ML20136B543
| ML20136B543 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/02/1997 |
| From: | Sparacino J, Streibich M DUKE ENGINEERING & SERVICES |
| To: | |
| Shared Package | |
| ML20136B539 | List: |
| References | |
| V609.000.00001, V609.000.00001-R01, V609.000.00001-R1, NUDOCS 9703110021 | |
| Download: ML20136B543 (18) | |
Text
.
' p CALCULATION
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DukeEmineeriq& Services enz;, no; ygoo,ooo,oooo1 PROJECT NAME:
CUENT:
Monticello Nuclear Station Northern States Power Co.
CALCULATION TITLE Low Pressure Emergency Core Cooling System (ECCS) Net Positive Suction Head (NPSH).
PROBLEM STATEMENT OR OBJECTIVE OF CALCULATION The Objective of this Calculation is to calculate the Net Positive Suction Head (NPSH) to the IOw pressure ECCS pumps for the specified Torus conditions.
C DOCUMENT AFFECTED REVISION PROJECT ENGINEER NAMES AND INITIALS REVISION PAGES DESCRIPTION APPROVAL /DpTE OF PREPARERS AND CHECKERS 1
Client PREPARER / DATE s,
Comments seph D.
Sparacino Revised model
/@
D, h4 to incorporate
$ /d/2//76 pipe inner diameter changes.
CHECKER / DATE Matthew W.
Streibich 1
$224/ - h a d.' -: < -
9%M
.; r 7 9703110021 970304 PDR ADOCK 05000263 P
PDR PAGE 1 0F ;/
Page 1-2
DE&S Naperville, Illinois 1
oEOJECT Monticello Nuclear Station siaJ No:
V609.000.00001 I
NMER Northern States Power Co.
Calc No:
V609.000.00001 CLIENT Monticello Nuclear Station 4
REFERENCE TABLE OF CONTENTS Section Description Pace No.
l 1.0 Purpose / Objective 4
2.0 Methodology / Acceptance Cnteria 4
3.0 Assumptions 5
4.0 Design input 6
5.0 References 8
6.0 Calculations 10 7.0 Summary and Conclusions 11 List Of Attachments i
A.
CORE SPRAY, RESIDUAL HEAT REMOVAL, AND RING HEADER MODELED i page 12 PIPING INFORMATION LIST.
B. FLO-SERIES MODEL PIPING AND NETWORK INPUT, AND OUTPUT FOR page 26 CASE #1.
C. FLO-SERIES MODEL PIPING AND NETWORK INPUT, AND OUTPUT FOR page 52 CASE #2.
D. FLO-SERIES MODEL PIPING AND NETWORK INPUT, AND OUTPUT FOR p,g, 7g CASE #3.
i E, FLO-SERIES MODEL PIPING AND NETWORK INPUT, AND OUTPUT FOR page 105 CASE #4.
4 REVISION 1
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DE&S Naperville, Illinois
. PROJECT Monticello Nuclear Station File No:
V609.000.00001 j
OWNER Northern States Power Co.
Calc no:
v609.000.00001 CLIENT Monticello Nuclear Station I
REFERENCE LIST of ATTACHMENTS 1
Attachments Title No Of Paaes e
i A
CORE SPRAY, RESIDUAL HEAT REMOVAL.
14 AND RING HEADER MODELED PIPING
'l l
INFORMATION LIST.
8 FLO-SERIES MODEL PIPING AND NETWORK 26 INPUT, AND OUTPUT FOR CASE #1.
i C
FLO-SERIES MODEL PIPING AND NETWORK 26 l
INPUT, AND OUTPUT FOR CASE #2.
D FLO-SERIES MODEL PIPING AND NETWORK 27 INPUT, AND OUTPUT FOR CASE #3.
E FLO-SERIES MODEL PIPING AND NETWORK 27 j
!NPUT AND OUTPUT FOR CASE #4.
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REVISION 1
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DE&S Naperville, Illinois l
PROJECT Monticello Nuclear Station File No:
V609.000.00001 j
cwnER Northern States Power Co.
calc No:
V609.000.00001 CLIENT Monticello Nuclear Station 4
REFERENCE 1.0 PURPOSE / OBJECTIVE The purpose of this calculation is to determine the NPSH delivered to the Residual Heat Removal (RHR) and Core Spray (CS) pumps for specified pump flows, Torus pressures, temperatures, and water levels.
2.0 METHODOLOGY /ACCEF ' AivCE CRITERIA 2.1 Analysis Method 2.1.1 Introduction This analysis will use FLO-SERIES software, version 5.01, to create a model of the Torus ring header and suction piping for both loops of the Core Spray and Residual Heat Removal systems. Specific data, which is representative of a particular Torus and plant conditions, will be the input into these models. The model output data will be used to calculate the NPSH of each running pump.
2.1.2 FLO SERIES results The FLO-SERIES model of the Torus ring header and ECCS suction pipir.g allows the specified initial Torus conditions, such as pressure and temperature, to be input into the model with the purpose of receiving the outputs needed for the NPSH calculations. The specific FLO-SERIES outputs required for the NPSH calculations are fluid velocity and pressure at the suction of the particular pump.
2.1.3 NPSH Calculations Given the inputs of fluid velocity and pressure at the pump suction, the NPSH available to any of the RHR or CS loops (A or B)is calculated using the following equation.
2 NPSHA = Hb/ - Hva/y + Ps/ + Z + Vs /2g
- 5.22 7
7 where; NPSHA = net positive suction head available, ft.
2 Hb = atmospheric pressure,2116.8 lb/ft.
5.23 REVISION 1
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Naperville, Illinois PROJECT Monticello Nuclear Station File No:
V609.000.00001 CWNER Northern States Power Co.
Calc No:
V609.000.00001 CLIENT Monticello Nuclear Station REFERENCE 2
Hva = vapor pressure at the specified fluid temp, Ib/ft,
Ps = pressure of the fluid at the suction of the pump,Ib/ft'.
3 y = specific weight of fluid at pumping temperature, Ib/ft = (1/ specific volume).
Vs = average velocity of the fluid at the suction of the pump, ft/s.
Z = vertical distance between centerline of pump and indication of Ps, = 0.0 ft.
g = 32.2 ft/s'.
2.1.4 Acceptance Criteria The calculated NPSHA meets or exceeds the NPSH required (NPSHR) given for
)
each particular pump.
4 i
3.0 ASSUMPTIONS 4
)
3.1 The Torus ring header suction strainer differential pressure is equivalent to 1.0 ft, of 5.14, 5.27 water at 10,000 gpm. The friction loss due to the strainers was modeled as a fixed friction loss.
3.2 The Torus ring header suction strainer ("D" strainer) located at 315' on the ring 5.13, 5.27 header is modeled as completely plugged.
3.3 The majority of the Torus ring headeris modeled as a 20" diameter schedule 20 pipe :
with an inner diameter of 19.250". The parts of the Torus ring header around the 5.1, 5.23 penetrations are conservatively modeled as a 20" diameter schedule 60 pipe with an j
inner diameter of 18.376". The actual inner diameter of the reinforced area of the penetrations is 18. 500". This is the closest this pipe could be modeled to its actual size. By using a smaller diameter pipe than actual, the results will be conservative since higher line losses will result. The reduction and expansion of inner diameters due to the reinforced penetrations were modeled as fixed resistance.
34 The Torus ring header suction strainer piping is conservatively modeled as a 20" diameter schedule 60 pipe with an inner diameter of 18.376". The actual inner 5.8, 5.23, diameter is 18.500". This is the closest that this pipe could be modeled to its actual 5.28 size. By using a smaller diameter pipe than actual, the results will be conservative since higher line losses will result.
3.5 The Core Spray piping was conservatively modeled as 12" diameter schedule 40 piping which has an inner diameter of 11.938" The actual piping schedule is 40S,
- 5. 9, 5.23, which has an inner diameter of 12.000". This is the closest that this pipe could be 5.28 F.IVIS Ion
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DE&S Naperville, Illinois i
PROJECT Monticello Nuclear Station File No; V609.000.00001 OWNER Northern States Power Co.
Calc No:
V609.000.00001 CLIENT Monticello Nuclear Station i
i 1
REFERENCE modeled to its actual size. By using a smaller diameter pipe than actual, the results Will be conservative since higher line losses will result.
l 1
4.0 DESIGN INPUT 4.1 All pipe lengths, fittings, and elevations used to model the system, were obtained from the applicable drawings listed in references 5.1 through 5.10. This information is i
listed in Attachment A, and stated in the modeled pipe lists in Attachments B, C, D, and E.
4.2 Case #1 j
PARAMETERS VALUE Primary Containment Pressure 24.7 psia 5.11, 5.27 l
i Torus Water Temperature 147'F 5.11, 5.27 Torus Water Level 907'-9.3" 5.12,5.27 "A" CS Pump Flow 4600 gpm 5.15,5.27 "B" CS Pump Flow 4600 gpm 5.15,5.27 "A" RHR Pump Flow 3788 gpm 5.15,5.27 "B" RHR Pump Flow 3788 gpm 5.15,5.27
' 'C" RHR Pump Flow 3788 gpm 5.15,5.27 "D" RHR Pump Flow 3788 gpm 5.15,5.27 "A" CS Required NPSH 33.0 ft.
5.24,5.27 "B" CS Required NPSH 33.0 ft.
5.24, 5.27 "A" RHR Required NPSH 25.7 ft 5.25,5.27 "B" RHR Required NPSH 25.7 ft.
5.25,5.27 "C" RHR Required NPSH 25.7 ft.
5.25,5.27 "D" RHR Required NPSH 25.7 ft.
5.25,5.27 Hva 496.944 lb/ft*
5.26 3
y = 1/ specific volume 61.248 lb/ft 5.26 4.3 Case #2 PARAMETERS VALUE Primary Containment Pressure 18.82 psia 5.20,5.27 Torus Water Temperature 147'F 5.21, 5.27 Torus Water Level 907'-9.3" 5.12,5.27 "A" CS Pump Flow 4600 gpm 5.15,5.27 "B" CS Pump Flow 4600 gpm 5.15,5.27 "A" RHR Pump Flow 3788 gpm 5.15,5.27 "B" RHR Pump Flow 3788 gprn 5.15, 5.27 "C" RHR Pump Flow 3788 gpm 5.15,5.27 "O" RHR Pump Flow 3788 gpm 5.15,5.27 REVISION 1
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DE&S
~
l Naperville, Illinois 1
?ROJECT Monticello Nuclear Station File No:
V609.000.00001 CwMER Northern States Power Co.
calc No:
V609.000.00001 CLIENT Monticello Nuclear Station J
REFERENCE i,
5.24, 5.27 I
5.24, 5.27 "A" RHR Required NPSH 25.7 ft.
5.25, 5.27 "B" RHR Required NPSH 25.7 ft, 5.25.5.27 "C" RHR Required NPSH 25.7 ft.
5.25,5.27 l
"D" RHR Required NPSH 25.7 ft.
5.25.5.27 2
4 Hva 496.944 lb/ft 5.26 4
3 y = 1/ specific volume 61.248 lb/ft 5.26 i
1 4.4 Case #3 PARAMETERS VALUE Primary Containment Pressure 22.5 psia 5.11, 5.27 Torus Water Temperatu,e 184 'F 5.11, 5.27 Torus Water Level 907'-9.3" 5.12, 5.27 1
"A" CS Pump Flow 3310 gpm 5.15,5.27 i
"B" CS Pump Flow 0gpm 5.15,5.27 "A" RHR Pump Flow 4000 gpm 5.15,5.27 l
"B" RHR Pump Flow 0 gpm 5.15,5.27 "C" RHR Pump Fiow 0gpm 5.15,5.27 "D" RHR Purr.p Flow 0gpm 5.15,5.27 "A" CS Required NPSH 29.0 ft.
5.24, 5.27 "A" RHR Required NPSH 26.0 ft.
5.25,5.27
)
2 Hva 1181.232 lb/ft 5.26 4
y = 1/ specific volume 60.481 lb/ft' 5.26 j
4 4.5 Case #4 i
PARAMETERS VALUE Pnmary Containment Pressure 22.85 psia 5.20,5.27 Torus Water Temperature 184 *F 5.21,5.27
+
Torus Water Level 907' 9.3" 5.12,5.27 "A" CS Pump Flow 3310 gpm 5.15, 5.27 4
l "B" CS Pump Flow 0gpm 5.15, 5.27 "A" RHR Pump Flow 4000 gpm 5.15,5.27 "B" RHR Pump Flow 0gpm 5.15, 5.27 "C" RHR Pump Flow 0gpm 5.15,5.27 "D' RHR Pump Flow 0 gpm 5.15,5.27 "A" CS Required NPSH 29.0 ft.
5.24, 5.27 "A" RHR Required NPSH 26.0 ft.
5.25, 5.27 2
Hva 1181.232 lb/ft 5.26 y : 1/ specific volume 60.481 lb/ft' 5.26 4
i i
REVISION 1
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Naperville, Illinois PROJECT Monticello Nuclear Station File No:
V609.000.00001 CWER Northern States Power Co.
Calc No:
V609.000.00001 OIENT Monticello Nuclear Station
]
REFERENCE J
j
5.0 REFERENCES
j i
I 5.1 Monticello Station Drawing No. NX-8291-51 rev. A. "20 Dia. Header for Suppression chamber."
i 5.2 Monticello Station Drawing No. NF-36371 rev. B. " Area 3 - Piping Drawing Plan 1
Below EL. 923'-0"."
i 5.3 Monticello Station Drawing No. NH-36246 rev. BC. "P & ID Residual Heat Removal 1
j System."
l 5.4 Monticello Station Drawing No. NH-36247 rev. BB. "P & ID Residual Heat Removal l
4 System."
l l
5.5 Monticello Station Drawing No. NH-36248 rev. AF. "P & ID Core Spray System."
j 5.6 Monticello Station Drawing No. NH-36250 rev. W. "P & ID High Pressure Coolant injection System."
i 5.7 Monticello Station Drawing No. NH-36252 rev. X. "P & ID RCIC (Water Side) i System."
5.8 Monticello Station Drawing No. NX-13142-17 rev. E. " Torus Water?
l 5.9 Monticello Station Drawing No. NX-13142-20 rev. H. " Torus Water."
1 5.10 Monticello Station Drawing No. NX-13142-41 rev. D. "RCIC and CRD Suction Piping."
l 5.11 GE-NE-T2300731-1, Containment Response Evaluation, Task 6.0, Table 6.
5.12 CA 93-05B Supp)ression Pool Drawdown Calculation, Rev. O, (instrument zero elevation, d10 0 - 14.7"= 907'-9. 3" 5.13 NX 8291-51' C RHR pumps, conservative. Location of strainer on suction header between penetration Spray and A/
e 5.14 NX=8291-51, Design requirement for the suction strainers.
EqDE-34-0687, CS and RHR maximum flow to reactor (short term) values.
5.1.5 1
c.16 USAR, Revision 14 Figure 5.2-15,50% of containment pressure at 600 seconds plusi 14.7 psia, (8.24X0. )
.7=18.82 psia.
l 5.17 USAR, Revision 14, Figure 5.2-17 Suppression pool temperature at 600 sec.
5.18 PPT - Peak Pool Temperature, NPSH available is minimum at peak pool temperature-1 REVISION 1
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DE&S Naperville, Illinois PPoJECT Monticello Nuclear Station File No:
V609.000.00001 Nf 3 ER Northern States Power Co.
cale No:
V609.000.00001 CLIENT Monticello Nuclear Station REFERENCE 5.19 EqDE-34-0687, CS and RHR flow values for long term case 1, (OOS - Out Of Service) 5.20 USAR, Revision 14 Fi ure 5.2-15. 50% of containment pressure at time of PPT plus 14.7 psia, (16.3X0.6)+ f4.7=22.85 psia.
5.21 USAR, Revision 14, Figure 5?.-17, Peak suppression pool temperature.
5.22 Pump Handbook 2nd edition,1986, Igor J. Karassik.
5.23 Mark's Standard Handbook for Mechanical Engineers,9th edition, Eugene A.
Avallone.
5.24 NX-7833-23&24 5.25 NX-7905-51 &52 5.26 ASME Steam Tables, Fifth Edition, The American Society of Mechanical Engineers.
5.27 Monticello Nuclear Generatino Plant Fax Transmittal from: Pat Tobin, To: Steve Huodleston, subject: "Inout da'la and references for NPSH calculation for CS and RHR pumps" dated 12/23/96. This fax also contains a letter explaining NPSH at higher fluid temperatures.
5.28 Monticello piping specification 5828-M-40, NL-88187, Rev. 2.
i REVISION 1
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PROJECT Monticello Nuclear Station File Nr
- V609.000.00001 CWNER Northern States Power Co.
Calc No:
V609.000.00001
]
CLIENT Monticello Nuclear Station i,
l t
4 REFERENCE
- 6.0 CALCULATION j
6.1 FLO-SERIES Modeling' The nodal diagram of each modelis contained in each of the model attachments.
These diagrams depict the ring header and all of the ECCS suction piping. 'in each of the model runs, the High Pressure Core injection (HPCI) system, and Reactor Core Isolation Cooiing (RCIC) system have been isolated from the ri g header.
6.2 The table below lists the fluid pressure and velocity at the suction of each pump, as well as the results of the NPSHA calculations. The fluid pressures and velocities are results of the FLO-SERIES MODEL runs and can be seen in attachments B, C, D, and E.
l PUMP PRESSURE VELOCITY NPSHA l
"A" CS 8.847 psi 13.200 ft/sec 49.954 ft.
"B" CS 9.231 psi 13.200 ft/sec 50.814 ft.
Case #1 "A" RHR 10.94 psi 8.821 ft/sec 53.377 ft.
"B" RHR 10.55 psi 8.821 ft/sec 52.460 ft.
"C" RHR 10.64 psi 8.821 ft/sec 52.672 ft.
"D" RHR 10.85 psi 8.821 ft/sec 53.165 ft.
"A" CS 2.967 psi 13.200 ft/sec 36.129 ft.
"B" CS 3.333 psi 13.200 ft/sec 36.990 ft.
Case #2 "A" RHR 5.061 psi 8.821 ft/sec 39.555 ft.
"B" RHR 4.669 psi 8.821 ft/sec 38.633 ft.
"C" RHR 4.760 psi 8.821 ft/sec 38.847 ft.
"D" RHR 4.970 psi 8.821 ft/sec 39.341 ft.
Case #3 "A" CS 9.608 psi 9.495 ft/sec 39.744 ft.
"A" RHR 10.73 psi 9.315 ft/sec 42.363 ft.
Case #4 "A" CS 9.958 psi 9.495 ft/sec 40.578 ft.
"A" RHR 11.08 psi 9.316 ft/sec 43.196 ft.
TABLE 6.1 FLO-SEREIS DATA AND NPSHA RESULTS REVISION 1
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DE&S Naperville, Illinois T
File No:
V609.000,00001 PROJECT Monticello Nuclear Station CW!!ER Northern States Power Co.
Calc :::
V609.000.00001 CLIENT Monticello Nuclear Station d
.i a
REFERENCE 1
i 7.0
SUMMARY
& CONCLUSIONS
7.1 Results
The resuits of the FLO-SERIES model runs for each case are attached to this calculation in each of attachments B,C, D, and E. The results of the NPSHA calculations are listed in section 6 for each case. The NPSHA
,l results are also shown below with the NPSHR for each of the pumps in each case, i
7.1.1 The calculated NPSHA and the associated NPSHR is shown below for each pump in
~
each of the cases.
l PUMP NPSHA NPSHR l
3 "A" CS 49.954 ft.
33.0 ft.
4 "B" CS 50.814 ft.
33.0 ft.
4 Case #1 "A" RHR 53.377 ft.
25.7 ft.
"B" RHR 52.460 ft.
25.7 ft.
"C" RHR 52.672 ft.
25.7 ft.
"D" RHR 53.165 ft.
25.7 ft.
"A" CS 36.129 ft.
33.0 ft.
i "B" CS 36.990 ft.
33.0 ft.
Case #2 "A" RHR 39.555 ft.
25.7 ft.
"B" RHR 38.633 ft.
25.7 ft.
"C" RHR 38.847 ft.
25.7 ft.
}
"D" RHR 39.341 ft.
25.7 ft.
i Case #3 "A" CS 39.744 ft.
29.0 ft.
"A" RHR 42.363 ft.
26.0ft.
_ Case #4 "A" CS 40.578 ft.
29.0 ft.
"A" RHR 43.196 ft.
26.0 ft.
I TABLE 7.1 NPSH COMPARISCN 4
7.2
==
Conclusion:==
The NPSHA that was calculated for each pump is shown to be larger than its associated NPSHR in Table 7.1. Therefore, one can conclude that for the given initial conditions that there will be adequate NPSH for each of the CS and RHR pumps.
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l Response to USNRC Request For AdditionalInformation 2
March 4,1997 4
c SUPPLEMENTAL INFORMATION FOR LICENSE AMENDMENT REOUEST DATED JANUARY 23.1997 TORUS ATTACHED PIPING EVALUATION Torus attached piping has been conservatively analyzed for 195'F by Duke Engineering and Services. The results of this analysis are summarized below.
Acceptance Criteria In accordance with Section 12.2.2.12.2 of the Updated Safety Analysis Report (USARL Torus attached piping is analyzed to withstand loss of coolant accident (LOCA) related loads and safety relief valve (SRV) discharge-related loads postulated to occur during a LOCA or a SRV discharge event defined by Nuclear Regulatory Commission (NRC) Safety Evaluation Report NUREG-0561.
NUREG-0661 requires compliance with the American Society of Mechanical Engineers (ASME) Code,Section III 1977 Edition to evaluate the acceptability of torus attached piping designs.
' Analysis Assumptions Torus attached piping was analyzed for a peak suppression pool temperature of 184 F in March,1995, following the update of the Monticello design basis accident containment pressure and temperature response analysis reported in NEDO-32418.
For anticipated future increases in licensed thermal power of up to 1880 MWT, the current peak suppression pool temperature of 184 F will increase to 193 F with river water temperature at the upper limit of 90 F specified in Section 6.2.3.2.4.c of the USAR.
For conservatism, the piping analysis was reperformed for a suppression pool tempercure of 195 F.
Methodoloav The postulated Mark I containment hydrodynamic loads for operation up to 1880 MWT are bounded by those determined fe: the original Mark I containment Page 2 - 1
_ ~..
_ Response to USNRC Request For AdditionalInformation
. March 4,1997
' Attachment 2 l
analysis. This is primarily due to the fact that the reactor operating pressure will
[
not be increased for power rerate and improved models were used for calculating the vessel blowdown. Therefore, there will be no increase in piping stresses, j
support loads and stresses in the torus penetrations due to the Mark I hydmdynamic loads.
1 The SRV discharge line dynamic loads are influenced by parameters such as SRV discharge line geometry, torus geometry, water leg length, and SRV flow rate which is linearly proportional to the SRV opening pressure. Of these parameters, only the SRV opening set point pressure is potentially affected by rerate. The SRV opening setpoint pressures will not be changed since the reactor pressure will remain unchanged for operation up to 1880 MWT. Therefore, the SRV discharge dynamic loads will remain within the original SRV loaa definition.
- There will be no increase in piping stresses, support loads and stresses in the torus
. penetrations due to the SRV discharge line dynamic loads Torus attached piping was designed for the suppression pool design pressure of 56 psig. The suppression pool pressure under rerate conditions will still be bounded by the current suppression pool design pressure.
Based on the above, it is concluded that the impact of power rerate on' torus attached piping will be limited to the impact of higher peak suppression pool temperature. The affected torus attached lines which require evaluation include large bore and small bore lines of the following systems.
SRV Discharge Lines Residual Heat Removal Reactor Core Isolation Cooling offgas line (steam side) a Primary Containment & Atmospheric Control Core Spray High Pressure Coolant Injection (water side)
Reactor Core Isolation Cooling (water side)
Primary Containn:ent Nitrogen Control Miscellaneous torus attached piping (construction drains and cantilevered lines)
Results The torus attached piping has been evaluated for the peak suppression pool temperature of 195 F and concurrent Mark I containment hydrodynamic loads.
Evaluations concluded that all pipe stresses, pipe supports and torus penetrations are in compliance with the requirements of NUREG-0661 and the ASME Code.
Page 2 - 2
Response to USNRC Request For Additional Information March 4,1997 This evaluation is conservative since Mark I containment hydrodynamic loads will
.enninate before the design peak suppression pool temperature of 195 F is reached. Termination of DBA Mark I containment hydrodynamic loads will occur
'at a suppression pool temperature of about 134 F for the design basis loss of j
coolant accident.
The detailed the torus attached piping calculations are currently being transferred l
from the contractor to the site. They will be available on site for inspection.
b i
a i
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Response to USNRC Request For Additional Information March 4,1997
-l SUPPLEMENTAL INFORMATION FOR LICENSE AMENDMENT REOUEST DATED JANUARY 23.1997 -
EVALUATION OF RHR ROOM TEMPERATURE DURING DBA LOCA Maximum RHR room temperature under long term design basis accident loss of coolant -
accident (LOCA) conditions was determined in Calculations CA-96-113 and CA-97-074.
The results of these calculations are summarized below.
Acceptance Criteria Maximum long term ambient temperature in the Residual Heat Removal (RHR) pump rooms is specified as 140 F in Section 6.2.2.2.1 of the Updated Safety Analysis Report (USAR).
i Analysis Assumptions Calculations were performed for two scenarios:
Case 1 - One RHR pump and one RHR service water (RHRSW) pump are running in suppression pool cooling mode. One Core Spray pump is running for reactor makeup. Suppression pool temperature is maximized at 191 F. This is the peak long term suppression pool temperature for this pump combination at a future maximum anticipated operating power level of 1880 MWT.
Case 2 - Two RHR pumps and two RHRSW pumps are running in suppression pool cooling mode. One Core Spray pump is running for reactor makeup. Heat added to the RHR room from motor losses is maximized. Peak suppression pool temperature is 182 F for this combination of RHR and RHRSW pumps as reported in NEDO-30485,"Monticello Design Basis Accident Containment Pressure and Temperature Response for FSAR Update," December,1983.
Summer time steady state room temperatures were assumed before the LOCA.
At the time the LOCA is initiated, loss of all normal air conditioning and ventilation is assumed.
Page 3 - 1
Response to USNRC Request For Additional Information March 4,1997 Service water (Mississippi River) temperature was assumed to be at the maximum analyzed value of 90"F for the entire period. Ground water temperature was determined to be 48 F for purposes of this analysis.
All sources of heat input to the room were modeled. Significant sources included:
Core Spray and RHR piping RHR Heat Exchanger Electrical cables, lighting, and misc electrical loads e
Core Spray and RHR pump motors Concrete and steel heat sinks were conservatively modeled.
Heat removal by room coolers and heat transfer to the building structure and adjacent soil and groundwater was conservatively modeled.
While both A and B pump rooms are similar, RHR pump room B was chosen for modeling. Because of the layout of the plant, RHR pump room B requires longer pipe runs for cooling water. Tests have shown that the A pump room receives approximately 30% more cooling water flow.
Methodology A lumped parameter model was used to calculate the average bulk air temperature in the RHR pump room following initiation of the LOCA and loss of normal ventilation and air conditioning equipment. If possible, the computation was carried on long enough until room temperature peaks and begins trending down with time. Equations were formulated on a Lotus 123 spreadsheet.
Maximum service water temperature was assumed for piping temperature upstream of heat exchangers. Classic heat exchanger formulas were used for temperature of piping downstream of heat exchangers.
Convective surface to air and radiation surface to surface heat transfer was modeled for all significant heat transfer surfaces. For conservatism, all nominal convective heat transfer coefficients were reduced by 25%. The extra dampening capacity of steel and reinforcing bar buried in concrete walls was conservatively ignored.
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a Response to USNRC Request For AdditionalInformation j
March 4,1997 Results The maximum RHR pump room temperatures during a design basis LOCA were determined to be-Case.
' Assumptions).
- Maximum RHR 1
' ~
Pu'mp Room..
i Temperature ( F)'
1 Suppression Pool 130.55 Temperature Maximized 4
2 Heat Added to RHR Pump 139.84*
j Room From Motor Losses i
Maximized I
6
- RHR room temperature at 10 seconds _(about 11.5 days) after the beginning of the accident - this is the maximum time for which suppression pool temperature is plotted in NEDO-30485. Two RHR pumps would no longer be running at this point in the accident.
Details of the RHR pump room heatup calculations are available on site for inspection.
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4-Response to USNRC Request For Additional Information l
March 4,1997
?
l SUPPLEMENTAL INFORMATION FOR LICENSE AMENDMENT REOUEST DATED JANUARY 23.1997 -
4 EVALUATION OF NUCLEAR CONTAINMENT INSULATION SYSTEM In 1981 Monticello began replacing the originalinsulation inside the drywell with Nukon i,
blankets under Modification 81M058. The safety evaluation supporting this modification cited Owens-Corning Fiberglas Corporation Topical Report OCF-1 as providing the basis for acceptability of Nukon insulation for boiling water reactor applications.
j OCF-1 was submitted to the NRC for review and approval. NRC completed a detailed review of OCF-1 and accepted it for referencing in licensing applications. Refer to NRC letter dated December 8,1978, from Mr Robert L Baer, Program Manager, Light Water Reactors Branch No. 2, Office of NRR, " Final Staff Evaluation of Topical Report OCF-1, Nuclear Containment Insulation System," to Mr Gordon Pinsky, Owens-Corning Fiberglas Corporation.
l In 1992, Nukon insulation was selected as replacement insulation for reactor building closed cooling water (RBCCW) system piping inside the drywell under Modification j
92Q020. It was determined that Nukon insulation remained the most viable product available to withstand a post accident environment inside containment.. As part of this 1
modification, an updated evaluation of the suitability of Nukon was performed by the 4
supplier, Performance Contracting,Inc. This evaluation, Calculation No. NSP-ECCS-1, Rev.1, "ECCS Analysis for the Monticello Nuclear Plant With the Nukon Insulation System," is attached.
The calculations presented here are being superseded by actions taken in response to NRC Bulletin 96-03," Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water-Reactors." New, larger, suction. strainers are planned for installation during the next refueling outage.
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