ML070600611

Request for Additional Information Regarding Request for Extended Power Uprate

ML070600611
Person / Time
Site:	Hope Creek
Issue date:	03/02/2007
From:	James Shea Plant Licensing Branch III-2
To:	Levis W Public Service Enterprise Group
SHea J, NRR/DORL, 415-1388
References
TAC MD3002
Download: ML070600611 (8)

Text

March 2, 2007 Mr. William Levis Senior Vice President & Chief Nuclear Officer PSEG Nuclear LLC-X04 Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION - REQUEST FOR ADDITIONAL INFORMATION REGARDING REQUEST FOR EXTENDED POWER UPRATE (TAC NO. MD3002)

Dear Mr. Levis:

By letter dated September 18, 2006 (Agencywide Documents and Management System (ADAMS) Accession No. ML062680451), as supplemented on October 10, 2006 (Accession No. ML062920092), October 20, 2006 (Accession No. ML063110164), February 14, 2007 (Accession No. ML070530099), and February 16, 2007 (Accession No. ML070590178), PSEG Nuclear, LLC submitted an amendment request for an extended power uprate for Hope Creek Nuclear Generating Station. The proposed amendment would increase the authorized maximum power level by approximately 15 percent, from 3339 megawatts thermal (MWt) to 3840 MWt.

The Nuclear Regulatory Commission (NRC) staff has been reviewing the submittal and has determined that additional information is needed to complete its review. The specific questions are found in the enclosed request for additional information. The questions were sent by e-mail to you on February 13, 2007 (Accession No. ML070540508), to ensure that the questions were understandable, the regulatory basis was clear and to determine if the information was previously docketed. In subsequent discussions with your staff, questions were revised for further clarification or deleted. Mr. Paul Duke of your staff agreed to respond within 30 days from the date of this letter for all the enclosed questions.

Please note that if you do not respond to this letter within the prescribed response times or provide an acceptable alternate date in writing, we may reject your application for amendment under the provisions of Title 10 of the Code of Federal Regulations, Section 2.108. If you have any questions, I can be reached at (301) 415-1388.

Sincerely,

/ra/

James J. Shea, Project Manager Project Directorate I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosure:

As Stated cc w/encl: See next page

ML070600611 OFFICE PDI-2/PM PDI-1/LA APLA/BC EICB/BC SBWB/BC PD1-2/BC NAME JShea SLittle MRubin AHowe GCramston HChernoff DATE 2/28/07 3/1/07 2/09/07 2/08/07 2/07/07 3/2/07 Hope Creek Generating Station cc:

Mr. Michael P. Gallagher Jeffrie J. Keenan, Esquire Vice President - Eng/Tech Support PSEG Nuclear - N21 PSEG Nuclear P.O. Box 236 P.O. Box 236 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Ms. R. A. Kankus Mr. Michael Brothers Joint Owner Affairs Vice President - Nuclear Assessments Exelon Generation Company, LLC PSEG Nuclear Nuclear Group Headquarters KSA1-E P.O. Box 236 200 Exelon Way Hancocks Bridge, NJ 08038 Kennett Square, PA 19348 Mr. George P. Barnes Lower Alloways Creek Township Site Vice President - Hope Creek c/o Mary O. Henderson, Clerk PSEG Nuclear Municipal Building, P.O. Box 157 P.O. Box 236 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Dr. Jill Lipoti, Asst. Director Mr. George H. Gellrich Radiation Protection Programs Plant Support Manager NJ Department of Environmental PSEG Nuclear Protection and Energy P.O. Box 236 CN 415 Hancocks Bridge, NJ 08038 Trenton, NJ 08625-0415 Mr. Michael J. Massaro Brian Beam Plant Manager - Hope Creek Board of Public Utilities PSEG Nuclear 2 Gateway Center, Tenth Floor P.O. Box 236 Newark, NJ 07102 Hancocks Bridge, NJ 08038 Regional Administrator, Region I Ms. Christina L. Perino U.S. Nuclear Regulatory Commission Director - Regulatory Assurance 475 Allendale Road PSEG Nuclear - N21 King of Prussia, PA 19406 P.O. Box 236 Hancocks Bridge, NJ 08038 Senior Resident Inspector Hope Creek Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038

REQUEST FOR ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION CHANGES FOR EXTENDED POWER UPRATE HOPE CREEK GENERATING STATION DOCKET NO. 50-354 By letter dated September 18, 2006 (Agencywide Documents and Management System (ADAMS) Accession No. ML062680451), as supplemented on October 10, 2006 (Accession No. ML062920092), October 20, 2006 (Accession No. ML063110164), February 14, 2007 (Accession No. ML070530099), and February 16, 2007 (Accession No. ML070590178), PSEG Nuclear, LLC submitted an amendment request for an extended power uprate (EPU) for Hope Creek Nuclear Generating Station. The proposed amendment would increase the authorized maximum power level by approximately 15%, from 3339 megawatts thermal (MWt) to 3840 MWt.

The Nuclear Regulatory Commission (NRC) staff has been reviewing the submittal and has determined that additional information is needed to complete its review.

9) PRA Licensing Branch (APLA) 9.1 Based on the Hope Creek Power Uprate Safety Analysis Report (PUSAR), Section 10.5, Pages 10-9 and 10-10: The NRC staff infers that a complete Level 2 probabilistic risk assessment (PRA) exists for the constant pressure power uprate (CPPU) plant and the current licensed thermal power (CLTP) plant. The NRC staff observes that a complete Level 2 PRA is different (i.e., more detailed) than a simplified PRA model used to estimate large early release frequency (LERF), e.g., NUREG/CR-6595. Please confirm that the NRC staffs inference is correct. If the NRC staffs inference is correct, please provide a summary of the Level 2 PRA results for both the CPPU and CLTP plants that includes a breakdown by release type (LERF, large late releases, core-damage events that do not result in any release, etc.).

9.2 In the Hope Creek PUSAR, Section 10.5, Pages 10-9 and 10-10: It is stated that the change in LERF is primarily due to the change in core damage frequency (CDF). Please provide the definition of LERF used in the PRA, specifically discussing the distinction between an early release and a late release. In addition, confirm that none of the late releases were reclassified as early releases as a result of the proposed EPU.

9.3 In the Hope Creek PUSAR, Section 10.5, Page 10-13: It is stated that the proposed power uprate would increase the reactor thermal power from 3339 MWt to 3840 MWt, which is approximately a 15% increase in thermal power. However, it is further stated that the CPPU PRA is based on an assumed 20% increase in thermal power. In addition, Page 10-20 and Table 10-10 indicate that calculations performed to estimate the timing of some operator actions were based on a decay heat that is 12.3% greater ENCLOSURE

than original licensed thermal power (OLTP). Please explain why different thermal power levels were used as inputs to the PRA. Justify the use of the 12.3% increase in decay heat, which is lower than the proposed EPU and, therefore, non-conservative.

9.4 In the Hope Creek PUSAR, Section 10.5.3, Page 10-19: It is stated that ...changes in the response of the SACS system (the intermediate safety system cooling loops) were evaluated as they influence crew actions. These changes are not described in Pages 10-11 through 10-13. Please describe what changes have been (or will be) made to the SACS system, and how these changes have been reflected in the PRA.

9.5 In the Hope Creek PUSAR, Section 10.5.3, Page 10-20: It is stated that, in general, the cognitive portions of the post-initiator human error probabilities (HEPs) were estimated using the Cause-Based Decision Tree Method (CBDTM). However, it is further stated that some post-initiator HEPs were estimated using a combination of the CBDTM and the Accident Sequence Evaluation Program (ASEP) time reliability correlation. What criteria or guidelines were used to determine the appropriate human reliability quantification method to be used for each HEP?

9.6 In the Hope Creek PUSAR, Section 10.5.3, Page 10-20: What method was used to estimate the implementation portion of the post-initiator HEPs?

9.7 Please augment Table 10-10 page 10-50 of the Hope Creek PUSAR to include the following information:

a) The HEPs for the OLTP plant and the CPPU plant, b) The human reliability quantification method that was used (e.g., CBDTM or a combination of CBDTM and the ASEP time reliability correlation), and c) The risk achievement worth (RAW) of the human action for the CPPU plant, as determined from the CDF calculation. (Note: The NRC staff will use this information, along with the previous reported Fussell-Vesely importance measures, to determine the appropriate amount of review to perform in accordance with NUREG/CR-1764, Guidance for the Review of Changes to Human Actions.)

9.8 In the Hope Creek PUSAR, Sections 10.5.5.1 and 10.5.5.2, Pages 10-23 through 10-25:

The NRC staff understands that the seismic PRA and the Fire Induced Vulnerability Evaluation (FIVE), which were performed as part of the Individual Plant Examination -

External Events (IPEEE), have not been updated to reflect the Revision 2005B PRA model. Confirm that the changes made to the PRAs logic model since the IPEEE was submitted do not significantly affect the IPEEE conclusions concerning seismic and internal fire risk.

9.9 In the Hope Creek PUSAR, Section 10.5.7.2, Pages 10-31 and 10-32, and Figure 10-2:

It is stated that a self-assessment of PRA quality was performed against the American Society of Mechanical Engineers (ASME) PRA standard. Please provide documentation of the self-assessment. Which addendum to the original ASME PRA standard was considered during the self-assessment? Were the NRC staffs clarifications and

qualifications to the ASME PRA standard, which are provided in Appendix A of Regulatory Guide (RG) 1.200, incorporated into the PRA quality self-assessment process? Note: The NRC staff understands that the request for EPU is not risk-informed, that Revision 0 of RG 1.200 was in effect when the request for EPU was made, and that Revision 0 to RG 1.200 was only issued for trial use. The intent of the above questions is to help determine whether or not the PRA has sufficient technical adequacy to support the EPU application, specifically whether or not an onsite audit of the PRA is warranted.

9.10 Please provide a parametric uncertainty analysis of the OLTP CDF and the CPPU CDF.

10) Instrumentation & Controls Branch (EICB) 10.1 The license amendment request (LAR) proposes Technical Specifications (TS) changes associated with instrument set point(s) for the EPU, please provide the following for each set point to be added or modified:

a) Setpoint Calculation Methodology: Provide documentation (including sample calculations) of the methodology used for establishing the limiting nominal set point and the limiting acceptable values for the As-Found and As-Left set points as measured in periodic surveillance testing as described below. Indicate the related Analytical Limits and other limiting design values (and the sources of these values) for each set point.

b) For set points that are not determined to be Safety Limit (SL)-related: Describe the measures to be taken to ensure that the associated instrument channel is capable of performing its specified safety functions in accordance with applicable design requirements and associated analyses. Include in your discussion information on the controls you employ to ensure that the as left trip setting after completion of periodic surveillance is consistent with your set point methodology.

If the controls are located in a document other than the TS (e.g., plant test procedure), describe how it is ensured that the controls will be implemented.

10.2 Provide the justification for removal of Turbine First Stage Pressure from the TSs. This justification should be based on how this instrumentation function does not meet the four criteria provided in Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(2)(ii).

10.3 Section 5.1 of the NRC staffs safety evaluation of General Electric Nuclear Energy Licensing Topical Report NEDC-33004P, Constant Pressure Power Uprate, dated March 31, 2003, require that a plant-specific submittal should address all CPPU related changes to instrumentation & controls, such as scaling changes, changes to upgrade obsolescent instruments and changes to control philosophy. Provide this information for staffs review.

3) BWR Systems Branch (SBWB)

The NRC staff plans to perform a limited set of audit calculations for Hope Creek Chapter 15 safety analyses at the proposed increased power rating using the RELAP5 computer code. The

computer model for a Boiling Water Reactor (BWR) type 4 plant will be modified to represent a mixed core loss-of-coolant accident (LOCA) analysis for Hope Creek. In order to enable the NRC staff to adequately perform this task, please provide the following information:

3.47 For postulated large and small recirculation line LOCAs for Cycle 15 (initial EPU core),

please provide and justify the limiting axial power shapes employed in the Appendix K evaluation determining PCT. For different exposures, select bundles with limiting axial power peaking operating with bottom peaked, double-hump or mid-peaked, and top peaked axial power distributions. Provide the peak fuel bundle to average fuel bundle power ratio (radial peaking factor). Provide the peak fuel rod to peak bundle power ratio (local peaking factor). Provide average and hot bundle exit void fraction.

Please provide above information for General Electric (GE14) fuel and Westinghouse (SVEA-96+) nuclear fuel. For SVEA-96+ fuel you alternatively could provide detailed justification demonstrating that the fuel would not be limiting in regards to peak cladding temperatures (PCT) in your LOCA analysis for Cycle 15. In addition, a) for SVEA-96+ fuel, if determined to be PCT limiting, provide the following information:

Fuel rod diameter for an average and a hot rod Cladding thickness Gap gas mole fractions for an average and a hot rod Gap thickness for an average and a hot rod Gap internal pressure for an average rod and for the hot rod gap conductance Cladding heat capacity vs temperature Cladding thermal conductivity vs temperature Fuel heat capacity vs temperature Fuel thermal conductivity vs temperature Channel box heat capacity vs temperature Channel box thermal conductivity vs temperature Temperature distribution within average and hot channels. Temperature distribution within a hot rod.

Channel box dimensions and thickness b) for GE 14 fuel, provide the following information:

Gap gas mole fractions for an average and a hot rod Gap internal pressure for an average rod and for the hot rod gap conductance Temperature distribution within average and hot channels. Temperature distribution within a hot rod.

c) Reactor Kinetics Information as follows:

Total power histories (include GE and SVEA in the mixed core of Cycle 15) after scram in the limiting LOCA analysis.

d) Fuel Bundle Information for GE-14 and SVEA-96+ fuel as follows:

Cross sectional drawing of the fuel bundles showing rod spacing and pitch.

Location of the highest power rod Location and dimensions of water rods e) Fuel Bundle Pressure drop information as follows:

Provide flow loss coefficients as a function of axial height for the GE14 and SVEA-96+ fuel bundles.

3.48 For the maximum power fuel bundles, provide the thermal radiation emissivites and view factors to be used in evaluation of radiation heat transfer during recovery from a LOCA at the CPPU conditions.

3.49 For a postulated recirculation pump suction break, at the CPPU conditions, provide the equivalent heat transfer coefficient for radiation heat transfer as a function of time for the highest temperature location of the hottest fuel rod. This information is contained in Figure B-2g of GE Nuclear Energy, Topical Report, (NEDC-33172), "GE LOCA analysis for Hope Creek EPU," but the figure is difficult to read. Please provide a more legible figure.

3.50 For a postulated recirculation pump suction break at the CPPU conditions, provide a graph of drywell pressure as a function of time.

3.51 Figures B-2e and B-5e of NEDC-33172 provide the Emergency Core Cooling (ECC) flows for the limiting large and small break sizes at the CPPU conditions. The figures do not distinguish how much Low Pressure Core Injection (LPCI) flow reaches each recirculation loop. Please provide this information. In addition, provide LPCI and High Pressure Core Injection (HPCI) head-flow curves assumed in the LOCA analyses.

Provide the capacity of the Automatic Depressurization System (ADS) valves assumed in the analyses in pounds mass per hour (lbs/hr) and pounds per square inch absolute (psia).

3.52 Provide the sequence events table for the Appendix K limiting Design Basis Accident large-break and small-break LOCAs at the CPPU conditions. They should identify all trip signals and delays such as reactor scram and Emergency Core Cooling Systems injection.

3.53 Provide the reactor vessel level setpoints used for reactor scram, ADS, Core spray, HPCI and LPCI in terms of height above the core at the CPPU conditions.

3.54 NEDC-33172 provides the results of LOCA analyses for Hope Creek at the uprate power level for a mixed core of GE14 and SVEA-96+ fuel. Please justify that the fuel burnup and power peaking assumed in these analyses for both fuel types bound those which will be experienced for cycle 14 of Hope Creek.

3.55 Provide a table of steady state initial conditions at the CPPU conditions. The table should include reactor power, reactor pressure, water level in the RPV, total core mass flow, feedwater flow, steam flow, recirculation flow rates, core inlet temperatures, etc.

3.56 Question Deleted.