ML103190745
| ML103190745 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/19/2010 |
| From: | Bennett Brady License Renewal Projects Branch 1 |
| To: | Public Service Enterprise Group |
| BRADY B, NRR/DLR/RPB1, 415-2981 | |
| References | |
| Download: ML103190745 (9) | |
Text
~p.f\ REG(Jl UNITED STATES ro+~p-.; :0",.1. NUCLEAR REGULATORY COMMISSION
~ C'I WASHINGTON, D.C. 20555-0001
<< 0 lh :
~;; November 19, 2010
~ l'
"'" ~o
- .
LICENSEE: PSEG Nuclear, LLC FACILITY: Hope Creek Generating Station
SUBJECT:
SUMMARY
OF TELEPHONE CONFERENCE CALL HELD ON OCTOBER 21,2010. BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND PSEG NUCLEAR, LLC, CONCERNING HOPE CREEK GENERATING STATION, LICENSE RENEWAL APPLICATION, ENVIRONMENTALLY ASSISTED FATIGUE ANALYSIS The U.S. Nuclear Regulatory Commission and representatives of PSEG Nuclear. LLC (the applicant). and Exelon, held a telephone conference call on October 21,2010, to discuss and clarify the applicant's license renewal application for Hope Creek Generating Station. The telephone conference call was useful in clarifying the intent of the applicant's license renewal application draft supplement.
Enclosure 1 provides a listing of the participants and Enclosure 2 contains a brief summary of the discussion and status of the items. Enclosure 3 contains the applicant's proposed response to the draft request for additional information (D-RAI).
The applicant had an opportunity to comment on this summary.
~#~
Bennett M. Brady, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-354
Enclosures:
- 1. List of participants
- 2. Summary of meeting discussion
- 3. D-RAI response cc w/encls: Distribution via Listserv
TELEPHONE CONFERENCE CALL HOPE CREEK GENERATING STATION LICENSE RENEWAL APPLICATION LIST OF PARTICIPANTS OCTOBER 21,2010 PARTICIPANTS AFFILIATIONS Bennett Brady U.S. Nuclear Regulatory Commission (NRC)
Bo Pham NRC OnYee NRC Allen Hiser NRC John Hufnagel Exelon AI Fulvio Exelon Tom Quintenz Exelon Randal Schmidt PSEG Nuclear, LLC Terry Herrmann ENCLOSURE 1
SUMMARY
OF MEETING ON THE LICENSE RENEWAL APPLICATION SUPPLEMENT FOR HOPE CREEK GENERATING STATION LICENSE RENEWAL APPLICATION ON ENVIRONMENTALLY ASSISTED FATIGUE OCTOBER 21, 2010 The U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of PSEG Nuclear, LLC, (the applicant) held a telephone conference call on October 21, 2010, to discuss and clarify the staff's concern with the locations selected for the environmentally assisted fatigue analysis in the Hope Creek Generating Station license renewal application (LRA).
During the discussion, the applicant discussed the proposed response shown in Enclosure 3.
The applicant's updated analysis resulted in a change in one plant-specific location (Le., Feedwater Line No. AE-036) that should have been used to determine the effects of environmentally assisted fatigue for the feedwater Class 1 piping.
The applicant also stated that it had verified that the data for the locations in LRA Table 4.3.5-1 were correct.
The NRC agreed that the locations that are equivalent to the locations suggested in NUREG/CR-6260 may be limiting and bounding for that set. However, the NRC needs to see an explanation of how one can show that these locations were limiting and bounding for all other plant-specific locations.
The NRC also asked that Table 4.3.5-1 in the LRA be updated.
This item remains a confirmatory item.
ENCLOSURE 2
DRAFT RESPONSE ON CONFIRMATORY ITEM NRC Confirmatory Item 4.3.5.2-1, Pages 1-8 of the Hope Creek Generating Station Safety Evaluation Review:
CI4.3.5.2-1: (Safety Evaluation Review Section 4.3.5 - Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)
License renewal application (LRA) Section 4.3.5 summarizes the evaluation of the environmentally assisted fatigue (EAF) analyses for the period of extended operation. This time-limited aging analysis (TLAA) is based on the analysis in NUREG/CR 6260, '1\pplication 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 of demonstrating 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 was 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 CI 4.3.5.2-1.
Section 4.3.5, on pages 4-31 of the SER further states:
The staff noted that Feedwater Line No. AE-036, Node 200 has a 60-year cumulative usage factor (CUF) of 0.841. However the reactor pressure vessel (RPV) feedwater nozzle (safe end and nozzle forging) and feedwater Class 1 piping (tee on header to RPV Nozzle N4E) which were evaluated for reactor water environmental effects, consistent with NUREG/CR-6260, have lower 60-year CUF values. The staff is unclear whether Feedwater Line No. AE-036, Node 200, should be evaluated for reactor water environmental effects or if it is bounded by the locations that have already been evaluated for reactor water environmental effects. In addition, the staff needs confirmation that the applicant has verified that the estimated 60-year CUF values for LRA Table 4.3.3-1 are still conservative and bounded, as compared to those of components in LRA Table 4.3.5-1, when adjusted for environmental effects. This is identified as Confirmatory Item CI 4.3.5.2-1.
ENCLOSURE 3
- 2
Response
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). 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.
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 Table 4.3.3-1, should have been used to determine the environmentally assisted fatigue for feedwater piping in Table 4.3.5-1, based on using the highest design basis 40-year 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 Generating Station LRA.
As a result of this discovery, a comprehensive review was completed. The review concluded that the other component locations in Table 4.3.5-1 were the plant-specific locations at Hope Creek Generating Station that are the limiting locations and equivalent to those identified in NUREG/CR-6260 using the 40-year design basis CUF. In addition, the review concluded that the values in LRA Table 4.3.3-1 and LRA Table 4.3.5-1 are correct, based on stress report inputs.
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 Node 200. Node 200 is at the terminal end of the piping system where it attaches to the RPV feedwater nozzle safe end. 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 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 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.
Attached is the revised LRA Table 4.3.5-1 showing the necessary changes that were a result of this discovery. Added text is shown in Bold Italics, and deletions are shown with strikethrough text.
The staff 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 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 because the estimated 40 and 60-year CUF
- 3 calculations were based on projected transients which did not assume an aBE occurred consistent with LRA Table 4.3.1-1. Feedwater Line No. AE-035, Node 200, CUF analysis has a higher impact due to the aBE transient than Feedwater Line No. AE-036, Node 200. The NB 3200 analysis of the feedwater nozzle was performed using bounding loads 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-35 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.
Therefore, the estimated 60-year CUF values for NUREG/CR-6260 components in LRA Table 4.3.3-1 are still conservative and bounded, as compared to those components in the revised LRA Table 4.3.5-1, when adjusted for environmental effects.
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) 11) Factor 12) Environmental Multiplier (4)
Effects (3)
Reactor pressure CRD Penetration Drive Stainless Steel 0.0393 0.5615 14.30 vessel shell and Housing lower head CRD Penetration with Alloy 600 0.2765 0.4119 1.49 Excavation Reactor pressure Safe End Stainless Steel 0.1982 2.3810 (5) 12.01 vessel feedwater Nozzle Forging Low Alloy Steel 0.1031 0.8096 7.85 nozzle 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 AlloV Steel 0.0701 0.2457 I 3.51 Core spray line Core Spray Nozzle Low Alloy Steel 0.1063 0.7678 7.22 reactor pressure Core Spray Nozzle Safe Alloy 600 0.0202 0.0301 1.49 vessel nozzle and associated Class 1 End piping Residual heat RHR Supply Piping Stainless Steel 0.0252 0.2105 8.36 removal nozzles and associated Class 1 RHR Supply Piping Carbon Steel 0.0547 0.3551 6.49 piping Feedwater Class 1 +ee SA i=leaEieF te RPV Carbon Steel (),Q.74 g,aag 4.73 piping N0221e N4E 0.1982 0.9375 Terminal End of Piping at Feedwater Nozzle I
Safe End I ~
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 carbonllow 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.
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.
November 19, 2010 LICENSEE: PSEG Nuclear, LLC FACILITY: Hope Creek Generating Station
SUBJECT:
SUMMARY
OF TELEPHONE CONFERENCE CALL HELD ON OCTOBER 21, 2010, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND PSEG NUCLEAR, LLC, CONCERNING HOPE CREEK GENERATING STATION, LICENSE RENEWAL APPLICATION, ENVIRONMENTALLY ASSISTED FATIGUE ANALYSIS The U.S. Nuclear Regulatory Commission and representatives of PSEG Nuclear. LLC (the applicant). and Exelon held a telephone conference call on October 21,2010, to discuss and clarify the applicant's license renewal application for Hope Creek Generating Station. The telephone conference call was useful in clarifying the intent of the applicant's license renewal application draft supplement. provides a listing of the participants and Enclosure 2 contains a brief summary of the discussion and status of the items. Enclosure 3 contains the applicant's proposed response to the draft request for additional information.
The applicant had an opportunity to comment on this summary.
IRA!
Bennett M. Brady, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-354
Enclosures:
- 1. List of participants
- 2. Summary of meeting discussion
- 3. D-RAI response cc w/encls: Distribution via Listserv DISTRIBUTION:
See next page ADAMS Accession No'.. ML103190745 *concurrence via e-mail OFFICE PM:RPB1 :DLR LA:DLR BC:RPB1:DLR PM:RPB1:DLR NAME BBrady IKing* BPham BBrady DATE 11/17/10 11/18/10 11119110 11/19/10 OFFICIAL RECORD COpy
Memorandum to PSEG Nuclear, LLC from B. Brady, dated November 19,2010
SUBJECT:
SUMMARY
OF TELEPHONE CONFERENCE CALL HELD ON OCTOBER 21,2010, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND PSEG NUCLEAR, LLC, CONCERNING THE HOPE CREEK GENERATING STATION, LICENSE RENEWAL APPLICATION DISTRIBUTION:
HARDCOPY:
DLRRF E-MAIL:
PUBLIC RidsNrrDlr Resource RidsNrrDlrRpb1 Resource RidsNrrDlrRpb2 Resource RidsNrrDlrRarb Resource RidsNrrDlrRasb Resource RidsNrrDlrRapb Resource RidsOgcMailCenter Resource BPham BBrady LPerkins REnnis CSanders BHarris, OGC ABurritt, RI RConte, RI MModes, RI DTifft, RI NMcNamara, RI