2CAN100902, Request for Alternative ANO2-ISI-004 - Proposed Alternative to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examinations

Request for Alternative ANO2-ISI-004 - Proposed Alternative to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examinations

ML093030136
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/29/2009
From:	David Bice Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN100902, TAC MC9768, TAC MD9934
Download: ML093030136 (9)

Text

2CAN100902 October 29, 2009 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Request for Alternative ANO2-ISI-004 Proposed Alternative to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examinations Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6 REFERENCES

1. Letter from NRC to Mr. Gordon Bischoff, Manager, Owners Group Program Management Office, Final Safety Evaluation of Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR)

WCAP-16168-NP, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, (TAC No. MC9768), dated May 8, 2008.

2. NRC letter to Indiana Michigan Power Company, dated June 8, 2009, Donald C. Cook Nuclear Plant, Unit 2 (CNP-2) - Evaluation of Relief Request (ISIR-29) to Extend the Third 10-Year Inservice Inspection (ISI)

Interval for Reactor Vessel Weld Examination (TAC No. MD9934).

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Operations, Inc. (Entergy) requests an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Code,Section XI, paragraph IWB-2412, Inspection Program B, for Arkansas Nuclear One, Unit 2 (ANO-2). In particular the requirement that volumetric examination of essentially 100% of the reactor vessel pressure retaining Examination Category B-A and B-D welds be performed once each ten-year interval.

The proposed alternative is to extend this examination frequency to once every 20 years. This alternative is based upon the methodology outlined in Reference 1. Request for Alternative ANO2-ISI-004 is provided as the attachment to this letter.

The NRC approved the Reference 1 WCAP which provides for extension of the inservice inspection (ISI) interval for certain pressure retaining welds in the reactor vessel from 10 to Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 David B. Bice Acting Manager, Licensing Arkansas Nuclear One

2CAN100902 Page 2 of 3 20 years. Entergy proposes to implement this extended ISI interval for ANO-2. The plant-specific information identified by Reference 1 needed to support this request is included in the attached request. Entergy has concluded that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i).

At the time of the issuance of Reference 1, it was the intent of the NRC, as noted in Reference 2, to establish a process by which licensees could receive approval to implement 20-year ISI intervals for the subject component examinations through the end of the facilitys current operating license. This objective lead to the provision established in Reference 1 for licensees to submit a license condition requiring them to evaluate future volumetric ISI data in accordance with the criteria in the draft and/or final alternative Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61a. Subsequent guidance from the NRCs Office of General Counsel has resulted in a modification of this NRC position, as discussed below.

Based on the current guidance, the NRC staff grants ISI interval extensions for the subject components on an interval-by-interval basis (i.e., only a facilitys current ISI interval may be extended for up to 20 years). Licensees will have to submit subsequent requested alternatives for NRC review and approval to extend each following ISI interval from 10 years to 20 years, as needed. Based on this new NRC position, the requirement in Reference 1 for a license condition to address the evaluation of future ISI data is no longer necessary. Subsequently, Entergy will not submit a proposed license condition for ANO-2 as part of this request.

This letter contains no new commitments.

Entergy requests approval of the proposed request for alternative by January 15, 2011, in order to support the spring 2011 refueling outage for ANO-2. Although this request is neither exigent nor emergency, your prompt review is requested.

If you have any questions or require additional information, please contact me.

Sincerely, DBB/rwc

Attachment:

Request for Alternative ANO2-ISI-004

2CAN100902 Page 3 of 3 cc:

Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 U.S. Nuclear Regulatory Commission Attn: Mr. Kaly Kalyanam MS O-8B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852

ATTACHMENT TO 2CAN100902 REQUEST FOR ALTERNATIVE ANO2-ISI-004

Attachment to 2CAN100902 Page 1 of 5 Proposed Alternative ANO2-ISI-004 In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety 1 ASME Code Component(s) Affected The affected component is the Arkansas Nuclear One, Unit 2 (ANO-2) reactor vessel, specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1) examination categories and item numbers covering examinations of the reactor vessel (RV). These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.

Examination Category Item No.

Description B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Shell Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas Throughout this request the above examination categories are referred to as the subject examinations and the ASME BPV Code,Section XI, is referred to as the Code.

2 Applicable Code Edition and Addenda

ASME Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Code 1992 Edition and the 1995 Edition with the 1996 Addenda for the ultrasonic testing (UT).

3 Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each 10-year interval. The ANO Unit 2 third 10-year inservice inspection (ISI) interval is scheduled to end March 25, 2011. In accordance with IWA-2430 D.1, each inspection interval may be reduced or extended by as much as one year. Entergy has utilized this one-year extension for this ANO-2 inspection.

4 Reason for Request

An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor vessel pressure retaining, Examination Category B-A and B-D welds, be performed once each 10-year interval. Extension of time frame between Examination Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.

Attachment to 2CAN100902 Page 2 of 5 5 Proposed Alternative and Basis for Use Entergy proposes to not perform the ASME Code required volumetric examination of the ANO-2 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the third ISI interval, currently scheduled for 2011. Entergy will perform the ASME Code required volumetric examination of the ANO-2 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the fourth ISI interval in 2018. These dates are consistent with the information provided to the Staff in Pressurized Water Reactor Owners Group (PWROG) letter OG-06-356 (Reference 2).

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current time frame can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 2, Risk Informed Extension of the Reactor Vessel In-Service Inspection Interval (Reference 4). This study focuses on risk assessments of materials within beltline region of the RV wall. The results of the time frame calculations for ANO-2 were compared to those obtained from the Combustion Engineering (CE) pilot plant evaluated in WCAP-16168-NP-A, Revision 2. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for ANO-2 are bounded by the results of the CE pilot plant qualifies ANO-2 for an ISI interval extension. Table 1 below lists these critical parameters investigated in the WCAP and compares the results of the CE pilot plant to that of ANO-2. Tables 2 and 3 provide additional information that was requested by NRC and included in Appendix A of Reference 4.

Removing the requirement to perform these examinations in the third interval and scheduling the examination in the fourth interval allows Entergy to use the Appendix VIII rules in effect during the fourth 10-year interval in lieu of those in effect at the beginning of the third 10-year interval. Therefore, Entergy will not be required to maintain both sets of non-destructive examination (NDE) procedures, personnel and procedure qualifications, and acceptance criteria (third and fourth intervals).

Table 1: Critical Parameters for Application of Bounding Analysis - ANO-2 Parameter Pilot Plant Basis Plant Specific Basis Additional Evaluation Required?

Dominant Pressurized Thermal Shock (PTS)

Transients in the NRC PTS Risk Study are Applicable NRC PTS Risk Study (Reference 5)

PTS Generalization Study (Reference 6)

No

Attachment to 2CAN100902 Page 3 of 5 Through-Wall Cracking Frequency (TWCF) 3.16E-7 Events per year (Reference 4) 5.59E-14 Events per year (Calculated per Reference 4)

No Frequency and Severity of Design Basis Transients 13 heatup / cooldown cycles per year (Reference 4)

Bounded by 13 heatup / cooldown cycles per year No Cladding Layers (Single/Multiple)

Single Layer (Reference 4)

Single Layer No Table 2 below provides a summary of the latest reactor vessel inspection for ANO-2 and evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the ANO-2 reactor vessel.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection - ANO-2 Inspection methodology:

The most recent ISI of the Examination Category B-A and B-D welds was performed to Regulation Guide 1.150, Revision 1 (Reference 7) requirements. Future ISI will be performed to ASME Section XI Appendix VIII requirements.

Number of past inspections:

Two 10-Year ISIs have been performed.

Number of indications found:

Eight indications were identified in the beltline region during the most recent ISI. These indications were acceptable per Table IWB-3510-1 of Section XI of the ASME Code.

Proposed inspection schedule for balance of plant life:

The third ISI is scheduled for 2011. This inspection is scheduled to be performed in 2018. This information is consistent with the PWROG implementation plan (Reference 2).

Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Attachment to 2CAN100902 Page 4 of 5 Table 3: Details of TWCF Calculation - ANO Unit 2 at 48 Effective Full Power Years (EFPY)

Inputs Reactor Coolant System Temperature, TRCS[ºF]:

N/A Twall [inches]:

8.09375 No Region/Component Description Material

/ Heat No.

Cu

[wt%]

Ni

[wt%]

R.G.

1.99 Pos.

CF

(ºF)

Un-irradiated RTNDT [ºF]

Fluence [1019 Neutron/cm2, E > 1.0 MeV]

1 Inter. Shell Long.

Weld 2-203 A 10120 0.046 0.082 2.1 14.9

-56 5.08 2

Inter. Shell Long.

Weld 2-203 B 10120 0.046 0.082 2.1 14.9

-56 3.92 3

Inter. Shell Long.

Weld 2-203 C 10120 0.046 0.082 2.1 14.9

-56 3.92 4

Lower Shell Long.

Weld 3-203 A 10120 0.046 0.082 2.1 14.9

-56 5.10 5

Lower Shell Long.

Weld 3-203 B 10120 0.046 0.082 2.1 14.9

-56 3.94 6

Lower Shell Long.

Weld 3-203 C 10120 0.046 0.082 2.1 14.9

-56 3.94 7

Lower/Inter. Shell Girth Weld 9-203 83650 0.045 0.087 1.1 34.1

-10 5.32 8

Intermediate Shell C-8009-1 SA-533B Cl.1 0.098 0.605 1.1 63.6

-26 5.32 9

Intermediate Shell C-8009-2 SA-533B Cl.1 0.085 0.600 1.1 54.5 0

5.32 10 Intermediate Shell C-8009-3 SA-533B Cl.1 0.096 0.580 2.1 40.7 0

5.32 11 Lower Shell C-8010-1 SA-533B Cl.1 0.085 0.585 1.1 54.5 12 5.34 12 Lower Shell C-8010-2 SA-533B Cl.1 0.083 0.668 1.1 53.1

-28 5.34 13 Lower Shell C-8010-3 SA-533B Cl.1 0.080 0.653 1.1 51.0

-30 5.34 Outputs Methodology Used to Calculate T30:

Regulatory Guide 1.99, Revision 2 (Reference 8)

Controlling Material Region No. (From Above)

RTMAX-XX

[ºR]

Fluence

[1019 Neutron/cm2 E > 1.0 MeV]

FF (Fluence Factor)

T30

[ºF]

TWCF95-XX Limiting Axial Weld - AW 11 545.4 3.94 1.353 73.73 0.00E+00 Limiting Plate - PL 11 548.8 5.34 1.415 77.12 2.24E-14 Circumferential Weld -

CW 11 548.8 5.32 1.414 77.08 0.00E+00 TWCF95-TOTAL(AWTWCF95-AW + AWTWCF95-PL + AWTWCF95-CW):

5.59E-14

Attachment to 2CAN100902 Page 5 of 5 6 Duration of Proposed Alternative This request is applicable to the ANO-2 ISI program for the third 10-year inspection interval.

7 References

1. ASME Boiler and Pressure Vessel Code,Section XI, 1992 Edition, American Society of Mechanical Engineers, New York.

2. OG-06-356, Plan for Plant-Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval. MUHP 5097-99, Task 2059, October 31, 2006.

3. NRC Regulatory Guide 1.174, Revision 1, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, November 2002.

4. WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, June 2008.

5. NUREG-1874, 10 CFR Part 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 10/3/07 (ADAMS Accession Number ML070860156).

6. NRC Letter Report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS)

Risk Results to Additional Plants, December 14, 2004 (ADAMS Accession Number ML042880482).

7. NRC Regulatory Guide 1.150, Revision 1, Ultrasonic Testing of the Reactor Vessel Welds During Preservice and Inservice Examinations, February 1983.

8. NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.