ML20247E389

Proposed Tech Specs,Revising Power Range Neutron Flux High Trip Setpoints in Event of Inoperable MSSVs & Deleting References to 3-loop Operation in Current TS Section 3.7.1.1(b) & Table 3.7-2

ML20247E389
Person / Time
Site:	Mcguire, McGuire
Issue date:	05/08/1998
From:	DUKE POWER CO.
To:
Shared Package
ML20247E387	List: ML20247E383 ML20247E389
References
NUDOCS 9805180326
Download: ML20247E389 (58)

Text

_ _ _ - _ _ _ _ - - _ _ _ _ - _ _ _ _ - _ - . . - _ _ _ _ _ - - - . . _ - -_ - - _ - . - _ - - .

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Attachment I Marked-up Changes'to Technical Specifications A. Current TS~and Bases 1~. Unit 1

-2. Unit 2' B. Improved Standard TS and Bases

1. Unit 1

2. Unit 2-C. 5/27/97 ISTS Markups with Current Additional Changes Noted

1. CTS markups 2.:NUREG-1431 markups (Unit 1 only; unit 2 is identical) i l

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9905190324 990506 F  ;

PDR ADOCK 05000369,  :

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Attachment I.A Technical Specification Markups - Current TS and Bases 1-l

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4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES 1

l ,

LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam p generator shall be OPERABLE with lift settings as specified in Table 3.7 .

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With four reactor coolant loops and associated steam generaturs in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With three rewitor coolant loop and associated steam enerators in )

peration and one or more ma steam line code sa valves a ociated with an erating loop in rable, operation i ODES 1, 2, d 3 may proceed ovided, that wit 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either t inop ble valve is res red to OPERABLE s tus or the Power R e Neutro lux High Trip Se int is reduced pe Table 3.7-2; other-at least HOT STA Y within the next ours and in CO (wise,be SHUTDOWN wi in the followino 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />s-3 /. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. Following testing, lift settings shall be within i 1%.

i j

McGUIRE - UNIT 1 3/4 7-1 Amendment No. 166

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TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION Haximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operatino Steam Generator (Percent of RATED THERMAL POWER)

-, I '

g 88 2

3 f.E 39 M 19 TABLE 3.7M '

MAXIMUMAh0WABLEPOWERRANG NEUTRO FLUX h1GH SETPOI k WITH INDPERABLE STEAM LINE SAFETY VAtVES DURMG THRBE LOOP OPEAATION Maxim Numbe of Ino .rable Maximum lowab Power Ra e Safe Valve on Any Neutron ux Hi Setpoint Operati

\

Stea Genera . r* (P rcent of TED T RMAL POWE

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3- **

TABLE 3.7 Jf E STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (* 3%)* ORIFICE SIZE

/

Loop A Loop B Loop C Loop D

1. SV 20 SV 14 SV 8 SV 2 1170 psig

2. SV 21 SV 15 12.174 in2 SV 9 SV 3 1190 psig

3. SV 22 12.174 in2 SV 16 SV 10 SV 4 1205 psig

4. SV 23 SV 17 16.00 in2 SV 11 SV S

5. SV 24 SV 18 SV 12 1220 psig 16.00 in 2 SV 6 1225 psig 16.00 in 2

• AtJ a= s + + wn ::fety /:1 c: Ch:ll be OPE m rr an theasn-epcidt4=g-cte::

,ganerater,

1. Ih000- V0lu 3 Icii. uldpendiflg RC Oppis. I si i.lsr ee IOop Upti ci40A he lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

McGUIRE - UNii 1 3/4 7-2 Amendment No. 166 l

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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 Coolant System pressure will be limited to within 110% of its design pressure of 1185 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine tri) from 100% RATED THERMAL POWER coincident with an assumed loss of condenser .1 eat sink (i.e., no steam bypass to the condenser).

rA The specified valve lift settings and relieving capacities are in accord-ance with the requirements o i Section III of the ASME Boiler and Pressure Code, 1971 Edition. Table 3.7 lows a

• 3% setpoint tolerance for OPERABILITY; however, the valves are reset to

• 1% during surveillance testing to allow for drift. The gotal relieving capacity for all valves on all of the steam lines is15.9xg0 lbs/hr which is 105% of the total secondary steam flow of.

15.14 x 10 lbs/hr at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is a ailable for the allowable THERMAL POWER restriction in Table / 3.7-1 emf =

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 Coolant System steam flow and THERMAL POWER required by the reduced Reactor Trip Settings of the Power Range Neutron Flux channels. The d "a

^-M'~'; '..:

Reactor Trip Setpoint reductions aree.ederive on AhL$& l tour loop operati

.g

, (X) - (Y)(V) x 09)

X Wegg.ghngs Wdsw hs$

For three 1 operation: 4 (,;go Lt 3p , (X) - .Y)(U) x p) (fJ$hL) %-091 -

Where: A SP = educed Reactor Trip tpoint in percent f RATED

.RMAL POWER, V = Maxi m number of inoperab safety valves per teamk line, U = Maximum n mber of inoperable sa ty valves per

( operating, steam line McGUIRE - UNIT 1 8 3/4 7-1

PLANT SYSTEMS

) BASES SAFETYVAbS(Continued) 109 =

ower Range Neutro Flux-High Trip Set int for four loop o ration,

= Maxim percent of RATED HERMAL POWER perm sible by P-8 Setpoin or three loop o) ation. This valu left blank pending N approval of tire loop operation, X Total relievin capacity of all afety valves per eam line in

/ hour, and Y = Maxim relieving c acity of any ong safety valve in s/ hour.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM J

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-offsite power.

Each electric motor-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 450 gpm at a pressure of 1210 psig to the

-) entrance of the steam generators. The steam-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 900 gpm at a pressure of 1210 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to ren.ove decay heat and reduce the Reactor Coolant System tem System may be placed into operation. perature to less than 350*F when the RHR Verification of the steam turbine-driven pump discharge pressure should be deferred until suitable test conditions are established (i.e., greater than or equal to 900 psig in the secondary side of the steam generator). This deferral is required because until 900 psig is reached, there is insufficient steam pressure to perform the test.

3/4.7.1.3 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

This dose also includes the effects of a coincident 1.0 gpm reactor to second-ary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses.

b

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES l

LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steamj generatorshallbeOPERABLEwithliftsettingsasspecifiedinTable3.7-7.g APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in ac least HOT STANDBY J.

within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With ree reactor coo nt loops and associ ted steam gener tors Y operati and with one o are main steam li code safety va associate with an operati loop inoperable, eration in MODES 2, and 3 ma proceed provide , that within 4 hou finoperableva e is restored to either the I ERABLE status or e Power Range \

Neutron Flux Hi Trip Setpoint is =duced per Table -2; other- '

wise, be in at le HOT STANDBY witn the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> nd in COLD 110 wing 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />s- J GHUTDOWNwithinthe f /. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. Following testing, lift settings shall be within 1%.

E i

i McGUIRE - UNIT 2 3/4 7-1 Amendment No. 148

(

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i TABLE 3.7-1 q

)

l MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINTj WIT ,

l JNOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION

}

l Maximum Number of Inoperable Maximum Allowable Power Range '

Safety Valves on Any Neutron Flux High Setpoint Operatino Steam Generator (Percent of RATED THERMAL POWER) 1 2

.# IE 3 39 l9

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TABLE h2-2 h sMAXIMU LOWABL ER RA NEUT FLUX HIG ETPOINT WITH Ih0PERABLE MEAM LINE $AFETY VAtVES DURhlG THREE lbOP OPERATION i Maxi um Number f Inop able Ma

• um A owable Pow Range f Saf ty Valves Any Ne ron F x High Set Opera 'no Steam nerato s*

int '. , .

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(Percen of RA ED THERMAL WER) -

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(*

4 TABLE 3.7-K2-STEAM LINE SAFETY VALVES PER LOOP

.k.f VALVE NUMBER LIFTSETTING(*3%)*8 ORIFICE SIZE j, Looo A LooD B LooD C L0oD D *

1. SV 20 SV 14 SV 8 SV 2 1170 psig 12.174 in2 tj

2. SV 21 SV 15 SV 9 SV 3 1190 psig

3. SV 22 SV 16 12.174 in2 SV 10 SV 4 1205 psig

4. SV 23 SV 17 16.00 in2 SV 11 SV 5 1220 psig

5. SV 24 SV 18 16.00 in 2 g SV 12 SV 6 1225 psig 16.00 in2 2

f

% '.es:t tu: :sfety valvc; 3l. oil Lc OP "f3LE cr. the nr.,perating steam. '

9e

=Tnese voiue:, icil blod pendiiig NRC appivv-d of three leep cperation.

N*Theliftsettingpressureshallcorrespondtoambientconditionsofthe valve at nominal operating temperature and pressure.

McGUIRE - UNIT 2 3/4 7-2 Amendment No. N l

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3/4.7 PLANT SYSTEMS l

BASES l

l l 3/4.7.1 TURBINE CYCLE 1

3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to within 110% of its design pressure of 1185 psig during the most severe anticipated system operational transi er.t. The maximum relieving capacity is associated with a Turbine tri) from 100% RATED THERMAL POWER coincident with an assumed loss of condenser leat sink (i.e., no steam bypass to the condenser).

The specified valve liftkt gs and relieving capacities are in accord-ance with the requirements TSection III of the ASME Boiler and Pressure Code, 1971 Edition. Table 3.7- a llows a i 3% setpoint tolerance for OPERABILITY; however, the valves are reset to i 1% during surveillance testing to allow for drift. The gotal relieving capacity for all valves on all of the steam lines is 15.9 x 4

15.14x10g0 lbs/hr which is 105% of the total secondary steam flow of lbs/hr at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity l isavailablefortheallowableTHERMALPOWERrestrictioninTablej3.7-1*nd-

-c.

'3 STARTR 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 Coolant System steam flow and THERMAL POWER required by the reduced Reactor Trip Settings of the Power Range Neutron Flux channels. The j Reactor Trip Setpoint reductions are derive c- $ f:ll;.. :.., L.,a :

r four loop ope. tion based om M

.g g j P = (X) X- (Y)( is We4NwarInife #8 (109) 4 y pg kdro'$o r bN#

For three op operation: [AffAL)y % 88I

SP = (X) (Y)(U) x g) i Where.

SP = educed Reactor Trip tpoint in percen of RATED THERMAL 3

WER, V = Ma ' mum number of inopera safety valves steam line, U = Maxim number of inoperable s ty valves per o rating, steam line McGUIRE - UNIT 2 B 3/4 7-1

hY b $ $ $ $ 0 ;l;;; , i j k & . h($

PLANT SYSTEMS

!l

)

)( BASES SAFET ALVES (Continue 109 Power Range He on Flux-High Tri o eration. etpoint for four loop

= Max um percent of RA THERMAL POWER pe issible by P-8 Setpohq for three loop eration. This va e left blank pending RC approval of th e loop operation, 3

= 1bs/

Total relie ng capacity of al safety valves per hour, an eam line in Y = Ma imum relievin capacity of any onssafety valve in Ib. our.

)

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-offsite power.

Each electric motor-driven auxiliary feedwater pump is capable of j delivering a total feedwater flow of 450 gpm at a pressure of 1210 psig to the

) entrance of the steam generators.

is capable of delivering a total feedwater flow of 900 gpm at a pressure of 1210 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System tem System may be placed into operation. perature to less than 350*F when the RHR Verification of the steam turbine-driven pump discharge pressure should be deferred until suitable test conditions are established (i.e.,

equal to 900 psig in the secondary side of the steam generator) greater

. This than or deferral is required because until 900 psig is reached, there is insufficient steam pressure to perform the test. .

3/4.7.1.3 SPECIFIC ACTIVITY I

The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

.This dose also includes the effects of a coincident 1.0 gpm reactor to second-ary tube leak in the steam generator of the affected steam line. These values  !

are consistent with the assumptions used in the accident analyses. 4 l

McGUIRE - UNIT 2 B 3/4 7-2

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Attachment I.B l Technical Specification Markups - Improved Standard TS and Bases l

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MSSVs 3.7.1

)

Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Maximum Allowable Power Range Neutron Flux High Setpoints in Percent of RATED THERMAL POWER MINIMUM NUMBER OF MSSVs MAXIMUM ALLOWABLE POWER PER SiEAM GENERATOR RANGE NEUTRON FLUX REQUIRED OPERABLE HIGH SETPOINTS (% RTP) 4 s f 5$

3 s % 39 2 s 96 14

)

Table 3.7.1-2 (page 1 of 1)

Main Steam Safety Valve Lift Settings VALVE NUMBER LIFT SETTING STEAM GENERATOR (psig

• 3%)

A B C D SV-20 SV-14 SV-8 SV-2 1170 SV-21 SV-15 SV-9 SV-3 1190 SV-22 SV-16 SV-10 SV-4 1205 SV-23 SV-17 SV-11 SV-5 1220 SV-24 SV-18 SV-12 SV-6 - 1225

)

McGuire Unit 1 3.7-3 5/20/97

• MSSVs B 3.7.1

).:*~

BASES ~.(continued)

APPLICABILITY In MODES 4 and 5, there are no credible transients requiring' i

.the MSSVs. The steam generators are not normally used for

heat removal ~in_ MODES 5 and 6, and thus cannot be overpressurized;.there is no requirement for the MSSVs to be OPERABLE in these MODES.

i ACTIONS The ACTIONS table'is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

A.1 Aa,1 A.R With one or more MSSVs inoperable, reduce power so that the available MSSV relieving capacity meets Reference 2 requirements for the applicable THERMAL POWER.

Operation with less than all five MSSVs OPERABLE for each -

steam generator is pennissible, if. THERMAL POWER is proportionally limited to the relief capacity of the

. remaining MSSVs. This is accomplished by restricting THERMAL POWER so that the energy transfer to the most limiting steam generator is not greater than the available relief capacity in that steam generator. For example, if one MSSV is inoperable in one-steam generator -the relief capacity of that steam generator is. reduced by approximately 20%. To offset- this reduction in relief capacity, energy transfer to that steam generator must be similarly reduced.

This is accomplished by reducing THERMAL POWER by the

'necessary amount to conservatively limit the energy transfer to all steam generators, consistent with.the relief capacity of the most limiting steam generator.

,, perati able wi the safety valveshinMODES1,2and3isal erable within he limitati s of the tion i

quirc;ien of A.1 on the asis of the reduction n the in' Steam stem steam flo and THERMAL OWER req red by t reduced ctor trip sett gs of the ower Rang Neutron F1 channels. The reactor tr setpoint ductions re deri ed on.the 11owing bases:

SP = x 09),where:

L SP = Redu reactor t setpoint'ih rcent RTP, l:

(continued) iMcGuire Unit.1- B 3.7-3 Supplement 1 l L

c ,

t SW $

L l The maximum power level specified for the power range

neutron flux high trip setpoint with inoperable MSSVs must

ensure that. power is limited to less than the heat removal capacity of the remaining OPERABLE MSSVs. The reduced.high flux trip setpoint also ensures that the reactor trip occurs early enough in the loss of load / turbine trip event to limit primary to secondary heat transfer and preclude overpressurization of the primary and secondary systems. To calculate this power level, the governing equation is the relationship q = m Ah, where q is the heat input from the primary side, m'is the steam flow rate and Ah is the heat of vaporization at the steam relief pressure (assuming no subcooled feedwater). The algorithm use is consistent with j the recommendations of the Westinghouse Nuclear Safety 1 Advisory Letter, NSAL-94-001, dated January 20, 1994 (Ref. 7). Additionally, the calculated values are reduced by 98s to account for instrument and channel uncertainties.

I 1

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• L* MSSVs B 3.7.1

~ BASES ACTIONS (continue V= Neximum nu.. r of inopera le safety val s per s m line.

109 = ower Range Ne tron Flux Hi setpoint,

'X= To 1 relieving pacity of al safety valves per ste line (in 1bs hour), and Y= Maxi relieving ca city of any sne safety val (in 1bs/ho ).

W1 one or more M Vs inoperable, reduce the wer range neu on flux high tr setpoints to he percent f RATED THER L POWER identif d in Table 3. 1-1.

The turb ne trip event i licitly takes redit for e power ran neutron flux h h trip setpoi to termi te the event- us ensuring tha primary and s ondary overpressure ation are preclu ed when operat ng at reduced power levels with inoperable M Vs.

3.1 and B.2 If the MSSVs cannot be restored to OPERABLE status within the associated Completion Time, or if one or more steam generators have less than two MSSVs OPERABLE, the unit must be placed in a MODE in which the LC0 does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full l

power conditions in an orderly manner and without challenging unit systems.

L SURVEILLANCE SR 3.7.1.1 '

L REQUIREMENTS

! This SR verifies the OPERABILITY of the MSSVs by the

( verification of each MSSV lift setpoint in accordance with (continued)

McGuire Unit 1 B 3.7-4 5/20/97 L

MSSVs B 3.7.1

' BASES .(continued)-'

RE RE ES 5. ASME, Boiler and Pressure Vessel Code,Section XI.

6. ANSI /ASMEOM-1-1987.

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7 b)CSTip)(, HOUSE NJMA/8 SMETY G Awiscay wrme, /USAL-W%,

Mfdb $t)dr\ fly 2q /W</.

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McGuire. Unit' 1 B 3.7-6 5/20/97

, . [* -

MSSVs i

3.7.1 Table 3.7.1-1 (page 1.of 1)

OPERABLE Main Steam Safety Valves versus-Maximum Allowable Power Range Neutron Flux High l Setpoints'in Percent of RATED THERMAL POWER- i MINIMUM NUMBER OF MSSVs I MAXIMUM ALLOWABLE POWER

-PER STEAM GENERATOR  :

RANGE. NEUTRON FLUX REQUIRED OPERABLE  ;

HIGH SETPOINTS (% RTP) '

4 s 51 5$

3 sfs 39

'2  ;

s PS l8

)

Table 3.7.1-2 (page 1 of 1)

Main Steam Safety Valve Lift Settings VALVE NUMBER LIFT SETTING STEAM GENERATOR A B (psig i 3%)

C D SV-20 SV-14 SV-8 -SV-2 1170 SV-21 SV-15 Sy-g SV-3 1190 SV-22. SV-16 SV-10 SV-4 1205 SV-23 SV-17 SV-11 SV-5 1220 SV-24 SV-18 SV-12 SV-6 '

1225 L

I' McGuire Unit 2 3.7-3 5/20/97

____ ____-___-________--=-

't

.- MSSVs l - f *1 3 8 3.7.1

l. BASES ~(continued)

APPLICABILITY In MODES 4 and .5, there are no credible transients requiring the MSSVs. The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be u overpressurized; there is no requirement.for the MSSVs to be OPERABLE in these MODES.

ACTIONS- The ACTIONS table is modified by a Note indicating that

. separate Condition entry is allowed for each MSSV.

A.1 M N. 1 d With one or more MSSVs inoperable, reduce power so that the available MSSV relieving capacity. meets Reference 2 requirements for the applicable THERMAL POWER.

Operation with less than all five MSSVs OPERABLE. for each steam generator is permissible, if THERMAL POWER is

. proportionally limited to the relief capacity of the remaining MSSVs. This is accomplished by restricting

. THERMAL POWER so that the energy transfer to the most limiting steam generator is not greater than the available relief capacity in that steam generator. For example, .if one'MSSV is inoperable in one steam generator, the relief capacity of that steam generator is reduced by approximately 20%. To offset this reduction in relief capacity. energy transfer to that steam generator must be similarly reduced.

This is accomplished by reducing THERMAL POWER by the necessary amount to conservatively limit the energy transfer- 1 to all. steam generators, consistent with the relief capacity l of the most limiting steam generator.

1. peration i MODES 1, 2, a 3 is allowabl with the safety lves inoper le within the imitations of. he' Action uirements o A.1 on the bas s of the reduc ion in the Ma Steam Syste steam flow an THERMAL POWER equired by the duced react trip' settings f the Power nge Neutron Flux hannels. The reactor trip s oint-reducti ns are derive on the follo ng bases:

SP<= (X1 (Y)(V) x (109), where: l

(

SP = Reduced reactor trip se point in percen RTP, p

(continued) ,

j McGuire Unit 2 B 3.7-3 Supplement 1 l

%M 1 A~

The maximum power level specified for the power range neutron flux high trip setpoint with inoperable MSSVs must ensure that power is limited to less than the heat removal.

capacity of the remaining OPERABLE MSSVs. 'The reduced high flux trip setpoint also ensures that the reactor trip occurs early enough in the loss of. load / turbine trip event to limit primary to secondary heat transfer and preclude overpressurization of the primary and secondary systems. To calculate this power level, the governing equation is the relationship q = m Ah, where q is the heat input from the primary side, m is the steam flow rate and Ah is the heat of vaporization at the steam relief pressure (assuming no subcooled feedwater). The~ algorithm use is consistent with the recommendations of the Westinghouse Nuclear Safety Advisory Letter, NSAL-94-001, dated January 20, 1994 (Ref. 7). Additionally, the calculated values are reduced by 9% to account for instrument and channel uncertainties.

i .* <

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. - , MSSVs B 3.7.1 l-f ' BASES U ACTIONS (continued)

V Maximum nu r of inoperabi safety valves er steam line, L -109 = wer Range' Ne ron Flux High tpoint,

X= To 1 relieving c acity of all fety valves p ste line (in 1bs our), and

-Y = Maxi relieving. cap ity of any on safety valve in 1bs/hou .

I W h one or more SVs inoperable, reduce the po range ne ron flux high t ip setpoints to he percent of TED ,

THE POWER identi ied in Table 3. 1-1. j i- ,

l , The tu ine trip event mplicitly takes redit for the l power r ge neutron flu high trip setpo t to terminat the even thus ensuring at primary and s ondary -

overpress ization are p luded when opera ng at reduced

power leve with inoperabi MSSVs. j B.1 and B.2

,, If the MSSVs cannot be restored to OPERABLE status within

the _ associated Completion Time, or if one or more steam generators have less than two MSSVs OPERABLE, the unit must be placed in a MODE in which the LCO does not apply. To ,

achieve this status, the unit must be placed in at least 1

i. MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience,-to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS- '

This SR verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with (continued)

McGuire Unit 2 - B 3.7-4 5/20/97

~.t* -

'MSSVs l' B 3.7.1 BASES (continued)'

REFERENCES 5.

(continued). ASME, Boiler and Pressure Vessel Code,Section XI.

6.; ANSI /ASMEOM-1-1987.

P 9., b)f6~Tp>fWtE6 fllClAl % MBTY AbV).Sollf LW6Sy /199 L'94 00lj M186 'JArjuS/2f' 24, N9Y,

)

-i e

i

~

McGuire Unit 2 B 3.7-6  !

5/20/97 .,

________m___ _ - _ - _ _ m_ -_

Attachment I.c Improved Technical Specification Markups -

with Current Additional Changes Noted l

i i L

( l

$WmwY L

l 3.'7. l TABLE 3.7'1 l (CI[$58tf A4h S1D LGbe(esv'ces J ,

MAXIMUM ALLOWAB.E POWER RANGE NEUTRON FLUX HIGH SETPOINfEMDT m dHOPERA8CE STEM' LINE SAFETY VAtRS DURIWFOUR LODV0PERAIKO n 3,,

t sum Number of perable Maximus Allowable Power Range SafetyValveson@Any (R)_ Neutron Flux High Set)oint b! Ooeratino Steam Generator TPemeenu of R/TEtb TME130 PODPM

< 6 .

M

'. 3 b

f TABLE 3.7-2 )

/ MAXINUM ALLOWABLE POWER RANGE NEUTRON FLUK HIGH SETPOINT W INOPERABLE STEAM LINE SAFETY _y_ALVES DURIlfG THREE LOOP OPERATION /

Maximus Number of I perable Maxi Allowable Power Ran Safety Valves on Any utron Flux High Setpoi ' ,-

Operatina Steam enerator* (P cent of RATED THERMAL ER) ,

I. 1 **)

i 2 **) )

l k 3

• • ) /

I TABLE 3.

Q STEAM LINE SAFETY VALVES WOFrSedwyQ VALVE NUMBER LIFT SETTING (* 3% N #ii!FICESTfE)

Loop A Loop ..B Loop C kgop_R o

1. SV 20 SV 14 SV 8 SV 2 1170 psig .174 i l 2. SV 21 SV 15 SV 9 SV 3 1190 psig 12.174 n 2 l 3. SV 22 SV 16 SV 10 SV 4 1205 psig 16.0 in-

4. SV 23 SV 17 SV 11 SV 5 1220 psig 16. in2

5. SV 24 SV 13 SV 12 SV 6 1225 psig g 1 00 in 2

(*Atleastt safety valves shal be OPERABLE on t non-operating steam generator.

'These ues left blank pe ing NRC approval three loop operation.

• "The lif t settin ressure shall cor ond to ambient con tions of th valve'at nomina operating temperat and pressure. ,

McGUIRE - UNIT 1 3/4 7-2 Amendment !?o, 166

[,. g y _

mM3

h Tantr 1.7.1 Tp u ' . ,

3 "' . I CNN"*4 22Lf.!? W va a an.

glput ALLOb AB.E PWER'RANGF $ EUTR0y%rLJX he  % .D m r H IGH 5t l PU INT FR T . , p e r < < ^*

l amrtMABLE 5LEMP LIP E SAFETY Yh:VE5 [W1P i FOUR LDOP OPERaff 0N) i 641 hl Aw.

mun Number of hoperable ximum Allowable Power Range

] Safety Valves on Any

~. eutron Flux High Setpoint Operatinc Steam Generator of RKTM TR1511B PdDFB) a sJ c (

TABLE 3.7-2 EX lMLM ALL LE POWER ItANGE NEU"R M UX ifGH SETPOINT W 'TH IN0Pl3 TABLE SMAN LINE SAFRY VALVE!i 11 RING TiREE LOOP OPERA"I Maximus Number o Inoperable inum Allowable power R ge Safety Valve on Any Neutron Flux High Set nt Doeratine 5 am Generator

• Percent of RATED TH POWER) hA

. 1 **1 2 **

r TABLE 3.74@

Q STEAM LINE SAFETY VALVES M N N

• N VALVE NUMBER LIFT SETTING (* 3HEEP RIFICE L22LA Loop B Loop C L22L.D

1. SV M SV 14 SV 8 SV 2 1170 psig 12.174 in2

2. SV i.1 SV 15 SV 9 SV 3 1190 psig 12.174 1 2

3. SV 22 SV 16 SV 10 SV 4 1205 psig 16.00 n

4. SV 23 SV 17 SV 11 SV 5 1220 psig 16 in2

5. SV 24 SV 18 SV 12 SV 6 1225 psig .00 in2

[*Atleasttwosa ty valves shall be 0 LE on the non-opera 'ng steam generator.

% se v es left blank pending C approval of three 1 p operation.

(***Theliftse ng pressure shall co spondtoambientcondigionsofthe (n2

( valve at inal operating t ture and pressure. /

McGUIRE - UNIT 2 3/4 7-2 Amendment No. 148 y I;a,ce dwdJf& 'N" Psy 3/3

f

. j

( INSERTg Table 3.7.1-1 (page 1 of 1) J OPERABLE Main Steam Safety Valves versus  !

Maximum Allowable Power Range Neutron Flux High Setpoints in '

Percent of RATED THERMAL POWER MINIMUM NUMBER OF MSSVs MAXIMUM ALLOWABLE POWER PER STEAM GENERATOR RANGE NEUTRON FLUX HIGH REQUIRED OPERACLE SETPOINTS (% RTP) i 4

sf a

'O #

r l

$l/CMse Ant AIM $- Nb*

g INSERT Page 3.7-3 l

McGuire

r

..g HSSVs

, 8'3.7.1 BASES LCO l

(continued) This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences

! of accidents that could result in a challenge to the RCPB. .

APPLICABILITY v In H00E 1(abe'eJer1ffD, the ntaber of HSSVs per steam f i

i penerator Table H00ES 7.

2 required to be OPERABLE must in the accompanying (40. Selow dar eTW be

3. only two HSSVs per steam generator are j h

according to/n requi t be OPERABLE.

In H00ES 4 and 5. there are no credible transients tring the MSSVs. lhe steam generators are not normally u for heat removal in H00ES 5 and 6 and thus cannot be overpmssurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each NSSV.

r

. 4 g JAE i .

i

• With one or more MSSVs inoperable, reduce power so that the \

available MSSV relieving capacity meets Referenm 2 i requirements for the applicable THERNAL POWER. -

1 Operation with less than all five MSSVs OPERABLE for each steam generator is permissible. if TERNAL POWER is proportionally limited to the mitef capacity of the remaint HSSVs. This is accomplished by restricting THERHAL so that the energy transfer to the most  :

• . - limiting steam generator ts not greater than the available j

. relief capacity in that steam generator. For example. if .

i

. one MSSV is inoperable in one steam generator, the relief  !

capacity of that steam generator is reduced by approximately - *

- 20r.

transferTo offset to :iatthis reduction in relief capacity, energy

.the - steam Oy atA nas ;M13. Tr s is generator must be similarly reduced.

accomplished by reducing ilERNAL

@wwwJ '0WERfby a VMs;t2r1 wtuc0 conservatively limitsthe t&ses*(4*j)energy G 1

transfer no a' I steam generators an_

er ==n imate W tI

==mn consistent with the relief' capacity of h) the most limiting steam generator. i (continued)

AcGlu.-

EXLSai 8 3.7 3 Rev 1 ,04/07/95 h henget hswbh Nejua8h DAN

i ,

MSSVs 8 3.7.1 ,

l ACTIONS ,A_d (continued)

For each steam generator, at a s ified fractional relief capacity (FRC of each bV is determined ssure, the )

as follows:

^

FRC

' where:

A = the reli capacity of.the MSSV: and l B = the al relief capacity of all the Vs of the I ste generator.

The FRC is relief cit necessary to ddress

! operation th reduced POWER.

The THERMAL POWER levels in the prevent operat at power levels greater tha relief capacity of t remaining MSSVs. The reduced POWER is det ined as follows:

l =

1 - f, x FRC, + N, x FRC, + + N, x FRC,)) x 100t where:

RP = Reduced for the most limiting s am generator exp ssed as a percent of RTP:

N,, repre t the status of the MSSV 1.

N , .respect

.....$. . . . N, ively .

= 0 if MSSV is OPERABLE,

= 1 if t MSSV is inoperable:

of the MSSV 1, d

\' '

FRC,,

2. . . . .FRC',,

fi .. . FRC, ive ly, as defined a= the relief capact GJ \ - , ,

l MSCM{ h (continued) l MG M B 3.7-4 Rev 1. 04/07/95

,gg f%

l t  !

l l

i a

g.

INSERT fOpe ion in MODES 1, 2, and 3 is allowable with the safety valves inoperable within limitations of the Action requirements of A.1 on the basis of the reduction the Main Steam System steam flow and THERMAL POWER required by the reduced re or trip settings of the Power Range Neutron Flux channels.

The reactor trip point reductions are derived on the following bases:

SP = X) - (Y)(V) x (109), ere:

SP = Reduced reactor trip setpc t in percent RTP, V= Maximum number of inoperable sa ty valves per steam line, 109 = Power Range Neutron Flux High setpo t, X= Total relieving capacity of all safety v ves per steam line (in 1bs/ hour), and Y= Maximum relieving capacity of any one safety va e (in 1bs/ hour).

L1 I With one or more MSSVs f noperable, reduce the power range neu on flux high

( trip setpoints to the percent of RATED THERMAL POWER identifie in Table 3.7.1-1.

The turbine trip event implicitly takes credit for the power range n tron flux high trip setpoint to terminate the event thus ensuring that pri ry and secondary overpressurization are precluded when operating at reduced po r levels with inoperable MSSYs.

) i L

---y See- neu) Insert MW @'

l l

Y b/C6ttSe kMC Midb U

( N'$"'"&

INSERT Page B 3.7 - 4 McGuire l

l

INSERT The maximum power level specified for the power range neutron flux high trip I setpoint with inoperable MSSVs must ensure that power is limited to less than the heat removal capacity of the remaining OPERABLE MSSVs. The reduced high flux trip setpoint also ensures that the reactor trip occurs early enough in the loss of load / turbine trip event to limit primary to secondary heat transfer and preclude overpressurization of the primary and secondary systems. y To calculate this power level, the governing equation is the relationship q = n m Ah, where q is the heat input from the primary side, m is the steam flow rate and Ah is the heat of vaporization at the steam relief pressure (assuming no subcooled feedwater).. The algorithm use is consistent with the reconinendations of the Westinghouse Nuclear Safety Advisory Letter, NSAL 001, dated January 20, 1994 (Ref. 7). Additionally, the calculated values are reduced by 9% to account for instrument and channel uncertainties.

I A license Antefeat' $8h8 (Re7aest) INSERT Page B 3.7 - 4 l

McGuire L____.________._

.s ,

i .

MSSVs B 3.7.1 ,

! BASES SURVEILLANCE SR 3.7.1.1 (continued)

REQUIREMENTS conditions using an assist device to simulate lift pressure.

If the MSSVs are not tested at hot conditions. the lift settiny pressure sha?1 be corrected to ambient conditions of i

tihe valve at operating temperature and pressure.

, REFERENCES

! 1. h . Section 10.3.1[

( 2. . ASE. Boiler and Pressure Vessel Code.Section III.

l Article NC 7000. Class 2 Components. l

3. h . Sectionk 15.2 [

l 7 ASE. Boiler and Pressure Vessel Code.Section XI.

@ ANSUASE OH 11987.

Q so cFA $o.%, Tsokmch< Wpwrw5[c)(OCi).

\( <

f 7. We&fana sateesafeqMaos/

Le#er, usAL-%oof bded ,

January .14 M94, (

L --

t Hiceme Ameobed r/she

$!i""0 ,

W0lPS13 B 3.7-6 Rev 1. 04/07/95 k

.s Attachment II New Originals A. Current TS and Bases

1. Unit 1 I

2. Unit 2 i

B. Improved Standard TS and Bases

1. Unit 1 1

2. Unit 2 l

i i

l l

l l

i i

l t__._______

l Attachment II.A i

New Original Pages - Current TS and Bases I

l l

i I

. m 3/4 J ' PLANT SYSTEMS

~

g 3/4.7.1JTURBINE CYCLE SAFETY VALVES l LIMITING CONDITION FOR OPERATION ~

i .

l' . 3.7.1.1 ~ All main steam line . Code safety valves associated with each steam

( generator shall tHe OPERABLE with lift settings as'specified in Table. 3.7-2.

APPLICABILITY:_ MODES 1, 2, and 3.

ACTION:

! a. With four: reactor. coolant loops'and associated steam generators in operation and with one or more main steam line code safety. valves inoperable,' operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is-restored lto OPERABLE status or the Power Range Neutron Flux High Trip Setpoint L is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY i

within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 )

hours.

b._ The provisions of Specification 3.0.4 are. not applicable.

SURVEILLANCE REQUIREMENTS ,, 4.7.1.1. No additional requirements other than those required by Specification 4.0.5. Following testing, lift settings shall be within i 1%.

1 l

l 4

I I

L l

McGUIRE - UNIT 1- 3/4'7-1 Amendment No.

hg

l .

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Saf(ty Valves on Any Neutron Flux High Setpoint Operatina Steam Generator (Percent of RATED THERMAL POWER) 1 58 2 39 3 19 JABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (* 3%)* DRIFICE SIZE Looo A loon B Looo C Looo 0

1. SV 20 SV 14 SV 8 SV 2 1170 psig 12.174 inz

2. SV 21 SV 15 SV 9 SV 3 1190 psig 12.174 in2 l 3. SV 22 SV 16 SV 10 SV 4 1205 psig 16.00 inz

4. SV 23 SV 17 SV 11 SV 5 1220 psig 16.00 in2

5. SV 24 SV 18 SV 12 SV 6. 1225 psig 16.00 in2

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

McGUIRE - UNIT 1 3/47-2 Amendment No.

l

3/4.7 PLANT SYSTEMS BASES 3 /4. 7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES l

I The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to within 110% of. its design pressure of 1185 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 accord-ance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. Table 3.7-2 allows a i 3% setpoint tolerance for OPERABILITY; j however, the valves are reset to i 1% during surveillance testing to allow for drift. The gotal relieving capacity for all valves on all of the steam lines is15.9xl0 lbs/hr which is 105% of the total secondary steam flow of 15.14 x 10 lbs/hr at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1. l 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 Coolant 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 derived based on the algorithm contained in Westinghouse's Nuclear Safety Advisory Letter (NSAL)94-001.

McGUIRE - UNIT 1 8 3/4 7-1

l .

Il PLANT SYSTEMS.

' BASES i

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM L The'0PERABILITY_of the Auxiliary Feedwater System ensures that the Reactor.

Coolant: System can be cooled down to less than 350*F from normal operating j conditions in the event'of a total-loss-of-offsite power. )

' Each electric motor-driven auxiliary feedwater pump'is capable. of delivering a, total l feedwater flow.of 450 gpm at a. pressure..of 1210 psig to the entrance of the steam generators. The steam-driven auxiliary feedwater pump.is-capable of delivering a total; feedwater flow of 900 gpm at a pressure of ,

1210 psig to the entrance.of the steam generators.- This capacity is' sufficient.

to' ensure that adequate feedwater flow is 'available to remove. decay heat' and.

n reduce.the Reactor Coolant System temperature'to less than 350*F when the RHR

. System may be placed into operation.

Verification of the" steam turbine-driven pump discharge pressure should be

deferred until

suitable test conditions are established (i.e., greater than or equal to 900 psig in the secondary side of- the steam generator). This deferral

..is required because until 900 psig.is reached, there is insufficient steam opressure to perform the test..

3/4.7.1.3 SPECIFIC ACTIVITY'

.The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of  ;

10 CFR Part 100 dose' guideline values 'in the event of a. steam line rupture. i

_This dose also includes the' effects of a coincident 1.0 gpm reactor to second-

_ary tube leak in the steam generator of-the affected steam.line. These values are consistent with the assumptions used in the accident analyses.

-I

, -i

]

l 4

t i I

q r

i McGUIRE - UNIT.1 B3/47-2 i

'~

3/4.7 PLANT SYSTEMS j 3/4.7.I TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION L.

' )

~3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.

APPLICABILITY:. MODES 1, 2, and 3.

ACTION:

l

a. With four reactor coolant loops and associated steam generators in l operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and 3 may. proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint

'is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30

hours.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification .

4.0.5. Following testing, lift settings shall be within i 1%.

L L

McGUIRE . UNIT 2 3/4 7-1 Amendment No.

l .

o TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint l Ooeratino Steam Generator (Percent of RATED THERMAL POWER) 1 58 2 39 3 19 TABLE 3.7-2 l STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (* 3%)* ORIFICE SIZE Looo A Loon B Looo C Looo D

! 1. SV-20 SV 14 SV 8 .SV 2 1170 psig 12.174 inz

2. SV 21 SV 15 SV 9 SV 3 1190 psig 12.174 inz

3. SV 22 SV 16 SV 10 SV 4 1205 psig 16.00 in2

4. SV 23 SV 17 SV 11 SV 5 1220 psig 16.00 in2

5. SV 24 SV 18 SV 12 SV 6- 1225 psig 16.00 inz

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

i l

l l

I 4

McGUIRE - UNIT 2- 3/47-2 Amendrent No.

( .

3/4.7 PLANT SYSTEMS l

. BASES' 3/4.7.1 -TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES

The.0PERABILITY of the main steam line Code safety valves ensures that the l Secondary Coolant System pressure will be limited to within 110% of its design

-pressure of 1185 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 accord-

' ance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971' Edition. Table 3.7-2 allows a i 3% setpoint tolerance for OPERABILITY; however. -the nives are reset to i 1% during surveillance testing to allow for l i

drift. The gotal relieving capacity for all valves on all of the steam lines i

.is 15.9-x go Ib:;/hr which is 105% of the total secondary ' steam flow of i- 15.14 x 10 lbs/hr at 100% RATED THERMAL- POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity l is available. for the allowable THERMAL POWER restriction in Table 3.7-1. l 1

! 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 Coolant System steam flow and THERMAL POWER required by the 1 l reduced Reactor Trip Settings of the Power Range Neutron Flux channels. The L  ; Reactor Trip Setpoint reductions are derived based on the algorithm contained L in Westinghouse's Nuclear Safety Advisory Letter (NSAL)94-001.

l-l L

l l- I J

L l

McGUIRE - UNIT 2 B3/47-1 l l

PLANT SYSTEMS v,

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 a1 total. loss-of-offsite power.

'Each electric' motor-driven auxiliary feedwater pump is capable of-- 1

_ delivering a total feedwater flow of 450:gpm'at a pressure.of 1210 psig to the u

' entrance of.the steam generators. The steam-driven auxiliary feedwater pump-

is capable of delivering a total feedwater flow of 900 gpm at a pressure of 1210 psig.to the entrance of the' steam generators. This capacity is sufficient '

to ensure that adequate feedwater flow is available to remove decay heat and - I reduce the Reactor Coolant System temperature to less than 350*F:when the RHR-

' System may be~placed into operation..

Verification of the steam turbine-driven' pump discharge pressure should be deferred until _ suitable test' conditions are established (i.e.,

equal to 900 psig in the secondary side of the steam generator) greater.

than or This deferral is required because until 900 psig -is reached, there is insufficient steam pressure to perform the_ test.-

3/4.7.1.3 SPECIFIC ACTIVITY The limitations on' Secondary Coolant. System specific activity ensure that

.the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

This dose also includes the effects of a coincident 1.0 gpm reactor to second-ary tube leak in'the steam generator.of the affected steam line. These values L are consistent with the assumptions used in the accident analyses.-

u

'l l

i I

L McGUIRE;- UNIT 2 B3/47-2 l

Attachment II.B New Original Pages - Improved Standard TS and Bases i

t l .

I-

, . MSSVs 3.7.1 w

Tab 1e 3.7.1-l'(page 1 of 1)

OPERABLE, Main Steam Safety Valves versus-l Maximum Allowable Power Range Neutron Flux High i

Setpoints in Percent of RATED THERMAL POWER MINIMUM NUMBER OF MSSVs MAXIMUM ALLOWABLE POWER l PER STEAM GENERATOR RANGE NEUTRON FLUX  ;

REQUIRED OPERABLE -HIGH SETPOINTS (% RTP)

\

4 s 58 .l l 3 s 39 l-2 s 19 l I

.1 1

l l-Table 3.7.1-2 (page 1 of 1)

Main Steam Safety Valve Lift Settings-t VALVE NUMBER l LIFT. SETTING {

STEAM GENERATOR (Psig

• 3%) j A B C D SV SV-14 SV-8 SV-2 1170 SV-21 SV-15 SV-9 SV-3 1190 SV-22 SV-16 SV-10 SV-4 1205 SV-23 SV-17 SV-11 SV-5, 1220 SV-24 ~SV-18 SV-12 SV-6' 1225 I

p b McGuire Unit 1 3.7-3 3/30/98 l

.c

_1__u_____________ _ _ _ _ .

• ~

MSSVs B 3.7.1 p'

(BASES (continued) r l' ' APPLICABILITY. In MODES 4 and 5,l there are no credible transients requiring the MSSVs. The. steam generators are not normally used for heat removal.in MODES 5 and 6, and thus cannot be.

, . overpressurized; there is no requirement for the MSSVs to be .

t OPERABLE in these MODES..

L ACTIONS' The ACTIONS table'is modified by a Note indicating that separate Condition ' entry is allowed for each MSSV.

A.1 and A.2 l I

With one or. more MSSVs l'noperable,. reduce power so that the available MSSV relieving capacity meets Reference 2 requirements for the applicable THERMAL POWER.

L Operation with .less than all- five MSSVs OPERABLE'- for each steam generator is permissible, if THERMAL POWER is proportionally limited to the relief capacity of-the remaining MSSVs. This is accomplished by restricting THERMAL POWER so that:the energy transfer to the most limiting steam generator is~ not greater than the available -

' relief-capacity in that steam generator. For example, if one MSSV is inoperable in one steam generator, the relief capacity of that steam generator is reduced by approximately 20%. To offset this reduction in relief capacity, energy transfer to that steam generator must be similarly reduced.

This is accomplished by reducing THERMAL POWER by the -

necessary amount to conservatively limit the energy transfer to all steam generators, consistent with the relief capacity of the most limiting steam _ generator.

The maximum power level specified for the power range

- neutron flux high trip setpoint with inoperable MSSVs must ensure that power is limited to less than the heat removal capacity of the remaining OPERABLE MSSVs. The reduced high flux trip setpoint also ensures that~ the reactor trip occurs early enough. in the loss of load / turbine trip event to limit primary to secondary heat transfer and preclude overpressurization of the primary and secondary systems. To calculate this power level, the governing equation is the-relationship q = m Ah, where q is the heat input from the primary side, m is the steam flow rate and Ah is the heat of' u vaporization at the steam reliaf pressure (assuming no ,

l b (continued) l McGuire Unit 1 B 3.7-?, 3/30/98 l

,s

MSSVs L",

B 3.7.1 BASES I i

ACTIONS A.1 and A.2 (continued) subcooled feedwater). The algorithm use is consistent with

' the recommendations of the Westinghouse Nuclear Safety li Advisory Letter, NSAL-94-001, dated January 20, 1994 (Ref. 7). Additionally, the calculated values are reduced by 9% to account for instrument and channel uncertainties. 1 I

B.1 and B.2 If the MSSVs cannot be restored to OPERABLE status within the associated Completion Time, or if one or more steam generators have less than two MSSVs OPERABLE, the unit nust be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience,- to reach the required unit conditions from full power conditions in an orderly nenner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITY of. the MSSVs by the verification of each MSSV lift setpoint in accordance with 1

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(continued) l McGuire Unit 1 B 3.7-4 3/30/98

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L' / . , MSSVs B 3.7.1 s ' BASES (continued).

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!- REFERENCES 5. ASME, Boiler and Pressure Vessel- Code Section XI.

(continued)

-6.- ANSI /ASMEOM-1-1987.

7. Westinghouse Nuclear Safety Advisory' Letter, NSAL 001, Dated January 20, 1994.-

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MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Maximum Allowable Power Range Neutron Flux High Setpoints in Percent of RATEJ THERMAL POWER l

l MINIMUM NUMBER OF MSSVs MAXIMUM ALLOWABLE POWER PER STEAM GENERATOR RANGE NEUTRON FLUX REQUIRED OPERABLE HIGH SETPOINTS (% RTP) 4 s 58 l j 3 s 39 l l 2 s 19 l l j l

Table 3.7.1-2 (page 1 of 1)  !

Main Steam Safety Valve Lift Settings VALVE NUMBER LIFT SETTING STEAM GENERATOR (psig i 3%)

A B C D SV-20 SV-14 SV-8 SV-2 1170 SV-21 SV-15 SV-9 SV-3 1190 SV-22 SV-16 SV-10 SV-4 1205 SV-23 E'v-17 SV-11 SV-5 1220 SV-24 SV-18 SV-12 SV-6 1225 l

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,. 1 ., MSSVs

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.B'3.7.1 BASES (continued) l l - .

APPLICABILITY In MODES _4 and.5, fhere are no credible transients requiring the MSSVs. .The steam generators _ are not normally used for:

heat removal in MODES 5 and 6,,and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE:in these MODES.

1 ACTIONS = '

The ACTIONS' table ;is modified by a Note indicating that

! . separate Condition' entry is allowed for each MSSV..

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A.1 and A.2 l With one or more MSSVs inoperable, reduce power so.that the-available MSSV relieving capacity meets Reference 2 requirements for the applicable THERMAL POWER.

Operation with less than' all five MSSVs OPERABLE for each steam generator ~1s permissible, if . THERMAL POWER is proportionally.. limited to the relief capacity of the remaining MSSVs. - This is accomplished by restricting THERMAL POWER so that the energy transfer to the most.

l limiting steam generator is not greater than the available relief capacity in that steam generator. fFor example, if  ;

I, one MSSV is inoperable in one steam generator, the relief  ;

capacity. of that steam generator is reduced by approximately 20%.. To offset this reduction in relief capacity, energy transfer to that steam generator must be similarly reduced.

p

, This is accomplished by reducing THERMAL POWER by the

! necessary amount to conservatively limit the energy transfer L to all steam generators,. consistent with the relief capacity

'of the most limiting steam generator.

The maximum power level specified for the' power range neutron' flux high trip setpoint with inoperable MSSVs must ensure that power is limited to less than the heat removal .

capacity of.the remaining OPERABLE MSSVs. .The reduced high flux trip setpoint also ensures.that the reactor trip occurs early enough in the loss of load / turbine trip event to limit primary to secondary heat transfer and preclude overpressurization of the primary and secondary systems. To calculate.this power level, the governing equation is the relationship q = m Ah, where q is the heat input from the primary side, m is~ the steam flow rate and Ah is the heat of vaporization at the steam relief ~ pressure (assuming no j

(continued) i Mc'Guire Unit 2 B 3.7-3 3/30/98 l u

MSSVs B 3.7.1 BASES

_ ACTIONS A.1 and A.2 (continued) subcooled feedwater). The algorithm use is consistent with the reconsnendations of the Westinghouse Nuclear Safety Advisory Letter, NSAL-94-001, dated January 20, 1994 (Ref. 7). Additionally, the calculated values are reduced by 9% to account for instrument and channel uncertainties.

B.1 and B.2 If the MSSVs cannot be restored to OPERABLE status within the associated Completion Time, or if one or more steam generators have less than two MSSYs OPERABLE, the unit must be placed in a MODE in which the LCO does not apply. To

! achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are' reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE $R 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with t

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(continued) ll McGuire Unit 2 B 3.7-4 3/30/98 i

[T .,

. MSSVs

~B 3.7.1 l -.

BASES (continued)

REFERENCES '5. ASME, Boiler and Pressure Vessel Code,Section XI.

(continued) 1

6. - ANSI /ASMEOH-1-1987.

'7. Westinghouse Nuclear Safety Advisory Letter, NSAL 001. Dated January 20, 1994.

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U.S. Nuclear Regulatory Commission Attachment III May 8, _1998 Page 1 of 6

Background

On January 20, 1994, Westinghouse issued Nuclear Safety

' Advisory Letter (NSAL)94-001 informing utilities that the algorithm used to initially calculate the Power Range

. Neutron Flux High. Trip'Setpoint for inoperable Main Steam Safety Valves (MSSVs) was not correct. As a result of the advisory letter, McGuire issued an interpretation to Technical Specification (TS) Table 3.7.1 to address the concern. Also, the station staff decided to pursue a proposed TS change that would allow higher reactor power

. operation with inoperable MSSVs. During the review of the proposed TS change, Engineering questioned the bases for the existing power level restrictions. A notification to the Nuclenr Regulatory Commission (NRC) was conservatively made on March 20, 1997 to report a possible error in the reactor trip setpoint values contained in TS Table 3.7-1.

Subsequent plant specific analysis confirmed that existing power level restrictions for inoperable MSSVs were non-conservative. Licensee' Event Report (LER) 369/97-04, Revision 0 was submitted to the NRC on April 18, 1997, and Revision 1 was submitted-to the NRC on June 2, 1997. The planned corrective action of Revision 1 of this LER was to submit to the NRC a revision to TS 3.7.1 to incorporate the appropriate setpoint values for inoperable MSSVs.

Description of the Proposed Changes TS 3.7.1.1 allows for plant operation with a reduced number of operable MSSVs at a reduced core power level. This amendment revises power level limits for one, two, and three inoperable MSSV(s) per steam generator (SG).

In order to correct an identified deficiency in the basis L for.the reduced power level limits in this TS, Westinghouse -

-issued NSAL 94-001. -Per the NSAL, the revised algorithm for use in defining the TS'setpoint values is given below- l l

Hi & = (100w.hrgN)/(QK) l where:

L Hi @ = S'afety Analysis power range high neutron flux setpoint, percent l

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U U.S. Nuclear Regulatory Commission Attachment III i May b, 1998 Page 2 of 6 Q = Nominal NSSS power rating of the plant including reactor coolant pump heat, MWt K = Conversion factor, 947.82 Btu /sec/MWt w, = Minimum total steam flow rate capability of the operable MSSVs on any one SG at the highest operable MSSV opening pressure including tolerance and accumulation, as appropriate, lbm/sec he, = Heat of vaporization for steam at the highest MSSV opening pressure including tolerance and accumulation, as appropriate, Btu /lbm N = Number of loops in plant (4 for McGuire)

The values calculated from this algorithm must then be adjusted lower for use in TS 3.7.1.1 to account for instrument and channel uncertainties (typically 9% power).

The maximum plant operating power level would then be lower than the reactor protection system setpoint by an appropriate operating margin.

Based on the above algorithm, revised high neutron flux setpoints are determined for 1, 2, and 3 inoperable MSSV(s) per SG. Following is a sample calculation.

For the single inoperable MSSV per SG, the heat of vaporization is taken at 1314.3 psia and the minimum total  !

steam flow rate capability is the sum of flow rates for the remaining operable MSSVs per SG. The total heat load is the nominal full power rating of 3411 MWt plus an assumed pump l heat of 20 MWt.

H e, (1314.3) = 592.03 Btu /lbm  :

W. (1,2,4,5) = 201 + 201 + 264 + 264 = 930 lbm/sec 1

l Q = 3411 + 20 = 3431 MWt Applying the algorithm from NSAL 94-001, the safety analysis high flux trip setpoint is calculated:

Hi @ = (100(930)(592.03)(4))/(3431)(947.82) = 67.7% RTP L

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U.S. Nuclear Regulatory Commission Attachment III May 8, 1998 Page 3 of 6 Applying the recommended 9% reduction yields the high flux trip TS setpoint of 58% RTP (conservatively truncated).

This calculation is repeated for the combinations of two and three valves per SG having the highest steam flow rate.

Technical Justification I The results of all applicable UFSAR Chapter 15 transients remain within their respective acceptance criteria with the i MSSVs inoperable and the setpoints appropriately modified.

l A discussion of the effect on each of these trannients is l provided.

The limiting transient for peak secondary system pressure is turbine trip (UFSAR 15.2.3). This transient has been I conservatively analyzed with one inoperable MSSV per SG from l 70% power, with two inoperable MSSVs per SG from 51% power, and with three inoperable MSSVs per SG from 31% power. The results of the analyses show that the peak secondary pressure remains below the acceptance criterion of 110% of the design pressure with the inoperable MSSVs.

The transients that are evaluated to determine peak primary system pressure include turbine. trip (UFSAR 15.2.3), partial loss of flow (UFSAR 15.3.1), complete loss of flow (UFSAR 15.3.2), locked rotur (UFSAR 15.3.3), and uncontrolled bank withdrawal at power (UFSAR 15.4.2). In each of these transients, the MSSVs are not challenged before the peak primary pressure is reached. Thus, inoperable MSSVs will not adversely impact these transients.

The transients that are evaluated to demonstrate short-term core cooling capability include excessive increase in feedwater flow (UFSAR 15.1.2), increase in steam flow (UFSAR 15.1.3), feedwater line break (UFSAR 15.2.8), partial loss of flow (UFSAR 15.3.1), complete loss of flow (UFSAR 15.3.2), locked rotor (UFSAR 15.3.3), dropped rod (UFSAR 15.4.3), startup of an inactive reactor coolant pump at an j incorrect temperature (UFSAR 15.4.4), inadvertent opening of a pressurizer safety or relief valve (UFSAR 15.6.1), and steam generator tube rupture (UFSAR 15.6.3). Minimum DNBR '

occurs prior to the lifting of any MSSV in each of these transients, and thus these transients are not affected by inoperable MSSVs. In the uncontrolled bank withdrawal at

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,9 j U.S. Nuclear Regulatory Commission Attachment III May 8, 1998 Page 4 of 6 power (UFSAR 15.4.2) and single uncontrolled rod withdrawal (UFSAR 15.4.3d)' transients, the MSSVs do not lift prior to reactor trip. Since the minimum DNBR occurs shortly after reactor trip, there is not sufficient time for any thermal effects due to delayed MSSV lift to reach the core.

Therefore, these transients are not affected by the inoperable MSSVs.

The transients that are evaluated to demonstrate long-term core cooling capability are the loss of AC power (UFSAR 15.2.6) and feedwater line break (UFSAR 15.2.8). Steam line pressure is controlled by MSSVs with the lowest setpoint during these transients. Maintaining steam line pressure at a higher value would slightly reduce primary-to-secondary heat transfer and increase primary temperature by the same amount as the increase in the secondary side saturation temperature. However, the increase in the secondary side saturation temperature with inoperable MSSVs with the lowest i lift setpoints is much less than the available margin in the primary side subcooled margin. In addition, the long-term heat load would be significantly reduced, since decay heat is directly proportional to the pre-trip power level.

Therefore, hot leg boiling will not occur with the

' inoperable MSSVs.

The steam generator tube rupture transient (UFSAR 15.6.3) is evaluated to ensure that the fission product release to the environment is within the established dose acceptance criteria. If the inoperable MSSVs are those with the lowest lift setpoints, the lifting of the operable MSSys could i occur at a pressure slightly higher than the accumulation i pressure assumed in the original analysis. However, due to )

the significantly reduced heat load resulting from the core power level limitations, the steam line PORVs and operable MSSVs would be able to more rapidly reduce the main steam pressure to the PORV reseat setpoint. Therefore, the operator identification and closure of the failed-open main steam PORV would be accelerated. Another major benefit  !

would be the reduction in the long-term steam release, since '

this is driven primarily by the core decay heat generation, which is directly proportional to the pre-trip power level.

Therefore, the licensing basis analysis remains valid with L up to three inoperable MSSVs per loop.

l.

The locked rotor (UFSAR 15.3.3), rod ( jection (UFSAR 15.4.8), and single uncontrolled rod withdrawal (UFSAR i l

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l U.S. Nuclear Regulatory Commission Attachment III May 8, 1998 Page 5 of 6 15.4.3d) transients are evaluated to ensure that the fission product release to the environment is within the established dose acceptance criteria. In each of these transients, the MSSVs lift for a short period of time relative to the overall length of the transient. In addition, the effect of inoperable MSSVs would be to slightly increase secondary side pressure, which would reduce primary-to-secondary leakage and thus reduce offsite dose. Another major benefit would be the reduction in the long-term steam release, since this is driven primarily by the core decay heat generation, which is directly proportional to the pre-trip power level.

Therefore, the licensing basis analyses remain valid with as many as three inoperable MSSVs per loop.

The small break LOCA transient (UFSAR 15.6.5) is evaluated to ensure that the peak fuel element cladding temperature remains below 2200 F. An evaluation has been performed which shows that the peak fuel element cladding temperature would remain below 2200oF with three inoperable MSSVs per loop. Thus, the inoperable MSSVs are consistent with the licensing basis analysis for the small break LOCA event.

Several transients are not affected by inoperable MSSVs.

The steam line break DNBR (UFSAR 15.1.5), steam line break mass and energy release (UFSAR 6.2.1.4), large break LOCA (UFSAR 15.6.5), and LOCA mass and energy release (UFSAR 6.2.1.3) analyses involve secondary side depressurization, so the MSSVs are not challenged. Transients that do not involve a system thermal-hydraulic analysis are also not impacted by inoperable MSSVs. These transients include statically misaligned rod (UFSAR 15.4.3), boron dilution (UFSAR 15.4.6), inadvertent loading and operation with a fuel assembly in an improper position (UFSAR 15.4.7),

inadvertent ECCS actuation pressurizer overfill (UFSAR 15.5.1), and a break in an instrument line from the reactor coolant pressure boundary that penetrates containment (UFSAR 15.6.2). The MSSVs do not lift during the rod ejection transient core cooling and peak primary pressure evaluations (UFSAR 15.4.8) due to the very short duration of the simulation. The uncontrolled bank withdrawal from zero power (UFSAR 15.4.1) is not affected by inoperable MSSVs since no significant heat transfer occurs across the steam j generator tubes prior to the time of minimum DNBR or peak 1 l primary system pressure.

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4 C' U.S. Nuclear Regulatory Commission Attachment III May 8, 1998 page 6 of 6 In summary, for all applicable UFSAR Chapter 15 transients, all acceptance criteria are met with the inoperable MSSVs.

In order to avoid secondary system overpressurization following a turbine trip event, reductions in the high neutron flux reactor trip setpoint are required. The

! setpoints are listed in Table 3.7-1 of Technical

! Specifications.

In addition to the setpoint changes, references to 3-loop operation (CN3 3. 7.1.1 (b) and Table 3.7-2) are being deleted.

Approval for 3-loop operation was never granted. This deletion is consistent with ISTS.

The appropriate sections of UFSAR Chapter 15 will be revised to incorporate NSAL 94-001 methodology. These revisions will be done following approval of this amendment by the Nuclear Regulatory Commission.

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I U.S. Nuclear Regulatory Commission Attachment IV May 8, 1998 Page 1 of 2 l

No Significant Hazards Evaluation The following analysis is presented, pursuant to 10 CFR 50.92, to demonstrate that the proposed change will not involve significant hazards.

1. Will operation of the facility in accordance with the propoued amendment involve an increase in the probability or consequences of an accident previously evaluated?

The proposed amendment-involves a reduction in the maximum allowable power range neutron flux high setpoints in case of inoperable. main steam safety valves. All applicable UFSAR Chapter 15 transient acceptance criteria are met with the proposed change. Therefore, operation of the facility in accordance with the proposed amendment will not involve an increase in the probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with the proposed ==an h nt create the possibility of a new or different kind of accident from any accident previously evaluated?

No new equipment or operating practice is involved with this proposed amendment. No alteration to any existing hardware is involved with this proposed mmendment. Power Range high neutron flux setpoint calibration is continued to be performed by the same approved procedure. Therefore, operation of the facility in accordance with the proposed amendment will not create the possibility of any new or different. kind of accident from any accident previously evaluated.

.3. Will operation of the facility in accordance with the proposed aman h nt involve a. reduction in a margin of safety? ,

The proposed change is in a more-conservative direction.

All applicable UFSAR Chapter 15 transient acceptance criteria'are met with the proposed amendment. Therefore, '

operation of the facility in accordance with the proposed l

amendment will not involve-a reduction in a margin of safety.

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--U.S.~ Nuclear.~ Regulatory Commission. Attachment IV-May.87.'1998 Page 2 of'2

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'BasedIon the-foreg'oing analysis, it is concluded"that operation Lof]the'. facility in accordance with the proposed amendment will not~ involve significant hazards.

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U. S. Nuclear Regulatory Commission Attachment V May 8, 1998 Page 1 of 1

' Environmental-Impact Assessm'ent

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Pursuant to 10~CFR.-51.22(b), the following~ evaluation is performed to determine if this proposed amendment meets the

-criteria for. categorical exclusion from an environmental impact assessment set forth in 10 CFR 51.22 (c) (9) .

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1) 'As-evaluated in Attachment IV, the proposed amendment does not involve significant hazards.

.- 2 ) No new radioactive. effluents released offsite are created by this proposed amendment. Previously. identified effluents remain within acceptance criteria as evaluated in Attachment III, Technical Justification.

3)' .No new effluents or increase in previously identified effluents are created by this proposed amendment; therefore, there will.not be any increase in either individual or cumulative radiation exposure.

.Therefore', this proposed amendment meets the criteria of 10 CFR 51.22 (c) (9) .for categorical exclusion from an environmental

impact assessment.

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