ML20087A117

Proposed Tech Specs,Increasing Pressurizer Safety Valve Lift Setpoint Tolerance as Well as Reduce Pressurizer High Pressure Rt Setpoint & Allowable Value

ML20087A117
Person / Time
Site:	North Anna
Issue date:	07/26/1995
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20087A114	List: ML20087A112 ML20087A117
References
NUDOCS 9508040178
Download: ML20087A117 (14)

Text

t

. '@$ Z i

O TABLE 2.2-1 -

o -j REACTOR TRIP SYSTEM INSTRUMENTATION TR'P SETPOINTS

xm z

$o "E RGS FUNCTIONAL UNIT TRIP SETPOINT

' ALLOWABLE VALUES

! 0 oo c 8 >- 1. Manual ReactorTrip Not Applicable Not Applicable

y8d 2E 2. . Power Range, Neutron Flux Low Setpoint- s 25% of RAEiD low Setpoint - s 26% of RATED

! 37 ,H

_, EIERMAL POWER TIIERMAL POWER m

Iligh Setpoint - s 109%** of RATED liigh Setpoint - s 110%*** of RATED

, TiiERMAL POWER TIIERMAL POWER *

3. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER s5.5% of RATEDTilERMAL POWER Iligh 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 T11ERMAL POWER 55.5% of RATEDTHERMAL POWER

, liigh Negative Rate with a time constant 2 2 seconds with a time constant 2 2 seconds y 5. Intermediate Range, Neutron s 25% of RATED TilERMAL POWER s 30% of RATED TIIERMAL POWER

& Flux '

s

6. Source Range, Neutron Flux s lo counts per second s 1.3 x 10 5counts per second i

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 3

9. Pressurizer Pressum - Low 21870 psig 21860 psig

10. Pmssurizer Pressum - Iligh s 2360 psig s 2370 psig iL Pressurizer Water I2 vel-liigh s 92% ofinstrument span s 93% ofinstrument span E 12. Loss of Flow 2 90% ei dic flow a per loop

• 2 89% of design flow perloop*

E Z

$m Design flow per loop is one-third of the minimum allowable Reactor Coolant System Total Flow Rate as specified in Table 3.2-1.

a **

The high trip setpoint for Power Range, Neutron Flux, shall be 5103% RATED TIIERM AL POWER for the period of operation

{ until steam generator replacement.

j0 "* The allowable value for the high trip setpoint for Power Range, Neutron Flux,is required to be s 104% RATED TIIERMAL POWER for the period of operation until steam generator replacement. t

. - _ _ _ _ __--- _ - __.---- - ..- -.- -- - .~.. -- .-. , . . , - - - - . - - - _ - - , - - - - - ~ . . , . - . . - - - , . --

l 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 .

of 2485 PSIGi3% as-found and 1% as-left.' l APPLICABILITY: MODE 4.

ACTION:

5 With no pmssurizer 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 required by Specification i 4.0.5.

l l

l l

The lift setting pressum shall conespond to ambient condition of the valve at nominal j operating temperatum and pressure.

l NORTH ANNA - UNIT 1 3/44-6 Amendment No. 444, i i

I

REACTOR COOLANT SYSTEM SAFETY AND RELIEF 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 2485 PSIG + 2% / - 3% average as-found with no single valve outside i 3%, and i 1% per valve as-left.

APPLICABILITYi MODES 1,2 and 3.

ACTION:

With one pmssurizer code safety valve inoperable, either mstom 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 l

l l

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal temperature and pressure.

NORTil ANNA - UNIT 1 3/44-7 Amendment No. 489,

l l

t ,

3144 REACTOR COOLANT SYSTEM  !

BASES within 20 F of the operating loops. Making the reactor subcritical prior to loop startup prevents any l l power spike which could result from this cool water induced reactivity transient.  !

3/4.4.2 AND 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized I 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 OPERABl.E. 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 i 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 ofload 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 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.

That is, the as-found error (expressed as a positive or negative percentage) of each tested safety valve is summed and divided by the numbe of valves tested. This average as-found value is compared to the acceptable range of + 27c to 37c. In addition, no single valve is allowed to be outside ofi 37c.

An average tolerance of + 2% / - 37c 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 + 27c (with no valve outside of 37c). The case of a + 2% tolerance on each of the three valves provided the most limiting results. The - 37c 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 positise shutoff capability should a relief valve become inoperable.

NORTH ANNA - UNIT 1 B 3/4 4-2 Amendment No. 32,111 !89,

4 3/4A REACTOR COOLANT SYSTEM BASES The OPERABILITY of the PORVs and block valves is determined on the basis of their l being capable of performing the following functions: i 1

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  ;

perfonn 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 IlOT SHUTDOWN is required in order to simulate the temperature and pressure environmental effects on PORVs. Testing at COLD SilUTDOWN is not considered to be a representative test for assessing PORV perfonnance under nonnal 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 SAR 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.

r NORTH ANNA - UNIT I B 3/4 4-2a Amendment No. #9,

^

Z O

c TABLE 2.2-1 ~

d REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS k

Z FUNCTIONAL UNrr TRIP SETPOINT ALLOWABLE VALUES

l. Manual Reactor Trip Not Applicable Not Applicable C

3 2. Power Range, Neutron Flux Low Setpoint - $ 25% of RATED Low Setpoint - s 26% of RATED

[ THERMAL POWER THERMAL POWER High Setpoint- s 109% of RATED High Setpoint - s 110% of RATED THERMAL POWER THERMAL POWER

3. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER 65.5% of RATEDTHERMAL POWER Ifigh Positive Rate with a time constant 2 2 seconds with a time constant 2 2 seconds

4. Power Range, Neutmn Flux, s 5% of RATED THERMAL POWER s5.5% of RATEDTHERMAL POWER High Negative Rate with a time constant 2 2 seconds with a time constant 2 2 seconds g 5. Intermediate Range, Neutmn s 25% of RATED THERMAL POWER s 30% of RATED THERMAL POWER

& Flux

6. Source Range, Neutron Flux s 10 5counts per second s 1.3 x 105counts per second

7. Overtemperature AT See Note 1 See Note 3 -

8. Overpower AT See Note 2 See Note 3

9. Pressurizer Pressure - Low 21870 psig 21860 psig

10. Pressurizer Pressure - High s 2360 psig s 2370 psig i 1. Pressurizer Water Ixvel- High 5 92% ofinstrument span s 93% ofinstrument span

12. Loss of Flow 2 90% of design flow perloop* 2 89% of design flow perloop*

9 E

o.

$ Design flow per loop is one-third of the minimum allowable Reactor Coolant System Total Flow Rate as specified in Table 3.2-1.

E Z

P u

REACTOR COOLANT SYSTEM SAFETY VA.LVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 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 andi 1% as-left.

l APPLICABILITY: 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 required by Specification 4.0.5.

l l

The lift setting pressum shall correspond to ambient condition of the valve at nominal operating temperatum and pmssum.

NORTH ANNA - UNIT 2 3/44-6 Amendment No. 424,

REACTOR COOLANT SYSTEM SAFETY AND RELIEF 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.

APPLICABILITY: MODES 1,2 and 3.

ACTION; With one pressurizer code safety valve inoperable, either mstore 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.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal -

temperature and pressure.

l NORTH ANNA - UNIT 2 3/44-7 Amendment No.

REACTOR COOLANT SYSTEM BASES 3/4.4.2 AND 3/4.4.3 SAFETY AND RELIEF 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 RIIR 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 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 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.

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 + 27c to - 37c. In addition, no single valve is allowed to be outside of 37c.

I An average tolerance of + 27c /- 37c was confirmed to be adequate for Modes 1-3 accident i analyses. For the overpressure events, the analyses considered several combinations of valve l tolerance with the arithmetic average of the three valves' tolerance equal to + 27c (with no valve outside ofi 37c). The case of a + 27c tolerance on each of the three valves provided the most limiting results. The - 37c 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 valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:

NORTil ANNA - UNIT 2 B 3/4 4 2 Amendment No. !21,170,

{

l

3L4A 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 cenain accidents and for plant shutdown.

b) Maintaining the integrity of the reactor coolant pressure boundary. This function is related to contro?. ling 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 Requirernents 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 1 addresses the block valves. The block valves are exempt from the surveillance requirements to l 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 nn inoperable PORV to operable status. j 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 PORY 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. He 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. He maximum water 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 cimulation.

NORTH ANNA - UNIT 2 B 3/4 4-2a Amendment No. MO,

1 n

L Attachment 3 Significant Hazards Consideration i

l l

I I

i l

SIGNIFICANT HAZARDS CONSIDERATIQN Virginia Electric and Power Company requests changes to the following Technical Specifications for North Anna Power Station Units 1 and 2:

. Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, item 10, Pressurizer Pressure - High

. 3.4.2, Reactor Coolant System Safety Valves - Shutdown 3.4.3.1, Reactor Coolant System Safety and Relief Valves - Operating

. 3/4.4.2 AND 3/4.4.3 Bases, Reactor Coolant System Safety and Relief Valves A safety evaluation has been performed which justifies increasing the current Technical Specification pressurizer safety valve (PSV) at-power (Modes 1-3) lift setpoint tolerance from 11% as-found and i1% as-left to +2%/-3% average as-found with no single valve outside 3% as-found and 1% per valve as-left. The as-found value is based on testing, the results of which are expressed as an error (i.e., positive or negative percentage deviation from the nominal lift setpoint). The errors of the tested valves are summed and the result divided by the number of valves tested. This result is compared to the acceptable range of + 2% to -3%. No single valve is allowed be outside of the i3% tolerance.

The safety evaluation also supports an increase to the Hot Shutdown (Mode 4) required PSV lift setpoint tolerance from 1% as-found and 1% as-left to 3% per valve as-found and 1% per valve as-left. These proposed changes will provide greater operational flexibility in meeting periodic test requirements established by the safety analyses.

A concurrent reduction in the pressurizer high pressure reactor trip setpoint and allowable value of TS Table 2.2-1 are also proposed. These changes ensure that the analysis results for the loss of external load accident continue to meet the acceptance criteria with the higher PSV tolerance.

The Loss of Load, Locked Rotor, and Rod Withdrawal event analyses demonstrate that increasing the at-power PSV lift setpoint tolerance to

+ 2%/-3% averaae as-found with no single valve outside 3% as-found and 1% per valve as-left does not result in a transient pressure in excess of the overpressure safety limit. Further, the increased setpoint tolerance does not adversely impact the DNBR results of any North Anna UFSAR Chapter 15 transient analysis. Mode 4 overpressure protection is adequate with one PSV with a tolerance of i3%.

Finally, the increased PSV setpoint tolerances and reduction of the high -(

pressurizer pressure reactor trip setpoint do not present any operational considerations which would significantly impact the performance of the plant during normal operatico or during postulated accident conditions. In summary, each pertinent safety criteria was evaluated for the proposed Technical Specification changes, and all were found to be acceptable.

Virginia Electric and Power Company has reviewed the proposed changes against the criteria of 10 CFR 50.92 and has concluded that the changes as  :

proposed do not pose a significant hazards consideration. Specifically,  !

operation of North Anna power Station in accordance with the proposed Technical Specifications changes will not:

j

1. Involve a significant increase in the probability or consequence of an  ;

accident previously evaluated.  !

Affected safety related parameters were analyzed for a change to North Anna 1 and 2 Technical Specifications 3.4.2 and 3.4.3 and Table 2.2-1  :

item 10. It was determined that the overpressure safety limits would not be exceeded in the most limiting overpressure transients (Loss of Load, ,

Locked Rotor, and Rod Withdrawal events) with the as-found pressurizer i safety valve lift setpoint tolerance increased to an average of + 2%/-3%, j no single valve outside of 3%, and the 25 psi reduction in the Pressurizer High Pressure Reactor Trip setpoint. The DNBR results of transients impacted by the proposed setpoint tolerance increase meet the i

acceptance criterion after accounting for the impact of the proposed  !

changes. The increased setpoint tolerance will not result in an inadvertent opening of the pressurizer safety valves. Mode 4  !

overpressure protection is adequate with one PSV with a tolerance of l 3%.

2. Create the possibility of a new or different kind of accident from any accident previously identified.

The proposed change to North Anna 1 and 2 Technical Specifications 3.4.2 and 3.4.3 and Table 2.2-1 item 10 does not involve any changes )

which would introduce any new or unique operational modes or accident l precursors. Only the allowable tolerance about the existing PSV lift setpoint will be changed, along with a reduction in the pressurizer high 1 pressure reactor trip setpoint.

3. Involve a significant reduction in a margin of safety.

l

. ._. .. .~ _. . . - , .

l It was determined that the rnost limiting overpressure transients do not'  !

result in maximum pressures in excess of the overpressure safety limits. l The DNBR results of ' transients impacted by the proposed setpoint i tolerance increase meet the acceptance criterion after accounting for the ,

impact of the ' proposed changes. Therefore, the margin of safety is unchanged by the proposed- increase in the - safety valve setpoint '

tolerances.

Virginia Electric and Power ' Company concludes that the activities associated with these proposed Technical Specification changes satisfy the no significant hazards consideration of the criteria of 10 CFR 50.92 and, accordingly, a no significant hazards consideration finding is justified. l 1

l t

1 i

e