ML20116N864
| ML20116N864 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 01/28/1985 |
| From: | STONE & WEBSTER ENGINEERING CORP. |
| To: | |
| Shared Package | |
| ML20116N845 | List: |
| References | |
| 14799.02, NUDOCS 8505070381 | |
| Download: ML20116N864 (66) | |
Text
. _ - .
ENCLOSURE 2 J.O.No. 14799.02 NORTH ANNA UNITS 1 AND 2 CORE POWER UPRATING STONE & WEBSTER / BOP LICENSING
SUMMARY
i F
i STONE & WEBSTER ENGINEERING CORPORATION JANUARY 1985 hg k b38' P PM .- ,
.. . . .. . . . _ . ~
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i NORTH ANNA UNITS 1 AND 2 CORE POWER UPRATING TABII 0F CONTENTS Section Title P,a,ge
, 1.0 Objective 1 2.0 Conclusions 2 3.0 Accident Analyses 3 4.0 Environmental Qualification 40 5.0 Transient Analyses 44 6.0 BOP Systems Review 45 7.0 BOP Components Review . 63 8.0 Structures 65 9.0 Radiation Protection 66 i
10.0 Review of Technical Specifications 67 i
1 i
B1-1479902-6263 January 28, 1985
i-NORTH ANNA UNITS 1 AND 2 CORE POWER UPRATING LIST OF TABLES Table Title 3.2-1 Containment Analysis Results 3.2-2 Accident Chronology 3.2-3 Mass and Energy Releases to Containment Pump Suction DER with Failure of a Diesel Generator 3.2-4 Energy Distribution Pump Suction DER with Failure of Diesel Generator
, 3.2-5 Recirculation Spray and Low Head Safety Injection Pumps NPSHA i 3.2-6 Mass and Energy Releases Pump Suction-DER-Min EST Limiting Case for LHSI Pump NPSHA 3.2-7 Mass and Energy Releases Hot Leg DER-Normal ESF. Limiting Case for RS Pump NPSH 6.2-1 . Design Conditions, Extraction Steam-6.2-2 Design Conditions, Feedwater Heaters 3 6.2-3 Component Cooling System Heat Loads 4
4 1
B1-1479902-6263 January 28,19S5
- . . .. __- . . = . . ~ _ . . - . _ . - . - - - -
NORTH ANNA UNITS 1 AND 2 CORE POWER UPRATING a
LIST OF FIGURES Fiaure- Title 3.2-1 Containment Pressure Vs. Time Worst Case Peak Pressure 3.2-2 Steam Condensing Coefficient vs. Time Pump Suction DER Minimum ESF 4
3.2-3 Containment Temperature Worst Case Peak Pressure 3.2-4 Pressure Transient Worst Case Depressurization Min.
l ESF PSDER 3.2-5 Condensing Coefficient Worst Case Depressurization 3.2-6 Quench Spray Flow Rate Worst Case Depressurization 3.2-7 Temperature Transient Worst Case Depressurization 3.2-8 Recirculation Spray Cooler Duty Worst Case Depressurization 3.2-9 NPSH Transients Worst Case LHSI Pump NPSH 3.2-10 Pressure Transients Worst Case LHSI Pump NPSH 3.2-11 Temperature Transients Worst Case LHSI Pump NPSH 3.2-12 Recirculation Spray Cooler Duty Worst Case LHSI Pump NPSH 3.2-13 Quench Spray Flow Rate Worst Case LHSI Pump NPSH 3.2-14 Condensing Coefficient Worst Case LHSI Pump NPSH a
3.2-15 NPSHA Transients, Worst Case RS Pump NPSH 3.2-16 Press. Trans. Worst Case RS Pump NPSH i
3.2-17 Temp. Trans. Worst case RS Pump NPSH 3.2'18 Recire. Spray Cooler Duty Worst Case RS Pump NPSH 3.2-19 Quench Spray Flow Rate Worst Case RS Pump NPSH 3.2-20 St. Condensing Coefficient Worst Case RS Pump NPSH
- 6.2-1 Heat Balance - Current Conditions 6.2 Heat Balance - Bounding Conditions for Uprate i
6.2-3 Heat Balance - Normal Operation for Uprate 1
81-1479902-6263 January 28, 1985 y . -
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I 1.0 OBJECTIVE The objective of this report is to provide a technical basis for
' determining that the proposed core power uprating does not involve an unreviewed safety question in accordance with the requirements of 10CFR50.59. The review summarized here was limited to systems within i Stone & Webster Engineering Corporation's original scope of work.
The evaluation used the following parameters, except where noted in the text, which bound the proposed uprated conditions:
5 Main Steam Pressure 1007. Power 880 psia Main Steam Temp. NoLoad 547*F
! Main Steam Pressure NoLoad 1020 psia RCS T 586.8'F ava 6 Steam Flow 10 lb/hr Total 13.15 Reactor Power MWt 2898.
4 NSSS Power MWt 2910.
This review encompassed all systems and previous analyses potentially affected by a core uprating. Systems and analyses were reviewed to verify continued compliance to the licensing criteria and standards currently applicable to North Anna Power Station.
The review was performed to verify that plant systems continue, af ter the uprating, to meet the functional requirements specified in the UFSAR.
B1-1479902-(263 January 28, 1985
J 2
2.0 CONCLUSION
S The preposed change in reactor core power has been reviewed and evalu-ated with respect to the following:
I
- 1. Accident Analyses
- 2. Environmental Qualification
- 3. Transient Analyses
- 4. BOP Systems Review
- 5. BOP Components Review
- 6. Structures
- 7. Radiation Protection
- 8. Review of Technical Specifications The station will continue to meet the existing design and performance objectives and the system functional requirements.
Based on the results of this review, it has been concluded that the proposed core uprating does not represent an unreviewed safety ques-tion as defined in 10CFR50.59. These conclusions can be summarized as follows:
- 1. It has been determined that the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the UFSAR has not been increased.
- 2. The possibility for an accident or malfunction of a different type than any evaluated previously in the UFSAR has not been created.
- 3. The margin of safety as defined in the basis for any Technical Specification has not been reduced. >
f B1-1479902-6263 January 28, 1985
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3.0 ACCIDENT ANALYSES I
3.1 Accidents From UFSAR I
+
The UFSAR has been reviewed and accidents and analyses that were evaluated are listed with their UFSAR section below:
i 1.
Containment Loss of Coolant Accident (LOCA), Section 6.2.1.3.1.1
- 2. Containment Main Steam Line Break (P.SLB) , Section 6.2.1.3.1.2
- 3. Containment Inadvertent CDA, Section 6.2.6.3
- 4. Volume Control Tank (VCT) Rupture, Section 15.3.6 ,
- 5. Waste Gas Decay Tank (WGDT) Rupture, Section 15.3.5 I '6. LOCA Doses, Section 15.4.1.7
! 7. MSLB Doses, Section 15.4.2.1.3
- 8. Fuel-Handling Accident Outside Containment, Section 15.4.5
- 9. Steam Generator Tube Rupture, Section 15.4.3
- 10. Containment Subcompartment Pressure Analysis, Section 6.2.1.3.2 I
From the above listed analyses, only the containment LOCA is affected.
3.2 Revised Accident Analyses
! 3.2.1 Containment Loss of Coolant Accident (LOCa, i
i j
The current containment LOCA Analysis was performed for a NSSS power level of 2910 MWt. The containment LOCA Analysis was redone for an NSSS power level of 2910 MWt + 2 percent to account .
for uncertainty in core thermal power. Although Westinghouse has I
confirmed that the current initial RCS mass and energy release rates remain applicable for the 2 percent increase in power, the long term blowdown, as calculated by LOCTIC, takes into account a. '
2 percent increase in decay heat. The analysis was performed in accordance with the methods described in 11FSAR 6.2.1.3.1.1. The results showed that the increased power level has a negligible effect on the peak calculated containment pressure for the acci-dent. The depressurization time is lengthened somewhat . at the B1-1479902-6263 January 28, 1985 4
4 y - wn - . - - -,-,
4 4
proposed power level but is still within the design basis of one hour. The subatmospheric peak pressure remains below zero at the
- proposed power level. These results are compared to the current license results in Table 3.2-1.
Reanalysis of available Net Positive Suction Head (NPSH) for the 1 recirculation spray pumps and low head safety injection pumps shows that the available NPSH is unaffected by the proposed power uprate.
Additional containment analysis input data and calculation [
results are shown in Tables 3.2-3 through 3.2-7 and Figures 3.2-2 through 3.2-20.
1 3.2.2 Containment MSLB Analysis Equipment qualification inside the containment is based on the Main Steam Line Break (MSLB) and LOCA and on the containment design pressure. The current Main Steam Line Break analysis and containment design pressure are unchanged for the uprate condi-I tions.
I
- The basis of the steam line break analysis is the full guillotine j main steam line break at the no-load (hot shutdown) condition.
The no-load Reactor Coolant System and Steam Generator tempera-ture and pressure remain unchanged subsequent to the uprate.
l Although there are changes in energy releases at uprated power i operating conditions, the no-load condition remains the limiting case. In addition, the current analysis assumes dry steam is released from the break. Therefore, the Main Steam Line Break post-accident conditions remain as previously analysed.
4 A
t B1-1479902-6263 January 26, 1985
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-3.2.3 Containment Inadvertent Containment Depressurization Actuation (CDA) Minimum Credible Pressure There are no changes required for the CDA system due to the uprate and the results of the accident do not change. The limiting technical specification conditions remain unchanged.
The spray system input parameters remain unchanged.
3.2.4 Volume Control Tank (VCT) Rupture The accident doses currently reported in WSAR Section 15.3.6 are for a core power of 2900 Wt, which bounds the proposed uprated core power of 2898 Wt. These accident doses are based on source terms (reference WSAR Section 15.1.7), which Virginia Power has '
determined to be applicable for the core uprating up to a batch I
average fuel burnup of 45,000 WD/!frU.
3.2.5 Waste Cas Decay Tank The accident doses currently reported in WSARQction 15.3.5 are for a core power of 2900 Wt, which bounds the proposed uprated core power of 2898 W t. These accident doses are based on source terms (reference WSAR Section 15.1.7) which Virginia Power has determined to be applicable for the core uprating up to a batch average fuel burnup of 45,000 MWD /MTU.
l 3.2.6 LOCA Doses The accident doses currently reported in WSAR Section 15.4.1.7 are for a core power of 2900 Wt, which bounds the proposed uprated core power of 2898 Wt. These accident doses are based
, on source terms (reference WSAR Section 15.1.7) which Virginia Power has determined to be applicable for the core uprating up to a batch average fuel burnup of 45,000 MWD /MTU.
l 51-1479902-6263 January 28, 1985 i
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$4 3.2.7 MSLB Doses' l l
The accident doses currently reported in UFSAR Section 15.4.2.1.3 are for a core power .of 2900 Wt, which bounds the proposed uprated core power of 2898 Wt. These accident doses are based on source terms (reference. UFSAR Section 15.1.7) which Virginia
- j. Power has determined to be applicable for the core uprating up to a batch average fuel burnup of 45,000 MWD /MTU.
1 F
3.2.8 Fuel Handling Accident 1
The accident doses currently reported in UFSAR Section 15.4.5 are for a core power of 2900 MWt, which bounds the proposed uprated
- core power of 2898 Wt. These accident doses are based on source-terms (reference UFSAR Section 15.1.7) which Virginia Power has determined to be applicable for the core uprating up to a batch j average fuel burnup of 45,000 WD/MTU.
t 3.2.9 Steam Generator Tube Rupture The accident dose's currently _ reported in UFSAR Section 15.4.3 are
- for a core power of 2900 Wt, which bounds the proposed uprated core power of 2898 W t. These accident doses are based on source i
terms - (reference UFSAR Section 15.1.7) which Virginia Power has -
determined to be applicable for the - core uprating up to a batch average fuel burnup of 45,000 MWD /MTU.
j 3.2.10 Containment Subcompartment Pressure Analysis l The subcompartment pressure analyses for the reactor cavity,
! steam generator compartment, and -pressurizer cubicle are pre-sented in the UFSAR. A review of these analyses indicates that l
i they are applicable to the proposed core upratie as they were j originally performed .t the Engineered Safeguards Design Rating (ESDR) - 2900 Wt.
i B1-1479902-6263 January 28, 1985 4
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3.3 ' Accident Analyses Conclusions The review encompassed all UFSAR analyses potentially affected by a core uprating.
Only the containment LOCA analyses. were redone and the results were within the applicable licensing and design criteria.
4 l-f' i
T B1-1479902-6263 January 28, 1985 1
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8 TABLE 3.2-1 CONTAINMENT ANALYSIS RESULTS 2900 MWt Design Worst Case Uprate Limit Containment Peak Pressure (psig) 40.94 45.0 Time of Peak Pressure (sec) 302 -
Containment Depressurization Time (sec) 3400 3600
~
Containment Subatmospheric Peak (psig) .01 0 Containment' Atmosphere Peak Temperature 268 -
4 Notes: 1. Initial containment pressure and temperature are 12.2 psia and 105*F, respectively. '
- 2. Peak containment pressure and peak containment temperature occur concurrently.
3 d
J j B14/41V January 28, 1985 A e
9 TABLE 3.2-2 ACCIDENT CHRONOLOGY PUMP SUCTION DER WITH FAILURE OF A DIESEL GENERATOR (LIMITING CASE FOR MAXIMUM POST-LOCA CONTAINMENT PRESSURE)
Time (sec) Event 0.0 Accident occurs.
3.0 Containment depressurization actuation signal (10 psig).
15.5 First containment peak pressure occurs.
25.0 End of blowdown; core reflooding begins.
30.0 Safety injection pumps become effective.
37.6 Accumulators empty.
63.5 Quench spray subsystem and casing cooling become effective.
204.0 Core reflooding ends; post-reflood frothing begins.
302 Second peak containment pressure occurs; recirculation j spray system becomes effective.
1283. Post-reflood frothing ends.
3160. Containment pressure becomes su'o atmospheric.
3580. Safety injection pumps switch to recirculation mode.
l B14/41V ' January 28, 1985
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F Il
12 TABLE 3.2-4' ENERGY DISTRIBUTION PUMP SUCTION DER WITH FAILURE OF DIESEL GENERATOR (LIMITING CASE FOR MAXIMUM POST-LOCA CONTAINMENT PRESSLTE IN UNITS OF MILLIONS OF BTUs)
Time'(sec)*
. 0.0 25.0 204.0 1300.0 3610.0 Neat sources Primary coolant 216.06 35.15 13.09 - -
Primary hot- metal 24.71 24.01 23.41 - -
Piping, pumps, valves 61.95 60.07 54.02 - -
i reactor vessel, and internals Pressurizer metal and 13.76 12.98 12.91 - -
- lines i Steam generator metal 117.56 117.11 105.10 - -
Steam generator secondary 155.30 160.38 116.68 - -
. water 4
Pressurizer water 27.47 0.0 0.0 - -
- Core sensible heat 26.31 7.92 4.63 - -
Accumulator contents 8.89 4.20 0.01 - -
a External water tanks 87.81 87.81 84.90 77.58 6i.18 a
(RWST and CCT) i Neat sinks Containment atmosphere 6.27 175.7 188.36 72.27 9.86 j water The significance of the chosen times is as follows:
0.0 - Accident occurs 25.0 - End of blowdown 204.0 - End of core reflooding 1300.0 - End of post-reflood frothing (approximate) 3610.0 - One hour af ter accident (approximate) - containment has depres-surized and is subatmospheric B14/41V January 28, 1985 4
w~ - .- -
1 13 TABLE 3.2-4 (Cont'd)
ENERGY DISTRIBUTION PUMP SUCTION DER WITH FAILURE OF DIESEL GENERATOR (LIMITING CASE FOR MAXIMUM POST-LOCA CONTAINMENT PRESSURE IN. UNITS OF MILLIONS OF BTUs)
Time (sec)*
0.0 25.0 204.0 1300.0 3610.0 Heat sinks (Cont'd)
Containment atmosphere 1.22 3.82 4.04 2.96' 1.56 air Containment floor water b 0.0 41.64 87.33 306.75 430.96 Concrete sinks 0.0 6.00- 25.50 69.80 75.31
- Liner and metal sinks 0.0 18.10' 56.72 57.30 35.75 Heat inputs" Delayed fissions 0.0 3.50 3.50 - -
Decay heat 0.0 4.25 26.18 - -
. Pump heat 0.0 0.0 0.15 0.69 1.93 j Input blowdown - -
0.0 141.10 291.42 Input spillage - -
0.0 92.02 120.63 i
Heat outputs Safety injection sump - -
0.0 0.0 3.47 i suction Recirculation coolers 0.0 0.0 0.0 93.84 242.57 The significance of the chosen times is as follows: ,
4 0.0 - Accident occurs 25.0 - End of blowdown 204.0 - End of core reflooding 1300.0 - End of post-reflood frothing (approximate) 3610.0 - One hour after accident (approximate) - containment has depres-i surized and is subatmospheric The energy store in the water on the containment floor is calculated as a function of time af ter an accident by the LOCTIC computer program. The water on the containment floor is treated as a separate node, apart from
, the containment atmosphere and apart from the concrete floor of the contain-ment structure. The mass and energy inventories are determined by performing mass and energy balances on the node for each-time step of the calculation.
B14/41V January 28, 1985
I 4 . .- 14 TABLE 3.2-4 (Cont'd)
ENERGY' DISTRIBUTION PUMP SUCTION DER WITH FAILURE OF DIESEL GENERATOR (LIMITING CASE FOR MAXIMUM POST-LOCA CONTAINMENT PRESSURE I
IN UNITS OF MILLIONS OF BTUs)
C After the end of core reflooding, k'estinghouse froth data are used for mass and energy release calculations. Heat sources and inputs associated with the primary coolant system are not individually accounted for after this time in the LOCTIC calculation. Energy removed from these sources after the end of core reflooding is included in " Input Blowdown" and " Input Spillage."
i 9
1 1
i
[
B14/41V Janua ry 28, 1985
15
- , TABLE 3.2-5 y RECIRCULATION SPRAY AND LOW HEAD SAFETY INJECTION PUMPS NPSHA
y Initial Water Minimus NPSHA, ft-
'..\'
Worst Temp of (at time, sec)
Case Break ESF RWST Service IRS ORS LHSI NPSHR*
- IRS ELLER Norm 30 35 11.6 17.0 25.3 9.4 ORS (550) (630) (1910)-11.0 LHSI .PSDER Min. 50 95 13.9 21.5 15.9 13.4 (1810) (1990) (2980)
- At design flow b
.}
l B14/41V 'Janua ry 28, 1985
Mass and Energy Release Pump Suction DER-Min-ESF Limiting Case for I.llSI pump NPSH.
HASS Ate EHERGY RELEASE RATES. SPILLAGE RATES. Ate INTECRATED RELEASES AfD SPILLAGE s
HDTEt RATE DATA IS CONSTANT OVER THE TillE INTERVAt. Are 15 EQUAL TO THE CHANGE IN THE INTEGRATED SATA DVER THE TIllE INTERVAL DIVIDED BY THE DistATIOH OF Tite INTERVAI. -
......................----...-RATE DATA.- ""IHTEGRATED DATA------------
.611HE INIERVAL.-. --...--CLEK 50084....---- --------.----------------------"""_"""BLOHD0184--------
SPILLAGE-------. . --TillE.- --
..-SP IL L AGE ----- --
..STAPT-- ---Ele-- --.HASS--- ..EHERGY.- .---HASS-- --EllERGY- . ---II45$-.- --El8ERGY-- ---IIASS -. --Er8ERGY-.
E SEc l ISECD ELSH/SECB (STtuSECD EL8H/SECD IBTU/SEC) . (SEC) ILDHS tbTUI ELLHB (DTut 8.9 9.1 S.1212E+04 4.7944Ee87 8.8 0.0 . 8.1 4.1212E+03 4.7044Ee64 0.0 8.0 0.1 1.0 3.4815Ee84 2.0439E*07 e.8 0.8 . 1.0 4.0535Ee04 2.3459Ee07 e.S e.9 1.8 2.0 3.5027Eeet 2.0505Ee87 4.8 0.0 . 2.0 1.5551Ee04 4.3744EeO7 0.0 0.0 2.s 3.0 3.5101EeD4 2.8741E*e7 8.8 0.0 . 3.0 1.1044E605 4.4705E*e7 0.0 0.0 3.0 4.0 3.4440EeD4 2.8723E*07 4.4 0.0 . - 4.0 1.4550Ee05 8.5429Ee87 , 0.0- 0.0 4.8 5.0 3.4359Ee84 '2.4517E,07 8.0 0.0 I 5.0 1.79*4Ee05 1.0595Ee06 e.0 3.0 5.0 4.0 3.3490E+04 2.0175Eee? 8.9 4.4 ' . 4.0 2.1355Ee05 1.2412Ee08 8.8 0.0 4.0 7.0 3.2447Ee64 ' 1.9417E*07 8.8 0.0 . 7.0 -2.4424Ee05 1.4574Ee04 0.0 0.0 7.9 - 4. 8 3.1479Ee04 1.4847Ee07 0.0 0.0 . 8.0' 2.7772Ee05 1.4458Ee08 0.0 0.0 4.4 9.8 3.0439Ee64 1.0140Ee07 4.8 0.8 . '9.0 3.0815Ee05 1.8272Ee04 0.0 0.0 9.9- 18.8 '2.9545Ee04 1.7454Ee07 8.0 0.0 10.0 3.3744Ee05 2.8018Ee88 0.0 0.0 10.0 11.0 1.4334Ee84 1.2542Ee07 0.4 e.e . 11.0 3.5599Ee05 2.1274Ee04 0.0 0.0 g 11.8 12.0 4.4535E,03 7.0552Ee04 0.0 '0.0 . 12.8 3.4245EeOS 2.1980Ecos 0.0 0.0 g 12.0 13.0 4.4814Ee03 5.3437Ee04 ' O.4 8.0 . 13.0 3.4713E*05 2.2515Ee08 8.0 . 0.0 e 13.0 14.0 3.2954E+03 3.9451E+04 0.0 0.0 , 14.8 3.7043Ee05 2.2913Ee06 0.0 0.0 tis y L.s 14.8 15.0 2.1552E+03 2.5942E*04 4.0 0.8 . 15.8 3.7258Ee05 2.3172Ee08 0.0 0.0
- 15.8 14.0 1.414SEe03 1.7024Ee04 8.8 0.0 . 14.0 3.7400E*05 2.3342Ee04 4.8 0.0 Y 14.5 17.0 9.3437Ee82 1.1238Ee04 0.0 0.8 . 17.0 3.7493Ee05 2.3455Ee06 e.0 0.0 cb 17.0 18.0 : 4.4442E+02 7.9507Eee5 0.0 0.0- . 18.4 3.754CIe05 2.3534Eo08 0.0 0.0 18.0 19.8 5.1012Ee02 4.0419Ee05 S.8 0.0 . 19.0 3.7411E*05 2.3595Ee08 0.0 0.0
'19.8 28.8 4.8525Ee02 9.4154Ee95 0.0 0.0 20.0 3.7451Ee05 2.3443Esoe 0.0 0.0 28.8 21.8 3.353?Ee02 3.9738E*05 0.0 0.4 . 21.0 3.7405E*05 2.3483Ee06 0.0 0.0 21.0 22.0 2.8541Eeet 3.3794Ee05 0.0 0.0 . 22.0 3.7713Ee05 2.3717Ee08 0.0 0.0 22.0 23.0 2.5412E+02 3.8221Ee05 , 0. 8 8.0 . 23.0 3.7739Ee05 2.3747Ee03 0.0 0.0 23.0 24.0 2.3294Ee02 2.7438Ee05 8.0 0.0 . 24.0 3. 7742E e 05 2.3774Ee08 0.0 0.0 24.8 '25.0 2.1325Ee02 2.5894Ee05 S.0 0.0 25.0 3.7783Ee05 2.3800Eee8 e.0 0.e
'25.0 24.0 1.2392Ee43 4.0904Ee05 8.0 ' 4.8 . 24.0 3.7907Ee05 2.3840Ee06 0.0 0.0
~
24.0 27.8 '4.8411Ee03 1.3919Ee04 0.9 4.0 . 27.0 3.8394Ee05 2.3980Ee06 0.0 0.0 27.0 28.4 't.3434Ee83 9.4424Ee05 S.8 0.e. . 28.0 3.8530Ee95. 2.4078Ee06 0.0 0.0 24.0 29.0 2.4412Ee03 5.7275Ee05 ' 8.8 0.8 . 29.0 3.8774Ee05 2.4135Ee08 0.0 0.0 29.4 .30.0 1.5533Ee03 4.9896Ee05' O.0 0.0 . 30.0 3.8931E*05 2.4185Ee06 0.0 0.0
.30.0 35.0 4.34S4Ee03 8.1108Ee05 S.e 0.0 . 35.0 4.1104Ee05 2.459 tee 08 0.0 0.0 35.0 40.0 3.1493Ee03 7.e393Ee05 0.0 0.0 . 40.0 4.240 lee 05 2.4963Es00 0.0 0.0 40.0 45.8 4.9302Ee02 4.3942Ee05 S.0 0.0 . 45.0 '4.2927Ee05 2.5305Ee08 0.0 0.0 45.0 '50.0 5.3819Ee02 4.4792Ee05 0.8 0.0 . 50.0 4.3197Ee05 2.542 7E e 08 0.0 0.0 O
17 TABLE 3.2-6 (cont.)
1 l
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8 W>
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g 4
a= s' g N3 ii e af M e
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P e in c .g4--,e ede ' is oe==
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me W e=
4 tap 0 == l i
9 9 W W IA &eMMM NNNMM e e M se M eeee
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seeem NNNNN
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4 ee43 'O e i eft # 4 8
r* e in N eOMt ut e == ae as Nmema sein e m in e m e in 9m ed er e e p's es sw ed emm-,
a es-- os .
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s 4 e.#.US
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g 0 9 ansm It se 60 '. =m.
Ins e .= 8 eZt 6
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TABLE 3.2-7 1
4 h
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- o* =c c.a o o. o. o. ca = =c*o
> - @ == h tag w taa and a.s las taJ ted isJ taa les bJ 68 lad a.J tad W ens taa had has and W ing taJ W 4.4 ans :.ad tas W ina taa e.d 4 i C. 3 ii Pb F M ea3 F N O g A P= O W is qs* F F f'3 M ed tJ C4 N M se o ed me .1 ?" 4: O Q "* F* *= Fa i
R N N o N L* m F f* N M a'sP* N M M M M & ed . C wl M l
g U O P* e4 o me F t J. O e e U . , ,S= =U f
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- e. EE.2 * **
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= * * *
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= zM u
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wg - 't g e ==. m. o. to. e. e. e. N. m. e. ..... e. e. e. e. e. e. e. u. e. o. e. ce. cJ. M. M. e. w. e. r=. r*.
M 3 8 0 8 e se in ce et ce WNMMM. MMMMM MMMMM MMMM& WW#g7 FFgFF g -
s= 2" e s Q t 0 O es e-o e
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a a 4 ew - e e4NMe in e P* e e e-NMe m .,=e r* e e o - eJ M e in e P* e e eoemo De 8- U e s=s eme as amo e=e ce u ame sue s=0 N N N tJ tJ to fd N N u M M F F in f,"J* g 0 se - tad i vs B 4= M e 0 8 - e T g WW G t 8
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EEI O tad e i'd 4= 8 eseee eeOeO < eOeee eeoee OOOOe OOOOO OOOOe ena t= I!4 i em =J i. 8 m.m .
4 m et e .Ja 0 9 k *> t me 6 -9 g Q ' g C; - 9 C. 8 ** I
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4.0 ENVIRONMENTAL QUALITICATION (EQ) !
i 4.I' General l l
l~
The EQ limits for North Anna are presented in the Station Environ-eental Zone Descriptions (EZD) for the following environments: High l
l Energy Line Break (MELB), normal, LOCA and MSLB. A review has been l performed to assess the impact on the EZD and associated EQ limits for j the core uprating. l 4.2 EQ Review , ;
4.2.3 LOCA & MSLB Environmentj ;
I The limiting equipment qualification conditions inside the con- -
t talement are based on the Main Steam Line Break, the LOCA and the '
containment design pressure. The current Maia steam Line Break analysis and containment design pressure are unchanged for the uprate conditions. ;
9 i The containment LOCA analyses were redone, but did not effect the ;
envelope in the EZD.
The basis of the stese line break analysis is the full guillotine main steam line break at the no-load (hot shutdown) condition. .
The so-lead Reactor Coolant System and Steae Generator tempera-ture and pressure reesta unchanged subsequest to the uprate.
Although ease and energy releases change at power operating conditions due to decreased (uprated) Steam Generator pressure j and temperature, and increased core power, the me load condition reesias the limiting case. Therefore, the Nata Steam Line Break l
post accident conditions reesta as previously analysed, t
i 81 1479902 6263 January 28, 1983 1
l 61 As stated in Section 9.0, LOCA & MSLB radiation limits have already been based on a bounding core power of 2900 MW. There-fore, the radiation EQ limits are unaffected by the core uprate.
Chemical spray environments within the containment are based on the volumes and concentrations of boron and NA0H in the reactor coolant system, refueling water storage tank, boton injection tank, accumulators, chemical addition tank, casing cooling tank, etc. These volumes and concentrations remain unchanged for the core uprate confirming that the existing chemical spray EQ limits remain valid.
4.2.2 Normal Environments Normal environments both inside and outside the containment remain unaffected by the uprate. Technical Specifications for containment pressure and temperature will not change for the uprate. As discussed in Section 9.0, radiation limits both inside and outside tt.. containment have already been based on a bounding core power of 2900 MWt, and temperatures of systems outside the containment have not changed significantly, there-fore, the temperature ar.d pressure limits outside the containment will also remain valid.
4.2.3 HELB Environments Post-accident environments outside containment which are used to generate equipment qualification envelopes are based on the following high energy line breaks.
4.2.3.1 Primary Systen Branch Line treek The Letdown Line treak forms part of the basis for the environ-mental qualification in the charging pump cubicle in the aux-iliary building. Tor the uprate, the letdown line temperature 31-1479902 6263 ,
January 28, 1985
. -. . = - -. . .. -
.+ . :.2 l- ,
from the regenerative heat exchanger is bounde1 by the current analysis.
1 t
I . . .
4.2.3.2 Secondary System Break The Main Steam Line Break affects the Main Steam Valve House, the l Service Water Valve Pit and the Turbine Building. The Main Steam Valve House environmental envelope is based on no-load power l
condition which is unaffected by the uprate. Equipment quali-
~
fication temperature and pressure in the Service Water Valve Pit and Turbine Building is limited by the Turbine Building siding pressure-retaining capability. Any changes to break effluent duc l
l to the uprate has no ef fect on this pressure or temperature, and l therefore, on equipment qualification.
l l 4.2.3.3 Auxiliary Steam Line Break i
This break affects the Auxiliary Building and the Service Water l The releases are based on the Auxiliary Steam Line Valve Pit.
l I relief valve pressure setting which is unchanged by the uprate.
Additionally. the Auxiliary Steam System pressure is controlled by a pressure reduction valve tied into the Main Steam header.
The decreased Main Steam operating pressure will not affect the >
t i operation of this valve, and therefore, the Auxiliary Steam l System pressure will remain unchanged.
l
(' 4.2.3.4 Steam Generator Blowdown Line Break This affects the Pipe Tunnel and the Auxiliary Building. The releases are based on the bounding condition of no-load steam l generator pressure which is unchanged by the uprate.
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, 1 81-1479902-6263 January 2$. '.+
_ _ =
43 4.3 EQ Conclusions The EQ limits as presented in the station EZD remain bounding for the core uprating.
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l l 81-1479902-6263 1 January 28,-1985 k
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r 44 5.0. TRANSIENT ANALYSES i
Transient analyses have been performed for the uprated condition and have verified the following UFSAR performance objectives:
4
- 1) On 50 percent load rejection, there will be no reactor trip or atmospheric dump through the main steam safety valves.
- 2) On turbine trip with a reactor trip, there will be no atmospheric
, dump through the main steam safety valves.
These analyses were completed using a NSSS power of 2968 MWt.
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l B1-1479902-6263 January 28, 1985 I
_ . - - . ~ . _ . _ _ ._ . .__ _ _ . . ._ . . __ _ _ . .
4 Y
45 6.0 DOP SYSTEMS REVIEW i
8 6.1 General 1
a t
The objectives of the review of Balance of Plant (BOP) systems were to 4
determine the effect of the core uprating on system normal and acci-I dent operations. Also, compliance with the design basis and per-formance criteria were reviewed.
, The- BOP system interfaces were reviewed to determine the effect the
] uprated conditions have on system operation, flow, pressure, tempera-ture, and heat load. The existing system component capacities were c
- , used to determine the effect on the system operation, pressure, and j temperature.
s
); The following systems have been reviewed:
i
- 1. . Main Steam System
- 2. Auxiliary Feedwater System t
- 3. Extraction Steam System, Feedwater Heaters
- 4. Condensate and Feedwater Systems I
- 5. Low Pressure and High Pressure Heater Drain Systes
- 6. Steam Generator Blowdown Systes
- 7. Condensate Polishing System 4
- 8. Component Cooling System I
- 9. Spest Fuel Pool Cooling Systes i
- 10. Containment Air Recirculation System
- 11. Centrol Systems and Instrumentation '
t
- 12. Electrical Systems
- 13. Service Water System 1' ;
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l 31-1479902-6263 Janusry 28, 1985 t' l
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46 i
- 14. Radioactive Waste Systems
- 15. Boron Recovery System
- 17. Bearing Cooling System
- 18. Quench Spray Subsystem
- 19. Recirculation Spray Subsystem
- 20. Auxiliary Steam System 6.2 Results of BOP Systems Review 6.2.1 Main Steam System The Main Steam System piping is designed for 560*F and 1100 psia.
These conditions bound the uprated Main Steam conditions of 547*F/ 1020 psia at no-load and $30*F/880 psia at 100 percent power.
The Main Steam Safety Valves (MSSV) were designed for an NSSS rating of 2910 MWt and the - capability is in accordance with applicable codes.
The MSSV's have a total relieving capacity of 12,826,260 lb/hr.
Based on Heat Balance Figure 6.2-3, the total Main Steam flow rate will be 12,766,627 lb/hr at 2905 MWt.
f The Main Steam Trip and Non-Return Valves were evaluated for rapid closure impact loads applied subsequent to Main Steam System Pipe rupture at uprated conditions. The results of computer runs - that modeled the transient's effect on the valves showed that they would close as required during a Main Steam Pipe break without jeopardizing the integrity of the pressure boundary.
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B1-1479902-6263- ~ January 28,_1985
47 6.2.2 Auxiliarv Feedwater System 1
The Auxiliary Feedwater pumps are designed to deliver rated flow to the Steam Generator at a static discharge head equivalent to the set pressure of the lowest set Main Steam Safety Valve, 1085 psig. Because this set point will not change at the uprated conditions, it is concluded that the resistance parameters asso-ciated with the Auxiliary Feedwater System are unchanged. There-fore, since the . flow requirements are unchanged at the uprated
- conditions, the system is considered adequate.
Westinghouse supplied the Auxiliary Feedwater flow requirements (based on 2910 MWt core power plus 2 percent) for North Anna as follows:
AFWS Actua-tion Signal Time Delay j Transient For AFWS- Flow Rate
- 1) Loss of Feedwater 60 Seconds 680 spa total to (Station Blackout) 2 of 3 S/G's.
- 2) Main Feedline Break 60 Seconds 340 gpa to one intact S/G.
680 gpa to 2 intact S/G after 30 minutes, o
The conclusion- is that the ' existing Auxiliary Feedwater System is adequate at the uprated conditions as the flow requirements and system resistance parameters are unchanged as a result of the uprating.
B1-1479902-6263 January'28,.1985
, ~ _
..m- ,
48 6.2;3
-Extraction Steam System, Feedwater Heaters and Flash Evaporator A heat balance was developed that bounds the proposed Core Uprate conditions. From the bounding Heat Balance (Figure 6.2-2), the values were taken for the temperatures and pressures of the
. Extraction Steam lines. j
! l It was observed that the uprated operating temperatures of the firsts second, and fif th point extraction lines and -pressure of the fifth' point . extraction line on both units were above the design values. The third point extraction temperature is above design for the Core Uprate (347'F), but was previously evaluated i
at a bounding temperature (356*F) for operation at the existing plants parameters. The maximum thermal stress in these extrac-tion lines were reviewed and found to be below the code allowable stress, confirming that all Extraction Steam lines are adequate to operate at the uprated conditions.
It has also been verified that the uprated extraction pressures A
are within the shell side design pressures for all of the Feed-water Heaters and the Flash Evaporator. It was conservatively assumed that the pressure at the turbine extraction nozzle was i
equivalent to the operating pressure at the Feedwater Heater shell without considering pressure drop in the extraction lines.
t Refer to Table 6.2-1 for a summary of design conditions and the proposed operating conditions corresponding to the bounding Core
- - -Uprate conditions for the Units 1 and ' 2 Extraction Steam lines.
Table 6.2.-2 lists a comparison 'of the Feedwater Heater shell
. design pressures and the uprated operating pressures. In both tables, _ absolute pressures were taken from the heat balance and
, converted to gage pressures.
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l i-t B1-1479902-6263
49
> 6.2.4 Condensate and Feedwater Systems To determine if the Condensate and Feedwater System piping would be adequate at the uprated conditions, a comparison was made between the current North Anna heat balance (Figure 6.2-1) and the bounding uprated heat balance .(Figure 6.2-2). It was shown that the entire Condensate /Feedwater piping train uprated temp-erature did not increase significantly from the current condi-tions.
Discharge pressure at the Condensate pump discharge decreases from 486.58 psia to 461.36 psia. Discharge pressure at the Feedwater pump discharge decreased from 1180.3' psia to 1109.93 psia. The decrease in pressure is due to pump / head characterstic at increased flow rates.
It-has been shown that there is an insignificant change in Con-densate and Feedwater System temperature and pressure parameters due to the uprating and these small changes are within the cap-ability of the current system.
The total Condensate and Feedwater System resistance was evalu-ated for the new flow rates and Steam Generator pressure pertain-ing to the Core Uprate. It has been determined from reviewing the existing Condensate /Feedwater System calculation that the existing pumps have sufficient head to overcome the total system resistance with two Condensate and two Steam Generator Feed Pumps in operation at the uprated condition.
i L
The NPSHA at the suction of the Condensate and Feedwater pumps i
, was evaluated at the uprated conditions. It was determined that sufficient NPSHA exists to allow acceptable operation at the uprated flow. -
B1-1479902-6263
l l
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-The main feedwater regulating valves have been reviewed for the uprated conditions and it has been determined that the Unit I valves are acceptable for the core uprating. A valve trim change is required for the Unit 2 valves to provide operational flex-ibility at the uprated condition.
6.2.5 Low Pressure and High Pressure Heater Drain System As with the Condensate and Feedwater systems, the uprated temp-erature and pressure conditions associated with the Low and High Pressure Heater Drain Systems did not change significantly from the current conditions.
The Low and High Pressure Heater Drain Pumps have been shown to be adequate at the uprated flow conditions. Uprated NPSHA has been evaluated at the pump suctions and has been determined to be acceptable for pump operation.
6.2.6 Steam Generator Blowdown System 4
A review of the Steam Generator Blowdown System has indicated that the core uprating will not affect the present safety aspects or operability of the system. Under Core Uprate conditions, the g-steam generator pressure at both no-load and 100 percent power remains equivalent to the current operation pressure. Therefore, the uprate does not affect any piping design pressure limita-tions. -
The design of the excess flow-high energy line break isolation valv.es was for an inlet pressure of 1100 psig which is higher than the lowest Main Steam Safety Valve setpoints and is there-fore acceptable with regard to the uprate.
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l B1-1479902-6263 January 28, 1955 l 1
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i All remaining portions of the Steam Generator Blowdown System including flow control valves, safety valves, tanks and pressure control valves were reviewed for any expected temperature and pressure changes and are unaffected by the uprate.
Additionally, the system is protected from the consequences of excess blowdown flow rates (greater than 100 gpm) and blowdown tank high pressure (greater than 125 psig) by instrumentation that terminates blowdown as required under these conditions.
These setpoints will not be changed as part of the Core Uprate program.
6.2.7 Condensate Polishing System The design conditions for the Condensate Polishing System were evaluated to determine the ability of the polishers to operate at the uprated conditions. The proposed temperature and pressure of the condensate at the polisher inlet (445 psig and 104*F) remain below the polishing system design conditions of 700 psig and 180*F.
From the bounding heat balance (Figure 6.2-2), the condensate flow rate is given as 8,644,000 lb/hr or 2.90 gym per sq ft of Condensate Polishing System filtering surface area. This flow rate is less than the vendors guaranteed filtering capacity of 4.0 spe per sq ft of surface area at design pressure drop.
l Therefore, it has been determined that the condensate polishing system is adequate to operate at the Core Uprate conditions.
Although, at - increased flow rates, filter differential pressure will increase at a faster rate and the backwash frequency will_be slightly higher.
l-B1-1479902-6263 January 28,-1985
+,-e- - ,,w- --,--r ,:- -w -,e, w- -
52 '
6.2.8 Component Cooling System The decreased RCS cold leg temperature decreases the heat load-ings on the Component Cooling Water Systes due to the Chemical and Volume Control System heat exchangers which are designed to remove heat from the letdown flow stream. The letdown line is tied into one RCS cold leg.
3 The Westinghouse supplied heat loads to the Non-Regenerative,
- . Excess Letdown, and Seal - Water Return heat exchangers at the current normal operating condition and associated Stone & Webster Engineering Corporation Design values are shown in Table 6.2-3 as Design Heat Loads. Because the heat loads resulting from the existing plant parameters (555.5'T cold leg temperature) are greater than those as a result -of the Core Uprate program (552.3*F cold les temperature), the current heat load is bounding for the the Core Uprating.
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.u 53-Table 6.2-1 Desian Conditions, Extraction Steam Unit 1
Ist point 450 460 455 428 2nd point 395 260 402 237 3rd point 324 110 347 80 4th point 293 75 292 44 5th point 230 6 233 7 6th point 188 -5. 9 187 -8 Unit 2 1st point 450 460 455 428 2nd point 395 260 402 237 3rd point 324 110 347 80 4th point 293 75 292 44 Sth point 230 6 233 7 6th point 188 -5.9 187 -8
- Maximum Operating Conditions used for pipe stress analysis.
1 i
B1-1479902-6263 m --
i 54 ,
l Table 6.2-2 Design Conditions, Feedwater Heaters 4
Units 1 and 2 Shell Design Pressure- Core Uprated System Pressure Heater (Psia) (Psia)
Ist point 475 -428 2nd point 250 237
- 3rd point 100 80
.1 4th point 50 44 j- 5th point 50 7 6th point 50 -8 Flash Evaporator 15 -8 i
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3 t B1-1479902-6263', January.28, 1955-k
55 Table 6.2-3 Component Coolina System Heat Loads Current Heat Load
- Design Heat Load Heat Exchanger (BTU /HR) (BTU /HR)
Non-Regenerative 16.1 x 10 0 15.72 x 10 6 Excess Letdown 3.26 x 10 0 3.23 x 10 6 0
Seal Water Return 1.02 x 10 1.45 x 10 0 Total 20.38 x 10 6 20.40 x 10 6 4
- Based on a bounding condition associated with the current plant parameters.
B1-1479902-6263 Janua ry 28,_1985
56 6.2.9 Spent Fuel Pool Cooling System There is no impact on the Spent Fuel Pit Heat Loads as a result of the uprating as core power and associated decay levels were based on a core power of 2910 MVt.
6.2.10 Containment Air Recirculation System The heat input into the containment will not increase because of the power uprating and the heat loads for the existing plant conditions are bounding. Since Technical Specifications limit the ambient temperature inside the containment to 105*F, no changes in the operating procedures are necessary.
6.2.11 Control Systems and Instrumentation The bounding Core Uprate heat balance diagram (Figure 6.2-2) was reviewed to determine the effect of the uprate on Balance of Plant (BOP) instrumentation and control valves in the Feedwater, Condensate, Main Steam, and Heater Drain systems. It was deter-mined that all equipment has sufficient margin to be adequate for use at the uprated conditions. No setpoint changes are required as a result of this uprating other than those already specified by Westinghouse.
6.2.12 Electrical Systems An uprated generator output of 1,007,726 KV from the bounding Heat Balance (Figure 6.2-2) at .9 PF yields 29,363 amps to the Isophase Bus Duct. The existing bus duct is rated at 30,500 amps continuous and is adequate for the uprated conditions.
B1-1479902-6263 January 28, 1985
- .- . - - _ .- _. .. __= .
1 57 A review was performed to evaluate the increased loading on the Feedwater, Condensate, and Heater Drain Pump motors resulting from increases in fluid flow rate. As a result of this review, it was determined that the motors and their associated power feeds are i
adequate for the uprated conditions.
Because the increased loadings on the above-mentioned pumps does not exceed the rated horsepower for the respective motors, there is no impact on the Station Service Transformers, Normal Buses, or connecting cables due to the uprate.
The Emergency Buses are not affected because the uprating causes none of the emergency loads to increase.
i The GDC-17 Confirmatory Analysis Studies have been evaluated with respect to the uprating and it has been determined that the GDC-17 confirmatory analyses bound the ' core uprating conditions.
6.2.13 Service Water System The service water flow rate and design heat loads of each com-ponent serviced is not changed due to the core uprate, therefore, the system design capacity and design pressure and temperature i remain unaffected for the core uprate. Existing system para-meters I
were used in the accident analysir..
6.2.14 Radioactive Waste Systems The liquid, solid, and gaseous waste systems have a design based on 2900 MWt. The proposed core uprate conditions are bounded by the design of the current system. The design of the systems are
- based on source terms (reference UFSAR Section 11.1) which Vepco has determined to be applicable for the core uprating up to a
- . batch average fuel burn-up of 43,000 MWD /MTU.
B1-1479902-6263. January 28, 1985 l .
i
58 6.2.15 Boron Recovery Systems The flow rate, temperature,,and pressure of the Boron R'ecovery System are not affected by the core uprating and are bounded by 4
the current' design.
6.2.16 Circulating Water System The heat balance performed for the core uprate conditions utilized
- existing flow rates and inlet temperatures. The outlet ,
temperature increased slightly, but is within the current system i design.
6.2.17 Bearing Cooling System i
- Operating heat loads at the turbine may increase due to the core I
uprate, however, the bearing cooling water flow, temperatures, and pressure are based on the turbine generator design which is
., bounding for the core uprating. The other bearing cooling water requirements are unaffected by the core uprate.
6.2.18 Quench Spray Subsystem
( The quench spray subsystem flow rates _ and fluid temperature are l
l unaffected by the core uprating. Since the containment analysis reported herein gives acceptable results, the current quench spray subsystem design is adequate.
-6.2.19 Recirculation Spray Subsystem a
p Since the containment analyses reported herein give acceptable
, results, the current recirculation spray systes design is l adequate.
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l B1-1479902-6263 Janua ry 28, 1985 i
59
.6.2.20 Auxiliary Steam System l
The Auxiliary Steam Pressure Control Valve (PCV), PCV-AS-105, has been reviewed for the uprated pressure conditions. The valve is designed for a maximum inlet pressure of 1200 psig. This design pressure is higher than the maximum possible upstream pressure of 1020 psia at-no-load conditions. It is, therefore, concluded that the Auxiliary Steam PCV is adequate for the Core Uprate conditions and will reduce uprated Main Steam pressure to 150 psig as originally designed.
6.2.21 Heat Balance Calculations Heat balance diagram Figure 6.2-2 was developed and represents-the Core Uprate program at 2968 MWt (2898 MWt core power x 1.02 plus 12 NWt pump heat) and 880 psia steam pressure. This heat balance was developed for analytical use.
Heat balance Figure 6.2-3 was also developed and represents the Core Uprate program at 2905 MWt (2893 MWt core power plus 12 MWt pump heat) and 850 psia steam pressure. This heat balance reflects the normal operating condition expected at the uprated condition.
6.3 BOP Systems Review Conclusions In conclusion, all BOP systems reviewed are adequate for the uprated conditions.
81-1479902-6263 January 28, 1985
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7.0 BOP COMPONENTS REVIEW 7.1 Genera; The objectives of the review for BOP Cotspenents were to determine the effect of~the core uprating for normal and accident conditions on
.[ .
stress levels and fatigue analysis of all systems within the scope of
.the uprating program. Also, the design calculations were reviewed for major equipment supports, seismic tanks, vessels, pipe rupture re-straints and shields and miscellaneous mecahnical equipment. The components of the following systems were evaluated for the uprated conditions:
System Safety Class Main Steam 2 Feedwater 2 Reactor Coolant 1 RC Loop Letdown 1 RC Loop Excess Letdown 1 Pressurizer Spray 1
- Residual Heat Removal 1,2 4
Low Head Safety Injection 1 High Head Safety Injection 1 CVCS Seal Water Inlet 1 CVCS Seal Water Outlet 1 RC Loop Fill 1 Resistance Temperature Detection 1 RC Loop Drain 1 RC Loop Charging 1 Steam Generator Blowdown 2 Steam Generator Wet Layup- 2 -
Component Cooling to RC Pumps 2 Reactor Vessel Level Indication System 2 B1-1479902-6263 .Janua ry 28, 1985
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7.2 Results of BOP Components Review t
7.2.1 Pipe Stress and Supports l
For all systems within the scope of the uprating program the appropriate maximum stress levels and fatigue analyses results have been reviewed for the plant uprate condr.tions. The review indicated that in all cases' the associated piping stress and fatigue allowables would not be exceeded as a result of the uprate. It is concluded that all existing pipe support and equipment nozzle design loads remain valid.
7.2.2 Major Equipment Supports and Pipe Rupture Restraints No-load and 100 percent power conditions were considered in the review of design calculations for major equipment supports, i seismic tanks, vessels, pipe rupture restraints and shield and miscellaneous mechanical equipment. With the exception of the Main Steam and Feedwater flow rates, the parameters associated with the existing plant parameters bound the respective Core Uprate values. See Sections 6.2.1 and 6.2.4 for evaluation of Main Steam and Feedwater flow rates.
The data used to assess the effect of LOCA loadings on major equipment supports remain applicable for the uprated conditions.
The amplitudes and time history of the LOCA loads data previously supplied by Westinghouse were verified as being applicable.
7.3 BOP Components Review Conclusions In conclusion,' and because the existing plant parameters represent a bounding condition, all existing pipe stress and s'upport analyses within the scope of this evaluation have been determined to remain
- valid under the conditions of the Core Uprate.
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B1-1479902-6263 Janua ry 28, 1985 i
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65 8.0 STRUCTURES l
The only structural loads subject to change as a result of the uprate were those resulting from system parameter changes such as pressure, temperature, and flow and subsequent postulated pipe breaks as
- developed in the review in Section 7.0. The new post LOCA peak con-tainment pressure remained essentially the same, and is still less than th'e original containment design pressure. Subcompartment pressures are expected to decrease from the UFSAR values. There is no effect on the structural integrity of the Reactor Containment or other safety-related structures (Auxiliary Building, Main Steam Valve House, Fuel Building, Safeguards Building) due to the uprate, i
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B1-1479902-6263 Janua ry 28, 1985
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66 9.0 RADIATION PROTECTION 9.1 Operational Releases All operational releases reported in the UFSAR are based on a core power level of 2900 NWt. These releases are based on source terms (reference UFSAR Section 11.1) which Virginia Power has determined to be applicable for the core-uprating up to a batch average fuel burnup of 45,000 MWD /NTU.
9.2 Accident Doses All UFSAR accident doses are bounding for the core uprate. See See-tion 3.2.4 through 3.2.9.
9.3 Radiation Levels In EZD The doses calculated for the North Anna Environmental Zone Description are calculated based on a core. power level of 2910 MWt. This being greater than the proposed core uprate power level of 2898 MWt pre-cludes having to revise these doses.
10.0 REVIEW 0F THE TECHNICAL SPECIFICATIONS The Technical Specifications have been reviewed to determine if any sections could be affected by the proposed increase in power from 2775 MWt to 2893 NWt. With the exception of the Technical Specifica-tion revisions recommended by Westinghouse Electric Corporation, no additional sections require NRC approval.
The Bases section, 3/4.7 Plant Systems, will need revision to reflect the uprated total secondary steam flow.
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