|
|
(2 intermediate revisions by the same user not shown) |
Line 2: |
Line 2: |
| | number = ML070600611 | | | number = ML070600611 |
| | issue date = 03/02/2007 | | | issue date = 03/02/2007 |
| | title = Hope Creek Generating Station - Request for Additional Information Regarding Request for Extended Power Uprate | | | title = Request for Additional Information Regarding Request for Extended Power Uprate |
| | author name = Shea J J | | | author name = Shea J |
| | author affiliation = NRC/NRR/ADRO/DORL | | | author affiliation = NRC/NRR/ADRO/DORL |
| | addressee name = Levis W | | | addressee name = Levis W |
Line 9: |
Line 9: |
| | docket = 05000354 | | | docket = 05000354 |
| | license number = | | | license number = |
| | contact person = SHea J J, NRR/DORL, 415-1388 | | | contact person = SHea J, NRR/DORL, 415-1388 |
| | case reference number = TAC MD3002 | | | case reference number = TAC MD3002 |
| | document type = Letter, Request for Additional Information (RAI) | | | document type = Letter, Request for Additional Information (RAI) |
Line 18: |
Line 18: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:March 2, 2007Mr. William LevisSenior Vice President & Chief Nuclear Officer PSEG Nuclear LLC-X04 Post Office Box 236 Hancocks Bridge, NJ 08038 | | {{#Wiki_filter: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:== | | ==SUBJECT:== |
| HOPE CREEK GENERATING STATION - REQUEST FOR ADDITIONALINFORMATION REGARDING REQUEST FOR EXTENDED POWER UPRATE (TAC NO. MD3002) | | HOPE CREEK GENERATING STATION - REQUEST FOR ADDITIONAL INFORMATION REGARDING REQUEST FOR EXTENDED POWER UPRATE (TAC NO. MD3002) |
|
| |
|
| ==Dear Mr. Levis:== | | ==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 hasdetermined 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 orprovide 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 haveany questions, I can be reached at (301) 415-1388. Sincerely,/ra/James J. Shea, Project ManagerProject Directorate I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket 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 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:== | | ==Enclosure:== |
| As Stated cc w/encl: See next page | | As Stated cc w/encl: See next page |
|
| |
|
| ML070600611OFFICEPDI-2/PMPDI-1/LAAPLA/BCEICB/BC SBWB/BCPD1-2/BCNAMEJSheaSLittleMRubinAHoweGCramstonHChernoffDATE2/28/073/1/072/09/072/08/072/07/073/2/07 Hope Creek Generating Station cc:
| | 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. GallagherVice President - Eng/Tech Support PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038Mr. Michael BrothersVice President - Nuclear Assessments PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038Mr. George P. BarnesSite Vice President - Hope Creek PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038Mr. George H. GellrichPlant Support Manager PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038Mr. Michael J. MassaroPlant Manager - Hope Creek PSEG Nuclear P.O. Box 236 Hancocks Bridge, NJ 08038Ms. Christina L. PerinoDirector - Regulatory Assurance PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038Jeffrie J. Keenan, EsquirePSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038Ms. R. A. KankusJoint Owner Affairs Exelon Generation Company, LLC Nuclear Group Headquarters KSA1-E 200 Exelon Way Kennett Square, PA 19348Lower Alloways Creek Townshipc/o Mary O. Henderson, Clerk Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038Dr. Jill Lipoti, Asst. DirectorRadiation Protection Programs NJ Department of Environmental Protection and Energy | | 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 |
|
| |
|
| CN 415 Trenton, NJ 08625-0415Brian BeamBoard of Public Utilities 2 Gateway Center, Tenth Floor Newark, NJ 07102Regional Administrator, Region IU.S. Nuclear Regulatory Commission
| | 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 |
|
| |
|
| 475 Allendale Road King of Prussia, PA 19406Senior Resident InspectorHope Creek Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038 ENCLOSUREREQUEST FOR ADDITIONAL INFORMATIONREGARDING TECHNICAL SPECIFICATION CHANGES FOREXTENDED POWER UPRATEHOPE CREEK GENERATING STATIONDOCKET NO. 50-354By letter dated September 18, 2006 (Agencywide Documents and Management System(ADAMS) Accession No. ML062680451), as supplemented on October 10, 2006 (Accession No.
| | 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. |
| 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 hasdetermined that additional information is needed to complete its review. 9)PRA Licensing Branch (APLA) 9.1Based 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 staff's inference is correct. If the NRC staff's 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.).9.2In the Hope Creek PUSAR, Section 10.5, Pages 10-9 and 10-10: It is stated that thechange 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.9.3In the Hope Creek PUSAR, Section 10.5, Page 10-13: It is stated that the proposedpower 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 than original licensed thermal power (OLTP). Please explain why different thermal powerlevels 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.9.4In the Hope Creek PUSAR, Section 10.5.3, Page 10-19: It is stated that "...changes inthe 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.9.5In the Hope Creek PUSAR, Section 10.5.3, Page 10-20: It is stated that, in general, thecognitive 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?9.6In the Hope Creek PUSAR, Section 10.5.3, Page 10-20: What method was used toestimate the implementation portion of the post-initiator HEPs?9.7Please 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., CBDTMor a combination of CBDTM and the ASEP time reliability correlation), andc)The risk achievement worth (RAW) of the human action for the CPPUplant, 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.")9.8In 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 -
| | 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. |
| External Events (IPEEE), have not been updated to reflect the Revision 2005B PRA model. Confirm that the changes made to the PRA's logic model since the IPEEE was submitted do not significantly affect the IPEEE conclusions concerning seismic and internal fire risk.9.9In 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 staff's clarifications and qualifications to the ASME PRA standard, which are provided in Appendix A ofRegulatory 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.9.10Please provide a parametric uncertainty analysis of the OLTP CDF and the CPPU CDF.10)Instrumentation & Controls Branch (EICB) 10.1The license amendment request (LAR) proposes Technical Specifications (TS) changesassociated 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 samplecalculations) 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: Describethe 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.
| | 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? |
| 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. 10.2Provide the justification for removal of Turbine First Stage Pressure from the TSs. Thisjustification 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).10.3Section 5.1 of the NRC staff's safety evaluation of General Electric Nuclear EnergyLicensing 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 staff's review.3) BWR Systems Branch (SBWB)The NRC staff plans to perform a limited set of audit calculations for Hope Creek Chapter 15safety 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 amixed 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:3.47For 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 rodCladding 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.
| | 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? |
| Channel box dimensions and thicknessb)for GE 14 fuel, provide the following information:Gap gas mole fractions for an average and a hot rodGap 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.
| | 9.7 Please augment Table 10-10 page 10-50 of the Hope Creek PUSAR to include the following information: |
| 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 rodse)Fuel Bundle Pressure drop information as follows:Provide flow loss coefficients as a function of axial height for the GE14 andSVEA-96+ fuel bundles.3.48For the maximum power fuel bundles, provide the thermal radiation emissivites and viewfactors to be used in evaluation of radiation heat transfer during recovery from a LOCA at the CPPU conditions.3.49For a postulated recirculation pump suction break, at the CPPU conditions, provide theequivalent 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.3.50For a postulated recirculation pump suction break at the CPPU conditions, provide agraph of drywell pressure as a function of time.3.51Figures 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 HighPressure Core Injection (HPCI) head-flow curves assumed in the LOCA analyses.
| | 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.) |
| 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).3.52Provide the sequence events table for the Appendix K limiting Design Basis Accidentlarge-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. 3.53Provide the reactor vessel level setpoints used for reactor scram, ADS, Core spray, HPCIand LPCI in terms of height above the core at the CPPU conditions.3.54NEDC-33172 provides the results of LOCA analyses for Hope Creek at the uprate powerlevel 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.3.55Provide a table of steady state initial conditions at the CPPU conditions. The tableshould 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. 3.56Question Deleted.}}
| | 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.}} |
Letter Sequence RAI |
---|
TAC:MD3002, Control Room Habitability (Approved, Closed) |
Initiation
- Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request
- Acceptance, Acceptance
- Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement
Administration
- Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance
- Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting
|
MONTHYEARML0733004532003-11-27027 November 2003 Schedule and Response to PSEG Letter (LR-N07-0274) Dated October 23, 2007, Plans Related to Steam Dryer Evaluation Regarding Request for Extended Power Uprate Project stage: Other ML0626900642006-09-0101 September 2006 Attachment 26, C.D.I. Report No. 06-16NP, Rev 1, Estimating High Frequency Flow Induced Vibration in the Main Stream Lines at Hope Creek Unit 1: a Subscale Four Line Investigation of Standpipe Behavior. Project stage: Request LR-N06-0286, Request for License Amendment Extended Power Uprate2006-09-18018 September 2006 Request for License Amendment Extended Power Uprate Project stage: Request ML0626900442006-09-30030 September 2006 Attachment 21, C.D.I. Report No. 06-27, Rev 0, Stress Analysis of Hope Creek Unit 1 Steam Dryer at CLTP and EPU Conditions Using 1/8th Scale Model Pressure Measurement Data. Project stage: Request ML0631101672006-09-30030 September 2006 Attachment 2 - CDI Technical Memo 06-23P (Non-Proprietary) Comparison of Hope Creek and Quad Cities Steam Dryer Loads at EPU Conditions, Revision 0, Dated September 2006 Project stage: Request LR-N06-0413, Supplement to License Amendment Request for Extended Power Uprate2006-10-10010 October 2006 Supplement to License Amendment Request for Extended Power Uprate Project stage: Supplement ML0628303692006-10-13013 October 2006 Supplement to Application for Extended Power Uprate Project stage: Other ML0629003222006-10-18018 October 2006 Extended Power Uprate Accpetance Review Results Project stage: Other LR-N06-0418, Supplement to License Amendment Request for Extended Power Uprate, to Increase the Maximum Authorized Power Level to 3840 Megawatts Thermal2006-10-20020 October 2006 Supplement to License Amendment Request for Extended Power Uprate, to Increase the Maximum Authorized Power Level to 3840 Megawatts Thermal Project stage: Supplement ML0628304722006-11-0808 November 2006 PSEG Nuclear LLC, Withholding from Public Disclosure, NEDC-33076P, Revision 2, Safety Analysis Report for Hope Creek Constant Pressure Power Uprate, Class III, MD3002 Project stage: Other ML0628304532006-11-21021 November 2006 PSEG Nuclear LLC, Request for Withholding Information from Public Disclosure for Hope Creek Generating Station Project stage: Withholding Request Acceptance ML0633301432006-12-15015 December 2006 11/13-14/2006 Summary of Category 1 Meeting with PSEG Nuclear LLC, Regarding Application for Extended Power Uprate for Hope Creek Project stage: Meeting ML0703002772007-01-26026 January 2007 E-Mail Shea-NRR,to Duke, PSEG, Group1 Draft EPU RAI Project stage: Draft Other ML0716303072007-01-31031 January 2007 C.D.I. Technical Note No. 07-01NP, EPU Conditions in the Main Steam Lines at Hope Creek Unit 1: Additional Subscale Four Line Tests. Project stage: Request LR-N07-0034, C.D.I. Technical Memorandum No. 06-23NP, Revision 1, Comparison of the Hope Creek and Quad Cities Steam Dryer Loads at EPU Conditions.2007-01-31031 January 2007 C.D.I. Technical Memorandum No. 06-23NP, Revision 1, Comparison of the Hope Creek and Quad Cities Steam Dryer Loads at EPU Conditions. Project stage: Request LR-N07-0026, Supplemental to License Amendment Request for Extended Power Uprate2007-02-14014 February 2007 Supplemental to License Amendment Request for Extended Power Uprate Project stage: Supplement LR-N07-0029, Supplement to License Amendment Request for Extended Power Uprate2007-02-16016 February 2007 Supplement to License Amendment Request for Extended Power Uprate Project stage: Supplement ML0703304152007-02-16016 February 2007 Request for Additional Information Regarding Request for Extended Power Uprate Project stage: RAI ML0706606192007-02-20020 February 2007 Draft Request for Additional Information Hope Creek EPU Grp 5 Project stage: Draft RAI ML0706601992007-02-20020 February 2007 Draft Request for Additional Information Hope Creek EPU Grp 4 Project stage: Draft RAI ML0704602432007-02-23023 February 2007 Ltr Request for Additional Information Related to the Request for Extended Power Uprate Project stage: RAI ML0706606382007-02-23023 February 2007 Draft Request for Additional Information Hope Creek EPU Grp 6 Project stage: Draft RAI ML0706803142007-02-28028 February 2007 Hope Creek, Supplement to License Amendment Request for Extended Power Uprate Project stage: Supplement ML0708103652007-02-28028 February 2007 C.D.I. Report No. 07-01NP, Rev. 0, Revised Hydrodynamic Loads on Hope Creek Unit 1 Steam Dryer to 200 Hz. Project stage: Request ML0706006112007-03-0202 March 2007 Request for Additional Information Regarding Request for Extended Power Uprate Project stage: RAI ML0709200252007-03-0202 March 2007 Slides for Summary of March 2, 2007 Meeting with PSEG Nuclear LLC on an Application for Extended Power Uprate for Hope Creek Generating Station Regarding Steam Dryer Margin Project stage: Meeting ML0706803062007-03-13013 March 2007 Request for Additional Information Regarding Request for Extended Power Uprate Project stage: RAI LR-N07-0035, Response to Request for Additional Information Request for License Amendment - Extended Power Uprate2007-03-13013 March 2007 Response to Request for Additional Information Request for License Amendment - Extended Power Uprate Project stage: Response to RAI LR-N07-0055, Supplement to License Amendment Request for Extended Power Uprate2007-03-13013 March 2007 Supplement to License Amendment Request for Extended Power Uprate Project stage: Supplement ML0708004302007-03-29029 March 2007 Summary of March 2, 2007, Meeting with PSEG Nuclear, LLC, on an Application for Extended Power Uprate for Hope Creek Generating Station Regarding Steam Dryer Margin Project stage: Meeting LR-N07-0060, Response to Request for Additional Information Request for License Amendment - Extended Power Uprate2007-03-30030 March 2007 Response to Request for Additional Information Request for License Amendment - Extended Power Uprate Project stage: Response to RAI LR-N07-0069, Response to Request for Additional Information Request for License Amendment - Extended Power Uprate2007-03-30030 March 2007 Response to Request for Additional Information Request for License Amendment - Extended Power Uprate Project stage: Response to RAI LR-N07-0070, Response to Request for Additional Information on Request for License Amendment - Extended Power Uprate2007-04-13013 April 2007 Response to Request for Additional Information on Request for License Amendment - Extended Power Uprate Project stage: Response to RAI LR-N07-0084, Supplement to License Amendment Request for Extended Power Uprate, Revised No Significant Hazards Consideration2007-04-18018 April 2007 Supplement to License Amendment Request for Extended Power Uprate, Revised No Significant Hazards Consideration Project stage: Supplement ML0711400912007-04-20020 April 2007 Request for Additional Information Regarding Request for Extended Power Uprate (TAC MD3002) - NON-PROPRIETARY Project stage: RAI ML0713603772007-04-30030 April 2007 C.D.I. Report No. 06-16NP, Rev. 2, Estimating High Frequency Flow Induced Vibration in the Main Steam Lines at Hope Creek, Unit 1: a Subscale Four Line Investigation of Standpipe Behavior. Project stage: Request LR-N07-0099, Response to Request for Additional Information Request for License Amendment, Extended Power Uprate2007-04-30030 April 2007 Response to Request for Additional Information Request for License Amendment, Extended Power Uprate Project stage: Response to RAI ML0710104492007-05-0303 May 2007 Explanation of Hope Creek Nuclear Generating Station Extended Power Uprate and Conclusion of Informal Consultation Project stage: Other ML0712404872007-05-0909 May 2007 PSEG Nuclear LLC, Request for Withholding Information from Public Disclosure for Hope Creek Generating Station Project stage: Withholding Request Acceptance LR-N07-0102, Response to Request for Additional Information Request for License Amendment - Extended Power Uprate2007-05-10010 May 2007 Response to Request for Additional Information Request for License Amendment - Extended Power Uprate Project stage: Response to RAI ML0712404112007-05-14014 May 2007 RAI, Request for Additional Information Regarding Request for Extended Power Uprate Project stage: RAI ML0712405052007-05-15015 May 2007 Summary of Meeting with PSEG Nuclear LLC, Regarding Technical Aspects of the Licensee'S Application for an Extended Power Uprate (EPU) at the Hope Creek Generating Station (Hope Creek) Project stage: Meeting ML0712404612007-05-17017 May 2007 PSEG Nuclear LLC, Request for Withholding Information from Public Disclosure for Hope Creek Generating Station Project stage: Withholding Request Acceptance ML0628605302007-05-17017 May 2007 Nuclear LLC, Request for Withholding Information from Public Disclosure for Hope Creek Generating Station Project stage: Withholding Request Acceptance ML0713707002007-05-17017 May 2007 Request for Additional Information (RAI) Regarding Request for Extended Power Uprate Project stage: RAI ML0712405102007-05-17017 May 2007 PSEG Nuclear LLC, Request for Withholding Information from Public Disclosure for Hope Creek Generating Station Project stage: Withholding Request Acceptance LR-N07-0123, Response to Request for Additional Information, Request for License Amendment - Extended Power Uprate2007-05-18018 May 2007 Response to Request for Additional Information, Request for License Amendment - Extended Power Uprate Project stage: Response to RAI LR-N07-0113, Supplement to Request for License Amendment Extended Power Uprate Steam Dryer Limit Curves2007-05-18018 May 2007 Supplement to Request for License Amendment Extended Power Uprate Steam Dryer Limit Curves Project stage: Supplement LR-N07-0122, Supplement to Request for License Amendment Re Extended Power Uprate & C.D.I. Technical Note No. 07-012007-05-24024 May 2007 Supplement to Request for License Amendment Re Extended Power Uprate & C.D.I. Technical Note No. 07-01 Project stage: Supplement ML0717203702007-05-31031 May 2007 Attachment 2, C.D.I. Technical Note No. 07-19NP, Revision 0, Hope Creek, Unit 1, Extended Power Uprate Limit Curve Analysis for Dryer Stress Prediction During Power Ascension Project stage: Request 2007-02-28
[Table View] |
|
---|
Category:Letter
MONTHYEARIR 05000354/20230042024-02-0101 February 2024 Integrated Inspection Report 05000354/2023004 ML24030A8752024-02-0101 February 2024 Operator Licensing Examination Approval ML24009A1022024-01-26026 January 2024 Exemption from Select Requirements of 10 CF Part 73 (EPID L-2023-LLE-0045 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) IR 05000354/20234012024-01-22022 January 2024 Material Control and Accounting Program Inspection Report 05000354/2023401 ML23341A1372024-01-16016 January 2024 Issuance of Amendment No. 235 Revise Trip and Standby Auto-Start Logic Associated with Safety Related Heating, Ventilation and Air Conditioning ML23335A1122023-12-15015 December 2023 Retest Schedule for Drywell to Suppression Chamber Vacuum Breakers ML23307A1532023-12-15015 December 2023 NRC Investigation Report No. 1-2023-001 ML23270C0072023-11-29029 November 2023 Notice of Proposed Amendment to Decommissioning Trust Agreement ML23324A3072023-11-17017 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000354/20230032023-11-0707 November 2023 Integrated Inspection Report 05000354/2023003 IR 05000272/20234022023-10-12012 October 2023 and Salem Nuclear Generating Station, Units 1 and 2 - Security Baseline Inspection Report 05000354/2023402, 05000272/2023402 and 05000311/2023402 (Cover Letter Only) LR-N23-0065, Submittal of 2023 Annual 10 CFR 50.46 Report2023-10-0202 October 2023 Submittal of 2023 Annual 10 CFR 50.46 Report LR-N23-0045, and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement2023-09-0808 September 2023 and Peach Bottom Atomic Power Station, Units 2 and 3 - Notice of Proposed Amendment to Decommissioning Trust Agreement ML23249A2612023-09-0606 September 2023 License Amendment Request to Modify the Salem and Hope Creek Exclusion Area Boundary IR 05000354/20230052023-08-31031 August 2023 Updated Inspection Plan for Hope Creek Generating Station (Report 05000354/2023005) ML23192A8212023-08-14014 August 2023 and Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 234, 347, and 329 Revise Technical Specifications to Delete Meteorological Tower Location IR 05000354/20230022023-08-0303 August 2023 Integrated Inspection Report 05000354/2023002 and Independent Spent Fuel Storage Installation Inspection Report 07200048/2023001 IR 05000354/20230102023-08-0303 August 2023 Biennial Problem Identification and Resolution Inspection Report 05000354/2023010 LR-N23-0052, Retest Schedule for Drywell to Suppression Chamber Vacuum Breakers Per Technical Specification 4.6.2.12023-07-31031 July 2023 Retest Schedule for Drywell to Suppression Chamber Vacuum Breakers Per Technical Specification 4.6.2.1 LR-N23-0042, Spent Fuel Cask Registration2023-07-12012 July 2023 Spent Fuel Cask Registration LR-N23-0046, Emergency Plan Document Revisions Implemented June 28, 20232023-07-10010 July 2023 Emergency Plan Document Revisions Implemented June 28, 2023 IR 05000354/20230112023-05-0101 May 2023 Commercial Grade Dedication Report 05000354/2023011 ML23121A1412023-05-0101 May 2023 Senior Reactor and Reactor Operator Initial License Examinations LR-N23-0034, 2022 Annual Radiological Environmental Operating Report (AREOR) - Salem Nuclear Generating Station, Unit Nos. 1 and 2 and Hope Creek Generating Station2023-04-27027 April 2023 2022 Annual Radiological Environmental Operating Report (AREOR) - Salem Nuclear Generating Station, Unit Nos. 1 and 2 and Hope Creek Generating Station LR-N23-0035, 2022 Annual Radioactive Effluent Release Report (ARERR)2023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report (ARERR) IR 05000354/20230012023-04-26026 April 2023 Integrated Inspection Report 05000354/2023001 LR-N23-0010, License Amendment Request Revision of Technical Specification (TS) to Delete TS Section 5.5 - Meteorological Tower Location2023-04-21021 April 2023 License Amendment Request Revision of Technical Specification (TS) to Delete TS Section 5.5 - Meteorological Tower Location LR-N23-0009, License Amendment Request (LAR) to Revise the Hope Creek Trip and Standby Auto-start Logic Associated with Safety Related Heating, Ventilation and Air Conditioning (HVAC) Trains2023-04-18018 April 2023 License Amendment Request (LAR) to Revise the Hope Creek Trip and Standby Auto-start Logic Associated with Safety Related Heating, Ventilation and Air Conditioning (HVAC) Trains ML23087A1492023-04-17017 April 2023 NRC to PSEG Salem, Transmittal of the National Marine Fisheries Service'S March 24, 2023, Biological Opinion GAR-2020-02842 Concerning Salem and Hope Creek ML23089A0942023-04-17017 April 2023 NRC to PSEG Hope Creek, Transmittal of the National Marine Fisheries Service'S March 24, 2023, Biological Opinion GAR-2020-02842 Concerning Salem and Hope Creek ML23103A3232023-04-13013 April 2023 Submittal of Updated Final Safety Analysis Report, Rev. 26, Summary of Revised Regulatory Commitments for Hope Creek, Summary of Changes to PSEG Nuclear LLC, Quality Assurance Topical Report, NO-AA-10, Rev. 89 ML23095A3682023-04-12012 April 2023 and Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Threshold Determination for Proposed Transfer of Land Ownership LR-N23-0024, Submittal of Hope Creek Generating Station Technical Specification Bases Changes2023-03-29029 March 2023 Submittal of Hope Creek Generating Station Technical Specification Bases Changes LR-N23-0006, Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations2023-03-24024 March 2023 Report on Status of Decommissioning Funding for Reactors and Independent Spent Fuel Storage Installations ML23086A0912023-03-24024 March 2023 NMFS to NRC, Transmittal of Biological Opinion for Continued Operations of Salem and Hope Creek Nuclear Generating Stations LR-N23-0019, and Salem Generating Station, Units 1 and 2 - Guarantees of Payment of Deferred Premiums2023-03-21021 March 2023 and Salem Generating Station, Units 1 and 2 - Guarantees of Payment of Deferred Premiums ML23037A9712023-03-0909 March 2023 and Salem Nuclear, Unit Nos. 1 and 2 Issuance of Amendment Nos. 233, 344, and 325 Relocate Technical Specification Staff Qualification Requirements to the PSEG Quality Assurance Topical Report IR 05000354/20220062023-03-0101 March 2023 Annual Assessment Letter for Hope Creek Generating Station (Report 05000354/2022006) LR-N23-0016, and Salem Generating Station, Units 1 and 2 - Report of Changes, Tests, and Experiments2023-02-28028 February 2023 and Salem Generating Station, Units 1 and 2 - Report of Changes, Tests, and Experiments LR-N23-0018, Technical Specification 6.9.1.5.b - 2022 Annual Report of SRV Challenges2023-02-27027 February 2023 Technical Specification 6.9.1.5.b - 2022 Annual Report of SRV Challenges LR-N23-0012, Annual Property Insurance Status Report2023-02-24024 February 2023 Annual Property Insurance Status Report LR-N23-0014, Stations Submittal of 2022 Annual Report of Fitness for Duty Performance Data Per 10 CFR 26.203(e) and 10 CFR 26.7172023-02-23023 February 2023 Stations Submittal of 2022 Annual Report of Fitness for Duty Performance Data Per 10 CFR 26.203(e) and 10 CFR 26.717 IR 05000354/20220042023-01-24024 January 2023 Integrated Inspection Report 05000354/2022004 LR-N23-0011, In-Service Inspection Activities - 90 Day Report: Twenty-Fourth Refueling Outage2023-01-19019 January 2023 In-Service Inspection Activities - 90 Day Report: Twenty-Fourth Refueling Outage LR-N22-0096, and Salem Generating Station, Units 1 and 2 - Request for Threshold Determination2023-01-0505 January 2023 and Salem Generating Station, Units 1 and 2 - Request for Threshold Determination LR-N22-0094, Emergency Plan Document Revisions Implemented November 21, 20222022-12-14014 December 2022 Emergency Plan Document Revisions Implemented November 21, 2022 LR-N22-0091, Independent Spent Fuel Storage Installation, Report of 10 CFR 72.48 Changes, Tests, and Experiments2022-12-0202 December 2022 Independent Spent Fuel Storage Installation, Report of 10 CFR 72.48 Changes, Tests, and Experiments ML22335A0412022-12-0101 December 2022 Notification of Commercial Grade Dedication Inspection (05000354/2023011) and Request for Information IR 05000354/20220032022-11-0303 November 2022 Integrated Inspection Report 05000354/2022003 2024-02-01
[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML22006A3212022-01-0606 January 2022 NRR E-mail Capture - Final RAI - Hope Creek - Revise TS Limits for Ultimate Heat Sink ML21300A0222021-10-27027 October 2021 Notification of Conduct of a Fire Protection Team Inspection ML21041A3972021-02-18018 February 2021 Request for Additional Information Revise Low Pressure Safety Limit to Address General Electric Part-21 Safety Communications (Non-Proprietary) ML20295A4922020-10-21021 October 2020 NRR E-mail Capture - Hope Creek - Final RAI Revise ECCS TS with Respect to HPCI System Inoperability (L-2020-LLA-0131) ML18263A1442018-09-12012 September 2018 NRR E-mail Capture - Final RAI from Apla: Revise Technical Specifications to Increase Inverter AOT Extension (L-2018-LLA-0101) ML18250A3142018-09-0606 September 2018 NRR E-mail Capture - Final RAI: Revise Technical Specifications to Increase Inverter AOT Extension (L-2018-LLA-0101) ML18150A6912018-05-30030 May 2018 NRR E-mail Capture - Hope Creek - Final RAI Revise TS to Adopt TSTF-542 ML18120A2482018-04-27027 April 2018 Information Request for the Cyber-Security Inspection, Notification to Perform Inspection 05000354/2018403 ML18094B0832018-04-0505 April 2018 Enclosurequest for Additional Information (Letter to P. Duke RAI Regarding Entergy Operations, Inc.'S Decommissioning Funding Plan Update for Salem and Hope Creek, and Peach Bottom ISFSIs Docket Nos. 72-48 and 72-29) ML17349A0812017-12-14014 December 2017 NRR E-mail Capture - Final Request for Additional Information for Reactor Systems Branch (Srxb) - Hope Creek Mur ML17348A9972017-12-14014 December 2017 NRR E-mail Capture - Final Request for Additional Information for Steam Dryer Analysis with the Hope Creek Measurement Uncertainty Recapture Uprate Request (L-2017-LLS-002) ML17348A6242017-12-14014 December 2017 NRR E-mail Capture - Final Request for Additional Information for Human Factors Associated with the Hope Creek Measurement Uncertainty Recapture Uprate Request (L-2017-LLS-002) ML17331A0072017-11-21021 November 2017 E-mail Regarding Final Eicb Request for Additional Information - Hope Creek Power Uprate ML17348A6282017-11-21021 November 2017 NRR E-mail Capture - Draft Request for Additional Information for Steam Dryer Analysis with the Hope Creek Measurement Uncertainty Recapture Uprate Request (L-2017-LLS-002) ML17324A2602017-11-17017 November 2017 NRR E-mail Capture - Final Request for Additional Information - Hope Creek Mur Electrical Instrumentation and Controls ML17320A2152017-11-16016 November 2017 NRR E-mail Capture - Hope Creek Measurement Uncertainty Recapture Application Review - Electrical Engineering (Eeob) Request for Additional Information (L-2017-LLS-0002) ML17290B0132017-10-17017 October 2017 NRR E-mail Capture - Hope Creek Mur Request for Additional Information for Apla and Snpb ML17222A1652017-08-10010 August 2017 NRR E-mail Capture - Hope Creek: Final Request for Information, Pressure-Temperature Limit Report Request (MF9502) ML17181A1352017-06-29029 June 2017 NRR E-mail Capture - Request for Additional Information Related to Hope Creek and Salem Emergency Action Level Scheme Change License Amendment Request (MF9268/69/70) ML17164A4062017-06-13013 June 2017 NRR E-mail Capture - Hope Creek - Final Request for Additional Information Concerning ILRT (MF8462) ML17027A3282017-02-27027 February 2017 Request for Additional Information Regarding License Amendment Request to Permanently Extend Type a and Type C Leak Rate Test Frequencies ML16321A4642016-11-21021 November 2016 and Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Request for Additional Information Regarding License Amendment Request to Remove Certain Training Requirements ML16277A5142016-11-0303 November 2016 Request for Additional Information Regarding License Amendment Request to Permit Operability of Low Pressure Coolant Injection While Aligned to Shutdown Cooling ML16271A2052016-11-0202 November 2016 Request for Additional Information Regarding Request to Delete Technical Specifications Action Statement 3.4.2.1.b Associated with Stuck Open Safety/Relief Valves ML16231A4272016-08-24024 August 2016 Request for Additional Information Regarding Relief Request VR-02, Associated with the Fourth 10-Year Inservice Test Interval ML16217A4302016-08-17017 August 2016 Request for Additional Information Regarding Digital Power Range Neutron Monitoring System Upgrade ML16007A1742016-05-31031 May 2016 Request for Additional Information Regarding Review of Post-EPU Steam Dryer Stress Calculation Acoustic Circuit Model Software Error ML16089A0792016-05-0505 May 2016 Request for Additional Information Regarding Relief Requests GR-01, PR-01, PR-02, VR-01, and VR-02, Associated with the Fourth 10-Year Inservice Test Interval ML16124A5802016-05-0303 May 2016 and Salem Nuclear Generating Station, Unit Nos. 1 and 2 - Redacted - Request for Additional Information Review of Security Plan Revision 17 Changes ML15055A3772015-03-10010 March 2015 and Salem Nuclear Generating Station, Unit Nos. 1 and 2, Request for Additional Information Request to Update Appendix B to Renewed Facility Operating License DPR-70, DPR-75 and NPF-57 ML14197A6192014-07-24024 July 2014 and Salem Nuclear Generating Station, Units 1 and 2 - Request for Additional Information Regarding Cyber Security Plan Milestone 8 Implementation Schedule (Tac Nos. MF3384, MF3385 and MF3386) ML14168A2352014-07-0202 July 2014 Request for Additional Information Regarding Flooding Hazard Reevaluation ML13309B5922013-11-22022 November 2013 Interim Staff Evaluation and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Pool Instrumentation ML13304B4182013-11-0101 November 2013 Request for Additional Information Associated with Near-Term Task Force Recommendation 2.3, Seismic Walkdowns ML13168A2762013-08-12012 August 2013 Peach Bottom, Units 2 and 3, Request for Additional Information Decommissioning Funding Status Report (TAC No. MF2202; MF2227; MF2228) ML13193A2912013-07-22022 July 2013 Request for Additional Information Regarding Overall Integrated Plan for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) TAC MF1031 ML13204A2532013-07-19019 July 2013 Draft RAIs - Decommissioning Funding Status Report for Hope Creek and Peach Bottom ML13190A2212013-07-0202 July 2013 RAIs for Hope Creek Regarding Response to Order EA-12-051 ML13154A4952013-06-0404 June 2013 Rai'S Following Ifib Analysis of Pseg'S 2013 Decommissioning Funding Status Report for Hope Creek Generating Station ML12334A3102012-12-0505 December 2012 Request for Additional Information Request to Correct Technical Specification and Facility Operating License Editorial Items ML12310A1672012-11-0101 November 2012 Draft RAI to License Amendment Request Regarding Average Power Range Monitoring Operability Requirements in Operational Condition 5 ML1227901012012-10-0505 October 2012 Clarifying Information for Question No. 4 (with Attachment) ML12212A1032012-08-0808 August 2012 and Salem Nuclear Generating Station, Unit Nos. 1 and 2, Request for Additional Information ML12216A1792012-08-0303 August 2012 Draft RAI Related to Request for Exemption from Certain Requirements of the Fitness for Duty Rule ML12213A6162012-07-30030 July 2012 Draft Request for Additional Information Related to Request for Exemption from Certain Requirements of the Fitness for Duty Rule ML1209002372012-03-29029 March 2012 and Salem Nuclear Generating Station, Unit Nos. 1 and 2, Draft Request for Additional Information ML12056A0512012-03-12012 March 2012 Enclosure 5 - Recommendation 9.3: Emergency Preparedness ML12056A0482012-03-12012 March 2012 Enclosure 2 - Recommendation 2.1: Flooding ML12056A0492012-03-12012 March 2012 Enclosure 3 - Recommendation 2.3: Seismic ML12056A0472012-03-12012 March 2012 Enclosure 1 - Recommendation 2.1: Seismic 2022-01-06
[Table view] |
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.