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{{Adams
#REDIRECT [[PLA-7389, Flood Hazards Reevaluation Report, Information to Support Audit]]
| number = ML15267A600
| issue date = 09/24/2015
| title = Flood Hazards Reevaluation Report, Information to Support Audit
| author name = Rausch T S
| author affiliation = Susquehanna Nuclear, LLC, Talen Energy
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000387, 05000388
| license number =
| contact person =
| case reference number = PLA-7389
| document type = Letter, Response to Request for Additional Information (RAI)
| page count = 9
}}
 
=Text=
{{#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
@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 Willow Island cooling tower failure (NIST, 2014),
* 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:
* 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}}

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