ML20101L681
| ML20101L681 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/01/1996 |
| From: | Imbro E NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20101L684 | List: |
| References | |
| NUDOCS 9604040172 | |
| Download: ML20101L681 (19) | |
Text
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UNITED STATES s
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2065&m01
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VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-338 NORTH ANNA POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 200 License No. NPF-4 l
1.
The Nuclear Regulatory Commission (the Commission) has found that:
l A.
The application for amendment by Virginia Electric and Power Company l
et al., (the licensee) dated July 26, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as i
amended (the Act), and the Commission's rules and regulations set l
forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; l
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have l
been satisfied.
l l
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9604040172 960401 PDR ADOCK 05000338 P
PDR l
j 2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.D.(2) of Facility Operating License No. NPF-4 is hereby 1
amended to read as follows:
(2) Technical Soecifications i
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 200, are hereby incorporated in the license. The licensee shall operate the facility in accordance with l
the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
l FOR THE NUCLEAR REGULATORY COMMISSION A
q~ct Eugene V. Imbro, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 1, 1996 l
1 l
i a
ATTACHMENT TO LICENSE AMENDMENT NO. 200 TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET N0. 50-338 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.*
Remove Paaes Insert Paaes 2-5 2-5*
2-6 2-6 3/4 4-5 3/4 4-5*
3/4 4-6 3/4 4-6 3/4 4-7 3/4 4-7 i
B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-2a 1
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
I With a reactor trip system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3.1.1 until the. channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
i NORTH ANNA - UNIT 1 2-5
%O
- o TABLE 2.2-1 N
REAClDR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SElPOINT ALLOWABLE VALUES 1.
Manual Reactor Trip Not Applicable Not Applicable C
2.
Power Range, Neutron Flux Imw Setpoint-s 25% of RATED Imw Setpoint - 5 26% of RATED THERMAL POWER TilERMAL POWER High Setpoint - s 109%** of RATED liigh Setpoint - 5110%"* of RATED DIERMAL POWER DIERMAL POWER 3.
Power Range, Neutron Flux, s 5% of RATED THERMAL POWER 55.5% of RATED DIERMAL POWER liigh Positive Rate with a time constant 2 2 seconds with a time constant 2 2 seconds 4.
Power Range, Neutron Flux, s 5% of RATED THERMAL POWER 55.5% of RATEDDfERMAL POWER liigh Negative Rate with a time constant 2 2 seconds with a time constant 2 2 seconds 5.
Intermediate Range, Neutron s 25% of RATED THERMAL POWER s 30% of RATED THERMAL POWER Flux 5
5 6.
Source Range, Neutron Flux s 10 counts persecond s 1.3 x 10 counts per second 7.
Overnemperature AT See Note 1 See Note 3 8.
Overpower AT See Note 2 See Note 3 9.
Pressurizer Pressum -Iow 21870 psig 21860 psig 10.
Pressurizer Pressure - High s 2360 psig s 2370 psig l
11.
Pressurizer Water level-High 592% ofinstrument span s 93% ofinstrument span 12.
Loss of Flow R
2 90% of design flow per loop
- 2 89% of design flow perloop*
a i
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Design flow per loop is one-third of the minimum allowable Reactor Coolant System Total Flow Rate as specified in Table 3.2-1.
The high trip setpoint for Power Range, Neutron Flux, shall be s 103% RATED THERMAL POWER for the period of operation
{'
until steam generator replacement.
K
"* The allowable value for the high trip setpoint for Power Range, Neutron Flux,is required to be s 104% RATED TilERMAL POWER for the period of operation until steam generator replacement.
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REACTOR COOLANT SYSTEM ISOLATED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.5 A reactor coolant loop cold leg stop valve shall remain closed until:
a.
The isolated loop has been operating on a recirculation flow of greater than or equal to 125 gpm for at least 90 minutes and the temperature at the cold leg of the isolated loop is within 20'F of the highest cold leg temperature of the operating loops.
l b.
The reactor is subcritical by at least 1.77 percent Ak/k.
APPLICABILITY:
ALL H0 DES.
ACTION:
With the requirements of the above specification not satisfied, suspend startup of the isolated loop.
SURVEILLANCE REQUIREMENTS 4.4.1.5.1 The isolated loop cold leg temperature shall be determined to be l
within 20*F wr the highest cold leg temperature of the operating loops within 30 minutes 9* ar to opening the cold leg stop valve.
4.4.1.5.2 (he reactor shall be determined to be subcritical by at least 1.77 percent av./k within 10 minutes prior to opening the cold leg stop valve.
NORTH ANNA - UNIT 1 3/4 4-5 Amendment No. 32
REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting i
of 2485 PSIG i 3% as-found and 1% as-left."
l AEPLICABILITY-MODE 4.
ACTION:
With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation.
SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those reqmred by Specification 4.0.5.
The lift setting pressure shall correspond to ambient condition of the valve at nominal operating temperature and pressure.
NORTH ANNA - UNIT 1 3/44-6 Amendment No. 444. 200
l 1
REACTOR COOLANT SYSTEM SAFETY AND RFI TFF VALVES - OPERATING SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.4.3.1 All pressurizer code safety valves shall be OPERABLE with a lift setting of j
2485 PSIG + 2% / - 3% average as-found with no single valve outside i 3%, and i 1% per valve as-left.
APPLICABTUTY-MODES 1,2 and 3.
ACTION:
With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.3.1 No additional Surveillance Requirements other than those required by Specifration 4.0.5.
1 The lift setting pressure shall correspond to ambient conditions of the valve at nominal temperature and pressure.
NORTH ANNA - UNIT 1 3/44-7 Amendment No. -189, 200
3/4A REACTOR COOLANT SYSTEM BASES within 20 F of the operating loops. Making the reactor subcritical prior to loop startup prevents any j
power spike which could result from this cool water induced reactivity transient.
l l
3/4.4.2 AND 3/4.4.3 S AFETY AND RFI IFF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 380,000 lbs per hour l
l of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during hot shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, or the power operated relief valves (PORVs) will provide overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load l
assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e., no credit is taken for a direct reactor trip on the loss ofload) and also assuming no operation of the power operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings will occur only during shutdown and will l
be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
The safety valve tolerance requirement for Modes 1-3 is expressed as an average value.
l That is, the as-found error (expressed as a positive or negative percentage) of each tested safety l
valve is summed and divided by the number of valves tested. This average as-found value is compared to the acceptable range of + 2% to - 3%. In addition, no single valve is allowed to be l
outside of 3%.
l l
An average tolerance of + 2% /- 3% was confirmed to be adequate for Modes 1-3 accident analyses. For the overpressure events, the analyses considered several combinations of valve tolerance with the arithmetic average of the three valves' tolerance equal to + 2% (with no valve outside of 3%). The case of a + 2% tolerance on each of the three valves provided the most limiting results. The - 3% tolerance is limiting for the DNB acceptance criterion.
The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. Each PORV has a remotely operated block valve to provide a j
positive shutoff capability should a relief valve become inoperable.
NORTH ANNA - UNIT I B 3/4 4-2 Amendment No. 32, "!,189. ?'"'
3/4 4 REACTOR COOLANT SYSTEM BASES The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:
a) Manual control of PORVs to control reactor coolant system pressure. This is a function that may be used to mitigate certain accidents and for plant shutdown.'
b) Maintaining the integrity of the reactor coolant pressure boundary. This function is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.
c) Manual control of the block valve to (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a, above), and (2) iso-late a PORV with excessive seat leakage (Item b, above).
d) Automatic control of PORVs to control reactor coolant system pressure. This function reduces challenges to the code safety valves for overpressurization events.
e) Manual control of a block valve to isolate a stuck-open PORV.
Surveillance Requirements provide the assurance that the PORVs and block valves can perform their functions. Specification 4.4.3.2.1 addresses the PORVs and Specification 4.4.3.2.2 addresses the block valves. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORV to operable status.
Surveillance Requirement 4.4.3.2.1.b.2 provides for the testing of the mechanical and electrical aspects of control systems for the PORVs.
Testing of PORVs in HOT STANDBY or HOT SHUTDOWN is required in order to simulate the temperature and pressure environmental effects on PORVs. Testing at COLD SHUTDOWN is not considered to be a representative test for assessing PORV performance under normal plant operating conditions.
3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial S AR assumptions. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.
NORTH ANNA - UNIT I B 3/4 4-2a Amend ent No. -149. 206
,/ " '%
4 UNITED STATES 1
^
NUCLEAR REGULATORY COMMISSION 3
s,.....j WASHINGTON, D.C. 20006-00%
VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.181 License No. NPF-7 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated July 26, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
i
(
_ _. - 2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-7 is hereby i
amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 181, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Tec.hnical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMUiSlSN ftd-Eugene V. Imbro, Director i
Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 1,1996 l
i O
l ATTACHMENT TO LICENSE AMENDMENT NO.181 TO FACILITY OPERATING _ LICENSE NO. NPF-7 DOCKET NO. 50-339 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.*
Remove Paaes Insert Paaes 2-6 2-6 3/4 4-5 3/4 4-5*
3/4 4-6 3/4 4-6 3/4 4-7 3/4 4-7 B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-2a l
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TABLE 2.2-1 N
Y REAC1DR TRIP SYSTEM INSTRUMENTATION 1 RIP SETPOINTS
(
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES l
1.
Manual ReactorTrip Not Applicable Not Applicable 2.
Power Range, Neutron Flux Imw Setpoint-s 25% of RAED IAw Setpoint - 5 26% of RATED 111ERMAL POWER 111ERMAL POWER y
High Setpoint -s 109% of RATED High Setpoint-5110% of RATED DIERMAL POWER 111ERMAL POWER t
3.
Power Range, Neutron Flux, s 5% of RATED THERMAL POWER 55.5% ofRATED THERMAL POWER liigh Positive Rate with a time constant 2 2 seconds with a time constant 2 2 seconds 4.
Power Range, Neutron Flux, s 5% of RATED THERMAL POWER s5.5% of RATED 111ERMAL POWER liigh Negative Rate with a time constant 2 2 seconds with a time constant 2 2 seconds 5.
Intermediate Range, Neutron s 25% of RATED THERMAL POWER s 30% of RATED THERMAL POWER Flux 5
5 6.
Source Range, Neutron Flux s 10 counts persecond s 1.3 x 10 counts per second 7.
Ovenemperature AT See Note 1 See Note 3 8.
Overpower AT See Note 2 See Note 3 9.
Pressurizer Pressure -IAw 21870 psig 21860 psig i
10.
Pressurizer Pressure - High s 2360 psig s 2370 psig 11.
Pressurizer Water level-High s 92% ofinstrument span s 93% ofinstrument span 12.
Loss of Flow 2 90% of design flow perloop*
2 89% of design flow per loop
- Ra i
Design flow per loop is one-third of the minimum allowable Reactor Coolant System Total Flow Rate as specified in Table 3.2-1.
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REACTORCOOLANTSWTEM ISOLATED LOOPSTARTUP I
UMITING CONDITION FOR OPERATK)N 3.4.1.5 A reactor coolant loop cold leg stop valve shall remain closed until:
- a. The isolated loop has been operating on a recirculation flow greater than or equal to 125 gpm for at least 90 minutes and the temperature at the cold leg of the isolated loop is within 20'F of the highest cold leg temperature of the operating loops.
I
- b. The reactor is subcritical by at least 1.77 percent ak/k.
l
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APPLICABILITY: ALL MODES.
ACTION With the requirements of the above specification not satisfied, suspend startup of the isolated loop-SURVEILLANCE REQUIREMENTS 4.4.1.5.1 The isolated loop cold leg temperature shall be determined to be within 20*F of the j
highest cold leg temperature of the operating loops within 30 minutes prior to oponing the cold leg stop vahre.
F
[
4.4.1.5.2 The reactor shan be determined to be subcritical by at least 1.77 percent ak/k l
j within 30 minutes prior to opening the cold leg stop vMio.
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l NORTH ANNA-UNIT 2 3/4 45 Amendment No. 129 1
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SAFETY VALVES - SHUTDOWN l
LIMITING CONDITION FOR OPERATION l
3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 PSIG i 3% as-found and i 1% as-left.
l APPLICABILITY
- MODE 4.
ACTION:
l 1
With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive mactieity changes and place an OPERABLE RHR loop into operation.
SURVEILLANCE REQUIREMENTS l
4.4.2 No additional Surveillrace Requirements other than those required by Specification 4.0.5.
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The lift setting pressum shall correspond to ambient condition of the valve at nominal j
operating temperature and pressure.
NORTH ANNA - UNIT 2 3/44-6 Amendment No. 44.181
REACTOR COOLANT SYSTEM SAFETY AND RFI TFF VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.3.1 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 PSIG + 2% / - 3% average as-found with no single valve outside i 3%, and i 1% per valve as left.
APPLICABII. TTY:
MODES 1,2 and 3.
ACTION:
With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.3.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
1 The lift setting pressure shall correspond to ambient conditions of the valve at nominal temperature and pressure.
NORTH ANNA - UNTT 2 3/44-7 Amendment No. I"
- _ -. - =.
-~
REACTOR COOLANT SYSTEM BASES i
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3/4.4.2 AND 3/4.4.3 SAFETY AND RFI TFF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 380,000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during hot shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, or the power I
operated relief valves (PORVs) will provide overpressure relief capability and will prevent RCS l
overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load l
assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance w;th the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
The safety valve tolerance requirement for Modes 1-3 is expressed as an average value.
That is, the as-found error (expressed as a positive or negative percentage) of each tested safety valve is summed and divided by the number of valves tested. This average as-found value is compared to the acceptable range of + 2% to - 3%. In addition, no single valve is allowed to be outside of 3%.
l An average tolerance of + 2% /- 3% was confirmed to be adequate for Modes 1-3 accident l
analyses. For the overpressure events, the analyses considered several combinations of valve tolerance with the arithmetic average of the three valves' tolerance equal to + 2% (with no valve I
outside of 3%). The case of a + 2% tolerance on each of the three valves provided the most f
limiting results. The - 3% tolerance is limiting for the DNB acceptance criterion.
The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety l
valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability j
should a relief valve become inoperable.
J j
The OPERABILITY of the PORVs and block valves is determined on the basis of their j
being capable of performing the following functions:
i NORTH ANNA - UNIT 2 B 3/4 4-2 Amendment No.12 t, !~'^.
181
34M REACTOR COOLANT SYSTEM BASES a) Manual control of PORVs to control reactor coolant system pressure. This is a function that may be used to mitigate certain accidents and for plant shutdown.
b) Maintaming the integrity of the reactor coolant pressure boundary. This function is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.
c) Manual control of the block valve to (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressum (Item a, above), and (2) iso-late a PORV with excessive seat leakage (Item b, above).
d) Automatic control of PORVs to control reactor coolant system pressure. This function reduces challenges to the code safety valves for overpressurization events.
e) Manual control of a block valve to isolate a stuck-open PORV.
Surveillance Requirements provide the assurance that the PORVs and block valves can petform their functions. Specification 4.4.3.2.1 addresses the PORVs and Specification 4.4.3.2.2 addresses the block valves. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORV to operable status.
Surveillance Requirement 4.4.3.2.1.b.2 provides for the testing of the mechanical and electrical aspects of control systems for the PORVs.
Testing of PORVs in HOT STANDBY or HOT SHUTDOWN is required in order to simulate the temperature and pressure environmental effects on PORVs. Testing at COLD SHUTDOWN is not considered to be a representative test for assessing PORV performance under normal plant operating conditions.
3/4.4.4 Eltf.MURIZER Thu limit on the maximum water volume in the pressurtzer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. 'Ihe limit is consistent with the initial SAR assumptions. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum vrater volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be
~
OPERABLE ensures that the plant will be able to establish natural circulation.
NORTH ANNA - UNIT 2 B 3/4 4-2a Amendment No. 4M. W
. - - -. _