Letter Sequence Supplement |
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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
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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
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Category:Letter type:LR
MONTHYEARLR-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 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 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) 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 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 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 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 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 LR-N22-0075, 2022 Annual 10 CFR 50.46 Report2022-09-30030 September 2022 2022 Annual 10 CFR 50.46 Report LR-N22-0074, Emergency Plan Evacuation Time Estimate2022-09-15015 September 2022 Emergency Plan Evacuation Time Estimate LR-N22-0051, License Amendment Request to Relocate Technical Specification Facility/Unit Staff Qualification Requirements to Quality Assurance Topical Report2022-06-22022 June 2022 License Amendment Request to Relocate Technical Specification Facility/Unit Staff Qualification Requirements to Quality Assurance Topical Report LR-N22-0044, Emergency Plan Document Revisions Implemented November, 20212022-05-19019 May 2022 Emergency Plan Document Revisions Implemented November, 2021 LR-N22-0041, 2021 Annual Radioactive Effluent Release Report (Rerr)2022-04-28028 April 2022 2021 Annual Radioactive Effluent Release Report (Rerr) LR-N22-0040, 2021 Annual Radiological Environmental Operating Report2022-04-28028 April 2022 2021 Annual Radiological Environmental Operating Report LR-N22-0039, Emergency Plan Document Revisions Implemented March 24, 20222022-04-21021 April 2022 Emergency Plan Document Revisions Implemented March 24, 2022 LR-N22-0023, Guarantees of Payment of Deferred Premiums2022-03-21021 March 2022 Guarantees of Payment of Deferred Premiums LR-N22-0017, Submittal of 2021 Annual Report of Fitness for Duty (FFD) Performance Data2022-02-25025 February 2022 Submittal of 2021 Annual Report of Fitness for Duty (FFD) Performance Data LR-N22-0016, Radiological Survey of Site Property to Be Used for Offshore Wind Port Facility2022-02-24024 February 2022 Radiological Survey of Site Property to Be 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Revision 32, Summary of Revised Regulatory Commitments for Salem, 10 CFR 71.106 Review Results and 10 CFR 54.37(b) Review Results for Salem LR-N21-0040, Response to Request for Additional Information SNSB-RAI 1 License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication2021-05-27027 May 2021 Response to Request for Additional Information SNSB-RAI 1 License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication LR-N21-0042, Core Operating Limits Report, Reload 23, Cycle 24, Revision 212021-05-24024 May 2021 Core Operating Limits Report, Reload 23, Cycle 24, Revision 21 LR-N21-0025, License Amendment Request to Revise Technical Specification Limits for Ultimate Heat Sink2021-05-0707 May 2021 License Amendment Request to Revise Technical Specification Limits for Ultimate Heat Sink LR-N21-0039, Deviation from EPRI Document 33002012244 Inspection Requirements2021-04-30030 April 2021 Deviation from EPRI Document 33002012244 Inspection Requirements LR-N21-0018, Response to Requests for Additional Information SNSB-RAI 2 and SNSB-RAI 3 License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication2021-04-29029 April 2021 Response to Requests for Additional Information SNSB-RAI 2 and SNSB-RAI 3 License Amendment Request to Revise Low Pressure Safety Limit to Address General Electric Part 21 Safety Communication 2023-09-08
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Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 O PSEG Nuclear LLC 10 CFR 50.90 LR-N07-0084 LCR H05-01, Rev. 1 April 18, 2007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354
Subject:
Supplement to License Amendment Request for Extended Power Uprate, Revised No Significant Hazards Consideration
Reference:
- 1) PSEG letter LR-N06-0286, Request for License Amendment:
Extended Power Uprate, September 18, 2006
- 2) PSEG letter LR-N07-0034, Supplement to License Amendment Request for Extended Power Uprate, February 27, 2007
- 3) PSEG letter LR-N07-0055, Supplement to License Amendment Request for Extended Power Uprate, March 13, 2007
- 4) PSEG letter LR-N06-0413, Supplement to License Amendment:
Request for Extended Power Uprate, October 10, 2006
- 5) PSEG letter LR-N06-0418, Supplement to License Amendment:
Request for Extended Power Uprate, October 20, 2006
- 6) PSEG letters LR-N07-0035, LR-N07-0056, LR-N07-0060, LR-N07-0069, and LR-N07-0070, Response to Request for Additional Information, Request for License Amendment -
Extended Power Uprate, dated March 13, March 22, March 30, March 30, and April 13, 2007 In Reference 1, PSEG Nuclear LLC (PSEG) requested an amendment to Facility Operating License NPF-57 and the Technical Specifications (TS) for the Hope Creek Generating Station (HCGS) to increase the maximum authorized power level to 3840 megawatts thermal (MWt).
References 2 and 3 provided the results of evaluations performed to facilitate staff review, demonstrating the conservatism in the predicted loads on the HCGS steam dryer for extended power uprate (EPU) operation. References 4 and 5 kow 95-2168 REV. 7/99
LR-N07-0084 LCR H05-01, Rev. 1 April 18, 2007 provided additional documentation of evaluations performed in support of the requested amendment. Reference 6 provided PSEG responses to NRC Requests for Additional Information on the EPU amendment request.
Following discussions with the NRC staff, PSEG has revised the proposed No Significant Hazards Consideration (NSHC) determination that was provided in Reference 1. The revised NHSC (Attachment 1 to this submittal) is more detailed reflecting the additional information provided in References 2 through 6 and is generally more comprehensive consistent with the regulatory guidance provided in NRC Regulatory Issue Summary (RIS) 2001-22, "Attributes of a Proposed No Significant Hazards Consideration," dated November 20, 2001.
Should you have any questions regarding this submittal, please contact Mr. Paul Duke at 856-339-1466.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on_________ 7 (Date)
Sincerely, George Barnes Site Vice President Hope Creek Generating Station Attachments (1)
S. Collins, Regional Administrator - NRC Region I J. Shea, Project Manager - USNRC NRC Senior Resident Inspector - Hope Creek K. Tosch, Manager IV, NJBNE LR-N07-0084 LCR H05-01, Rev. 1 No Significant Hazards Consideration This proposed extended power uprate (EPU) amendment increases the Hope Creek Generating Station (HCGS) licensed thermal power level to 3840 megawatts thermal (MWt), approximately 15% above the current rated thermal power (RTP) of 3339 MWt and 16.6% above the original RTP of 3293 MWt. The requested increase in reactor thermal power level will allow operational changes to generate higher steam flow to the turbine generator, in turn permitting an increase in the electrical output of the plant.
A higher steam flow is achieved by increasing the reactor power along specified control rod and core flow lines. A limited number of operating parameters are changed, some setpoints are adjusted and instruments are recalibrated. Plant procedures are revised, and tests similar to some of the original startup tests are performed.
The technical bases for this request follow the guidelines contained in NRC-approved GE Nuclear Energy (GENE) Licensing Topical Reports (LTRs) for a Constant Pressure Power Uprate (CPPU). CPPU operation does not involve an increase in the maximum normal operating reactor dome pressure, or the maximum licensed core flow. This is possible because the plant, following modifications to non-power generating equipment, has sufficient capabilities to control turbine generator inlet pressure.
Detailed evaluations of the reactor, engineered safety features, emergency power, support systems, environmental issues, and design basis accidents have been performed. These evaluations demonstrate that Hope Creek can safely operate at 3840 MWt.
PSEG Nuclear LLC (PSEG) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment" as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The CPPU analyses, which were performed at or above CPPU power levels, included a review and evaluation of the structures, systems, and components (SSCs) that could be affected by the proposed change. The proposed amendment does not change the design function or operation of the affected SSCs.
1 LR-N07-0084 LCR H05-01, Rev. 1 Plant specific analyses were performed in the following areas: Reactor Core and Reactor Internals (e.g., steam dryer), Reactor Coolant System and associated systems, Containment, Emergency Core Cooling Systems, Control and Instrumentation Systems, Electrical Systems, Balance of Plant Systems, and Radwaste Systems. The results of the analyses, which included evaluating the increase in the likelihood of an SSC malfunction, concluded that the SSCs are capable of performing their design functions at CPPU conditions.
Comprehensive evaluations were performed on the steam dryer and other reactor internals for both operational and structural performance. Predicted steam dryer peak and alternating stress ratios remain within allowable levels.
The existing margins to steam dryer alternating stress limits and the steam dryer monitoring program during power ascension provide assurance that steam dryer integrity will be maintained.
Vibration evaluations at CPPU conditions were performed on the Reactor Internal components and Reactor Coolant and associated system piping. These included the Main Steam, Feedwater and Reactor Recirculation systems piping and supports. The results of the vibration analyses demonstrate that operation at CPPU conditions will not result in any detrimental effects. System values will remain within allowable American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) limits. In addition, the ASME Code and regulatory guidelines require vibration test data be taken on high-energy piping during initial CPPU startup. The vibration start-up test program will validate the vibration analyses that were performed, demonstrating adequate performance of the SSCs.
Engineered Safety Features (ESF) were evaluated at CPPU conditions using NRC-approved methods. The Emergency Core Cooling Systems (ECCS) were evaluated to ensure they are capable of performing their design function during loss-of-coolant-accidents (LOCA). Adequate net positive suction head is maintained without reliance on post-accident containment pressure. CPPU does not result in an increase or decrease in the available water sources, and does not result in any change in the maximum nominal reactor operating pressure. The CPPU evaluations demonstrate that the ECCS performance satisfy the requirements of 10 CFR 50.46 and 10 CFR 50 Appendix K.
Balance-of-plant (BOP) systems and equipment were also evaluated for CPPU operation. The resulting evaluations demonstrate adequate performance with limited modifications that were or will be made to BOP components.
These analyses, which included evaluating the increased likelihood of an SSC malfunction, confirm acceptable performance of plant SSCs under CPPU conditions. On this basis, PSEG concludes that there is no significant change in the ability of the SSCs to preclude or mitigate the consequences of accidents.
2 LR-N07-0084 LCR H05-01, Rev. 1 The probability (frequency of occurrence) of postulated Design Basis Accidents (DBA), and other Updated Final Safety Analysis Report (UFSAR) evaluated accidents, occurring is not affected by the increased power level, and Hope Creek continues to comply with the regulatory and design basis criteria established for plant equipment. The changes in consequences of hypothetical accidents, which are assumed to occur at 102% of the CPPU RTP, compared to those previously evaluated, are in all cases insignificant. The CPPU accident evaluations do not exceed any of the NRC-approved acceptance limits. The spectrum of hypothetical accidents and transients has been investigated, and is shown to meet the plant's currently licensed regulatory criteria. Consequently, there is no significant increase in the probability or consequences of an accident previously evaluated.
The impact of CPPU on the radiological consequences of postulated DBAs, operational transients and other UFSAR accidents was evaluated. The magnitude of the potential consequences is dependent upon the quantity of fission products released to the environment, the atmospheric dispersion factors and the dose exposure pathways. The atmospheric dispersion factors and the dose exposure pathways are not changed by CPPU operation. The only factor which could influence the magnitude of the consequences is the quantity of activity released to the environment. For CPPU, the Control Rod Drop Accident (CRDA), Loss-of-Coolant Accident (LOCA), Fuel Handling Accident (FHA), Main Steamline Break Accident (MSLBA) and instrument line break accident (ILBA) were reanalyzed.
The DBA that has historically been limiting from a radiological criterion is the LOCA, for which USNRC Regulatory Guide 1.183, Appendix A guidance was applied. Adherence to the guidance in RG 1.183, and the use of the specific values/limits contained in the Technical Specifications with as-tested post-accident performance of the safety grade engineered safety functions (ESF),
provide the assurance for sufficient safety margin, including a margin to account for analysis uncertainties. The CPPU LOCA evaluation results include the 2%
power uncertainty factor from Regulatory Guide 1.49.
The results of the CPPU radiological analyses remain below the allowable limits of 10 CFR 50.67 and Table 6 in Regulatory Guide 1.183; the CPPU impact is minimal and all radiological limits are met at CPPU conditions. Therefore, the proposed change does not involve a significant increase in the radiological consequences of an accident previously evaluated.
While the proposed CPPU amendment is not being submitted as a risk-informed licensing action, it was evaluated from a risk perspective using the NRC guidelines established in Regulatory Guide 1.174. Level 1 and Level 2 Probabilistic Risk Assessments (PRAs) were performed for the CPPU. When compared to the risk-acceptance guidelines presented in Regulatory Guide 1.174, the calculated changes in core damage frequency (CDF) and large early 3
LR-N07-0084 LCR H05-01, Rev. 1 release frequency (LERF) are insignificant. Based on these results, PSEG concludes that the proposed amendment would not involve a significant increase in the probability of an accident previously evaluated.
The impact of CPPU operation on plant operator actions and procedures was also evaluated. The operator action response times credited in the safety analyses in the UFSAR are not changed by CPPU. In addition, there is no change in Emergency Operating Procedure (EOP) strategy for CPPU operation.
Based on the above, PSEG concludes that the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
As discussed above, the evaluation of the proposed amendment included review of the SSCs that could be affected by the proposed change. The proposed amendment does not change the design function or operation of the affected SSCs. The proposed amendment does not introduce any new or different plant safety-related equipment, and only involves instrument set-point changes for CPPU conditions, and minimal modifications to plant BOP power generation equipment. The proposed amendment does not significantly impact the manner in which the plant is operated, and does not have any significant impact on the capability the SCCs involved to perform their design function.
No new operating mode, safety-related equipment lineup, accident scenario or equipment failure mode was identified. The CPPU evaluations also addressed the impact to postulated accidents, accident radiological consequences and operator response. No significant impacts were identified. The full spectrum of accident considerations has been evaluated, and no new, different, or limiting kind of accident has been identified. CPPU uses developed technology, and applies it within the capabilities of existing plant equipment in accordance with presently existing regulatory criteria to include NRC approved codes, standards and methods. The CPPU analyses results confirm acceptable performance of plant SSCs under CPPU conditions. Consequently, there are no new credible failure mechanisms, malfunctions, or accident initiators that were not previously evaluated in the plant design and licensing bases.
Based on the preceding, PSEG concludes that the proposed change would not introduce any new or different kind of accident, or failure mode, not previously analyzed.
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- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
Safety margins are applied to plant parameters to account for various uncertainties and to avoid exceeding regulatory and licensing limits. The proposed change does not involve a significant reduction in any margin of safety.
First, due to continuing improvements in the analytical techniques (computer codes and data) based on several decades of BWR safety technology, plant performance feedback, and improved fuel and core designs, a significant increase has resulted in the design and operating margins between calculated safety analysis results and the licensing limits. These available safety analyses differences, combined with the excess as-designed equipment, system and component capabilities, provide BWR plants the capability to achieve an increase in their thermal power ratings within the existing design and licensing basis. The proposed CPPU will reduce some of the existing design and operational margins.
However, safety margins are considered to not be significantly reduced if: (1) applicable regulatory requirements, codes and standards or their alternatives approved for use by the NRC, are met, and (2) if safety analysis acceptance criteria in the licensing basis are met, or if proposed revisions to the licensing basis provide sufficient margin to account for analysis and data uncertainty. This is the case for the proposed CPPU amendment.
Safety margin is related to the ability of the fission product barriers to limit the level of radiation dose to the public. The impact of the proposed CPPU amendment on the: (1) fuel cladding barrier, (2) reactor coolant pressure boundary (RCPB) barrier, and (3) containment fission product barrier is discussed below.
To assure that fuel cladding damage limits are not exceeded, the impact of the proposed amendment on fuel system design, nuclear system design, thermal and hydraulic design, accident and transient analyses, and fuel design limits was evaluated. No new fuel design, or change in the specified fuel design limits, is required for CPPU. The current fuel and core design limits will continue to be met; both the Safety Limit Minimum Critical Power Ratio (SLMCPR) and other applicable Specified Acceptable Fuel Design Limits (SAFDLs) are still met.
Analyses for each fuel reload will continue to meet the criteria accepted by the NRC. Continued compliance with the SLMCPR and other SAFDLs will be confirmed on a cycle specific basis consistent with the criteria accepted by the NRC as specified in NEDO-2401 1, "General Electric Standard Application for Reactor Fuel, GESTAR I1." The ECCS evaluation for CPPU demonstrates the continued conformance to the acceptance criteria of 10 CFR 50.46, for peak cladding temperature (PCT) and the other 10 CFR 50.46 parameters. The increased PCT consequences for CPPU are insignificant and remain 5
LR-N07-0084 LCR H05-01, Rev. 1 substantially below the regulatory criteria. Therefore, the ECCS safety margin and fuel cladding margin (PCT) are not significantly impacted by CPPU.
Challenges to the Reactor Coolant Pressure Boundary were evaluated at CPPU conditions (pressure, temperature, flow, and radiation) and were found to meet their acceptance criteria for allowable stresses and overpressure margin. These evaluations included (1) overpressure protection, (2) structural integrity of the RCPB piping, components, and supports, and (3) structural integrity of the reactor vessel. For the most limiting pressurization event, the peak calculated pressure remains below the ASME Code allowable peak pressure. The structural integrity of the RCPB piping, components, and supports was evaluated using NRC-approved methodology. The changes in flow, pressure and temperature associated with CPPU do not result in load limits being exceeded.
Sufficient margin remains between the calculated stresses and ASME Code limits. In addition, the ASME Code and regulatory guidelines require vibration test data be taken on high-energy piping during initial CPPU startup. The vibration start-up test program will validate the vibration analyses that were performed, demonstrating adequate performance.
The structural integrity of the reactor vessel was evaluated. The neutron fluence was re-analyzed in accordance with the requirements of 10 CFR 50 Appendix G.
The existing Pressure-Temperature (P-T) limit curves have been revised for CPPU conditions (a previous amendment to the Hope Creek license changed the P-T curves and included CPPU conditions). The reactor vessel materials surveillance program is unchanged by CPPU. The maximum normal operating reactor dome pressure for CPPU is unchanged and the vessel remains in compliance with regulatory requirements. Consequently, CPPU operation does not have an adverse effect on the reactor vessel fracture toughness. The structural evaluation of the vessel demonstrates that ASME Code requirements are met for normal, upset, emergency and accident conditions.
Based on the preceding, PSEG concludes that the RCPB structural integrity will be maintained and the licensing basis requirements will continue to be met following implementation of the proposed CPPU.
The impact of the proposed CPPU on the Containment was evaluated. The effect of CPPU on the peak values for containment pressure and temperature confirms the suitability of the plant for operation at CPPU RTP. Also, the effects of CPPU on the conditions that affect the containment dynamic loads were determined to be satisfactory for CPPU operation. Where plant conditions with CPPU are within the range of conditions used to define the current dynamic loads, current safety criteria are met and no further structural analysis was required. The change in short-term containment response is negligible.
Because there will be more residual heat with CPPU, the containment long-term response slightly increases. However, containment pressures and temperatures remain below their design limits following any design basis accident, and thus, 6
LR-N07-0084 LCR H05-01, Rev. 1 the containment and its cooling systems are satisfactory for CPPU operation.
The small increase in the calculated post LOCA suppression pool temperature above the currently assumed peak temperature was evaluated and determined to be acceptable. Based on the use of conservative assumptions in these evaluations, PSEG concludes that containment structural integrity will be maintained under the proposed CPPU conditions, and the containment parameters will remain below design limits. Therefore there is no significant reduction in safety margin.
In summary, challenges to the fuel, RCPB, and containment were evaluated for CPPU conditions. The structural integrity of the fission product barriers will be maintained under CPPU conditions. As such, the proposed amendment would not degrade confidence in the ability of the barriers to limit the level of radiation dose to the public. Fuel integrity is maintained by meeting existing design and regulatory limits. The calculated loads on all affected structures, systems and components, including the reactor coolant pressure boundary, will remain within their design allowables for all design basis event categories. The containment parameters remain below design limits. No NRC acceptance criterion will be exceeded. Because the Hope Creek configuration and responses to transients and hypothetical accidents do not result in exceeding the presently approved NRC acceptance limits, CPPU does not involve a significant reduction in a margin of safety.
Based on the above, PSEG concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.
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