ML20202G131

Proposed Tech Specs Re Pressure Temp Limits Rept

ML20202G131
Person / Time
Site:	Farley
Issue date:	02/12/1998
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20202G112	List: ML20202G106 ML20202G131
References
NUDOCS 9802200067
Download: ML20202G131 (45)

Text

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Enclosure 2 Revised Technical Specification Pages Unit 1 Page 3/4 4-3 New page, replace Page 3/4 4-4a New page, replace Page 3/4 4-32 New Page, replace Page B 3/4 4-1 New Page, replace 4 Page B 3/4 4-8 New Page, replace page in December 18,1997 submittal Unit 2 Page 3/4 4-3 New page, replace Page 3/4 4-4a New page, replace Page 3/4 4-32 New Page, replace Page B 3/4 4-1 New Page, replace Page B 3/4 4-8 New Page, replace page in Decembr 18,1997 subinittal h!R DO O O O 48 p PDR

Rf. ACTOR COOLANT SYSTEM H3T SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the Reactor Coolant and/or residual heat removal ( RHR) loops listed below shall be OPERABLE:

1. Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump,

2. Reactor Coolant Loop B and its associated steam generator and Reactor Coolant pump,*

3. Reactor Coolan*, Loop C and its associated steam generator and Reactor Coolant pump,

4. Residual Heat Removal Loop A,

5. Residual Heat Removal Lcop B.

b. At least one of the above Reactor Coolant and/or kHR ,

loops shall be in operation.**

APPLICABILITY: MODE 4.

ACTION:

a. With less than the above required Reactor Coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status ab soon as possible,

b. With no Reactor Coolant or RHR loop in operation, suspend all operations invclving a reduction in boron

. concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 325'r l unless 1) the pressurizer water volume is less than 770 cubic feet (24%

of wide range, cold, pressurizer level indication) or 2) the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

• • All Reactcr Coolant pumps and residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided 1) no operations are permitted that x would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10'F below i,

saturation temperature.

EARLEY-UNIT 1 3/4 4-3 AMENEMENT NO.

1

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 a. Two# residual heat removal (RHR) loops shall.be OPERABLE

• and at least one RHR loop shall be in operation.**

APPLICABILITY: MODE 5.NN UNU l

ACTION:

a. With less than the above required RHR/ Reactor Coolant loops OPERABLE, immediately initiate corrective action to return the required RHR/ Reactor coolant loops to OPERABLE status as soon as possible.

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant Lystem and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one RHR loop shall be determined to be in operation and circulating reacter coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The nornal or emergency power source may be inoperable in MODE 5.

The RHR loop may be removed f rom operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided (1) _no operations are permitted that would cause diluuien of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintainea at least 10* F below saturation temperature.

N Three filled Reactor Coolant 1 ops and at least two steam generators having levels greater than or eqial to 10% of wide range indication may i

be substituted for one RHR loop.

NN A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 325*F unless (1) the pressurizer water volume is less than 770 cubia l feet (24% of wide range, cold, pressurizer level indication) or (2) the secondary water temperature of each steam generator is less than 50' F I above each of the Reactor Coolant System _ cold leg temperatures.

NNN The number of operating Reactor Coolant pumps is limited to one at RCS temperatures less than 110'F with the exception that a second pump may be started tot the purpose of maintaining continuous flow while taking the operatit.g pump out of service.

FARLEY-UNIT 1 3/4 4.Aa AMENDMENT No.

PT. ACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.4.10.3 At least one of the following overpressure protection systems shall be OPERABLE:

a. Two RHR relief valves with:

1. A lift setting of less than or equal to 450 psig, and

2. The associated RHR relief valve isolation valves open; or

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.85 square inches.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 325'r, except when the reactor vessel head is l removed.

ACTION:

a. With one RHR relisf valve inoperable, restore the inoperable va?'e to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or perform the following:

1. Establish the following requirements:

1. Reduce pressurizer level to less than or equal to 30 percent (cold calibrated), and

11. Assign a dedicated operator for RCS pressure monitoring and control, and 111. Restore the inoperable valve to OPERABLE status within 7 days, or;

2. Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. With both RHR relief valves inoperable, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:

1. Restore at least one RHR relief valve to OPERABLE status, or

2. Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent.

c. In the event a RHR relief valve or a RCS vent is used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the RHR relief valves or vent on the transient and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

FARLEY-UNIT 1 3/4 4-32 AMENDMENT NO.

f 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all Reactor Coolant Loops in operation, and meet the DNB design criterion during all normal operations and anticipated transients. In MODCS 1 and 2 with one Reactor Coolant Loop not in operation this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, two Reactor Coolant Loops provide sufficient heat removal capability for removing core heat even in the event of a bank withdrawal accidents however, a single Reactor Coolant Loop provides sufficient decay heat removal capacity if a bank withdrawal accident can be prevented; i.e.,

by opening the Reactor Trip Breakers or shutting down the rod drive motor / generator sets.

In MODE 4, a singic reactor coolant or RHR loop provides suf ficient heat removal capability for removing decay b+at, but single failure considerations require that at least two loops be OPERABLE. Thus, if the Reactor Coolant Loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

In MODE 5, single failure considerations require two RHR loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and conttol.

The restrictions on starting a Reactor Coolant Pump with one or more Reactor Coolant System cold legs less than or equal to 325'r are provided l to prevent Reactor coolant system pressure transients, caused by energy additions f rom the secondary system, which could exceed the lindts of Appendix G to 10 CFR Part 50. The Reactor Coolant System will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volu'.e for the primary coolant to expand into, or (2) by restrictin,g starting of the Reactor Coolant Pumps to when the secondary water temperature of er.ch steam generator is less than 50'F above each of the Reactor Coolant dystem cold leg temperatures.

FARLEY-UNIT 1 B 3/4 4-1 AMENDMENT NO.

- _ _ _ _ _ _ _ - __- I

REACTOR COOLANT SYSTEM hASES Values of ART ndt determined in accordance with the NRC-approved methodology l may Le used until the next results from the material surveillance program, evaluated according to ASTM E185-82, are available. Capsules will be removed in accordance with the requi;ements of ASTM E185-82 and 10 CFR $0, Appendix H. The surveillance specimen withdrawal schedule is shown in the PTLR. The heatup and cooldown curves must be recalculated when the ART ndt determined f rom the surveillance capsule exceeds the calculated 6RT ndt ICE the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section XI of the ASME Boiler and Pressure Vessel Code as required by l Appendix G to 10 CFR Part 50 and these methods are discussed in detail in WCAP-14040-NP-A, Revision 2. l l

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided te assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of either RHR relief valve or an RCS vent opening of greater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS legs are less than or equal to 325'F. Either RHR relief valve has adequate relieving l 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 less than or equal to 50*r above the RCS cold leg temperatures provided measures are taken to cushion the overpressure ef fects at RCS temperatures above 250'r, or (2) the start of all operable charging pumps and their injection into a water solid RCS. In the case of the injection by the charging pumps, the analysis is based on the start of the maximum number of operable charging pumps allowed by the Technical Specifications.

FARLEY-UNIT 1 B 3/4 4-8 AMENDMENT NO.

_ . . . _ _ _ _ . . _ _ . .__...___._._-____.____m _ _ _ _ _ . _ , _ _ _ _ _ _ - ,

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the Reactor Coolant and/or residual heat l removal (RHR) loops listed below shall be OPERABLE: l

1. Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump, *

2. Reactor Coolant Loop D and its associated steam generator and Reactor Coolant pump,

• l

3. Reactor Coolant Loop C and its associated steam generator and Reactor Coolant pump, *

4. Residual Heat Removal Loop A,

5. Residual Heat Removal Loop B.

b. At least one of the above Reactor Coolant and/or RHR loops shall be in operation. **

APPLICABILITY: MODE 4.

ACTION:

a. With less than the above required Reactor Coolant and/or i RHR loops OPERABLE, immediately initiate corrective I action to return the required loops to OPERABLE status as soon as possible.

b. With no Reactor Coolant or RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 325'F l unless 1) the pressurizer water volume is less than 770 cubic feet (24%

of wide range, cold, pressurizer level indication) or 2) the secondary water temperature of each steam generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.

• • All Reactor Coolant pumps and residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10'F below saturation temperature.

FARLEY-UNIT 2 3/4 4-3 AMENDMENT NO.

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 a. Two# residual heat removal (RHR) loops shall be OPERABLE

• and at least one RHR loop shall be in operation. **

APPLICABILITY: MODE 5.8# #NN l ACTION:

a. With less than the above required RHR/ Reactor Coolant loops OPERABLE, immediately initiate corrective action to return the required RHR/ Reactor coolant loops to OPERABLE status as soon as possible.

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The normal or emergency power source may be inoperable in MODE 5.

The RHR loop may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

1. Three filled Reactor Coolant loops and at least two steam generators having levels greater than or equal to 10% of wide range indication may be substituted for one RHR loop.

1. 1. A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 325'F unless (1) the pressurizer water volume is less than 770 cubic l feet (24% of wide range, cold, pressurizer level indication) or (2) the secondary water temperature of each steam generator is less than 50'r above each of the Reactor Coolant System cold leg temperatures.

NN# The number of operating Reactor Coolant pumps is limited to one at RCS temperatures-less than 110*F with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

FARLEY-UNIT 2 3/4 4-4a AMENDMENT NO.

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATIONS ,

3.4.10.3 At least one of the following overpressure protection systems shall be OPERABLE:

a. Two RHR relief valves with:

1, A lift setting of less than or equal to 450 psig, and

2. The associated RHR relief valve isolation valves opens or

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.85 square inches.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 325'r, except when the reactor vessel head is l removed.

ACTION:

a. With one RHR relief valve inoperable, restore the inoperable valve to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or perform the following:

1. Establish the following requirements:

1. Reduce pressurizer level to less than or equal to 30 percent (cold calibrated), and

11. Assign a dedicated operator for RCS pressure monitoring and control, and 111. Restore the inoperable valve to OPERABLE status within 7 days, or;

2. Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. With both RHR relief valves inoperable, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:

1. Restore at least one RHR relief valve to OPERABLE status, or

2. Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent.

c. In the event a RHR relief valve or a RCS vent is used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the ef fect of the RHR relief valves or vent on the transient and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

FARLEY-UNIT 2 3/4 4-32 AMENDMENT NO.

3/4.4 REACTOR COOLANT SYSTM BASES 3/4.4.1 REACN R COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all Reactor Coolant Loops in operation, and meet the DNB design criterion during all normal operations and anticipated transients. In MODES I and 2 with one Reactor Coolant Loop not in operation this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, two Reactor Coolant Loops provide sufficient heat removal capability for removing core heat even in the event of a bank withdrawal accidents however, a single Reactor Coolant Loop provides sufficient decay heat removal capacity if a bank withdrawal accident can be prevented; i.e.,

by opening the Reactor Trip Breakers or shutting down the rod drive motor / generator sets.

In MODE 4, a single reactor coolant or RHR loop provides suf ficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be OPERABLE. Thus, if the Reactor Coolant Loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

In MODE 5, single failure considerations require two RHR loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boren concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more Reactor Coolant System cold legs less than or equal to 325'r are provided l to prevent Reactor Coolant System pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The Reactor coolant System will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.

EARLEY-UNIT 2 B 3/4 4-1 AMENDMENT NO.

i l

REACTOR COOI. ANT SYSTEM BASES Values of ART ndt determined in accordance with the NRC-approved methodology, l

may be used until the next results from the material surveillance program, evaluated according to ASTM E185-L", are available. Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the PTLR. The heatup and cooldown curves must be recalculated when the ART ndt determined from the next surveillance capsule exceeds the calculated ART ndt for the equivalent capsule radiation exposure.

I Allowable pressure-temperature relationships for various heatup and cooldown i rates are calculated using methods derived from Appendix G in Section XI of l the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR 50 and these methods are discussed in detail in WCAP-14040-NP-A, Revision 2.

I Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of either RHR relief valve or an RCS vent opening of greater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 325'F. l Either RHR relief valve has adequate relieving capability to protect the RCS f rom overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50'r above the RCS cold leg temperatures provided measures are taken to cushion the overpressure effects at RCS temperatures above 250'F, or (2) the start of all operable charging pumps and their injection into a water solid RCS. In the case of the injection by the charging pumps, the analysis is based on the start of the maximum number of operable charging pumps allowed by the Technical Specifications.

FARLEY-UNIT 2 B 3/4 4-8 AMENDMENT NO.

9 Pen and ink Technical Specification Page Markups

REACTOR C0OLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the Reactor Coolant and/or residual heat removal (RHR) loops listed below shall be OPERABLE:

1. Reactor Coolant Loop A and its associated st aa generator and Reactor Coolant pump,"

2. Reactor Coolant Loop 8 and its associated steam generator and Reactor Coolant pump,"

3. Reactor Coolant Loop C and its associated steam generator and Reactor Coolant pump,"

4. Residual Heat Removal Loop A,

5. Residual Heat Removal Loop 8.

b. At least one of the above Reactor Coolant and/or RHR loops shall be in operation.**

APPLICABILIT/: MODE 4.

ACTION:

a. With less than the above required Reactor Coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b. With no Reactor Coolant or RHR loop in operation', suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

325 "A Reactor Coolant pump shall not be started with one or morekf the Reactor Coolant System cold leg temperatures less than or equal to +1FF t.nless 1) the pressurizer water volume is less than 770 cubic feet (24% of wide range, cold, pressurizer level indication) or 2) the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

• • All Reactor Coolant pumps and residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature.

7-r7 3/4 4 4 ,

t3 ~+ f.Jl4 4'-2. f 4 b FARLEY-UNIT 1 , 3/4 4-3 AMENDMENT NO. 26

REACTOR COOLANT 3YSTEM COLD SHUTDOWN, LIMITING CONDITION FOR OPERATION 3.4.1.4 a. Two# residual heat removal (RHR) loops shall be OPERABLE

• and at least one RHR loop shall be in operation.**

APPLICABILITY: HODE 5." MM4 ACTION:

a. With less than the above required RHR/ Reactor Coolant loops OPERABLE, immediately initiate corrective action to return the required RHR/ Reactor coolant loops to OPERABLE status as soon as porsible.

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. .

• The normal or emergency power source may be inoperable in MODE 5.

• • The RHR loop may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

1. Three filled Reactor Coolant loops and at least two steam generators having levels greater than or equal to 10% of wide range indication may be substituted for one RHR loop.

32 5 A Reactor Coolant pump shall not be started with one or m e of the Reactor Coolant System cold leg temperatures less than or equal to S19'F unless 4 (1) the pressurizer water volume is less than 770 cubic feet (24% of wide range, ccid, pressurizer level indication) or (2) the secondary water tem-perature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures, j /Abd'6'

1. 1. 1. The number of operating Reactor Coolant pumps is limited to one at RCS temperatures less than 110'F FARLEY-UNIT 1 3/4 4-44 with the exception thai a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service

' ~

REACTOR COOLANT SYSTEM DVERPREssuRE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATIONS 3.4.10.3 At least one of 3.e following overpressure protection systems shall be OPERA 8LE:

a. Two RHR relief valves with:

1.

A lift setting of less than or equal to 450 psig, and

2. The associated RHR relief valve isolation valves open; or b.

The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.85 square inches.

APPLICARILITY:

less than or equal to When the temperature of one or more of the RCS cold legs is F, except when the reactor vessel head is removed.

ACT10M1 326

a. With one RHR relief valve inoperable, re. core the inoperable valve to CPERA8LE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or perfdth the following:

1. Establish the following requirements:

1. Reduce pressurizer level to less than or equal to 30 percent (coldcalibrated),and

11. kssignadedicatedoperatorforRCSpressuremonitoring and control, and . .

iii. Restore the inoperable valve to 0PERABLE status.within 7 days, or;

2. Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,

b. With both 15R relief valves inoperable, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:

1 Restere at least or.e RHR relief valve to OPERABLE status, or

2. Depressurire and vent the RCS through a greater than or equal to 2.85 square inch vent. ,

c. In the event a RHR relief valve or a RCS vent is used to mitigate a RCS pressure transt:nt, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 8.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the RHR relief valves or vent on the transient and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

, $ - -p f. 3l& H O FARLEY-UNIT 1 AMDEMENT NO. 26,108 3/4 4-32

3/a.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND C00LAFf CIRCULATION The plant operation. and is designed to operate with all Reactor Coolant Loops in meet anticipated transients.the DNB design criterion during all normal operations and ( i operation this specification requires thatIn MODES the plant 1beand in at2 least vith one HOTReactor STANDBY Coolant Lool vithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. I In MODE 3, two Reactor Coolant Loops provide suf ficient heat reseval capability for removing core heat even in the event of a bank vithdraval l accidents however, a single Reactor Coolant Loop provides sufficient decay heat removal capacity if a bank vithdraval accident can be prevented: 1.e., by opening the heactor Trip Breakers or shutting hvn the rod drive actor / generator sets.

l In F.0DE 4 a single reactor coolant or PER loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least tvo loops be OPERABLE. Thus, if the Reac';or Coolant Loops are not OPERABLE. this specification requires two RHR loops to be OPERABLE.

In MODE 5. single failure considerations require two RER loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one RER pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction vill, therefore, be within the capability of operator recognition and contro k CpC M The restrictions on starting a Reactor Coolant Pump /vith one or more Reactor Coolant Systes cold legs less than or equal to 3te*F are provided te prevent Reactor Coolant Systes pressure transients, caused by energy additions from the secondary systes, which could exceed the limits of Appendix G to 10 CFR Part 50. The Reactor Coolant Systen vill be protected against overpressure i

transients and vill not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the l

Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

l 15/

r m zr - unrT 1 5 3/4 4 1 AMENDMENT NO. 26, ii.

~

/_--...

REACTOR COOLANT SYSTEM ~ /b Odd *4@O#fM k!O N BASE 5 kVE[LV.Akh WIN A-U[$k_l.(l,[Wl.*&l ^s" -

.................................... .~.........................................~

\

' ----- may be used until the next f

Values results of ST,$e, from t determined in material surveillance program, evaluated according to I i

ASTM E185-82, are available. Capsules vill be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix 8. The surveillance specimen withdrawal schedule is shovn in !!!". 5:::!= 5.'

The heatup and cooldown curves must be recalculated when the RT , s determined f rom the surveillance capsule exceeds the calculated Mt"'

-for the equ valent capsule radiation exposure.

y fe FTL2 Allovable sure-temperature relationships for various heatup an F cooldownyates are calculated using methods derived from Appendix G in Section in of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 1 nd h Qods are discussed in detail in

'. 'Cl. " ~'. "..+ - C .G.

N A. 4/6tCd 7.7

• Th: .;;;ni method-for-ealcula tifig-heatup-end 00:13:r. 11:i: :::r:: !:

beeed 'gaa -the-pr4nc.1ples-of.-tche-. Linear--elas444-tree:ur . ::9r!:: ' !" )

• -"^'^ y I" e-calculatrion proceduses-a-semi-etkitt4e:1 erf r:

deft : ri %e-dep4h-of-ene-quar 4er-of-4he-wall-th4c"=:, T, d : !=;f

f 1/27-6s-assumed :c exte+-et-the-insid:-cf :h: r::::1 ::11 = r:11 =

:h: : tside-of-the-vessel-vall. Th: 44eenelon: ef e.!: ;::: 12::e tr e9 referred te != Append 4*-C-of AS".E 5 ::10: !!! : th: ::lesenee-f1:r, r;17 :::::d th: ::::::: ::peb!!!:ie: Of in:::vi:: in:p:::i:n teehniques. Ther:f=:, :h: ::::::: :pentien 1! i: reu e der:1e; d fer

- e'e nte: : rh n: = = =v::1:: =d p:::id: :::!ici::: nf::7 er;in: fer-pr+te+&&en-egeln+4-nen de::ile f eilr:. T: n:=: th : the

=dicti:n :ebr4++1eeen: :f f:::s-eee-ee::::::d != in th: =1::12:in f the-14mi4-eueves r -the-nost-.ktai+ tag-vele: ef th: all due:111:y nf= = r

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--FARLEY UNIT-1-REACTOR-VESSEL--TOUGHNE45-PROPERT4ES-t E

a Material Cu P Ni Tndt RTndt upper i Energy j Component Code No. Type (1) (%) (1) (*F) (*F) [C] NMW[d]

i --

Closure head dome 0690 A533 B.C1.1 0.16 0.009 0.50 -30 -20[a] 140 -

{

Closure head segment B6902-1 A533,B,C1.1 0.17 0.007 0.52 -20 -20[8] 138 - I j Closure head flange B6915-1 SQ8, C1.2 0.10 0.012 0.64 8] 60[a] 75[aj _

Yessel flange 06913-1 C 0.17 0.011 0.69 60[a] 60[a] 106[8] -

Inlet nozzle B6917-1 A50hyC (1.2 A508, .2 -

0.010 0. 60[8] 60[a] -

110 Inlet nozzle 0.008 .80 B5917-2 A508 C1.'2 -

60[a] 60[a] -

80 t

. o, Inlet nozzle 06917-3 A508 C1.2 -

0.00 0.87 60[a] 60[aj _

93 [

u Outlet nozzle 06916-1 A508, C1.2 - 0,007 0.77 60[8] 60[a] -

96.5 2 Outlet nozzle B6916-2 A508 C1.2 -

.011 0.78 60[a] 60[8] -

97.5  !

, Outlet nozzle 06916-3 A508 C1.2 -

0 009 0.78 60[a] 60[8] -

100 i j, Nozzie shell 86914-1 A508 C1.2 0% 0.68 30 30[aj 343 _

Inter. shell B6903-2 A533,B C1.1 0.13 0.01 0.60 0 0 151.5 97 Inter. shell B6903-3 A533,B,0111 0.12 0.014 0.56 10 10 134.5 100 tower shell B6919-1 A53),87C1.1 0.14 0.015 035 -

15 133 9C.5 tower shell B6919-2 Sf33,B,C1.1 0.14 0.015 0.56 \ 20 10 5 134 97 [

Bottom head ring B6912-1 A508, C1.2 -

0.010 0.72 ' 1.0 10[a] 163.5 -

j Bottom head segment B690 A533,B,C1.1 0.15 0.011 0.52 -30 -30[a] ;47 ,

j 00ttom head dome 6967-1 A533,B,C1.1 0.17 0.014 0.60 -30 3 143.5 - I j E Inter. shell long. M1.33 Sub Arc Weld 0.25 0.017 0.21 0[a] \sg)0[8] [aj _ _  !

! 9 weld seam -  !

i 3 Inter. to lowe Gl.18 Sub Arc Weld 0.22 0.011 <0.20[b] 0[a] 0[8] - -  !

i 9 shell weld ams i

[ Lower sh I long. 0[aj Gl.08 Sub Arc Weld 0.17 0.022 <0.20[b] 0[a] _ _ j

,6 weld seams  !

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FARLEY - UNIT 1 B 3/4 4-10 A AMENDMENT N

REACTOR COOLANT SYSTEM .-

BASE 3 e-MdW -

/

The ASME approach for calculating the allowable Ifmit curves for var us he up and cooldown rates specifies that the total stress intensity fa or, K,g r the combined the; mal and pressure stresses at any time durin heatup or cool wn cannot be greater than the reference stress intensity actor, Kgg, for the e al temperature at that t'me. K is obtained from t reference IR fracture to hness curve, defined in Appendix G to the ASME C e. The K yg curve is given y the equation:

Kgg = 26.78 + .223 exp [0.0145(T-RTNOT + 160)] (1) ,

where K IR is the refere e stress intensity factor as a function of the metal temperature T and the met nil ductility refer ice temperature RT NOT. Thus, the governing equation for t e heatup-cooldow analysis is defined in Appendix G of the ASME Code as follows:

CKgg + Kgg < Kgg (2)

Where, K gg is the stress'intrnsi caused t,y membrane (pressure) a fact stress.

/

K gg isthestressjntensityfactorcausedbythethermalgradients.

K IR is provide by the code as a function of temperature relative to the RT H of the material. ,

C = 2.0 for level A and B service limits, and C= .5 for inservice hydrostatic and leak test operat ns.

A y time durirg the beautp or cooldown transient K IR is determined by

• the tal temperature at the tip of the postulat'ed flaw, the appro ate l v ue for RTNOT and the reference fracture toughness curve. The the al '

stresses resulting from temperature grad'ients through the vessel wall are

---,e--- - - - r --

y-wi,-#- _m y e.

REACTOR COOLANT SYSTEM BASES H4\4 Phk inh $1&fLLY Lhi-T 6LALN,)

~

alculated and then the corresponding thernal stress inten3ity factor, K ,

fo the reference flaw is computed. From Equation (2) the pressure str ss inten are calcul(ty at d.

factors are obtained and from these, the allowable nressur COOLDOWN For the calculation of the allowable pressure versus colant temperature st at the inside of during the vessel cooldown, wall. Laur thg (Code reference ng cooldown, the controllingflawloc is ton assumed of the flaw to ise always at the inside o the wall because the thermal radients produce tensile stresses at the inside, ich increase with increas g cooldown rates. Allowable pressure-temperature rela ons are generated for th steady-state and finite cooldown rate situations, om these relations omposite limit curves are constructed for each cooldown ate of interest The use of the composite cur in th cooldown analysis is necessary because control of the cooldown proc u e is based on measurement of reactor -

coolant temperature, whereas the limi pressure is actually dependent on the material temperature at the tip of th assumed flaw. During cooldown, the 1/4T vessei Incation is at a hi her temperat e than the fluid adjacent to the vessel ID. This condition, o,f course, is not ue for the steady-state situation.

It follows that at any giv, arf reactor coolant temp (ature, the delta T developed during cooldown results 6 a' higher value of K at V e 1/4T location for IR finite cooldown rates an for steady-state operation. Furthermore, if conditions exist su that the increase in K exceeds K gg the calculated IR allowable pressur during cooldown will be greater than the teady-state value.

The pbove procedures are needed because there is no direct cent ol on tempera)cre at the 1/4T location; therefore, allowable pressures may unknoyingly be violated if the rate of cooling is decreased at various int vals along a cooldown ramp. The use of the composite curve eliminat t problem and assures conservative operation of the system for the enti c 1down period. .

FARLEY-UNIT 1 B 3/4 4-12 AMENOMENTNO.h

REACTOR COOLANT SYSTEM

. ._ __"w BASES m ~.

NOA-@ LW1%

HEATUp NThree separate calculations are required to determine the linil, curve for finite heatup rates. As is done in the cooldown tnalysis, allowable pressure-teeperature relationships are developed for steady-state cond tions as well as finite heatup rate conditions assuming the presence of a /4T defect at the 'inside of the vessel wall. The thermal gradients d ring heatup produce compress'ive stnsses at the inside of the wall that al)eviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the co'o(ant toe,arature; therefore, the Kgg!orthe1/4Tcrack f

during heatup is lower'than the E gg for die 1/4T cnc uring steady-state conditions at the same cobt tamperature. During estup, especially at the end of the transient, condit\ ns may exist such t the effects of compressive thermal stresses and differen Kg 's for steadyy tate and finite heatup rat'es

• do,.not offsst each other and the assure-teep'erature curve based on steady-state conditions no longer repnsent a lowe/r bound of all similar curves for finite heatup rates when the 1/4T flaw s /considend. Therefore, both cases have to be analyzed in order to assure'tt t at any coolant temperature the lower value of the allowable press,ure calcu ted for steady-stata and finite heatup rates is obtained. /

/

The second portion of the heatup analysis co erns the calculation of pressure-temperature limitations for the case in whih a 1/4T deep outside surface flaw is assumed / Unlike the situation at the Vessel inside surface, the thermal gradients,e'stablished at the outside surfac during heatup produce ,

stresses which are tensile in nature and thus temi to ret orce .any pressure stresses present. f These thermal stresses, of course, are ndent on both the rate of heatup and the time (or coolant temperature) alo the heatup ramp. Furthermore since the thermal stmsses, at the outside re tuns 11e and increase with ffncre,asing heatg rate, e lower bound curve canno e defined.

Rather, ea heatup rate of interest must be analyzed on M indi ual basis.

Follo' wing the generation of pressure-temperature curves for be the steady-5f. ate and fialte heatup rate situations, the final Itait curves re produce,d as follows. A composite curve is constructed based on a point-poi ( comparison of the steady state and finite heatup rate data. At any gi en temperatun, the allowable pressure is taken to be the lesser of the .

ree values taken from the curves under consideration.

FARLEY-UNIT 1 8 3/4 4-13 ,

W W MT MO 8

REACTOR COOLANT SYSTD4 ,

BASES The-use-of-the coepes4te-c.urve is-necessary te tet conserv+t4v: h::te 14mitat4cns-because 4t-ts-

-course-of--the hertup-camp- possible for.-condit4cns-to-ext-st-suc' '

" p-th: the-controlling-condit.ian-switches f.com-the insid ::

-t he-mest-c r% ital-**4-t : r i:r. .-c e t : t de-a nd-th e-p re s su r e-14mi t-au s t-a t-a ti-t F13:

11y, the 10 CFR-part 50c Append 4x-G-4.u14-which-addresses the et:1 tr r: tere -ef -the closur-e-head f44nge-and-vesset flange-must- be cons 44eeed.

Th+s-Rele-St a t es-tha t-t he-min imum-me tal-t empera ture-of-the-ele s ure-fl ange 7:

th: sten 5: :t least 4204-h4gher thea the 'imiW "% 'e- these mien: eh:r, pressure er.:::ds-20-percent-of-4he-preserv4ce-hydrostat4c-test-pressere-(624-pelg-for-Far4ey-Unit.4)-In-addition,- the-new.40-CFR #4r-t-50 ".el: :t:t::

that-a-p14nt--spec 4f4c-fracture-evaluation-may-boyer4cr=ad-te-je:tify 1:::

14*tt4ng F:rity r:qu+rements,---As-a-resul-tr-such-a fracture-analys4s-was-perf4d Sr Unit--2. These-Far4e F6May-Unit 1 :ince the-pery--Unit-2-fr-acture-analysts-results-are-appt4c:51: t Y"

DdC*

t4nent-parameters-are tdantica! 'er beth ht:d upon-th45-f-racture-anal-ys4sr-the-16--EFP4-heatup-and-co*1de= c?!:ett. ury:: :r:

t ;;;ted by the :: 10 Cr" Part 50 ", ele-a: :h n ce -

F!gur:: 3d 2 :p 2.t-3.

=

Although the pressurizar operates in tesperature ranges above those for there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements, yr The OPERA 81LITY of either RHR relief valve or an RCS vent opening ofater g than or equal to 2.85 square inches ensures that the RCS will be prot ted froe l pressdre transients which could exceed the limits of Appendix G to 1 CFRPart{

50 when one or more of the RCS cold less are less than or equal to F.

Either RHR relief valve has adequate relieving capability to protect the RCS

\, from overpressurization when the transient is limited to either (1) the start f{

of an idle RCP with the secondary water temperature of the steam generator less I than or equal to 50'F above the RCS cold leg temperatures provided measures are k

\ taken to cushion the overpressure effects at RCS temperatures above 250*F, or (2) the start of- <

and their injection into a water solid RCS.

)

[7(Til sTRucTUnn iwTrnairy -

IN .

The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operctional readiness of these of the components plant. will be maintained at an acceptable level throughout the life These programs are in accordance with Section II of the ASME 8ciler and Part 50.55a(g Pressure Vessel Code and applicable Addenda as required by 10 CFR Ceemission pur)suant to 10 CFR Part 50.55a(g)(6)(1).except where spec

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FARLEY-UNIT 1 B 3/4 4-14 AMDOMENT NO. 47,7h

)JCre', ~T'H: FAdt.6,bM lllLLdCL-D IL3 Y4 8MN /'II'W N "Io %e 6 ?/4 4-d RC CC@ddOr# 0: M S 3/4 iO.

4 9 9

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REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 4.1. 3 a. At least two of the Reactor- Coolant and/or residual heat removal (RHR) loops listad below shall be OPERABLE:

1. Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump,"

2. Reactor Coolant Loop 8 and its associated steam generator and Reactor Coolant pump,"

3. Reactor Coolant Loop C and its associated steam generator and Reactor Coolant pump,*

4 Residual Heat Removal loop A.

5. Residual Heat Removal Loop 8.

b. At least one of the above Reactor Coolant and/or RHR loops shall be in operation.**

APPLICA8ILITY: MODE 4. .

ACT, ION:

a. With less than the above required Reactor Coolant and/or RHR loops OPERA 8LE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b. With no Reactor Coolant or RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate correctiver action to

, =. return the required coolant loop to operation.

GAf "A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to-Me*F unless 1) the pressurizer water volume is less than 770 cubic feet (24,% of wide range, cold, pressurizer level indication) or 2) the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

• • All Reactor Coolant pumps an( residual heat removal pumps may be de-energized for up to Z hours provided 1) no operations are pemitted that would cause dilution of the Reactor Coolant System baron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature. .'

FARLEY-UNIT 2 3/4 4-3 i l

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 a. Two# residual heat removal (RHR) loops shall be OPERABLE

• and at least one RHR loop shall be in operation.**

APPLICABILITY: MODE 5. "

ACTION:

a. With less than the above required RHR/ Reactor Coolant loops OPERABLE, immediately initiate corrective action to return the required RHR/ Reactor coolant loops to OPERABLE status as soon as possible.

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one RHR loop shall be determined to be in operation and l circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

~ "The normal or emergency power source may be inoperable in MODE 5.

l

• • The RHR loop may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is m'intained a at

least 10'F below saturation temperature.

l Three filled Reactor Coolant loops and at least two steam generators having levels greater than or equal to 10% of wide range indication may be substituted for one RHR loop.

A Reactor Coolant pump shall not be started with one or morekf we Reactor Coolant System cold leg temperatures less than or equal to -Meef un' ss (1) the pressurizer water volume is less than 770 cubic feet (24% o. wide range, cold, pressurizer level indication) or (2) the secondary water tes-perature of each steam generator is iets than 50*F above each of the Reactor Coolant System cold leg temperatures.

N J SA&E8 Y ~

1. 1. 1. The number of operating Reactor Coolant pumps is limited to one at RCS temperatures less than 110*F FARLEY-UNIT 2 3/4 4-4a with the exception thaf a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

REACTOR COOLANT SYSTEM

~

OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATIONS 3.4.10.3 At least one of the following overpressure protection systems shall be OPERABLE:

a. Two RHR relief valves with:

1.

A lift setting of less than or equal to 450 psig, and 2.

The associated RHR relief valve isolation valves open; or

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.85 square inches.

APPLICABILITY

• less than or equal toWhen the temperature of one or more of the RCS cold legs is F, except when the reactor vessel head is removed.

ACTIONt U26

a. With one RHR relief valve inoperable, restore the inoperable valve to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or perform the following: -

1. Establish the following requirements:

1. Reduce pressurizer level to less than or equal to 30 percent (coldcalibrated),and

11. hssignadedicatedoperatorforRCSpressuremonitoring and control, and 111. Restore the inoperabic valve to OPERABLE status within 7 days, or;

2. Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. 3

b. With both RHR relief valves inoperable, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:

1. Restore at least one RHR relief valve to OPERABLE status, or

2. Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent.

c. In the event a RHR relief valve or a RCS vent is used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating t.te transient, the effect of the RHR relief valves or vent on the transient and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0/4 are not applicable.

l FARLEY-UNIT 2 ,

3/4 4-32 AMENOMENT NO.100

3/4.4 REACTOR C00LAffT SYSTEM BASES 3/4.4.1 REACTOR C00LAfff LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all Reactor Coolant Loops in operation, and meet the DNB design criterion during all normal operations and anticipated transients. In MODES 1 and 2 vith one Reactor Coolant Loop not in l operation vithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.this specification requires that the plant be in at least HOT STANDBY In H0DE 3, two Reactor Coolant Loops provide sufficient heat removal capability for removing core heat even in the event of a bank withdrawal l accidents however, a single Reactor Coolant Loop provides sufficient decay heat removal capacity if a bank withdrawal accident can be prevented: 1.e., by opening sets. the Reactor Trip Breakers or shutting dovn the rod drive motor / generator l

In MODE 4, a single reactor coolant or RER loop provides sufficient heat removal capsbility for removing decay heat, but single failure censiderations require that at least two loops be OPERABLE.

are not OPERABLE, this specification requires two RER loops to be OPERABLE.Thus, if the In MODE 5, single failurt considerations require two RER loops to be OPERABLE.

flov The operation to ensure of one Reactor Coolant Pump or one RER pump provides adequate mixing changes during boron c,oncentration reductions in the Reactor Coolant System. prevent The reactivity change rate associated with boron reduction vill, therefore, be within the capability of operator recognition and conte The restrictions on starting a Reactor Coolant Pumpyvith one or more Reactor Coolant System cold legs less than or equal to 4Fte*F are provided to prevent Reactor Coolant Systes pressure transients, caused by energy additions from 50.

Part the secondary system, which could exceed the limits of Appcndix G to 10 CFR The Reactor Coolant Systen vill be protected against overpressure transients and vill not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator temperatures. is less than 50'F above each of the Reactor Coolant System cold leg b

FARLEY - UNIT 2 B 3/4 4-1 AMENDMENT NO. II. 85

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• t'e recuirements of A$7t! E185-82 anc 10 CFA 50 Ascencia fee et. Capsules j elll :e re Surveillance specimen withdrawa t senecule is shown in - u n'; '

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ngg tne t.Rf

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tetermines fecm the next surveillance capsule esceeds the calcu'ated 14 7 #

ndt 3r *e. ecuf valent capsule radiatten esposure.

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.. :2. : I,aISJ nw :. ..-aa I.3. .'I *t FARI.EY. UNIT 2 B3/4 4-9 AMENDMINT NO -

REACTOR COOLANT SY3 TEM _-

BASES

& % M n c & u.) L2 =r S l u k. y

\ The ASME approach for calculating the allowable limit curves for vario heatu and cooldown rates specifies that the total stress intensity facto ,

Kg , for the combined thermal and pressure stresses at any time during atup or coold'own cannot be greater than the reference stress intensity fa or, M 3, for the metal temperature at that time. K IR is obtained from the eference fracture toughness curve, defined in Appendix G to the ASME Cod . The K IR curve is given by the equation:

Kgg = 26.78 + 1 N 3 exp [0.0145(T-RTNDT + 160)] (1) where K IR is the referen stress intensity factor ., a function of the metal temperature T and the metal il ductility refere e temperature RT Thus, NDT.

the governing couation for th heatup-cooldown nalysis is defined in Appendix G of the ASME Cv e es-follows:

CKyg + kit 5Kgg (2)

Wh t e. L g is the stress intensity act caused by membrane (pressure) dress,

\

K It is the stress intensity factor cause}hby the thermal gradients.

K IR is provided y the code as 'a function of emperature relative

o the RT NOT f the material.

C = 2.0 or level A and B service limits, and C=1 for inservice hydrostatic and leak test operati s.

At/ny time during the heautp or cooldown transient, KIR is d ermined by the metal temperature at the tip of the postulated flaw, the 'appropt te val e for RTHDT, and ti.e reference fracture toughness curve. The therm 1 resses resulting from temperature gradients through the vessel wall are FARLEY-UNIT 2 B 3/4 4-11 AM6dMhdT' do.

REACTOR COOLANT SYSTEM BASES s

/%.s th uJrdnoJM

= - -

-- g ca culated .ind then the corresponding thermal stress intensity factor, K for reference flaw is computed. From Equation (2) the pressure str ss intens1 factors are obtained and from these, the allowable pressur are calculated.

N COOLDOWN \

For the cal tion of the allowable pressure versus colant temperature during cooldown, th(encode reference flaw is assumed to e 1st at the inside of the vessel wall. Durlqg cooldown, the controlling loc, ion of the flaw is always at the inside of\the wall because the thermal radients produce tensile stresses at the inside, Aich increase with increas' g cooldown rates. Allowable pressure-temperature relations tre generated for th steady-state and finite cooldown rate situations. F'rgm these relations omposite limit curves are constructed for each cooldowniate of interast i N

, , The use of the composite curve in th ooldown analysis is necessary be-cause control of the cooldown proce re s s based on measurement of recctor -

coolant temperature, whereas the limi ing pressure is actually dependent on I -

the material temperature at the ti' ofth)sassumedflaw. Durino cooldown, the 1/4T vessel location is at a hiyher temperatQre vessel ID.

s than the fluid adjacent to the Thiscondition,ofcourse,isnotSrueforthesteady-statesituation.

It follows that at any givedeactor coolant temhirature, the delta T developed during cooldown results jn'a higher value of K IR at the 1/4T location for finite cooldown rates ,than for steady-state operatf or:. Furthermore, if conditions exist su that the increase in K cxceeds K IR It the calculated allowable pressur during cooldown will be greater than the teady-state value. ,

The 3bove procedures are needed 'oecause there is no direct con gol on unkno n' gly at temperature the 1/4T location; therefore, allowable pressures may be violated if the rate of cooling is decreased at various inte thi is along a cooldown ramp. The use of the composite curve elimina qs problem and assures conservative operation of the system for the entirq oldown period.

\

6 FARLEY-UNIT 2 B 3/4 4-12 AW: G& T b-

REACTOR COOLANT SYSTEM m me- .

BASES -

HEATUP ree separate calculations are required to determine the limit cur es forfi(teheatuprates. Asisdoneinthecooldownanalysis,allowape pressure-temperature relationships are developed for steady-state e ditions as well as f(nite heatup rate conditions assuming the presence of a 1/4T defect at the 'iqside of the vessel wall. The thermal gradient, during heatup produce compressi stresses at the inside of the wall that 'lleviate the tensile stresses produced by internal pressure. The meta temperature at the crack tip lags tha co'olant temperature; therefore, the IR for the 1/4T crack during heatup is lower'than the K g for the 1/4T crack during steady-state conditions at the same coo nt temperature. Durin heatup, especially at the end of the transient, condi pns may exist suchf at the effects of compressive thermal stresses and different g's for steapy-state and finite heatup rates do not offset each other and the pressure-t,emperature curve based on steady-state conditions no longer represen'ts a pwer bound of all similar curves for finite heatup rates when the 1/4T f1 is considered. Therefore, beth cases.

have to be analyzed in order to assu t(at at any coolant temperature the lower value of the allowable pres [re calch ated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis co erns the calculation of pressure-temperature limit,ations for the case in w h a 1/4T deep'outside surface flaw is assumed / Unlike the situation at th essel inside surface, thethermalgradients/stablishedattheoutsidesurf during heatup produce stresses which are t,e'nsile in nature and thus tend to r nforce any pressure stresses present. hese thermal stresses, of course, are ependent on both the rate of hea and the time (or coolant temperature) a g the heatup ramp. Furthe e since the thermal stresses, at the outsi are tensile and increase with dncre,asing heatup rate, a lower bound curve cann t be defined.

Rather, eacty'heatup rate of interest must be analyzed on an ind idual basu Foi o ing the generation of pressure-temperature curves for b h the steady ate and finite heatup rate situations, the final limit curv are produ d as follows. A composite curve is constructed based on a pain -by-poi comparison of the steady-state and finite heatup rate data. At a gi en temperature, the allowable pressure is taken to be the lesser of th(

ree values taken from the curves under consideration. A FARLEY-UNIT 2 B 3/4 4-13 AM6k O,

REACTOR COOLANT SYSTEM

• BASES P

l The-use-of-the-composite-cueve. is-necessary te-sat caaremtive 5::tu-14mi-te64cas-because-4t-ts-possible for-scondit. tons-to--exist such-.that-over-the course of the heatup-ramp-the-control.l.ing conditton-switches fros-the-hside-to- '

the-outs 4de-and-the-pressure-.1 tatt-must-at -al.l.-times -be based-on .analys4e-d th: =:t critica! cr4.tep4em.

re- 11o +w. in ren n.. cn ano..ai, e on1. wk t ,.h oda ~.... +6- ..i tempe/eture-o(-the-closuredea"d 4.1angs-.and ves?ai d=5a= -fit 55E5E55d::d.

B+s-Rele-States-that-the-minimum-seta1-temperature-of- the ele:ure :rg:

re9 ton: 5: Obleast4208F-h4gher--th:n the 1!eitiaa 8 % fer t'e:c regten: -5::

the-pressure-exceeds-20-percent ef the pate~4ce hydra?+ etic test ; eesure .

4421-pstg for-F-ar44y-Unit-2-)_  != edditiony the-new-10-CFA-Pe-t -M Rel: :t:t::

-thet-e-plant-specif t: fracture-evaluette e y -be-perfe---d te jectify 1 ::

44*tting r:quirements.--Base 4-upon-such-a--fracture-analys45-for-Fer4ey-Un44-Er

$ ds. the44-EFP-Y-heatup-and-cooldown-curves-are-impacted-by-the-rew-M-GFR-P:-t 50 kg] Rule:: he= en Figur:: 3.4-2 :nd 3.?-3.

--- ~~ ~--~ < % _ , - - - - - ,

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile-failure, operating lletts are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of either RHR relief valve or an RCS vent opening of reater than or equal to 2.85 square inches ensures that the RCS will be pr acted from h[

pressure transients which could exceed the limits of Appendix G to CFR Part 50 when one or more of the RCS cold legs are less than or equal to ~

j Either RHR relief valve has adequate relieving capability to protect the F.

I 7

RCS from overpressurization when the transient is limited to either (1) the start h than or equal to 50*F above the RCS cold leg temperatures provided y me taken to cushion the overpressure effects at RCS temperatures above 250*F. or 4,

( (2) the start- of geharging pumps and their injection into a water solid RCS.

- .) ;

3/4.4.11

[Iphbc7~4 STRUCTURAtAh~ aa~m T u cFecs @" s ^

The inservice inspection and testing programs for ASME Cede Class 1, 2 and 3 components ensure that the structural integrity and operatanal readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section II of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g Commission pur)suant to 10 CFR Part 50.55a(g)(6)(1).except where specific w deet 4 1d T4e G44- cf THe. IALT6CbJ WT-6OfAb& fDMPb TW6 AdLC'5tT!5 EM60 OdM $7AE.7 CF T* MA4MdM WMrh2 a: OFedadd: CH464 PdMFS Atu>hbD W LeTeo4Jrcatspr_F,c4 Tied 5, 1

FARLEY-UNIT 2 8 3/4 4-14 AMENDMENT No. 38 K0iB Im' 84%f4.%S idDJC60 WTW6 &lMLe, M MCOIFfCO l.46 Md60 T'Q Pb6 5 M d-8 Fec CcdTiddlW CF EAM3 5/4.4.10.

Enclosure 3 Significant Hazards Evaluation

Joseph M. Farley Nuclear Plant - Units I and 2 Pressure Temperature Limits Report Technical Specification Changes 10 CFR 50.92 Evaluation Pursuant to 10 CFR 50.92, SNC has evaluated the proposed amendments and has determined that operation of the facility in accordance with the proposed amendments would not involve a significant hazards consideration. The basis for this determination is as follows.

1. The proposed changes do not involve a significant increase in the probability or cuasequences of an accident previously evaluated.

The proposed removal of the Reactor Coolant System (RCS) pressure temperature (P-T) limits from the Technical Specifications (TSs) and relocation to the proposed Pressure Temperature Limits Report (PTLR) in accordance with the guidance provided by Generic Letter (GL) 96-03 is administrative in that the requirements for the P-T limits are unchanged. The P-T limits proposed for inclusion in the PTLR are based on the fluence associated with 2775 MW thermal power and operation through 21.9 effective full power years (EFPY) for Unit I and 33.8 EFPY for Unit 2. GL 96-03 requires that the P-T limits be generated in accordance with the requirements of 10 CFR 50, Appendices G and H, and be documented in an NRC-approved methodology incorporated by reference in the TSs.

Accordingly, the proposed curves have been generated using the NRC-approved methods described in WCAP-14040-NP-A, Revision 2, as modified at the direction of the NRC Staff, and meet the requirements of 10 CFR 50, Appendices G and H. TS 3.4.10.1 will continue to require that the RCS pressure and temperature be limited in accordance with the limits specified in the PTLR. The NRC-approval document will be specified in TS 6.9.1.15, and NRC approval will be required in the form of a TS Amendment prior to f changing the methodology. Use of P-T limit curves generated using the NRC-approved methods will provide additional protection for the integrity of the reactor vessel, thereby assuring that the reactor vessel is capable of providing its function as a radiological barrier.

TS 3.4.10.3 for Farley Nuclear Plant (FNP) Unit I and Unit 2 provides the operability requirements for RCS low temperature overpressure protection (LTOP). Specifically, TS 3.4.10.3 will be revised to require that two residual heat removal (RHR) system suction relief valves (RHRRVs) be operable or that the RCS be vented at RCS indicated cold leg temperatures less than or equal to 325 F. The higher temperature requirement for LTOP will provide additional assurance that overpressure protection will be available at low temperatures. Consistent with GL 96-03, the Farley Unit I and Unit 2 requirements for LTOP will be retained in TS 3.4.10.3 and will be evaluated in accordance with the proposed methodology.

Based on the above evaluation, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Enclosure 3 Page 2 Significant Hazards Evaluation

2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

As stated above, the proposed changes to remove the RCS P-T limits from the TSs and relocate them to the proposed PTLR are administrative in nature. Consistent with the guidance provided by GL 96-03, the proposed P-T limits contained in the proposed PTLR meet the requirements of 10 CFR 50, Appendices G and H, and were generated using the NRC-approved methods described in WCAP-14040-NP-A, Revision 2, as modified at the direction of the NRC Staff. The proposed changes do not result in a physical change to the plant or add any new or different operating requirements on plant systems, structures, or components with the exception oflimiting the number of operating RCPs at RCS temperatures below 110 F, increasing the temperature requirement at which the RHR relief valves are required to be operational, and establishing a higher minimum boltup temperature. Limiting the number of opera'ing RCPs below 110 F results in a reduction in the AP between the reactor vessel beltline and the RHRRVs, thereby providing additional margin to limits of Appendix G. Provisions are made to allow the stan of a second RCP at temperatures below 110 F in order to secure the pump that was originally operating without interrupting RCS flow. The LTOP enable temperature will be increased and will exceed the minimum LTOP enable temperature determined as described in WCAP-14040-NP-A, Rev. 2, thereby providing additional assurance that the LTOP system will be available to protect the RCS in the event of an overpressure transient at RCS temperatures at or below 325 F.

As stated in the above response, implementation of the proposed changes do not result in a signiEcant increase in the probability of a new or different accident (i.e., loss of reactor vessel integrity). The RCS P-T limits will continue to meet the requirements of 10 CFR 50, Appendices G and H, and will be generated in accordance with the NRC approved methodology described in WCAP-14040-NP-A, Revision 2, as modified at the direction of the NRC Staff. Therefore, the proposed changes do not result in a significant increase in the possibility of a new or different accident from any previously evaluated.

3. The proposed changes do not involve a significant reduction in a margin of safety.

The margin of safety is not affected by the removal of the RCS P-T !imits from the TSs and relocating them to the proposed PTLR. The RCS P-T limits will continue to meet the requirements of 10 CFR 50, Appendices G and H. To provide additional assurance that the P-T limits continue to meet the requirements of Appendices G and H, TS 6.9.1.15 will require the use of the NRC-approved methodology to generate P-T limits. The RCS LTOP requirements will be retained in TS 3.4.10.3 due to use of the RHRRVs for LTOP, consistent with the guidance provided by GL 96-03, and will be verified to provide adequate protection of the reactor coolant system against the limits of Appendix G. The LTOP enable temperature will be increased to 325 F and will exceed the LTOP enable temperature determined in accordance with the NRC-approved methodology, thus protecting the RCS in the event of a low temperature overpressure transient over a broader I

Enclosure 3 Page 3 Significant Hazards Evalt.ation range of temperatures than required by WCAP-14040-NP-A, Rev. 2. Administrative procedures will preclude operation of the RCS at temperatures below the minimum boltup temperature for the reactor vessel head, thus precluding the possibility of tensioning the >

reactor vessel head at RCS temperatures below the minimum boltup temperature.

Operation of the plant in accordance with the RCS P-T !!ndts specified in the PTLR and continued operation of the LTOP system in accordance with TS 3.4.10.3 will continue to meet the requirements of 10 CFR 50, Appendices G and H, and will, therefore, assure that a margin of safety is not significantly decreased as the result of the proposed changes.

Based on the preceding analysis, SNC has determined that removal of the RCS P-T limits from the TS and relocation to the proposed PTLR will not significantly increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety. SNC therefore concludes that the proposed changes meet the requirements of 10 CFR 50.92(c) and does not involve a significant hazards consideration.

1 i

Enclosure 4 Revised PTLR Methodology I

?

4 4 JOSEPli M. FARLEY NUCLEAR PLANT METilODOLOGY FOR DETERMINATION OF REACTOR COOLANT SYSEM PRESSURE TEMPERATURE LIMITS AND LOW TEMPERARIRE OVERPRESSURE PROTECTION SYSEM i

The methodology for determining the reactor coolant system pressure temperature limits includes the determination oflow temperature overpressure protection setpoints and is best described by addressing the seven " Requirements for Methodology and PTLR" found in

, Generic Letter 96-03.

1. Describe the transport calculation methods including computer codes and formulas used to calculate neutron fluence. Provide references.

Section 2.2 of WCAP-14040-NP-A, Revision 2, provides the methodology for determining the neutron fluence for the surveillance capsules and the reactor vessel with the exception j that, as requested by the NRC, calculated fluence values ($cm) are used in lieu of best-i estimate fluence (4 Bat Est) described in WCAP-14040-NP-A, Revision 2.

2. Briefly describe the surveillance program. Licensee transmittalletter should identify by title and number report containing the Reactor Vessel Surveillance Program and surveillance capsule reports. Topical / generic report contains place nolder only.

Reference Appendix H to 10 CFR 50.

The reactor vessel material surveillance program for Farley Nuclear Plant Unit 1 is described in WCAP-8810, Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program, dated December 1976. To date, four surveillance capsules have been removed from Farley Nuclear Plant Unit I as documented in the following test repons submitted to the NRC in accordance with 10 CFR

50, Appendix H

, WCAP-14196, Analysis of Capsule W from the ^ 1abama Power Company Farley Unit i Reactor Vessel Radiation Surveillance Program, dated February 1995.

.

• WCAP-11563, Revision 1, Analysis of Capsule X from the Alabama Power i

Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program, dated September 1987.

• WCAP-10474, Analysis of Capsule U from the Alabama Power Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program, dated February 1984.

i WCAP-9717, Analysis of Capsule Y from the Alatama Power Conipany Farley

, Unit No.1 Reactor Vessel Radiation Surveillance Program, dated June 1980.

Enclosure 4 Page 2 i Methodology For Determination Of RCS Pressure Temperature Limite And LTOP Limits The reactor vessel material surveillance program for Farley Nuclear Plant Unit 2 is described in WCAP-8956, Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program, dated August 1977. To date, three surveillance capsules have been removed from Farley Nuclear Plant Unit 2 as documented in the following test reports submitted to the NRC in accordance with 10 CFR 50, Appendix H:

WCAP-12471, Analysis of Capsule X from the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, dated December 1989. .

WCAP-11418, Analysis of Capsule W from the Alabama Power Company Joseph M. Farley Unh 2 Reactor Vessel Radiation Surveillance Program, dated April 1987.

WCAP-10425, Analysis of Capsule U from the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, dated October 1983.

To assure continued compliance with the requirements of 10 CFR 50, Appendix H, Surveillance Requirement 4.4.10.1.2 for Farley Nuclear Plant Units 1 and 2 associated with the P-T limits requires that the reactor vessel material irradiation surveillance specimens be removed and examined in accordance with 10 CFR 50, Appendix H.

3. Describe how the LTOP system limits are calculated applying system / thermal hydraulics and fracture mechanics. Reference SRP Section 5.2.2; ASME Code Case N-514; ASME Code, Appendix G;Section XI as applied in accordance with 10 CFR 50.55.

Farley Nuclear Plant utilizes the residual heat removal system relief valves (RHRRVs) for low temperature overpressure protection (LTOP) of the RCS from brittle fracture by assuring that the limits of Appendix G are not exceeded. The RHRRVs are spring loaded, bellows-type valves which have a setpoint of 450 psig and are designed to provide rated flow at 495 psig (i.e.,10% accumulation). In order to assure that the RHRRVs are available to protect the RCS from an LTOP event, Technical Specification (TS) 3.4.10.3 is revised to require that the RHR suction valves be open and the RHRRVs operable with a lift setting less than or equal to 450 psig or that the RCS be depressurized with a vent of greater than or equal to 2.85 square inches at RCS temperatures les.s than or equal to ,

325 F.

l The design basis transients for the Farley Nuclear Plant LTOP system consist of a heat input transient and a mass input transient with the RCS in a water-solid condition. The worst-case heat input transient assumes the start of a single reactor coolant pump with a temperature differential of 50 F existing between the RCS and any one steam generator.

At RCS temperatures less than or equal to 180 F, the worst-case mass input transient is

Enclosure 4 Page 3 Methodology For Determinatica Of RCS Pressure Temperature Limit:: And LTOP Limits assumed to be the inadvertent start of one high head safety injection (HHSI) pump with a maximum flow rate of 590 gallons per minute based on the maximum number ofoperable HHSI pumps allowed by TS 3.1.2.3. For RCS temperatures greater than 180*F, the worst-case mass input transient assumes the inadvertent operation of three HHSI pumps with a maximum total flow rate of 1000 gallons per minute at zero backpressure. These three transients discussed above are utilized to determine the RCS pressure for further analysis.

The Farley Nuclear Plant LTOP analysis consists of a determination of RCS pressures resulting from each of the design basis LTOP transients based on the reliefcapacity of the RHRRVs and the following conservative assumptions:

• Credit is taken for flow through only one RHRRV due to single failure of the other RHRRV;

• No flow through the RHRRVs is credited in the analysis until RCS pressure achieves the 10% accumulation pressure for the RHRRVs of 495 psig;

• Flashing is assumed ic, occur at the valve discharge; e No credit is taken for a bubble in the pressurizer; and

• The analysis is performed at isothermal conditions in the RCS and provides protection against the steady-state Appendix G limit.

At RCS temperatures less than or equal to 180 F, the most-limiting design basis transient results in an RCS pressure of 495 psig. The resulting pressure is compared to the proposed Appendix G steady-state limit curve to assure that the resulting RCS pressure of 495 psig does not exceed the allowable RCS pressure. The following table provides an Example of Comparison of Limiting Design Basis Transient (LDBT) to Appendix G Steady State Limit Curve.

Example of Comparison of Limiting Design Basis Transient to Appendix G Steady State Limit Curve for Farley Unit 2 RCS Temperature RCS Pressure Appendix G Steady

('F) (LDBT)(psig) State Limit Curve (psig) 75 495 501 180 495 626 181 56? 5 629 260 562.5 1070 261 795 1080 310 795 1749

i Enclosure 4 Page 4 Methodology For Determinat;on Of RCS Pressure l Temperature Limits And LTOP Limits l

I As stated above, the RCS pressure for each of the above temperatures are compared to the proposed steady-state Appendix G curve to assure that the RCS pressure does not exceed the Appendix G allowable pressure for the corresponding temperature. If this criteria is met, the Farley Nuclear Plant LTOP system provides adequate protection for the proposed Appendix G curves. As can be seen from the above comparison, the Farley Nuclear Plant LTOP system provides adequate protection for the Appendix G curves.

If the projected RCS pressure exceeds the Appendix G allowable pressure for the corresponding temperature, changes to the RHRRV characteristics, e.g., capacity, relief setpoint, accumulation, may be required. If the projected RCS pressure exceeds the Appendix G allowable pressure using this methodology, the issue must be resolved with NRC StafTreview and approval.

The Farley Nuclear Plant LTOP enable temperature is the temperature below which the LTOP system is required to be operable in accordance with Section 3.4 of WCAP-14040-NP-A, Revision 2. The LTOP enable temperature is compared to the RCS cold leg temperature stated in the applicability statement of TS 3.4.10.3 to assure the RCS l overpressure protection systems are available at temperatures below the LTOP enable temperature. The LTOP enable temperature will include an allowance for indicated temperature measurement uncertainty. If 325'F is not an acceptable LTOP enable temperature, a change to Technical Specification 3.4.10.3 will be required. Since an LTOP event can not occur with the vessel head removed, the minimum boltup temperature will also include an allowance for indicated temperature measurement uncertainty.

In order to minimize setpoint uncertainties and dnft, Farley Nuclear Plant tests the RHRRVs on an accelerated basis from that required by the ASME Code. Bench tests are performed at 18 month intervals on a rotating basis for at least one of the RHRRVs to verify the setpoint in accordance with TS Surveillance Requirement 4.4.10.3.l(c). This frequency is more stringent than that required by the AShE Code for class 2 relief valves.

. Additionally, Farley Nuclear Plant surveillance test procedures currently use an RHR relief valve setpoint of 436 13 psig. Use of a maximum RHRRV setpoint less than 450 psig coupled with the 10% accumulation provides adequate protection against setpoint drift The increased surveillance test frequency, the reduced RHRRV setpoint, coupled with the analysis assumption that flow does not start untilinlet pressure reaches 450 psig + 10% accumulation, i.e.,495 psig, provide assurance that the RHR relief valves will provide adequate protection against the limits of Appendix G.

AShE Code Case N-514 is not used for Farley calculations.

l

Enclosure 4 Page5 Methodology For Determination Of RCS Pressure Temperature Limits And LTOP Limits

4. Describe the method for calculating the ART using Regulatory Guide 1.99, Revision 2.

Section 2.4 of WCAP-14040-NP-A, Revision 2, provides the methodology for calculating the adjusted reference temperature in accordance with Regulatory Guide 1.99, Revision 2.

5. Describe the application of fracture mechanics in constructing P-T curves based on ASME Code, Appendix G,Section XI, and SRP Section 5.3.2.

Sections 2.5 and 2.6 of WCAP-14040'-NP-A, Revision 2, provides the application of fracture mechanics in constructing P-T curves. The resulting P-T limit curves are adjusted to account for the 60 psi AP between the reactor vessel beltline and the RHRRVs associated with the operation of three reactor coolant pumps (RCPs) at RCS temperatures greater than or equal to 110 F. At RCS temperatures less than 110'F, the number of operating RCPs is limited to one and the resulting AP correction of 27 psiis applied. The above AP corrections include the static and dynamic effects of RHR system operation.

6. Describe how the minimum temperature requirements in Appendix G to 10 CFR 50 are applied to P-T curves.

Section 2.7 of WCAP-14040-NP-A, Revision 2, provides the methodology for determination of the minimum temperature requirements in 10 CFR 50, Appendix G. The minimum temperature requirement is adjusted as necessary to assure the RCS pressure resulting from design basis LTOP transients does not exceed the steady state Appendix G limit.

7. Describe how the data from multiple surveillance capsules are used in the ART calculation.

Section 2.4 of WCAP-14040-NP-A, Revision 2, provides the methodology for calculating the adjusted reference temperature with multiple surveillance capsules.

Describe procedure if measured value exceeds predicted value.

As stated in Section 2.4 of WCAP-14040-NP-A, Revision 2, if the measured value exceeds the predicted value, a supplement to the PTLR must be provided to demonstrate how the results affect the approved methodology.

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[ Methodology For Determination Of RCS Pressure Temperature Limits And LTOP Limits WIIEN OTIIER PLANT DATA ARE USED

1. Identify the source (s) of data when other plant data are used.

Farley Nuclear Plant does not rely on surveillance data from other licensees for its reactor vessel integrity analysis. Therefore, this item is not applicable to Farley Nuclear Plant.

2a. Identify by title and number the safety evaluation report that approved the use of data for the plant. Justify applicability.

Farley Nuclear Plant does not rely on surveillance data from other licensees for its reactor vessel integrity analysis, Therefore, this item is not applicable to Farley Nuclear Plant.

OR 2b. Compare licensee data with other plant data for both the radiation environments (e.g., neutron spectrum, irradiation temperature) and the suneillance test results.

Farley Nuclear Plant does not rely on surveillance data from other licensees for its reactor vessel integrity analysis. Therefore, this item is not applicable to Farley Nuclear Plant.

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