LR-N10-0440, Response to Confirmatory Item Ci 4.3.5.2-1 Associated with NRC Safety Evaluation Report, Related to License Renewal Application

Response to Confirmatory Item Ci 4.3.5.2-1 Associated with NRC Safety Evaluation Report, Related to License Renewal Application

ML110110426
Person / Time
Site:	Hope Creek
Issue date:	01/06/2011
From:	Davison P Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N10-0440
Download: ML110110426 (10)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC 10 CFR 50 JAN 06 2011 10 CFR 51 10 CFR 54 LR-N 10-0440 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

Response ,to Confirmatory Item Cl 4.3.5.2-1 associated with the NRC Safety Evaluation Report, related to the Hope Creek Generating Station License Renewal Application

References:

1. Letter from Mr. Brian E. Holian, USNRC to Mr. Thomas Joyce, PSEG Nuclear, LLC "SAFETY EVALUATION REPORT RELATED TO THE LICENSE RENEWAL OF HOPE CREEK GENERATING STATION,"

dated September 30, 2010

2. Letter from Mr. Paul J. Davison, PSEG Nuclear to USNRC "PSEG Nuclear, LLC Review of the Safety Evaluation Report with Open Items Associated with the Hope Creek Generating Station License Renewal Application," dated November 15, 2010 In Reference 1, the NRC issued the Safety Evaluation Report with Open Items (Safety Evaluation Report) related to renewal of the operating license for the Hope Creek Generating Station. In Reference 2, PSEG Nuclear, LLC responded to the NRC with the results of its review of the Safety Evaluation Report, including responses to Open Item 01 3.0.3.1.2-1 and Confirmatory Item Cl 3.0.3.1.20-1.

In the Enclosure to this letter, PSEG Nuclear, LLC provides the response to Confirmatory Item Cl 4.3.5.2-1. As a result of this response, commitment 54 is added to the License Renewal Commitment List, as shown on page 8 of the Enclosure.

There are no other new or revised regulatory commitments contained in this letter.

If you have any questions, please contact Mr. Ali Fakhar, PSEG Manager - License Renewal, at 856-339-1646.

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

95-2168 REV. 7/99

Document Control Desk LR-N 10-0440 Page 2 of 2 If you have any questions, please contact Mr. Ali Fakhar, PSEG Manager - License Renewal, at 856-339-1646.

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

Executed on 14 Sincerely, Paul J. aio Vice President, Operations Support PSEG Nuclear LLC

Enclosure:

Response to SER Confirmatory Item Cl 4.3.5.2-1 cc: William M. Dean, Regional Administrator - USNRC Region I B. Brady, Project Manager, License Renewal - USNRC R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector- Hope Creek P. Mulligan, Manager IV, NJBNE L. Marabella, Corporate Commitment Tracking Coordinator P. Duca, Hope Creek Commitment Tracking Coordinator

Enclosure LR-N 10-0440 Page 1 of 8 Enclosure Response to NRC Safety Evaluation Report Confirmatory Item Cl 4.3.5.2-1 Associated with Hope Creek Generating Station License Renewal Application

Enclosure LR-N 10-0440 Page 2 of 8 Confirmatory Item Cl 4.3.5.2-1: (SER Section 4.3.5 - Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)

LRA Section 4.3.5 summarizes the evaluation of the environmentally assisted fatigue (EAF) analyses for the period of extended operation. This TLAA is based on the analysis in NUREG/CR 6260, "Application of NUREG/CR 5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components." The applicant stated that the effects of the reactor coolant system environment on fatigue life were evaluated for certain representative components that are identified in NUREG/CR 6260 for newer vintage General Electric plants. As part of its analysis, the applicant identified plant specific limiting locations per NUREG/CR 6260, and performed EAF calculations using guidance in NUREG/CR 6583, "Effects of LWR Coolant Environments on Fatigue Curves of Carbon and Low Alloy Steels," for components made of carbon and low alloy steels and the guidance of NUREG/CR 5704, "Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels," for components made of austenitic stainless steel. The applicant dispositioned its TLAA for EAF analyses based on the criterion in 10 CFR 54.21(c)(1)(iii), with the intention to demonstration that the effects of aging associated with the analysis will be adequately managed for the period of extended operation.

During its review, the staff was concerned whether the applicant had verified that the limiting location per NUREG/CR 6260 were bounding as compared to other plant-specific locations, (e.g., feedwater line no. AE-036, node 200/130), and requested confirmation from the applicant. This is identified as Confirmatory Item Cl 4.3.5.2-1.

PSEG Response:

Thetresponse includes three subject areas related to NRC Confirmatory Item 4.3.5.2-1.

a. The review of the selection of the limiting NUREG/CR-6260 Feedwater Class 1 Piping location.

b. The review of the values presented in LRA Tables.

c. A commitment to perform additional reviews to confirm the limiting Hope Creek locations per NUREG/CR-6260 are bounding as compared to other plant-specific locations, in response to Confirmatory Item 4.3.5.2-1.

Each of these areas is addressed in separate sections, as shown below:

a. Selection of the Limiting NUREG/CR-6260 Feedwater Class 1 Piping Location In response to this Confirmatory Item, actions have been taken to perform a verification to confirm that the limiting locations evaluated per NUREG/CR-6260 are bounding as compared to other plant specific locations, (e.g. feedwater line no. AE-036, node 200/130) for the Feedwater Class 1 piping. We have determined that one of the plant-specific locations, Feedwater Line No. AE-035 Node 200, is bounding, and should have been used to determine the effects of environmentally assisted fatigue for the Feedwater Class 1 piping instead of AE-036, Node 130.

Enclosure LR-N 10-0440 Page 3 of 8 As part of this verification, a review was conducted of the feedwater piping values of LRA Table 4.3.3-1 and the basis documents which support the table. This review concluded that Feedwater Line No. AE-035 Node 200 instead of Feedwater Line No.

AE-036 Node 130 listed in LRA Table 4.3.3-1 should have been used to determine the environmentally assisted fatigue (EAF) for feedwater piping in LRA Table 4.3.5-1 based on using the highest design basis 40-year cumulative usage factor (CUF). The use of Node 130 in the LRA instead of Node 200 was determined to be caused by an error in the stress report input during the preparation of the calculations for the Hope Creek LRA.

Node 200 is at the terminal end of the piping system where it attaches to the RPV feedwater nozzle safe end.

To address this limiting feedwater piping location for the effects of environmentally assisted fatigue, the feedwater nozzle analysis was reviewed, since it includes the bounding terminal end of the piping. The feedwater nozzle analysis results shown in LRA Table 4.3.5-1 were obtained from an ASME Section III NB-3200 analysis that used a finite element model that included the low alloy steel nozzle forging, the carbon steel safe end (with a stainless steel inlay), and the terminal end of the carbon steel pipe that correlates with AE-035 Node 200. This NB-3200 analysis of the feedwater nozzle was performed using loads bounding all the feedwater nozzles. The finite element model showed that the highest stress location within the nozzle assembly was in the safe end, resulting in the highest fatigue usage. Therefore, the 60-year CUF value of 0.1982 shown for the safe end is also bounding for AE-035 Node 200. By applying the maximum carbon steel Fen multiplier of 4.73 for the feedwater piping, an environmentally-adjusted CUF value of 0.9375 is determined for AE-035 Node 200 that is also bounding for the remainder of the feedwater piping. The treatment of the feedwater safe end as a feedwater piping location is consistent with the treatment of the core spray nozzle safe end as an equivalent core spray piping location.

In Section 4.3.5 of the Draft SER, the staff also noted that Feedwater Line No. AE-036 Node 200 has a higher estimated 60-year CUF of 0.841. Feedwater Line No. AE-036 Node 200 is also a terminal end of the piping system where it attaches to the RPV feedwater nozzle safe end. The node selected with the higher design basis 40 year CUF compared to other Feedwater Class 1 piping locations is Feedwater Line No. AE-035 Node 200, and has a lower estimated 60-year CUF of 0.808. This is because the design basis CUF calculation includes the operating basis earthquake (OBE) transient and the estimated 40 and 60-year CUF calculations were based on projected transients which did not assume an OBE occurred consistent with LRA Table 4.3.1-1. The OBE transient has a higher impact on Feedwater Line No. AE-035 Node 200 CUF analysis than it does on Feedwater Line No. AE-036 Node 200 analysis.

The NB-3200 analysis of the feedwater nozzle was performed using loads bounding all the feedwater nozzles, and therefore bounds Feedwater Line No. AE-035 Node 200 as well as AE-036 Node 200. Similarly, Nodes 315 and 265 for both Feedwater Lines No AE-035 and -036 are the terminal ends of the piping system and are also bounded by the NB-3200 analysis. The terminal end location used in the revised LRA Table 4.3.5-1, is the terminal end location which bounds all feedwater piping locations. The ASME Section III NB-3200 analysis used to assess the environmental effects bounds all Table 4.3.3-1 feedwater terminal end locations.

Enclosure LR-N 10-0440 Page 4 of 8 Shown below is the revised LRA Table 4.3.5-1 showing the necessary changes as a result of this discovery. Added text is shown in Bold Italics, and deletions are shown with strikethrough text.

It should be noted in its response to RAI 4.3-07, in Letter LR-N10-0344 dated September 9, 2010 Hope Creek reported the EAF evaluation of the Alloy 600 locations in LRA Table 4.3.5-1 using the NUREG-6909 methodology. In that transmittal, Hope Creek also committed to use the data and methodology described in NUREG-6909 or later revisions/reports for any future environmental fatigue calculations or updates for Ni-Cr-Fe alloys.

Enclosure LR-N 10-0440 Page 5 of 8 Table 4.3.5-1 Environmental Fatigue Results for HCGS for NUREG/CR-6260 Components 60-Year 60-Year Overall Fatigue Fatigue Usage Environmental NUREG/CR-6260 Equivalent HCGS Material Usage Factor with Fatigue Location Location(s) (1) Factor (2) Environmental Multiplier (4)

Effects (3)

Reactor pressure CRD Penetration Drive Stainless Steel 0.0393 0.5615 14.30 vessel shell and Housing lowerhead CRD Penetration with Alloy 600 0.2765 0.4119 1.49 Excavation Reactor pressure vessel feedwater Safe End Stainless Steel 0.1982 2.3810 (5) 12.01 nozzle Nozzle Forging Low Alloy Steel 0.1031 0.8096 (6) 7.85 Reactor recirculation RHR Return Tee Stainless Steel 0.2405 0.6250 2.60 piping (including RPV inlet nozzle forging Low Alloy Steel 0.1033 0.3589 3.48 inlet and outlet nozzles) RPV outlet nozzle forging Low Alloy Steel .0.0701 0.2457 3.51 Core spray line Core Spray Nozzle Low Alloy Steel 0.1063 0.7678 7.22 reactor pressure vessel nozzle and Core Spray Nozzle Safe Alloy 600 0.0202 0.0301 1.49 associated Class 1 End piping Residual heat RHR Supply Piping Stainless Steel 0.0252 0.2105 8.36 removal nozzles and RHR Supply Piping Carbon Steel 0.0547 0.3551 6.49 associated Class 1 piping Feedwater Class 1 Tee on header to RPV Carbon Steel 0074 023,= 4.73 piping Nle- N4E 0.1982 0.9375 Terminal End of Piping at FeedwaterNozzle Safe End Notes:

1. Locations shown are the bounding locations for HCGS.

2. Revised fatigue usage factors were computed for all of the NUREG/CR-6260 components based on the assumed number of cycles for 60 years of plant operation.

3. Environmental fatigue usage was computed using the methodology of NUREG/CR-6583 (for carbon/low alloy steels) and NUREG/CR-5704 (for stainless steels), as appropriate for the material for each location.

4. Environmental multipliers (Fens) were calculated based on the assumption that Hydrogen Water Chemistry (HWC) conditions exist for 85% of the overall 60-year operating period, and Normal Water Chemistry (NWC) conditions exist for 15% of the overall 60-year operating period. The following dissolved oxygen (DO) conditions were used based on review of historical DO data:

• Feedwater line DO is 31 ppb for pre-HWC and 86 ppb for post-HWC conditions.

• Recirculation line DO is 266 ppb pre-HWC and 57 ppb post-HWC.

• RPV Upper Region DO is 103 ppb pre-HWC and 81 ppb post-HWC.

• RPV Beltline DO is 106 ppb pre-HWC and 4 ppb post-HWC.

• RPV Bottom Head Region DO is 109 ppb pre-HWC and 33 ppb post-HWC.

5. The estimated 60-year CUF with environmental effects exceeds the allowable value of 1.0. As discussed in Section 4.3.5, corrective action will be taken prior to exceeding the environmental assisted fatigue CUF value of 1.0.

6. Rapid cycling effects are included in the 60-year Environmental CUF value, but are not multiplied by the Overall Environmental Multiplier based on the guidance of Section 4.2.6 of MRP-47, Revision 1.

Enclosure LR-N 10-0440 Page 6 of 8

b. Review of LRA Tables As a result of the stress report input error identified for the Feedwater Class 1 piping component location, a comprehensive review of the stress reports supporting the LRA Tables was completed. During this review, the stress report for Reactor Water Clean Up (RWCU) piping system was found to have a second 40-year design CUF value at Node 905, which is higher than that reported in LRA Table 4.3.3-1. RWCU Node 905 is located at the transition between carbon steel piping and stainless steel piping. The higher CUF value of 0.573 is attributed to the evaluation of stainless steel material at this location, in addition to the CUF value of 0.523 for carbon steel already presented in Table 4.3.3-1. The weld material used at this transition is stainless steel.

As a result, LRA Table 4.3.3-1 is updated as shown below with added text, including Note 10, shown in Bold Italics:

Table 4.3.3-1 Fatigue Monitoring Locations for HCGS RCPB Piping Components and Estimated CUFs Design Basis Estimated Estimated Estimated Component 40-Year Allowable CUF as of 40-Year 60-Year Monitoring CUF "I CUF (2) 12/31/07 "1 CUF 14) CUF 14) Technique R,6)

Reactor Water Cleanup Node 570 0.609 1.0 0.309 0.492 0.689 CBF Node 575 0.639 1.0 0.321 0.512 0.717 CBF Node 905 (CS)(1 °) 0.523 1.0 0.268 0.430 0.603 CBF 1

Node 905 (SS)" °) 0.573 1.0 0.307 0.490 0.685 CBF Notes:

10. CS is Carbon Steel materialand SS is Stainless Steel material.

The review concluded the remaining values in LRA Table 4.3.3-1 and LRA Table 4.3.5-1 are correct based on stress report inputs. Other than the Feedwater Class 1 piping, the review also concluded the component locations in Table 4.3.5-1 are the limiting Hope Creek Generating Station plant-specific locations that correlate with the NUREG/CR-6260 components. However, additional plant-specific locations may exist which are more limiting than those considered in NUREG/CR-6260. Therefore, an additional review will be performed as described in the Commitment discussed below.

c. Commitment In Response To Confirmatory Item 4.3.5.2-1 In response to Confirmatory Item 4.3.5.2-1, PSEG will also perform a review of design basis ASME Class 1 fatigue evaluations to determine whether the NUREG/CR-6260 based locations that have been evaluated for the effects of the reactor coolant environment on the fatigue usage are the limiting locations for the Hope Creek plant configuration. If more limiting locations are identified, the most limiting location will be evaluated for the effects of the reactor coolant environment on fatigue usage. If any of the limiting locations consist of nickel alloy, NUREG/CR-6909 methodology for nickel alloy will be used in the evaluation. These additional evaluations will be performed through the Metal Fatigue of Reactor Coolant Pressure Boundary program and the most limiting location will be monitored in accordance with 10 CFR 54.21 (c)(1)(iii).

Enclosure LR-N 10-0440 Page 7 of 8 To address this Confirmatory Item Hope Creek adds commitment 54 to the LRA, Appendix A, A.5 License Renewal Commitment List, as follows. Only the added commitment is shown:

Enclosure LR-N10-0440 Page 8 of 8 Update to License Renewal Commitment List As a result of this RAI response, the commitment discussed above is added to LRA Table A.5, License Renewal Commitment List as commitment number 54, as shown below. This new commitment is displayed in bold, italicized font. Any other actions described in this letter are not regulatory commitments and are described for the NRC staff's information:

A.5 License Renewal Commitment List UFSAR ENHANCEMENT PROGRAM OR COMMITMENT SUPPLEMENT OR TOPIC LOCATION IMPLEMENTATION (LRA APP. A) SCHEDULE 54 Metal Fatigue PSEG will perform a review of N/A Priorto the period Hope Creek of Reactor design basis ASME Code Class I of extended Letter Number Coolant fatigue evaluationsto determine operation. LR-NIO-0440; Pressure whether the NUREGICR-6260 Confirmatory Boundary based locations that have been Item 4.3.5.2-1 evaluated for the effects of the reactor coolantenvironment on fatigue usage are the limiting locations for the Hope Creek plant configuration. If more limiting locations are identified, the most limiting location will be evaluated for the effects of the reactor coolant environment on fatigue usage. If any of the limiting locations consist of nickel alloy, NUREGICR-6909 methodology for nickel alloy will be used in the evaluation.