ML20078S756

Proposed Tech Specs Re Pages for MSSV

ML20078S756
Person / Time
Site:	Farley
Issue date:	12/19/1994
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20078S754	List: ML20078S753 ML20078S756
References
NUDOCS 9412290197
Download: ML20078S756 (41)

Text

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j Enclosure 1 i

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I Revised Technical Specification Pages for the MSSV Amendment

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3/47-2 Replace i 3/47-3 Replace j B 3/4 7-1 Replace

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Changas Marked with Brid, it;//cized Print and Strikethroughs Table 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER) 1 87 2 65 48 3 4-3 28 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator * (Percent of RATED THERMAL POWER) 1 **

2 **

3 **

• At least two safety valves shall be OPERABLE on the non-operating steam generator.

• • These values left blank pending NRC approval of 2 loop operation.

FARLEY-UNIT 1 3/4 7-2 AMENDMENT NO.

Changes Marked with B Id, Italicized Print and Strikethroughe

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i N1

> TABLE 3.7-3 W

h STEAM LINE VALVES PER LOOP K

[ VALVE NUMBER LIFT SETTING M143 * (136) *

• ORIFICE SIZE (SQ. IN.) I I

Z w

g a. Q1N11VO - 10A, 11A, 12A 1075 psig 16

" 01N11VO - 10B, 11B, 12B

b. 1088 psig 16

c. OlN11VO - 10C, 11C, 12C 1102 psig 16 i

01N11VO - 10D, 11D, 12D

d. 1115 psig 16

e. 01N11VO - 10E, llE, 12E 1129 psig 16 ta N

b

-J E *The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

• • M ter testing, the valves will be left at 116. l Z

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tn 2

8 Z

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Chang s Marked with B:Id,it //cizedPrint and Strikethroughs 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1194 psig) of its design pressure of 1085 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser) .

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 12,923,371 at least 22,984,660 lbs/hr which is 444 112 percent of the total secondary steam flow of 11,565,792 22,613,849 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are consistent with the asstaptions used in the accident analysis. derived == the f=11=uing besee+

Fer-3 1ccp cperation cp - (X) (Y)(V) (199; M

Fc 2 lecp cp : ti-en W) X (Y)(U) , (55)

M Wh::::

SP - Reduced ::::ter trip-setpcint in percent cf nATED THEnMAL POWER W aw4m= nu-50: cf inoperable safety valves p : steam 44ee U-=-Maxim = n=ber cf incperable :sfety valve p:: cperet4ng

=te - line FARLEY-UNIT 1 B 3/4 7-1 AMENDMENT NO.

Ch:ng:s Marked with Beld, it:lleized Print and Stdkethroughe PLANT SYSTEMS EASES 100 - 0;ucr n ng: Neutron riux-Migh Trip S:tpcint for 3 1 0p epee *bl+n 55 - Mexi=u pere:nt of n.'.TED TMESMAL PONER p:rmieeible by 0 9 S:tpcint for 2 1:0p Oper: tion.

X - Tctal relieving : pacity Of all : fety valve; per :tes.T line in 2b:/ hour Y - M:ni=u= relieving ::p::ity Of :ny en :sfety valv in Ib:/heur 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM j The OPERABILITY of the auxiliary feedwater system ensures that the l Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of off-site power.

1 Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 330 gpm at a pressure of 1133 peig psia to the entrance of the steam generators. The steem driven auxiliary feedwater pump is capable.of delivering a total feedwater flow of450gpmatapressureof1133peigpelatotheentranceofthesteaml generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat l Removal System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.

This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the accident I analyses.

FARLEY-UNIT 1 B 3/4 7-2 AMENDMENT NO.

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Changss Marked with Beld, italicized Print and Strikethroughe Table 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER) 1 87 2 45 48 3 4-3 28 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITl!

INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator * (Percent of RATED THERMAL POWER) 1 **

2 **

3 **

• At least two safety valves shall be OPERABLE on the non-operating steam generator.

• • These values left blank pending NRC approval of 2 loop operation.

FARLEY-UNIT 2 3/4 7-2 AMENDMENT NO.

i Changes Marked with Brid, It licized Print and Strikethroughe ,

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$ TABLE 3.7-3 y STEAM LINE VALVES PER LOOP k

VALVE NUMBER LIFT SETTING M * (138) *

• ORIFICE SIZE (SO. IN.) l

[

Z y a. 02N11VO - 10A, 11A, 12A 1075 psig 16

" b. 02N11VO - 10B, 11B, 12B 1088 psig 16

c. Q2N11VO - 10C, lic, 12C 1102 psig 16

d. 02N11VO - 10D, 11D, 12D 1115 psig 16

e. 02N11VO - 10E, 11E, 12E 1129 psig 16 w

N A

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

• • After testing, the valves will be left at lit. l Z

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I Chang s Markcd with Brid, Italicized Print and Strikethroughs 3/4.7 PIANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1194 psig) of its design pressure of 1085 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i . e . , no steam bypass to the condenser) .

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 12,'75,234 at least 22,984,660 lbs/hr which is 444 112 percent of the total secondary steam flow of 11,613,849 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER l required by the reduced reactor trip settings of the Power Range Neutron l Flux channels. The reactor trip setpoint reductions are consistent with i

the assuxptions used in the accident analysis. d::ived =n th:-fo14sukng besee+

i I N: 31 Op Operation l  %(X) (Y) (V) ,, (109)

M F0: 2 locp ep :stien

%) X (Y)(" W _.g g.

M Wh::::

SP - Reduced ::::ter trip ::tpcint in pt: cent o f Pf.TE,9 THEPJ'I.L POWER l

l v- -wax =u= nu-se ef in:perme-.:fety valve: p: ::::=

Liwe t' - Maid =ur number cf inoperable :sfety valvez p : eperating s..- ,s..

wwwun ....w

! FARLEY-UNIT 2 B 3/4 7-1 AMENDMENT NO.

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I Chang:is Marked with Beld, Italicized Print and Stdkethroughe )

, . i PLANT SYSTEMS BASES 400 - P uer n:ng: M:utron Flun-Migh Trip S:tpoint for 3 loop operation f f - M:nimum p;rcent of P*.TSO TMS."'.L PO"En per .iccible by P-S S:tpcint f+e 2 loop sp:: tien Y - Total relieving : p: ity of all ::fety valve per ste =

line in lb / hour Y - M:nimum relieving : p :ity of any ont :sfety v:lve in Ib:/ hour 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350'F from normal operating conditions in the event of a total loss of off-site power.

Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 330 gpm at a pressure of 1133 psia l to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 450 gpm at a pressure of 1133 psia to the entrance of the steam generators. l This capacity is sufficient to ensure that adequate feedwater flow is  ;

available to remove decay heat and reduce the Reactor Coolant System '

temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.

j This dose also includes the effects of a coincident 1.0 GPM primary to I secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the accident analyses.

FARLEY-UNIT 2 B 3/4 7-2 AMENDMENT NO.

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0 4 Technical Specification Pages t

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l Table 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Pt reent of RATED THERMAL POWER) 1 87 2 48 3 28 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator * (Percent of RATED THERMAL POWER) 1 **

2 **

3 **

• At least two safety valves shall be OPERABLE on the non-operating steam generator.

• • These values left blank pending NRC approval of 2 loop operation.

FARLEY-UNIT 1 3/4 7-2 AMENDMENT NO.

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m h TABLE 3.7-3 g STEAM LINE VALVES PER LOOP VALVE NUMBER LIFT SETTING * (13%)** ORIFICE SIZE (50. IN.)

f l Z

y a. Q1N11VO - 10A, 11A, 12A 1075 psig 16

b. Q1N11VO - 10B, 11B, 12B 1088 psig 16

c. 01N11VO - 10C, 11C, 12C 1102 psig 16

d. 01N11VO - 10D, llD, 12D 1115 psig 16

e. 01N11VO - 10E, 11E, 12E 1129 psig 16 w

N b

9 0 *The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

• • After testing, the valves will be left at 11%. l

=

0 Z

H Z

.O

l 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures J

, that the secondary system pressure will be limited to within 110% (1194 psig) of its design pressure of 1085 psig during the most severe anticipated system operational transient. The maximum relieving ,

capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in ac'm edance with the requirements of Section III of the ASME Boiler and l

Pr.3sure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is at least 12,984,660 lbs/hr which is 112 percent of the total secondary steam flow of 11,613,849 lbs/hr at 100% PATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the

basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are consistent with the assumptions used in the accident analysis, i

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FARLEY-UNIT 1 B 3/4 7-1 AMENDMENT NO.

PLANT SYSTEMS BASES I

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of off-site power.

Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 330 gpm at a pressure of 1133 psia l to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 450 gpm at a pressure of 1133 psia to the entrance of the steam generators. l This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimwm water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.

This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the accident analyses.

FARLEY-UNIT 1 B 3/4 7-2 AMENDMENT NO.

Table 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER) 1 87 2 48 3 28 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER) 1 **

2 **

3 **

• At least two safety valves shall be OPERABLE on the non-operating steam generator.

• • These values left blank pending NRC approval of 2 loop operation.

FARLEY-UNIT 2 3/4 7-2 AMENDMENT No.

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$ TABLE 3.7-3 STEAM LINE VALVES PER LOOP VALVE NUMBER LIFT SETTING * (i3%)** ORIFICE SIZE (SQ. IN.)

f l Z

y a. Q2NilVO - 10A, 11A, 12A 1075 psig 16

" Q2N11VO - 10B, llB, 128

b. 1088 psig 16

c. 02N11VO - 10C, llc, 12C 1102 psig 16

d. 02NilVO - 10D, llD, 12D 1115 psig 16

e. Q2N11VO - 10E, llE, 12E 1129 psig 16 ta s

A a

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

• • After testing, the valves will be left at 11%. l k

m Z

O Z

m Z

H Z

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i 3/4.7 PLANT SYSTEMS i

BASES

{ l l I Il l II I

3/4.7.1 TURBINE CYCLE

! 3/4.7.1.1 SAFETY VALVES 4 The OPERABILITY of the main steam line code safety valves ensures

! that the secondary system pressure will be limited to within 110% (1194 psig) of its design pressure of 1085 psig during the most severe

! anticipated system operational transient. The maximum relieving l capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam j bypass to the condenser) .

The specified valve lift settings and relieving capacities are in  !

accordance with the requirements of Section III of the ASME Boiler and l Pressure Code, 1971 Edition. The total relieving capacity for all i valves on all of the steam lines is at least 12,984,f60 lbs/hr which is l 112 percent of the total secondary steam flow of 11,613,849 lbs/hr at

100% RATED THERMAL POWER. A ndnimum of 2 OPERABLE safety valves per i steam generator ensures that sufficient relieving capacity is available i for the allowable THERMAL POWER restriction in Table 3.7-2.

i i STARTUP and/or POWER OPERATION is allowable with safety valves

! inoperable within the limitations of the ACTION requirements on the

basis of the reduction in secondary system steam flow and THERMAL POWER 4 required by the reduced reactor trip settings of the Power Range Neutron j Flux channels. The reactor trip setpoint reductions are consistent with the assumptions used in the accident analysis, d

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j FARLEY-UNIT 2 B 3/4 7-1 AMENDMENT NO.

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PLANT SYSTEMS i

BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350'F from normal operating conditions in the event of a total loss of off-site power.

Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 330 gpm at a pressure of 1133 psia l to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 450 gpm at a pressure of 1133 psia to the entrance of the steam generators. l This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the ndnimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.

This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the accident analyses.

FARLEY-UNIT 2 B 3/4 7-2 AMENDMENT NO.

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l Enclosure 2 l

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1 Significant Hazards Evaluation 1

for  ;

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Main Steam Safety Valve 1 i

Technical Specification Amendment 4

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Joseph M. Farley Nuclear Plant Main Steam Safety Valve Technical Specification Amendments 1 l

10 CFR 50.92 EVALUATION As required by 10 CFR 50.91 (a)(1), an analysis is provided to demonstrate that the proposed license amendment to increase the main steam safety valve (MSSV) lift setting 1% tolerance from to 3% and to change high nuclear flux setpoints for multiple MSSVs out of service involves no significant hazards consideration.

Proposed Change ne propoced change for Table 3.7-3 of the Farley Nuclear Plant Technical Specifications includes the revision to the MSSV setpoint tolerance from 11% to 3% and also modifies the bases to Technical Specification 3/4.7.1.1 to increase the relieving capacity of the MSSVs to at least 12,984,660 pounds per hour which corresponds to approximately 112% of total secondary s flow at 100% rated thermal power. In addition, modifications to Table 3.7-1 are proposed to reduce the allowable power range neutron flux high setpoints for multiple inoperable steam generator safety valves. %ese changes are consistent with the current analyses for Farley Units 1 and 2. In addition, this amendment includes an editorial correction to Bases 3/4.7.1.2 which shou indicate required auxiliary feedwater flow at i133 psia rather than 1133 psig.

Backaround He MSSVs provide overpressure protection for the Farley steam generators. These automatic pressure relieving devices are designed to pass at least 110% of the maximum guaranteed steam flow at a steam generator shell pressure not greater than 110% of the design pressure of the generator. This is the nuximum pressure allowed by the code.

He current Technical Specifications contain a 11% tolerance on the MSSV actuation is the intent of the proposed change to expand the tolerance of the MSSVs to 3% When measuring the MSSV setpoint for surveillance, the valves will be set to a 11% measured, as-le tolerance; however, should a valve be found outside the il% tolerance band but within the proposed 3% tolerance band, it is acceptable since 3% tolerance is the new analyzed condition Other or proposed more MSSVs changes inoperable. reflect the current analysis basis for Farley Nuclear Plant for The Bases 3/4.7.1.2 correction is editorial since the Farley analyses have consistently cons auxiliary feedwater capability of 330 gpm at 1133 psia.

Ac.ahiis Ac following evaluations and analys= address the effect of 3% MSSV setpoint tolerance on mechanical aspects, design transients, non-LOCA accident analysis, LOCA and LOCA related analyses, procedures. steam generator tube rupture, radiological consequences, and emergency ope

1' Main Steam Safety Velves Page 2 j Significant Hazards Analysis 1

j Mechanical Asoccts and Desian Transients i

a An evaluation of the MSSVs has determined that if the MSSVs are set to within 1% of the i nominal set-pressure and later found to be outside this toicrance but within 3%, there are no valve i operational concerns. The valves will continue to function as designed.

i j

j ne design transients used for the component fatigue stress analyses applicable to the Faricy units j have been evaluated to support an increase in MSSV setpoint tolerances to *3%. Increasing the

] tolerance on these setpoints must be evaluated to confirm that no design transient is created which j is more limiting than those currently applicable.

i i

) An increase of the MSSV setpoint tolerance could affect the peak steam generator pressure during i the loss ofload and loss of power transients that are used to define the design transients for Farley

{ Nuclear Plant. Any increase in pressure associated with the tolerance increase will not increase the  !

! maximum postulated pressure to a value greater than the design basis of 110% of the steam  !

] generator shell design pressure. He lowest steam pressure remains above the design opening l j pressure for the steam generator PORVs. Adequate margin exists in the analytical assumptions for i 1 the current design trarsients such that no change to the component fatigue analyses is required. i ne c:: rent design transients remain valid for all applications.

Non-LOCA Accident Analysis 4

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j He non-LOCA accident analyscs can be placed into two categories with respect to the proposed

increase in MSSV setpoint tolerance to 3%. Dese are (1) non-limiting transients or transients j which do not actuate MSSVs, or (2) limiting transients which may actuate MSSVs.

i l The first set of analyses, i.e., non-limiting transients or transients which do not actuate MSSVs,

! include uncontrolled RCCA bank withdrawal from a suberitical condition, main steamline break, j- main steam line break mass and energy releases (inside and outside containment), RCCA

misalignment, uncontrolled boron dilution, loss of flow / locked rotor analyses, startup of an inactive loop, feedwater system malfunction, excessive load increase, accidental depressurization of the RCS, inadvertent loading and operation of a fuel assemi ly in an improper position, single RCCA withdrawal at full power, and RCCA ejection. This group of transients is unaffected by the increase in the MSSV setpoint tolerance and therefore all FSAR conclusions remain valid.

i ,

I ne second set of analyses, i.e., limiting transients that may actuate MSSVs include uncontrolled  ;

RCCA bank withdrawal at power, loss of external electrical load and/or turbine trip, loss of l normal fccdwater and loss of non emergency AC powcr to plant auxiliarics, inadvertent operation

of ECCS during power operation, and major rupture of a mai ! feedwater pipe. For all of these

events, the current analys s of record has accounted for the inweased tolerance on MSSVs and all l l acceptance criteria continue to be met.

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I Main Steam Safety Valves Page 3 Significant Hazards Analysis l

Additional analyses were performed to determine the maxunum allowable power range neutron flux high setpoints with inoperable main steam safety valves. He setpoint (87%) for 1 valve / loop remams the same. He setpoint for 2 valves / loop is modified to 48%, and the setpoint for 3 valves / loop is modified to 28%. Rese high nuclear flux setpoints reflect the 3% MSSV setpoint j tolerance and appropriate accumulation.

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i LOCA and LOCA Related Evaluations 1

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Large Break LOCA 1

l He current large break LOCA analyses for J. M. Farley Units 1 and 2 were performed with the j NRC approved Evaluation Model using BASH. After a postulated large break LOCA occurs, the heat transfer between the reactor coolant system (RCS) and the =~=d y system may be in either j direction, Awynding on the relative temperatures. In the case of continued heat addition to the i aer tary system, the secondary pressure increases and the MSSVs may actuate to limit pressure.

] However, this does not occur in the large break evaluation model since no credit is taken for auxiliary feedwater actuation. Caas~;naady, the =ar=tary system acts as a heat source in the j postulated large break LOCA transient and the secondary pressure does not increase. Since the j secondary system pressure does not increase, it is not necessary to model the MSSV setpoint in the

! large break evaluation model. Herefore, an increase in the allowable MSSV setpoint tolerance for

J. M. Farley Units 1 and 2 will not impact the current FSAR large break LOCA analyses.

2 l Small Break LOCA i

l The small break LOCA analyses for J. M. Farley Units I and 2 were perfomed with the NRC l approved Evaluation Model using the NOTRUMP code. After a postulated small break LOCA 1 j occurs, the heat transfer between the RCS and the secondary system may be in either direction

, dependmg on the relative temperatures. In the case of continued heat addition to the secondary j system, the secondary system pressure increases which leads to steam relief via the MSSVs. In the s small break LOCA, the secondary flow aids in the reduction of RCS pressure. The current j licensing basis small break LOCA analysis for J. M. Farley Nuclear Plant was performed using a l MSSV setpoint tolerance ofi3%. Derefore, the increased setpoint tolerance has already been j included in the analysis of record and all acceptance criteria continue to be met.

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LOCA Related Events l

i l Post-LOCA cooling, hot leg switchover to prevent potential boron precipitation, and LOCA a hydraulic forcing functions have been reviewed and it has been deternuned that the MSSV setpoint j tolerance incuse has no effect on the results or conclusions for these events.

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i Main Steam Safety Vclves Page 4 Significant Hazards Analysis Steam Generator Tube Ruoture (SGTR)

A design basis failure of a single steam generator tube was evaluated using the assumptions which were utilized in the FSAR SGTR analysis. An SGTR results in a loss of coolant inventory, and reactor trip and safety injection (SI) are assumed to occur on a low pressurizer pressure signal.

After reactor trip and SI actuation, it is assumed that the manA=ry side pressure stabilizes at the MSSV setpoint minus the tolerance (3%). It is assumed that the RCS pressure stabilizes at the equilibrium value where the incommg SI flow rate balances the tube rupture break flow rate, which is dependent on the primary to mnadary side pressure differential. The resultant equilibrium break flow rate is assumed to persist from the time of reactor trip and SI actuation until 30 minutes after the accident. A maximum SI flowrate is conservatively assumed for the design basis SGTR analysis in order to maxim i ze the break flow.

Le results of the thermal and hydraulic evaluation of the SGTR for the increased MSSV tolerance indicate that the primary to enndary break flow of 137,811 pounds would increase by 0.9% to 139,100 pounds The amount of steam released is increased by 1.6% to 65,500 pounds. Ecse results were used in calculating the effect of the increased MSSV tolerance on the offsite radiological consequences.

He radiological consequences were reanalyzed using the results of the previously noted thermal and hydraulic evaluations, ne calculated doses remain with the NRC acceptance criteria of 10CFR100.

Radiological Consequences The radiological consequences for FSAR analyses for loss ofload/ turbine trip, loss of offsite power, locked rotor. main steam line break and rod ejection remain bounding for the proposed MSSV tolerance inenw:. l EOPs Emergency Operating Procedures, which could be affected by the proposed change, have been reviewed and no changes in any setpoints are rauired.

10CFR50.92 Conclusions Conformance of the proposed amendment to the standards for a determination of a No Significant ,

Hazards as performed in 10CFR50.92 is shown in the following:

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! Main Steam Safety Valves Page5

Significant Hazards Analysis 1

1. The proposed license amendment does not involve a significant increase in the probability J or consequences of an accident previously evaluated.

l l i l i These proposed changes to the Farley Technical Specifications do not result in a condition where the design, material and construction standards of the MSSVs that were applicable l

j prior to the proposed change are altered. The valves will continue to function as designed.

l All applicable safety analyses have been reviewed, evalusted or reanalyzed and all

! applicable safety criteria continue to be met. No accident quences are altered because of the proposed amcr4T. cat. The radiological consequences for the Steam Generator Tube Rupture were reanalyzed and 10CFR100 criteria continue to be met. All other FSAR radiological analyses remam bounding Analyses have been performed tojustify the 1 proposed high nuclear flux setpoint changes All acceptance criteria for these analyses j continue to be met. Therefore, the proposed amendment does not result in a significant j increase in the probability or consequences of an accident previously evaluated.

4 l 2. The proposed license amendment does not create the possibility of a new or different

accident from any accident previously evaluated.

j The MSSVs continue to have the required pressure relieving capacity to ensure that system

! design pressure remams below 110% of shell design pressure. The proposed changes are j not accident initiators nor do they create any new accident scenarios or any new limiting j single failures. The ability of the MSSVs to respond to an accident condition is not j impaired by the proposed changes. The proposed high nuclear flux setpoints for multiple i valves out of service ensure all applicable safety criteria for accident analyses are met. No j new accident scenarios are created by these proposed changes. Therefore, the proposed amcr4T.cnt does not create the possibility of a new or different kind of accident from any

accident previously evaluated.

! 3. The proposed license amendment does not involve a significant reduction in the margin of l safety. l

, l Acceptance criteria for accident analysis continue to be met. Radiological consequences I for the affected Chapter 15 analysis remain within 10CFR100 acceptance criteria. No j safety limits or safety system setpoint requires modification due to the proposed changes.

j 1he current secondary side over-pressure limit of i10% of steam generator shell design j pressure is not violated. Analysis for the high nuclear flux setpoints have verified that  ;

i there is no reduction in margin for the events analyzed. Therefore, there is not significant reduction in the margin of safety.

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l Enclosure 3 l

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1 Safety Analysis for Main Steam Safety Valve Technical Specification Amendment

1 Joseph M. Farley Nuclear Plant l Main Steam Safety Valves )

Safety Analysis )

EXECUTIVE

SUMMARY

Southem Nuclear is proposing a change to the Main Steam Safety Valve (MSSV) setpoint I tolerance from 1%to 3%. If the as-found set-pressure of a MSSV is found to be within the tolerance of 3%, the measured, as-left set-pressure will be reset to be within 1%. He analyses indicate that the change does not adversely affect the performance of the MSSV and acceptance criteria for all accidents continue to be met. In addit;on, analyses to support changes to high nuclear flux setpoints for multiple MSSVs out of service were performed. Acceptance criteria for these cases was also met.

BACKGROUND He MSSVs provide overpressure protection for the Farley steam generators. Dese automatic pressure relieving devices are designed to pass at least 110% of the maximum guaranteed steam flow at a steam generator shell pressure not greater than 110% of the design pressure of the steam generator. His is the maximum design pressure allowed by the applicable code.

The Technical Specifications' setpoint tolerance of11% was based on an ASME Section III requirement. To allow increased flexibility, a Technical Specification setpoint tolerance change to i3% is proposed. Prior to implementing such a change, the applicable safety analyses have to be i reviewed to show that the plant would not be overpressurized if a valve (s) is determined to have a l

set-pressure within the 13% tolerance. All analyses have been reviewed and all applicable criteria continue to be met.

Westinghouse identified in Nuclear Safety Advisory Letter NSAL-94-001 a deficiency in the basis I for Technical Specification 3.7.1.1. His Technical Specification allows the plant to operate at a reduced power level with a reduced number of operable Main Steam Safety Vahrs (MSSVs). The deficiency is in the assumption that the maximum allowable initial power level is a linear function of the available MSSV capacity. The linear function is identified in the bases section for this Technical Specification.

One of the options presented in NSAL-94-001 is to perform plant-specific analyses of the Loss of Loadfrurbine Trip (LOI/IT) event to analytically determme the maximum allowed power level for a given number ofinoperable MSSVs. Dependmg on key specific plant parameters, these analyses may be able tojustify the continued validity of the current Technical Specification. A plant-specific analysis has been performed for Farley Nuclear Plant.

i E- Main Steam Safety Valves Page 2 j Safety Analysis i

EVALUATION

} Mechanic =1

The MSSVs at Farley Units 1 and 2 are Dresser Model 3707R. The valve enluation has shown j that if the MSSVs are set to within 1% of the nominal set-pressure and later found to be outside j this tolerance but within i 3%, there are no ASME valve operational concerns. If a valve is found j to have a set-pressure outside the 1% tolerance, it should be reset to within 1% to minmuze the

potential for valve leakage and set-pressure overlap problems.

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! The MSSVs design basis set-pressures are staggered so that they open at different pressures. The

! first valve set-pressure is 1075 psig, which corresponds to the steam generator shell design l pressure minus the pressure loss from the steam generator to the valve Each of the remaining valves is set at a higher pressure, such that all valves are open and at full relief without exceedmg 110% of the steam generator shell design pressure. The design steam pressure for the steam j generator is 1085 psig.

Desian Transients l

The design transient curves used for the component fatigue stress analyses applicable to the Farley units have been evaluated to support an increase in MSSV setpoint tolerances to *3%. Increasing the tolerance on these setpoints must be evaluated to confirm that no design transient is created which is more limiting than those currently applicable.

An increase of the MSSV setpoint tolerance could affect the peak steam generator pressure during the loss ofload and loss of power transients that are used to define the design transients for the Farley Nuclear Plant. Any increase in pn nure associated with the tolerance increase will not increase the maximum pa=tuhtad pressure to a value greater than the design basis of 110% of the steam generator shell design pressure. The lowest steam pressure remams above the design opening pressure for the steam generator PORVs. Adequate margin exists in the analytical assumptions for the current design transients such that no change to the component fatigue analyses is required. The current design transients remain valid for all applications.

Non-LOCA l

Currently the Farley Unit I and 2 Technical Specifications require that the main steam safety valves (MSSV) be verified to be operable within a

• 1% tolerance of the corresponding nominal lift set-pressure. The analyses and evaluations for non-LOCA transient documented within demna=trate that the MSSV setpoint tolerance can be increased to i3%.

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i Main Steam Safety Valves Page 3 l Safety Analysis l Evaluation l

4 i The following evaluation discusses how the MSSVs have been historically modeled and how the j safety valves are currently modeled in the non-LOCA safety analyus. This is followed by a j discussion of how the increase in the MSSV tolerance has been incorporated into the non-LOCA j safety analyses.

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4 Historic Safety Analyses Safety Valve Model 4

I Historically, the *1% tolerance of the MSSVs has not been explicitly modeled in the non-LOCA j safety analyses. Rather, the safety analyses assumed that the safety valves were open when the valves reach a pressure 3% above the nommal set-pressure. The 3% value is the point where the j valves had accumulated such that the valves were relieving at their full capacity. No relief was

assumed prior to reaching the accumulation pressure. In addition, the safety valves were modeled i at one pressure, corresponding to the steam generator design pressure plus 3% for accumulation.

I This approach was acceptable because S transients presented in the FSAR only required the i

) opemng of the safety valves for a post-reactor trip condition in which the total stored energy and  !

decay heat level would be less than the relief capacity of one bank of safety valves. 'Ihe MSSVs  !

! " closed" when the pressure dropped below the accumulation pressure.

J Justification of the Increased Safety Valve Tolerance i

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Tojustify an increase in the MSSV tolerance, it is assumed that the MSSVs open at the set-

! pressure plus 3% accumulation plus 3% tolerance. The safety analyses assume that the valves only relieve enough steam to maintain pressure at the set-pressure. It is also assumed that the j

j secondary pressure will drop well below the openmg pressure of the valve before the valve will

! close. Therefore, the " average" secondary pressure during the time when the valves are open will j be less than the set-pressure plus the tolerance.

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Two MSSV models are used in the safety analyses. The first model assumes that the safety valves

open at a pressure greater than the lowest set-pressure plus 3% accumulation plus 3% tolerance.

The second model assmnes staggered MSSV banks at nommal set-pressures plus an accumulation allowance plus 3% tolerance. The second MSSV model models the actual operation of the MSSVs

more accurately. Both MSSV models support an increase in the safety valve tolerance to i3%.

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! The following section addresses the individual non-LOCA accidents with respect to the MSSV i

tolerance, as discussed above.

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l Main Steam Safety Vcjves Page 4 Safety Analysis i

Non-LOCA Safety Analyses Event Discussion  !

Steamline Break Mass and Energy Releases (FSAR 6.2.1.3.11)

He steamline break mass and energy releases are not affected because the transient results in a decrease in the secondary pressure. Herefore, the MSSVs are not challenged and the mass and energy releases remain valid.

Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition (FSAR 15.2.1) he uncontrolled RCCA withdrawal from a subcritical condition event is analyzed to demonstrate that the DNB design basis is met. Since this is a rapid reactivity excursion event that is promptly termmated by a reactor trip at 35% rate thermal power, the peak thermal power is much less than that associated with full power operation (typically, less than 50% rated thermal power). His event is not limiting with respect to secondary side pressure. Herefore, an MSSV tolerance of i3% can be supported and the conclusions of the FSAR remam valid.

Uncontrolled RCCA Bank Withdrawal at Power (FSAR 15.2.2) he uncontrolled RCCA withdrawal at power event is analyzed to demonstrate that the DNB design basis is met. He MSSVs are modeix! at a pressure greater than the lowest set-pressure plus 3% accumulation plus 3% tolerance. He results show that the per.k secondary pressure is not exceeded, the DNB design basis is met and the pressurizer does not overfill for the Farley plant.

Herefore, an MSSV tolerance ofi3% can be supported and the conclusions of the FS AR remain valid.

RCCA Misalignment (FSAR 15.2.3)

The dropped RCCA(s) event is analyzed to demonstrate that the DNB design basis is met.

Changes to the MSSVs tolerance do not impact the event.  ;

Uncontrolled Boron Dilution (FSAR 15.2.4)

He chemical and volume control system malfunction that results in a decrease in the boron concentration in the RCS event is analyzed to demonstrate that sufficient operator action time is available to terminate the event prior to losing shutdown margin. He event does not examme secondary pressure since secondary pressure is bounded by other Condition II events. Herefore, an MSSV tolerance of13% can be supported and the conclusions of the FSAR remain valid.

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Main Steam Safety Valves Page 5 Safety Analysis Partial Loss of Forced Reactor Coolant Flow (FSAR 15.2.5 )

The partial loss of reactor coolant flow event is analyzed to demonstrate that the DNB design basis is met. The MSSVs are not challenged during this transient. Therefore, an MSSV tolerance of i3% can be supported and the conclusions of the FSAR remam valid.

Startup of an Inactive Reactor Coolant Loop FSAR (15.2.6)

The startup of an inactive reactor coolant loop event is analyzed to demonstrate that the DNB design basis is satisfied. The analysis of the event results in a very slight increase in the RCS pressure and mandary system pressure never approaches the MSSV set-pressure. Therefore, an MSSV tolerance ofi3% can be supported and the conclusions of the FSAR remain valid.

Loss of External Electrical Load and/or Turbine Trip (FSAR 15.2.7)

The loss of external electrical load / turbine trip event is analyzed with MSSVs at a pressure greater than the lowest set-pressure plus 3% accumulation plus 3% tolerance. All four cases presented in the FSAR are analyzed, i.e., the nummum feedback cases with and without pressure control and the rnaximum feedback cases with and without pressure control. For the minimum feedback cases, a 0 pcm/*F moderator temperature coefficient (MTC) is assumed instead of the +7 MTC presented in the FSAR. This is justified since the event is analyzed at full power conditions and the plant must have a 0 pcm/ F (or negative) MTC at full power. The analysis with this combination bounds any combination oflower power level with a positive MTC (based on the Technical Specification's MTC requirement that ramps from +7 pcm/*F at 70% rated thermal power to 0 pcm/ F at full power). The results of the analysis show that the mnadary pressure limit is not exceeded Based on the analyses, an increase on the MSSV tolerance to i3% can be supported and the conclusions of the FSAR remam valid.

i Loss of Normal Feedwater events and Loss ofNon-Emergency AC Power to the Plant Auxiliaries (FSAR 15.2.8 and 15.2.9) i i

l The loss of normal feedwater event and loss of non-emergency AC power to the plant auxiliaries are analyzed to demonstrate that the pressurizer does not become water solid. This ensures that these Condition II transients do not create a more limiting event (i.e., Condition III or IV event) due to water relief out the safety valves. With respect to DNB, these events are bounded by the loss of flow event and loss ofload/ turbine trip events. For these events, the staggered MSSV banks are modeled with each bank at its nonunal set-pressure plus 3% accumulation plus 3% tolerance. The staggered MSSV bank model reduced the peak steam generator pressure for the loss of normal feedwater and loss of non-emergency AC power events. Therefore, the MSSV tolerance ofi3%

can be suppoited and the conclusions of the FSAR remam valid.

a Main Steam Safety Valves Page 6 Safety Analysis Excessive Heat Removal Due to r edwater System Malfunctons (FSAR 15.2.10)

The feedwater malfunction event is unaffected by any changes to the MSSV tolerance because this event results in a decrease in the waad=ry pressure. 'Iherefore, the MSSV tolerance ofi3% can be supported and the conclusions of the FSAR remam valid.

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Excessive Load Increase Incident (FSAR 15.2.11) -l I

'Ihe excessive load increase event is unaffected by any changes to the MSSV tolerance because this I event results in a decrease in the mnad=ry pressure. Therefore, the MSSV tolerance ofi3% can l l

be supported and the conclusions of the FSAR remam valid. j l I l

l Accidental Depressurization of the RCS (FSAR 15.2.12) 1  :

The inadvertent cpemng of a pressurizer safety or relief valve event results in a depressurization of j RCS. 'Ihe secondary pressure rises for this event; however, the MSSVs are not challenged.

] 'Iherefore, an MSSV tolerance of13% can be supported and the conclusions of the FSAR remain i

1 valid. 1 1

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Accidental Depressurization of the Main Steam System (FSAR 15.2.13)

The inadvertent opening of a safety valve, relief valve or steam dump in the main steam system

results in a depressurization of main steam system. 'Ihis event results in an initial increase in j

steam flow that decreases during the accident as the steam pressure falls. The MSSVs are not l challenged for this event. Therefore, an MSSV tolerance ofi3% can be supported and the

! conclusions of the FSAR remam valid.

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j Inadvertent Operation of the Emergency Core Cooling System During Power Operation (FSAR 15.2.14) l The inadvertent operation of the ECCS during power operation event is analyzecf to support a non-l conservatism identified in the current licensing basis analysis. 'Ihe event is analyzed to a

demonstrate that sufficient operator action time exists to terminate ECCS flow following a

. spurious safety injection signal to prevent water relief out the pressurizer safety valves. This j ensures that this Condition II trartsient does not create a more limiting event (i.e., Condition III or

IV event) due to water relief out the safety valves. For this transient, the staggered MSSV banks

! are modeled with each bank at its nominal set-pressure plus 3% accumulation plus 3% tolerance.

1 The results show that the pressurizer does not fill before the operator can termmate the ECCS

flow. Peak wnadary pressure is not exceeded for the transient. Therefore, an MSSV tolerance ofi3% can be supported and the conclusions of the FSAR renmin valid.

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l Main Steam Safety Valves Page 7 Safety Analysis - .

l Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (FSAR 15.3.3)

De inadvertent loadmg and operation of a fuel assembly in an improper position does not result in any RCS transient, rather it resuhs in peakmg factor concerns. Therefore, it is not affected by the proposed change in the MSSV tolerance. He MSSV tolerance ofi3% can be supported and the conclusions of the FSAR remam valid.

Complete Loss of Forced Reactor Coolant Flow (FSAR 15.3.4 )

%e complete loss of reactor coolant flow event is analyzed to demonstrate that the DNB design basis is met. The MSSVs are not challenged during this transient. %erefore, an MSSV tolerance ofi3% can be supported and the conclusions of the FSAR remam valid.

Single RCCA Withdrawal at Full Power (FSAR 15.3.6) ne single rod withdrawal at full power event is a DNB event. Changes to the MSSV tolerance do not impact the event.

Rupture of Main Steam Line (FSAR 15.4.2.1) ne steam line break event is unaffected by any changes to the MSSV tolerance because this event results in a decrease in the emndey pressure. %erefore, the MSSV tolerance ofi3% can be supported and the conclusions of the FSAR remain valid.

Major Rupture of a Main Feedwater Pipe (FSAR 15.4.2.2) l De feedline break transient is analyzed to demonstrate that the core remams in a coolable i geometry. This is ensured by demonstrating that no hot leg boiling occurs prior to the time that the l auxiliary feaiwater system heat removal capability exceeds the stored energy, decay heat and RCS

pump heat (for the case with offsite power) levels. For this event, the staggered MSSV banks are i modeled with each bank at its nommal set-pressure plus 3% accumulation plus 3% tolerance. The l 1 results indicate that no hot leg boiling occurs prior to event turn around and peak secondary 1 pressure is not exceeded Therefore, an MSSV tolerance ofi3% can be supported and the .

conclusions of the FSAR remam valid.

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Main Steam Safety Valves Page 8 i Safety Analysis Reactor Coolant Pump Shan Seizure (Locked Rotor) and Reactor Coolant Pump ShaR Break i

I (FSAR 15.4.4) 4 l 1 1 'Ihe reactor coolant pump shaft seizure and shaA break are analyzed as one event. It is assumed I

! that when RCS flow in the faulted loop is positive, the rotor is locked, and when RCS flow in the 1

_ faulted loop is negative, the rotor is f.x to spin (shaft break). The MSSVs are not challenged

during this transient. Therefore, an MSSV tolerance ofi3% can be supported and the conclusions J of the FSAR remam valid.

l l Rupture of a Control Rod Drive Mechanism Housing (RCCA Ejection) (FSAR 15.4.6) i, j The rod ejection accident is analyzed to hmhne that applicable Condition IV criteria are i

! satisfied. The analysis presented in the FSAR exammes the peak clad temperature, fuel enthalpics, l 1 fuel melting and zirc-water reactions at the " hot spot" in the core. The MSSVs are not modeled as )

a part of the RCS overpressure analysis since the heat addition and overpressurization occur so i

. rapidly. However, should the MSSVs activate following a reactor trip, adequate relief capacity j exists to prevent overpressurization of the emadary side. Therefore, an MSSV tolerance ofi3%

can be supported and the conclusions of the FSAR remam valid.

l Steam Line Break Mass and Energy Releases - Outside Containment a

'Ihe steamline break mass and energy releases outside contamment are not affected because the

transient results in a decrease in the secondary pressure. The MSSVs are not challenged during  !

I l this transient. 'Iherefore, an MSSV tolerance ofi3% can be supported and the conclusions of the FSAR remain valid.

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j Evaluation for Inoperable MSSV(s) i l Introduction i

I Westinghouse identified in Nuclear Safety Advisory Letter NSAL-94-001 a deficiency in the basis i for Technical Specification 3.7.1.1. The Technical Specification allows the plant to operate at a l l reduced power level with a reduced number of operable MSSVs. The deficiency is in the 3 assumption that the maximum allowable initial power level is a linear function of the available

] MSSV capacity. The linear function is identified in the bases section for this Technical l Specification.

One of the options presented in NSAL-94-001 is to perform plant-specific analyses of the Loss of Loadffurbine Trip (LOI/IT) event to analytically determine the maximum allowed power level for a given number ofinoperable MSSVs. Depending on key specific plant parameters, these j analyses may be able tojustify the continud ralidity of the current Technical Specification. A i plant-specific analysis has be performed for Farley Nuclear Plant. This evaluation prosides the j results of this effort.

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i Main Steam Safety Valves Page 9 Safety Analysis i Evaluation i

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{ Technical Specification Table 3.7-1 indicates the maxunum power level at which a plant can j operate, based on a reduction in the high neutron flux reactor trip setpoint, for 1, 2, or 3 inoperable

! MSSVs on any loop. The most limiting transient for this condition for the Farley units is the i LOIIIT event presented in Section 15.2.7 of the FSAR since this transient resuks in the greatest

! demand on the relief capacity of the MSSVs. Therefore, the maximum power level at which the l plant may safely operate, with inoperable MSSVs, may be deternuned by i.woi h ting that a j LOLflT incident initiated from the assumed power level, will not result in the secondary steam a pressure exceedmg 110% of the design pressure. For Farley Nuclear Plant, the pressure limit is j 1208.5 pia ((1085 psig

• 1.1) + 15 psi}.

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! The analysis supporting revisions to Table 3.7-1 uses a more detailed MSSV relief model l

} compared to the lump model historically employed. The Farley units each have 5 MSSVs per l

} loop, or 5 banks For each bank, the analysis conservatively models steam relief to begin at the l j valve set-pressure plus 3% of the set-pressure (tolerance) plus 20 psi due to the pressure drop from

! the steam generator to the location of the MSSVs on the main steamlines. The latter is to account

] for the difference between the actual bank location and the analysis model which uses the steam

generator pressure as the reference pressure for MSSV bank actuation. Valve accumulation is also

} typically modeled by a linear ramp in flow from the initial valve opening pressure to full flow 3%

i above the openmg pressure. However, when considering a 3% tolerance,3% accumulation, and 20 i psi pressure drop, full valve flow will not be attained in banks 4 and 5 prior to exceeding the

, analysis pressure limit. For banks 4 and 5 a revised accumulation requirement is modeled. Bank 4 j is assumed to be fully accumulated over a 2% (of the valve set-pressure) ramp and bank 5 is

! assumed to fully accumulate over a 10 psi ramp. This modeling is consistent with MSSV test l resuhs (Reference A) for actuation at higher pressures.

l i Multiple cases of the LOl/IT event, initiated from less than full power, and assummg either 1,2 i or 3 inoperable MSSVs per loop have been analyzed. The moderator temperature coefficient (MTC) was also varied according to the Technical Specification requirement (+7 pcm/'F 170%

rated thermal power and ramping to O pcm/*F at 100% rated thermal power) to bound the allowed j operation conditions of the plant. The LOI/IT analysis credits reactor trip signals on high neutron i flux (HNF), OTAT, and Low-Low Steam Generator Level (LLSOL). In general, when the MTC

. is greater than +3 pcm/* F, the power level rises until the assumed HNF setpoint is reached. This

occurs relatively early in the transient. For values of MTC less than +3 pcm/ F, a reactor trip signal is generated by either the OTAT or LLSGL functions. Note the OTAT function prosides a l

j reactor trip at power levels supported with only 1 inoperable MSSV per loop. At the further l reduced power levels required with 2 or 3 inoperable MSSVs per loop, an OTAT signal may not be j generated and protection for the lower MTC cases (<+3) is provided by the LLSGL function.

j The analysis assumes a HNF setpoint equal to the proposed Technical Specification limit plus 9%

uncertainty and an initial power level equal to the proposed Technical Specification HNF setpoint.

The reduced power LOI/IT analysis, perfor ned with 1,2, or 3 inoperable MSSVs supports the i

following revised HNF setpoints for Technical Specification Table 3.7-1 by demonstrating that the i

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Main Steam Safety Valves Page 10 Safety Analysis peak steam generator and RCS pressures do not exceed 110% of their respective design values, the pressurizer will not reach a water solid condition, and the DNBR limit is not violated.

Maximum Number ofInoperable Maxunum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operatina Steam Generator (Percent of Rated Hermal Power) 1 87 2 48 3 28 This evaluation, consistent with the Technical Specifications, does not support opention with 4 or more inoperable MSSV on any single loop.

Conclusions A plant-specific analysis of the LOI/IT event has been performed for the Farley units to determme acceptable maximum reduced power levels for operation with 1,2, or 3 inoperable MSSVs per loop. De LOI/IT event is the limiting transient for Farley Units 1 and 2 with respect to MSSV relief capacity and over-pressurization of the mad =y steam system. The analysis demonstrates that operation at or below power levels governed by the above reduced HNF setpoints will not result in exceeding the safety analysis criteiia should a LOL/IT event be initiated from these conditions.

Non-LOCA Conclusions The analyses and evaluations documented within demonstrate that with respect to the non-LOCA safety analyses, an MSSV tolerance ofi3% and modification to the high nuclear flux setpoints with MSSVs out of service can be supported for the Farley Units.

LOCA and LOCA Related Evaluations Large Break LOCA i

The current large break LOCA analyses for J. M. Farley Units 1 and 2 were performed with the

NRC approved Evaluation Model using BASH. After a postulated large break LOCA occurs, the heat transfer between the reactor coolant system (Reactor Coolant System) and the secondary i system may be in either direction, depending on the relative temperatures. In the case of continued i heat addition to the secondary system, the secondary system pressure increases and the MSSVs

! may actuate to limit the pressure. However, this does not occur in the large break evaluation I

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Safety Analysis

} model since no credit is taken for auxiliary feedwater actuation. Consequently, the secondary I system acts as a heat source in the plat ~i large break LOCA transient and the secondary

] pressure does not increase. Since the ==iary system pressure does not increase, it is not j necessary to model the MSSV setpomt in the large break evaluation model. Therefore, an increase

in the allowable MSSV setpoint tolerance for Farley Units 1 and 2 will not impact the current FSARlarge break LOCA analyses.

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! Small Break LOCA t

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] The small break LOCA analyses for Farley Units 1 and 2 were performed with the NRC approved Evaluation Model using the NOTRUMP code. After a postulated small break LOCA occurs, the 4

j heat transfer between the RCS and the Wary system may be in either direction dependmg on j the relative temperatures. In the case of continued heat addition to the secondary system, the 1 secondary system pressure increases which leads to steam relief via the MSSVs. In the small

( break LOCA, the ==iary flow aids in the reduction of RCS pressure. He current licensing-i basis small break LOCA analysis for Farley was performed using a MSSV setpoint tolerance of 3% Therefore, the increase has already been included in the analysis.

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t Post-LOCA Long Term Core Cooling i 1

! l The Westmghouse licensing position for satisfying the requirements of 10 CFR 50.46 Paragraph l

(b), Item (5), "Long Term Cooling," concludes that the reactor will remam shut down by borated i 4

ECCS water residing in the RCS/ sump after a LOCA. Since credit for the control rods is not l taken for a large break LOCA, the borated ECCS water provided by the accumulators and the 1 RWST must have a boron concentration that, when mixed with other water sources, will result in j the reactor core remammg suberitical assummg all control rods out. De calculation is based upon j the reactor steady state conditions at the initiation of a LOCA and considers sources of both borated and unborated fluid in the post-LOCA containment sump. The steady state conditions are l obtamed from the large break LOCA analysis which, as stated above, does not take credit for j MSSV actuation. Dus the post-LOCA long-term core cooling evaluation is independent of the l l MSSV setpoint tolerance, and there will be no change in the calculated RCS/ sump boron

! concentration after a postulated LOCA for Farley Units I and 2.

Hot Leg Switchover to Prevent Potential Boron Precipitation i

i Post-LOCA hot leg recirculation time is determined for inclusion in emergency operating j procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. His j time is dependent on power level, and the RCS, RWST, and accumulator water volumes and their j associated boron concentrations. He proposed i3% increased MSSV setpoint tolerance does not j affect either the power level or the boron concentrations assumed for the RCS, RWST and j accumulator in the hot leg switchover calculation for J. M. Farley Units 1 and 2.

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Main Steam Scfety Valves Page 12 SafetyAnalysis LOCA Hydraulic Forcing Functions l The peak hydraulic forcing functions on the reactor vessel and internals occur very early in the large break LOCA transient. Typically, the peak forcing functions occur between 10 and 50 milliseconds (0.01 and 0.05 seconds) and have subsided well before 500 milliseconds (0.50 . l seconds). Any change in time associated with an increased MSSV setpomt tolerance would occur j several seconds into the transient. Since the LOCA hydraulic forcing functions have peaked and subsided before the time at wiuch the MSSV may actuate, the increase in the MSSV setpoint tolerance to 3% will not impact the LOCA hydraulic forcing functions calculation for Farley Units 1 and 2. .l 1

LOCA Conclusions The effect ofincreasing the MSSV setpoint tolerance to i3% for J. M. Farley Units I and 2 has <

been evaluated for each of the LOCA related analyses addressed in the FSAR. It was shown that l the i3% MSSV setpoint tolerance does not result in any design or regulatory limit being exceeded for operation. Therefore, with respect to the LOCA analyses, it can be concluded that increasing the MSSV setpoint tolerance to 13% for J. M. Farley Units 1 and 2 will be acceptable from the l standpoint of the FSAR accident analysis discussed in the safety evaluation.

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! The impact ofincreasing the Technical Specification allowable drift tolerance to 3% on the l MSSVs was evaluated with respect to the Emergency Operatmg Procedures (EOPs) since lift i

setpoints for the MSSVs are utilized as EOP setpoints. It is noted that although this change will increase the allowable drift of the MSSVs, the valves will continue to be set at a 1% measured, as-left tolerance. This evaluation was based on the setpomt requirements (as footnoted) of the Westinghouse Owners Group Emergency Response Guidelines.

The following EOP setpoints which utilize MSSV lift setpoints were reviewed and evaluated:

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1. Steam generator pressure for highest steamline safety valve; i

! 2. Steam generator pressure for lowest steamline safety valve; l 3. Setpoint for steam generator PORV contraller (typically 25 psig below lowest safety valve set-j pressure);

)- 4. Setpoint for steam generator PORV controller (typically 25 psig below lowest safety valve set-4 pressure), minus 25 psi; and

! 5. RCP trip parameter and setpoint.

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!... i Main Steam Safety Vches Page 13 Safety Analysis  ;

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For the steam generator pressure values correspondmg to the highest and lowest steamline safety valves, the MSSV set-pressures are used on the Heat Sink Critical Safety Function Status Tree to determme the appropriate procedure for implementation. 'Ihe nonunal value for the highest steamline safety valve setpoint is used to determme if a steam generator overpressure condition exists while the nominal value for the lonst steamline safety valve setpoint is used to identify a loss of normal steam release capability. For each of these setpoints, the increase in the allowable setpoint drift tolerances (i 3%) between the EOP MSSV setpoints and the MSSV in-plant lift l pressures will not impact the use of the setpoints in the EOPs. If the set-pressures are within 3%,

no modifications to the setpoints used in the EOPs will be required.

The setpoint of the steam generator PORV controller is used in the E-3, Steam Generator Tube Rupture guideline to isolate flow from the ruptured steam generators. The 25 psi margin below the lowest safety valve set-pressure is a typical generic value that is low enough to allow for the opening of the steam generator PORV prior to lifting the safety valve and high enough to stay above the no-load steam generator pressure value in order to mimnuze atmospheric releases from the ruptured steam generator. Even if the MSSV lift setpoint drifts to the new allowable limit from the 1% tolerance to which they are set, the actual Farley EOP setpoint value for the steam generator PORV controller will ensure that the steam generator PORV will be opened prior to lifting the MSSV. Therefore, an increase in the MSSV setpoint drift tolerance from 1%to 3%

will not impact this setpoint.

The setpoint for the steam generator PORV controller minus 25 psi is used in the steam generator tube rupture recovery guidelines as a setpoint for stopping feed flow to a ruptured steam generator as the steam bubble is compressed. The 25 psi margin in addition to the setpoint discussed above allows for operator action without opening the steam generator PORV. With these operating margins, an increase in the MSSV setpoint drift tolerance from 1% toi 3% will not impact this setpoint.

For the RCP trip parameter and setpoint, the lowest MSSV lift pressure is used as input for the determination of when to trip the RCPs in the EOPs based on RCS pressure. This determination is conservative since a generic 3% partial open and tolerance error is assumed and instrument uncertainties are taken into account. With this conservatism and a small difference between the MSSV pressure to determine the RCP trip setpoint and an in-plant first lift pressure ofless than 3%, there is no impact on the EOPs in this area.

In summary, the revision to the Technical Specification allowable drift tolerance of 3% on the I MSSV lift setpoints will not effect the EOP setpoints listed above. The current EOP setpoints remain valid.

Main Steam Safety Valves Page 14 Safety Analysis

't SGTA

< Steam Generator Tube Rupture Evaluation An evaluation of the steam generator tube rupture (SGTR) has been performed to determine the effect of the increase in MSSV tolerance.

De SGTR analysis performed for Farley Units I and 2, presented in Section 15.4 of the FSAR, was performed to ensure that the*offsite radiation doses remam below ("a small fraction") the limits defmed in 10CFR100. The primary thermal and hydraulic parameters which affect the calculation of the offsite radiation doses for an SGTR are the amount of radioactivity assumed to be available  ;

in the reactor coolant, the amount of reactor coolant transferred to the secondary side of the ruptured steam generator through the ruptured tube, and the amount of steam released from the ruptured steam generator to the atmosphere.

1 For the FSAR SGTR analysis, the activity in the reactor coolant is based on an assumption of 1% l defective fuel, and this assumption will not be affected by the increase in MSSV tolerance. Thus,  !

1 an evaluation was performed to determme the effect of the increased MSSV tolerance on the primary to mnadary break flow and the amount of steam released to the atmosphere.

A design basis failure of a single steam generator tube was evaluated using the assumptions which were utilized in the FSAR SGTR analysis. An SGTR results in a loss of coolant inventory, and reactor trip and safety injection (SI) are assumed to occur on a low pressurizer pressure signal. I After reactor trip and SI actuation, it is assumed that the manary side pressure stabilizes at the 1 MSSV setpoint minus the tolerance (3%). It is assumed that the RCS pressure stabilizes at the equilibrium value where the incoming SI flow rate balances the tube rupture break flow rate, which is dependent on the pnmary to wnadary side pressure differential, ne resultant equilibrium

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break flow rate is assumed to persist from the time of reactor trip and SI actuation until 30 minutes after the accident. A maximum Si flowrate is conservatively assumed for the design basis SGTR analysis in order to maximize the break flow.

The results of the thermal and hydraulic evaluation of the SGTR for the increased MSSV tolerance l indicate that the primary to wnnary break flow of 137,811 pounds would increase by 0.9% to 139,100 pounds. The amount of steam released for the reduction in auxiliary feedwater flow of 64,488 pounds is increased by 1.6% to 65,500 pounds These results were used in calculating the effect of the increased MSSV tolerance on the offsite radiological consequences (see below).

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Main Steam Safety Velves Page 15 l Safety Analysis

! P=dialnaimi Consequences

An analysis for the SGTR and evaluations for other radiological events were performed based on a 13% tolerance on the MSSVs. He FSAR analyses for Loss of Load / Turbine Trip, Loss of Offsite

! Power, Locked Rotor, Main Steamline Break, and Rod Ejection remam boundmg for this proposed

configuration. Only the SGTR required reanalysis.

4 l The Steam Generator Tube Rupture offsite doses were reanalyzed for the change in faulted steam '

generator steam releases and break flow due to the increase in tolerance. De offsite doses were l calculated to be

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H yroid Whole S_ kin Dgic Body Dose Dgs; l

i' SB 3.8 rem 0.14 rem 0.17 rem i 4 LPZ l.6 rem 0.05 rem 0.06 rem I i

he doses remain within the NRC dose acceptance criteria of 10CFR100.

j CONCLUSION l

ne evaluation shows that allowing a setpoint tolerance of 3% does not result in any ASME code I

related problems with the MSSVs. Resetting the setpoint to il% reduces the concerns for valve j leakage, operating in an unanalyzed condition, and set-pressure overlap. In addition, the steam

.; pressures resulting from a i3% setpoint tolerance are enveloped by existing design transients. All FSAR accident analyses continue to meet their respective acceptance criteria.

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REFERENCE

! a. Full Flow Certification Testing of 15 Dresser Main Steam Safety Valves for Farley Nuclear Plant, Unit II, Wyle Test Report J/N 42539.

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