Text

Proposed Tech Specs Re PORV & Block Valve Reliability & LTOP for LWRs
	ML20115C890
Person / Time
Site:	Farley
Issue date:	10/13/1992
From:	; SOUTHERN NUCLEAR OPERATING CO.
To:	;
Shared Package
ML20115C874	List: ML20115C869; ML20115C890;
References
	NUDOCS 9210200124
	Download: ML20115C890 (52)
	v • d • e

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REACTOR COOLANT SYSTEM l

1/,,,4 a 4. 5 RELIEF VALVES-LIMITING CONDITION FOR OPERATION OIW Ibw&R.canito 3.4.5 All me relief valves (PORVs) and thstr associated block valves shall be OPERABLE.

APPLICA8ILITY: MODES 1, 2, and 3.

ACTION:

M e C/ Eifh55<d J64Tltde:Ec With one orBoTHeeee-PORV(s) inoperable { within -1 hour either restore-a.

the PORV(s) to OPERA 8LE status-or close the--associated block valve (s) AWEa6e -

A ddddb D and 7:::v: ;:=r fn: the-b_ lock valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in 40@ fHUTDOWN within the following 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months . WoT 0Ahtti ja w 6 _ _,

t. t!Rh ca er ::n block-valv:(:) i=;:r:bl*r-withir.1 M2" +4%eFI rester the bleek v:1v:(:)- te OPE"J"L&44+te: _ r c!Me th b!Mh t AtRE( , yrs valve (s) eM rerve peeer f.~= the b1ME velve(s); etMMee, M w. seer. >

z6 in :t leset TT ST.'?DSY eithi- Se ret 4-heure W 1- COLA

-6HUTRM-within t% fellwing 30 h:;rev ,

f. + The provisions of Specification 3.0.4 are not aplicable.

SURVEILLANCE REQUIREMENTS 4.4.5.1 Each.PORY:shall be demonstrated 0PERA8LE at-least once per 18 months by g-fe-ence ef_.aquayun cit;gpf7;cy g pereggy tw ye.)ye t%;;g en M ele ef fe!! trew),.

=-

4.4.5.2 Each block valve shall be demonstrated CPERABLE at least.once per 92' days by operating the valve through one. complete cycle of full travel unless the block valve is closed et2 the ; =r anve4 in order to meet the ACTIOM re;Uiterrete ef 2. Hv:. MaI46M6V73 or la7a/ 6 Oc C. /A/: -

SRt.JFK4770n] Sd.5.

( d. Eharn ka 7% RW%eoJ6H Ok 6MRGb [k& QC Rh.L Tt2pk LkJLtLha NobG 3 *R 4 j AA/O g b. OPGL9npl) 77tE PnRVTttRov4H ONE C&Pt ETE CYCLE or= Pol. t- 712AV5L l VSINC3 THE 8AcicGP PoAW CONTRCL. SYSTeAA, AND

[0. 5FetM1A1:. A CM$AJANL OLi/SEATtodoc 7.s tizac,ovlitsitutasi./isricvl FARLEY-UNIT-1 3/4 4-8 MENDMENT NO. 46-9210200124 921013 PDR ADOCK 05000348

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INSERT la O

b. With one PORV inoperable due to_causes other than excessive seat- -

leakage, within-1 hour either-restore the PORV to OPERABLE status or close its associated block valve- and-remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . within the next 6-hours.and in HOT SHUTDOWN _within the

c. With both PORVs inoperable due to causes-other than excessive seat-leakage, wi,hin I hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the ncxt 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 'and in HOT-SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

' INSERT lh

d. With one block valve -inoperable,.within I hour restore the block valve to OPERABLE status-or place its associated PORV in manual control.

Restore the inoperable block valve to OPERABLE status within 72' houre or be in at least HOT STAND 2Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT-SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

e. With both block valves inoperable, within _1 hour restore both block .

valves to OPERABLE-status or-place their associated PORV-in manual-control.

Restors at least one block valve to OPERABLE-status withini

-the next hour.and restore the remainino inoperable block valve-to OPERABLE status within-72 hours; othercue, .bc in at least HOT! STANDBY within hours.

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHU100WN within the-following 6 4

S 1

} _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ " ~

-l l

REACTOR COOLANT SYSTEM 3

BASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES ,_

The pressurizer code safety valves operate to prevent the RCS frne being pressurized above'its Safety Limit of 2735 psig. Each safety valve is designed to relieve 345,000 lbs per hour of saturated steam at tto valve set point. The relief capacity of a single safety valve is adequate' to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability-and will-prevent RCS overpres- --

surization. In addition, the Overpressure Protection System provides a

diverse means of: protection against RCS overpressurization at low temper;tures.'

During operation, all pressuriz'er code safety valves must be OPERABLE te prevent the RCS from being pressurized above its safety limit of 2735 psig. -

The combined relief capacity-of all of .these valves'is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached:

(i.e. , no credit is taken for a' direct reactor _ trip-on the loss of load) and -

also assuming no operation of the power operated relief valves or steam dump i valves,

. Demonstration of the safety valves' lift _ settings will occer only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Soiler and Pressure Code.

3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintaine( within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance is sufficient to ensure that the parameter -

is restored to within-its limit following expected transient operation. The

maximum water volume also ensures that a steen bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a ainimue number of pressurizer heaters be OPERA 8LE enhances the capability of the plant.

to control Reactor Coolant Pressure and establish natural circulation.-

3/4.4.5 RELIEF VALVES (PORV's)

The power operated relief valves- and steam bubble function to relieve RCS :

pressure during all design transients up ta and including.the design step load -

? decrease with steam dump. Operation of the PORV's minimizes the undesirable opening of the: spring-loaded pressurizer code safety valves. Esch N" 5e? e

\

E

z: tely ^pected blech velve u ; ev44e e pe:itive : buff ::-fility 20:1d 4-me14*f valve br- * :;;-210.D / .

lh{$&eT 2 FARLEY-UNIT 1 8 3/4 4-2 m g, y

INSERT 2 The OPERABILITY of the PORVs and-block valves is determined on the basis of their being capable of performing the following functions:

f A. Manual control of PORVs to control reactor coolant system pressure. This-is a function that is used for the steam generator tube rupture accident and for plant shutdown.

B. Maintaining the integrity of the reactcr coolant pressure boundary. This is a function that is related to controlling ideiitified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage..

C. Manual control of the block valve to: (1) unblock. an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item A), and (2) isolate a PORV with excessive seat leakage (Item B).

D. Manual control of a block valve to isolate a stuck-open PORV.

The Surveillance Requirements found.in Specification 4.4.5.1 for the PORVs and Specification 4.4.5.2 for tLe block valves-improve reliability and provide the assurance that the PORVs and block valves can perform their functions. The PORVs are stroke tested during MODES 3 or 4 with the associated block valves closed in order to eliminate uncertainty introduced by testing the PORVs at lesser system temperatures than ex)ected during actual operating conditions.

The block valves are exempt from tie surveillance requirements to cycle the valves when they have- been closed to comply with the ACTION requirements. This.

precludes the need to cycle .the' valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORV to operable status. Surveillance requirement 4.4.5.1.b includes testing which demonstrates.

the functionality of the backup _ PORV control system.

Unit 1 Typed Pages t

v - - - - ~ , - e . ,

.i REACTOR COOLANT SYSTEM 3/4.4.5 REllEF VALVES LIMITING CONDITION FOR OPERATION 3.4.5 Both power-operated relief valves (PORVs) and their associated block l valves shall be OPERABLE.

APPLICABILITY: H0 DES 1, 2, and 3.

ACTION:

a. With one or both PORVs inoperable because of excessive seat leakage,.

within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with p%er maintained to the block valve (s); otherwise, be in at least iiOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With one PORV inoperable due to causes other than excessive seat leakage, within I hour either restore the PORV to OPERABLE status or-close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in. HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With both PORVs in'perable due to causes other than excessive Feat leakage, within I hour either restore at le':st one PORV.to OPERABLE status or close its asscciated block' valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With one block valve inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore -the Llock valve to.0PERABLE status or place its associated PORV. in manual control. Restore the inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at 'teast H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

c. With both block valves inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore hth block valves to CPERACLE status or place their-associated PORV % manual control. Restore at least one block valve to OPERABLE status within the next' hour and restore the remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

f. The provisions of Specification 3.0.4 are not applicable.

FARLEY-UNIT 1 3/4 4-8 AMENDMENT N0.

REACTOR COOLANT SYSTEM SVRVElllANCE RE0VIREMENTS 4.4.5.1 Each PORY shall be demonst. stan 0PERABLE at least once per 18 months by:

a. Operating the PORV through one complete cycle of full travel during MODE 3 or 4, and

b. Operating.the PORV through one complete cycle of full travel using

. the backup PORV control system, and

c. Performing a CHANNEL CAllBRATION of the actuation' instrumentation.

4.4.5.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by. operating the valve through one complete cycle of full- travel unless the block valve is closed in order to meet the requirements of ACTION b or c in Specification 3.4.5.

l r

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

. . - . . . . .- .- - .~ . . .. .

y a

RL@ COOL ANT SYST@

, BASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS' from being

pressurized above its Safety Limit of 2735 psig. Each safety- valve is -

designed to relieve 345,000 -lbs per hour of saturated steam at the valve set

! point. The relief capacity _of a single: safety valve is adequate to- relieve-

' any overpressure condition which-could occur during-shutdown. In the event that-no safety valves are OPERABL'i, an operating RHR loop, connected.to the:

RCS, provides overpressure relief _ tigtbility and will prevent RCS overpres-surization. In addition. the.0verpressure Protection System provides:a-diverse means of protection against RCS overpressurization attlow-temperatures.

During operation, allipressurizer code safety valves must be OPERABLE to e revent the RCS from being pressurized above its safety limit of-2735'psig.

, The combined relief _ capacity of all of these, valves is-greater than the maximum surge _ rate resulting: from a. complete loss of load assuming no reactor

' trip until the first Reactor Protective-System-trip set point is reached !

'(i.e., no credit'is taken for a Ldirect reactor trip on the loss of load) and _

also assuming no operation off the power operatud 're'ief valves lor. steam dump.

valves. ,

Demonstration of the safety valves' lift-settings will. occur only during i

~

shutdown and will be performed in accordance with the provisions of.Section XI of the ASME Boiler and Pressure Code.

3/4.4.4 PRESSURIZER-The limit on the maximum water-volume'in the pressurizerf assures that:the parameter is maintained within the -normal steady- state envelope of operation assumed in the SAR. The limit is-consistent with!the C 'tial SAR assumptions, j- The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillanceLis sufficient to r e that the parameter

is restored to within its. limit'following expected ..istent operation. The maximum water volume also ensures that a steam bubble is formed and thus.the i; RCS is not a hydraulically solid system. The requirement that a. minimum number of pressurizer heaters be OPERABLE enhances the capability of the-plant-1 to control Reactor Coolant Pressure and establish natural 1 circulation.

J' 3/4.4.5 RELIEF VALVES (PORVs)-

4 The power operated relief valves and steam bubbie function to relieve RCS i-pressure during all design' transients up ~torand -ircluding the design step load c decrease with steam dump.- Operation of.the PORVs minimizes; the undesirable

opening of the spring-loaded pressurizer code safety valves. l 5

FARLEY-UNIT 1 B 3/4 4-2 AMENDMENT N0.

e o

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

REACTOR COOLANT' SYSTEM BASES __

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

A. Manual control of PORVs to control reactor coolant system pressure. This is a function that is used for the steam generator _ tube rupture accident and for plant stiutdown.

B. Maintaining the integrity of the reactor. coolant pressure boundary. This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

C. Manual control of the block valve to: (1) unblock an isolated PORV to allow it to be used for manual control of- reactor coolant system pressure (Item A), and 0) isolate a PORV with excessive seat leakage (Item B).

D. Manual control of a block valve to isolate e stuck-open-PORV.

The Surveillance Requirements found in Specification 4.4.5.1 for the PORVs -

and Specification 4.4.5.2 for the block valves improve reliability and provide the assurance that the PORVs and block valves can 3erform their functions. The-PORVs are stroke tested during MODES 3 or 4 with tie associated block valves closed in order to eliminate uncertainty introduced by testing the PORVs at lesser system temperatures than expected during actual operating conditions.

The block valves are exempt from the surveillance requirements to cycle the valves when they_have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance-is being performed to restore an -inoperable PORV to operable

status. Surveillance requirement 4.4.5.1.b includes testing which demonstrates the functionality-of the backup PORV control system.

FARLEY-UNIT 1 B 3/4 4-2a AMENDMENT N0.

l

_ _ _ _ --- _ _ _ _ - - -- - - - - - - - - J

l-4 Farley Unit 2 Proposed Changed Technical Specification ~ Pages Remove Paog Insert Pace 3/4.4-6 3/4 4-8 3/4 4-8a 0 3/4 4-2 B 3/4 4-2 B 3/4 4-2a ,

=

k l

Unit 2 Marked Pages t

l l

l 1

i l

.I 1

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-r. _ _ _ _ _ _ . . _ . . . . _ _ _ _ _ _ _. . _ . _ _. _ _ _ _ _ _ _

' 1 REACT 081 C00LANT SYSTEN

. 3/4.4.5 RELIEF VALVt5-1 1

LIMITIECONDITION FOR OPERATION

&VW Rwl2.d%2aTED 3.4.5 *11 per relief valves (PORVs) and thir associated block valves

shall t>e OPERA 8LE.

APPL 1CABILITY: MDE51, 2, and 3.

ACTION: )

l &b.uhl O Eft.as'tVL 5.C LL4C2bc '

l EctrH V t

a. With one or-eeee-PORV(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore l

the PORV(s) to CPERABLE status or close the associated block valve (s)eW64 the block valve (s); otherwise, be in-at least l MIA/ AID 7D eM mere ;ner f-^

' SHUTDOWN within the HOT following STAND 8Y.within the next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and 40 hcurs. N in@T

- ItLET .fa 6 ,

+.

W4% 'e .: er ee e blect Yelve(s) *^epereble,M9'a .1 He? ef tM--i RA44 w,Tre

! resters th: bh:k :h:r-) t- CPEPRLE et:te: er el :: t% 5!91 '

/hC.f6 t'alve(s) :nd :::v: en of :- th blect Y !ve(s); ee:We, h l

- :t le::t "0? STAMOSY eitM.: t% next 5 Mer: --d ta-COLD.

i 6"'5W ett'h th f:lledq 30 Mert. .

s b The provisions of Specification 3.0.4 are 'not applicable.

l $URVEILLANCE REQUIREMENTS i 4.4.5.1 Each PORY shall be demonstrated CPERABLE at least once per 18 months

. by:perfe.-a-ce ef e cy>wrt cALI! PAT!09 eM eperati ; te velve er egh no "ycle of full-travel,

---o-F 4.4.5.2 Each block vc1ve shall be demonstrated OPERA 8LE at least once per 92 days by operating the valve through one cos91ste cycle of full travel -

unless the block valve is closed Mt' Se p v r nrv:d in order to meet

! the ACT!cM r: pin:::*: ef e he. St.kjftt.xw73 0 4c77tx/ f, ce d, fAf .

SkCif tGtTiod 9.4. 5.

L f 8. Mfs77A/6 I)%L UVTihC&Ok O.MREt% bkih & S./LL INAL/hd.

) DJ21Als MCM 3 o< 4 blu 4

, - f b. opees ristc, rae pouv raeoosi+ one c0Aretera cycts oproli rynet l VSINf, Tite BA cevP Pod C oNTROL SYSTEM, AtJD

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0 MIAk, L C4Alist 63LIBRATic.x/ oc 7),s hjgizyj /d372JAftd%Tici) i i-A l FARLEY-UNIT 2 3/4 4-8 Anendment No. -M-i

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4 f

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

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

. _. ._ ~ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ . _ . _ - __ _ ___ ..-.._-

N' JX$ERT la-s b. With one PORV inoperable due to causes other than excessive seat leakage, within I hour either restore the PORY to OPERABLE status or close its associated block valve and remove )ower from the block valve; restore the PORV to OPERABLE status within tie following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT _ SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

~

c. With both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in-HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

,LNSERT lb

d. With one block valve inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block valve to OPERABLE status or place its associated PORV in manual centrol.

Restore the inoperable block valve to OPERABLE status withi.n /2 hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e. With both block valves inoperable, within I hour restore both block-valves tr OPERABLE status or place their as weiated PORV in manual control, Restore st least one block valve to OPERABLE status within the next hour and restore the remaining-inoperable block valve to ,

OPERABLE status Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ;'otherwise, be in at least HOT STANDBV within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT-SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i l

! l atacTOR C00 tant $YSTEM i SA$f5 _

l 3/4.4.2 and 1/4.4.3 5AFETY V4LVE5

! The pressurizer code safety valves operate to prevent the RC5 from being L pressurized above its Sahty Limit of *2735 psig. .f ach safety valve is j designed to relieve 345,000 lbs per hour of saturated steam at the val"1 set j

p9 ins. T1.a r*11ef capacity of a sin le safety valve is adequate tr reiteve l i any everpressure condition which cou d occur during thutdown.. In,the event l 1 that no safety valves are OPERABLE, an operating RHR loop, connected to the i RC5, provides ovsrpressura relief capability and will prevent RC5 overpres- l suritation. In addition the Overpressure Protection Systes provides a diverse seans of protection against RC5 overpressurization at low'tasperatures. l

! During operation, all pressuriser code safety valves sust be CPtaA8LE to

! prevent the RC5 from being pressurized above its safety limit of 2735 psig.

l Tha t.osbined relief capacity of all of these valves is greater than the *

! saximum surge rate resulting from a complete loss of lead assuming ne reactor !

l trip until the first Reactor Protective Systes trip set peint is reached l (i.e., no credit is taken for a direct reactor trip on the less of lead) and e also assuming no operation of the power operated relief valves or steam dusp valves.

I Desonstration of the safety valves' lift settings will oc:ur only during *

shutdown and will be performed in accordanca with the provisions of Section XI -

a l of the A5ME Boiler and Pressure Code.

I !

) 3/a.a.4 PetiSUR12tR ,

i *

! The Itait on the saximus water volume in the v.essurizer assures that the l parameter is sainte.ine( within the normal steedy state anvelope of operation assumed. in Le 5AR. The limit is consistent with the initici SAR assumptions.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance is sufficient to ensure that the parameter l

is restored to within its Itait following expected transient operatten. The

+

.maximus water volume also ensures that a steam bubble is formed and thus the l RC5 is net a hydraulically solid systes. The requirement that a sinisus ,

~

i number of pressuriser heaters be OPERABLE enhances the capability of ths. plant i

to control Reacter Coelant Pressure and establish natural,circulati6n.

3/4.4.5 RELIEF VALVE 5 (PORV's)

The power operated relief valves and stvas bubble function to relieve RC5 pressure during pil design transientr, up to and inclcdin the design step load- 4 decrease with steam dump. Operation of the'PORV's sinf a zes the undesirable e ;

j opening of the :pring-loaded pressurizer t ;:^^ 'd: p:-'th: code safety Cut ^'" valves. M he:

d'Mt, 00 M

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t'-d ti:

FARLEY-UNIT 2 8 3/4 4-2 Ascrdsent No.' 4-i

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. 3 l

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

1 INSERi 2 The OPERABILITY of the PORVs and block valves is detennined on the basis of their being capable of performing the following functions:

A. Manual control of PORVs to control reactor coolant system pressure. This is a function that ds used for the steta generator tube rupture accident 4

and for plant shutdown.

I B. Maintaining the integrity of the reactor coolant pressure boundary. This t

is a function that is related to controlling identified leakage and

ensuring tho ability tu detect unidentified reactor coolant pressure

, bour.d:ry leakage.

C. Manual control of the block valve to: (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure 4

(Item A), and (2) isolate a PORV with excessive seat leakage (Item B).

D. Manual control of a block valve to isolate a stuck-open PORV.

1' The Surveillance Requirements found in Specification 4.4.5.1 for the PORVs and Specification 4.4.5.2 for the block valves improve reliability and provide

, the assurance t'.st the PORVs and block valves can 3erform their functions. The 4

PORVs are stroke tested during MODES 3 or 4 with t1e associated block valves closed in order to eliminate uncertainty introduced by testing the PORVs at lesser syctem temperatures than ex)ected during actual operating conditions.

The block valves are exempt from tie serveillance requirements to cycle the valves when they have been closed to comply with the ACT10!1 requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORV to operable status. Surveillance requiropont 4.4.5.1 b includes testing which demonstrates the functionality of the backup PORV control system.

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REACTOR COOLANT SYSTEM 3/4.4.5 REllEF VALVES LIMITING CONDITION FOR OPERATION 3.4.5 Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABIt111t H0 DES 1, 2, and 3.

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a. With one or both PORVs inoperable because of excessive seat leakage, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block nive(s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN Mthin the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

h. With one PORV inoperable duc to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to OPERADLE status or close its associated block valvo and remove )ower frota the block valve; reatore the PORV to OPERABLE status within tie following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 h e s,

c. With both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With one block valve inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block valve to OPERABLE status or place its associated PORV in manual control. ,

Restore the inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e. With both block valves inoperable, within I hour restore both block valves to OPERABLE status or place their associated PORV in manual control. Restore at least one block valve to OPERABLE status within

, the next hour and restore the remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following_6 hours.

f. The provisions of Specification 3.0.4 are not applicable.

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

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4 REACTOR COOLANT SYST[8 :

SURVEllLANCE REQUIREMENTS ,

4.4.5.1 Each PORV shall be demonstrated OPERABLE at least once per 18 months by:

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a. Operating the PORV through one complete cycle of full travel during MODES 3 or 4, and

b. Operating the PORV through one complete cycle of full travel using the backup PORV control system, and

c. Performing a CHANNEL CAllBRATION of the actuation instrumentation.

4.4.5.2 E:ch block valve shall be demonstrated OPERABLE at least or.ce per 92 days by operating the valve through one complete cycle of full travel 4

unless the bic.k valve is closed in order to meet the requirements of ACTION b or c in Specification 3.4.5.

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FARLEY-UNii 2 3/4 4 8a -AMENDMENT NO..

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1 REACTOR COOLANT SYSTEM

! BASES i.

j 3/4.4.2 and 3/4.4.3 SAFETY VALVES i: The pressurizer code safety valves operate to prevent the RCS from being

pressurized above its Safety Limit of 2735 psig. Each safety valve is

. designed to relitve 345,000 lbs per hour of saturated steam at the valve set 3

point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop,L connected to the-

{ RCS, provides overpressure relief- capability and will prevent RCS overpres-

6. surization, in addition, the Overpressure Protection System provides a

diverse means of protection-against RCS overpressurization at low temperatures.

During operation, all pressurizer. code safety valves must be-OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the 3 maximum surge rate resulting from a complete loss of. load assuminC no reactor- l

trip until the first Reactor Protective-System trip set )oint is reached -

i (i.e., no credit is taken for a direct reactor trip on tie loss of load) and

also assuming no operation of the power operated relief valves or steam dump -

valves.

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Demonstration of the safety. valves' lift settings will occur only during shutdcwn and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. .

[ i 3/4.4.4 PRESSURIZER i t

The limit on the maximum water volume in the pressurizer assures that the: [

parameter is maintained within ti,e normal steady state-envelope of operation 1 assumed in the SAR. The limit is consistent with the initia! SAR assumptions. .

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance is sufficient to ensure that-the parameter

! is restored to within its limit following expected transient operation. The i maximum water volume also ensures that a steam bubble is formed and thus the. '

l RCS is not s hydraulically solid system. The requirement that a minimum '

l number-of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant Pressure and establish natural circulation. r

j. .3/4.4.5 -RELIEF VALVES (PORVs)

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The power. operated relief valves and steam. bubble function' to relieve RCS pressure during-all design ;ransients. up to and-including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable wening of .the spring-1mded pressurizer code safety valves. l-FARLEY-UNIT 2. B 3/4 4-2' AMENDMENT NO.-

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

A. Manual control of PORVs to cuntrol reactor coolant system pressure. This is a function that is used for the steam generator tube rupture accident and for plant shutdown. This function has been classified as safety related for more recent plant designs.

B. Maintaining the integrity of the reactor coolant pressure boundary. This is a function that is related to controlling identified leakage and 4 ensuring the ability to detect unidentified reactor coolant pressure boun4 ry leakage, r '

i C. Manual control of.the block valve to: (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item A), and (2) isolate a PORV with excessive seat leakage (Item B).

D. Manual control of a block valve to isolate a stuck-open PORV.

4 The Surveillance Regtiirements found in Specification 4.4.5.1 for the PORVs ,

i and Specification 4.4.5.2 for the block valves-improve reliability and provide

the assurance that the PORVs and block valves can 3erform their functions. The PORVs are stroke tested during MODES 3 or 4 with tie associated block valves closed in order to eliminato uncertainty introduced by testing the PORVs at

lesser system temperatures than ex;'ected durii.g actual operating conditions.

The block valves are exempt from tie surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with tull system differential pressure or when maintenance is being performed to restore an inoperable PORV to operable status. Surveillance requirement 4.4.5.1.b includes testing which demonstrates the functionality of the backup PORV control system.-

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

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4 Attachment 2 Safety Analysis i for Proposed Technical Specification Changes Associated With Power-herated Relief Valve And Block Valve Reliability 4

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Joseph H. f arley Nuclea- Plant Units 1 and 2

Technical Specification Changes Associated With Generic letter 90-06 Power-Operated Relief Valve and Block Valve Reliability l Safety Analysis i

i Proposed Chanae

Revise Farley Limiting Condition for Operation 3.4.5, Surveillance Requirements 4.4.5.1 and 4.4.5.2, and Bases 3/4.4.5 to incorporate the following changes

l. Revised Limiting Conditi m for Operation (LCO) Action Statement "a."

i ~ for Technical Specification 3.4.5 to specify that power be maintained ,

to a block valve which:is closed due-to-its associated PORY being

inoperable due to excessive seat leakage. Additionally, revise the shutdown requirement for Action Statement "a." to require that tne urit 1 be placed in %t Shutdown within the f9110 wing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after reaching.

i Hot Stondby.

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2. Added LCO Action Statement "b." for Technical Specification 3.4.5 to-l specify that power be removed from a block valve that is closed due to its associated PORV.being inoperable for rcasons other than excessive seat leakage. The Action-Statement is applicable when only one PORV is ino)erable and requirss that the PORV be. restored to operable status 3 wit 1in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in Hot Standby within the ne.tt _6 hours.and-in Hot j Shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3. Added LCO Action Statement "c." for Technical Specification 3.4.5 to specify that in the event that both PORVs are inoperable for reasons- .

other than excessive-seat leakage,. restore at'least one PORV to operable status or close its associated block valve and remove power from the block valve and be in Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and: '

l in Hot Shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ..

i i- 4. Added LCO Action Statement "d " for Technical Specification 3.4.5 to specify that in the event that one block valve is inoperable, restore the block valve to operable status within I hour or place the associated PORY in manual: control. -The Action Stater.ent requires:that l the inoperable block valve be restored to operable status within 72-hours or be. in at least Hot- Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Hot :

Shutdown within the following_6 hours, 2-

5. AddedLCOAction-Statement"e."forTechnicall Specification 3.4.5to-specil'y that in the event that both' block valves are inoperable, L -restore both block valves to operable status or place their associated

-PORV-in manual Tontrol:within I hour.- Restore at least one< block valve ,

l to'o)erable'within the next hour and restore the remaining' inoperable

bloc ( valve to operable _ status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at-L least Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months andzin. Hot-Shutdown within L the following 6_ hours. ,

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Safety Analysis Page 2

6. Revised Surveillance Requirement 4.4.5.1 to reflect the requirement of Generic Letter 90-06 to operate the PORVs through one full cycle of travel at least once per 18 months while in Modes 3 or 4. In addition, added Surveillance Requirement 4.4.5.1.b requiring that the backup PORV control system be stroke tested at least once per 18 months.

7. Revised Surveillance Requirement 4.4.5.2 in order to prevent surveillance testing of the block valves when they are closed as a result of Action Statement b.- or c. of Specification 3.4.5.

8. Revised the associated technical specification Bases to reflect the proposed changes and to better define the basis for operability of the PORVs and block valves. Clarified that the PORVs are stroke tested during MODES 3 or 4 with the associated block valves closed.

Basis and Justification The proposed changes to the Farley Technical Specifications are consistent with the guidance pravided by Generic Letter 90-0C. The proposea changes to Specification 3.4.5 and the associated Bases increase ths probab.lity that the PORVs would be available in the event they were called upon to elieve RCS pressure. The proposed changes do not eliminate any function previously required by the PORVs, do not increase the probability of inadvertent opening (,f the PORVs, and do not create any new challenges to the RCS pressure boundary.

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3 Attachment 3 S4gnificant Hazards Evaluation

) Pursuant to 10 CFR 50.92 >

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Joseph M. f arley Nuclear Plant Units I and 2 Technical Specification Changes Associated With Generic Letter 90 06 3 Power-Operated Pelief Valve and Block Valve Reliability 10 CFR 50.92 Evaluation Erspased ChEqu The proposed changes to the farley Unit I and Unit 2 Technical Spaifications are required in order to improve the reliability of the PORVs and block valves to ensure that the PORVs are available when called upon to perform their function. These changes involve the revision of Specification 3.4.5 and the associated Bases and are consistent with the recommendations of Generic Letter s

90 06.

Backaround The Nuclear Regulatory Commissicn (NRC) issued Generic Letter 90 06, Resolution of Generic Issue 70, " Power-Operated Relief Valve and Block Valve Reliabi1Hv,"

and Generic lssue 94, ' Additional Low-Temperature Overpressurization Prs for Light-Water Reactors," June 25, 1990. The Generic letter provided go. ..te to licensees regarding the allowable outage time for inoperable PORVs and block valves in order to increase the reliability and availability of the PORVs and block valves. The proposed changes are consistent with the guidance provided by Generic Letter 90-06.

At Farley Nuclear Plant, the PORV's function is to automatically relieve RCS pressure below the pressurizer safety valve setpoint and to reduco RCS pressure upon demand by the operator. Automatic actuation of the PORVs is not assumed to mitigate the consequences of a design basis accident as described in Chapter 15 of the FSAR. The safety functions performed by the PORVs are: 1) inactive valves which form part of the RCS boundary, and 2) manual operation as required by emergency operating procedures. The PORVs are utilized to depressurize the ,

RCS in the event of a steam generator tube rupture and during natural circulation; however, automatic actuation is not relied upon by the emergency r.perating procedures. Additionally, the PORys are not utilized for low temperature overpressure protections instead, the residual heat remcyal suction relief valves perform this function.

Anal ysis The proposed change to the Technical Specifications will increase the availability of the PORVs and their associated block valves. The proposed change will allow continued operation with PORVs inoperable due to excessive seat leakage by closing the assor.iated bicek valve with power maintained to the block valve. This change will allow cperators to contir.ue to have a readily available path for relieving the RCS pressure by opening the PORY block valves.

In addition, the proposed change also revises the shutdown requirement from told Shutdown to Hot Shutdown to be consistent with the mode appilcability requirements for PORY operability.

10 CFR 50.92 Evaluation Page 2 The above changes increase the probability that the PORVs would be available if needed to mitigate the consequences of a steam generator tube rupture or fur RCS cooldown. No function previously rrauired of the PORVs has been deleted nor has the probability of the inadvertent opening of the PORVs been increased.

Additionally, these changes are consistent with the guidance provided by Generic 1.etter 90 06. Therefore, this change will not have an adverse affect on the health and safety of the public.

Southern Nuclear Operating Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed changes and has made the following determination:

1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. Automati; actuation of the PORVs at Farley Nuclear Plant is not assumed to mitigate the consequences of a design basis accident as described in Chapter 15 of the FSAR. The safety functions performed by the PORVs are: 1) inactive valves which form part of the RCS boundary, and 2) manual operation as required by emergency operating procedures. The proposed t.1anges will increase the reliability of the PORVs thus ensuring they are available to perform their function when required to do so. Therefore, it can be concluded that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not involve any physical changes to the PORVs or their setpoints. The proposed changes do not delete any function previously provided by the PORVs nor has the probability of inadvertent opening been increased. Accordingly, no new failure modes have been defined for any plan + system or component important to safety nor has any new limiting single failure been identified as a result of the )roposed changes.

Therefore, it can be concluded that the proposed c.anges do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction in a margin of safety. The proposed changes increase the reliability of the PORVs thus ensuring their availability when callad upon to perforn their function and will not impact any safety arealysis assumptions. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Conclusion Based on the preceding analysis, Southern Nuclear Operating Company has determined that the proposed changes to the technical specifications 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. Therefore, Southern Nuclear Operating Company has determined that the proposed changes meet the requirements of 10 CFR 50.92(c) and do nut involve a significant hazards consideration.

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4 Attachment 4 Environmental Evaluation

, For Proposed Technica's Specification Changes Associated With Power-Operated Relief Valves-And Block Valve Reliatiility

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Joseph M. Farley Nuclear Plant Units 1 and 2 Technical Specification Changes Associated With Generic Letter 90-06 Power-0perated Relief Valve and Block Valve Reliability Environmental Evaluation Pursuant to 10 CFR 51.22(c)(9), the proposed license amendment can be categorically excluded from the requirement to perform an environmen+si arsessment or an environmental impact statement based on the following evaluation:

Southern Nuclear Operating Company has determined that the proposed changes to the Farley Unit 1 and Unit 2 Technical Specifications, to increase the reliability of the PORVs and block valves consistent with the guidance provided by Generic letter 90-06, do not affect the types or amounts of any radiological or non-radiological effluents that may be released offsite. No increase in individual or cumulative occupational radiation exposure will result from these changes. Additionally, these changes do not involve the use of any resources not previously considered in the Final Environmental statement related to the operation of Farley Nuclear Plant.

Based upon this evaluation, it can be concluded pursuant to 10 CFR SI.22(b) that it is not necessary to perform an environmental assessment or an environmental impact statement.

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Attachment 5 Safety Analysis for Technical Specification Changes Associated With-Low-Temperature Overpressure Protection m

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Joseph H. Farley Nuclear Plant Units 1 and 2 Technical Specification Changes Associated With Generic Letter 90-06 Low-Temperature Overpressure Protection Safety Analysis Proposed Chanae Revise farley Limiting Concition for Operation 3.4.10.3 and Basis 3/a.4.10 to incorporate the following thanges:

1. Revise Limiting Condition for Operation (LCO) Action Statement "a." for Technical Specification 3.4.10.3 to reduce the allowed outage time for one RHR relief valve from the current 7 days to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless: 1) the pressurizar level is reduced to 30 percent (cold calibrated), and

2) a dedicated operator is assigned to perform RCS pressure monitor and control functions.

2. Revise Basis for Technical Specification 3/4.4.10 to clarify the means of providing low-temperature overpressure protection for the limiting heat addition transient.

Basis and Justification in response to NRC Generic letter 90-06, SNC submitted a proposed change to the technical specifications to revise the allowable. outage time (A0T) for having one inoperable low-temperature overpressure protection (LTOP) channel from the current 7-day requirement to a 24-hour requirement for water-solid conditions.

For non-water-solid conditions, the A0T would remain at 7 days.

According.to FSAR Section 5.2.2, RCS LTOP is provided, durtre, startup and shutdown when the RCS is in a water-solid condition, by !be two independent RHR suction relief valves. The Farlay LTOP system anu supporting analysis is based on the fact that there is sufficient capacity provided by one RHR relief valve-to limit the effects of: 1) the worst case i;ast. input transient (inadvertent

' start of charging pumps) and 2) the limiting heat addition transient (RCP start) provided measures are taken to cushion the overpressure effects-at.RCS

. temperatures above 2500F.

A pressurizer level of 30 percent (cold calibrated) was selected as the definition of water-solid conditions. TL level was chosen tc allow the operator sufficient time to respond 10 the overpressure event so that the limits of Appendix G are not violated. In uder to evaluate the risk while operating under an LCO for or;e inoperable LTOP channel, the postulatcJ failure of the other LTOP channel must be considered.

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Saf6Ly Analysis Page 2 An analysis of the consequences of the inadvertent start of two charging pumps was performed assuming both LTOP channels are inoperable with no other RCS vents

! available and an initial pressurizer level of 30 percent. The analysis demonstrated that the limits of Appendix G would be exceeded within

approximately 3.5 minutes.

! Prior analysis of the limiting heat addition transient of an RCP with a temperature difference between the steam generators and the RCS primary side of

less than 500F concluded that an initial pressurizer level of 30 percent

provides sufficient capacity for water expansion to prevent the limits of i Appendix G from being exceeded.

! Due to the short period of time in which an operator must respond to an 4

inadvertent charging pump start, SNC proposas to dedicate an operator to the task of monitoring and controlling RCS pressure whenever an RHR suction relief l valve is inoperab)e. 1his will provide greater assurance that the overpressure l protection system is not challenged uuring the 7-day allowed outage time.

Based upon the above analysis, the proposed technical specification changes do not pose a safety concern.

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Attachment 6 Proposed Changes to the farley Units 1 and 2 Reactor Coolant System Overpressure Protection Technical Specifications

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j Farley Unit 1 Proposed Changed Technical Specification Pages t Remove Paae Insert Paae i

3/4 4-32 3/4 4-32 B 3/4 4-14 8 3/4 4-14 B 3/4 4-14a -

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REACTOP COOLANT SYSTD$

OVERPRESSURE PROTECTION SYSTDt5 LIMITING CONDITIONS FOR OPERATION 3.4.10.3 At least one of the following overpressure protection systems shall be OPERA 8LE:

a. Two RHR relief valves with:

1. A lift setting of less than or equal to 430 psig, and

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

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

APPLICASL!"!: .When the temperature of one er more of the RC5 cold legs is less than or equal to 310'r, except when the reactor vessel head is removed.

I N SE R__A;

e. Wie = .J.-rs14ef-valve-inoperable -restore r 4ns-inoposable-valve-

.. nornine r .....= 4 thin-? dey er de rossurdze-anCven* +5e 8CS th ch apeter th== er eg3e! --te L a5 saers Jach-vent-within

h e t a h -s.-.

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

1. Restore at least one RHR reitef valve to OPERA 8LE -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 R65 vent is used to mitigate a-RCS pressure transient, a Special Report shall be prepared and submitted to the Consission pursuant to Specificar. ion 6.9.2 Mthin 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 Speci "> . tion 3.0.4 are not applicable.

FARLEY-UNIT 1 3/4 4-32 AMEN 0HENT NO.-- 26

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REACTOR COOLANT SYSTEM BASES The use of the composite curve is necessary to set conservative heatup

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limitations because it is possible for conditions to exist such that over the j 4 course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit nust at all times be based on analysis of i the most critical criterion.

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Finally, the 10 CFR Part 50. Appendix G Rule sich addresses the metal temperature of the closure head flange and vessel flange must be considered.

This Rule states that the minimum metal temperature of the closure flange regions be at least 120'F higher than the limiting Rindt for these regions when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Farley Unit 1). In addition, the new 10 CFR Part 50 Rule states that a plant specific- fracture evaluation may be performed to justify less '

limiting requirements. As a result, such a fracture analysis Ws nerformed for Farley Unit 2. These Farley Unit 2 fracture analysis results are epplicable to '

Farley Unit 1 since the pertinent parameters are identical for both- plants.-

Based upon this fracture analysis, the 16 EFPY heatup and cooldown curves are impacted by the new 10 CFR Part 50 Rule as shown on Figures 3.4-2 and 3.4-3.

Although the pressurizer operates in temperature ranges above those for which 2

there is rtason for concert of non-ductile failure, operating ' limits are provided to assure compatibility of.0peration with the fatigue analysis perfonned in acco sith the ASME Code requirements. *

, c. Ver The OPERASILITY o -two-RHR relief valve or an RCS vent opening of greater than or eaual to 2.85 square inches ensures at 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 310'F.

Either RilR relief valve has adequate relieving capability to pruect the F.CS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of tne steam generator less than or equal to 50*F above the RCS cold icg temperatures 4or (2) the start of 3 charging pumps and their injection into a water solid RCS A -- _

3/4.4.11 STRUCTURAL INTEGRITY b 1he inservice inspection- and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of i these components will be mintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME

, Boiler and Pressure Venel Code and applicah?e Addende as required by 10CFR-Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10CFR Part 50.55a(g)(6)(i).

3/4.4.12 REACTOR VESSEL HEAD VENTS The OPERABILITY of the Reactor Head Vent-System ensurts that adequate core to Ing can be maintained in the event of the accumulation of non-condensable gases in the reactor vessel. This system is in accordance with 10CFR50.44(c)(3)(fii).

FARLEY-UNIT 1 B 3/4 4-14 AMENDMENT NO. 47, - h y , - . _ - ~ . ~ , . . , , . - , - . , , , ,. ,,..r..- . , - , , _ , ,

INSERT A ACTION: ,

a. With one RHR relief valve ino)erable, restore the inoperable valve to OPERABLE status within 24 Tours 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 iii. 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , t

INSERT B provided-measures are taken to ct:shion the overpressure effects at RCS temperatures above 2500F, s

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i i Unit 1 i Typed Page.

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M ACTOR COOLANT SYSTEM OVERPRESSURE FROTECTION 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 Syste . (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 3100F, except when the reactor vessel head is removed.

ACTION:

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

1. Establish the following requirements:

1. Reduce pressurizer level to less than or equal to 30 percent (coid calibratri), 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

b. With both RHR relief valves inoperable, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 l 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 i transient, the effect of the RHR-relief valves or vent on the transient and any corrective action necessary to preverit recurrence,

d. The provisions of Specification 3<0.4 are not applicable.

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

REACTOR COOLANT SYSTEM BASES _

lhe use of the ccmposite curve is necessary to set conservative heatop limitations because it is possible for conditions to exist such that over ti.e course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the 10 CFF Part 50, Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange must be considered.

This Rule states that the minimum metal temperature of the closure flange regions be at least 120'F higher than the limiting RTndt for these rcJions when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Farley Unit 1). In addition -the new 10 CFR Part 50 Rule states that a plant specific fracture evaluation may be performed to justify less -

limiting requirements. As a result, such a fracture analysis was performed for farley Unit 2. These Farley Unit 2 fracture analysis results a o applicable to Farley Unit I since the pertinent parameters are identical for both plants.

Based upon this fracture analysis, the 16 EFPY heatup and cooldown curves are impacted by the new 10 CFR Part 50 Rule as shown on Figures 3.4-2 and 3.4-3.

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 l than or equal to 2.85 square ir.ches ensures that the RCS will be protected from pressure transients which could exceed the li. nits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are.less than or equal to 310 F.

Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is liinited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50'F 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 3 charging pumps and their injection into a water solid RCS.

3/4.4.11 STRUCTURAL INTEGRITY Yhe inservice inspectior, and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level.-throughout the life g of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specifi.: written relief. has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(1).

FARLEY-UNIT 1 B 3/4 4-14 AMENDMENT NO.

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

( -.

REACTOR COOLANT SYSTEM BASES ,,

3/4.f. 12 REACTOR VESSEL HEAD VENIS 1 The OPERABil'TY of the Reactor Head Vent System ensures that adeqtiate core cooling can be maintained in the event of the accumu'lation of non-condensable gases in the reactor vessel. This system is in accordance with 10 CFR 50.44(c)(3)(iii).

f d

l FARLEY-UNIT 1 B 3/4 4-14a AMENDMENT NO.

_ _ - - - _ - _ _ - _ _ _ _ _ - - - - = _ _ - - - _ - - - - - - - -

c ,

+

)

Farley Unit -

. Proposed-Changed Technical- Specification f"'

Eggye Pace ~ Insert Paqc 3/4 4 3/4 4-32=

B 3/4 4-14 B 3/4 4 ---

B 3/4 4-14a 4

4 t

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Unit 2 Marked Page f

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REACTOR CCOLANT SYSTEM OVERPRESSURS PROTECTION SYSTEMS LIMITING CON 0!TIONS FOR OPERATION 3.4.10.3 At least one of the following overpressure protecs.on 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.

AP Pl.ICAB LITY: When the temperature of one or more of the RCS cold legs is less taan or equal to 310*F, except when the reactor vessel head is removed.

Ac m N:

INSERT A

. 9 th ene RMR r: lief--ve4ve-4*opvab!:, re: tore-the-4*opeab-10 volve te OPERABL4-status-w4%4e7-4 y: or depee+seite tod-vent- th: RCS theough : getatar-than-4Wal t0 2.SS square-loch vent e&M4r the next-8-heesv-

b. 'ditn both RHR relief valves inoperabie, within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months either:

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

2. Deprtssurize and vent the RCS through a greater than or epual 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 cransient, the effect of the RHR relief valves or vent on the transient and any corrective action necessary to prevent recur-ence.

d. The provisions of Specification 3.0.4 are not applicable.

l l

art.Ev-UNI ~ 2 3/1 4-32 l

} J

REACTOR COOLANT SYSTEM B.A.SE..S.......*...

. 22..........................................................

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the hentup ramp the contrrlling condition svita.hes from the inside to the outside and the pressure' limit must at all times be based on analysis-of the most critical criterion.

Finally, the 10CFR Part 50, Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange must be considered.

This Rule states that the minimum metal temperature of the closure flange regions be at least 120'T higher than the limiting RT for these regions when the pressure exceeds 20 percent of the preservic,,hydrot e 3 tic test pressure (621 psig for Farley Unit 2). In addition, the nev 10CTR Part 50 t Rule states that a plant specific fracture evaluation may be performed to justify less limiting requirements. Based upon such a fracture analysis for Farley Unit 2, the 14 EFPY heatup and cooldovn curves are impacted by the l nev 10CFR Part 50 Rule as shown on Figures 3.4-2 and 3.4-3.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductible failure, operating limits are provided to assure compatiLility of operation with the fatigue analysis performed in acco 4 -

vith the ASME Code requirements.

t eNW .

The OPERABILITY o +vo-RRR relief valve or an RCS vent opening of greater than or equal to 2,85 square inches ensWes that the RCS vill be protected from pressure transients which could exceed the limits of Appendix G to 10CFR Part 50 when one or more of the PC3 cold legs are less than or equal to 310'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 of an idle RCP vith the secondary vater temperature of the steam generator less than or equal to 50'F above the RCS cold leg tamperaturestor bhEIR Eb (2) the start of 3-charging pumps and their injection into a vater solid RCS.

3/4.4.11 STRUCTURAL INTEGRITY The inservice inspection end. testing progrars for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components vill be maintained at an acceptable level throughout the life of the plant. -These programs are in.accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable' Addenda-as required by 10CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10CFR Part 50.55a(g)(6)(1).

3/4.4.12 REACTOR VESSEL HEAD VENTS I The OPERABILITY of the Reactor Head Vent System ensures that adequate core cooling can be maintained in the event of the accumulation of non-condensable gases in the reactor vessel. This system is in accordance with 10CFR50. 44(c)( 3)(iii) .-

\

FARLEY-UNIT 2 -

B 3/4 4-14 AMEN 0 MENT NO. 38, 55, 13 I

INSERT A

ACTION

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

1. Establish the following requirements:

d

i. Reduce pressurizer level to less than or equal to 30 percent ~

(cold calibrated), and ii. Assign a dedicated operator for RCS pressure monitoring and control, and iii. Restore the inopcrable v51ve 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 E hours, INSERT B provided measures are taken to cushion the overpressure effects at RCS temperatures above 2500F,

Unit 2 L

Typed Page b

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ' _ _ _ _ _ _____z_______ _ ____ _ __ _ __ _ _ ___.__ _ ____ _

{,

ig

.. REACTOR' COOLANT SYSTEM: .:

r 3-OVERPRESSURE' PROTECTION SYSTEMS ,

~ .

l- LIMITING CONDITIONS FOR OPERATIONS I 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 j b. The~ Reactor Coolant System (RCS)--depressurized with- anL RCS-vent.cfi

[ greater than~ or equal- to 2.85 square inches.

4 APPLICABILITY: .When the temperature of.one.or more of the-RCS cold legs is-

[ less than or equal to 3100F, except when the reactor vessel head is removed. .

ACTION:

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

f l '. Establish the- following. requirements:

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

e ii. Assign:a dedicated operator for.RCS_ pressure monitoring; and control, and 1

[ iii. Restore the inoperable valve to OPERABLE: status within 7 days, or;

2. Depressurize and vent the RCS throughla greater tha_n or equal

to 2.85' square inch vent within: .the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ._ _

[ b; With both RHR relief-valves-inoperable, _within:8 hours either:

, 1. -- Restore at least one RHR relief valve to;0PERABLE status, or' s i

i 12.- 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 orsa 'RCS vent is Lused toimitigate a.-

RCS pressure transient,-'a SpecialJReport shall be prepared and-submitted.to the Commission pursuant-to Specification 6.9.2;within; g 30 days. :The report shallEdescribe the circumstances initiating the -

, transient,. the' effect of the P.HR relief' valves or: vent on the .

j transient and any corrective action necessary to-prevent: recurrence.

d. The provisions of Specification _3.0.4 are' not applicable.

L p FARLEY-UNIT 2 3/4 4-32 : AMENDMENT NO.

REACTOR COOLANT SYSTEM BASES The use of the composite curve is necessary to set conservative heatop )

limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the 10 CFR Part 50 Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange must be considered.

This Rule states that the minimum metal temperature of the closure flange regions be at least 120 F higher than the limiting RT ndt for these regions when-the pressure exceeds 20 percent of ine preservice hydrostatic test pressure (621 psig for Farley Unit 2). In addition, tha new 10 CFR Part 50 Rule states that a plant sp:.cific fracture evaluation may be performed to justify less limiting requirements. Based upon such a fracture analysis for farley Unit 2, the 14 EFPY heatup and cooldown curves are impacced by the new 10 CFR Part 50 Rule as shown on Figures 3.4-2.-and 3.4-3.

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 l than or aqual to 2.85 square -inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix C to 10 CFR Part 50 when one or more of the RCS cold legs are'less than or equal to 310 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 of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50 F 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 3 charging pumps and their injection into a water solid RCS.

4 3/4.4.11 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class-1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 59.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).

('

i FARLEY-UNIT 2 B 3/4 4-14 AMENDMENT N0.

REACTOR COOLANT SYSTEM BASES 3/4.4.12 REACTOR VESSEL HE&D VENTS The OPERABILITY of the Reactor Head Vent System ensures that adequate core '

cooling can be maintained in the event of the accumulation of non-condensable gases in the reactor vessel. This system is in accordance with 10 CFR 50.44(c)(3)(iii).

f

.c FARLEY-UNIT 2 B 3/4 4-14a AMENDMENT N0.

e

. Attachment-7 Environmental Evaluation For Propo',ed Technical Specification Changes Associated With Low Temperature Overpressure Protection

_____._________.___________.__o

4 Joseph M. Farley Nuclear = PlantL Units 1 and 2 Technical Specification Changes Associated With Gereric Letter 90-06 Low-Temperature Overpressure Protection Environmental Evaluation Pursuant to 10 CFR 51.22(c)(9), the proposed license amendment can be categorically excluded from the requiremer,t to perform an environmental assessment or an nvironmental impact statement br sd on the'following evaluation:

Southern NJclear Operating Company has determined that the proposed changes to the Farley Unit I and Unit 2 Technical Specifications, to increase the reliability of the RCS low-temperature overpressure protection system (LTOP) consistent with the guidance provided by Generic Letter 90-06, do not affect the types or amounts of any radiological or non-radiological effluents that may be released offsite. No increase in individual or cuniulative cccupational radiation exposure will result from these changes. Additionally, these changes do not involve the_ use of any resources not previously considered in the Final Environmental statement related' to the operation of Farley Nuclear Plant.

Based upon this evaluatior, it can be concluded pursuant to 10 CFR 51.22(b) that it is not necessary to perform an environmental assessment or an environmental impact statement.

i r

D

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ML20115C890

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