ML110680115
| ML110680115 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 03/08/2011 |
| From: | Richard Ennis Plant Licensing Branch 1 |
| To: | Chernoff H Plant Licensing Branch 1 |
| Ennis R, NRR/DORL, 415-1420 | |
| References | |
| TAC ME4786 | |
| Download: ML110680115 (6) | |
Text
March 8, 2011 MEMORANDUM TO:
Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation FROM:
Richard B. Ennis, Senior Project Manager /ra/
Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
SUBJECT:
HOPE CREEK GENERATING STATION, DRAFT REQUEST FOR ADDITIONAL INFORMATION (TAC NO. ME4786)
The attached draft request for additional information (RAI) was transmitted on March 8, 2011, to Mr. Paul Duke of PSEG Nuclear LLC (the licensee). This information was transmitted to facilitate an upcoming conference call in order to clarify the licensees amendment request for Hope Creek Generating Station (HCGS) dated September 22, 2010. The proposed amendment would modify the Facility Operating License and Technical Specifications to allow HCGS to operate at a reduced feedwater temperature for purposes of extending the normal fuel cycle.
The amendment would also allow operation with feedwater heaters out-of-service at any time during the operating cycle.
This memorandum and the attachment do not convey or represent an NRC staff position regarding the licensees request.
Docket No. 50-354
Attachment:
Draft RAI
March 8, 2011 MEMORANDUM TO:
Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation FROM:
Richard B. Ennis, Senior Project Manager /ra/
Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
SUBJECT:
HOPE CREEK GENERATING STATION, DRAFT REQUEST FOR ADDITIONAL INFORMATION (TAC NO. ME4786)
The attached draft request for additional information (RAI) was transmitted on March 8, 2011, to Mr. Paul Duke of PSEG Nuclear LLC (the licensee). This information was transmitted to facilitate an upcoming conference call in order to clarify the licensees amendment request for Hope Creek Generating Station (HCGS) dated September 22, 2010. The proposed amendment would modify the Facility Operating License and Technical Specifications to allow HCGS to operate at a reduced feedwater temperature for purposes of extending the normal fuel cycle.
The amendment would also allow operation with feedwater heaters out-of-service at any time during the operating cycle.
This memorandum and the attachment do not convey or represent an NRC staff position regarding the licensee's request.
Docket No. 50-354
Attachment:
Draft RAI DISTRIBUTION PUBLIC ATsirigotis, NRR/DE/EMCB LPL1-2 R/F OHopkins, NRR/DSS/SBPB RidsNrrDorlLpl1-2 Resource SGardocki, NRR/DSS/SBPB RidsNrrDorlDpr Resource JGall, NRR/DSS/SRXB RidsNrrPMREnnis Resource MRazzaque, NRR/DSS/SRXB RLobel, NRR/DSS/SCVB MPanicker, NRR/DSS/SNPB ACCESSION NO.: ML110680115 OFFICE LPL1-2/PM NAME REnnis DATE 03/08/2011 OFFICIAL RECORD COPY
DRAFT REQUEST FOR ADDITIONAL INFORMATION (RAI)
REGARDING PROPOSED LICENSE AMENDMENT OPERATION WITH FINAL FEEDWATER TEMPERATURE REDUCTION AND FEEDWATER HEATERS OUT-OF-SERVIICE HOPE CREEK GENERATING STATION DOCKET NO. 50-354 By application dated September 22, 2010 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML102790111), PSEG Nuclear LLC (PSEG or the licensee) submitted a license amendment request for the Hope Creek Generating Station (HCGS). The proposed amendment would modify the Facility Operating License (FOL) and Technical Specifications (TSs) to allow HCGS to operate at a reduced feedwater temperature for purposes of extending the normal fuel cycle. The amendment would also allow operation with feedwater heaters out-of-service at any time during the operating cycle.
The Nuclear Regulatory Commission (NRC) staff has reviewed the information the licensee provided that supports the proposed amendment and would like to discuss the following issues to clarify the submittal. The following questions are from the Mechanical and Civil Engineering Branch (EMCB):
EMCB-RAI-1 It appears that, as proposed in the license amendment request (LAR), a final feedwater temperature reduction (FFWTR) of 102 °F and 60 °F temperature reduction for feedwater heaters out-of-service (FWHOOS) would reduce the feedwater (FW) design temperature from 431.6 °F to 329.6 °F and from 431.6 °F to 371.6 °F, respectively. Please provide similar information which shows how the minimum operating FW temperature will be reduced by the LAR and verify whether (or not and why) the minimum operating temperatures reflecting the FFWTR and FWHOOS have been utilized in the evaluations described in the LAR.
EMCB-RAI-2 For the structural evaluations performed for the proposed LAR, please provide references for the codes utilized. If different than the original code(s) of construction that the evaluated systems, structures, and components (SSCs) were ordered or built to, please discuss how the later codes have been reconciled to the original codes and address whether the original code allowable values have been utilized for these analyses.
EMCB-RAI-3 Annulus pressurization (AP) loads are determined from breaks postulated in the nuclear steam supply system (NSSS) piping. Attachment 4 to the LAR, GE-Hitachi Nuclear Energy Americas LLC (GEH) report NEDC-33506P1, states in a number of places that, The AP loads remain bounded by the original design basis AP loads. Please specify which pipe breaks the AP loads 1 Attachment 3 to the LAR, GEH report NEDO-33506 (ADAMS Accession No. ML102790106), is a publicly available version of Attachment 4 to the LAR, GEH report NEDC-33506P which is proprietary and non-publicly available.
in the first part of the quote are referring to and which pipe breaks the AP loads in the second part of the quote are referring to.
EMCB-RAI-4 As discussed in Section 4.1 of Attachment 1 to the LAR, GEH issued a 10 CFR Part 21 Safety Information Communication, SC09-01, titled, Annulus Pressurization Loads Evaluation on June 8, 2009. As described by the licensee, SC09-01 identified a potential issue with the methodology that developed the AP loads. As a corrective action, SC09-01 recommended that affected plants review their licensing and design bases in light of the issues presented and consider reevaluating the AP loads to ensure that they are consistent with the plants design and licensing basis. Please provide a list of reference(s) of these reviews and or evaluations and discuss the results and conclusions.
EMCB-RAI-5 In reference to the issues identified in GEH SC09-01, Section 4.1 of Attachment 1 of the LAR states that GEH and PSEG evaluated the effect of FWTR on AP loads. In addition, this section of the LAR states that:
The structural responses and amplified response spectra (ARS) of the RPV
[reactor pressure vessel], reactor internals, piping, biological shield wall (BSW),
and RPV pedestal (drywell inner skirt) were evaluated due to the application of Recirculation Suction Line Break (RSLB) and Feedwater Line Break (FWLB) loads.
These evaluations are further discussed in Attachment 4 to the LAR, NEDC-33506P, which only considers RSLB and FWLB. Part of the issue identified by GEH SC09-01 is that plants had not analyzed all major break locations from their design and licensing basis but only considered the recirculation line and feedwater line breaks.
a)
Please provide a technical justification for the omission of the loads due to breaks in the recirculation discharge, low pressure coolant injection, core spray, and main steam for the AP analyses.
b)
Breaks from the omitted lines mentioned in (a) above are included in the HCGS Updated Final Safety Analysis Report (UFSAR) Section 3.6. Please provide an explanation for the apparent inconsistency between the UFSAR and Section 4.1.3 of Attachment 1 of the LAR which indicates that breaks other than RSLB and FWLB are outside of the existing HCGS design basis.
c)
Please provide a discussion which addresses how the ARS was developed.
EMCB-RAI-6 In reference to RSLB and FWLB, NEDC-33506P Sections 4.1.1 and 4.1.2 state that, the current analysis structural loads on the RPV and reactor internals are bounded by the design basis loads. This statement is not clear. Please explain what the term current analysis structural loads means and whether it includes the proposed 102 °F FFWTR effects.
EMCB-RAI-7 The LAR indicates that the AP loads have been evaluated for their affect on the primary and containment structures, but it only discusses the effects on the RPV, RPV internals, BSW and RPV pedestal.
a)
Have the effects of the AP loads, when considering GEH SC09-01 for the proposed amendment, been evaluated for all of the containment structures including the drywell structure and its attachments?
b)
If the answer to (a) above is affirmative, please discuss the results and conclusions of these evaluations. If not, provide a technical justification for not performing these evaluations.
EMCB-RAI-8 PSEGs letter LR-N10-0356 dated September 20, 2010 (ADAMS Accession No. ML102660024),
provided an RAI response supporting the HCGS license renewal application. Table 4.3.1-2, Fatigue Monitoring Locations for HCGS Reactor Pressure Vessel Components and Estimated CUFs, on page 10 of Enclosure A to the letter shows that the design basis 40-year fatigue cumulative usage factor (CUF) for the shroud support is 0.672. The table also shows that the estimated 60-year CUF for the shroud support is 0.465.
As discussed in Section 5.3 of NEDC-33506P, for the proposed amendment, the fatigue life of the reactor internals (including the shroud support) was evaluated based on 60-years, taking into consideration the proposed plant life extension (PLEX). Please provide additional information to address how the 60-year CUF for the shroud support was determined.
EMCB-RAI-9 List the references which approve the exclusion of the FW Sparger from the licensing renewal evaluation, as implied on pages 5-4 through 5-5 of NEDC-33506P.
EMCB-RAI-10 Page 9 of Attachment 1 of the LAR indicates that an analysis has demonstrated that HCGS operation with the proposed feedwater temperature reductions (FWTR) will not exceed the design limits for the design basis accident (DBA) loss-of-coolant accident (LOCA) peak drywell pressure and temperature. Please state the design limits for the DBA-LOCA peak drywell pressure and temperature and compare these limits with those derived from the analysis for proposed FWTR. In addition, please list the loads considered and governed for the existing design limits and for the FFWTR analysis values.
EMCB-RAI-11 With respect to NEDC-33506P, Section 5.4, Reactor Coolant Pressure Boundary Piping, please discuss how the AP loads have changed based on the proposed amendment. In addition, provide summaries for the reactor coolant pressure boundary pipe stresses and fatigue CUFs and pipe supports/restraints and compare these values to the Code allowable values to support your statement that, The results of those evaluations showed with the change in AP loads, the stresses on the piping, supports, and restraints will continue to meet the applicable ASME Code requirements.
EMCB-RAI-12 NEDC-33506P, Section 5.2, discusses flow-induced and acoustic loads resulting on key RPV internals for the FWTR of 102 oF and states that The acoustic and flow-induced loads will be used for further structural evaluation. Please provide stress and fatigue summaries compared to Code allowable values which demonstrate how these loads have affected the structural integrity of the RPV internal structures. In addition, please explain why these loads have not been developed for the FWHOOS condition.
EMCB-RAI-13 Please discuss the limiting conditions/events that bound the FFWTR and FWHOOS conditions for the system cycling fatigue usage for the FW nozzles and FW piping.
EMCB-RAI-14 NEDC-33506P, Section 6.5, shows the 40-year rapid cycling fatigue usage for the FW nozzle safe end for the FFWTR duty, the FWHOOS duty and the original duty. For the FFWTR and FWHOOS duties, a newer ASME Code (2001 with 2003 Addenda) has been utilized for the values of the instantaneous coefficient of thermal expansion in the fatigue evaluation.
a)
Please explain why the original design basis Code was not used for the fatigue evaluation and provide a technical justification which reconciles the newer Code to the original Code values used for the FW nozzles.
b)
Please describe the methodology used in the NEDC-33506P Reference 25 for the feedwater nozzle rapid cycling analysis.
c)
In addition, to the safe end values shown in NEDC-33506P please provide 40-year and 60-year max CUF values for the FW nozzle safe end, blend radius area and the FW sparger. Include system cycling and rapid cycling.
EMCB-RAI-15 For FFWTR and FWHOOS please provide the following:
a)
Cycle summaries for the FFWTR and FWHOOS which show the plant life associated permissible frequencies of occurrence. Also, if different, please provide the number of cycles used for the fatigue evaluations.
b)
Discuss the permissible number of days per year that the plant can operate with the proposed FFWTR and FWHOOS and how this was derived.
c)
Also, please address how the FFWTR and FWHOOS limits (temperature and cycles) will be implemented and protected or monitored.