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| {{#Wiki_filter:Timothy S. Rausch President and Chief Nuclear Officer SEP 2 4 l015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Susquehanna | | {{#Wiki_filter:Timothy S. Rausch President and Chief Nuclear Officer SEP 2 4 l015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, P A 18603 Tel. 570.542.3445 F a x 570.542.1504 Timothy.Rausch |
| : Nuclear, LLC 769 Salem Boulevard
| | @t a lenenergy.com SUSQUEHANNA STEAM ELECTRIC STATION FLOOD HAZARDS REEVALUATION REPORT INFORMATION TO SUPPORT AUDIT PLA-7389 EN ERG Y 10 CFR2.202 Docket Nos. 50-387 and 50-388 |
| : Berwick, P A 18603 Tel. 570.542.3445 Fax 570.542.1504 Timothy.Rausch
| |
| @talenenergy | |
| .com SUSQUEHANNA STEAM ELECTRIC STATION FLOOD HAZARDS REEVALUATION REPORT INFORMATION TO SUPPORT AUDIT PLA-7389 ENERGY 10 CFR2.202 Docket Nos. 50-387 and 50-388 | |
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| ==References:== | | ==References:== |
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| I. NRC Letter, Request for Information Pursuant to Title IO of the Code of Federal Regulations 50.54(/) Regarding Recommendations 2.I, 2.3, and 9.3, of the Term Task Force Review of Insights from the Fukushima Dai-Ichi | | I. NRC Letter, Request for Information Pursuant to Title IO of the Code of Federal Regulations 50.54(/) Regarding Recommendations 2.I, 2.3, and 9.3, of the Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, dated March I2, 20I2 2. PPL Letter (PLA-6867), Response to Request for Information Pursuant to Title IO of the Code of F e deral Regulations 50.54(/) Regarding the Flooding Aspects of Recommendations 2.I and 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated June II, 20I2 3. PPL Letter (PPL-7287), Flood Hazards Reevaluation Report, dated March 3, 20I5 The United States Nuclear Regulatory Commission (NRC) issued Reference 1 on March 12, 2012, pursuant to Title 10 of the Code of Federal Regulations (CFR), Section 50.54(f), related to the implementation of Recommendations 2.1, 2.3, and 9.3 from the Near-Tetm Task Force, a portion of which called for performing flood hazard reevaluations at all nuclear power plants in the United States. Reference 2 indicated plans to comply with the requested response date of March 12, 2015 for flood hazard evaluation. |
| : Accident, dated March I2, 20I2 2. PPL Letter (PLA-6867),
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| Response to Request for Information Pursuant to Title IO of the Code of Federal Regulations 50.54(/) | |
| Regarding the Flooding Aspects of Recommendations 2.I and 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi | |
| : Accident, dated June II, 20I2 3. PPL Letter (PPL-7287),
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| Flood Hazards Reevaluation Report, dated March 3, 20I5 The United States Nuclear Regulatory Commission (NRC) issued Reference 1 on March 12, 2012, pursuant to Title 10 of the Code of Federal Regulations (CFR), Section 50.54(f), | |
| related to the implementation of Recommendations 2.1, 2.3, and 9.3 from the Near-Tetm Task Force, a portion of which called for performing flood hazard reevaluations at all nuclear power plants in the United States. Reference 2 indicated plans to comply with the requested response date of March 12, 2015 for flood hazard evaluation. | |
| Reference 3 provided the required Flood Hazard Reevaluation Report for the Susquehanna Steam Electric Station (SSES), Units 1 and 2. The NRC Audit of the SSES Flood Hazard Reevaluation Report identified additional information needed by the staff to complete the Audit. The information request was discussed during an Audit teleconference held on September 9, 2015. The request and SSES response to each request is provided in the Enclosure. | | Reference 3 provided the required Flood Hazard Reevaluation Report for the Susquehanna Steam Electric Station (SSES), Units 1 and 2. The NRC Audit of the SSES Flood Hazard Reevaluation Report identified additional information needed by the staff to complete the Audit. The information request was discussed during an Audit teleconference held on September 9, 2015. The request and SSES response to each request is provided in the Enclosure. |
| There are no new or revised regulatory commitments contained in this submittal. | | There are no new or revised regulatory commitments contained in this submittal. |
| If you have any questions regarding this submittal, please contact Mr. Jeffery N. Grisewood, | | If you have any questions regarding this submittal, please contact Mr. Jeffery N. Grisewood, Manager, Nuclear Regulatory Affairs, at (570) 542-1330. Document Control Desk PLA-7389 I declare under penalty of perjury that the foregoing is tme and correct. Executed on: |
| : Manager, Nuclear Regulatory
| |
| : Affairs, at (570) 542-1330. Document Control Desk PLA-7389 I declare under penalty of perjury that the foregoing is tme and correct.
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| Executed on: | |
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| ==Enclosure:== | | ==Enclosure:== |
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| These specific structure CDB values applicable to both the LIP and Cooling Tower Basin Rupture are delineated in FHRR Table 2-1 (Page 56) in the "Flood Banier Elevation" Column. The Peak Event LIP and Cooling Tower Basin Rupture flood levels are provided in FHRR Tables 3-1 and 3-2. The results of the 3 FHRR Tables are combined in the below Table to show, for those structures with flood barriers, that the flood CDB is greater than the peak re-evaluated flood levels for the LIP and the Cooling Tower Basin rupture events. Structure LIP and Cooling LIP Peak Flood Cooling Tower Basin Tower Basin Level Rupture Peak flood Rupture CDB Level (ftNGVD29) | | These specific structure CDB values applicable to both the LIP and Cooling Tower Basin Rupture are delineated in FHRR Table 2-1 (Page 56) in the "Flood Banier Elevation" Column. The Peak Event LIP and Cooling Tower Basin Rupture flood levels are provided in FHRR Tables 3-1 and 3-2. The results of the 3 FHRR Tables are combined in the below Table to show, for those structures with flood barriers, that the flood CDB is greater than the peak re-evaluated flood levels for the LIP and the Cooling Tower Basin rupture events. Structure LIP and Cooling LIP Peak Flood Cooling Tower Basin Tower Basin Level Rupture Peak flood Rupture CDB Level (ftNGVD29) |
| Source: FHRR Table Source: FHRR Source: FHRR 3-2 Table 2-1 Table 3-1 ESSW 694.80 685.74 686.42 Pump house (south side) Common Diesel 679.0 676.30 N/A Generator Building Unit 1 Reactor 672.0 671.36 N/A Building Unit 2 Reactor 672.0 670.91 N/A Building Common Diesel 678.0 675.27 NIA 'E' Building Enclosure to PLA-7389 Page 2 of6 For LIP and Cooling Tower Basin Rupture, Tables 3-1 and 3-2 of the FHRR report the margin (using the term "Freeboard") | | Source: FHRR Table Source: FHRR Source: FHRR 3-2 Table 2-1 Table 3-1 ESSW 694.80 685.74 686.42 Pump house (south side) Common Diesel 679.0 676.30 N/A Generator Building Unit 1 Reactor 672.0 671.36 N/A Building Unit 2 Reactor 672.0 670.91 N/A Building Common Diesel 678.0 675.27 NIA 'E' Building Enclosure to PLA-7389 Page 2 of6 For LIP and Cooling Tower Basin Rupture, Tables 3-1 and 3-2 of the FHRR report the margin (using the term "Freeboard") |
| between the CDB (Tables 2-1) and the Re-Evaluation results. | | between the CDB (Tables 2-1) and the Re-Evaluation results. These are shown in the last column ofTables 3-1 and 3-2. For example, for the ESSW Pumphouse (south side): 694.8(Table 2-1)-685.74 (Table 3-1) = 9.06. Audit Request 3 The use of CLB and CDB appear to be interchangeable-need licensee to confirm this. Response 3: The use of CLB and CDB are interchangeable. |
| These are shown in the last column ofTables 3-1 and 3-2. For example, for the ESSW Pumphouse (south side): 694.8(Table 2-1)-685.74 (Table 3-1) = 9.06. Audit Request 3 The use of CLB and CDB appear to be interchangeable-need licensee to confirm this. Response 3: The use of CLB and CDB are interchangeable. | | Note that the PLA-6938 provides a comprehensive description of the SSES Design Basis Flood Hazards. Audit Request 4 In Section 3.3.8.1, the licensee discusses failure modes for the Cooling Tower basins, but does not discuss how they were developed. |
| Note that the PLA-6938 provides a comprehensive description of the SSES Design Basis Flood Hazards. | |
| Audit Request 4 In Section 3.3.8.1, the licensee discusses failure modes for the Cooling Tower basins, but does not discuss how they were developed. | |
| Response 4: In order to determine the most credible cooling tower basin failure modes, historical cooling tower failures were reviewed, including the following: | | Response 4: In order to determine the most credible cooling tower basin failure modes, historical cooling tower failures were reviewed, including the following: |
| * The Willow Island cooling tower failure (NIST, 2014), | | * The Willow Island cooling tower failure (NIST, 2014), |
| * The Vermont Yankee Nuclear cooling tower failure (NRC, 2007), and | | * The Vermont Yankee Nuclear cooling tower failure (NRC, 2007), and |
| * The Ferrybridge cooling tower failure (Ford, 1965). Based on review of historical failure modes and an understanding that this investigation is primarily concerned with failure of the basins beneath the cooling towers rather than structural failure of the actual hyperbolic tower structures, the following two failure modes are considered to be the most credible flood-causing failures for the SSES site: | | * The Ferrybridge cooling tower failure (Ford, 1965). Based on review of historical failure modes and an understanding that this investigation is primarily concerned with failure of the basins beneath the cooling towers rather than structural failure of the actual hyperbolic tower structures, the following two failure modes are considered to be the most credible flood-causing failures for the SSES site: |
| * A collapse of one or more panels along the perimeter of one (or both) cooling tower basin(s), | | * A collapse of one or more panels along the perimeter of one (or both) cooling tower basin(s), and |
| and | | * A collapse of the headwall of the Cold Water Outlet Chamber (CWOC) of one or both of the cooling towers. The CWOCs are the intake boxes for the two nine-foot interior diameter pipes (PPL, 1982) providing suction to the cooling water circulation pumps. Panel failure hydro graphs were independently computed, regardless of panel location, which is reasonable because all panels for both Unit 1 and Unit 2 have the same dimensions (i.e. panel length of24 feet [PPL, 1981]). It is understood that failure in these modes would most likely occur due to seismic activity, with a relatively short failure time (i.e., a maximum of approximately 1 minute). |
| * A collapse of the headwall of the Cold Water Outlet Chamber (CWOC) of one or both of the cooling towers. The CWOCs are the intake boxes for the two nine-foot interior diameter pipes (PPL, 1982) providing suction to the cooling water circulation pumps. Panel failure hydro graphs were independently | |
| : computed, regardless of panel location, which is reasonable because all panels for both Unit 1 and Unit 2 have the same dimensions (i.e. panel length of24 feet [PPL, 1981]). It is understood that failure in these modes would most likely occur due to seismic activity, with a relatively short failure time (i.e., a maximum of approximately 1 minute).
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| ==References:== | | ==References:== |
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| Enclosure to PLA-7389 Page 3 of6 1. Ford, 1965, Ford, David N., "Ferrybridge Cooling Towers Collapse," | | Enclosure to PLA-7389 Page 3 of6 1. Ford, 1965, Ford, David N., "Ferrybridge Cooling Towers Collapse," "When Technology Fails: Significant Technological Disasters, Accidents, and Failures of the Twentieth Century," Edited by Neil Schlager, Gate Research, Inc., Detroit, MI, 1994. 2. NIST, 2014, National Institute of Standards and Technology, Willow Island Cooling Tower Failure, West Virginia, 1978, Website: http :I lwww .nist. gov I ell disasterstudiesl construction/failure_ |
| "When Technology Fails: Significant Technological Disasters, Accidents, and Failures of the Twentieth Century," | | cooling_ tower _1978 .cfm, Date of Publication, August 2011, Date Accessed, October 7, 2014. 3. NRC, 2007, Preliminary Notification of Event or Unusual Occunence-PNO-I-07-008, September 14, 2007. 4. PPL, 1981, Pennsylvania Power & Light Company (PPL) Susquehanna LLC, Cottrell, Inc. Hamon Cooling Tower Division, General Arrangement Plan, Drawing No. CT-220-60 1. 5. PPL, 1982, Pennsylvania Power & Light Company (PPL) Susquehanna LLC, Cottrell, Inc. Hamon Cooling Tower Division, Cold Water Outlet II, Drawing No. CT-220-317. |
| Edited by Neil Schlager, Gate Research, Inc., Detroit, MI, 1994. 2. NIST, 2014, National Institute of Standards and Technology, Willow Island Cooling Tower Failure, West Virginia, 1978, Website: | | Audit Request 5 Enclosure to PLA-7389 Page 4 of6 Provide on the docket the following schematics from RIZZO Calculation 12-4834 F-06, Revision 0 Response 5: Bas i n Wa ll Foundat i on P i e r s , 33ft on center a l ong t ota l pe ri mete r /A A._ * -/o = 205.25 ft Bas i n Pane l to brea k H Y; h=l.S ft I Bas i n W a ll Founda ti on 1 P i er ,_ S o i l w edgea SE u m ed to be s c oured out b yf a il u r e flcv; SECTION A-A APPLICATION OF EQUATION 2-1 TO CT BASIN PANEL COLLAPSE-SCHEMATIC DRAWING NOT TO ANY SCALE F ill D i rt Around Bas i n (notto s ca l e) Top o f Bas i n Pe r i meter Wa ll (E l evat i on= 26 ft) Bonom o f Bas i n E l ev. = 1S.5 ft Foundat i on block Re f e r ence l eve i (E i evat i on=O) T op of Ou tl et Bo x Section A View Enclosure to PLA-7389 Page 5 of6 Co l d Water Out l et Chamber Fa il ed I nvert o f 9ft D i ameter Out l et P i pes (E l evat i on =0) SCHEMATIC DIAGRAM OF HYDRAULIC PARAMETERS FOR CALCULATION OF COLD WATER OUTLET FAILURE HYDROGRAPH |
| http :I lwww .nist. gov I ell disasterstudiesl construction/failure_ | |
| cooling_ | |
| tower _1978 .cfm, Date of Publication, August 2011, Date Accessed, October 7, 2014. 3. NRC, 2007, Preliminary Notification of Event or Unusual Occunence-PNO-I-07-008, September 14, 2007. 4. PPL, 1981, Pennsylvania Power & Light Company (PPL) Susquehanna | |
| : LLC, Cottrell, Inc. Hamon Cooling Tower Division, General Arrangement Plan, Drawing No. CT-220-60 1. 5. PPL, 1982, Pennsylvania Power & Light Company (PPL) Susquehanna
| |
| : LLC, Cottrell, Inc. Hamon Cooling Tower Division, Cold Water Outlet II, Drawing No. CT-220-317.
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| Audit Request 5 Enclosure to PLA-7389 Page 4 of6 Provide on the docket the following schematics from RIZZO Calculation 12-4834 F-06, Revision 0 Response 5: Basin Wall Foundation Piers, 33ft on center along total perimeter /A A._ * -/o = 205.25 ft Basin Panel to break H Y; h=l.Sft I BasinWall Foundation 1 Pier ,_ SoilwedgeaSEumed to be scoured out byfailureflcv; SECTION A-A APPLICATION OF EQUATION 2-1 TO CT BASIN PANEL COLLAPSE-SCHEMATIC DRAWING NOT TO ANY SCALE Fill Dirt Around Basin (notto scale) Top of Basin Perimeter Wall (Elevation= 26ft) Bonom of Basin Elev. = 1S.5 ft Foundation block Reference levei(Eievation=O) Top of Outlet Box Section A View Enclosure to PLA-7389 Page 5 of6 Cold Water Outlet Chamber Failed Invert of 9ft Diameter Outlet Pipes (Elevation =0) SCHEMATIC DIAGRAM OF HYDRAULIC PARAMETERS FOR CALCULATION OF COLD WATER OUTLET FAILURE HYDROGRAPH | |
| ,. t | | ,. t |
| ' WS !Ja..:W3 .. .. I " I Headwall | | ' WS !Ja..:W3 .. .. I " I Headwall \ CTBasin Failure l cwoc I I " ' I 1S ;; I i ! .. I 205.25 ft to I I CTBasin I Dummy ,. Center Reach ' s _. / -,_ ----* , ___ i r-274.00 ft to CTBasin I Center ., *S 0 ,. .. .. .. ... , ,. ... 160 11.J nCtl.l nnri OGta a cctfl1 HEC-RAS PROFILE AND LOCATION OF PRINCIPAL MODEL ELEMENTS Enclosure to PLA-7389 Page 6 of6 _,__ ii,_. ___ J " I I I I '*: Panel Failure Locations for Critical Failure Scenarios |
| \ CTBasin Failure l cwoc I I " ' I 1S ;; I i ! .. I 205.25 ftto I I CTBasin I Dummy ,. Center Reach ' s _. / -,_ ----* , ___ i r-274.00ftto CTBasin I Center ., *S 0 ,. .. .. .. ... ,,. ... 160 11.JnCtl.lnnriOGtaacctfl1 HEC-RAS PROFILE AND LOCATION OF PRINCIPAL MODEL ELEMENTS Enclosure to PLA-7389 Page 6 of6 _,__ ii,_. ___ J "I I I I '*: Panel Failure Locations for Critical Failure Scenarios | |
| -scenario1 | | -scenario1 |
| -scenario2 =Scenario 3 0 =Scenario4 150 300 Feet LOCATION OF FAILURES FOR SCENARIOS 1 THROUGH 4 N A}} | | -scena r io2 =Scena r io 3 0 =Scenario4 150 300 Feet LOCATION OF FAILURES FOR SCENARIOS 1 THROUGH 4 N A}} |
Letter Sequence Request |
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MONTHYEARPLA-7287, Flood Hazards Reevaluation Report PLA-72872015-03-0303 March 2015
Flood Hazards Reevaluation Report PLA-7287
Project stage: Request
ML15153A1632015-06-0404 June 2015
Nuclear Regulatory Commission Plan for the Audit of PPL Susquehanna Llc'S Flood Hazard Reevaluation Report Submittal Relating to the Near-Term Task Force Recommendation 2.1 - Flooding for Susquehanna Steam Electric Station, Units 1 and 2.
Project stage: Other
PLA-7389, Flood Hazards Reevaluation Report, Information to Support Audit2015-09-24024 September 2015
Flood Hazards Reevaluation Report, Information to Support Audit
Project stage: Request
ML15288A5632015-10-0202 October 2015
NRR E-mail Capture - (External_Sender) Susquehanna: Follow-on Audit Question
Project stage: Request
ML15281A1562015-10-21021 October 2015
U. S. Nuclear Regulatory Commission Report for the Audit of Susquehanna Nuclear, Llc'S Flood Hazard Reevaluation Report Submittals Relating to the Near-Term Task Force Recommendation 2.1 - Flooding for Susquehanna Units 1 and 2 (Tac. MF6037
Project stage: Other
ML15303A3142015-11-0303 November 2015
Interim Staff Response to Reevaluated Flood Hazards Submitted in Response to 10CFR 50.54(f) Information Request -Flood Causing Mechanism Reevaluation
Project stage: Other
ML15239B2212015-11-0303 November 2015
Master Report - Tables 1 and 2
Project stage: Request
ML15314A7472015-11-12012 November 2015
Correction to Interim Staff Response to Reevaluated Flood Hazards Submitted in Response to 10 CFR 50.54(f) Information Request Flood-Causing Mechanism Reevaluation
Project stage: Other
2015-11-03
[Table View] |
|
|---|
Category:Letter
MONTHYEARPLA-8145, Biennial 10 CFR 50.59 and 72.48 Summary Report and Changes to Regulatory Commitments - PLA-81452024-10-21021 October 2024
Biennial 10 CFR 50.59 and 72.48 Summary Report and Changes to Regulatory Commitments - PLA-8145
ML24291A1562024-10-16016 October 2024
Missed Annual Inventory Required by 40 CFR 266, Subpart in PLE 0026645
PLA-8148, Registration for the Use of Spent Fuel Storage Casks 311, 308, and 3102024-10-15015 October 2024
Registration for the Use of Spent Fuel Storage Casks 311, 308, and 310
PLA-8141, Response to Request for Additional Information Regarding Relief Request 1RR06, PLA-81412024-09-18018 September 2024
Response to Request for Additional Information Regarding Relief Request 1RR06, PLA-8141
PLA-8142, Registration for the Use of Spent Fuel Storage Casks 306, 309, and 307 - PLA-81422024-09-18018 September 2024
Registration for the Use of Spent Fuel Storage Casks 306, 309, and 307 - PLA-8142
ML24260A2312024-09-17017 September 2024
Senior Reactor and Reactor Operator Initial License Examinations
05000387/LER-2024-002, B Diesel Generator Inoperable Due to Failed Excitation System Linear Reactor2024-09-16016 September 2024
B Diesel Generator Inoperable Due to Failed Excitation System Linear Reactor
ML24255A8642024-09-0606 September 2024
Rscc Wire & Cable LLC Dba Marmon Industrial Energy & Infrastructure - Part 21 Retraction of Final Notification
ML24249A1362024-09-0404 September 2024
EN 57304 - Westinghouse Electric Company, LLC, Final Report - No Embedded Files. Notification of the Potential Existence of Defects Pursuant to 10 CFR Part 21
ML24233A2192024-09-0303 September 2024
– Authorized Alternative to Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code
IR 05000387/20240052024-08-29029 August 2024
Updated Inspection Plan for Susquehanna Steam Electric Station, Units 1 and 2 (Report 05000387/2024005 and 05000388/2024005)
05000387/LER-2023-004-01, Manual Reactor Scram Due to Degraded Main Condenser Vacuum2024-08-26026 August 2024
Manual Reactor Scram Due to Degraded Main Condenser Vacuum
05000387/LER-2024-001-01, Main Steam Isolation Valve Leakage Due to Valve Body Seat Wear2024-08-21021 August 2024
Main Steam Isolation Valve Leakage Due to Valve Body Seat Wear
IR 05000387/20240022024-08-12012 August 2024
Integrated Inspection Report 05000387/2024002 and 05000388/2024002
ML24208A0962024-07-25025 July 2024
57243-EN 57243 - Rssc Wire & Cable LLC, Dba Marmon - Part 21 Notification
PLA-8117, 23rd Refueling Outage Owners Activity Report (PLA-8117)2024-07-23023 July 2024
23rd Refueling Outage Owners Activity Report (PLA-8117)
ML24197A0982024-07-15015 July 2024
Request for Information for a Biennial Problem Identification and Resolution Inspection; Inspection Report 05000387/2024010 and 05000388/2024010
ML24127A2262024-05-29029 May 2024
Issuance of Amendment Nos. 288 and 272 Adoption of TSTF-563
PLA-8126, Response to Request for Confirmation of Information Regarding Relief Request 1RR062024-05-29029 May 2024
Response to Request for Confirmation of Information Regarding Relief Request 1RR06
PLA-8122, Annual Radiological Environmental Operating Report (PLA-8122)2024-05-28028 May 2024
Annual Radiological Environmental Operating Report (PLA-8122)
PLA-8123, Main Steam Isolation Valve Leakage2024-05-23023 May 2024
Main Steam Isolation Valve Leakage
PLA-8115, Relief Request IRR06 One Time Extension to the Fourth 10-Year Inservice Testing Program Interval (PLA-8115)2024-05-23023 May 2024
Relief Request IRR06 One Time Extension to the Fourth 10-Year Inservice Testing Program Interval (PLA-8115)
IR 05000387/20240012024-05-13013 May 2024
Integrated Inspection Report 05000387/2024001 and 05000388/2024001
IR 05000387/20244042024-05-0707 May 2024
Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000387/2024404 and 05000388/2024404
ML24082A1372024-04-22022 April 2024
Issuance of Amendment Nos. 287 and 271 Adoption of TSTF-568, Revision 2 and Associated Technical Specification Changes
PLA-8094, Radioactive Effluent Release Report and Offsite Dose Calculation Manual PLA-80942024-04-22022 April 2024
Radioactive Effluent Release Report and Offsite Dose Calculation Manual PLA-8094
PLA-8095, 2023 Annual Radiological Environmental Operating Report (PLA-8095)2024-04-22022 April 2024
2023 Annual Radiological Environmental Operating Report (PLA-8095)
PLA-8101, Re 2023 Annual Report of Radiation Exposure2024-04-22022 April 2024
Re 2023 Annual Report of Radiation Exposure
PLA-8102, Annual Environmental Operating Report (Nonradiological) PLA-81022024-04-11011 April 2024
Annual Environmental Operating Report (Nonradiological) PLA-8102
PLA-8113, Response to Request for Additional Information Regarding Relief Request 4RR-11 (PLA-8113)2024-04-11011 April 2024
Response to Request for Additional Information Regarding Relief Request 4RR-11 (PLA-8113)
PLA-8112, Relief Request 4RR-11 Relief from End of Interval Boundary Leakage Test PLA-81122024-04-0909 April 2024
Relief Request 4RR-11 Relief from End of Interval Boundary Leakage Test PLA-8112
PLA-8110, Submittal of Unit 1 Cycle 24 Core Operating License Report (Pla 8110)2024-04-0404 April 2024
Submittal of Unit 1 Cycle 24 Core Operating License Report (Pla 8110)
PLA-8100, Property Insurance Program (PLA-8100)2024-04-0101 April 2024
Property Insurance Program (PLA-8100)
ML24092A4022024-04-0101 April 2024
Annual Financial Report (PLA-8098)
PLA-8109, Supplement to Request for Exemption from Certain Requirements of 10 CR 72.212 and 10 CFR 72.214 Resulting from Fuel Basket Design Control Compliance (PLA-8109)2024-03-21021 March 2024
Supplement to Request for Exemption from Certain Requirements of 10 CR 72.212 and 10 CFR 72.214 Resulting from Fuel Basket Design Control Compliance (PLA-8109)
ML24067A2512024-03-19019 March 2024
Authorized Alternative to Requirements of the ASME Code
PLA-8107, Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 Resulting from Fuel Basket Design Control Compliance2024-03-19019 March 2024
Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 Resulting from Fuel Basket Design Control Compliance
ML24044A2532024-03-14014 March 2024
Associated Independent Spent Fuel Storage Installation – Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0077 (Security Notifications, Reports, and Recordkeeping & Suspicious Activity Reporting))
IR 05000387/20240112024-02-29029 February 2024
Commercial Grade Dedication Inspection Report 05000387/2024011 and 05000388/2024011
IR 05000387/20230062024-02-28028 February 2024
Annual Assessment Letter for Susquehanna Steam Electric Station, Units 1 and 2 (Reports 05000387/2023006 and 05000388/2023006)
ML24039A1882024-02-27027 February 2024
Issuance of Amendment Nos. 286 and 270 Changes to Technical Specifications for Control Rods
ML24037A3072024-02-22022 February 2024
Summary of Regulatory Audit in Support of Relief Request 5RR-02
PLA-8099, Proof of Financial Protection and Guarantee of Payment of Deferred Premiums (PLA-8099)2024-02-13013 February 2024
Proof of Financial Protection and Guarantee of Payment of Deferred Premiums (PLA-8099)
IR 05000387/20230042024-02-0707 February 2024
Integrated Inspection Report 05000387/2023004 and 05000388/2023004
PLA-8096, Response to Request for Additional Information Regarding Proposed Relief Request for the Fifth 10-Year Inservice Test Program Interval2024-01-0404 January 2024
Response to Request for Additional Information Regarding Proposed Relief Request for the Fifth 10-Year Inservice Test Program Interval
PLA-8077, Emergency Plan Revision 67 (PLA-8077)2023-12-27027 December 2023
Emergency Plan Revision 67 (PLA-8077)
IR 05000387/20230102023-12-11011 December 2023
Fire Protection Team Inspection Report 05000387/2023010 and 05000388/2023010
PLA-8088, Request for Exemption from Enhanced Weapons, Firearms Background Checks and Security Event Notifications Implementation (PLA-8088)2023-12-0505 December 2023
Request for Exemption from Enhanced Weapons, Firearms Background Checks and Security Event Notifications Implementation (PLA-8088)
PLA-8089, Submittal of Revision to Inservice Testing Program Plan2023-12-0505 December 2023
Submittal of Revision to Inservice Testing Program Plan
PLA-8084, Application to Revise Technical Specifications to Adopt TSTF-568, Revise Applicability of BWR TS 3.6.2.5 and TS 3.6.3.22023-11-29029 November 2023
Application to Revise Technical Specifications to Adopt TSTF-568, Revise Applicability of BWR TS 3.6.2.5 and TS 3.6.3.2
2024-09-06
[Table view]
Category:Response to Request for Additional Information (RAI)
MONTHYEARPLA-8141, Response to Request for Additional Information Regarding Relief Request 1RR06, PLA-81412024-09-18018 September 2024
Response to Request for Additional Information Regarding Relief Request 1RR06, PLA-8141
PLA-8113, Response to Request for Additional Information Regarding Relief Request 4RR-11 (PLA-8113)2024-04-11011 April 2024
Response to Request for Additional Information Regarding Relief Request 4RR-11 (PLA-8113)
PLA-8096, Response to Request for Additional Information Regarding Proposed Relief Request for the Fifth 10-Year Inservice Test Program Interval2024-01-0404 January 2024
Response to Request for Additional Information Regarding Proposed Relief Request for the Fifth 10-Year Inservice Test Program Interval
PLA-8091, Response to Request for Additional Information Regarding Proposed Relief Request for the Fifth 10-Year Inseervice Inspection Interval (PLA-8091)2023-11-0808 November 2023
Response to Request for Additional Information Regarding Proposed Relief Request for the Fifth 10-Year Inseervice Inspection Interval (PLA-8091)
PLA-8048, Response to Request for Additional Information Regarding License Amendment Requesting Temporary Addition of Analyzed Rod Position Sequence2023-01-14014 January 2023
Response to Request for Additional Information Regarding License Amendment Requesting Temporary Addition of Analyzed Rod Position Sequence
PLA-8013, Response to Request for Additional Information Regarding License Amendment Requesting Adoption of TSTF-505, Revision 2 PLA-80132022-06-27027 June 2022
Response to Request for Additional Information Regarding License Amendment Requesting Adoption of TSTF-505, Revision 2 PLA-8013
PLA-8005, Response to Request for Additional Information Regarding License Amendment to Revise Reactor Steam Dome Pressure - Low Instrumentation Function Allowable Value (PLA-8005)2022-05-23023 May 2022
Response to Request for Additional Information Regarding License Amendment to Revise Reactor Steam Dome Pressure - Low Instrumentation Function Allowable Value (PLA-8005)
PLA-7984, Supplement to License Amendment Requesting Adoption of TSTF-505, Revision 22022-03-0808 March 2022
Supplement to License Amendment Requesting Adoption of TSTF-505, Revision 2
PLA-7865, Response to Request for Additional Information Regarding Proposed License Amendment Requesting Revision to the Dose Consequence Analysis for a Loss of Coolant Accident (PLA-7865) Loss of Coolant Accident2020-06-0202 June 2020
Response to Request for Additional Information Regarding Proposed License Amendment Requesting Revision to the Dose Consequence Analysis for a Loss of Coolant Accident (PLA-7865) Loss of Coolant Accident
PLA-7853, Ninety-Day Response to Request for Additional Information Regarding Proposed License Amendment Requesting Application of Advanced Framatome Methodologies PLA-78532020-04-0101 April 2020
Ninety-Day Response to Request for Additional Information Regarding Proposed License Amendment Requesting Application of Advanced Framatome Methodologies PLA-7853
PLA-7841, Thirty-Day Response to Request for Additional Information Regarding Proposed License Amendment Requesting Application of Advanced Framatome Methodologies2020-02-0606 February 2020
Thirty-Day Response to Request for Additional Information Regarding Proposed License Amendment Requesting Application of Advanced Framatome Methodologies
PLA-7830, Response to Second Request for Additional Information Regarding Proposed License Amendment Requesting a Temporary Change to the Technical Specifications to Allow Replacement of Emergency Service Water System Piping2019-12-0909 December 2019
Response to Second Request for Additional Information Regarding Proposed License Amendment Requesting a Temporary Change to the Technical Specifications to Allow Replacement of Emergency Service Water System Piping
PLA-7793, Response to Request for Additional Information Regarding Proposed License Amendment Requesting a Temporary Change to the Technical Specifications to Allow Replacement of Emergency Service Water System Piping2019-06-0303 June 2019
Response to Request for Additional Information Regarding Proposed License Amendment Requesting a Temporary Change to the Technical Specifications to Allow Replacement of Emergency Service Water System Piping
PLA-7704, Response to Generic Letter 2016-01, Request for Supplemental Information2018-05-24024 May 2018
Response to Generic Letter 2016-01, Request for Supplemental Information
PLA-7701, Response to NRC Issue Summary 2018-02, Preparation and Scheduling of Operator Licensing Examination.2018-04-11011 April 2018
Response to NRC Issue Summary 2018-02, Preparation and Scheduling of Operator Licensing Examination.
PLA-7673, Proposed Amendment to Licenses NPF-14 and NPF-22: Response to Request for Additional Information and Supplement to Application to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control PLA-76732018-02-16016 February 2018
Proposed Amendment to Licenses NPF-14 and NPF-22: Response to Request for Additional Information and Supplement to Application to Adopt TSTF-542, Reactor Pressure Vessel Water Inventory Control PLA-7673
PLA-7655, Response to Request for Additional Information, (CAC Nos. MF9131 and MF9132) PLA-76552017-12-0404 December 2017
Response to Request for Additional Information, (CAC Nos. MF9131 and MF9132) PLA-7655
PLA-7619, Response to Request for Additional Information Regarding License Amendment Request to Revise Diesel Generator Surveillance Requirements with New Steady State Voltage and Frequency Limits2017-08-0404 August 2017
Response to Request for Additional Information Regarding License Amendment Request to Revise Diesel Generator Surveillance Requirements with New Steady State Voltage and Frequency Limits
PLA-7583, Response to NRC Request for Supplemental Information for License Amendment Request to Revise Diesel Generator Surveillance Requirements with New Steady State Voltage and Frequency Limits (PLA-7583)2017-03-21021 March 2017
Response to NRC Request for Supplemental Information for License Amendment Request to Revise Diesel Generator Surveillance Requirements with New Steady State Voltage and Frequency Limits (PLA-7583)
PLA-7518, Response to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools.2016-10-31031 October 2016
Response to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools.
PLA-7537, Response to Request for Additional Information License Amendment Request Extending Completion Times in Support of 480V Ess Load Center Transformer Replacements2016-10-10010 October 2016
Response to Request for Additional Information License Amendment Request Extending Completion Times in Support of 480V Ess Load Center Transformer Replacements
ML16097A4872016-04-0606 April 2016
Supplemental Information Needed for Acceptance of Requested Licensing Action Amendment Request for Temporary Change of Technical Specifications 3.7.1 and 3.8.7, Applicable Drawings
PLA-7418, Response to Request for Additional Information for the Third Interval Relief Requests 3RR-19, 3RR-20 and 3RR-212015-12-10010 December 2015
Response to Request for Additional Information for the Third Interval Relief Requests 3RR-19, 3RR-20 and 3RR-21
PLA-7406, Response to Request for Additional Information on Technical Specification Changes to Adopt Traveler TSTF-4252015-11-11011 November 2015
Response to Request for Additional Information on Technical Specification Changes to Adopt Traveler TSTF-425
ML15296A0602015-10-16016 October 2015
Enclosure 3, Revised (Clean) Copy of the SSES EAL Basis Document Which Includes the Changes Made in Enclosure 2
ML15296A0592015-10-16016 October 2015
Enclosure 1, Response to NRC Request for Additional Information Regarding License Amendment Request to Adopt Nuclear Energy Institute 99-0 1, Revision 6 and Enclosure 2, Mark-up of the Changes Made to the SSES EAL Basis..
ML15296A0532015-10-15015 October 2015
Enclosure 4 to PLA-7399 Revised (Clean) Copy of the SSES EAL Basis Document
ML15296A0522015-10-15015 October 2015
Attachments 1 Through 9 - EP-RM-004 to Enclosure 3 to PLA-7399 Mark-up of Proposed Additional Changes Made to the SSES EAL Basis Document
ML15296A0502015-10-15015 October 2015
Enclosures 1 and 2 to PLA-7399 - List of Proposed Additional Changes to the Susquehanna Steam Electric Station Emergency Action Level Basis Document and Mark-up of Proposed Additional Changes Made to the SSES EAL Comparison Matrix (Revision
PLA-7399, Proposed Additional Changes to the SSES Emergency Plan Basis Document Since Submittal of Response to NRC Request for Additional Information PLA-73992015-10-15015 October 2015
Proposed Additional Changes to the SSES Emergency Plan Basis Document Since Submittal of Response to NRC Request for Additional Information PLA-7399
PLA-7389, Flood Hazards Reevaluation Report, Information to Support Audit2015-09-24024 September 2015
Flood Hazards Reevaluation Report, Information to Support Audit
PLA-7381, Response to Request for Additional Information on Technical Specification Changes to Adopt Traveler TSTF-4252015-09-21021 September 2015
Response to Request for Additional Information on Technical Specification Changes to Adopt Traveler TSTF-425
PLA-7371, Response to Request for Supplemental Information for the Third Interval Relief Requests 3RR-19, 3RR-20, and 3RR-212015-08-0606 August 2015
Response to Request for Supplemental Information for the Third Interval Relief Requests 3RR-19, 3RR-20, and 3RR-21
PLA-7334, Response to Request for Additional Information on Technical Specification Changes to Adopt Traveler TSTF-4252015-07-0202 July 2015
Response to Request for Additional Information on Technical Specification Changes to Adopt Traveler TSTF-425
ML18024A1091978-09-0808 September 1978
Letter Attaching a Schedule for Responses to NRC Question List of June 5, 1978
ML18026A1161978-08-31031 August 1978
Letter Responding to the Letter of May 17, 1978 Enclosing Information for Antitrust Review of Operating License.
ML18025A5151978-08-31031 August 1978
Letter Regarding Information for Antitrust Review of Operating License Application
ML18024A0921978-04-14014 April 1978
Letter Responding to the Commission Request for Additional Information with a Schedule for Response on Containment
ML18024A0891978-03-31031 March 1978
Letter Replying to the Questionnaire Contained in the December 6, 1977 Letter and Attaching an Updated Response for Unit 1 and a Complete Response for Unit 2
ML18024A0851978-03-0707 March 1978
Letter Regarding the Commission Letter of February 15, 1978 Containing the Staff Position on the Use of Austenitic Stainless Steel and Advising That PP&L Will Provide a Response by September 1, 1978
ML18025A2481977-06-14014 June 1977
Letter Submitting Additional Information Relative to the Commission'S Previous Request to Establish a Cold Weather Concrete Freeze-Protection Period of Three Days
ML18025A2611977-01-19019 January 1977
Letter Documenting Responses to Questions Given to Mr. Singh Bawa of the NRC from Pp&L'S Mr. E.D. Testa in Telephone Conversations on 1/17/1977 and 01/18/1977
ML18025A2641976-12-30030 December 1976
Letter Responding to Letters Requesting That Certain Information Be Submitted to Address Anticipated Transients Without Scram (ATWS) for SSES Including Analysis and Justification of the GE Analysis Model ...
ML18023A9051976-11-18018 November 1976
Response to Four Questions on the Susquehanna-Siegfried 500 Kv Line
ML18023B4951976-11-18018 November 1976
Letter Responding the November 3, 1976 Letter with Answer to Four Questions on the Susquehanna-Siegfried 500 Kv Line as Indicated in Amendment 5 and Attaching PA Dept. of Environmental Resources Approval for the Crossing ...
ML18025A4871976-10-15015 October 1976
Responses to Questions with Attached Drawings and Maps
ML18023B4961976-10-15015 October 1976
Letter Responding to the October 6, 1976 Letter Requesting Additional Information on SSES Transmission Lines
ML18025A2741976-09-0909 September 1976
Letter Responding to the Commission Letter of August 9, 1976 Relating to Annulus Pressurization and Cracks in the Feedwater Nozzle Blend Radii
ML18025A4951975-06-0505 June 1975
Letter Responding to the April 17, 1975 Letter Requesting Additional Information Relative to the Design of the Containment for SSES and Attaching the Mark II Containment Program and Schedule
ML18023A7651975-05-13013 May 1975
Letter Responding to NRC Letters Dated February 18, 1975 and March 14, 1975 Containing NRC Staff Positions
2024-09-18
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Timothy S. Rausch President and Chief Nuclear Officer SEP 2 4 l015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, P A 18603 Tel. 570.542.3445 F a x 570.542.1504 Timothy.Rausch
@t a lenenergy.com SUSQUEHANNA STEAM ELECTRIC STATION FLOOD HAZARDS REEVALUATION REPORT INFORMATION TO SUPPORT AUDIT PLA-7389 EN ERG Y 10 CFR2.202 Docket Nos. 50-387 and 50-388
References:
I. NRC Letter, Request for Information Pursuant to Title IO of the Code of Federal Regulations 50.54(/) Regarding Recommendations 2.I, 2.3, and 9.3, of the Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, dated March I2, 20I2 2. PPL Letter (PLA-6867), Response to Request for Information Pursuant to Title IO of the Code of F e deral Regulations 50.54(/) Regarding the Flooding Aspects of Recommendations 2.I and 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated June II, 20I2 3. PPL Letter (PPL-7287), Flood Hazards Reevaluation Report, dated March 3, 20I5 The United States Nuclear Regulatory Commission (NRC) issued Reference 1 on March 12, 2012, pursuant to Title 10 of the Code of Federal Regulations (CFR), Section 50.54(f), related to the implementation of Recommendations 2.1, 2.3, and 9.3 from the Near-Tetm Task Force, a portion of which called for performing flood hazard reevaluations at all nuclear power plants in the United States. Reference 2 indicated plans to comply with the requested response date of March 12, 2015 for flood hazard evaluation.
Reference 3 provided the required Flood Hazard Reevaluation Report for the Susquehanna Steam Electric Station (SSES), Units 1 and 2. The NRC Audit of the SSES Flood Hazard Reevaluation Report identified additional information needed by the staff to complete the Audit. The information request was discussed during an Audit teleconference held on September 9, 2015. The request and SSES response to each request is provided in the Enclosure.
There are no new or revised regulatory commitments contained in this submittal.
If you have any questions regarding this submittal, please contact Mr. Jeffery N. Grisewood, Manager, Nuclear Regulatory Affairs, at (570) 542-1330. Document Control Desk PLA-7389 I declare under penalty of perjury that the foregoing is tme and correct. Executed on:
Enclosure:
Susquehanna Steam Electric Station Flood Hazards Reevaluation Report Copy: NRC Region I Mr. J. E. Greives, NRC Sr. Resident Inspector Mr. J.D. Hughey, NRC Project Manager Mr. M. Shields, PA DEP/BRP Mr. J. A. Whited, NRC Project Manager Tekia Govan NRC Project Manager Enclosure to PLA-7389 Susquehanna Steam Electric Station Flood Hazards Reevaluation Report Audit Request 1 Provide the LIP CDB values. Audit Request 2 Provide the Cooling Tower Basin CDB values Response 1 and 2: Enclosure to PLA-7389 Page 1 of6 The Susquehanna Steam Electric Station (SSES) current design base (CDB) credited flood barrier values for Local Intense Precipitation (LIP) and Cooling Tower Basin Rupture were developed in response to the Fukushima event and described briefly in the last two paragraphs in the "Design Basis Flood Hazards" section of the SSES letter to NRC PLA-6938 dated November 12, 2012. PLA-6938 indicates that flood barrier design capability was established and documented in station flooding analyses.
Flood barrier drawings document the external passive flood design barrier capability.
Table 2-3 and 4-3 of the Flood Hazard Reevaluation Report (FHRR) identify that the CLB does not report the Flood levels/depths for LIP and Cooling Tower Basin Rupture flood causing mechanisms.
Though not reported in the CLB, the flood barrier drawings and analyses do report and establish the CDB external passive flood design barrier capability.
These specific structure CDB values applicable to both the LIP and Cooling Tower Basin Rupture are delineated in FHRR Table 2-1 (Page 56) in the "Flood Banier Elevation" Column. The Peak Event LIP and Cooling Tower Basin Rupture flood levels are provided in FHRR Tables 3-1 and 3-2. The results of the 3 FHRR Tables are combined in the below Table to show, for those structures with flood barriers, that the flood CDB is greater than the peak re-evaluated flood levels for the LIP and the Cooling Tower Basin rupture events. Structure LIP and Cooling LIP Peak Flood Cooling Tower Basin Tower Basin Level Rupture Peak flood Rupture CDB Level (ftNGVD29)
Source: FHRR Table Source: FHRR Source: FHRR 3-2 Table 2-1 Table 3-1 ESSW 694.80 685.74 686.42 Pump house (south side) Common Diesel 679.0 676.30 N/A Generator Building Unit 1 Reactor 672.0 671.36 N/A Building Unit 2 Reactor 672.0 670.91 N/A Building Common Diesel 678.0 675.27 NIA 'E' Building Enclosure to PLA-7389 Page 2 of6 For LIP and Cooling Tower Basin Rupture, Tables 3-1 and 3-2 of the FHRR report the margin (using the term "Freeboard")
between the CDB (Tables 2-1) and the Re-Evaluation results. These are shown in the last column ofTables 3-1 and 3-2. For example, for the ESSW Pumphouse (south side): 694.8(Table 2-1)-685.74 (Table 3-1) = 9.06. Audit Request 3 The use of CLB and CDB appear to be interchangeable-need licensee to confirm this. Response 3: The use of CLB and CDB are interchangeable.
Note that the PLA-6938 provides a comprehensive description of the SSES Design Basis Flood Hazards. Audit Request 4 In Section 3.3.8.1, the licensee discusses failure modes for the Cooling Tower basins, but does not discuss how they were developed.
Response 4: In order to determine the most credible cooling tower basin failure modes, historical cooling tower failures were reviewed, including the following:
- The Ferrybridge cooling tower failure (Ford, 1965). Based on review of historical failure modes and an understanding that this investigation is primarily concerned with failure of the basins beneath the cooling towers rather than structural failure of the actual hyperbolic tower structures, the following two failure modes are considered to be the most credible flood-causing failures for the SSES site:
- A collapse of one or more panels along the perimeter of one (or both) cooling tower basin(s), and
- A collapse of the headwall of the Cold Water Outlet Chamber (CWOC) of one or both of the cooling towers. The CWOCs are the intake boxes for the two nine-foot interior diameter pipes (PPL, 1982) providing suction to the cooling water circulation pumps. Panel failure hydro graphs were independently computed, regardless of panel location, which is reasonable because all panels for both Unit 1 and Unit 2 have the same dimensions (i.e. panel length of24 feet [PPL, 1981]). It is understood that failure in these modes would most likely occur due to seismic activity, with a relatively short failure time (i.e., a maximum of approximately 1 minute).
References:
Enclosure to PLA-7389 Page 3 of6 1. Ford, 1965, Ford, David N., "Ferrybridge Cooling Towers Collapse," "When Technology Fails: Significant Technological Disasters, Accidents, and Failures of the Twentieth Century," Edited by Neil Schlager, Gate Research, Inc., Detroit, MI, 1994. 2. NIST, 2014, National Institute of Standards and Technology, Willow Island Cooling Tower Failure, West Virginia, 1978, Website: http :I lwww .nist. gov I ell disasterstudiesl construction/failure_
cooling_ tower _1978 .cfm, Date of Publication, August 2011, Date Accessed, October 7, 2014. 3. NRC, 2007, Preliminary Notification of Event or Unusual Occunence-PNO-I-07-008, September 14, 2007. 4. PPL, 1981, Pennsylvania Power & Light Company (PPL) Susquehanna LLC, Cottrell, Inc. Hamon Cooling Tower Division, General Arrangement Plan, Drawing No. CT-220-60 1. 5. PPL, 1982, Pennsylvania Power & Light Company (PPL) Susquehanna LLC, Cottrell, Inc. Hamon Cooling Tower Division, Cold Water Outlet II, Drawing No. CT-220-317.
Audit Request 5 Enclosure to PLA-7389 Page 4 of6 Provide on the docket the following schematics from RIZZO Calculation 12-4834 F-06, Revision 0 Response 5: Bas i n Wa ll Foundat i on P i e r s , 33ft on center a l ong t ota l pe ri mete r /A A._ * -/o = 205.25 ft Bas i n Pane l to brea k H Y; h=l.S ft I Bas i n W a ll Founda ti on 1 P i er ,_ S o i l w edgea SE u m ed to be s c oured out b yf a il u r e flcv; SECTION A-A APPLICATION OF EQUATION 2-1 TO CT BASIN PANEL COLLAPSE-SCHEMATIC DRAWING NOT TO ANY SCALE F ill D i rt Around Bas i n (notto s ca l e) Top o f Bas i n Pe r i meter Wa ll (E l evat i on= 26 ft) Bonom o f Bas i n E l ev. = 1S.5 ft Foundat i on block Re f e r ence l eve i (E i evat i on=O) T op of Ou tl et Bo x Section A View Enclosure to PLA-7389 Page 5 of6 Co l d Water Out l et Chamber Fa il ed I nvert o f 9ft D i ameter Out l et P i pes (E l evat i on =0) SCHEMATIC DIAGRAM OF HYDRAULIC PARAMETERS FOR CALCULATION OF COLD WATER OUTLET FAILURE HYDROGRAPH
,. t
' WS !Ja..:W3 .. .. I " I Headwall \ CTBasin Failure l cwoc I I " ' I 1S ;; I i ! .. I 205.25 ft to I I CTBasin I Dummy ,. Center Reach ' s _. / -,_ ----* , ___ i r-274.00 ft to CTBasin I Center ., *S 0 ,. .. .. .. ... , ,. ... 160 11.J nCtl.l nnri OGta a cctfl1 HEC-RAS PROFILE AND LOCATION OF PRINCIPAL MODEL ELEMENTS Enclosure to PLA-7389 Page 6 of6 _,__ ii,_. ___ J " I I I I '*: Panel Failure Locations for Critical Failure Scenarios
-scenario1
-scena r io2 =Scena r io 3 0 =Scenario4 150 300 Feet LOCATION OF FAILURES FOR SCENARIOS 1 THROUGH 4 N A
