View source for NL-13-1257, DOEJ- FRSNC419117 - M001 - Southern Nuclear Operating Company - Documentation of Engineering Judgment, CST Volume Requirements Per Plant Transient Events, Enclosure 2 to NL-13-1257

{{Adams
| number = ML13198A450
| issue date = 09/04/2012
| title = DOEJ- FRSNC419117 - M001 - Southern Nuclear Operating Company - Documentation of Engineering Judgment, CST Volume Requirements Per Plant Transient Events, Enclosure 2 to NL-13-1257
| author name = Jennings W
| author affiliation = Southern Co, Southern Nuclear Operating Co, Inc
| addressee name = 
| addressee affiliation = NRC/NRR
| docket = 05000348, 05000364
| license number = 
| contact person = 
| case reference number = NL-13-1257
| document type = Report, Technical
| page count = 28
}}

=Text=
{{#Wiki_filter:Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117 
-MOOI Southern Nuclear Operating Company Southern Nuclear Operating Company DOCUMENTATION OF ENGINEERING JUDGMENT DOEJ-FRSNC419117 
-M001 CST Volume Requirements per Plant Transient Events Version Record Version No. 2 Orlg inatorlDate 5i nature Willie Jennings I 8-14-2012 NMP*ES*039
-002. Version 2.0 Reviewer/Date 5i nature Andy Palko I 8-14-2012 DoclITIentation of Enginee ri ng Judgment, E2 -1 Enclosure 2 to NL-13-l257 DOEJ-FRSNC4t9117 
-MOOt Southern Nuclear Operating Company Purpose: The purpose of this analysis is to determine the CST volume required for the plant events that credit the AFW system to cool down the RCS or maintain the plant in a hot standby condition. Design Inputs (Reference NMP-ES-(42):
See References Below 

==References:==
: 1. Auxiliary Feedwater Functional System Description (FSD), A-181010, Ver. 23 2. Calculation BM-95-096l-00l, Version 5 -"CST Sizing Verification" 3. Technical Specification , Un it 1 -Amendment 188, Unit 2 -Amendment 183 4. Mechanical Engineering Review Manual, 8 th Edition. 5. Cameron Hydraulic Data , 16 t h Edition 6. Calculation 38.04 , Revision 4, Verif i cation of AFW Flow Bases -Unit 2 7. Calculation 40.02, Revslon 4, Verification of AFW Flow Bases -Unit 1 8. U-161693, Version 2.0 -CST General Plan 9. SCS Calculation SM-95-0721
-005 , Revision 1 -Design Basis Temperature Site/Condensate Storage Tank/Refueling Water Storage Tank. 10. Pump Suction Piping Isometric Drawings D-514547, Rev.l and D-514548, Rev. 1. 11. MDAFW Pump Suction Piping, D-518847, Sheet 2, Rev.O 12. TDAFW Pump Suction Piping, D-518847, Sheet I, Rev. 0 13. AFW Suction Line Flow Instrument Drawings , A-175856, Rev. 2 and A-205856, Rev. 3 14. AFW System P&ID, D-175007 , Rev. 31; D-205007 , Rev. 25 15. WCAP-15097 , Revision 1: FNP Units 1 and 2 Replacements Steam Genera t or Program NSSS Engineering Report Book I, March 2001. 16. U-161703B, Ver. 0.3; U-161694C, Ver. 0.4; U-213481 Ver.3.0 -CST Interna l Dimensions 
: 17. SM-SNC335993-001, Vers i on 2.0 -CST AFW Pump Suction -Submergence Analysis 18. FSAR Section 15.2.8 , Rev. 24 , June 2012 -Loss of Normal Feedwater 
: 19. WCAP -14722 , Nov. 1997: FNP Units 1 and 2 Power Uprate Project NSSS Engineering Report 2 NMP*ES*039-002. Version 2 , 0 Documentation of Eng i nee ri ng Judgment , E2 -2 Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117 
-MOO1 Southern Nuclear Operating Company 20. FSAR Section 15.2.9, Rev. 24 , June 2012 -Loss of Offsite Power 21. FSAR Section 15.3.1, Rev. 24 June 2012, -Small Break Loss of Coolant Accident 22. FSAR Section 15.4.2.1 , Rev. 24, June 2012 -Main Steam Line Break 23. FSAR Section 15.2.13 , Rev. 24, June 2012 -Depressurization of Main Steam 24. FSAR Section 15.4.2.2, Rev. 24 , June 2012, -Main Feedwater Line Break 25. FSAR Section 9.2.6 , Rev. 24, June 2012, -Condensate Storage Tank 26. Technical Specification Basis, Revision 55, Section B.3.7.6 27. Calculation 01.10, Rev. 0, Condensate Storage, 3/10f72 and Engineering Judgement M1. 10R)/01/CN 96-030 Rev. A 3 NMp*ES*039-002. Vars i oo 2.0 Documentation of Engineering Judgment , E2 -3 Enclosure 2 to NL-13-1257 DOEJ* FRSNC419117 
-MOO1 Southern Nuclear Operating Company Assumptions: 
: 1. Net RCP heat added to RCS,1 OMWt, (Reference 
: 15) during power operations is decayed over a 1 hour period in determining CST volume required for RCP heat input. 2. Assume CST inventory loss during Normal cooldown event due to AFW Recirc. Line and Instrumentation line rupture thus requiring a minimum submergence level to prevent vortex formation. 
: 3. Assume CST volume required for core decay heat removal is based on 102% of 2775 MWt (100% power) if not specified in FSAR Chapter 15 -Accident Analysis. 
: 4. To envelop all possible scenarios of decay heat removal, the worst case steam generator back pressure (highest) from Reference 1 will be used to determine the final enthalpy of steam released from the steam generators. 
: 5. During cooldown the steam generator pressure decreases from the hot standby condition, 1155 psia to 135 psia (saturation pressure at 350 OF). The enthalpy varies 1204 Btu/lb to an enthalpy of 1187 Btu/lb. The lower enthalpy, 1187 Btu/lb is conservatively used to determine the water volume required (See Attachment "C"). 6. The rupture of the pump recirculation line is assumed to occur at the location near the CST connection where the portion of the line is unprotected by the missile barrier (References 8 and 14). 7. The four 3/8" instrumentation lines for the Auxiliary Feedback Pump suction lines are assumed to rupture at the same elevation as of the pump suction nozzle at the CST (el. 156-9") for conservative purpose. This elevation is the lowest point above the ground level for the portion of the instrumentation lines that are unprotected from the missile impact. 8. The time allowed for FNP Operations to manually isolate a faulted SG during a MSLB event is 15 mins. and the time allowed for FNP Operations to manually isolate a faulted SG during a MFWLB event is 30 minutes. 9. It is assumed in the MSLB event that no heat removal from the RCS is achieved by the AFW flow through the faulted SG; thus, the AFW is considered lost and accounted for in the total volume required of the CST. This assumation is considered conservative, in 4 NMP*ES*039*002.
Vers i on 2.0 Documentation of Engineering Judgment , E2 -4 Enclosure 2 to NL-13-1257 DOEJ* FRSNC4t9117 
-MOOt Southern Nuclear Operating Company that, the AFW flow through the faulted SG during this event is removing heat but at an unknown rate since the SG is at atmospheric pressure due to the break. Evaluation:
Decay Heat 1. Determine CST volume required for Core Decay Heat removal during the following plant events, where decay heat is based on 102% power at 2775 MWt, 9 hours in Mode 3 steaming to atmosphere.
* Loss of Off-Site Power(LOSPl wi Tornado (Reference 
: 20)
* Loss of Off-Site Power(LOSPl wi Seismic Event (Reference 
: 20)
* Loss of Off-Site Power(LOSPl ( Reference 
: 20)
* Loss of Normal Feedwater wlo LOSP ( Reference 
: 18)
* Depressurization of Main Steam ( Reference 
: 23)
* Technical Specification Basis (Reference 
: 26) Adjust decay load from Ref. 2, Attachment "B" which represents 2775 MWt (100% power) plus 10 MWt pump heat to values that represents only decay heat due to fisson products decay heat. The decay heat from Attachment 
" B" at 9 hrs after shutdown is 0.1055 x 10'&deg; Btu which i ncludes 10 MWt pump heat. The portion of decay heat results that is due to pump heat is: 10 MWt I 2785 MWt = 0.00359 Therefore , the decay heat at 9 hrs. after shutdown from flsson products ( minus pump heat) is: 0.1055 X 10 10 -0.00359(0.1055 X 10'&deg;) = 0.10512 x 10'&deg; Btu Since i t is conservatively assumed that the reactor trips at 102% of rated reactor thermal power (2775 MWt) for this event, the fission product decay heat at 9 hrs after shutdown becomes: 0.10512 x 10'&deg; Btus x 1.02 = 0.10722 x 10'&deg; Btu 5 NMP*ES*Q39-002. Vers i on 2.0 Documentation of Engineering Judgment , E2 -5 Enclosure 2 to NL-13-1257 DOEJ* FRSNC4t9117 MOOt Southern Nuclear Operating Company Therefore, the CST Volume Required for Decay Heat is: W(9hrs.)=
0.10722 x 1010Btu (0.016165ifllbm)(gaVO
.1337 ft3) 1109 Btullbm W(9hrs.) = 116,893 aallons Note: Properties of the AFW are taken from calculation SM-95-0721-005 Ver. 1.0, i.e., AFW temperature of 110&deg;F (Reference 9). 2. Determine CST volume required for Core Decay Heat removal during the following plant events, where decay heat is based on 102% power at 2775 MWt, 2 hours at hot standby and 4 hours to hot shutdown.
* Small Break LOCA (Reference 21 )
* Main Feedwater Line Break w/LOSP (Reference 
: 24)
* Normal Cooldown (References
: 1 , 25 and 26)
* Main Steam Line Break wI LOSP (Reference 
: 22) Adjust decay load from Ref. 2 , Attachment 
" B" which represents 2775 MWt (100% power) plus 10 MWt pump heat to values that represents only decay heat due to fisson products decay heat. The port i on of decay heat results that is due to pump heat is: 10 MWt 12785 MWt = 0.00359 The decay heat from Attachment 
" B" at 2 hrs after shutdown is 0.35713 x 10* Btu. Therefore , the decay heat at 2 hrs. after shutdown from fisson products alone is: 0.35713 X 10* -0.00359(0.35713 x 10 9) = 0.35584 X 10 9 Btu Since it is conservatively assumed that the reactor trips at 102% of rated reactor thermal power for this event, the fission product decay heat at 2 hrs after shutdown becomes: 0.35584 x 10* Btu x 1.02 = 0.3630 x 10* Btu The decay heat from Attachment "B" at 6 hrs after shutdown is 0.7878 x 10* Btu. Therefore, the decay heat at 6 hrs. from fisson product heat alone is: 6 NMp*ES*039*002. Version 2.0 Documentation of Engineering Judgment. E2 -6 Enclosure 2 to NL-13-1257 DOEJ-FRSNC4t9117 
-MOOt Southern Nuclear Operating Company 0.7878 X 10 9 -0.00359(0.7878 X 10 9) '" 0.7849 X 10 9 Btu Since it is conservatively assumed that the reactor trips at 102% of rated reactor thermal power for this event, the fission product decay heat at 6 hrs after shutdown becomes: 0.7849 x 10 9 Btu x 1.02", 0.8006 x 10 9 Btu Volume of water required for heat removal is determined by: W", (Q I t.h) x v, (110&deg;F) x (gal/ft 3) Where, t.h '" h g (1155 psia) -hi (110&deg;F) (Reference 1 and 9, typical) t.h", 1187 -7B '" 1109 Btu/lb (Reference 5, typical) Calculating the required volume of condensate water for the 2 hour holding period results in: W(2 hrs) '" [0.3630 x 10 9 Btu 11109 Btullb] (0.016165 ft3/1b) (1gaI/0.1337 tt") W(2 hrs) '" 39575 aallons Calculating the required volume of condensate water for the 4 hour holding period results in: W(4 hrs) ",[(0.B006 x 10 9 Btu -0.3630 x 10 9 Btu) 11109 Btu/lb] (0.016165 tt"/lb) (1gall 0.1337 ft 3) W(4 hrs) '" 47,708 aallons 3. Determine core decay heat for the MSLB wlo LOSP event based on 102% power , 2831 MWT; at 2 hr and 5 hr. Based on FNP operations input, with the all Reps available the duration of the event should be set at a maximum of 5 hours, L e., 583&deg;F to 350&deg;F at a conservative cooldown rate of 50 &deg;F/hr. Duration is determined by: (583&deg;F -350&deg;F)1 50 &deg;F/hr = 4.66 hr. = 5 hr.
* Main Steam Line Break wlo LOSP (Reference 
: 22) Adjust decay load from Ref. 2, Attachment "B" which represents 2775 MWt (100% power) plus 10 MWt pump heat to values that represents only decay heat due to fisson products decay heat. 7 NMp*ES-D39*002. Vers i on 2.0 Documentation of Engineering Judgment, E2 -7 Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117 MOO1 Southern Nuclear Operating Company The portion of decay heat results that is due to pump heat is: 10 MWt 12785 MWt = 0.00359 The decay heat from Attachment "B" at 2 hrs after shutdown is 0.35713 x 10 9 Btu. Therefore, the decay heat at 2 hrs. after shutdown from fisson products alone is: 0.35713 x 10 9 -0.00359(0.35713 x 10 9) = 0.35584 x 10 9 Btu Since FSAR Table 15.4-23 (Sheet 1 of 2) conservatively assumes the reactor thermal power is at 102% prior to event, the fission product decay heat at 2 hrs after shutdown becomes: 0.35584 x 10 9 Btu x 1.02 -.3630 x 10 9 Btu The decay heat from Attachment "B" at 5 hrs after shutdown is .69009 x 10 9 Btu. Therefore, the decay heat at 5 hrs. from fisson product heat alone is: 0.69009 X 10 9 -0.00359(0.69009 x 10 9) = 0.68761 X 10 9 Btu Since FSAR Table 15.4-23 (Sheet 1 of 2) conservatively assumes the reactor thermal power is at 102% prior to event, the fission product decay heat at 5 hrs after shutdown becomes: 0.68761 X 10 9 Btu x 1.02 = 0.7014 x 10 9 Btu
* Volume of water required for deay heat removal is determined by: W = (Q 1 t.h) x VI (11 O"F) x (gaI1ft 3) Where, t.h = hg(1155 psia) -hi (11 O"F) t.h = 1187 -78 = 1109 Btullb Calculating the required volume of condensate water for the 2 hour holding period results in: W(2 hrs) = [.3630 x 10 9 Btu 11109 Btullbl (0.016165 ft3/1b) (lgaI/O.I337f e) 8 NMP-ES-039-002. Version 2.0 Documentation of Engineering Judgment. E2 -8 Enclosure 2 to NL-13-1257 DOEJ-FRSNC4t9117 
-MOOt Southern Nuclear Operating Company W(@2hrs)=
39575 gallons Calculating the required volume of condensate water for the 5 hour holding period results in: W(5hrs) =[( 0.7014 x 10 9 Btu -0.3630 x 10 9 Btu) 11109 Btullbl (0.016165 ft3/lb) (1gal I 0.1337ftl)
W(@5 hrs) = 36893 gallons Sensible Heat 1. Determine CST volume required for Sensible Heat removal during the following plant events, where sensible heat is based on Tno*load of 547 of (Ref. 15), 9 hours in Mode 3 steaming to atmosphere.
* Loss of Off-Site Power (LOSP) wI Tomado Loss of Off-Site Power CLOSP) wI Seismic Event
* Loss of Off-Site Power CLOSP)
* Loss of Normal Feedwater wlo LOSP
* Depressurization of Main Steam
* Technical Specification Basis No CST volume is required for the removal of RCS sensible heat, since the unit stays at hot standby conditions through-out the duration of above events. 2. Determine required CST volume for removal of Sensible Heat in the RCS at Tavg (hot) of 577 .2&deg;F(Ref. 
: 15) for the following events: Main Feedwater Line Break w/LOSP Normal Cooldown Small Break LOCA Main Steam Line Break wlo LOSP Main Steam Line Break wI LOSP An uncertainty of 6&deg;F on the initial reactor average coolant temperature is conservatively assumed. Thus, sensible heat calculation is based on aRCS Tavg (temperature of 577.2 of + 6&deg;F = 583.2 of for 9 NMp*ES*039*002.
Version 2.0 Documentation of Engineering Judgment. E2 -9 Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117 M001 Southern Nuclear Operating Company the thermal capacity of the system metal and the reactor coolant fluid enthalpy change from 583.2 of to 350&deg;F. The equation used to calculate the sensible heat load is based upon the conservation of mass and energy: dU = m x c p x .toT + .toh x M where; dU = sensible heat load from ReS temperature of 583.2 OF to 350&deg;F, Btu m = mass of Res metal , Ibm = 5.0 x 10 6 1bs .toT = 583.2&deg;F -350&deg;F c p = specific heat of metal, 0.11 Btu/oF-Ibm (from Reference 
: 4) .toh = enthalpy change of reactor coolant , Btu/lb @ 583.2 OF; hI, = 593.02 BtU/lb; VI' = 0.022918 ft3/lb @ 350&deg;F; hf2 = 321.80 Btu/lb; Vf2 = 0.01799 ft3/lb M = total reactor coolant mass , Ib = V / VI. where V = total reactor coolant volume , ft3 = 9,723 ft3 (Refererence 
: 27) VI = specific volume of reactor coolant, where VI' = Vf2 for conservatism. = 0.Q18 ft3/lb; M = 9,723 ft 3/O.018 W/lb = 540,167 Ibs Thus, dU = m x c p x.toT +.toh x M = 5.0 x 1 061bs. x 0.11 Btu/OF-Ibm x (583.2&deg;F -350&deg;F) + (593.02 Btullb -321.80 Btullb) x 540,167Ibs. dU -2.7476 X 10 8 Btus Volume of eST required for Sensible Heat Load W (sensible ht.) -2.7476 x 10 8 Btu x 0.016165 ft3/lbm 1109 Btu/Ibm x 0.1337 ft3tgal. W (sensible ht.> = 29.955 gallons reguired for sensible heat load 10 NMP*ES-039-002. Vers i on 2.0 Documentation of Eng i neering Judgment , E2 -10 Enclosure 2 to NL-13-1257 DOEJ-FRSNC4t9117 
-MOOt Southern Nuclear Operating Company Rep Heat 1. Determine CST volume required for removal of the RCP heat added to the RCS.
* Loss of Off-Site Power(LOSP) wI Tornado
* Loss of Off-Site Power(LOSP) wI Seismic Event
* Loss of Off-Site Power(LOSP)
* Small Break LOCA
* Main Steam Line Break wI LOSP
* Main Feedwater Line Break w/LOSP
* Tech. Spec. Basis (Mode 3 for 9 hours) These events are concurrent with a LOSP where no RCPs are in operation; however , the net heat input of 10 MWt due to the RCPs at beginning of the event exists in the RCS and it i s assumed this heat decays over a 1 hour period. (Reference 
: 15) 10 MWt = 3.41297 x 10 7 Btulhr (based on 0.293 watts = 1 Btulhr) Therefore, CST volume required due to 10 MWt RCP net heat input at beginning of the event and assumed to decay in 1 hour is; W(1 OMWt decayed over 1 hr.) = 3.41297 x 10 7 Btu/hr (0.016165ft 3/Ibm)(gaI/0
.1337 ft3) x 1 hr. 11 09 Btu/lbm W (10 MWt decayed over 1 hr.> = 3721 gallons 2. Determine CST volume required for removal of the RCP heat added to the RCS for the following event:
* Main Steam Line Break w/o LOSP This RCP heat is based on net heat input of 10 MWt per WCAp*15097 and operation of all RCPs from Tavg(583&deg;F) to 350&deg;F at a cooldown rate of 50&deg;F/hr. The FSAR Section 15.4.2.1 does not assume a conservatively large RCP heat of 15MWt. However, the initial net heat input of 10MWt is assumed to 11 NMp*ES-039-002.
Vers i on 2.0 Documentation o f Engineering Judgment. E2 -11 Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117 MOO1 Southern Nuclear Operating Company decay over 1 hour and the operation of all three RCPs for the duration of the event is supported by input from Westinghouse and FNP Ops.
* Assume aRCS cooldown rate of 50&deg;F/hr.
* Assume a Tavg of 583 &deg;F(57JOF + 6 OF) at the start of the event cooling down to 350 of in determining the duration of the event at the cooldown rate. Tavg of 583&deg;F is assumed plant is initially at 102% core thermal power per FSAR Table 15.4-23 (Sheet 1 of 2) Duration of Event is determined by : (Tavg --350&deg;F) /50&deg;F/hr.(cooldown rate) Set duration of the event at 5 hours. 10 MWt pump heat input = 3.41297 x 10 7 Btu/hr. Determine No. of gallons per hour required remove 10MWt pump heat is: W (gal/hr.) = (3.41297 x 10 7 Btu/hr /11 09 Btu/lb) x 0.16165 ft3/1bm x ga1/0.1337ft 3 W = 3721 gal/hr to remove 1 OMWt pump heat Thus for the 5 hr duration of the event, the CST volume required to remove heat input of all RCPs operating is: W = 3721 gallhr x 5.0 hr. = 18605 gallons. For the initial net RCP heat input of 1 OMWt that is assumed to decay over 1 hour from the start of the event, the CST volume required is: 3721 gallons. Therefore, the total CST volume required to remove the initial heat input and operating heat input during the event is: 18605 gallons + 3721 gallons = 22326 gallons 3. Determine CST volume required for removal of the RCP heat added to the RCS for the following event: 12 NMP-ES-039-002.
Version 2.0 Documentation of Engineering Judgment.
E2 -12 Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117 
-MOO1 Southern Nuclear Operating Company
* Loss of Normal Feedwater w/o LOSP This RCP heat is based on net heat input of 15 MWt per WCAP-15097 and operation of all RCPs from Tavg(full power) to 350&deg;F at a cooldown rate of 50&deg;F/hr. The FSAR Section 15.2.8.2.1.A assumes a conservatively large RCP heat input of 15MWt. This initial net heat input of 15 MWt is assumed to decay over 1 hour. Per this FSAR sect i on , all three RCPs operate at the initiation of the event for 10 mins and one(l) RCP operates for the duration of the event; i.e. 9 hrs. in hot standby.
* Per WCAP -15097( Ref. 15), Tavg of 583&deg;F (577&deg;F + 6&deg;F) at the start of the event cooling down to 350 OF in determining the durat i on of the event at the cooldown rate. 15 MWt pump heat input = 5.119 x 10 7 Btu/hr. Determine number of gallons per hour required to remove 15MWt pump heat is: W (gal/hr.) 
= (5.119 x 10 7 Btu/hr 111 09 Btu/lb) x 0.016165 fe/lbm x gall 0.1337fe W = 5580 gaVhr to remove 15MWt pump heat. Thus , assumming 1 hr to decay 15 MWt , the net heat input yields: 5581 gallons Per FSAR 15.2.8.2.1.A, three RCPs operate for 10 mins at start of event: 5581 gals/hr. x 1 hr.!60mins x 10 min. = 930 gallons. Thus, for the remaining duration of the event which is; (9 hrs -10 mins), the CST volume required to remove heat input of one RCPs operating Is: 1 RCP heat input = 5 MWt = 1.7065 x 10 7 Btulhr. W (gaVhr) = (1.7065 x 10 7 Btu/hr 11109 Btullb) x 0.016165 ft 3 11bm x gal.! 0.1337 ft 3 W (gaVhr) = 1860 gaVhr Now, timing remaining 
= 9 hrs. -(10min x 1 hrlSO min) = 8.83 hr. Therefore , volume required for timing remaining during event is: Volume = 1860 gaVhr x 8.83 hr. = 16,430 gallons 13 NMp*ES*039*002 , Version 2.0 Documentation of Engineering Judgment , E2 -13 Enclosure 2 to NL-13-1257 DOEJ* FRSNC419117 
-MOot Southern Nuclear Operating Company Total CST volume required for RCP heat added during LNFW event is: 5581 + 930 + 16,430 = 22941 gallons 4. Determine the CST volume required for removal of the RCP heat added to the RCS for the following event:
* Depressurization of Main Steam FSAR 15.2.13 (Ref. 23) norWCAP 14722 (Ref. 19). section 6.2.13 makes assumptions regarding RCP operation.
thus it is assumed that 3 RCPs operating for approx. 10 mins to reach RCS equilibrium than 1 RCP operates for remainder of 9 hours in hot standby condition. Also assumed. RCP heat includes the initial 10 MWt net RCP heat input before reactor trip and this heat is assumed to decay in one (1) hour. Assume Tavg is based on reactor full power operation. Per WCAP -15097( Ref. 15). Tavg of 583&deg;F (577&deg;F + 6&deg;F) at the start of the event cooling down to 350 of in determining the duration of the event at the cooldown rate. 10 MWt pump heat input = 3.41297 x 10 7 Btu/hr. Determine number of gallons per hour required to remove 10MWt pump heat is: W (gallhr.) 
= (3.41297 x 10 7 Btulhr 111 09 Btullb) x 0.016165 If/lbm x gall 0.1337ft 3 W = 3721 gallhr to remove 1 OMWt pump heat. Thus. assumming 1 hr to decay 10 MWt. the net heat input yields: 3721 gallons Assuming three RCPs operate for 10 mins at start of event where the net heat input for 3 RCPs operating is 10 MWt. determine the CST volume required: 3721 gals/hr. x 1 hr.l60mins x 10 min. = 620 gallons. Thus. for the remaining duration of the event which is; 9 hrs -10 mins. the CST volume required to remove heat input of one RCPs operating is: 1 RCP heat input = 5 MWt = 1.7065 x 10 7 Btu/hr. 14 NMp*ES*039*002. VelS i on 2.0 Documentation of Engineering Judgment , E2 -14 Enclosure 2 10 NL-13-1257 DOEJ* FRSNC419117 
-MOO1 Southern Nuclear Operating Company W (gallhr) = (1.7065 x 10 7 Btulhr /1109 Btullb) x 0.016165 fellbm x gal.l 0.1337 fe W (gal/hr) = 1860 gal/hr Now, timing remaining 
= 9 hrs. -(10min x 1 hr/60 min) = 8.83 hr. Therefore, volume required for timing remaining during event is: Volume = 1860 gallhr x 8.83 hr. = 16,430 gallons Total CST volume required for RCP heat added during the Depressurization of Main Steam event Is: 3721 + 620 + 16,430 = 20771 gallons 5. Determine the CST volume required for removal of the RCP heat added to the ReS for the following event:
* Normal Cooldown For th i s event, normal cooldown , the heat added to the ReS is based on 1 Rep in operation for the duration of the event , 6 hrs. plus 10MWI net Rep heat added and assumed to decay over 1 hr. after shutdown.
Per Reference 15, the heat added to RCS per pump is assumed at 5 MWt and the net heat input added for all three RCPs is 10 MWt. For 5 MWt (1 RCP), the CST volume required per hour is: 5 MWt = 5 x 10" watts and 1 Btulhr = 0.293 watts, therefore, 5 MWt = 5 x 10 "watts x (1 Btu/hr / 0.293 watts) = 1.7065 x 10 7 Btulhr. W/hr ( 1 Rep) = 1.7065 x 10 7 Btu/hr x 0.016165 ft 3 l1b: 1109 Btullb x 0.1337 fe/gal where 1109 Btu/lb is the enthalpy heat sink of the AFW, from 110&deg;F to 1155 psia, max i mum backpressure of the SlG secondary side, Wlhr ( 1 RCP) -1860 galslhr. For 10 MWt the CST volume required per hour is: 15 NMP*ES*039-002. Vers i on 2.0 Documenta ti on of Engineering Judgment. E2 -15 Enclosure 2 to NL-13-1257 nOEJ-FRSNC419117 
-M001 Southern Nuclear Operating Company 10 MWt", 3.4129 x 10 7 Btu/hr W/hr (10 MWt net heat added) '" 3.4129 x 10 7 Btu/hr x 0.016165 ft 3/1b; 1109 Btu/lb x 0.1337 ft3/ga l. W/hr (10 MWt net heat added) -3721 gal/hr. Therefore , for this normal cooldown event , the CST volume required for removal of RCP heat is; W (RCP heat) '" (6 hr. x 1860 gaVhr) + (1hr x 3721 gaVhr) W (Rep heat) '" 14881 gallons Pressurizer Heaters Input to ReS Per FSAR Section 15.2.8, during a Loss of Normal Feedwater (LON F) event, the pressurizer proportional and backup heaters are assumed operable. The total capacity of the pressurizer heaters is 1.4 MWt. The heaters output represents an addition to the RCS energy which must be removed by the AFW system. Therefore, determine CST volume required for removal of this energy during the event duration of 9 hours in Mode 3, steaming to atmosphere .
* Loss of Normal Feedwater Pressurizer Heater capacity in Btu/hr. 1.4 MWt = 4.7781 x 10 6 Btulhr (based on 0.293 watts = 1 Btulhr) W/hr (PzrHtrs.) '" 4.7781 x 10 6 Btulhr x 0.016165 ft3/lb; 1109 Btullb x 0.1337 W/gal where 1109 Btullb is the enthalpy heat sink of the AFW, from 11 oaF to 1155 psia, maximum backpressure of the SG secondary side , W/hr (pzr Htrs.) = 521 gals/hr. Therefore , for this LONF event, the CST volume required for removal of Pressurizer Heater energy input to the RCS: W (pzr Htrs.) = (9 hr. x 521gallhr) 16 NMp*ES*039*002.
Version 2.0 Documentation of Engineering Judgmen t, E2 -16 Enclosure 2 to NL-13-1257 DOEj* FRSNC4t9117 
-MOOt Southern Nuclear Operating Company W CPzr Htrs.l = 4689 aallons Steam Generator Refill Volume Per calculation BM*0961*001 ,Version 4.0, a steam generator refill volume was added to the CST based on initiallow*low steam generator water level for added margin. Per WCAP *15097 Section 4.2.2.4.1 , the CST volume conservatively considers refill of the SG to the no-load programmed level. Since it has been clearly defined that the steam generator refill volume added conservative margin to the CST sizing in calculation BM-0961-001, Version 4.0, this conservatism has been removed per the latest revision of the calculation. Per email from FNP Operations, Thomas Nesbit, dated 3/12/2012 , there would be no need to refill the faulted SG Refilling of the SG would be more for equipment preservation rather than RCS cooling once the RHR is aligned. Based on the above documentation that the SG refill volume is conservative and to add design margin to the CST Volume, this refill volume has been removed from the CST volume sizing verification since other conservatism, such as, instrument and pump recirculation line failures as well as vortex prevention have been added to provide design margin to the CST volume. An LDCR will be developed to remove this assumption that the SG is refilled to no-load programmed level at the completion of cooldown to 350&deg;F per FNP FSAR-9.2.6.3. AFW Pumps Recirc. Line and Instrumentation Lines Rupture 1. Determine the eST inventory loss due to the AFW pump's recirculation line and instrumentation line failure during the following events:
* Loss of Off-Site PowerCLOSPl wI Tornado
* Loss of Off-Site PowerCLOSPl wI Seismic Event During a LOSP event assumed for a duration of 9 hours, two MDAFW pumps or the TDAFW pump is started. Heat removal from the ReS i s maintaind by natural circulation while the unit is maintained in hot standby. The natural c i rculation capab i lity of the ReS will remove decay heat from the core aided by the AFW flow in the secondary system. As the steam system pressure rises following a trip, the steam system PORV's are automatically opened to the atmosphere. It is assumed per FSAR 15.2.9.2.1.E for this event that two MDAFW pumps are available to supply a minimum of 350 gpm to three steam generators. 17 NMp*ES*039*002. Vers i on 2.0 Documentat i on of Eng i neering Judgment, E2 -17 Enclosure 2 to NL-13-1257 DOEJ* FRSNC419117-MQ!!!
Southern Nuclear Operating Company Due to a tornado missile or seismic event, it is assumed that the pump's recirculation line fails. The volume of water lost out from a break in the pump recirculation line will be calculated with an assumption that the ruptured minimum flow recirculation will be isolated within 30 minutes. Three recirculation lines of the pumps form one 6-inch line to the CST. It is assumed that, even though th is s i ngle 6-inch return line would rupture at a location near the CST , all three pump recirculation lines would be isolated. The MDAFW pumps will be operated w i th the minimum flow lines isolated during the remaining time of 8.5-hour period and this will not affect the pump periormance for this operation mode. Since the valves for the recirculation lines can be closed manually inside the plant, 30-minute span for the action i s cons i dered sufficient (Reference 3 Section 6.3.7.6). From References 6 and 7, the maximum recirculation flow for the TDAFW pumps is 100 gpm, and 50 gpm for each MDAFW pump. These recirculation flow values are slightly higher than calculated to conservatively accommodate the slight decrease in frictional loss due to a line break. Thus, the volume of water that w i ll be lost from two MDAFW pumps during the 30 minute span for isolation of the recirculation lines: (50 + 50)gpm x 0.5 hour = 3000 galions. Since the nozzle of the AFW pump recirculat i on line is located at 19 feet above the CST base, which i s well above the height (13.25 feet, see below calculation) of the protected 164 , 832 gallon volume, or even above the protected height of the tank (16') (Reference 11), a possibility of water coming out from the CST th r ough the ruptured recirculation line is not cons i dered. From References 10,11,12,13,14, four flow instrumentation lines branch out from the two AFW pump suction lines a t EI. 150 ft. location. Due to a tornado missile or seismic event , it is assumed that these instrument lines fa i l and are not i solated during the du r at i on of the event. Since the locations of the branch-out are in the pipe trench, for a conservative calculation purpose , the break point of the instrumentation lines is assumed to happen at the lowest location above the ground level (EI.156' -9") where the instrumentation lines are exposed to missile impact. This is the same elevation as of the pump suction nozzle at the CST. The instrument tubing is 318" size with 0.065 minimum wall with ID 0.245" (Reference 13). In order to calculate the flow rate at the rupture point, assume all the pressure losses in the 8" pump suction pipe and the instrument tubing are ignored. This will give a conservative result for the flow rate. Since the lower 164,000 gallon volume of the tank shall be reserved for handling decay heat and cooldown, the height of this water volume in the tank is: Tank bottom area: Tr r2 = Tr (2311)2 = 1662 (Reference 
: 8) Height of volume (164 , 832 gallons) = (164832 gals 17.48galslft 3) 11662 = 13.25 ft. (equivalent to a pressure of 5.73 psi). From Reference 5, page 2-9, the water flow from the break point is calculated by extrapolation.
The break size of the tubing (0.245" ID) is approximately equivalent to 1/4". The discharges of 1/4" nozzle at 10 and 15 psi are 5.91 and 7.24 gpm, respectively (Reference 6). 18 NMP*ES*Q39-002.
Version 2.0 D oc umenta ti on of Eng i nee ri ng Judgment. E2 -18 Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117 
-MOO! Southern Nuclear Operating Company Taking a middle po i nt of the CST water level for the fluid pressure at the instrument line break pOint , or 2.9 psi, and ignoring all the line losses and the effect of the AFW pump suction pressures for conservatism, then the flow at the break points is calculated as: 5.91 gpm -{[(7.24 -5.91 )/(15-1 0)] x (1 0-2.9)} = 5.91 -1.88 = 4.03 gpm per line Total outflow for four instrumentation lines: 4.03 x 4 lines = 16.1 gpm If we consider that these instrumentation lines will not be isolated during the 9-hour cooldown period, than the total volume lost from the ruptured instrument lines for 9 hours is 16.1 x 9 x 60 = 8694 gallons. 2. Determine the CST inventory loss due to the AFW pump's recirculation line and instrumentation line failure during the following event:
* Normal Cooldown During normal plant cooldown, operating one MDAFW pump will permit a maximum initial cooldown rate of 100&deg;F/Hr.
Each of the MDAFW pumps is sized to supply the steam generators with 100 percent of the required feedwater flow for a nonnal safe cooldown of the reactor coolant system (References 1 & 3). Therefore, one MDAFW pump is sufficient for a nonnal cooling operation with a cool down rate of 50 &deg;F/Hr assumed for this ca l culation.
For conservative approach for this calculation, all three AFW pumps (the TDAFW and two MDAFW pumps, each at the design flow rate of 700 and 350 gpm, respectively) operate for the first 30 minutes before the TDAFW pump and one MDAFW pump are secured. For nonnal plant cooldown, the AFWS is placed under manual control to supply feedwater to the steam generators for removal of decay and sensible heat from the reactor system. The TDAFW pump is designed to be manually or automatically initiated to its rated capacity and head within 1 minute from starting at rest for at least 2 hours independent of any ac power (Reference 1). Operator action to secure the TDAFW and one MDAFW pump within 30 minutes will be adequate for this calculation purpose. The volume of water lost out from a break in the pump recirculation line will be calculated with an assumption that the ruptured minimum flow recirculation w i ll be isolated within 30 minutes. Three recirculation lines of the pumps fonn one 6-inch line to the CST. It is assumed that , even though this single 6-inch return line would rupture at a location near I the CST, all three pump recirculation lines would be isolated. The MDAFW pumps will be operated with the minimum flow lines isolated during the remaining time of 5.5-hour period and this will not affect the pump perfonnance for this operat i on mode. Since the valves for the recirculation lines can be closed manually inside the plant, 30-minute span for the action is considered sufficient (Reference 26 , Section 8.3.7.6). 19 NMP-ES-039-Q02. Version 2.0 Documentation of Eng i neering Judgment. E2 -19 Enclosure 2 to NL-1 3-1257 nOEJ-FRSNC4I9117 
-MOOI Southern Nuclear Operating Company From References 6 and 7, the maximum recirculation flow for the TDAFW pumps is 100 gpm, and 50 gpm for each MDAFW pump. These recirculation flow values are slightly higher than calculated to conservatively accommodate the slight decrease in frictional loss due to a line break. Thus, the maximum volume of water that will be lost during the 3D-minute span from the recirculation line failure while all three AFW pumps are in operation is: (100 + 50 + 50)gpm x 0.5 hour = 6,000 gallons. Since the nozzle of the AFW pump recirculation line is located at 19 feet above the CST base, which is well above the height (13.25 feet , see below calculation) of the protected 164,832 gallon volume, or even above the protected height of the tank (16') (Reference 8), a possibility of water coming out from the CST through the ruptured recirculation line is not considered.
The basis for the inventory loss from the failure of the instrument tubing for the normal cool down event is the same as that provided for the LSOP events. See sheets 18 and 19 under the LOSP events for this basis. The total outflow for four instrumentation lines: 4.03 x 4 lines = 16.1 gpm If we consider that these instrumentation tubing will not be isolated during the 6-hour cooldown period, than the total volume lost from the ruptured instrument lines for 6 hours is 16.1 x 6 x 60 = 5803 gallons. The total volume inventory lost from all the ruptured lines is, assuming all the line ruptures occurred at time = 0 sec. of the cooldown mode: 6,000 + 5,803 = 11,803 gallons Vortex Prevention 
-CST Minimum Submergence Level Determine the volume required to prevent the potential of a vortex formation in the CST assuming air enters the tank undemeath the tank's diaphragm for the following events:
* Loss of Off-Site PowerILOSP) wI Tomado
* Loss of Off-Site PowerILOSP) wI Seismic Event
* Normal Cooldown ( 2 hours at hot standby and 4 hours to hot shutdown)
* Determine the CST volume required based on the minimum submergence water level required in the CST to prevent vortexing at the AFW pump's suction inlet. Per Reference 20 NMp*ES-D3!H lO 2. Ve rsi on 2.0 Documentati o n of Engineering Judgment , E2 -2 0 Enclosure 2 to NL-13-1257 DOE,J* FRSNC419117 
-MOO1 Southern Nuclear Operating Company 17, the minimum submergence level required in the CST based on the operation of two MDAFW pumps or the TDAFW pumps shall be greater than or equal 9.78 inches from the bottom of the tank. From the intemal dimensions of the tank per Reference 16, the volume required is: Volume Req'd (gallons) 
= Tank Area (tf) x Min. Submergence (ft.) x 7.48 gallons ft.3 Tank Area (ft.2) = TT radius 2 , where the radius is equal to 23 ft. Tank Area (ft.2) = TT (23)2 = 1661.9 ft.2, use 1662 ft2 Min. Submergence(ft.) 
= 9.78 inches x (1ft.l12inches) 
= 0.815 ft. Thus, Volume Req'd for Vortex Prevention=
1662 ff x 0.815 ft. x 7.48 gal = 10.132 gallons ft.' Note: This volume includes the unusable volume that is 4 inches from the bottom of the tank below the pump suction nozzles. Unusable Volume -CST Volume below AFW pump Suction Nozzles Determine the unusable volume in the CST due to the 4 inches from the AFW pump's suction nozzles to the bottom of the tank for the the following events:
* Loss of Off*Site PowerlLOSPl
* Loss of Normal Feedwater w/o LOSP
* Small Break LOCA
* Main Steam Line Break wand w/o LOSP
* Depressurization of Main Steam
* Main Feedwater Line Break w/LOSP
* Technical Specification Basis 21 NMP*ES*039-002. Vers i on 2.0 Documentation of Engineering Judgment , E2 -21 Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117-MOOl Southern Nuclear Operating Company Un-usable Volume in CST is equivalent to the 4 inches (0.33 It) of CST volume below the pumps' suction nozzles since no vortex prevention is required while bringing the unit to hot shutdown, RCS at 350&deg;F. That is, during these events no attached lines to the CST are assumed to fail allowing air to enter the tank below the diaphragm thus creating the potential for vortex formation. Unusable Volume (gallons) 
= Tank Area (ft2) x 0.33 ft. x 7.48 gallonslft.
3 Tank Area (ft.2) = 1T radius 2 , where the radius is equal to 23 ft. Tank Area (ft.2) = 1T (23)2 = 1661.9 ft.2 , use 1662 tt2 Thus, Unusable Volume (gallons) 
= 1662 ff x 0.33 ft. x 7.48 gal/fe = 4102 gallons Inventory Loss 1. Determine the CST volume inventory loss during the following event:
* Main Feedwater Line Break From FLB wlo Isolation, S/G "A" , "B" and " C" Faulted, sheets 32 thru sheet 40 of Calculation 40.02, Attachment 8, the inventory loss is summed from pipes 15C, 16C and 17C. The results show the following:
* FLB wlo Isolation , SlG "A" Faulted Pipes Faulted S/G Flow(gpm) 15C A 496.50
* FLB wlo Isolation , S/G " B" Faulted Pipes Faulted S/G Flow(gpm) 16C B 495.80 22 NMp*ES*039
*002. Version 2.0 Documentation of Eng i neering Judgment, E2 -22 Enclosure 2 to NL-13-1257 DOEJ-FRSNC4I9117 
-MOOI Southern Nuclear Operating Company
* FLS w/o Isolation, S/G "C" Faulted Pipes Faulted S/G Flow(gpm) 17C C 496.60 Taking the worst CST inventory loss of the faulted S/Gs w/o isolation for 30 minutes during a feedwater line break yields: Inventory Loss (gals) = 496.60 gpm x 30 mins = 14898 gallons However, a 10% margin is added to this value for uncertainity until this value is documented and verified in a new version of the Calculation 40.02. Thus, the new version of calc. 40.02 will be an input to this calculation. Including the 10% margin: 14898 + 1490 = 16388 gallons 2. Determine the CST volume inventory loss during the following event:
* Main Steam Line Break From MSLS w/SG Pressure = 300 psia and S/G "A", " S" and "C" Faulted, sheets 14 thru sheet 26 of calculation 40.02, Attachment 8, the inventory loss is taken from pipes 15C, l6C and l7C. The results show the following:
* SLS w/o Isolation, S/G " A" Faulted Pipes Faulted S/G Flow(gpm) l5C A 764.2
* SLS w/o Isolation , S/G " S" Faulted Pipes Faulted S/G Flow(gpm) 23 NMp*ES'()39'()()2. Version 2.0 Documentation of Eng i neering Judgment.
E2 -23 Enclosure 2 to NL-13-1257 DOEJ-FRSNC419117 -Moo1 Southern Nuclear Operating Company 16C B 763.0
* SLB w/o Isolation, S/G "C" Faulted Pipes Faulted S/G Flow(gpm) 17C C 763.5 Taking the worst CST i nventory loss of the faulted S/Gs w/o isolation for 15 minutes during a main steam line break yields: Inventory Loss (gals) = 764.2 gpm x 15 mins = 11463 gallons However , a 10% margin is added to this value for uncertainity until this value is documented and verified in a new version of the calculation 40.02. Thus, the new version of calc. 40.02 will be an input this this calculation.
including the 10% margin: 11463 + 1146 = 12609 gallons

== Conclusion:==

Summing up the volumes required for each sizing component as applicable
; such as , decay heat, sensible heat , etc., the required volumes of the CST for each plant event is summarized below. CST Volume Required (Gallons) per Plant Event Pllnt LOSPlTornado LOSP/Selsmlc LOSP SBLOCA MSLBw/o MSLB Events LOSP wlLOSP CST Volume 142 , 440 142 , 440 124,718 125,081 145,480 137,870 Roqu l rod Plln t LNFWw/o Depressurlzartlon MFWLB Normal Tech. Spec Events LOSP of Main Steam w/LOSP Cooldown Basis CST Volume 148 , 625 141 , 766 141 , 449 154 , 054 124,716 Reau l rod 24 NMP*ES-039-002, V.rs;on 2.0 Documentation of E n g i neering Judgmen t, E2 -2 4 Enclosure 2 to NL-13-1257 DOEJ* FRSNC4t9117 MOOt Southern Nuclear Operating Company List of Attachments:
A. Excel Spreadsheets; "Condensate Storage Tank Volume Requirements per Plant Events" and "Condensate Storage Tank Volume Sizing Basis" B. MAP*121, Decay Heat Calculation Code, Version V01; Copyright 1979, 1988 Bechtel Power Corporation C. Properties of Saturated Steam D. Email dated8/10/2012.From
: WillieJennings.To:
Thomas H. Nesbit (FNP Operations); 

==Subject:==
CST Sizing Verification 
-MSLB and MFWLB Events (Mass Flow Question)
E. Email dated3l19/2012.From
: JamesD.Andrachek(Westinghouse).To
: Willie Jennings; Sublect: Farley CDBI -MSLB Case F. Email dated3/18/2012.From
: JamesD.Andrachek(Westinghouse).To:
Willie Jennings; Sublect: Farley CDBI -MSLB Case G. Email dated3/1212012.FromThomasH.Nesbit(FNPOperations)
.To: Willie Jennings; 

==Subject:==
RCP Heat Addition during Shutdown -CST Sizing for CDBI Response 25 NMp*ES*039-002. Version 2.0 Documentation of Engineering Judgment, E2 -25 Enclosure 2 to NL-13-1257 iSQ[J11 link as:gLtlrfOls:nUi III' P1i!!l1 b!:OIJi -FB&sect;N!:;4H111I-M OO1 Version 2 6/14/2012 Depressurization MFWl8w/LOS MFWlBw/lDSP Tech Spec. Bas is lOSP/Tornado lOSP/Seismic LOSP SBlOCA MSLB W 0 lOSP MSlBw/lOSP lNFW w/o lOSP Main Steam Case (ase"S" Mode 3 for 9 hrs. ,-'$I Slrln. Components Deuy Heat (@ 102" Power) 116893 116893 116893 116893 116893 116893 Decav Heat ( @102"power)
Trip to Mode 3 39575 39575 39575 39575 39575 Trip to Mode 3 nme to RHR Initiation 47708 36893 47708 47708 47708 47708 nme to RHR Inltialion ---Sens ible Heat@) T aVi 583 f, 100% power 29955 29955 29955 29955 29955 29955 SenSIble Heat @I T aVI 583 F. lOO%power 5ftUl bl e Heat@TavaS47F.NoLoad Sensible Hut #I T av, 547 F, No load Rep H eat 3721 3721 372 1 3721 22326 3721 22941 20771 3721 3721 14881 3721 RCPHeat P ZRHea terlnput '68' PZR Heater Input SG Re fill No load SG Refill No load AFW Pum ps Re cire Une Rupture 3000 3000 6000 AFW Pumps Reclre Line Rupture l nst. Line R upture 86.' 8694 sao3 lnst lme Rupture C ondHl ser H o tweU Makeup Condenser Hotwell MJkel,lp Min. Submersenee 
-Vortex 10132 10132 10112 Min. Submersence Unusable Volume 4 Inches from BOT 4102 4102 4102 41 0 2 4102 4102 4102 4102 4102 Unusable Volume 4 inches from BOT 1-In'nntory loss dUI! to MS/FW break 11463 114 6 3 14898 14898 Inventory loss due: to MS/FW line I nven tory loss dUI! to MS/FW line break 10% marl i n 1146.3 1146.3 1489.8 1489.8 Inventory loss due line break 10% marlin Total CST Volume Required:
14244 0 1 4 2 440 124716 125 0 6 1 145460 1]7670 148625 141766 141449 141449 154054 124716 T a ta l CST V olume Required:
W (max i mum prgtected volume of the CSD -16 4 832 ga ll9 n, (Refere nc e 2 01 DOEJ) E2 -26 Enclosure 2 to NL-13-1257 Per FSAR IS.2.9 lOSP/Tornado OKay heat based on 102% power at 2775MWt.9 hours In Mode 3 steam!", to the atmosphere. FSAR 15.2.9 and WCAP14722 section 6 , 2.9 makes no assumption regardlns power level; therefore. it is assumed the unit Is at 102" of InS MWt with 10 MWt of net RCP heat considered under RCP Heat. Pe r FS A R IS.2.9 l OS P/Sefs m i c Deca y heat based on 102% power ilt 277S MWt, 9 hours In Mode 3 steamin, to the atmosphere.
FSAR 15.2.9 ilnd WCAP14722 section 6.2.9 makes no assumption relilrdlng power level; therefore, It Is assumed the unit Is <lit 101% of 2775 MWt with 10 MWt of net RCP heat considered under Rep He a l Per FSAR 15.2.9 lOSP D eca y heal based 00 102% power 177SMWt. 9 hourS" In Mode 3 steaming t o the atmosphere.
FSAR 15.2.9 and W CAP14722 section 6.2.9 makes no assump tion re,arding power level; t herefore, It 15 assumed the unilis at 102% of 2775 MWt with 10 MWt of net Rep heat conSidered under Rep HeOilit. No sensible hut slna unit wilt be held at No sensible heal since unit will be held No sensible heat since unit will be held at hot sundby conditions Ihrou,h this duration. hot standby conditions throush this Sensible Heal. T av& 583 F, 100% power duration. Senslbl&Heat "same as above" at hot standby condit i ons throu," th i s duration. No Rep w/LDSP; however, Initial 10 MWt N o RC Pw/L0 5P , how ever, Il"It tiallO MWt N o RCP wILOSP; however, Initial 10 MWt net pump heat Input Is assumed to dl!COilly net pu mp heat i npu t Is auume d 10 n et pump heat Input is assumed to RCPHeat in one hour. decay in one hour_ deay In one hour. SG Refill Nolootd Pressurber Heat Input AFW Pumps Retire Une Ruptu re Inst. Un. Rupture Condenser Hotwell Makeup Min. Submel'lence . Vortea Per email from T homa s Nesbit, FNP Op s; from an acdd e nt perspective, there is no reason to r e f i ll the faul t ed S6. The faulted SG s hould be r e filled at some point for equipment perservilt l on rath er than RCS cool i ns once RHR is aI/l n e d. Pl!r WCAP -15097 Section 4.2.2.4.1, CST voluml! determination should conservatively cons i der SG re fill to n/Gild proarammed, N/A Only applies to sei sm Ic event Of torn il d o missae. Volum e per al e. 8M*9S.()g61
-OOlV e r.S Onlv applies to seis mic eV!!nt or tornado m iss Jl e. Volum e p er c a lc. BM-9 S*0961-001 V e r.S Only appli es to seism ic event or torn a do mI ss il e. Failure of lin e establlshe s th e maximum prot ecte d CST volume of 166 25 0 K ill s. Credibl e to th is filiture concurr e nt wllh tornado or se.is m i c e vent. Due to any breach of t he tank or connect i n! line that allows a i , bene a th h'a dd e r. App Hca h'e to TOlnado E v e nt Unusable Volume 4 Inches from BOT N/A I nventory lOu dUll to MS/FW line break N/A N o te s! S. FSA R 15_4.1 , Sensi tiv i ty s tudy assumes lOSP, no reference to AFW 6. FSA R IS.3.i , lOSP assumed, I-MOAFW pump N/A N/A Only applie s to se ismic event or tornado No t credible tD ponulate thIS event m i ssile. Volum e pe r cil l e. BM*95-0961-con current with lornado or seismiC 00 1 Ver. S event_ On l y a pplies to s eismic event or tornado Not credib le to postulate this ellent m issil e. Volume per calc. BM-9S-0961-concurr o n t with tornado or seismic 00 1 Ver. S e v ent. Due t o any breach oflh e tank or conn e ctinA line th.1t allows air beneilth bl a d d er. App l icable to Seis mic E v e nt Not ued l bll! to pastulall!
this event COfIctJrrl!ll1 with lornildo or leismic event. Vortex N/A N/A Appficabil!
wino min submersente*
lIortex N/A N/' E2 -27 PeIF$AR 15.3.1 t Note 6) Decay Heat 011102% pow", 2nS MWT(lOO ")j OilIt2hrand6hrofevent.
Sensible heal based on Ta"lofS83'F(S77'F f 6'FJ N/A 14722 anum!! LOSP therefore no Reps w/LOSP; however , Initia l lO MWt net pump hut lnpulis. auumed to decay in one hour. N/' Not uedlble to poslulatl!
this eonnt concurrent with tornado or seismic event Not (r!:dlble 10 postulate this event concurrent with tornado or seismic !:lIent.
as Not credible to postulate this event (ono.urent wilh lornildo a, se i smic evenL Applicable wI no mltl. submerlence-vortex N/A Attachment -A-DOEJ *FRSNC419111-MOOl Veralon 2 Shee 2 of 3 Enclosure 2 to NL-13-1257 MSllw/olO!>l' powcr,11l1;d21I1a"dShrol M5nw/lOSl' 
....
pGWl!t,lSll'tlh.and6h'of nent.FSARTableIS4
*Uj,hI. '.,fSAItIS.2.a 1.Hrw..,/oLOSP 
... U.AdK..,h 
.. ;dbutd powefl Z115 MWI 10k" 10 MW'Inttpuml>h, ..
* EVcnt,ith." **
ISMWt. Unit IIpl!r;de<
In hOI .undby Ootp.nW,I"IIQ11 ,"ul on 102%powtr on H15MWt. I>ooIr,.,Modellt 
.. mI"Ilothealm01.phH.
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.. tor .. , Ith."umedthounlth,tIOl%oll17SMWt HOI.2 'floedonIOl"paw 
.. at2715MWI.
2h".lhot, I MO d llV<0"dlllo",.IId4 MFWlI ... flOSP Nol.) In,.,j 01'\ 102"poWCl.1 111SMWI. 2tI, ... 1 
...... 12115MWtI0l2"'.'hol O.a,t.cltb,. 
... onIOl"p-..
aI217SMWt,9hou 
** (nMo"') 101 )) 
..... "'''' 1 01 21 co"ditlo".
10" hau" 'lu",ln , 10 Ih. with 10 MWlllf neillep hUI conflde.ed und .. , hOlln 10 IIoI.hutciown I"HIII C\lHn It holl1andby cOfIdUlonlMd 41>""fllo hot .tuw:lby (l>ndltlol1$.IId" 1>0 .. " 10 hot .I'Iutdo_ .lnml"f1O 
, h. >Imo'phett per S..IibI." .. I.J ..... ... S."IibI.H 
** I * .,....,5A7f.HoLe.d M hl.$ubnt ..... II ... *VortU U.""'bI.VOI....,.
41nc1MItro.IQT cQI"e th.,m.' pOWef. cOte ttwnn., pI1fIIe,. .Iln!Kpho 
** HoHn..ole hut 'Ir\c.t unifwfl behetd.t HnMble Mat <o."eoo_fv.tIV S.n>llbleMIt oonun.ttilHly hoIlta/lliby conditlor11llvou&" thIa No 'efI,lbI.
hut oj"" ""'I ...
al hoi on T.., al SI?f (51rf. b .. 1td Oft T ...... 01 51?F ISn*r. eNr.llon. IIel. c.k. 8"'*!l5-(1961*001 ,tandby <ondlt!onllh 
...... lllhk du,.tion. 
".f. 6'fI:rSAR 154.2.1 2.i.F 6'F):fSARt5
.4.l.I.U.F 5 C*.IM.9S.09'H10IVef
.5 ... , edo" Netlie lllnpulfllIO M'Nt,er WCA9.I.SO,l .. d WCAP*1 47J l.fSAR5ItcUon IS **. 2.1doc1.noI'U$Umel c o m er utlwtyl u.lellCPhutfll lS MWI.A .... m. a <Go/down ** I.oI 5 0f/1u.'or.IIRCPI O pe"II" 1 I , a m I n. no*fiud 10 J5O'. Abo. include-INti'" M1 Pe. FSAII IS.l.I.2.1.,,) 
'OI"IPP'o 
... 10mimlorcachIlCS f5,o\1IIS2.U"",WC\PI47U,J.<<11o;"6.1
.IJ m.k ..... u"'plian".,.lJdj"'Repope 
** IIa .... Ihlll It k .. lUmeci 1h.1 1 RO>s ol',uUna I", h ** 1 ;"'pulo' 1 0 M\YtdKa.,.,..
eqlHlibor 
.... l1I.n IIICPop ..... I ... ,." aoppro ** lOmin,'O't.u.flCSeqlld 
....... ihln 1 owe. I hOUf. 0p"lI'IloII ol 0/1 rem.lnder CIt 9 ho .... 1n hot IICP OCIef;t\"e1.0r 
**
hou,* In "01 h.ft MCP"OI l lwd.,.t!ono' OfIOtIOfI.IICPh.' .... l ndulle1li>. 
... m."'...t.IICPh .. 1 I h.nl"lll ll..,pponedby.rnllll 
"",,",, . ducksthelnlrbllOMWtnecllCP"'Uinporl 1._ W<=-o tlnlh"".e md fH' 10MWt"et hutlnplll the FSAR 
'.Ict..-trlp.nd lhi.he.. belor. ,.adOl trip and hulls .u""" ... 'o OP' _'01\1(11'" II.swmtd 10 dKly In OIWI ItlIIou.. euylnoneU) 
........ P ...... ail lrom"ThomIdH 
.. br'. f H P Op.;I,,,",.nKddI"ll' pt fl pe<!l ve.lhe!.ltnorulofl torel , ftlht'.uItItdSG1".
f lUlt l d 5Goho"ldb 
* .-.IilIItd,1
_orne po inl 10. ,"uipmtl\l p .... IV't1onr.UrerthmflCS c ool ln$onteRHflIla/C"ed.Ptr WCAP*IS0I7StcUor!4.1
.1 *. I , C5 T....r"",.doe * .,ml,ulion hould<on
* p'Olllmmed.
NOIcr.dobl.eloponulltclhil Hal .. ...w.t'topo.IUlll.thlt 
* ""!\Iconcu.7enlwllhto",,do 
.... nICGflcu" .. lwllt.lo........
Hot u ..... W.I.pot' ...... lhlseve<lt 
.... le tfoil ....... oon< .... _ orH'krnk.........
_"tfllwlthltmldoOtl4!iunicnent. wl\htQmado",",",,*t""nt Hot credobl. 10 pcnluf.te lhi, NOluedibltlopo 
.. 
""n' <on<u".,,1 wilh lom.d.. .......1 (OIIcuru!M WIth 10m.do Nol uedobl.!o Uri' ev e nl Not ueclible.o pOllul.l" Ihlt ... "'1 01 Hltmlt; event. <01\(\1"1"111 WIth 100Ndo or t.dlmlt; cYeIIl. wllh lom.do 01 HlUllk L Ont, * ..,.I., oMiomlc...-or lor"..doonllIU 
** 1Ji1u'.0 ...... HUbhhH 1M m.odmcom protKledCST 
........ "'e01 16USO,;a!I
.Ctllfilrl.to poll .... lle I"'" ...... nl <ono"",' ",ilhlOl"..doortd\ll"lkcwnl. 
<Gnnfttlnc.M INIIhwI air bennlhblaoddtr
.w... .... bI. ,opolluln.thi, .... nt cortalrrcnlwittlto, ... door .. 11rn1elYl"l. *u"", ** adjKe"l" or<G_d-..:Iine'''''40W1 Olle , olllYb,uchof th. link or air bentllll trlldcf ... HOI <G1\IIIK1"'c lint th" 01""', Ilrbl"":llh " ... iblelop",Wlethb_," bladder. Hoi cred;ble 10 poIIl,IIal 
** hll conc .. ".nt wilh lOIn."" or .... nlconorfl'."'withloinadoCl'.tI,llllc Notcredib!elopoltullllthll.wefll<oncu 
.. ent 
.... enl. A,pp!QbI .... !flomin ApplicHIeW/nomfn. 
... bmHte .... I>.ppllc:lbll w!no min. 
........ rpno: .. *ytN'lt. :"'hc.MoI.w/,..mln
, lubrrw. ttlc.*...,ttu 
.. B.oedoncalwitllan40.oZ_. )1.11.''''"" 14. 26 ","h 1HP 11.II,I"'HI.14 H""lhfNI' Nofu-dwat 
** flrteb.lllciIUIIIIId 
... cI, Opltntionbol" .... AfWlO p.rallonlsolal...,AfWlo onfylollolno.rnalfeft!wale,du.,o Inv ... '...., Loullll. t o MS{fW .... bi n. 1 .. 1t ... SGwrlhirr 15 mlM. Ilulltd SG wllhln 15 mini. PIIfI"Ip, """IIt. [te. ","' ... nctlon LOlidul l oMS!fW .......... lrKlftm.... 
-"'P1Ie.blll leal>!. 2.U. Mlin 51 ...... Unt Oepr ... ul'l .. tIon* No ",,,,,,eo( MS .... Nt> lOSP. MWcrltdlttd and """heW" Mt lUndbotClClnNtionL Pto ...................
10 ... oI ... lltdanlft-hownlobe 
................ 
<G.t'_IO'I1I9II1 
.. oII.malnlil 
...... u... lSOI)IOf4 houro. ohutdowPl (J1HIINI-lnal l5OF)lor 4 III"'" IIIHRcut ...... l1SO')aI8hr FSAII.9.1.' 3 andT S,17.ti OKayH, ... lOon.) ,-l bI.hul b.aMCIon T ...... 01 SIJ'F !S7rf.6'F! . , . NofICPtw/lOSP , HJ.fI IMIatn .....,...a 
... "'<=-olOSP
.Ho ......... 10 MWtnetIlCPMallnpullt.6ded1lld
* ecayed .... orltw. ' .......... 
'.oml"horNoH 
.. bi!.fh'1' 
:::""'.:'0 
::::::::;::;ed SG.Thel;ault"'SGlnould,""liI.d .I>OIII.po/r!IIo.equ"' 
.... nl pr!l ...... lI'Ilon .. thlllha""C5(oolinf: 
.... ed. PlfWCN" lSOt15<<tion4.2
.2 .... 1.CSTwoIume dde-minltion"""_ld 
... " .. rv.tIveI, c .... sldI!tSGrelill'o" .. lo.d IIIOCrommltd. ", .. rJ",wlth SHe "1'1''''''
if WII d.te'lII .... dto rtmO\l.lhe5/G rdiI,"""" .... """'It is Indi ... "d ... <Gn.ef\<I1I 
.... In Iht Ho"n .. "bI.hulsI 
....... ni!wllbe htld .. hat II ..... ..,.,(ondill 
.... s"."dbl.hcll'lb.at.<<!on h .... ofS&l'fjSl1"f hrou,hlhhduratlon.llcfC.1c.. 
.6"f) 
..... "'SIl'flS17*f"'f) 1M*!lS-0961-GOIVe.
5 
.... Sn'.t.OO!r'ptw.
H II>. H/A s..tlIIae 
..... t.T .... S41F.N<rlo 
.. HoftCPlw/lOSP';FSA.Rinclat 
.. ..wY'/l 
.. lOSP.Howt""" 10 MWI net ACP hcll'l In""t It ,dd'" and dotayed o-r 1 hr . Pe.t",all._lho 
..... H.,bil.fNPOpo, .0"' .. 
.. I.1IO 'ilIOn 10 1.liI lh.'aulltd SG.Th.I.ultec!5G Ihouldbe.eliledilsom.poin,lo.
equlp",.nlp'"",IV.tk", .. tI.e,lh.nflC5 coolln.lo" ... IItMlsllit"ed I'erW(AP.
deIRlrnl"rMlon"'ould<Gme<V1tIvcIy comkl .. SG .. lilltono-load,... 
......... 
: d. P.,_U,..wlIhSNCIo<<min&.It_, d.t .. mi ..... lo ...... D ... th.S/G .. I.volu ... dnc.eill.lnditatedlli(onlllVllllmln,)o c WCAP1S097.
Jh ......
pIorIIOMWtnIllIlCP",,"lI'Iadd'" Ir"Ioddeu,,",ow.on.Ulhr.
P.' ...... II.o"" ..........
N .. bil,fNPOp.;t.om.n aca"de"lptnptcth 
** I)otreltno,e.sonlo,tfil hellulled SCi. Th.lllllied 
.. IOm.palnllo,tqulp"'.nlp'es.e 
..... atlan'.ttw' rhanRCScooIl,..onuItHllhltilnllf. P.,.WCAp*
150915t-c11on4.1
.2.4.I.C5T"" ....... "'t .. 
..
to"tJdct" 5G .. 
........ mcd. Perm .... ,.. .. i1h SHCliceltlin 
........ detannlne4to 
.........
lh. S/G ... i .............
_.,lhlndluted 
.. 1 (01'1 ....... 1'.'" In 'he WCAP 15Ul. ... S/G /WIiIVoI.llllld .. dedinf5AA9.l.6.
HoMW .... mp 1I.o.c. b ... bllure. HOI cre-dibl.,o p"rW9.2.61,,,,,nlcOll("''''' 
"""h tQ5.P; howwff.'nllmt nel hUI IrIput 01 10 Mwt d", 1o flCPI II beha of Ih. _III d,,,,,, In 1 Ilnunolln.I>,..khuoaunH'" m ..... II.mor,etdw.l 
... Noicr"'.we 10 po.lllbt.
this e ....... '
event ClCIII<Vr,enl 
... I)o lotflldo O. Not uedrbielO pc)I'UI.,.lh1 
* (orw::r.rrrelltlMthlotnldoOls.eitmic HC!tuedr.l>leIOP,"wlat.uu. 
.... nt ........ \ltnt.How 
..... r.Iir1e'allurelllndudtdln 
.1fII1loolKljrrtnt1lfllhlornlodo0l" Nolu ... 
< ... W"t"I ... ... r.l.CAn.' '''-tllJ*40 
... /FHPOptbol.tlono!
AfWlohlll,,.,jSG"';lhlnlO,,,InJ.
ApplIc,bll E2 -28 concu"",lwilhtomodoorMbmIc
__ ** fSAA9.2.'. !tIlmk L I>.fW._,.Rtclrr: 
...... .... N"AfWPlr",pAt<jn.M ...
pOItuf".thll 
.... nl ........ rr""I ... lthlo' ... doo. NOI u""bl.lo thl. nenl lellmlc IVII!nL lIoweve<. 1 .. Iu,e i.lndud<ed In '''efli <GII", ... "I with lom.do or OO .. CI ... enl ... lthtOlroadoo' 
........ k._1. 'SAIl 9.1.'. ...._ltlVll!"l. Inlt.U".II 
.. ptUfI Notu.d1blllopclllulalt,hllew-nt wl.hl""'.doOl!tllll1k 
....... r<t;howcv 
... mlnimu ... "'ltml<._nl:lh 
... eI"' * ..., ... 
... khlornadoOls.eltm l t ...... l. 
.. _rtdded .. m .. "". nI... MIIt.S\lbIl'lIIPM"VOI1 
.. pplltIlHw/nomin App!k.Wew/no 
..... 'ubmt'cenc.*vonn HIA .... enc;t'Woi,..
URlllllbi e Vol..,.,.
41n,1M 1 ,",'"lor ... 'ldon ..... 0.OJwr.10l>.""'lhHtI1l ... / 'HP 0,... 10",.110" 01 AFW 10 f.ulled SGwlthlnlOm 
.....
In ... e"toryloUdll.loMS/fW 
.... b' *** IfI"."loryI.O" plul"",mll 
--}}