|
---|
Category:Letter type:LR
MONTHYEARLR-N23-0079, Special Report 23-02-00 Pursuant to the Requirements of Salem Unit 1 Technical Specification 3.3.3.7, Action 10, for the Unit 1 Main Steam Line Rad Monitor Inoperable for Greater than Seven Days2023-12-0707 December 2023 Special Report 23-02-00 Pursuant to the Requirements of Salem Unit 1 Technical Specification 3.3.3.7, Action 10, for the Unit 1 Main Steam Line Rad Monitor Inoperable for Greater than Seven Days LR-N23-0077, Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report 2023 & 30 Day Report for Salem Unit 1 Upflow Conversion2023-11-29029 November 2023 Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report 2023 & 30 Day Report for Salem Unit 1 Upflow Conversion LR-N23-0072, Core Operating Limits Report Cycle 302023-11-0101 November 2023 Core Operating Limits Report Cycle 30 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 LR-N23-0055, Special Report 272/23-01-00, Pursuant to the Requirements of Technical Specification 3.3.3.1, Action 23, for the Vent Noble Gas Rad Monitor Inoperable for Greater than Seven Days2023-08-0303 August 2023 Special Report 272/23-01-00, Pursuant to the Requirements of Technical Specification 3.3.3.1, Action 23, for the Vent Noble Gas Rad Monitor Inoperable for Greater than Seven Days 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-N23-0054, In-Service Inspection Activities2023-07-26026 July 2023 In-Service Inspection Activities 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-N23-0005, License Amendment Request to Amend Technical Specifications (TS) 6.8.4.f for Permanent Extension of Type a and Type C Leak Rate Test Frequencies2023-06-23023 June 2023 License Amendment Request to Amend Technical Specifications (TS) 6.8.4.f for Permanent Extension of Type a and Type C Leak Rate Test Frequencies LR-N23-0035, 2022 Annual Radioactive Effluent Release Report (ARERR)2023-04-27027 April 2023 2022 Annual Radioactive Effluent Release Report (ARERR) 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-N23-0033, Core Operating Limits Report Cycle 272023-04-26026 April 2023 Core Operating Limits Report Cycle 27 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-N23-0003, Response to Requests for Additional Information Salem Unit 2 Relief Request S2-I4R-2112023-02-0101 February 2023 Response to Requests for Additional Information Salem Unit 2 Relief Request S2-I4R-211 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-N22-0095, Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report 20222022-12-20020 December 2022 Loss of Coolant Accident Peak Cladding Temperature Margin Tracking - Annual Report 2022 LR-N22-0094, Emergency Plan Document Revisions Implemented November 21, 20222022-12-14014 December 2022 Emergency Plan Document Revisions Implemented November 21, 2022 LR-N22-0092, Response to Final Iolb Request for Additional Information for Salem LAR to Revise TS to Extend Allowed Outage Time for Inoperable EDG2022-12-0909 December 2022 Response to Final Iolb Request for Additional Information for Salem LAR to Revise TS to Extend Allowed Outage Time for Inoperable EDG 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-N22-0084, Response to Final Request for Additional Information for Salem LAR to Revise TS to Extend Allowed Outage Time for Inoperable EDG (EPID L- 2022-LLA-0095)2022-11-17017 November 2022 Response to Final Request for Additional Information for Salem LAR to Revise TS to Extend Allowed Outage Time for Inoperable EDG (EPID L- 2022-LLA-0095) LR-N22-0090, Supplement to Submittal of Salem Generating Station Updated FSAR, Revision 33, 10 CFR 71.106 Review Results and 10 CFR 54.37(b) Review Results for Salem2022-11-10010 November 2022 Supplement to Submittal of Salem Generating Station Updated FSAR, Revision 33, 10 CFR 71.106 Review Results and 10 CFR 54.37(b) Review Results for Salem LR-N22-0075, 2022 Annual 10 CFR 50.46 Report2022-09-30030 September 2022 2022 Annual 10 CFR 50.46 Report LR-N22-0065, Submittal of Relief Request Associated with the Fourth Inservice Inspection (ISI) Interval Limited Examinations2022-09-27027 September 2022 Submittal of Relief Request Associated with the Fourth Inservice Inspection (ISI) Interval Limited Examinations LR-N22-0074, Emergency Plan Evacuation Time Estimate2022-09-15015 September 2022 Emergency Plan Evacuation Time Estimate LR-N22-0066, License Amendment Request (LAR) to Relocate Technical Specifications (TS) Requirements for Reactor Head Vents to the Technical Requirements Manual (TRM)2022-08-31031 August 2022 License Amendment Request (LAR) to Relocate Technical Specifications (TS) Requirements for Reactor Head Vents to the Technical Requirements Manual (TRM) LR-N22-0063, Spent Fuel Cask Registration2022-08-10010 August 2022 Spent Fuel Cask Registration LR-N22-0068, In-Service Inspection Activities - 90-Day Report2022-08-10010 August 2022 In-Service Inspection Activities - 90-Day Report LR-N22-0012, License Amendment Request to Amend the Technical Specifications to Revise and Relocate the Reactor Coolant System Pressure and Temperature Limits and Pressurizer Overpressure Protection System Limits to a Pressure and Temperature2022-08-0707 August 2022 License Amendment Request to Amend the Technical Specifications to Revise and Relocate the Reactor Coolant System Pressure and Temperature Limits and Pressurizer Overpressure Protection System Limits to a Pressure and Temperature LR-N22-0062, Spent Fuel Cask Registration2022-07-21021 July 2022 Spent Fuel Cask Registration LR-N22-0006, License Amendment Request (LAR) to Amend Salem Unit 1 and Unit 2 Technical Specifications (TS) to Extend the Allowed Outage Time for an Inoperable Emergency Diesel Generator from 72 Hours to 14 Days2022-06-29029 June 2022 License Amendment Request (LAR) to Amend Salem Unit 1 and Unit 2 Technical Specifications (TS) to Extend the Allowed Outage Time for an Inoperable Emergency Diesel Generator from 72 Hours to 14 Days 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-N22-0043, Core Operating Limits Report - Cycle 292022-05-0909 May 2022 Core Operating Limits Report - Cycle 29 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-N21-0052, Request for Relief from ASME Code Defect Removal for Service Water Buried Piping2022-04-0707 April 2022 Request for Relief from ASME Code Defect Removal for Service Water Buried Piping LR-N22-0023, Guarantees of Payment of Deferred Premiums2022-03-21021 March 2022 Guarantees of Payment of Deferred Premiums LR-N22-0022, Response to Request for Additional Information Relief Request S1-14R-210, Alternative Examination of Welds2022-03-21021 March 2022 Response to Request for Additional Information Relief Request S1-14R-210, Alternative Examination of Welds 2023-09-08
[Table view] |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC JUN 2 5 2012 10 CFR 72.236(f)
LR-N12-0194 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Director, Division of Spent Fuel Storage and Transportation Office of Nuclear Materials and Safeguards Washington, DC 20555-0001 Salem Generating Station Units 1 and 2 Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos, 50-272 and 50-311 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354 Salem/Hope Creek Generating Station Independent Spent Fuel Storage Installation NRC Docket No. 72-0048
Subject:
HI-STORM 100 Cask Supplemental Cooling System Validation Testing Using Air Mass Flow Rate
References:
- 1) Entergy Letter to NRC 0CAN090902, HI-STORM-100 Cask System Supplemental Cooling System Validation Test, Arkansas Nuclear One, September 29, 2009
- 2) Entergy Letter to NRC, GNRO-2011/00086, HI-STORM-1 00 Cask System Supplemental Cooling System Validation Test, Grand Gulf Nuclear Station, Unit 1, October 14, 2011
- 3) Holtec Report HI-2002444, "HI-STORM 100 Cask System Final Safety Analysis Report", Revision 7 Condition 9, Special Requirements for First Systems in Place, of the Holtec HI-STORM 100 System Certificate of Compliance (CoC), requires a report of the Supplemental Cooling System (SCS) validation test and analysis for each first time user of a HI-STORM 100 Cask System SCS that uses components or a system that is not essentially identical to components or a system that has been previously tested. The SCS was first utilized during the initial Salem Unit 1 dry fuel storage loading campaigns that took place in September and October 2010. Each of the four systems loaded had heat loads that exceeded 28.74kW and/or included high burnup fuel assemblies, thus required the use of SCS.
Document Control Desk LR-N12-0194 Page 2 Prior SCS submittals (References 1 and 2) were reviewed to determine applicability for the use of SCS at Salem. These were deemed not to be applicable as neither had a configuration that could be considered essentially identical to the Salem SCS.
A summary of the review performed of the SCS performance during its initial use at PSEG Nuclear is provided in the Attachment. The results demonstrate that the SCS performance limits the coolant temperature to below 180 degrees Fahrenheit under steady-state conditions for the design basis heat load at an ambient air temperature of 100 degrees Fahrenheit as required by Reference 3. Therefore, the results for the MPCs loaded in 2010 validate the thermal methods described in the HI-STORM FSAR used to determine the SCS requirements.
There are no commitments contained in this letter.
If you have any questions or require additional information, please contact Paul Bonnett at 856-339-1923.
Sincerely, John F. Perry *(
Site Vice President - Hope Creek Attachment - Validation of PSEG Nuclear Supplemental Cooling System (SCS)
Performance cc: Mr. W. Dean, Regional Administrator - Region I U. S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 U. S. Nuclear Regulatory Commission Mr. J. Hughey, Project Manager - Hope Creek Mail Stop 08B3 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Hope Creek (X24)
USNRC Senior Resident Inspector - Salem (X24)
Document Control Desk LR-N12-0194 Page 3 Mr. P. Mulligan, Manager IV Bureau of Nuclear Engineering PO Box 420 Trenton, New Jersey 08625 Hope Creek Commitment Coordinator (H02)
Salem Commitment Coordinator (X25)
Corporate Commitment Coordinator (N21)
Document Control Desk LR-N12-0194 Page 4 bcc: Vice-President - Salem (S05)
Director - Nuclear Fuels (N20)
Regulatory Affairs Manager - Salem (x25)
Regulatory Affairs manager - Hope Creek (H02)
Document Control Desk Attachment to LR-N 12-0194 Page 1 Attachment Validation of PSEG Nuclear Supplemental Cooling System (SCS) Performance
Document Control Desk Attachment to LR-N12-0194 Page 2 The following tables provide the Supplemental Cooling System (SCS) temperature and water flow rate information as recorded during the initial use of the system for the Salem Unit 1 2010 dry fuel storage campaign:
MPC-93 Time after SCS SCS SCS initiated T-inlet T-outlet Flow-rate Date/time (hours) (deg-F) (deg-F) (gpm) 10/7/10 14:30 0 162 136 14.49 10/7/10 15:30 1.0 156 130 14.38 10/7/10 17:30 3.0 152 120 11.00 10/7/10 20:30 6.0 150 130 13.54 10/7/10 22:30 8.0 142 118 13.26 10/8/10 0:30 10.0 138 114 13.37 10/8/10 2:30 12.0 132 110 13.37 10/8/10 4:30 14.0 130 108 13.35 10/8/10 6:30 16.0 126 106 13.35 10/8/10 8:30 18.0 123 102 13.33 MPC-94 Time after SCS SCS SCS initiated T-inlet T-outlet Flow-rate Date/time (hours) (deg-F) (deg-F) (gpm) 10/2/10 2:00 0 170 140 13.63 10/2/10 4:00 2.0 158 130 13.67 10/2/10 6:00 4.0 152 128 14.51 10/2/10 8:45 6.8 158 130 13.48 10/2/10 10:00 8.0 148 124 14.40 10/2/10 11:00 9.0 144 122 14.38 10/2/10 13:00 11.0 139 118 14.38 MPC-95 Time after SCS SCS SCS initiated T-inlet T-outlet Flow-rate Date/time (hours) (deg-F) (deg-F) (gpm) 9/23/10 15:35 0 150 128 14.10 9/23110 17:30 1.9 154 132 14.42 9/23/10 19:30 3.9 158 131 14.41 9/23/10 21:30 5.9 160 135 14.96 9/23/10 23:30 7.9 158 124 13.78 9/24/10 1:30 9.9 144 122 13.76 9/24/10 3:30 11.9 140 120 13.67 9/24/10 5:30 13.9 136 116 13.54 9/24/10 7:30 15.9 134 114 13.71
Document Control Desk Attachment to LR-N12-0194 Page 3 MPC-96 Time after SCS SCS SCS initiated T-inlet T-outlet Flow-rate Date/time (hours) (deg-F) (deg-F) (gpm) 9/17/10 4:00 0 170 145 14.10 9/17/10 6:00 2.0 163 137 14.42 9/17/10 8:00 4.0 157 132 14.41 9/17/10 9:05 5.1 157 132 14.96 9/17/10 10:00 6.0 162 138 13.78 9/17/10 12:00 8.0 150 126 13.76 9/17/10 14:00 10.0 144 124 13.67 Note: SCS T-inlet or T-outlet measurements are taken from the inlet and outlet of the SCS heat-exchanger.
MPC helium backfill data Final Helium Backfill Measured He Pressure Pressure Adjusted to 70 deg-F MPC (psig) (psig) 93 52.8 37.53 94 63.5 44.48 95 52.2 37.23 96 64.1 44.63 Fuel Handling Building temperature during the time of the 2010 dry storage campaign ranged from 67 to 81 degrees Fahrenheit.
The MPC heat generation was calculated initially as part of the fuel selection for cask loading effort.
MPC Heat Loads Qtotal(1) Qcoc(2)
(kW)
MPC (kW) 93 24.03 33.77 94 23.66 33.34 95 22.85 33.44 96 22.72 33.44 Notes:
(1) Qtotal is the summation of the decay heat loads in each of the MPC 32 cells. This includes the individual assembly and contained component decay heat values at the time the MPC was loaded.
Document Control Desk Attachment to LR-N12-0194 Page 4 (2) Qcoc is the MPC heat generation used to demonstrate CoC compliance per HI-STORM 100 FSAR Revision 7 Section 2.1.9.1.2. It is determined by taking the highest loaded fuel assembly heat generation (includes fuel and inserted component) multiplied by the total number of MPC cells.
The combined SCS performance results for the initial system use are shown below:
Salem Unit 1 2010 Dry Storage - SCS Measurements 180 170 160 150 U- 140
- 0) 130 120 110 100 90 0 2 4 6 8 10 12 14 16 18 20 Time from SCS initiation (hours) s MPC-93 inlet - -o- - MPC-93 outlet & MPC-94 inlet - - - MPC-94 outlet
-m-- MPC-95 inlet - .o- - MPC-95 outlet -- MPC-96 inlet - -o- - MPC-96 outlet During the 2010 Salem dry fuel storage campaign, the SCS maintained the coolant water temperate below 180 degrees Fahrenheit at all times the system was running for all four canisters. As previously noted, the ambient air temperature in the fuel handling building remained below 100 degrees Fahrenheit during these evolutions. An additional assessment was performed to determine the system capability over a range of ambient conditions for a number of SCS water temperatures. This provides an additional validation for this aspect of the CoC Condition 9 SCS requirement.
The heat exchanger utilized in the Salem SCS is sized to remove 110,000 BTU/hour from the cooling water to an ambient air temperature of 100 degrees Fahrenheit. Per Holtec SCS specification, the heat exchanger was sized to remove sufficient heat to maintain fuel cladding temperatures below 400 degrees Celsius at an ambient air
Document Control Desk Attachment to LR-N12-0194 Page 5 temperature of 100 degrees Fahrenheit (assuming maximum fouling factors for the process liquids).
To maintain a lower outlet water temperature, the heat exchanger needs to transfer more heat to ambient. Also, as the difference between the water temperature (inside the annulus between the HI-TRAC (transfer cask) and MPC) and ambient air temperature increases, more of the MPC heat is rejected through the HI-TRAC, thus reducing the required heat exchanger capability.
Heat exchanger required capacity is shown below over a given range of ambient air and SCS water temperatures:
Required heat transfer rate (BTU/hour)
T-ambient -> 80 90 100 110 T-scs I
V 140 93,210 97,020 100,820 104,620 150 88,930 92,740 96,540 100,340 160 84,360 88,460 92,260 96,060 170 80,090 84,180 87,980 91,780 180 75,610 79,420 83,320 87,500 Note, both T-ambient and T-SCS are in degrees Fahrenheit All the above cases consider a total MPC heat generation of 34 kW. In all cases, the heat exchanger capacity exceeds the system requirements over the expected range of ambient conditions and required SCS water temperature. This is consistent with the SCS recorded data that that shows a reduction in water temperature while the system was in operation. Thus, the SCS in use at PSEG Nuclear, as demonstrated during the initial Salem DCS campaign, meets the necessary HI-STORM FSAR Revision 7 Appendix 2.C design criteria.