ML20064H738
| ML20064H738 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/21/1978 |
| From: | Stallings C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Parr O Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7812260195 | |
| Download: ML20064H738 (26) | |
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9 VImoxxzA Ex.ncTazc Axn Powna CourAxr Rxcuxown.VrmotwrA sonet December 21, 1978 Mr. Harold R. Denton, Director Serial No. 693 Office of Nuclear Reactor Regulation LQA/ESG:esh y
Attn:
Mr. O. D. Parr, Chief
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Light Water Reactors Branch No. 3 Docket Nos.50-35d Division of Project Management 50-339 U. S. Nuclear Regulatory Commission License No. NPF-4 Washington,'DC 20555
Dear Mr. Denton:
Our response to FSAR Comment 5.81 describes the measures that are and will be used at North Anna Units 1 and 2 in order to reduce the probability and consequences of overpressurization transients of the Reactor Coolant System
/
when it is in a water-solid condition. These measures include administrative controls, as well as long term modifications, empjoying dual setpoints for pressurizer power operated relief valves (PORVs). The PORVs vill be installed prior to fuel loading of Unit 2, and prior to startup following the first scheduled refueling outage for Unit 1.
A Recently, Mr. A. W. Dromerick of the NRC Staff has requested that we provide additional details concerning the design and implementation of the long term modifications. This information, in the form of responses to each of the requests given to us by Mr. Domerick, is attached. These responses discuss the design features of the PORVs, as well as providing the setpoints which will provide assurance that the Appendix G limits will not be exceeded during the Ilmiting pressure transients.
Also attached are our proposed Technical Specifications regarding overpressure protection for Unit 2.
We believe that these specifications address the need for reliable overpressure protection, while recognizing the
. design features of the North Anna systems.
It should be noted that revised reactor coolant system heatup.and cooldown limitation curves have been included with the responses to NRC Staff questions. These curves, generated as part of this analysis, involve only minor revisions from the previous curves, and have already been subnitted to the NRC as part of Amendment 65 to the FSAR. The new curves should be included as part of the Technical Specifications for Unit 2 when the operating license is issued. We will request that the Unit I revisions be made when the Unit 1 overpressure protection specification is issued at the first refueling.
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o Vinoim Eucraic un Powan Commy to Mr. Harold R. Denton, Director 2
Should you have any questions concerning our proposals, we will be pleased to discuss them with you.
Very truly yours, d.7??..JaAy C. H. Stallings Vice President-Power Supply and Production Operations Attachments L
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ATTACHMENT 1 RESPONSES TO NRC STAFF QUESTIONS 1
C0letENT 1 Startup and Shutdown Overpressure Protection i
Identify and justify the most limiting pressure transients caused by mass input and heat input.
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RESPONSE
There are two credible n'eans of adding excess mass to the reactor coolant system 0
when the plant is in a relatively cold (100 F) solid water mode of operation, the first being a possible mismatch betwen the charging and letdown flows and the second an inadvertent start of a High Head Safety Injection Pump. The mismatch
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between the charging and letdown flow is certainly the most likely possibility yet the inadvertent start of a High Head Safety Injection pump has the capability of a grescer rate of mass input and subsequent pressure transient. Therefore, when the n.f is input analysis was conducted, the case considered as most limiting was an insdvertent start of a High Head Safety Injection Pump.
For the analysis of heat input, four possible heat additions were considered; pressurizer heater, core decay heat, and two types of reactor coolant loop tempera-ture asymmetry. The first asyummetry can occur when the reactor coolant is at a
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relatively uniform warm temperature and the cold reactor coolant pump seal injec-tica water continues to enter the system. The coolant injection water settles in the loop seal below the pump inlet. The coolant pressure transient is initiated upon starting one reactor coolant pump. As the pump starts, cold water is dis-charged into the cold leg and reactor vessel inlet plenum.
Simultaneously, cold pools of water in the inactive loops flow through their steam generators (reverse flow), the temperature wl'1. be increased by the heat transferred from the secondary side and, since the coolant cannot expand, the coolant pressure will increase.
The second type of temperatu e asynenetry can occur when the reactor coolant has been cooled down without sufficient circulation, and the steam generators remain at an average temperature higher than that of the reactor coolant. The pressure transient is initiated by starting a reactor coolant pump which can cause the warm water in the steam generators to be replaced by cold water from the loops. Heat from the steam gererators is transferred to the cooler reactor coolant and the pressure increases.
Of the four possible heat inputs considered, the first three are relatively small
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compared to the fourth. The potential excursions caused by a reactor coolant system steam generator asymmetry was considered to be the worst case for the heat input and is the basis of the heat input analysis.
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e C0leENT 2 Startup and Shutdown Overpressure Protection Show that overpressure protection is provided (do not violate Appendix G limits) over ette rance of conditions applicable to shutdown /heacup operation.
RESPONSE
The analysis for power operated re'.1ef valve (PORV) overshoot during solid water modes of operation has been completed on the basis cf a five year operating period for Unit 1 and an eight year operating period for Unic 2, per revised.
t Appendix G heatup and cooldown curves supplied by Westinghouse Electric Corpora-tion (Figures 2.1 through 2.4).
With the analysis now complete, it has been determined that PORV settings of 430 i
psig for Unit 1 and 405 peig for Unit 2, with calculated overshoots of 519.5 pois -
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for Unit 1 and 497 psig for Unit 2, are within acceptable limits of the Appendix l
G heatup and cooldown curves. The limiting pressure transient for both Unit 1 j
and Unit 2 is the assa input case.
However actual PORV settings will be as i llows:
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Unit 1 4
j PORV 1 425 pais open 410 psig close PORV 2 410 psig open 395 psig close 4
Unic 2 PORV 1 400 psig open 385 peig close PORV 2 385 peig open 370 psig close Peak pressure transients for actual PORV settings are listed below with corres-ponding Appendix G pressure limits at 100 F.
Peak pressure transient Appendix G limit Unit 1 PORV 1 515 1 peig 520 psig @ 100 F Unit 1 PORV 2 501.3 psig Unit 2 PORV 1 492.4 peig 500 psig @ 100 F Unit 2 PORV 2 478.9 psig
'l By lowering the PORV settings below the acceptable settings as specified, we are adding to the conservatism already present in the Appendix G heatup and cooldown curves for possible instrument error, which is 60 psig and 100 F.
It should further be noted that opening times of the PORV's have been improved since the completion of the analysis, once again adding to the conservatism.
PORV settings as specified will provide assurance that Appendix G limits will not be exceeded during the limiting pressure transients. Sufficient safety me.rgin is provided by these set points since transient analyses show that only one PORV, set to open at 430 psig for Unit 1 and 405 psig for Unic 2 is required
.to prevent exceeding the Appendix C limits, i.
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N1F 3000 MATERI AL PROPERTY 6 ASIS CONTROLLING MATERIAL: FORGED METAL COPPER CONTENT: 0.16 WT%
PH0SPHORUS CONTENT: 0.019 WTf.
RT iMITIAL: 38'F MDT RT AFTER 5 EF M.
NDT I /4T, 13 4*F 3 /4T, 104'F F 2000 E
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W UMACCEPTA BLE U
OPERATION OPERATION x
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1000 HEATUP RATE UP T0100 ('F/HR) 4 CRITICALITY LIMIT 100 ]
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100 200 300 400 500 INDICATEDTEMPERATURE(F)
Figure 2.1 North Anna Power Station No.1 Reactor Coolant System Heatup Limitations Applicable to 5 Effective Full Power Years and Contains Margins of 10 F and 60 PSIG for Possible Instrument Errors 0
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e N1F 3000 MATERI AL PROPERTY BA$l$
CONTROLLING MATERI AL: F9RGED METAL COPPER CONTENT: 0.16 WT/.
PHOSPHORUS CONTENT: 0.019 WT,(
RT INITIAL: 38'F NDT RT AFTER 5 EFPY:
NDT 1/47. 134'F 3/4T, 104*F g 2000 G
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UNACCEPTABLE OPERATION 80e E-1000 C00LDOWN RATES (*F/HR)
ACCEPTABLE 25 50 100 0
0 100 200 300 400 500 INDICATEDTEMPERATURE(F)
Figure 2.2 North Anna Power Station No.1 Reactor Coolant System Cooldown Limitations Applicable to 5 Effective Full Power Years and 0
Contains Margins of 10 F and 60 PSIG for Possible Instrument Errors
N1F 3000 MATERI AL PROPERTY BASIS CONTROLLING MATERIAL: FORGED HETAL COPPER CONTENT: 0.13 WT%
PHOSPHORUS CONTENT: 0.013 WT%
RT IMITIAL: 56*F NDT RT AFTER 8 EFPY:
M01 t/4T,148 2/4T,il9*F q
2000 t
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UNACCEPTABLE o$
OPERATION ACCEPTA BLE S
O OPERATION 1000 l
HEATUP RATE (*F/HR) 100 1
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100 200 300 400 500 INDICATEDTEMPERATURE(F)
Figure 2.3 North Anna Power Station No. 2 Reactor Coolant System Heatup Limitations Applicable to 8 Effective Full Power Years and Contains Margins of 10 F and 60 PSIG for Possible Instrument Errors
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3000 MATERI AL PROPERTY BAS f 3 CONTROLLING MATERIAL: FORGED METAL COPPER CONTENT: 0.13 WT%
PHOSPH0RUS CONTENT: 0.013 WT%
RT IMITIAL: 56 9 MOT RT AFTER 8 EFPY:
NDT Il4T,148'F
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UNACCEPTABLE CPERATION E
ACCEPTABLE 5
1000 OPERATION COOLDOWN RATES ('F/HR) 0 l
25 50 100 I
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O 100 200 300 400 500 INDICATED TEMPERATURE ( F)
F;gure 2.4 North Anna Power Station No. 2 Reactor Coolant System Cooldown Limitations Applicable to 8 Effective Full Power Years and Contains Margins of 100F and 60 PSIG for Possible Instrument Errors
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4 COMMENT 3 Startup and Shutdown Overpressure Protection
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Identify and justify that the equipment will meet pertinent parameters assumed in the scalyses (e.g., valve opening times, signal delay, valve capacity).
RESPONSE
1.
Valve opening times were established by field testing on North Anna Unit 2 installation using bottled nitrogen supply. Average time was 2.14 seconds to open and 1.1 seconds to close. The opening times are conservative since pressure was not under_ the seat during testing and the spring in the power operated relief valves opposes pressure under the seat.-
2.
The signal delay times include the pressure transmitter response time summed with the relay activation time. As stated in the response to FSAR Comment 7.8, the pressure transmitter response time is 0.4 seconds. Typical relay actuation times are 0.015 seconds. The total signal delay would be less than 0.5 seconds.
3.
The valve capacity at full open for water at 100 F and 300 psig drop through
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the valve at 400 psig inlet pressure is as follows: Cy of North Anna Unit 1 and 2 valves at full open is 46:
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300 46 3 0.9946 Q = 799 gym The reference relief valve used by Westinghouse in the " Pressure Mitigating Systems Transient Analysis Results" July 1977, use a valve having a C, of 50,-
I and flows for North Anna 1 and 2 are slightly lower since they have a Cy of 46.
This lower Cy reduces the flous shown in the figure by 8 percent..
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The capacity with 8 percent reduction (799 gpa) is greater than run out flow for one charging pump (600 spa).
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The Cy is proportional to valve opening and is linear for North Anna 1 and 2; j
for example, at 3/4 open valve position, the C y is 34.5.
1 1
4.
Backup Nitrogen Supply Calculation for North Anna 1 and 2 for 120 cycles for each PORV.
4 s.
Initial pressure 2,000 psig b.
Low pressure al.arn 1,950 peig c.
Valve acuator volume is 680 cubic inches, and 184 cubic inches is allowed k
for leakage for one stroke. A total of 864 cubic inches = 0.5 cubic' feet per stroke.
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55 psig is required to stroke the valve Standard cubie feet required to stroke valve 120 strokes at 55 peig e.-
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o COMMENT 3 (Cont'd) Startup and Shutdown Overpressure Protection PV
= Standard Cubic Feet 14.7 120 x 0.5 cu ft x 70 psia = 285.7 scf 14.7 psia f.
Volume of tank required starting at 2,000 psig and keeping a reserve of 585 psig at end of 120 cycles.
PV = Standard Cubic Feet 14.7 (2,015 psia - 600 psia) jV = 285.7 scf 14.7 psia V = 2.968 cu ft Tank volume will be 11.9 cubic feet for each valve.
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COMMENI 4 Startup and Shutdown Overpressure Protection Provide a description of the system including relevant P&I drawings and electrical schematics.
RESPONSE
Attached is the RCS Overpressure Protection System Description.
Also attached are the following figures:
4.1 Unit 1 Power Operated Rel; ef Valves, Nitrogen Supply System (Unit 2 is identical) i 4.2A,B Logic Diagrams - Pressurizer Press. Cont. RCS Overpress.
Protection 4.3 Z1ementary Diagram - Annunciator 4.4 Control Switch Contact Diagram 4.5 Elementary Diagram - Solenoid Operated Valves
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J.0.50. 120$0 NORTH ANNA POWER STATICIP VIRGINIA ELECTRIC AND POWER COMPANY.
PRESSURIZER PRESSURE CO:.' TROL AND RCS OVERPRESSURIZATION PROTECTION SYSTEM DESCRIPTION 25-9, A.
REFERENCE 12050-LSK-25-9A & 3 1.0 Logic Diagram 12050rESK-6NR,10C Sh 2 &
2.0 Elementary Diagram 10AAJ B.
DESIGN Two pressurizer power operated relief valves PCV-2456 and 1.0 2455C are provided to relieve pressure from the pressurizer RC-E-2 to the pressurizar relief tank RC-TK-2.
The valves j
PCV-2456 and 2455C are located downstream of the, isolation
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valves MOV-2535 and 2536, respectively.
2.0 The PORV's are provided to protect the reactor coolant system from overpres,sure during normal operation and during solid weette. modes of operation.
I The PORV's have dual setpoints and during normal plant-operation are supplied by containment instrument air.
During solid water modes of operation, a 3-way solenoid is automatically energized and the pneumatic supply is switched from instrument air to nitrogen from the gas supply area. Redundant nitrogen reserve tanks are also provided in the event of loss of bottled nitrogen supply.
See Figure 1 for nitrogen supply arrangement.
3.0 Alaras are provided to protect reactor coolant system I
against high pressure during start-up or shut-down and loss of reactor coolant system overpressurization protection which are annunciated in the main control room.
4.0 The relief valve solenoids are powered from the following circuits':
Valve _
Solenoid 125 V DC l
PCV-2456 50V-2456-1, 2,& 3 PNL-15 Ckt 23 PCV-2455C SOV-2455C-1, 2,& 3 PNL-1A Ckt 23 I
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OPERATION i
1.0 Logic Diagram 25-9A & B illustrates the operation of
' pressure control valve PCV-2456 pressurizer relief valve.
i During normal operation PCV-2456 is closed. Three f
solenoid valves 50V-2456-1 and -2.to admit air and -3 to admit nitrogen to PCV-2456 to open. The operation of PCV-2455C is similar.
1.1 SOV-2456-1 and -2 will energize to admit air to open PCV-2456 manually by placing the selector i
i switch in the OPEN position, provided 50V-2456-3 is not energized.
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i 1.2 507-2456-1 and -2 will energize to admit air to open PCV-2456 automatically when the selector switch is in the AUTO position and all the following conditions exist:
1.2.1 SOV-2456-3 is not energized 1.2.2 Pressurizer pressure high signal from two of three sources (Train 3) i 1.2.3 Pressurizer pressure ehme el high r
1.3 507-2456-1 and -2 will de-energize to vent air to close PCV-2456 when the selector switch is in the l
AUTO position and neither 1.2.2 nor 1.2.3 conditions exist or the selector switch is in the CLOSE position, provided SOV-2456-3 is not energized.
1.4 Sov-2456-3 will be de-energized to open air passage for SOV-2456-1 and -2 to open PCV-2456 when any of l
the following conditions are satisfied:
1.4.1 Keysvitch in OFF position
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1.4.2 Keysvitch to auto and reactor coolant system pressure not high i
1.5 Sov-2456-3 vill energize to admit nitrogen to open i
PCV-2456 when the plant is in a solid water condition sad all of the following conditions exist!
1.5.1 Kayswitch in AUTO position l
1.5.2 Reactor coolant, system press high w
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1.6 SOV-2456-3 will energize when its keyswitch is in the OPEN position.
D.
CONTROL AND MONITORINC DEVICE
SUMMARY
1.0 Main control Board 1.1 (2) OPEN-AUTO-CLOSE selector switch for pressurizar relief valves (PCV-2456 and 2455C) 1.2 (7) OPEN-AUTO-OFF keysvitch for pressurizer relief valves (PCV-2456 and 2455C) 1.3 (2) Red (open) and green (closed) indicating lights for valves (PCV-2456 and 2455C) 1.4 (2) Low pressure alarms'for nitrogen tanks (nos. 358 and 359) 1.5 (1) Alarm for RCS SOLID AND NDT HI PRESS (No. 156) 1.6 (1) Alarm for RCS SOLID AND NDT PRESSURE PROTECTION REQUIRED (No. 155)
E.
SPECIAL OPERATING PRECAUTIONS AND NOTES 1.0 Prior to testing of PCV-2456, motor-operated isolation valve MOV-2535 must be closed and reopened upon completion of the test.
Sf=11=r for testing PCV-2455C and MOV-2536.
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.f COMENT 5 Startup and Shutdown Overpressure Protection Discuss how the system meets the criteria.
RESPONSE
See FSAR page S5.81-8, item 2a for the reply to this comment which was answered previously. l ( \\ \\ i 1 D y
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COMMENT 6 Startup and Shutdown Overpressure Protection Discuss all administrative controls required to implement the protection system.
RESPONSE
Administrative controls consist of a three-position key lock switch located on the main control board bench section for each power-operated relief valve. The key lock switch has three positions: "0 PEN-AUTO-OFF." When the operator is alerted by the main control board annunciator that pressure protection is required, the operator will place both key lock switches in the "AUT0" position. This action will allow the power-operated relief valves to cycle open when the pressure set points are reached, thus preventing an overpressurization condition. Additional administrativa controls are discussed in the response to FSAR' Comment 5.81. L e e
F O e e ATTACHMENT 2 PROPOSED OVERPRESSURE PROTECTION SPECIFICATION NORTH ANNA UNIT 2 .s s I
. _ _ ~ _ s.s/ c PROPOSED OVERPRESSURE PROTECTION TECHNICAL SPECIFICATION NORTH ANNA 2 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9 3 Two power opsrated relief valves (PORVs) with a lift setting of j[ 405 psig shall be OPERABLE. APPLICA31LITY: 0 When the reactor coolant system average temperature is j[ 340 F, and the reactor vessel head is bolted. ACTIGN,: a. With one PORV inoperable, either restore the looperable PORV to OPERABLE status within 7 days or depressurize the RCS and open one PORV within the next 8 hours; maintain the RCS depressurized and verify one PORV is open at least once per 12 hours until both PORVs have been restored to OPERABLE status. b. With both PORVs inoperable: 1 1. Either maintain a bubble in the pressurizer with a maximum pressurizer level of 33%, and restore both PORVs to OPERABLE status with 72 hours, or 2. Depressurize the RCS within 8 hours. When the RCS has been depres-surized, open one PORV and verify that one PORV is open at least once k _/ per 12 hours, or establish the conditions listed below and verify their implementation at least once per 12 hours: 3 a. Maintain a maximum pressurizer level of 333. i b. Maintain the series RHR inlet valves open, or an alternate letdown path OPERABLE. i c. Limit charging flow to less than.150 gpm. d. Maintain safety injection accumulator discharge valves closed. Maintain the RCS depressurized until bott. POR'Js have bees; restored to OPERABLE status. c. In the event the PORVs are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant l to Specification 6.9 2 within 30 days. The report shall describe the cir-I t l l
cunstances initlaring the transient, the effect of the PORVs on the transient and any corrective action necessary to prevent recur'ence. d. The provisions of Specification 3 0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.4.9.3 Each PORV shall be demonstrated OPERABLE by: 1. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE. b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel, at least once per 18 months. c. Verifying the PORV isolation valve is open at least ones per 72 hours when the PORV is being used for overpressure protection. d. Testing pursuant to Specification 4.0.5 BASES The OPERABILITY of two PORVs, or the RCS vented through an opened PORV, ensures that the RCS will be protected from pressu.e transients which could exceed the limits of Appendix G to 10 CFR Part 50, when the Reactor Coolant System is water solid and the Reactor Vessel Head is bolted. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator < 20 F above the RCS cold leg temperature, or (2) the start of a cnarging pump and Tes injection into a water solid RCS. When the Reactor Vessel tread is unbolted, a RCS pressure of less than 100 psig will lift the head, thereby creating a relieving capability equivalent to at least one PORV. 0 340 F, overpressure protection When the Reactor Coolant average temperature is is provided by a bubble in the pressurizer and/or pressurizer safety valves. If both PORVs are Inoperable, a limited period of time (72 hours) is allowed to restore the valves to OPERABLE status, without necessitating depressurization of the RCS, if a bubble is maintained in the pressurizer. In such a case, a maximum pressurizer level of 33% has been selected to provide sufficient time, approximately 10 minutes, for orerator response in the event of a malfunction t resulting in maximum flow from one charging punp. When a bubble is not maintained, and the RCS is depressurized, and it is impossible to manually open a_t least one PORV, additional administrative controls are imple- _ mented to further reduce the possibility of an overpressure transient. l I i i .~. .}}