potentially grow into the vessel wall at the nozzle penetration.

Operating experience as documented in BWRVIP-60-A, BWRVIP-104 and BWRVIP-233-R1 have established the potentiallGSCC of the Alloy 182 weld metal and the cold worked Alloy 600 nozzle material, as well as establishing the insensitivity of the low alloy steel to SCC. Because this nozzle is located in the steam volume of the vessel, the severity of the operating transients is significantly reduced and the potential fatigue crack growth of this flaw is minimal. The initial flaw evaluation has demonstrated that this conservative hypothetical flaw would remain acceptable per the requirements of IWB-3600 and Appendix A of Section XI for a full operating cycle, with consideration of potential active fatigue crack growth. The subsequent elastic-plastic fracture mechanics (EPFM) evaluation is expected to show that a significantly larger flaw would be acceptable. The final fracture mechanics results will demonstrate flaw tolerance that meets acceptance criteria even when considering fatigue crack growth over the remaining extended operating life of the plant.

In summary, the EPFM evaluation is expected to demonstrate acceptability of a much larger flaw than would credibly occur due to growth of the present flaw, over the remaining life of the plant. Considering the limited number of transient events that may occur during an operating cycle and the small amount fatigue crack growth that is predicted to occur, an examination after one cycle of operation is not required to ensure the integrity of the vessel wall.

Page 2 of 4

ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs

d. Explain why the flaw evaluation requires two phases (i.e., one cycle calculation and beyond one cycle calculation),

Response

The ASME Section XI flaw evaluation requires a projection of crack growth for the flaw in the Alloy 182 weld of the remnant nozzle being abandoned in-place, and potentially into the reactor pressure vessel material. Since the limiting flaw geometry assumes that the Alloy 182 J-groove weld is already completely cracked, there is no credit taken for any remaining life of this material, and flaw growth in this material is therefore not evaluated.

There is the potential for the crack to propagate into the vessel low alloy steel, since propagation by fatigue crack growth is possible. Prediction of potential further growth requires the weld residual and design basis transient event through-wall stresses for the vessel shell at this nozzle. A refined prediction of the residual stress distributions and the deSign basis transient through-wall stress distributions requires detailed finite element analyses of the vessel shell and nozzle with the replacement nozzle. The initial evaluation uses a conservative approximation of the residual stress acting on the flaw and assumes a conservative amount of fatigue crack growth caused by the limited number of operating transients during a single operating cycle to demonstrate compliance with the ASME Section XI flaw evaluation requirements of IWB-3610.

The second phase is a more detailed evaluation using residual stresses determined by finite element analyses and includes the operating cycles projected for the remaining life of the plant when projecting the fatigue crack growth. Additionally, the second phase analyses will use more rigorous EPFM methods to establish additional margin for the flaw evaluation.

e. Confirm that flaw analysis will include or bound transient loads and conditions beyond normal operation,

Response

The flaw evaluation has determined the bounding transient events from the normal, upset, emergency and faulted conditions defined for plant design. The evaluation determined that the heat-up, cooldown, blowdown and SCRAM transients were the most limiting events. These events are included in the flaw evaluation. The heat-up, cooldown and SCRAM transients are defined as normal/upset conditions; and the blowdown transient is defined as an emergency/faulted condition.

f. Specify what load conditions the flaw analysis bound.

Response

The flaw evaluation combines pressure, thermal (including thermal transient stress) and residual stresses associated with the design basis plant operating condition. See Attachment 3 for additional details.

Page 3 of 4

ATTACHMENT 1 Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs

10. Provide a summary of the design analysis confirming that the design of the new nozzle will not eject from the reactor vessel under design conditions and that it satisfies the requirements of the ASME Code,Section III, NB-3200 and the plant's original instrument penetration design basis. For example, the original design relied on the j-groove with the weld on the reactor pressure vessel inside diameter and this repair process moves the pressure boundary weld to the outside diameter with new material for the nozzle. Does this new design adversely affect the design function of this pressure retaining component? Provide the projected date when the design analysis will be available for audit or inspection.

Response

The design analYSiS, confirming that the design of the new nozzle will not eject from the reactor vessel under design conditions, is being performed in accordance with the QCNPS current design and licensing basis of the instrument penetration which satisfies the requirements of ASME Code,Section III and the original instrument penetration design basis. The new design is reconciled to the original code of construction, addressing all design and transient loads and ensures all code requirements are met. Therefore, adequate design margin is maintained and the repair does not adversely affect the pressure retaining capability of the nozzle. The design analysis will be completed prior to Unit 2 startup.

Startup is currently projected for Monday, April 16, 2012.

Summary of Loose Parts Evaluation Flaw indications were identified in the original N-11 B nozzle. Given the original nozzle will not be entirely removed, Exelon completed a lost-parts evaluation to assess the potential for nozzle segments to enter the reactor vessel during power operation. Two evaluations were completed as follows: EC 388684 addressed the potential impact on the fuel, while EC388654 addressed the potential impact on internal vessel components. The evaluations, completed in accordance with approved Exelon procedures, determined the potential for lost parts did not pose any safety concerns. The evaluations considered interfacing systems and other reactor vessel internal components, flow blockage, and adverse chemical reactions.

REFERENCES

1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to NRC, "Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 6, 2012

2. Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Final RAls Regarding Quad Cities, Unit 2 Relief Request 14R-19," dated April 11, 2012

3. Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Response to Request for Additional Information Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs," dated April 11, 2012

4. Email fromJoeIWiebe(NRC)toD.M.Guliott (Exelon Generation Company, LLC), "Further Clarification of RAI Question 10 for Quad Cities Unit 2 Relief Request 14R-19," dated April 12, 2012 Page 4 of 4

ATTACHMENT 4 Quad Cities Nuclear Power Station Unit 2 AREVA NP Inc.

Affidavits (2)

Supporting Response to Request for Additional Information (RAI) #8 and #9 Regarding Relief Request 14R-19 Associated with the Reactor Pressure Vessel Nozzle Repairs

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. (AREVA NP) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in Engineering Information Record 51-9180975-001 entitled, "Corrosion Evaluation of the Quad Cities Unit 2 N11-B Reactor Vessel Nozzle Modification,* dated April 2012 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information":

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this dayof ~~ 2012.

Kathleen Ann Bennett NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31/15 Reg. # 110864 KATHLEEN ANN BENN£TT ~

Notary Pubtlc ~

Commamrelltlt of Virginia 110114 My CommiIIIofI &pIrtt Aut 31, 2015

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

COUNTY OF CAMPBELL )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. (AREVA NP) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in Calculation Summary Sheet 32-9181076-001, "Quad Cities Unit 2 Instrument Nozzle J-Groove Weld Flaw Evaluation," dated April, 2012 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.