ML20094G850

Station Blackout W/Rcic & RR Seal Leak

ML20094G850
Person / Time
Site:	Cooper
Issue date:	01/03/1992
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20094G841	List: ML20094G837 ML20094G844 ML20094G850
References
91-261, NSD920243, NUDOCS 9203030441
Download: ML20094G850 (25)

Text

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ATTACHMENT 2 TO NSD920243 COOPER NUCLEAR STATION NRC DOCKET NO. 50-298, DPR-46

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f,cm er%s af Nebraska Public Power District DESIGN CALCULATIONS COVER SHEET I

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Calculation No.

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x NPPL snerated Calculation Non NPPD Generated Calculation 1

Prepared By C [4

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Description:

l Modular Accident Anch.is Program was used to evaluate the timing of thermal-hydraulic phenomenon i

during a Station Ble.o - ut Scenario with RCIC available as the sole injection source. The results will be used to support the CNS response to 10CFR50.63 Station Blackout Rule.

References Attachtrents

1. SBO 10CFR50.63 submittals for CNS.

A. Archived ata Sets i

2. Letter EEll-9)-073, ENERCON Services to B. Input File Listing NPPD dated 10/17/91.

C. Summary File Listing

3. BWR Owners Group Emergency Procedure D. SBO Output Figures Guidelines, Rev 4.

l

4. CNS EP5.8 (EOPs), Rev. 5.

l

5. Accident Sequence Delineation Document, PRA-ET001, Rev. O, Scetion 3.2.1 (DRAPI').

3 i

6. PRA Procedure PRA-P0114D"APrfSl+9*

}

7. Terry Turbine Letter to IDCORE (GE), dated l

10/24/72 and GE Letter PFB-10-83, filed C25.1.5.3.

S. PRA System Notebook, PRA-SN010, Nuclear System Pressure Relief Systo.1, Rev. O.

Dessn Basis or

References:

Attauhments

..USAR NA A.

2. TECH. SPECS.

N/A B.

l

9. PRA System Notebook, PRA SN018, Reactor Core Isolation Cooling (RCIC) System, Rev. 0 l

10. MAAP 3.0B Rev.6 Users Manual.

~

11. NEDC 01-149, Rev. O.

12. MIPS Version L40 Users Manual.

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Rev.

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N1324389 NebraSk2 Public Pow:r District DESIGN CALCULATIONS SHEET ss..e / of 7 C h No.

9/ - 2fi/

NPPD Generated CAlculatbn Review of Non-NPPD STA n DA)

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19 1.0 Introdt.ction The effects of a station blackout (SBO) coupled with primary system leakage into the drywell were analyzed using a detailed deterministic simulation of a normal plant SCRAh! initiated from a main steam isolation valve (MSIV) closure.

2.0 Purpose A thermal. hydraulic analysis is r,eeded to support the CNS Station Blackeut Safety Evaluation Report (SER)[Ref.11. The CNS SBO for 10 CPR 50.63 will assume a small LOCA in combination with loss of offsite and failure of diesel AC power. This is to simulate the NRC assumed primary system leakage of 61 gpm for CNS [Ref. 21.

Information is needed on the drywell heatup rate and in particular the drywell temperature at the end of the CNS Station Blackout coping duration. The continued operation of RCIC under these containment conditions is also ofimportance.

t0 Descrintion The following sequence of events describes the SBO scenarie being analyzed. The operator actions modeled in this scenario are based on Rev. 4 of the BWR Owner's Group Emergency Procedure Guidelines [Ref. 31 which are inenrporated into Rev. 5 of the CNS EP5.8 (EOPs)[Ref. 41 A normal SCRAh! is initiated by MSIV closure with all forms of coolant injection including Control Rod Drive Pump SCRAM flow locked out. At level 2, RCIC and HPCI are allowed to autostart on low water level. Aner initial injection, HPCI is secured by the operator and RCIC continues to control level between level 2 and level

8. The operator maintains RCIC suction from Emergency Condensate Storage Tank for the duration of the event. The pressure in the vessel is controlled by the safety relief valves (D/F) operating on low. low set, which adds decay energy to the suppression pool. The two SRVs (D/F) are also used to reduce reactor pressure (emergency depressurization) according to the IIcat Capacity Temperature Limit (HCTL) of Graph 7, CNS EPS.8, Rev. 5 [Ref. 41. Since no low pressure injection systetas are available, the operator is directed not to depressurize the vessel to lower than 120 psig, per EOP.2 Step SP/I'.5. Throughout the event, primary containment heat removal systems (RHP Cont. Vent) are considered not available.

To prevent automatic : Jtiation of ADS in MAAP, ADS is locked out. This is not a required operator action for this sequence since neither LPCI or Core Spray systems are available. Therefore ADS inhibit is not required. Ilowever, this input is required due to the ADS system model in MAAP which initiates on high drywell pressure or low reactor water level.

i.0 Assumptions The T1 transient modeled in this analysis is representative of those delineated in the

(

loss of offsite power event tree, see Section 3.2.1 of the event tree document (Itef. 5].

The following system failures and requirements are needed to follow the event tree logic path.

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I N1324309 Nebraska Public Pocr District DESIGN CALCULATIONS SHEET 2 of_7_

ami Cale. No.

9/- 86/

NPPD Generated Calculation Review of Non-NPPD

. @A.7//h) Aggjcggg 7 Prepared By:

/hw Isf/)v, Generatts Calculations J/7N Ec/G.4dC Date; 3/ M 19k Company's Name m

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19 Offsite AC power and Diesel Generators are lost simultaneously at t=0, MSIVs are locked closed at time t=0.

Control rods insert upon scram signal.

Recirculation pumps f rip at t=0 and caastdown.

Core Spray, CRD, RilR and Drywell Vent systems are unavailable.

Several other assumptions were also needed to simulate the plant response and operator actions.

1.

Reset SRVs to low-low setpoint at t=0.

2.

The operator maintains RCIC suction from ECST, 3.

RCIC fails at 200"F suction temperature.

4.

Continuously monitor HCTL and emergency depressurize following EOPs using tw SRVs.

s 5.

IIPCI initially starts and then is manually secured after cycling off automatically.

6.

Drywell coolers are inoperable at the onset of the transient and remain off, 7.

A leak rate of 61 gpm exists from the RR pump seals at rated pressure.

8.

The convection heat losses from the RPV to the drywell are 4.8 MBtu/IIr at rated conditions.

9.

The numbar of reflective insulation plates is 11, each with a width of 0.292 ft, 10.

The initial bulk air temperature in the drywell and pedestal is 140"P at rated conditions.

11.

The initial bulk water temperature in the suppression pool and the emergency condensate storage tank is 90"F.

5.0 Edet hodolocv Deterministic calculations were performed using Modular Accident Analysis Program (MAAP) version 3.0B computer code [Ref.10,111. The executable code accessed is BMAAP.EXP and BMAAP.XMP with 3-121991,11:25a as the date and time tags for both these files.

5.1 Archive of MAAP Analysis The results of the analysis performed in this calculation were archived according to the

m324aes Nebraska Pub 0c Power Distnct DESIGN CALCULATIONS SHEET set.3_oi.2_

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19 PRA Procedure PRA-P011 [Ref. GL All input and output files are stored m 2Ti;007C90 WORM drive cartridge MAAP_#1, refer to Attachment A for specific file on information.

5.2 Computer Code input The MAAP input deck was prepared to utilize the flexibility of the MIPS preprocessor for organization and structure [Ref.121. The bases for the MAAP input is discussed below. The MAAP(MIPS) input deck construction reflects the requirements and assumpticns of Section 4.0. A brief discussion is provided for each item. Refer to Attachment C for particular items discussrd.

5.2.1 Bases For Assumption,s 1.

Once the first group of SRVs lift, the SRVs reset automatically to the Icw-low setpoint. [Ref. Sl. For this analysis, it is conservatively assumed that the SRVs go to low-low set immediately, at t=0. This will give the bounding suppression pool heatup rate for SRV low-low set operation.

2.

CNS IP5.8, Rev. 5 requires that 2C10 maintain uetion on the emergency condensate storage tank. Therefore no operato action is required since the normal lineup is from the ECST.

3.

The RCIC Terry Turbine lube oilis cooled with water taken from the suction line of the pump. The turbine design limits (Ref. 71 on the lube oil temperatnre indicate 200*F is the maximum oil temperature for continuous reliable operation of the turbine. Temperatures above this value would therefore cause RCIC failure (although not instantaneously). Note: this only beeoi..es a factor when taking suction from the torus.

4.

In CNS EP5.8, Rev. 5, EOP-2 Step SP/r-5 requires RPV pressure be reduced to stay within the constraints of Graph 7,IIcat Capacity Temperature Limit.

Tc depressurize the FPV, two SRVs are opened to simulate emergency depressurization, in accordance with EOP.1, Step RC.i-17. This is monitored continuously using MIPS to adjust SRV lift pressures.

5.

For the purposes of this study it is assumed that !!PCI is available initially, injecting coolant to the vessel for a single on.off cycle. After tripping oft on levet S. the operator secures llPCI and vessel level control is maintained using RCIC.

6.

The drywell coolers are dependent upon c BCCW for operation which in turn requires AC power to run, and AC is needed for.~an power also. Therefore it is assumed that the drywell coolers arc unavaihble for the duration of the scenario.

N1M4309 N:braska Public Power District DESIGN CALCULATIONS SHEET V or 7 she,i Calc, No.

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NPPD Generated C:lculation Review of Non.NPPD 47Miod_jft.Acftt)f Prepared Br.

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19 7.

The CNS Station Blackout as derined by the NRC assumes that the efTects of primary system leakage of 61 gpm be included in the thermal-hydraulic analysis. Each Reactor Recirculation Pump sealleaks at 18 gpm, for a total of 36 gpm from the primary system to the drywell. Then, including the Technical Specifications maximum allowable identified leakage adds another 25 gpm [Ref.11. To model the LOCA, the MAAP input parameter ZLOCA was given the RR suction nozzle elevation (930 fl, saturated liquid leak) and 8

ALOCA (leak area =0.00098 (1 ) was adjusted to obtain a single leak of 61 gpm from the reactor recirculation pump seals at rated RPV pressure (>1000 psia).

8.

The heat transfer rate between the reactor pressure vessel and drywell was calculated in NEDC 891439, Rev. 2 as 4,770,837 Btu /llr (or roughly 4.8 MBtu/Hr). Thin value is based on 1989-90 operational data.

9.

Tha reficctive insulation thickness and number of plates is listed iri the review comments of NEDC 90426, Rev. O as 11 plates at a thickness of 0.292 ft. This supr:edes the old value of 253 plates listed in the parameter file. The parameter file will be updated to reflect the change when Rev. O is approved.

10.

The initial bulk air temperature in the drywell and pedestal is based on NEDC 891439, Rev. 2 operating data from 1989-90. The maximum average air temperature is not expected to exceed 150*F during normal operation (includes zone 2A). For the purposes of the station blackout study, an average drywell temperature of 140"P is selected to prevent being overly conservative.

A separate perturbation oddresses the impact of 150*F initial temperature.

See Section 5.4.1.

11.

The initial suppression pool water temperature is 90*F which is the same initial value used in the design basis loss of coolant accident analysis reported in the USAR. The enthalpy of the condensate storage tank wateris also based on 90'F.

This is considered the upper bound for the expected ambient temperatures.

5.3 Computer Code Output The *. LOG file was reviewed for errors, warnings and diagnostic messages. No errors, warnings or diagnostics were found. Therefore program convergence is assured within all subroutines accessed.

No information was directed to the restart files. A summary of the sequence of events is given in the SUM file, refer to Attachment C. This file provides a chronological record of when trips ar.d setpoints were reached Important details of the analysis are summarized below, also refer to Section 5.4.

I

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N1324389 Nebraska Public Power Distnct DESIGN CALCULATIONS SHEET sn.1.i 5a 7 Catc. No.

9/ - 26 /

NPPD Generatedjalculation Review of Non-NPPD STAnod ELActeur vr.parea By:

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Time DMeription 3.5 see Reactor Scr m due to high pressure.

189 sec IIPCI starts to inject 194 see RCIC starts to inject.

307 see IIPCI locked out.

15 min Level re-estahlished, RCIC begins to cycle as needed to maintain level.

4.0 hrs End of run, core remains covered and containment is intact.

5.4 Graphical Results Attachment D shows the trended results of the analyses. VariaUes ofimportance are plot'ed. Iffurther iaformation is needed additional Ogures can be generated using the archive Oles. De6nitions of figure contents are provided below.

Figure 1:

TGDW -

Temperature of gas in the dgwell (F).

Figure 2:

PDW -

Pressure in the drywell (psia).

Figure 3:

XWSII -

fleight of water in the shroud (ref, to bottom of vessel, collapsed) (ft).

Figure 4:

TWSP Temperature of water in the suppression pool (F).

Figure 5:

PPS Pressure in primary system (psia).

l i

Figure 6:

WWLOCA - Flow rate ofliquid through the Recirculation l' ump Seals into l

drywell (Mibm/llr).

Figure 7:

QRVDW IIcat transfer rate between reactor pressure vessel and drywell j

(M Btu /If r).

l Figure 8:

PWW Pressure in the wetwell (psia),

l i

5.4.1 Goneral Observations The RPV reference elevations in MAAP are listed below:

i 1

Level 8 Ifigh Level Trip 47.65 ft i

Normal Water Level 46.0 ft i

Top of Active Fuel (TAF) 29.4 ft Vessel Bottom (I.D.)

0.0 ft (917 ft above Sea Leveh l

N1320389 Nebraska Public Power District DESIGN CALCULATIONS SHEET sseei 6 ot 7 Calc. No.

9/"28/

NPPD Generated Calculation Review of Non-NPPD STATios.1 PLMENAT Prepares ey; nid /$<>.Wm tJITH PZ/C 4A/D Date:

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19 The results after progressing 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the scenario, are summarized by the following conditions:

1.

Suppression pool temperature is roughly 160*F and IICTL limit is not reached, since depressurization is not required below 174*P.

2.

Drywell temperature increased to 263*F and wetwell pressure increased to 26 psia,11 psi below the RCIC turbine high exhaust pressure trip [Ref. 9].

3.

An additional case was executed which used identical input as Attachment B, but with an initial drywell (and pedestal) air tv.perature of 150*F (added 10*FT This perturbation demonstrated that le final drywell temperature a% a hours also increased 10*F.

10 Conclusion The progression of the 10 CFR 50.63 station blackout scenario, using primarily RCIC for core injection, indicates that containment gas temperatures do not exceed 281*F within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

E

N1324309 Nebraska Public Power District DESIGN CALCULATIONS SHEET sw.t 7 or 7 Calc. No.

9/ - 88 /

NPPD Generated Calculation Review of Non-NPPD S TAT 1on) BLACKOUT

• rieparea 01:

AS Aidfm Gene,atec caculatons

.B]/77/ MC MA/O /*f Date:

3 / /)db A s 19 Cornpany's Narrt EL (4AN

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19 NEDC 91-261 Attachment A This Attachment lists all of the MAAP (Rev. 6) input and output files which were part of this calculation. The date and time tags are used to determine the file revisionc used for a given analysis.

FILE NAME SIZE DATE TIME SBO-S3.AUW 1

12-31-91 2:29p S BO-S3.HUR 0

12-31-91 2:27p S BO-S3.H UW 870 12-31-91 2:51p SBO SS.INP 1442 12-31-91 2:34p r

SBO S3. LOG 33139 12-31 91 2:51p SHO-S3.PL1 64853 12-31 91 2:51p SBO S3.PL2 56805 12-31 91 2:51p S BO-S3.PL3 51445 12-31-91 2:51p SBO S3.PI/.

64231 12 31-91 2:51p S BO-S3. PIA 72067 12-31-91 2:51p SBO S3.PL6 64231 12-31-91 2:51p S BO-S3.PL7 125689 12-31-91 2:51p SBO-S3.RSR I

12 31-91 2:29p SBO-S3.RSW 50551 12 31 91 2:51p SBO-S3. SUM 8162 12-31-91 2:51p SBO-S3. TAB 59268 12-31-91 2:51p COOPR6DV. PAR 120959 12-31-91 2:33p COOPER.ISF 2221 09-28-91 11:39a COOPER.OSF 3912 09-28-91 11:45a f

l 1

NEDC 91-261 1

Attachment B VERBOSE SENSITIVITY TITLE

\\SBO-S3\\ CNS STATION BLACKOUT WITH RR SEAL LOCA COOPER REV 6 PARAMETER FILE 12-31-91, DV DRYWELL HEATUP CALCULATION - KES, NPPD 12-31-91 END SYMBOL TABLES INPUT SYMBOLS ARE IN C:\\MAAPEXEC\\MAAP\\MIPS\\ COOPER.ISE OUTPUT SYMBOLS ARE IN C:\\MAAPEXEC\\MAAP\\MIPS\\ COOPER.OSF END PARAMETER FILS COOPR6DV. PAR DO NOT LIST C

C INPUT FOR REACTOR DEPRESSURIZATION ALONG HCTL CURVE C

C PLEASE NOTE CONTROL BLOCK IS OVERRIDDEN BY ANY C

WHEN STATEMENTS THAT FOLLOW IN THE INPUT DECK, THUS C

MUST USE THIS HCTL WITHOUT WHEN BLOCK AS HERE. CANT USE C

AS ATTACH FILE HOWEVER.

C C

SRVS CONTROL PRESSURE BASED ON NOTE: MODIFIED HCTL C

HEAT CAPACITY TEMPERATURE LIMIT BASED ON LOW LOW SET, 2

FIGURE 2

-1, REV. 5 CNS EOP, ASSUME AN SRV WILL OPEN C

CHANGE SETPOINT AS TEMP INCREASES INITIALLY,-T.S. PG 59 C

REF. NCDC 89-1905, REV. 1 SRV/D_1029.7, SRV/F 1039.7 CONTROL SRV/D LOW SETPOINT USING POOL "'EMP 177.0 F 1029.7 PSIA 183.5 F 889.7 PSIA 190.7 F 739.7 PSIA 198.6 F 589.7 PSIA 207.5 F 439.7 PSIA 218.6 F 289.7 PSIA

.227.3 F 199.7 PSIA 22f.7 i 179.7 PSIA END CONTROL SRV/D HIGH SETPOINT USING POOL. TEMP

'177.0 F 1029.7 PSIA'

-183.5 F 889.7 PSIA 190.7 F 739.7 PSIA-198.6 F 589.7 PSIA 207.5 F 439.7 PSIA 218.6-F 289.7 PSIA 227.3 F 199.7 PSIA-229.7 F 179.7 PSIA END

-CONTROL SRV/F' LOW SETPOINT USING POOL TEMP 377.0 F 1039.7 PSIA

NEDC 91-261:-

2 Attachment B-

\\

.83.5 F 899.7 PSIA

\\

190.7 F 749.7 PSIA 198.6 F 599.7 PSIA 207.5 F 449.7 PSIA 218.6 F 299.7 PSIA 227.3 F 209.7 PSIA 229.7 F 189.7 PSIA END CONTROL SRV/F HIGH SETPOINT USING POOL TEMP 177.0 F 1039.7 PSIA 183.5 F 899.7 PSIA 190.7 F 749.7 PSIA

/

198.6 f*

599.7 PSIA 207.5 F

'449.7 PSIA 218.6 F 299.7 PSIA 227.3 F 209.7 P'IA 229.7 F 189.7 PSIA END C

C MAKE PARAMETER FILE CHANGES PRIOR C

TO BEGINNING CALCULATION C

PARAMETER CHANGES C

C SET INITIAL HEAT FLUX TO NED CALC INFO, 9

C ROUGHLY 4.8 MBTU/HR.

(NEDC 89-1439,REV 2)

C ALSO NOTE CHANGE TO #168 FROM PARAMETER C

FILE REV!EW OF NEDC 90-326.

01,167,4.8D6 01,168,11 C

C INITIAL BULK AIR TEMPERATURE IN THE

~

C DRYWELL AND PEDESTAL,-SEE NEDC 89-1439, C

REV. 02.

SET INITIAL BULK TEMPERATURE C

OF SUPP. POOL AND CONDENSATE STORAGE C

TANK TO DBA LOCA VI.:_UES (90 F).

03,07,0.01+;99 03,107,56.'zl8 09,08,140.

09,09,140.

09,10,90.

09,11,90.

C C

SMOOTH CONVERGENCE 0F SOLUTION

'C 11,3,0.004 11,4,0.005 C

C HIGH SUPP POOL LEVEL TO SWAP RCIC-

'C SUCTION - MAINTAIN SUCTION ON ECST C

03,206,885.0 C

)

d

t NEDC 91-261 3

Attachment B-C SMALL-SMALL LOCA-ON RECIRC PUMP SEALS IN DW C CHECK PLOT TO DFTERMINE FLOW RATE (AT LLSET C REACTOR PRESSURE.

C 01,47,930.0 01,48,0.00093 C

C ASSIGN BANDWIDTH FOR LOW LOW SET, C

ON F AND D.

C SRV/D DEADBAND IS 140.0 PSI SRV/F DEADE..ND IS 150.0 PSI END PARAMETER CHANGES NOLIST NOT A RESTART PRINT TIME 4 HRS FINAL TIME 4 HRS PARALLEL WHEN BEGIN LOCA ON AC POWER OFF DIESEL GEN OFF MSIVS CLOSE DW COOLERS OFF RHR HX1 OFF RHR HX2 OFF LPCII OFF LPCI2 OFF LPCI3 OFF LPCS OFF RCIC AUTO S S RCIC BLOCK HPCI AUTO S S HPCI BLOCK LPCS OFF ADS OFF CRD OFF-END C

'C CHANGE SRV GRP 1- (SRV/D) AND GRP.2 (SRV/F) TO C

NARROW ~ MARGIN A5 POOL: TEMP. INCREASES.

THIS WILL C

PREVENT DROPFING VESSEL PRESSURE BELOW 120 PSIA C

-WHEN POOL TEMP > 225 F PARAMETER CHANGES.

SRV/D DEADBAND 45.0 PSI 3RV/F DEADBAND.55.0 PSI END PARItMETER CHANGE END'

__l____,___.

---

i-------al-"

NEDC 91-261

.4 Attachment B C

C MANUALLY SErURE HPCI INJECTION C AFTER INIT1K.'aY AUTO STARTING C

WHEN EVENT CODE (78) IS FALSE AND EVENT CODE (39) IS TRUE

-HPCI TRIP END

p

_p d

NEDC 91-261 1

Attachment C 0.0 10 SCRAM SIGNAL RECIEVED 0.0 19 FEEDWATER CUMP TRIPPED 0.0 62 MSIV CLOSED 0.G 63 LOSS OF AC POWER (LOCKED) 0.0 65 RECIRC PUMP TRIPPED 0.0 66 TURBINE STOP VALVES CLOSED 0.0 70 PERMISSIBLE FOR RPT 0.0 71 CRD PUMP OFF 0.0 94 RESET HIGH DRYWELL PRESSURE TRIP FOR ADS 0.0 203 LPCI LOOP 1 LOCKED OFF 0.0 207 LPCS LOCKED OFF 0.0 215 MSIVS LOCKED CLOSED 0.0 226 ADS LOCKED CLOSED 0.0 228 RHR HTX #1 LOCKED OFF 0.0 230 RHR HTX #2 LOCKED OFF 0.0 232 LPCI LOOP 2 LOCKED OFF j

O.0 250 LOSS OF AC POWER 0.0 251 LOSS OF DIESEL POWER 0.0 256 BREAK IN PRIMARY SYSTEM (LOCA) 0.0 260 LPCI LOOP 3 LOCKED OFF 0.0 278 CRD PUMP LOCKED OFF 1.4 9

HIGH VESSEL PRESSURE SCRAM 3.5 64 REACTOR SCPJW2D 11.8 160 LOW LEVEL FOR SCRAM 38.0 78 LOW LEVEL TRIP FOR HPCI 38.0 82 LOW LEVEL TRIP FOR RCIC c

63.0 7

HFCI ON 67.9 29 RCIC ON 135.5 78 BESET LOW LEVEL TRIP FOR HPCI 135.5 82 RESET LOW LEVEL TRIP -~r'OR RCIC 135.6 78 LOW LEVEL TRIP FOR HPCI 135.6 82 LOW LEVEL TRIP FOR RCIC 136.8 78 RESET LO' LEVEL TRIP FOR HPCI 136.8 82 RESET 'sd LEVEL TRIP FOR RCIC 141.1 79 H1C" DRYWELL PRESSURE - FOR HPCI 141.1 161 HIGH DRYWELL PRESSURE FOR SCRAM 226.2 7

HPCI OFF-226.2 29 RCIC OFF 226.2 39 LEVEL 8 HIGH MATER LEVEL 227.1 79 RESET HIGH DW PRESS. E3R HPCI 227.1 201 HPCI LOCKED OFF 1025.3

-29 RCIC ON 1025.3 39 RESET LEVEL 8 TRIP 2159.2 29 RCIC OFF 2159.2 39

-LEVEL 8 HIGH WATER LEVEL 2168.2 29 RCIC ON 2168.2 39 RESET LEVEL 8 TRIP 2169.4 29 RCIC OFF 2169.4 39

-LEVEL 8 HIGH WATER LEVEL i

2169.8 29 RCIC ON 2169.8 39 RESET LEVEL 8 TRIP 2924.4 29 RCIC OFF-2924.4 39 LEVEL 8 HIGH WATER LEVEL 2979.7 29 RCIC ON 2979.7 39 RESET LEVEL 6 TRIP

____._____..J

NEDC 91-261 2

Attachment-C 3752.6 29 RCIC OFF 3752.6 39 LEVEL 8 H:GH WATER LEVEL 3884.7 29 RCIC ON 3884.7 39 RESET LEVEL 8 TRIP 4663.9 29 RCIC OFF 4663.9 39 LEVEL 8 HIGH WATER LEVEL 4861.6 29 RCIC ON 4861.6 39 RESET LEVEL 8 TRIP 4865.4 29 RCIC OFF 4865.4 39 LEVEL 8 HIGH WATER LEVEL 4865.4 29 RCIC ON 4865.4 39 RESET LEVEL 8 TRIP 5637.9 29 RCIC OFF 5637.9 39 LEVEL 8 HIGH WATER LEVEL 5896.4 29 RCIC ON 5896.4 39 RESET LEVEL 8 TRIP 5902.1 29 RCIC OFF 5902.1 39 LEVEL 8 HIGH WATER LEVEL 5904.2 29 RCIC ON 5904.2 39 RESET LEVEL 8 TRIP 6707.3 29 RCIC OFF 6707.3 39 LEVEL 8 HIGH WATER LEVEL 7024.7 29 RCIC ON 7024.7 39 RESET LEVEL 8 TRIP 7834.0 29 RCIC OFF 7834.0 39 LEVEL 8 HIGH WATER LEVEL 8197.2 79 RCIC ON 8197.2 39 RESET LEVEL 8 TRIP 8204.2 29 RCIC OFF 8204.2 39 LEVEL 8 HIGH WATER LEVEL 8204.5 29 RCIC ON 8204.5 39 RESET LEVEL 8 TRIP 9000.9 29 RCIC OFF 9000.9 39 LEVEL 8 HIGH WATER LEVEL 9416.0 29 RCIC ON 9416.0 39 RESET LEVEL 8 TRIP 10244.1 29 RCIC OFF

_10244.1 9

LEVEL 8 HIGH WATER LEVEL-3 10707.1 29 RCIC ON 10707.1 39 RESET LEVEL 8 TRIP

.11529.2 29 RCIC OFF 11529.2 39 LEVEL 8 HIGH WATER LEVEL 12023.5 29 RCIC ON 12023.5

-39 RESET LEVEL 8 TRIP 12837.8 29 RCIC OFF 12837.8 39 LEVEL 8 HIGH WATER LEVEL 13379.6 29 RCIC ON 13379.6 39 RESET LEVEL 8 TRIP 14188.1 29 RCIC OFF 14188.1 39 LEVEL 8 HIGH WATER LEVEL 14400.0 159 BATTERY POWER UMAVAILABLE i

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^

CALCULATION DESIGN VERIFICATION PACE t

OF 1

1.0 The following basic questions shall be addressed, as applicable during the performance of any design verification.

These questions are taken from ANSI N45.2.ll-1974, 1.1 Were the design inputs, per Procedure 3. 4. :', correctly selected and incorporated into the design?

1.2 Are the latest applicable revisions of design documents utilized?

1.3 Are assumptions necessary to perform the design activity adequately described and reasonable?

When necessary, are the assumptions identified for subsequent re-verifications when the detailed design activities are complete?

~

1.4 Are the applicable codes, standards and regulatory requirements, including issue and addenda, properly identified and are their requirements and design met?

1.5 Have applicable construction and operating experience been considered?

1.6 Have the design interface requirements been satisfied?

I 1.7 Was an appropriate design method used?

1.8 Is the output reasonable ccmpared to inputs?

?

1.9 Are the specified parts. equipment, and processes suitable for the required application?

1.10 Are the specified materials compatible with each other and the design environmental conditions to which the material vill be exposed?

1.11 Has the design properly considered radiation exposure to the public and plant personnel?

1.12 Has this design adequately considered hazards such as missiles, jet impingement, etc.?

1.13 Has the design adequately considered seismic requirements, barge impact, and Mark I loadings as appropriate?

Calculation Number: N d @ l-M (, Revision: O,- 1.as-been reviewed for the above design verification. 41ements and, the appropriate design verification elements have been ade untely a dressed.

Design Verifier:

i grs/

A[

Date:

I

/

i '

N j PROCEDURE NUMBER 3.4.7 l REVISION NUMBER 2

PACE 14 0F 14

-