L-PI-08-097, Responses to NRC Requests for Additional Information Dated November 4, 2008 Regarding Application for Renewed Operating Licenses

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Responses to NRC Requests for Additional Information Dated November 4, 2008 Regarding Application for Renewed Operating Licenses
ML083370202
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 11/12/2008
From: Wadley M
Northern States Power Co, Xcel Energy
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-08-097
Download: ML083370202 (11)
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Text

_ XcelEnergy-November 12, 2008 L-PI-08-097 10 CFR 54 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Responses to NRC Requests for Additional Information Dated November 4, 2008 Regarding Application for Renewed Operatingq Licenses By letter dated April 11, 2008, Northern States Power Company, a Minnesota corporation, (NSPM) submitted an Application for Renewed Operating Licenses (LRA) for the Prairie Island Nuclear Generating Plant (PINGP) Units I and 2. In a "L-PI-08-080, License Amendment Request (LAR) to Revise Emergency Diesel Generator (EDG) Test Loads in [[SR" contains a listed "[" character as part of the property label and has therefore been classified as invalid. and SR 3.8.1.9|letter dated November 4, 2008]], the NRC transmitted Requests for Additional Information (RAIs) regarding that application. This letter provides responses to those RAIs. provides the text of each RAI followed by the NSPM response.

If there are any questions or if additional information is needed, please contact Mr. Eugene Eckholt, License Renewal Project Manager.

Summary of Commitments This letter contains no new commitments or changes to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November 12, 2008.

Michael D. Wadley Site Vice President, Prairie Island Nuclear Generating Plant Units 1 and 2 Northern States Power Company - Minnesota 1717 Wakonade Drive East

  • Welch, Minnesota 55089-9642 Telephone: 651.388.1121

Document Control Desk Page 2 Enclosure (1) cc:

Administrator, Region III, USNRC License Renewal Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC Prairie Island Indian Community ATTN: Phil Mahowald Minnesota Department of Commerce Responses to NRC Requests for Additional Information Dated November 4, 2008 RAI B.2.1.34 (A) Section 6.0 in aging management program X1.31, "Reactor Vessel Surveillance," in NUREG-1 801, "Generic Aging Lessons Learned (GALL) Report," Volume 2, Revision 1 states that if an applicant has a surveillance program that consists of capsules with a projected fluence of less than the projected 60 year fluence at the end of 40 years, at least one capsule is to remain in the reactor vessel and tested during the extended period of operation.

Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2 have tested four of the six surveillance capsules in each unit to date, and the latest capsules of Units 1 and 2 were tested at projected fluence values which are less than 60 year fluence.

To ensure that the applicant is in compliance with the aforementioned GALL Report requirement, one of the two remaining capsules in each unit should be tested during the extended period of operation and the following relevant information regarding the testing of the capsules should be provided to the staff.

(1)

Applicant's plan to test an additional surveillance capsule from each unit (2)

The projected refueling outages of withdrawal for each unit (3)

Projected neutron fluence value for each capsule at the time of withdrawal (B) The staff requests that the applicant confirm that the withdrawal schedule of the capsules to be used for future tests during the extended period of operation is consistent with the requirements, specifically the limitations on lead factor, specified in paragraph 7.6.2 of the American Society of Testing Materials (ASTM)

E 185, "Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels."

(C) The staff requests that the applicant confirm that untested surveillance capsules (standby capsules) will be stored for future use at PINGP, Units 1 and 2.

NSPM Response to RAI B.2.1.34 Part (A)

(A)(1) Initially, six (6) surveillance capsules were installed in each Unit at PINGP. Four capsules were removed per an established withdrawal schedule. The last capsule to be removed from Unit 1 was Capsule'S' at 18.12 EFPY and the last capsule removed from Unit 2 was Capsule 'P' at 17.24 EFPY. Two spare capsules remain installed in each reactor vessel. To account for license renewal, one of the two remaining capsules from each unit will be withdrawn after the capsule has received a neutron fluence equivalent to the 60-year fluence. After capsule removal, the surveillance specimens will be tested in accordance with the requirements of 10 CFR 50, Appendix H and ASTM E 185-82.

The results of materials testing, fluence analysis, and effective full power years (EFPY) from startup are used to predict the effects of neutron embrittlement through the end of extended life. The remaining spare capsules will stay in the reactor vessel to provide meaningful metallurgical data for potential future license renewals.

I

Enclosure I Responses to NRC Requests for Additional Information Dated November 4, 2008 (A)(2) Prairie Island Unit 1 surveillance capsule withdrawal is planned for refueling outage 1 R27, expected to occur in 2011. Prairie Island Unit 2 surveillance capsule withdrawal is planned for refueling outage 2R27, expected to occur in 2012.

(A)(3) The maximum reactor vessel fluence at 54 EFPY for Unit 1 is projected to be 5.162E19 n/cm 2 (E>1.0 MeV), and for Unit 2 is projected to be 5.196E19 n/cm2 (E>1.0 MeV). Per the requirements of ASTM E 185-82, Section 7.6, the surveillance capsules should be removed when their neutron fluence exceeds the new peak End-Of-Life (EOL) vessel fluence (i.e.; 5.162E19 n/cm 2 for Unit 1 at 54 EFPY, and 5.196E19 n/cm 2 for Unit 2 at 54 EFPY), but prior to exceeding twice that fluence exposure (i.e.;

1.032E20 n/cm2 for Unit 1, and 1.039E20 n/cm 2 for Unit 2). Calculations have been performed to determine the earliest withdrawal times for the remaining capsules. The results are presented below:

Unit I Capsule T (Lead Factor = 1.89): Removal Time > 24.2 EFPY Capsule N (Lead Factor = 1.77): Removal Time > 26.5 EFPY Unit 2 Capsule N (Lead Factor = 1.72): Removal Time > 28.5 EFPY Capsule S (Lead Factor = 1.72): Removal Time > 28.5 EFPY As of July 1, 2008, the total lifetime performance of both Units is 29.0 EFPY.

Additional calculations have been performed to estimate the neutron fluence of each capsule at the time of next withdrawal. Current plans are to remove one capsule from each Unit during refueling outages 1 R27 and 2R27. The following are the projected fluence values for each capsule at that time:

Unit 1 (Next capsule removal planned for I R27, Withdrawal EFPY = 31.6)

Capsule T (Lead Factor = 1.89): Projected Neutron Fluence = 6.292E1 9 n/cm 2 Capsule N (Lead Factor = 1.77): Projected Neutron Fluence = 5.893E19 n/cm2 Unit 2 (Next capsule removal planned for 2R27, Withdrawal EFPY = 32.2)

Capsule N (Lead Factor = 1.72): Projected Neutron Fluence = 5.739E19 n/cm2 Capsule S (Lead Factor = 1.72): Projected Neutron Fluence = 5.739E1 9 n/cm2 Part (B)

(B) The surveillance capsule withdrawal schedule is generated based upon the requirements specified in ASTM E 185-82, Section 7.6. Six surveillance capsules were installed in each Unit at PINGP. Four capsules have been removed per an established withdrawal schedule. The last capsule to be removed from Unit I was Capsule 'S' at 18.12 EFPY and the last capsule removed from Unit 2 was Capsule 'P' at 17.24 EFPY.

In accordance with the requirements of ASTM E 185-82, Section 7.6.2, one of the remaining capsules will be withdrawn from each Unit when its neutron fluence exposure exceeds the new peak EOL (54 EFPY) vessel fluence, but prior to exceeding twice that 2

Enclosure I Responses to NRC Requests for Additional Information Dated November 4, 2008 fluence exposure. A revised capsule withdrawal schedule will be submitted for NRC approval prior to implementation in accordance with the requirements of 10 CFR 50, Appendix H.

Part (C)

(C) To account for license renewal, one of the two remaining capsules from each Unit will be withdrawn after the capsule receives a neutron fluence equivalent to the 60-year vessel fluence. The remaining spare capsules (one per Unit) will stay in the reactor vessel to provide meaningful metallurgical data for potential future license renewals. As discussed in LRA Section B2.1.34 (Page B-69) and Commitment No. 27, the PINGP Reactor Vessel Surveillance Program will be enhanced by adding a requirement to the program to ensure that in the event spare (standby) capsules are withdrawn, the untested capsules are placed in storage and maintained for future insertion at PINGP Units 1 and 2.

3

Enclosure I Responses to NRC Requests for Additional Information Dated November 4, 2008 RAI 4.2.2 Based on the beltline material "Circumferential Weld - Nozzle Shell Forging B to Intermediate Shell Forging C", the staff assumes that there is a small line nozzle penetration in the beltline region of the RPV. The staff requests the applicant to confirm this detail.

NSPM Response to RAI 4.2.2 The "Circumferential Weld - Nozzle Shell Forging B to Intermediate Shell Forging C," is a circumferential weld that connects the Reactor Pressure Vessel (RPV) nozzle (or upper) shell forging B to the RPV intermediate shell forging C. There is no small line nozzle penetration in the beltline region of the PINGP RPV. The RPV beltline region is depicted in PINGP USAR Figure 4.7-4.

4 Responses to NRC Requests for Additional Information Dated November 4, 2008 RAI 4.2.3 In Table 4.2-5 of the PINGP license renewal application (LRA), the RTNDT(u) for beltline material "Lower Shell Forging D (22642)" is listed as -4 OF. However, the value of RTNDT(u) for this beltline material is listed as 2 OF in the PINGP reactor vessel integrity database (RVID). Provide information which documents where the value of -4 OF comes from and demonstrate that is applicable to this forging.

NSPM Response to RAI 4.2.3 The initial RTNDT(u) of -4 OF for PINGP Unit 2, lower shell forging D (22642) is consistent with the RTNDT(u) value used in the current "PINGP Units One and Two Pressure and Temperature Limits Report", Revision 3 (Effective until 35 EFPY) and reported in WCAP-14637, "Prairie Island Unit 2 Heatup and Cooldown Limit Curves for Normal Operation", Revision 3, December 1999 (ML023230354 and ML003703560, respectively).

The initial RTNDT(u) of -4 IF for PINGP Unit 2, lower shell forging D (22642) is obtained from the PINGP USAR, Table 4.7-9, which references, as a data source, Structural Integrity Report SIR-99-075, Revision 2, "Update of the Response to Generic Letter 92-01 Reactor Vessel Structural Integrity for Prairie Island Units 1 and 2", November 1999 (ML993410414). This report was previously transmitted from NSP to the NRC via letter entitled "Comprehensive Revised Response to Generic Letter 92-01", dated November 10, 1999 (ML993330371). The initial RTNDT(u) of -4 IF for lower shell forging D (22642) is documented as a change on page viii, Table 2-4, and Table 2-5 of SIR-99-075, Revision 2. Section 7.0 of SIR-99-075 provided a summary of the PINGP information in the RVID2 database, and proposed a number of revisions to the database. As noted in Table 7-3 (page 7-4) of SIR-99-075, the proposed changes included a correction in RTNDT(u) from 2 OF to -4 OF for lower shell forging D (22642). Therefore, the -4 OF value of RTNDT(u) is applicable to lower shell forging D.

5 Responses to NRC Requests for Additional Information Dated November 4, 2008 RAI: 4.2.4 Per the requirements of Title 10 of the Code of Federal Regulations, Section 54.21(c)(1 )(i) and the staff guidance in NUREG 1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," the projected neutron fluence and subsequent adjusted reference temperature (ART) values at the end of the extended period of operation are reviewed to verify that they are bound by the embrittlement assumed in the existing P-T limit analysis. Therefore, the staff requests the applicant to provide current 1 T and 3/4 T ART values for all the beltline materials.

NSPM Response to RAI 4.2.4 As discussed in Section 4.2.4 (Page 4.2-9) of the PINGP LRA, PINGP has selected 10 CFR 54.21 (c)(1)(iii) and not (i) to manage the TLAA associated with P-T limits.

Updated P-T limits for the period of extended operation will be managed by the Reactor Vessel Surveillance Program. As stated in Appendix B2.1.34 (Page B-68) of the PINGP LRA, "The Reactor Vessel Surveillance Program manages updates of pressure-temperature operating limitations and the surveillance specimen withdrawal schedule, as needed, consistent with plant Technical Specifications, the Pressure and Temperature Limits Report, and 10 CFR 50.60 and 10 CFR 50, Appendix H."

In response to the NRC's request, %T and %T adjusted reference temperature (ART) values are provided below for all the beltline materials of PINGP Unit I and Unit 2. ART values are provided at 35 EFPY, which is the basis for the current P-T limits, as well as at 54 EFPY, which is assumed to define the end of the period of extended operation.

The 54 EFPY ART values were calculated using the procedures given in Regulatory Guide 1.99, Revision 2, and are provided for information only.

PINGP Unit I ART Values Unit 1-Beltline ID Material ART (°F) 1/4T ART (°F) 314T ART (°F) 1/4T ART (-F) 314T Type at 35 EFPY at 35 EFPY at 54 EFPY at 54 EFPY Nozzle Shell A 508-3 871 751 831 721 Forging B-21744/38384 Lower Shell A 508-3 85' 76T 798 Forging D-21887/38530 Int. Shell Forging A 508-3 1031 94' 1061 971 C-21918138566 1172 1052 1202 1092 Circ Weld-Int.

UM40, 1311 1161 135' 1201 Shell Forging C to Flux UM89 1452 1282 1492 1332 Lower Shell Forging D, 1752 Circ Weld-Nozzle
UM40, 154' 136' 149' 1311 Shell Forging B to Flux UM89 Int. Shell Forging C, 2269 1-Position (1.1) - No Surveillance Data 2-Position (2.1) - Surveillance Data 6

Responses to NRC Requests for Additional Information Dated November 4, 2008 PINGP Unit 2 ART Values Unit 2-Beltline ID Material ART (°F) 114T ART (°F) 314T ART (OF) 1/4T ART (°F) 314T Type at 35 EFPY at 35 EFPY at 54 EFPY at 54 EFPY Nozzle Shell A 508-3 71' 61' 67' 57' Forging B-22231139088 Lower Shell A 508-3 95' 84' 97' 871 Forging D-22642 1062 942 1092 972 Int. Shell Forging A 508-3 104' 951 1067 97' C-22829 Circ Weld-Int.

UM40, 91' 791 931 83' Shell Forging C Flux 992 822 1032 862 to Lower Shell UM89 Forging D, 2721 Circ Weld-Nozzle
UM40, 122' 106' 116' 1001 Shell Forging B Flux 1342 1162 1272 1092 to Int. Shell UM89 Forging C, 1752 1

1 1

1 1

1-Position (1.1) - No Surveillance Data 2-Position (2.1) - Surveillance Data The limiting Unit 1 material at 35 EFPY is the Circumferential Weld - Nozzle Shell Forging B to Intermediate Shell Forging C (2269) with a 1/4T ART of 154 OF and 3/4T ART of 136 OF. The limiting Unit 2 material at 35 EFPY is the Circumferential Weld -

Nozzle Shell Forging B to Intermediate Shell Forging C (1752) with a 1/4T ART of 134 OF and 3/4T ART of 116 OF. The 54 EFPY adjusted reference temperatures (1/4T and 3/4T) of these limiting materials have decreased relative to these 35 EFPY ART values.

Based on a comparison of the 54 EFPY ARTs and 35 EFPY ARTs for the current limiting materials, it is anticipated that PINGP will have sufficient operating margin to conduct plant heatups and cooldowns for the period of extended operation.

7 Responses to NRC Requests for Additional Information Dated November 4, 2008 RAI 4.7.2 In Section 4.7.2 of the PINGP LRA, the applicant states that the number of design cycles and transients assumed in topical report WCAP-1 5338, "A Review of Cracking with Weld Deposited Cladding in Operating PWR Plants," analyses bound the number of design cycles and transients projected for 60 years of operation as presented in Table 4.3-1 of the PINGP LRA. However, the WCAP-1 5338 report does not explicitly state the number of design cycles and transients used in the analyses of crack growth for 60 years of operation of the reactor pressure vessel. The staff requests the applicant to provide the bounding number of design cycles and transients that were used in the WCAP-1 5338 report analyses.

NSPM Response to RAI 4.7.2 The numbers of design cycles used in the fatigue crack growth evaluation is reported on page 9-10 of WCAP-15338-A. For convenience, the table is reproduced below as it appears in WCAP-1 5338-A.

REACTOR COOLANT SYSTEM TRANSIENTS FOR 40 YEARS Transient Identification WCAP-1 5338 Number for 40 Years*

Normal Conditions

1. Heatup and Cooldown at 10OF/hr 200 each
2. Load Follow Cycles (unit loading and unloading at 5% of full 18300 power / min)
3. Step load increase and decrease of 10% of full power 2000 each (20% step load changes)
4. Large step load decrease, with steam dump 200
5. Steady state fluctuations, initial / random 105E5 I 3.0E6
6. Feedwater Cycling at Hot Shutdown 2000
7. Loop Out of Service, shutdown / startup 80/70
8. Unit loading and unloading between 0% and 15% of full power
9. Boron Concentration Equalization 26400
10. Refueling 80 Upset Conditions
1. Loss of load, without immediate turbine or reactor trip 200
2. Loss of power (blackout with natural circulation in the RCS) 40
3. Loss of flow (partial loss of flow, one pump only) 80
4. Reactor trip

- No cooldown 230

- Cooldown, no safety injection 160

- Cooldown with SI 10

5. Inadvertent RCS depressurization 20
6. Inadvertent startup of an inactive loop 10
7. Control rod drop 80
8. Inadvertent Safety Injection 60 Excessive Feedwater Flow 30 8

Enclosure I Responses to NRC Requests for Additional Information Dated November 4, 2008 Test Conditions

1. Turbine roll test 80
2. Primary side hydrostatic test 10 1 3. Primary Side Leakage Test 280 primary/secondary
  • The 60-year number of transients is 1.5 times the 40-year number As indicated by the footnote, the 60-year fatigue crack growth analysis assumed a 50%

increase in the number of design transients compared to the standard 40-year set. The numbers of design cycles and transients assumed in the WCAP-15338 analysis bound the numbers of design cycles and transients projected for 60 years of PINGP operation, as reported in Table 4.3-1 of the LRA.

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