ML20217J061

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Proposing Changes Which Will Affect Nominal Trip Setpoints & Allowable Values
ML20217J061
Person / Time
Site: Millstone Dominion icon.png
Issue date: 10/15/1997
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML19313D007 List:
References
NUDOCS 9710200070
Download: ML20217J061 (58)
v • d • e


Text

.

.,=

Docket No. 50-423 E.tsft24 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification

~ Instrumentation Surveillances

-(PTSCR 3-30-97)

Marked Up Paaes October 1997 9710200070 971015 PDR ADOCK 050004 3 t%]

l U.S. Nuclear Regulatory Commission B16624\\ Attachment 2\\Page 1 9

MARKUP OF PROPOSED REVISION Refer to the attached _ markup of the proposed revision to the Technical Specifications.

-The attached markup: rollects - the currently issued version of the Technical

-Specifications listed below._ Pending Technical Specification revisions or Technical-Specification revisions issued subsequent to this submittal are not reflecte:I in the enclosed markup.

The following Technical Specification changes are 'mcluded in the attached markup.-

The Technical Specification is replaced in its entirety. The new wording maintains e:

the same Limiting Condition for Operation APPLICABILITY and incorporates the Westinghouse enhanced method of determining equipment operability without the restrictions of the 5 column technical specification.

i_

2.2.1 L

The table is revised to a two column format. The Trip setpoint column becomes a e-L

nominal trip setpoint column. The RTS trip setpoints remain unchanged with the exception of the RCP low shaft speed trip setpoint. Allowable values and Table Notes reflect the analysis of historical instrument data. Editorial changes have been made to certain Pressure values to reflect expression in PSIA instead of PSIG and for consistency in significant digits. Corrections for selected Westinghouse supplied -

Allowable Values are incorporated.

Table 2.2-1 The Technical Specification is replaced in'its entirety, The new wording maintains -

the same Limiting Condition for Operation APPLICABILITY and incorporates the Westinghouse enhanced method of determining equipment operability without the restrictions of the 5 column technical specification.

3.3.2 The table is revised to a two column format. The Trip setpoint column becomes a nominal trip - setpoint column. The ESFAS trip setpointa remain unchanged.

Allowable values reflect the analysis of historical instrument data. Editorial changeo have been made to certain Pressure values to reflect expression in PSIA instead of

- PSIG and for consistency in significant digits.

Table 3.3-4

'et-

____J

kJ.S. Nucicar Regulatory Commission B16624%ttachment 2\\Page 2

. The wording is changed to accommodate the new operability requirements.

Bases 2.2.1,3/4.3.1 and 3/4.3.2-i

_A

..,...~:----------

JAN 31 1988 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

-2.2 ' LIMITING SAFETY SYSTEM SETTINGS

' )

REACTOR TRIP SYSTEM INSTRUMbTATION SETPOINTS 2.2.1-c M rip System In

'nentation and In dSetpoint all u

F e sei consiptent with the Tri oint values sh Table 2.2.

y APPLICABILITY:

As shown for each channel in Table 3.3-1, t

ACTION' s

~

a.: [h,ith a Reactor Trip System Instrumentation or Interlock Setpoint

~

l 1ess conservative than the yarue shown in the Trip Setpoint column

/

but more conservative than"the valua shown in the Allowable Value M g(,M column of Table 2.2-Vaojust the Setpoint consisten fth the Trip' A-Setpoint value. /

t F

b.

With the Re or Trip System Instrumentati or Interlock Setpoint less cons vative than the value shown the Allowable Values column Table 2.2-1, either:

1 Adjust the Setpoint consi ont with the Trip Setpoi t'value of Table 2.2-1 and date a within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that ation 2.2-1 was satisfied for t affected channel, or i-2.

Declare the ch nel inope able and ap the applicable ACTION statement r frement of Specific 3.3.1 until the channel

I is restorpd to OPERABLE status wj itc Setpoint adjusted consisteftt with the Trip Setp hit value.

untion 2.2-1 Z + R + 5 t TA Where:

{

Z = The value from lumn Z of Table 2.2-1 for he affected channel, R = The "as se red" value (in percent an) of rack error for the i

affected annel.

5 = Either the "as measured" valu (in percent span) of e sensor error, or the value from C S (Sensor Error)

Table 2.2-1 for the affected channel and

-TA = The value from Column TA (Total Allowa of Table 2.2-1 for the affected channel.-

4

.)

MILLSTONE - UNIT 3 2-4

~o

I N S E R Y */) ",

PAG-E / 0F /

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 - LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS

- 2.2.1 The Reactor Trip System Instrumentation Channel and Interlock Channel shall be OPERABLE.,

APPLICABILITY: As shown for each channelin Toble 3.3-1.

AGyQB:

l o.

With o Reactor Trip System Instrumentation Channel or Interlock Channel Nominal Trip Setpoint inconsistent with ihe value shown in the_

NominalTrip Setpoint column of Table 2.2-1, adjust the Seipoint consistent with the Nominal Trip Setpoint value, b.

With a Reactor Trip System' instrumentation Channel or Interlock Channel found to be Inoperable, declare the channelinoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channelis restored to OPERABLE status.

A.2 j

h.i

  • T/9 6 L2 2 R-l r

~ ~,.'

g REACTOR TRIP SYSTER INSTRUNENTATION TRIP SETPOINTS SENSOR 3 f

OWANCE ERROR

/t/oMWAL FUNCTIONAL UNIT II I

f51 TRIP SETPOINT ALL0lm9tE YALUE 1.

Manual Reactor Trip N.A.

N.A.

N.A.

N.A.

N.A.

b g

2.

Power Range, Neutron Flux g #75 l

l

{

a.

High Setpoint

1) Four Loops Operating 7.

56 0

\\

< 109% of RTP**

1 Tit:iE of RTP**

g

~

2)ThreeLoopsOperating 7.5 4.

0 801 of11TP'*

(1ER of RTP**

g j

b.

Low Setpoint 8.3 4.56 0

25% of RTP**

s[2M5 of RTP**

/g gg l

3.

Power Range, Neutron Flux, 1.6 0.5 5% of RTP** with 110':3Elef RTP** with High Positive Rate A time constant E a time constant l

t 2 seccWs 1 2 seconds t

4.

Deleted 5.

Intermediate Range, 17.0 8.4 0

g 25% of RTP**

1 Neutron Flux

, f,og l

6.

Source Range, Neutron Flux 17.0 10.01 4

1 lX x 10** cps

+C

~

r 1

- 17 I i

7.

Overtemperature AT 7

a.

Four Loops Operating i

c_

C J

1)ChannelsI,II 10.0 8.14

.61 + 1 See Note 1 See Note 2 esp +

ssi

~

2) Channels III, IV 10.0 7.17 1.

+ 2.60 Note 1 See Note 2 (T

+ Press)

E i,

p

" RTP - RATED THERNAL POWER

TABLE 2.2-1 fCentlasedl i

REACTOR _ TRIP _ SYSTEM Ill5filllNDITATION TRIP _SETrelRf5 TOT SENSOR ALL RROR A/om /N/9 L l

FUNCTIONAL Ullli fiAl Z

TRIP SETFellli N TM.NE i

b.

Three Leops Operating

{

i

1) Channels I,11 10.0 6.

1.71 +

33 See Note I See Note 2 (Temy +

ss) r

2) Channels 111 IV 1 0 5.83 1.71 + 2.

See IIste 1 See IIste 1 1

(Teny+ Press) i

[

8.

C...;x;;- AT (Fear Loops operating) 4.8 1.28

1. I See IIste 3 See Note 4 l

I 1877.(

9.

Presseriter Pressure-Lew 5.0 1.77 3.3

' t I9OS psia Allegelpsia 7

2387.Y l

10. Presseriter Pressure-Migh

.8 1

3.3 d 2385 ps8a s[2385l psia i

87 3W

11. Pressurizer water Level-Migh 8.

5.1 2.7 i 891 of lastrument glge.3[of lastnnant

[

(

Ilian span i

W 87 #

i

12. Reactor Coolant Flow-tow 2!5 1.52

.78 901 of leep 199:14of1**p

^

a sign new*

design flow# g,g

13. Steam Generator Water 18.10 I 64 1.50

' @ 8.1 of narrow 1 17:431 of narrow l

Level low-Lew range lastrument range Instrement span span

14. General Warning Alaru N.A.

N.A.

N.A.

N.A.

72 9f N.A.

g 72,

15. Lew "' eft speed - Reacter 3.8 0.5 0

hb of rated 192:44 of rated so I

i Cool:

Pumps speed speed I

I E

l l

  • Minimen Measured Flow per Loop - 1/4 of the RCS Flow Rate Limit as listed in Sectlen 3.2.3.1.a (Fear Leops Operating);

[

l 1/3 of the RCS Flow Rate Limit as ilsted in Section 3.2.3.2.a (Three Loops Operating)

I ce ce ca.

i

BEACTOR TRIP s hita InsL M TATION TRIP SETPOINTS

~

TOTAL SENSORI VLLOWANCE ERROR A/om I"dL

~

FUNCTIOW1 UNIT NA)

Z f$1 TRIP SETPOINT '

ALLONA8LE YALUE

[

16. Turbine Trip a.

Low Fluid 011 Pressure N.A.

N.A.

N.

@500psig 1 450 psig b.

Turbine Stop Valve N.A.

N.A.

N.A.

@l%open 2 1% open Closure

17. Safety Injection Input.

N.A.

N.A N.A.

N.A.

N.A.

l from ESF

18. React 9r Trip System Interlocks 7.o a.

Intermediate Range N.A.

.N.A.

N.

@l x 10 ' amp 21x 10'" amp 4

Neutron Flux, P-6 l

b.

Low Power Reactor Trips Block, P-7 N' y l

1) P-10 input (Note 5)

N.A.

N.A.

.A.

11% ef RTP**

1 12>1$ of RTP**

l

2) P-13 input N.A.

N.A.

N...

10.T RTP** Turbine 5

RTP** Turbine 1se Pressure Impu s Pressure

? Equivalent Equivalent fo.f 7 c.

Power Range Neutron Flux, P-8 38if l

1) Four Loops Operating N.A.

N.A.

.A.

S 37.5% of RTP**

5f of RTP**

2) Three loops Operating N.A.

N.A.

'N.

37.5% of RTP**

$ M of RTP**

' 38./I

    • RTP = RATED THERNAL POWER 5

'i MILLSTONE - UNIT 3 2-7 Amendment No. pp.

}

~

ne r e>

85 3

mott 2.2-1 toontinueot

~

30/25/93 REACTOR TRIP 5Y5its INSTRUNENTATION TRIP SETPOINTS TOTAL

\\

SENSOR

^

ALLOWANCE ROR A/om/Not-

~

FUNCTIONAL UNIT

\\fTA) i TRIP SETPOINT ALLOWABLE VALUE N

51.8%

d.

Power Range Neutron N.A N.

N.A.

51% of RTP**

5 M of RTP**

Flux, P-9 r0Y[

e.

Power Range Neutron gN.A.

N.A.

.A 9% of RTP**

A lh91lof RTP**

Flux, P-10 (Note 6) 19.

Reactor Trip Breake.s N.A.

.A.

N.A N.A.

N.A.

20. Automatic Trip and Interlock N.A.

.N.A.

N.A.

N.A.

N.A.

Logic

21. Three Loop Operation N.

.A.

A.

N.A.

N.A Bypass Circuitry N

i RTP - RATED THERMAL POWER s

MILLSTONE - UNIT 3 2-8 f.nznk rit No. 17.

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

~

!,{t i

i Ek IABLE 2.2-1 frontinued)

@9 li-4 TA8i.E NOTATIONS 3_p NOTE 1: OVERTEMPERATURE AT I

I AT II + # 5)

.. SST IEl~E2

[T

- T'] +

(P - P') - f (AI))

R l

1 0

E (1 + r 5) I + r 5I 2

3 (1 + r 8I I + '65) g l

9 i

5 w

q M

g Where:

AT

- Measured AT by Reactor Coolant System Instrumentation; Q

'I+f3l

- Lead-lag compensator en measured AT; I+r52 r,r2

- Time constants utilized in lead-lag compensator for AT, r 2 8 s, r 5 3 83 !

g g

2 1+15

" N CN* " "

3 r,

- 1,me constants.ini,a in the ia, c-sato, r., A1, r,.. s; ATO Indicated AT at RAM ML M; I

K

=1.20(FourLoopsOperating);1.20(ThreeLoopsoperating);

g K

- 0.02456;

{

2 l+T5 4

- The Tunction generated by the lead-lag compensator for T'" dynaufc 1+y5 7

5 compensation; g

{

i r., r,

- Ti. costants.tniza i. u,e iew-ia, c-sator ror T,,,, r,1 a s,

=

5 s 4 s; l

7 l 'I Average temperature *F; T

1 I

- Lag compensator on measured T J;-

y I+73 6

I r,

- Time constant uttilzed in the measured T,q 1ag :ompensator, r6 - O s; f

=

=

t

.0-

l e

sG JMtf 2.2-1 fCentinuedl l

}%a 8

i Pli 1ABLE flDTAT10ftS fContinuedl g

NOTE 1:

(Continued)

}

T' s 587.l*f (flominal T,,, at RATED TifERMAL PUE R);

K 0.001311/ psi; 3

presseriter pressure, psla; p

p'

= 2250 psla (llaminal RCS operating pressure);

Laplace transform operator, s ';

,5

=

and f (AI) is a function of the Indicated difference between top and bottom detectors of th-I

'l*

power-range neutron ton chambers; with gains to be selected based on measured instrument response 5

during plant startup tests such that:

(1) For g

- gb between -261 and + 3%. f,(AI) - 0, where g and g+are percent RATED TH[lMAL' POWER in thI top ar.d bottom halves of the core respectively,g and g gb 'is total THElWIAL POWER in I

g percent of RAi[D THERMAL PO K R; (2) Foreachpercentthatthemagnitudeofg!ue-gexceeds-26%,theATTripSetpelatshallbe automatically reduced by 3.55% of Its va at RATED Tif[RMAL POWER; and (3) for each percent that the magnitude of g, - g exceeds +3%, the AT Tria Setpoint shall be automatically reduced by I.ge% of its value at RATED THERMAL POWER.'

I k

E h

NOTE 2:

hechanne15maximumTripSetpelatshallnotexceeditscomputedTrip5etpaintbymorethan g

T span (four loos Operation); 2.7% AT span (Three Loop operation). f lg THE in19%. Ion um c H A WNEl. 83 L EW TR1P SEWO!W W N

x

?

ur'CD 77Cif SETf;bsM By' incgE THow W R "0W'4

  • o o 7.s c_omy (I) 0.+X bT SPlys part rHc AT CHr9twEL 2

(1) o 4 jY, 6 r SPM" Fo t2 YHE Tyc C NfWWEL L3) oast-f g y SPnd foK THE PR&rl!Z6 MMW UWW

(

(.+) o g '4 6 Y SPAW Fo At THE "f (AI) C NA"""'

1

!$b IABLE 2.2-1 iContinsed)

[5

-p IABLE IIOTATICffS fCentinmed) 4 h N IIDTE 3:

DVERPOWER AT 7

gy (1 + f.5)

(

I

) g gy, (g,. g, ( r,S

)

(

I

) T - K, [T I I

I - T*] - f, (al E (1 + r,5) (1 + r,5) (1 + r,5) (1 + r,5) (1 + r,5) -4 Where: AT As deffned in flote 1 I + T'I As defined in Note 1, = I + r,5 = 3 D r,, ' r, - As defined in Note 1, R,. I 1 3 3, 7g - As defined in flote I, R, e D o g' O r, - As defined in llote 1 Rk ATO - As deffned in flote 1, K, I.09, K, ~ = 0.02/*F for increasing average temperature and 8 for decreasing average temperature, k

  1. 3 x

The function generated by the rate-lag compensator for T dynamic 7 f I + r,5 compensat1on,

s

= [ r, Time constants att11 red in the rate-lag compensator for T , r 2 18 s. l r o 1_ As deffned in flote 1, ye I + r,5_ = r, - As deffned in llote 1 W

I s nz ' h. TABLE 2.2-1 fContinuedl i g DN IABLE NOTAllolls fCentinmed)

6

.g 4 1r010 3.- (Continued) Kg

0. nlH/*F iw T > V and Kg = 0 Tw T $ T*,

= As defined in flote I, ,T

T*'

Indicated T,y at RAT [D Tit [PNLL POWER (Calibration tesiperature for AT instruentation, s 587.l*F), S As defined in Note I, and t f IAII

  • I'" *II II*

2 re 4;.echannel'smaximumTrioSetpointshalinotexceeditscomputedTripsetpointbymorethan NOTE 4: . QJ1 AT span. (Four toen n--ration) l 6 i NOTE 5: Setpoint is for increasing power. t NOTE 6: Setpoint is for decreasing power. i ~ ~,n ,,, n..u r e na ~~ o n or ~~ ~ ~ g bT C.1-Jr3 M N EL AND O*Y b Y 5# A" --m Ut. 0 f D t e ~, 5 [

2.2 LIMITING SAFETY SYSTEM SETTINGS l JAN 311986 DASES i I-2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip.Setpoint Li6specified i Sable 2.2-1 are the nominal) values at which the eactor trips, i set for ch functional unit. The Trip j Setpoints have b selected to ensure that core and Reactor Coo)elit System are pre nted from exceeding thei afety limits during npulal operation 4 and design b is anticipated operati occurrenc'es and to asMst the Engi-neered Safefy Features Actuation a in mitigating the e sequences of - e t accinen (. The Setpoint for a R tor Trip System or in lock function is i i consi red to be adjusted con tent with the nominal ue when the "as me red" Setpoint is withi he band allowed for cM bration accuracy and / I J strument drift. / { j To accommodate e instrument drift ass d to occur between oper fonal tests and the acc acy to which Setpoints n be measured and calibt ed, Allowable Value for the Reactor Trip S points have been specifi in Table 2.2-1, peration with Setpoin less conservative than Trip Set-point but hin the Allowable Val is acceptable since an Iowance has been made in e safety analysis to a onmodate this error, optional provision has be included for deterni g the OPERABILITY of a annel when its Trip j Setpp nt is found to exceed e Allowable Value. T. methodology of this j t option utilizes the "as sured" deviation from specified calibration i point for rack and sen componentsinconjun on with a statistical c n-4 ation of the other u ertainties of the ins ntation to measure the ocess ; variable and the u ertainties in calibrat ng the instrumentation. Equa-a tioi. 2.2-1, Z + + S < TA, the intera va effects of the error n the rack and the sensor and the "as measure " values of the errors a nsidered. Z, t as specifie n Table 2.2-1, in ont span, ir, the stati al summation of 1-r errors as J in the analysis cluding those associat ith the sensor and i rack dr and the accuracy their measurement. T r Total Allowance is thed}fference,inpercent pan, between the Trip oint and the value sed is the "as measure ' devia-j in he analysis for Rea r trip. RorRackEr(r tje)n, in percer.t span or the affected chan from the specified ip Set-1 point. S or Sensor tror is either the "a asured" deviation the sensor from its calibra n point or the value ecified in Table 2. , in percent span, from the alysis assumptions, se of Equation 2.2-allows for a sensor drift actor, an increased k drift factor, an rovides a threshold value for PORTABLE EVENTS. - e methodology to der e the Trip Setpoint s based upon combining all i 'of the uncertainties in channels. Inheren to the detemination of the l-Trip Setpoints are the agnitudes of these c nnel uncertainties. ensors and other instrumentati utilized in these c nels are expected capable of operating wit;.in. allowances of thes neertainty magnitu s. Rack drift in excess of t ' Allowable Value exhi its the behavior tha the rack has not-7 met its all e. Being that ther s a small statisti 1 chance that this will happe an infrequent excessJ e drift is expected Rack or sensor drift, in excess of the allowance that is more thah occasion 1, may be indicative of j ( more serious problems and should warrant further investigation. MILLSTONE - UNIT 3 8 2-3 MSM 6 i l d

. - -.. ~ i July 11,1995 i y joy;cf WI79/ LIMITING SAFETY SYSTEM SETTINGS //KS6/T7~ 8, sAsts / REACT 0 RIP SYSTEM INST ENTATION SET NTS(Continue The rious Reactor tr circuits auto tica11y open e Reactor tr breakers who ver a condition nitored by th eactor Trip S tem reaches reset or cale lated level. In dition to re ant channels d trains, th ~ sign approach rovides a Reacto Trip System w h monitors n rous system va tables, theraf a providing Trip stem function 1 diversity, e functional cap 111ty at the s acified trip sett g is required or those antic atory or diver Reactor trips for which no dire credit was a umed in the sa ty analysi to enhance th overall reliabil of the React Trip System. The eactor ip System init tes a Turbine tr signal whene Reactor trip s itiated. This prevents e reactivity ins tion that wou otherwise res t j fr excessi Reactor Coola t System cooldown nd thus avoid unnec:ssary actu tion of t e Engineered S ety Features Actu tion Systam. s t l Manual Reactor Trio The Reactor Trip System includes manual Reactor trip capability. 9 fewer Rance. Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting L. sed for a High and Low Range trip setting. The Low Setpoint trip provides protection during suberitical and low power operations to mitigate the consequences of a power excursio'n beginning i from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels. The High Setpoint trip is reduced during three loop operation to a value consistent with the safety analysis. The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated l below the p-10 Setpoint. l Power Rance. Neutron Flux. Hiah Positive Rate l The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing. s Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents. I MILLSTONE - UNIT 3 8 2-4 Amendment No.116 moe UM

. ~. INSERT "B" TO pas? B 2-3 L M1-REACTOR TRIP SYSTEN INSTRUMENTATION SETPOINTS The Nominal Trip Setpoints specified in Table 2.2-1 are the nominal values at which the reactor trips are set for each functional-unit. The Allowable Values (Nominal Trip Setpoints i the calibration tolerance) ire i considered the Limiting Safety System Settings as identified in 10CFR50.36 and have been selected to ensure that the core and Reactor Coolant System are prevented'from exceeding their. safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered j Safety Features Actuation Systwa in mitigating the consequences of accidents. i The Setpoint for a Reactor Trip System or interlock function is cansidered to be consistent with the nominal value when the measured "as left" setpoint is - t within the administrative 1y controlled (1) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint). Additionally, the Nominal Trip Setpoints may be j adjusted in the conservative direction provided the calibration tolerance j remains unchanged. Measurement and Test Equipment accuracy is administrative 1y controlled by plant procedures and is. included in the plant uncertainty calculations as defined in WCAP-10991. Operability determinations are based on the use of y Measurement and Test Equipment that conforms with the accuracy used in the plant uncertainty calculation. i The Allowable Value specified in Table 2.2-1 is the initial value for consideration of channel operability. If the process rack bistable setting is measured within the "as left" calibration tolerance, which specifies the { difference between the Allowable-Value and Nominal Trip Setpoint, then the channel is considered to be operable. Additionally. an administrative 1y controlled limit for operability of a j device is determined by device drift being less than the value-required for i-the surveillance interval. In the event the device exceeds the administratively controlled limit, operability of the device'may be evaluated by other device performance characteristics, e.g., comparison to historical device drift data, calibration characteristics, response characteristics, and short-term drift characteristics. A device (RTD, relay, transmitter, process-rack module, etc.), whose 'as found" value is_in excess of the calibration tolerance,- but within the operability criteria (administrative 1y controlled limit), is considered operable but must be recalibrated such that the "as left" value is within the two sided ( ) calibration tcierance. Plant procedures set administrative limits ("as left" and "as found" criteria) to control the determination of operability by setting minimum standards based on the methodology in WCAP-10991 and the uncertainty values included in the != determination of-the Nominal Trip Setpoint, and allow the use of other device characteristics to evaluate operability. REPORTABLE EVENTS are identified when the minimum number of channels required to be operable are not met. The methodology, as defined in WCAP-10991 to derive the Nominal Trip g Setpoints, is based upon combining all of-the uncer+ainties in the channels. Inherent in the determination of the Nominal Trip Setpoints are the magnitudes 4 of these channel uncertainties. -Sensors and other instrumentation utilized in I i

INSERT 'B' TO PAGE B 2-3 fcont'd) these channels should be capable of operating within the allowances of these uncertainty magnitudes. Occasional drift in excess of the allowance may be determined to be acceptable based on the other device performance 1 characteristics. Device drift in excess of the allowance that is more than I occasional, may be indicative of more serious problems and would warrant i further inv9stigation. 4 The various reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to the redundant channels and trains, the design approach provides Reactor Trip System functional diversity. The function,il capability at the specified trip setting is required for those 4 anticipatory or diverse reactor tri)s for which no direct credit was issumed in the safety analytis to enhance tie overall reliability of the Reactor Trip j System. The Reactor Trip System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actu..lon System. I s i i -g .a ,n,,-n. .-. --,. ---,,, ~. -,-v--,,-

i iI INETRLMfWTATIGN 6/28/94 1 1/4.1.7 enc 1NffRED 1AT~TY FfATURf1 ACTUATfDN SYSTEN INfTRLMfRTAT!DN LIMITINC CONDITION FOR DPERATION J ^^ l 3.3.2els and1mterleintered 5)stfiestures Actuatjen'$ stN ESFASMItr The E be OhEJARC with the i a hown in Table 3. 3 shal ifip se t consp nt with the valug hown in the T.rf Setpoint c,el i APPLICABILITY: As shown in Table 3.3 3. i j A W QN: y i (

a. [ With an ESFAS Instrumentation'or Interlock Trip,56tpoint trip less i

conservative than the vMe shown in' the Trip 4etpoint column but i \\ ff'the value shown in,He Allowable Value column more conservative t fass#f~ of Table 3.3 4, ust the Setpoint co istent with the Trip Sgtpoint valu p b. With anJ FAS Instrumentation Interlock Trip setpoint s conse ative than the valu own in the Allowable Val column of Ta)) 3.3 4, either: 1 I. Adjust the se int consistent with the J p Setpoint value of Table 3.3-and detemine within 12 bo1Trs that Equation 2.21 l was sati led for the affected ch el, or 2. De lire the channel inoperabi and apply the applicable ik i atoment requirements opTable 3.3 3 until the channe) { restored to OPERABLE (Titus with its setpoint adju d consistent with t rip Setpoint value. l Equation 2.2 1 R + $ 1 TA { Where: I = The ve from Column I of Tab .3 4 for the affecte i channe), = The "as sensured* va. (in-percent span) of error for the i affected channel, $ = Either the ' measured

  • value he value from Column (5in ppe ni spin) of the sensor error, o 5'nsor Error) of Table 3.3 4 e

for t affected channel, and } TA = e value from Column TA otal Allowance) of Table,) 4 for the affected channel. !6 c. With an ESFAS instr tation channel or inter 1 inoperable, take the ACTION shown in' Table 3.3 3. j i MILLSTONE - UNIT 3 3/4 3 15 Amendme nt No .91 i

d Insert C to Page 3/4 3-15 3.3.2 The Engineered Safety features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3 3 shall be OPERABLI:. APPLICA91L111: As shown in Table 3.3-3. ACTION: a. With an ESFAS Instrumentation Channel or Interlock Channel Nominal Trip Setpoint inconsistent with the value shown in the Nominal Tri) Setpoint column of Table 3.3 4, adjust the Setpoint consistent wit 1 the Nominal Trip Setpoint value. b. With an EsfAS Instrumentation Channel or Interlock Channel found to be inol>erable, declare the channel inoperable and apply the a>plicable ACTION statement requirements of Table 3.3 3 until the ciannel is restored to OPERABLE status.

March 11 1991 0 TABLE 3.3-4 D e *I: [.} g ENGINEERED SAFETY FEATURES ACTUATION 5T5 tin INSTRINIENTATION TRIP SETPOINTS o, u, ~ oM SENSOR. TOTAL A/o m No4-c FUNCTION 4L UNIT Alt 0WANC fTA) Z L TRIP SETPOINT ALLOWABLE VALUE 5 1. Safety Injection (Reactor Trip, Feedwater Isolation, Control Building Isolation (Nanual Initiation Only), Start Diesel Generators, and Service Mater) a. Manual Initiation M.A. N.A. N.A. N.A. N.A. b. Automatic Actuation Logic A. N.A. .A. N.A. 17.7 8S/6 N.A. 17 ?/'J/6 y c. Containment Pressure--Nigh 1 3.3 01 1. (t%44alg t* d. Pressurizer Pressure--tow 3 1891 & M MW,

1) Channels I and II

.22.16 20.1 1.5 .3 ps ( 1 9 R 5 psig

2) Channel III and IV 22.16 15.6 3.3

> 1877. 1 3 98'!L PSin IFL?T. fsp9 h658.6psig* A M sig* e. Steam Line Pressure--Low 17. 5.6 2. ssv.7 2. Containment Spray (CDA) a. Nanual Initiation M.A. N.A. N.A. N.A. N.A. b. Automatic Actuatiot logic A. N.A. A. N.A. N.A. and Actuation Relays y J J c. Containment Pressure--High-3 3.3

1. I 1.75 Ge&e-ps]ig Sg 3.

Containment Isolation o a. Phase "A" Isolation O

1) Nanual Initiation N.A.

.A. N. N.A. N.A. 9

TABLE _3.3-4 h ENEIIEEREN 5AFETT FEATURE 5 MTMTION SYSTIN IW5TNUMENTATISRJRIL3ETP91NTS r-SENSOR [ M '" # ' E TOTAL FUNCTIOML UNIT E fTA) f51 TRIP SETPOINT ALLENBLE VALUE 3. Contalament Iselatten (Continued)

2) Automatic Actuation Logic N.A.

.A. N.A. N.A. N.A. and Actuation Relays-

3) Safety injection See Item I. above for all Safety Injection Trip Setpoints and Allowable Values.
b. Phase 'B' Isolation
1) Manual Initiatten A.

N.A. cA. i N.A. N.A.

2) Astematic Actst'clon N.A.

N.A. N.A. N.A. M Logic and Actuation 2 2 ~1 &

g., ys z n
3) Centainment Pressure--

3.3 1.01 1 5 s .8 p Ts b 4_ 1(.8 psg Nigh-3

c. Purge Iselatien N

.A N.A. 1 I R/h i I R/h y{' j 4. Stema Line Isolatten

a. Manual Initiatten M.A.

N.A. .A. N.A. N.A.

b. Automatic Actuation Logic N.A

.A. N.A. N.A. N.A. 1 g and Actuation Relays / e-pp n.1 Pclu s?9r% i f {

c. Containment Pressure--High-2 j 3.3 1.01

.5 @.9 psID ST.Spsis g

d. Steam Line Pressure--Low 17.7

.6 2.2 @658.5psig* 2 Ig* I = l co w 5 0.5 0 ) S 100 psf /s** S h yst/s** 8 N-

e. Steam Line Pressure -

k.0 y Negative Rate--High / sfog y g 3

t July b 1990 n $y TABLE 3.3-4 (Continued) E ts M ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRtmenTATIO c: / di p SENSoll TOTAL \\ ERROR A/oth /A/AL FUNCTIONAL UNIT All0WANCE (141 I (S) TRIP SETPGiW7 ALLOWABLE VALUE 5. Turbine Trip and Feedwater isolation T Automatic Actuation logic N.A. N.A. N.A. M.A. N.A. a. Actualion Relays g,g *f, se gg b. Steam Generator, Water 5.25 3.76 1.50) (i so.45% of 5 of narrow Level--Iligh-High (P-14) l narrow range range' Instrument

  • f instrument span.

4 span. 'd

c..

Safety injection Actuation See Item I. above for all Safety injection Trip Setpoints,and ff Logic Allowabic Value.. d. 7,, Low Coincident with Reactor Trip (P-4) ) hbbeh f

1) Tour Loops Operating A.

N.A.I 564*F 2$60.6*F[ - l

2) Three loops Operating N.A.

N.. N.A. 564*F 2 h 6. Auxiliary Feedwater I M* -.q a. Manual Initiation .A. N.A. A. N.A. M.A. h. . ? ~- Automatic Actuation logic M.. .A. N.A. N.A. N.A. and Actuation Relays D c. Steam Generator Water Level--Low-Low l7-8b f j

1) Start Motor-Driven 18.10 16.64 1.50j

> 18.]OY. of 2 07.ll7def narrow Pumps L narrow range range lastrument instrument span. span. t 1

a GB/?"/94 l i TABLE 3.3-4(Coutinued) j

5 Emp SAFETY FEATURES ACTlikTIM SYSTDI INSTRIBENTATIM Tity SET 4glg3 i

i

F i
  • 3 SUISON i

9 TOTAL / A/0'"' A/'+ L - FIRETIlmlL IIIIT MiinmitE (TA) f51 TRIP SETFOINT ALLANELE INLEE E 5. Auxiliary Feedsater (Continued) i7.@ i ti / T' w

2) Start Turbine-18.10 16 1.50 i Q)lS.195of 1 7.1 of narrow i

i Driven Pumps j marrow' range trement [ Instrument. span. t span. i l d. Safety Injection See Item 1. above for all Safety Injectlen Trip Setpoints and A11emable 4 Values. [_ e. Less-of-Offsite Power N.A. N. N.A. 2 2006V 2 2720F i Start Motor-Ortwen Pumps wk f. Containment Depressurizatten See Item 2. above for all CDA Trip Setpoints and Alleunble Values. w Actuatten (CDA) Start l Motor-Driven Pumps 7. Centrol Building Isolation i a. Manual Actuation N.A. N.A. N.A. N.A N.A. b. Manual Safety Injection N. N. 4 .A. N.A N.A. k Actuation } l k c. Automatic Actuation N.A. N.A. N.A. N.A. N.A. j Logic and Actuatten g Re'.ays , 7,7 prh. b d. Centainment 3. .01 1,75 (3.0 psp $(3.8 pstj ? Pressure--Migh I e. Centrol But! ding N.A. N. A.' N.A $1.5 x 10-5pcl/cc $1.5x10-5,cgfcep I E g Inlet Ventilation j Radiation i {

~ g3 IABLE 3.3-4 (Contf d l ^ 11/30/94 h 4 ,E,, - INGINEERED SAFETY FEATURES ACT11ATION 5V$iut iiorm,.anATION TRIP 5Eirg FUNCTIONAL UNIT T0hLL. sEs.M ERROR AIO M /A A L / ALLOWANCE \\fTAl Z JS1 IRIP SETPOINr AlpfREJilE 8. Loss of Power

a. 4 kV Bus Undervoltage

.N.A. N.A.. N.. 2800 (Loss of Voltage) 1 2720 volts volts with with a i 2 a 5 2 second second time time delay. delay.

b. 4 kV Bus Undervoltage N.A.

N.A. N.A. 3730 volts 1 3706 volts l (Grid Degraded Voltage) R ith a S 8 with a S 8 second time second time Y delay with ESF dely wf th ESF 8 actuation or actuation or < 300 second ~ 300 second Ilmedelay Itsedelay without ESF without ESF actuation. actuation. 9. Er.gtneered Safety Features Actuation System Interlocks a. l919-71%!*-- 5 Pressurizer Pressure, P-Il N.A. N.A. N.A. W 2 o01. IPSI 4. $g b. Low-Low T,,,, P-12 = A. .A. N.A. p53*F 2@9.6*D g

c. Neactor Trip, P-4 N.A.

N.A. N.A. N.A. N.A. f f1.C 'F f

10. Emergency Generator Load N.A.

N.A. N.A. N.A. N.A. y Sequencer ? (

j ~ March 11, 1997 No C WNK-E F o S I N F O C> N W IABLE 3.3 4 (Centinued) a TABLE NOTATIONS 4 )

  • Time constants utilized in the lead lag controller for steam Line Pressure L are t 1 50 seconds and s. i l seconds.

ensure that these time consta,nts are adjusted to these values.CENNEL CAL

    • The ties constant utilized"in the rate lag controller for steam Line Pressure-Negative Rate High is greater than er equal to 50 seconds.

shall ensure that this time constant is adjusted to this value. CHANNEL CALIBRATIO ~ l 6 i i 4 s i i G ( l M" IIE

  • EII 3 3/4 3-31 Amendment No.134 we

January' 31, 1986 l ' -- 3/1.3 lN$78t*ENTAT10N pa5ES 3/4.3.1 and 3/4.3.2 REACTOR TRIP $YSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATUFE5 ACTUATION 5YSTEP ]N5TKvMENTAT10N The OPERAE LITY of the Reactor Trip System and the Engineered Safety Festures Actuation Systes instrumentation and intericcks ensures that: (1) the asscciatec ACTION and/or Reactor trip wl11 be initiated when the parameter monitored by each chant.e1 or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained. (3) sufficient redundancy is main-- ~ tained to permit a channel to be out-of service for testi or maintenance and (4) sufficient system functional capability is availab e from diverse I

  • o
  • --te rs.

..m The OPERABILITY of 'these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility . design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used iri the ssfety analyses. The Surveillance Requirements spect-fisc for these systems ensure that the overall systee functional capability is maintained comparable to tne original design standardsi. The periodic surveil. w ge nbc,'i,Mib.. lance tests performed at.the minimum frequencies are suffici,ent.,to. demonstrate 3f q Mgl . C. .'..- 76e Engineered safety Features A' Ncion Sysisam instiumenkItikTji[,, 1 1 J i ei: ,,utJ9 eW.2*.fste. 24 capabWty. .q M*.hidn T:s

  • Setpointa specified'in Table 3.3+are.the nominal values at which.the histables are set for each functional uMf. A Setpoint is considered te be adjustad,

' inc i s within t consistent with the nominMalue when the "as sensured* the bane allowed for ,1bration accuracy. To accommodat the instrument-drift ' ass to occur between enerationa19,. tests and theMeuracy to which Setpolnts be esasured and calibrated. I~ ". dc ~' + ' Allowable valuis'for the Setpoints ha've4een specified in Table 3.3 4. Ocora-tion with'Setooints less conservativd than the Trip Setpoint but within the A11oys61e Value is acceptable s,imfe an allowance has been made in the safety, apaTysis to accommodate this,mfror. An optional provision.has been,$ccluded f#( for determining the OPERA 41GTY of a channel when its Trip seteniet is found 'p to exceed the A11owab1g halue. The methodology of this option' utilizes the "asmeasured"deviapie'nfromthespecifiedcalibration ifit for rock and in conjunction with a statistical ination of..the other sensor components'the instrumentation to sensure process variable and the uncertainties of uncertaintio(in calibrating the instrumentatipn'. In Equation. 3.3-1, I+R5 , the interactive. effects of thVerrors in the rack.and the the "as measured" values of,the errors are considered; s sense spep ed in Table 3.3-4, in percent n, is the stat'istical s of errors assumed in the analysis cluding those associated"

th's sensor, their sensurement.

., otalillowance span, R or tack Error is,rTA,$he "as"seasured" fand rock drift and the accura is the difforence, in perc eeviation, in the percenp pan, for the affected channt) fron.the specif.ied i Trip 5etpoint. 5 or Slea ser Error is either the "af measug"g,devjation, St!LLSTONE - UNIT 3 5 3/4 3-1 s 4 u

6/28/94 INSTRUMENTATION - h

  • ^555

.A.y REACTOR TRLP SYS"EM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS"RUMEN"ATION-(Continued) ^ ( the sensor from its-ca ration point or the a 6 specified in Tab 1 X.f 4, 3 in percent span, f .the analysis assumpt s. Use of Equation

4. -1 allows for a sensor dri factor, an increased ck drift factor, an rovides a threshold val for REPORTABLE EVENT The thodology to derive e Trip setpoints i ased upon combinin 11 of th neertainties in the annels.

Inherent,te'the determination o e T Setpoints are the itudes of these c nel uncertainties, sor and instrumentation <zod in these ch els are expected to capable of l "" ' ' operating within t llowances of ther uncertainty magnit Rack drift ? in excess of 11owable Value exhW ts the behavior t the rack has not met its all ca. Being that t e is a small statt H al chance that this l } will hap , an infrequent exc sive drift is exp Rack or sensor drift, in exc s of the tilowance t is more than oc tonal, may be indicative of nor erious proolens an hould warrant fur r investigation. The measurene i of response time he specified fre uencies p ides -.-e. essurance that t Reactor. trip and e En ineered Safeti eature ctuation it'g:.m"

  1. is'sm:iated si each' channel'is c lated ithin' the time limir sumed in the
  • 1 h P dishfety' anal es. The RTS and respons6 times' are'inclsd

<n the;0perating V N'u .:Procedur P-3273 "Technica equirements--Su plementary hnical ~..... p,sigt,i. F:$'pecif ati6ns."?Any ch' es'to the RTS and SF re'sp a times :shall' be.ird., ^. 3 " ' acco anci sith' Sectio .59 of 10CFR50 and appro by the Plant Operations' h" ' Re ow Comeittee. credit was taken in the a yses for those channels wit ~ sponse times i cated as not a plicable, sponse time may be demonst by any series sequential, over apping total channel test measur ts . ".*"' " 'provided th ,ppch tests demonstrate total channel response ti s ' 'NM defined, nsoF Fesponse time veri ation may be demonstrated' tther:rg. 'in ace,'or. site, or offst est measurements, or (2)'u zing { '~(1)1 rep ement sensors with c f ied response time. Detec response times may be asured by the in si n line noise anal sis-res se time degradation method described in t estinghouse Topical epor, The Use of Process Noise Measurements To De ine Response Characteris s of Protection $4nsors in U.S. Plants,/ st 1983. p.wr T N.' NILLSTONE - UNIT 3 8 3/4 3-2 Amendment No. f'91 .m

___ _ _ _ _ _ _ _ m _ L I i j INSERT 'D' TO PARES R3/4 3 1 AND B3/4 3-2 } i The Engineered Safety Features Actuation System Nominal Trip Setpoints specified in Table 3.3-4 are the nominal values of which the bistables are set j .for each functional unit. The Allowable Values Nominal Trip Setpoints i the calibration tolerance) are considered the l.initi(ng Safety System Settings as identified in 10CFR50.36 and have been selected to mitigate the consequences of i accidents. - A Setpoint is considered to be consistent with the nominal value when i the measured "as left" Setpoint is within the administratively controlled (i) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip setpoint). Additionally. the Nominal-Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged. Measurement and Test Equipment accuracy is administrative 1y controlled by plant procedures and is included in the plant uncertainty calculations as defined in WCAP-10991. Operability determinations are based on the use of Measurement I and Test Equipment that conforms with the accuracy used in the plant uncertainty calculation. The Allowable Value specified-in Table 3.3-4 is the initial value for consideration of channel operability. If the process rack bistable setting is j measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nominal Trip Setpoint, then the channel is considereo to be operable. Additionally, an administrative 1y controlled limit for operability of a device is determined by device drift being less than the value required for the i surveillance interval. In the event the device exceeds the administratively i controlled limit, operability _ of the device may be evaluated by other device performance characteristics, e.g., comparison to historical device drift data, i calibration characteristics, response-characteristics, anc short-term drift characteristics. A device (RTD, relay, transmitter, process rack module, etc.), 4 whose "as found" value is in excess of tha calibration tolerance, but within the operability criteria (administrative 1y controlled limit), is considered operable but must be recalibrated such that the "as left" value is within the two sided (i) calibration tolerance. Plant procedures set administrative limits ("as left" i and "as found" criteria) to control the determination of operability by setting minimum standards based on the~ methodology in WCAP 10991 and the uncertainty - values included in the determination of tiie Nominal Trip Setpoint, and allow the i use of other device characteristics to evaluate operability. REPORTABLE EVENTS [ are identified when the minimum number of channels required to be operable are i not met. L h The methodology, as defined in WCAP-10991 to-derive the Nominal Trip Setpoints, is based upon combining all of the uncertainties in the channels. 1 Inherent in the ~ determination of the Nominal Trip Setpoints are the magnitudes l-of these channel uncertainties. Sensors and other instrumentation utilized in these channels should be capable of operating within _the allowances of these uncertainty magnitudes. Occasional drift in excess of the allowance may be determined to be acceptable based on the other device performance I l l 4 L ~.

I e INSERT D Cont'd. f characteristics.. Device drift in excess of the allowance that is more than occasional, may be indicative of more serious problems and would warrant further investigation. The above Bases do not apply to the two radiation monitors in the ESF Table (Item 30 and item 7E. For these radiation monitors the allowable values are essentially nominal v)alues.Due to the uncertainties involved ir radiological parameters, the methodologies of WCAP-10991 were not applied. Actual trip setpoints will be reestabitshed below the allowable value based on calibration accuracies and good practices. 4 4 4

j Qggket No. 50-423 B16624 ) h t Millstone Nuclear Power Station Unit No. 3 - Proposed Revision to Technical Specification Instrumentation Surveillances (PTSCR 3-30-97) Retvoed Paoes October 1997 4 -e' ,y re- -- w w .--+---r 7------- v- ,-r--~g* + - - -. - - ---wer- -~v-v-

ec.--,-w

-re-

U.S. Nuclear Regulatory Commission B16624%ttachment 3\\Page1 RETYPE OF PROPOSED REVISION Refer to the attached retype of the proposed revision to the Technical Specifications. The attached retype reflects the currently issued version of the Technical Specifications. Ponding Technical Specification revisions or Technical Specification revisions issued subsequent to this subraittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to issuance. Y

SAFETY LINITS MD LIMITIM SAFETY SYSTEM SETTIM5 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation Channel and Interlock Channel shall be OPERABLE. APPLICABillll: As showis for each channel in Table 3.3 1. ACTION: a. With a Reactor Trip System Instrumentation Channel or Intericek Channel Nominal Trip Setpoint inconsistent with the value shown in the Nominal Trip Setpoint column of Table 2.2-1, adjust the Setpoint consistent with the Nominal Trip Setpoint value, b. With a Reactor Trip System Instrumentation Channel or Interlock Channel found to be inoperable, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3~.3.1 until the channel is restored to OPERABLE status. MILLSTONE - UNIT 3 2-4 Amendment No. 0660

.= =. i TABLE 2.2-1

i g5 REACTOR TRIP SYSTEM INSTIMENTATION TRIP SETPOINTS F.

t i NOMIML 4 i E FUNCTIONAL UNIT TRIP SETPOINT ALLOMABLE VALUE m I 4 1. Manual Reactor Trip N.A. N.A. E Q 2. Power Range, Neutron Flux w j a. High Setpoint

1) Four Loops Operating 1095 of RTP**

s 109.6% of RTP** l 2

2) Three Loops Operating 80% of RTP**

5 80.6E of RTP** i b. Low Setpoint 25% of RTP** 1 25.65 of RTP** 3. Power Range, Neutron Flux, 5% of RTP** with 1 5.6E of RTP** with i High Positive Rate a time constant a time constant ~ 2 seconds 1 2 seconds

    • a 4.

Deleted l 1 j 5. Intermediate Range, 25% of RTP** 1 27.4% of RTP** Neutron Flux f i 6. Source Range, Neutron Flux 1 X IG** cps i 1.06 x 10+' cps i s k 7. Overtemperature AT s* F a. Four Loops Operating M

1) Channels I, II See Note 1 See Note 2

= i ?

2) Channels III, IV See Note 1 See Note 2 l

w E. w

    • RTP - RATED THERMAL POWER t

5 i. i I i

3 TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 85 s r-G NOMINAL FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE h b. Three Loops Operating

1) Channels I, II See Note 1 See Note 2 w
2) Channels III, IV See Note 1 See Note 2 F

8. Overpower AT (Four Loops Operating) See Note 3 See Note 4 9. Pressurizer Pressure-Low 1900 psia 1 1897.6 psia j

10. Pressurizer Pressure-High 2385 psia 1 2387.4 psia

[

11. Pressurizer Water Level-High 89% of instrument i 89.3% of instrument span span

{

12. Reactor Coolant Flow-Lcw 905 of loop 2 89.8% of loop design flow
  • design flow
  • i ao.
13. Steam Generator Water 18.1% of narrow 1 17.8% of narrow Level Low-Low range instrument range instrument f

span span M

14. General Warnirg Alarm N.A.

N.A. M

15. Low Shaft Speed - Reactor 92.4% of rated 1 92.2% of rated i

i Coolant Pumps speed speed ~ M E_ M

  • Minimum Measured Flow Per Loop - 1/4 of the RCS Flow Rate Limit as listed in Section 3.2.3.1.a (Four

? Loops Operating); 1/3 of the RCS Flow Rate Limit as listed in Section 3.2.3.2.a (Tnree Loops Operating) f i

TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRLNENTATION TRIP SET?GINTS NOMINAL FUNCTIOEL UNIT TRIP SETPOINT ALLOW 3LE_Y8LE G i w E

16. Turbine Trip a.

Low Fluid Oil Pressure 500 psig 1 450 psig E Q b. Turbine Stop Valve 1% open 1 1% open w Closure i

17. Safety Injection Input N.A.

N.A. from ESF

18. Reactor Trip System Interlocks a.

'Interisediate Range 1 x 10* amp 19.0 x 10-" amp i Neutron Flux, P-6 7 b. Low Power Reactor Trips Block, P-7 t

1) P-10 input (Note 5) 11% of RTP**

111.6% of RTP**

2) P-13 input 10% RTP** Turbine i 10.6% RTP** Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent g

c. Power Range Neutron g Flux, P-8 o.l

1) Four Loops Operating 37.5% of RTP**

5 38.1% of RTP**

2) Three Loops Operating 37.5% of RTP**

5 38.1% of RTP** N O E

    • RTP = RATED TliERMAL P0 lier I

E. 4

6 5 TABLE 2.2-1 (Continued) ~ i E - M REACTCR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E m \\ NOMINAL E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE G i w d. Power Range Neutron - 51% of RTP** 151.6% of RTP** Flux, P-9 e. Power Range Neutroe 95 of RTP** 1 8.4% of RTP** f Flux, P-10 (Note 6)

19. Reactor Trip Breakers M.A.

M.A. i

20. Automatic Trip and Interlock N.A.

N.A. Logic 1 m i f

21. Three Loop Operation N.A.

N.A Bypass Circuitry I 1 E i I ~ C i I i e g [

    • RTP - RATED THERMAL POWER E

l l l [ w, s . + - y -,+- - -

j i TABLE 2.2-1 (Continued) lz I r-g TABLE NOTATIONS (Continued) k NOTE 1: (Continued) I T' i 587.l*F INominal T,,9 at RATED THERMAL POWER); I ] K 0.G01311/pst 3 P Pressurizer pressure, psia; P' 2250 psia (Nominal RCS operating pressure); Laplace transform operator, s '; S i and f (AI) is a function of the indicated difference between top and bottom detectors of the 3 4 power-range neutron ion chanhers; with gains to be selected based on measured instrument response during plant startup tests such that: Y5 (1) - gb between -26% and + 3%, f,(AI) - 0, where q For qk top and bottom halves of the core respectively,t and qb+are percent RATED THERMAL POWER in th aruf q 9b is total THERMAL POWER in t percent of RATED THERMAL POWER; l (2) For each percent that the magnitude of q - qu exceeds -26%, the AT Trip Setpoint shall be automaticallyreducedby3.55%ofitsvaiveatRATEDTHERMALPOWER;and (3) Foreachpercentthatthemagnitudeofqkue exceeds +3%, the AT Trip Setpoint shall be automaticaliy reduced by 1.98% of its va RATED THERMAL POWER. I R. NOTE 2: The maximum channel as left trip setpoint shall not exceed its computed trip setpoiht by more than the 5 following: 5 (1) 0.4% AT span for the AT channel (2) 0.4% AT span for the T channel g (3) 0.4%ATspanforthepre$surizerpressurechannel av e (4) 0.8% AT span for the f(AI) channel .? D? 1 i - - ~

l h h TABLE 2.2-1 (Continued) E .m TABLE NOTATIONS (Continued) .E [ NOTE 3: (Continued) l 0.00180/*F for T > T* and K6 - O for T 1 T", K 6 As defined in Note 1, T Indicated T,,9 at RATED THERMAL POWER (Calibration temperature for AT l T" instrumentation, 5 587.1*F), As defined in Note 1, and S f %I) 0 for all AI. 2 b NOTE 4: The maximum channel as left trip setpoint shall not exceed its computed trip setpoint by more than 0.4% AT span for the AT channel and 0.4% AT span for the T channel. avg NOTE 5: Setpoint is for increasing power. NOTE 6: Setpoint is for decreasing power. a E ?

} e i j 2.2 LIMITING 1AFETY SYSTEN SETTIMS l BASES- [ f.2.1 RFACTOR TRIP SYSTEN INSTRLSENTATION SETPOINTS l-The Nominal Trip Setpoints specified in Table 2.2-1 are the nominal i 3 values at which the reactor trips are set for each functional unit. The Allowable Values Nominal Trip Setpoints i the calibration tolerance are i considered the Li iting Safety System Settings as identified in 10CF 50.36 and L have been selected to ensure that the core and Reactor Coolant System are i prevented from exceeding their safety limits during normal operation and-design basis anticipated operational occurrences and to assist the Engineered 1. -Safety Features Actuation System in mitigating the consequences of accidents. The Setpoint for a Reactor Trip stem or interlock function is considered to be consistent with the nominal y ue when the measured "as-left" Setpoint-is l within the administrative 1y: controlled (i) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint). Additionally, the Nominal Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged. Neasurement and Test Equipment accuracy is administrative 1y controlled by_ j plant procedures and is included in the plant uncertainty calculations as L . defined in WCAP-10991. Operability detereinations are based on the use of l Measurement and Test Equipment that-confonra with the accuracy used in the plant uncertainty calculation. The Allowable Value specified in Table 2.2-1 is the initial value for consideration of channel operability.- If the process rack bistable setting is measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nwinal Trip Setpoint, then the j channel is considered to be operable. 1 Additionally, an administratively controlled limit for operability of a l device is determined by device drift being less than the value required for i the surveillance interval. In the event the device exceeds the administrative 1y controlled limit, operability of the device may be evaluated by other device performance characteristics, e.g., comparison to historical device drift data, calibration characteristics, response characteristics, and. short-term drift characteristics. A device (RTD. relay, transmitter, process rack module - etc.), whose "as found" value is in excess of the calibration tolerance, but within the operability criteria (administrative 1y controlled limit),:is considered o>erable but must be recalibrated such that the "as left" value is within-tie two sided (i) calibration tolerance. Plant procedures set administrative limits ("as left" and "as found" criteria) to L control the determination of operability by setting minir:s standards based on the methodology in-WCAP-10991 and the uncertainty values included in the 4 i determination of the Nominal Trip Setpoint,-and allow the use of other device j characteristics to evaluate operability. - REPORTABLE EVENTS are identified j when the minimum number of channels required to be operable are not met. i L I -NILL $Y3NE - UNIT-3 32-3 Amendment No. nu aa- ......, - - -. - -.. ~. -. -, -. _. - _.. .... _ _. -., -.. -,. -.,, ~. -.. - -. - - - - -

1 2.2 LIMITIM 1AFETY SYSTEM SETTIMS l BASES l f t REACTOR TRIP SYSTEN INSTRUMENTATION SETPOINTS (Continued) I i The methodology, as defined in IdCAP 10991 to derive the Nominal Trip Setpoints, is based upon combining all of the uncertainties in the channels. Inherent in the determination of the Nominal Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels should be capable of operating within the allowances of these uncertainty magnitudes. Occasional drift in excess of the allowance may be determined.to h acceptable based on the other device performance characteristics. Device drift in excess of the allowance that is more than F occasional, may be indicative of more serious problems and would warrant further investigation. l The various reactor trip circuits automatically open the reactor trip i breakers whenever a condition monitored by the Reactor Trip System reaches a. preset or calculated level. -In addition to the redundant channels and trains, i the design approach provides Reactor Trip System functional diversity. The 4 functional capability at the specified trip setting is required for those anticipatory or diverse reactor trips for which no direct credit was assumed in i the safety analysis to enhance the overall reliability of the Reactor Trip System. The teactor Trip System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System. Manual Reactor Trin The Reactor Trip System includes manual Reactor t.atp capability. I -Power Ranoe. Neutron Flux In eat.h of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint tria provides protection during subcritical and low power operations to mitigate tle consequences of-a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of-a reactivity excursion from all power levels. The.High Setpoint trip is reduced during three loop operation i to a value consistent with the safety analysis.- The Low Setpoint trip may be manually blocked above P 10 (a power level of approximately 10% of RATED THER%L POWER) and is automatically reinstated below the P-10 Setpoint. i .NILLSTONE - UNIT 3 W2-4 Amendment No. JJp,

ow

I.2.LIMITIMR.1AFETY. SYSTEM SETTIM5 BASES REACTORTRIPSYSTEMINSTRUMENTATIONSETPOINTS(Continued) Power Ranae. Neutron Flux. Hiah Positive Rate The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to e,isure that the criteria are met for all rod ejection accidents, t t 1 gIgl5 TONE-UNIT 3 B 2 - 4a Amendment No. JJJ.

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LINITING CONDITION FOR OPERATION i 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3 3 shall be OPERABLE. APPLICABIL111: As shown in Table 3.3-3. ACTION: a. With an ESFAS Instrumentation Channel or Interlock Channel Nominal Trip Setpoint inconsistent with the value shown in the Nominal Trt) Setpoint column of Table 3.3 4, adjust the Setpoint consistent witi ~i the Nominal Trip Setpoint value. b. With an ESFAS Instrumentation Channel or Interlock Channel found to be inoperable, declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status. s i 4 MILLSTONE - UNIT 3 3/4 3-15 Amendment No. JJ. 0552

5 TABLE.3.3-4

~ 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEN ICSTIMENTATION TRIP SETPOINT5 a m NONINAL e q FUNCTIONAL LWIT TRIP SETPOINT ALLOWABLE VALUE 1. Safety Injection (Reactor Trip, Feedwater Isalation. Control Building Isolas mn (Manual Generators, and) Service Water) Initiation Only, Start Diesel a. Manual Initiation N.A. N.A. b. Automatic Actuation Logic N.A. M.A. c. Containment Pressure--High 1 17.7 psia $ 17.9 psia d. Pressurizer Pressure--Low k 1 Channels I and II 1892 psia 1 1889.6 psia 2 Channel III and IV 1892 psia .,1889.6 psia e. Steam Line Pressure--Low 658.6 psig* 2 654.7 psig* 2. Containment Spray (CDA) a. .ianual Initiation N.A. M.A. b. Automatic Actuation Logic N.A. M.A. and Actuation Relays c. Containment Pressure--High-3 22.7 psia i 22.9 psia a. g 3. Containment Isolation [ a. Phase "A" Isolation o;

1) Manual Initiation N.A.

N.A. ? U

TABLE 3.3-4 ~ l ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTERENTATION TRIP SETPOINTS NOMINAL-3 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE l 3. Containment Isolation (Continued) e

2) Automatic Actuation Logic N.A.

N.A and Actuation Relays w

3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

t

b. Phase "B" Isolation
1) Nanual Initiation N.A.

N.A. l

2) Automatic Actuation N.A.

N.A. Logic and Actuation l g Relays [ t s,

3) Containment Pressure--

22.7 psia 1 22.9 psia w 4, High-3

c. Purge Isolation i 1 R/h 1 1 R/h 4.

Steam Line Isolation

a. Manual Initiation N.A.

N.A.

b. Autmatic Actuation Logic M.A.

N.A. g and Actuation Relays f

c. Containment Pressure--High-2 17.7 psia i 17.9 psia 5
d. Steam Line Pressure--Low 658.6 psig*

1 654.7 psig* I f

e. Steam Line Pressure -

100 psi /s** 1 103.9 psi /r** Negative Rate--High ,? M? ) -w

e I M LE 3.3-4 $*P EPSINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUIENATION TRIP SETPOINS ",j NOMINAL' g FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE m 5. Turbine Trip.and Feedwater-Isol:..lon g 4 a. Automatic Actuation Logic M.A. N.A. Actuation Relays w b. Steam Generator Water 80.5% of narrow 1 80.8% of narrow Level--High-High (P-14). range instrument range instrument span.- span. c. Safety Injection Actuation See.Ites' 1. above for all Safety Injection Trip Logic-Setpoints. and Allowable Valves. d. T,,, Low Coincident with ,2 Reactor Trip (P-4)

1) Four Loops Operating -

564*F 1 563.6*F

2) Three Loops Operating 564*F 1 563.6*F 6.

Auxiliary Feedwater a. Manual Initiation N.A. N.A. a, b. Automatic Actuation Log c N.A. N.A. 3 and Actuation Relays E c. Steam Generator Water l Level--Low-Low '+

1) Start Notor-Driven 18.1% of 1 17.8% of narrow g

Pumps narrow range range instrument span. instrument span. D.

TABLE 3.3-4 (Cor tino!d) 3.3.. ENGINEERED SAFETY FEATURES ACTUATION SYST MI ISTRINENTATION TRIP SETPOINTS r-U NOMINAL I lL g FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE m 6. Auxiliary Feedwater (Continued) E

2) Start Turbine-18.1% of 2 17.8% of narrow

-Z Driven Pumps narrotr range' range instrument. w instrument span. span. d. Safety Injection See. Item 1.'above for all Safety Injection Trip Setpoints ,and Allowable Values. e. Loss-of-Offsite Power 2800V 2 2720V Start Motor-Driven Pumps f. Containment De ressurization' See Item 2. above for ali CDA Trip Setpoir.ts and Allowable Values. Actuation (CDA Start wi Motor-Driven s "4 7. Control Building Isolation e a. Manual Actuation N.A' N.A. b. Manual Safety' Injection M.A N.A. Actuatien c. Automatic Actuation N.A. N.A. k Logic-and Actuation-Relays = k s d. Containment 17.7 psia '< 17.9 psia Pressure--High 1 F w Inlet Ventilation -<1.5 x 10~5mi/cc -<1.5 x 10~'pci/cc e. Control Building ? Radiation D.

TABLE 3.3-4 (Continued) [5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS " P, v Y NOMINAL E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE h 8. Loss of Power 4 a. 4 kV Bus Undervoltage 2800 > 2720 volts (Loss of Voltage) volts with with a 5 2 w a i 2 second second time time delay. delay. b. 4 kV Bus Undervoltage 3730 volts 2 3706 volts (Grid Degraded Voltage) with a < 8 with a < 8 second Time second Time delay with ESF delay with ESF actuation or actuation or < 300 second < 300 second Time delay Time delay w1 without ESF without ESF actuation. actuation. b 9. Engineered Safety Features Actuation System Interlocks a. Pressurizer Pressure, P-11 1999.7 psia 5 2002.1 psia b. Low-Low T,yg, P-12 553*F 1 552.6*F k c. Reactor Trip, P-4 N.A. N.A.

(
10. Emergency Generator Load N.A.

N.A. g Sequencer .E .__.___m._____

3/4.3 INSTRIMENTATION BASES 3/4.3.1 and 3/0.3.2 - RMCTOR " RIP SYS"KN INSTRUMENTATION and ENGINEEEED SAFETY FEATURES ACTUA" ION SYS' 'EM INS"RUMENTA"..ON The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: 1 the associated action and/or Reactor ' trip will be initiated when the par (am)eter monitored by each channel or combination thereof reaches its setpoint (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse paremeters. The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.- The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The Engineered Safety Features Actuation System Nominal Trip Setpoints specified in Table 3.3-4 are the nominal values of which the bistables are set for each functional unit. The Allowable Values (Nominal Trip Setpoints i the calibration toleran:e) are considered the Limiting Safety System Settings as identified in 10CFR50.36 and have been selected to mitigate the consequences of accidents. A Setpoint is considered to be consistent with the nominal value when the measured "as left" Setpoint. is within the administratively controlled (1) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint). Additionally, the Nominal Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged. Measurement and Test Equipment accuracy is administratively controlled by plant procedures and is included in the plant uncertainty calculations as defined in WCAP-10991. Operability determinations are based on the use of Measurement and Test Equipment that conforms with the accuracy used in the plant uncertainty calculation. The Allowable-Value specified in Table 3.3-4 is the initial. value for consideration of channel operability. If the process rack bistable setting is-measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nominal Trip Setpoint, then the channel is considered to be operable. MILLSTONE - UNIT 3 5 3/4 3-1 Amendment No. 0655-

I INSTEMENTATION SASE5-4 3/4.3.1 and 3/4.3.2 REACTOR TRIP S"S"EM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUA"10N SYS'EM INSTRUMEN"A"10N (Continued) Additionally, an administrative 1y controlled limit for operability of a device is determined by device drift being less than the value required for the. surveillance interval.- In the-event the. device exceeds the administrative 1y controlled limit, operability of the - device may -be evaluated by other device performance. characteristics, e.g., comparison to historical device drift data, calibration characteristics, response characteristics, and short-term drift characteristics. A device (RTD, relay, transmitter, process rack module, etc.), whose "as found" value is in excess of the calibration tolerance, but within the operability criteria (administratively ;ontrolled limit), is considered-operable; but must be recalibrated such that the as left" value is within the two sided (i) calibration tolerance. Plant procedures set administrative limits ("as left" and "as found" criteria) to control the determination of operability by setting minimum standards based on the methodology in WCAP-10991 and the. uncertainty values included in;the determination of the Nominal Trip Setpoint, and allow the use of other - device characteristics -to evaluate operability. REPORTABLE EVENTS are identified when the minimum number of channels required to be operable are not met. - The methodology, as defined in WCAP-10991 to derive the Nominal Trip Setpoints, is based upon combining _ all of the uncertainties in the channels.- Inherent in the determination of the Nominal Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels should be capable of operating within the allowances 4f these uncertainty. - magnitudes. Occasional drift in excess of the allowance may be determined to be . acceptable based on the other device performance characteristics. Device drift in excess _ of the allowance that is more -than occasional, may be. indicative of more serious problems and would warrant further investigation. The above Bases do not apply to the two radiation monitors _ in the ESF Table (Item 3C and Item 7E). For these radiation monitors the allowable values are essentially nominal values. Due to the uncertainties involved in radiological parameters - the methodologies of WCAP-10991 were not applied. Actual trip V setpoints will be reestablished below the allowable 'value. based on-calibration accuracies and good practices, e NILL 5 TONE - UNIT 3 8 3/4 3-2 Amendment No. 7, U. ~ 0666

1 i l l Docket No. 50-423 B16624 e 1 1 Mi;: stone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Instrumentation Surveillances (PTSOR 3-30-97) Backaround and Safety Assessment October 1997

) U.S. Nuclear Regul: tory Commiselon B16624\\ Attachment 4\\Page1 Backaround Northeast Nuclear Energy Company (NNECO) is proposing to modify the Technical Specifications to reflect a two column format and to revise the allowable values based l on the Westinghouse analyses of historical instrument data contained in Westinghouse } Report Nos. WCAP-14354 (May 1995) and WCAP-10991 Rev. 5 (August 1997) Safety Assessment Currently Millstone Unit No. 3 Technical Specifications utilize the concept of an allowable value in combination with a total allowance for drift and other inaccuracies as a means of triggering an operability determination for a channel. This does not accurately satisfy the plant procedures and practices for determining equipment operability. Based on Millstme Unit No. 3 equipment performance, Westinghouse has developed an enhanced method of determining equipment operability without the restrictions of the 5 column technical specification. Based on the evaluation of equipment drift data for Millstone Unit No. 3 over the past 4 cycles, the need to reduce Technical Specification changes and the objective of the future increase of the surveillance intervals, from the current 18 month interval to a nominal 24 month interval, for RTS and ESFAS functions, NNECO proposes to use the allowable value as a means for establishing the initial acceptance criteria for the process rack bistable operability criteria in the Millstone Unit 3 Technical Specifications. Additional operability criteria for sensors and racks would be included in the Millstone Unit No. 3 administrative procedures (surveillance procedures), controlled by NNECO and be based on the following: The identification of trigger points based on actual channel instrumentation drift. Ability to recalibrate the hardware with acceptable calibration tolerances. The evaluation of equipment performance over time utilizing actual plant drift performance data as part of a monitoring program. To accomplish the Operability Criteria, NNECO has collected RTS/ESFAS function hardware data over the past four (4) operating cycles of Millstone Unit No. 3, in accordance with Generic Letter 91-04 Enclosure 2. This data was further analyzed by Westinghouse, utilizing enhanced methods of determining equipment operability. __________j

U.S. Nuctar Regulatory Commission B16624\\ Attachment 4\\Page 2 Analysis has demonstrated that adequate' margin for nominal trip setpoints exist and safety analysis limits are presened in all RTS/ESFAS functions. Based on thin analysis, NNECO is proposing that the Mi'btone Unit No. 3 Technical Specification reflect nominal trip setpoints and allowable values for RTS/ESFAS Table 2.2-1 and Table 3,3-4. In addition, the proposed change decreases the reactor trip setpoint for the reactor coolant pump (RCP) low shaft speed (underspeed trip setpoint) from 95.8 percent to 92.4 percent of rated speed. Changing the RCP low shaft speed trip setpoint will not change the probability of occurrence or consequence.s of a loss of forced RC flow event. The existing accident analysis (Millstone Unit No. 3 FSAR section 15.3.2) of the complete loss of forced reactor coolant flow remains valid for the proposed change. The Safety Analysis Limit, 92% of Rated Normal Speed (RNS), has not been changed. The Nominal Trip Setpoint was reduced from 95.8% RNS to 92.4% RNS to decrease the incidence of inadvertent . tripping caused by transient power line frequency variations. The Allowable Value has been chang. d to bring it closer to the Nominal Trip Setpoint. The reduction in the number of challenges to the Reactor Coolant Pump Underspeed Trip Channels caused by transient power line frequency variations increases the reliability of the trip channels. - Although the Nominal Trip Setpoint is now closer to the Safety Analysis .. Limit, the Margin between the Total Allowance and the Channel Statistical Allowance remains unchanged. Therefore, the change to the RCP low shaft speed trip setpoint does not increase the probability or consequences of any previously analyzed malfunction. NNECO is also requesting the Bases Section,2.2.1, for RTS and Bases Section 3/4.3 for ESFAS be changed to accommodate the operability requirements for the RTS and ESFAS respectively. Based on the above, the proposed Technical Specification change is safe. 4

1 Docket No. 50-423 B16624 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Instrumentation Surveillances (PTSCR 3-30-97) Sionificant Hazards Consideration and Environmental Considerations October 1997

U.S.- Nuclear Regul: tory Commission -

- B18624%ttachment 5\\Page 1 - Sionificant Hazards Consideration NNECO has reviewed the proposed revision in ~accordance with 10CFR50.92 and has concluded that the revision does not involve a signifm' ant hazards consideration (SHC). - The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve a SHC because the revision would not: 1. Involve a significant increase in the probabilit) or consequence of an accident previously evaluated. The proposed changes to Tables 2.2-1 and 3.3-4 involve changes from a five 1 column format to a two column format.- The RTS trip setpoints and ESFAS trip - setpoints remain unchanged with the _ exception of the RCP low shaft speed trip setpoint discussed below. _ Detailed operability criteria will be moved to surveillance procedures and--analysis has demonstrated that an. adequate . margin for normal trip setpoints exist and safety analysis limits are preserved in all RTS/ESFAS functions. Changing the RCP low shaft speed trip setpoint will not change the probability of occurrence of the event. The existing accident analysis (Millstone Unit No. 3 FSAR section 15.3.2) of the complete loss of forced reactor coolant flow remains valid for the proposed change. Therefore, the change to the RCP low shaft speed trip setpoint does not increase the probability or consequences of any previously analyzed accident. In addition, the proposed changes to Tables 2.2-1 and 3.3-4 do not alter the-intent or method _ by which the surveillances are conducted. Therefore, the scope of evaluation performed gives reasonable assurance that there will not be an adverse Impact on the consequences or the probability of any previously-analyzed accident. - Therefore, the proposed revision does not involve a significant increase in-the probability or consequence of an accident previously evaluated. 2. . Create'the possibility of a' new or different kind of accident from any accident . previously evaluated. The : existing ~ design basis adequately covers the plant response with the proposed change to the RCP low shaft speed trip setpoint. The change does-not introduce new failure modes The proposed changes to Tables 2.2-1 and 3.3-4 do not modify the design or operetion of any plant system. The proposed changes do not alter the intent or

4 c: U.S. Nuclear Regul: tory Commission B16624%ttachment 5\\Page 2 J method by which the surveillances are conducted, other than adjusting tho' j ' allowable values to reflect historical instrument performance data. Therefore, the proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Involve a significant reduction in a margin of safety. l The proposed changes to Tables 2.2-1 and 3.3-4 modify the existing five column format to a two column format to show-the RTS and ESFAS nominal trip setpoints~ and the-process' rack bistable allowable values for individual ~ functions. Detailed operability criteria will be moved to the surveillance procedures. With the exception of the low shaft speed trip discussed below, the RTS and ESFAS setpoints remain unchanged and analysis has demonstrated i-that an adequate margin for normal trip setpoints exist and safety analysis limits _ are preserved in all RTS/ESFAS functions. 4 l-Since the safety limits of the design are still met, the proposed change to _the RCP low shaft speed trip setpoint does not reduce the margin of safety. o Therefore, the proposed revision does not involve a significant reduction in a margin of safety. [ In conclusion, based on the information providod, it is determined that the proposed revision does not involve an SHC. f Environmental Considerations NNECO _ has reviewed the proposed license amendment against 'the criteria. of j 10CFR51.22 for environmental considerations. The proposed revision does not involve a SHC, does not significantly increase the type and amounts of effluents that may be [ released offsite,' nor significantly increase individual or. cumulative-occupational j radiation exposures. Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in-10CFR51.22(c)(9) for categorical exclusion. [ from the requirements for environmental review. I i

<f Docket No. 50-423 B16624 OCFR2 790 MMERIAL l Millstone Nuclear Powar Station Unit No. 3 i. Proposed Revision to Technical Specification Instrumentationjyryelliances (PTSCR 3-30-911 Westinohouse Sucoortino informatio_ n I p } i 2 i-4 1-October 1997 I; 4 3

q Proprietary Information Notice - h l. Transmitted herewith are proprietary and/or non-proprietary versions of documents fumished to the NRC in connection with requests for generic ardor plant specific review and approval. In order to conform to the requirements of 10 CFR 2.790 of the Commission's regulations conceming the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the p'oprietary information has been deleted in 'he non-proprietary versions, only the brackets remain (the information that was coritained within the brackets in the proprietary ve~Jons having been deleted).. The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) contained within parentheses located as a superscript immediately following the brackets enclosing each item of infctmation being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the 3

affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).

i l

a b ' e: Copyright Notice The reports transnitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its intemal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the extent such infamation has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its intemal use which are necessary in order to have one copy availaHe for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public L document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this t purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary, i I ---)}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20217J061&oldid=3962219"