ML20101F284

Proposed Tech Specs Re RTD Bypass Sys Elimination

ML20101F284
Person / Time
Site:	Seabrook
Issue date:	06/19/1992
From:	PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
Shared Package
ML20101F241	List: ML20101F238 ML20101F255 ML20101F284
References
NUDOCS 9206240360
Download: ML20101F284 (31)

Text

7 -. . .

. )

II. tiarkup of Pr onogyd Channes See attached markup of proposed changes to Technical specifications, t

5 9206240360 920619

.PDR ADOCK 05000443' P PDR

~

m IA8tE 2.2-1

,. g REACTOR'IRIP SYSTEN INSTRUpKNIATION IRIP SEIPOINIS E

o

$[MSOR l , TOTAL ERROR l .c IUNCTIONAL UNIT Att0WANCE (TA) Z (5) TRIP $EIPOINT All0WABLE VA8 DE h

~

1. Manual Reactor Trip M.A. N.A. M.A. N.A. M.A.

) 2. Power Range, Neutron Flux 1

a. High Setpoint 7. 5 4.56 0 <l11.1% of RIPa 1109% of RIP *

, b. Low Setpoint 8.3 4.56 0 <25% of RIP * <21.1% of P. ipa

3. Power Range, Neutron Flux, 1. 6 0. 5 0 <S1 of RIP

• with <6.3% of RIP" with High Positive Rate i time constant i time constant 12 seccads )2 seconds y 4. Power Range, Neutron Flux, 1.6 0.5 0 $5% of W * @ ;

• High Negative Rate <6.3% of RIP

• with j

a time m ut*' a time constant

>2 secv % >2 seconds I 5. Intermediate Range, 17.0 8.41 0 Neutron Flux -

123% of n's $31.1% of RIP *

6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps 4

<l.6 x 105 cps 3.5

7. Overtemperature AT 6.5

( ** See hte 1 See Note 2 l 8. Overpower al 4.4 'b

,, 3 See Note 4

9. Pressurizer Pressure - tow 3.12 0.86 0.99 h34Ygisig >1.931 psig

10. Pressurizer Pressure - Higts 3.12 1.00 0.99 12385 psig $2.198 psig

" RIP = RAllD IHERMAL POW (R I- b I OS I

• • Ihe sensor error for I is fl/% and time sensor error for Pressuriter Pressure 1s [4% "As me.asured" sensor errors may t>e used in lieu of eittier or lesitti of t tiese va l ue s , wis s a. In tisces maest t*e sessament t o eleter-4 mine the overtemperature al total (learistel valtie f or *>_

-, r ai _____ _

~

.~. ..

m IABLE 2.2-1 (Continued) h REACTOR TRIP 1 .IEN IN5fRtMENTATION TRIP SE1POINIS

" SENSOR IUTAL [RROR j

[ IUNCIIONAL UNIT ALLOWANCE (TA) Z (5) TRIP SEIPOINT At10WA81[ VAtUE l h II. Pressurizer Water Level - High 8.0 4.20 0.84 192% of instrument 193.15% of instr nt span spanX Erf4)lofloop 167.5 E g

12. Reactor Coolant flew - Low 2.5 , 0.6 390% of Iwp l.9 design flow" design flow *

13. Steam Generator Water 14.0 12.53 0.55 >l4.0% of narrow >l2.6% of narrow Level Low - Low range instrtment fange instrument span span

'14. Undervoltage - Reactor 15.0 1.39 0 310.200 volts >9,822 volts Coolant Pumps y 15. Underfrequency - Reactor 2. 9 0 0 355.5 Hz 355.3 Hz

= Coolant Pumps

16. Turbine Trip I

a. Low Fluid Oil Pressure N.A. N.A. N.A. 3500 psig 1450 psig

b. Turbine Stop Valve N.A. M.A. N.A. 31% open 31% open Closure

17. Safety Injection Input N.A. N.A. N.A. M.A. N.A.

from ESF

• ioop design flow = 95,700 gpa b - . . . _ r

. . ~..

IABLE 2.'-1-(Continued) '

M TASTE NOIAIIONS -

3 y NOIE 1: :OVERIEMPERAluRE AI-7 (1) ( ' (

gg ((1:I 1 +. 125) 5) (1 + s25) < gg 0 ga~K2 (1

• u s 5) II (1 * .5) ~ I* }

• K 2 (P - P') - i:(al)1 G

r Where: AT = MeasuredAlbyRfDlMif:!dl Instrumentation;

{*g- = lead-lag compensator on sneasured A1; I

=

u s , r2 Time constants utilized iss lead-lag compensator for AI, e, > 8 s, 12 $ 3 s; I

g, g

Lag compensator on measured ai; m

• r3 Time constants utilized in the lag compensator for af, 3 = 0 s; e

= Indicated AI at RATED THERMAL POW [R; AT, K = 1.0995; K2 = 0.0112/*F; I

g ,'*3 =

The function generated by the lead-lag compensator for i dynamic compensation; s., as =

Time constants utilized in the lead-lag compensator for I e , -> 33 5, as 1 4 s;. d'9, T = Average temperature, "F; 1

=

g, 3 Lag compensator on measured I,  ;

r. =

1ime constasit utilited i t the measured I lag compensator, s,= 0 s;

~.~ i IA8tE'2.2-1 (Continued) h

= TA8tE NOIATIONS

=

8 s

NOTE 1: (Continued) '

e T* - 588.5'F (Nominal T 859 at RATED THERMAL POWER);

x Q K3 = 0.000519/psig; e

P = ' Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure);

P S = laplace transform operator, s ';

and f (AI) is a function of the indicated dif ference between top and bottom detectors of the power range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests so that:

(1) For qg qbbetueen - 35% and + 8%, f (al) = 0, where q and g are percent RAIED IH[RMAL POWER in the top and bottom halves of the core respectively, and q i t * % s total IH[IW4mL POWER in percent of RATED THERNAL POWER;-

(7) For each percent that the magnitude of q q, exceeds - 35%, the ai Trip Setpoint shall be automatically reduced 'by 1.09% of its value at RATED THERMAL POWER; and (3) for each percent that the magnitude of q g exceeds

• 8%, the AT Trip 5etpoint shall be automatically reduced by 1.00% of its value at RATED IHERMAL POWER.

2.sz NOIE 2:

The of ATchannel's span. maximum Trip 5etpoint shall not exceed its computed Trip setpoint by more than[ _

.o . .,

TABLE 2.2-l'(Continued) h

. TABLE NOIAIIONS (Continued) 8 e

NOTE 3: (Continued) . o.oo/386/*F E-

-4 Ke - =l 0.""!?"/*j fer I > 1" and K = 0 for I i 1",

w =

T As defined in Note 1, I" =

Indicated I,, at RATED THENt4L POWER (Calibration temperature for al instrementation '$ 585.5'F),

S = As defined in Note 1, and fa(AI) =

0 for all AI.

7 NOTE 4:

5 channel's of AT span. maximun Trip Setpoint shall not exceed its computed Trip setpoint by more tsun I .o r, e

-M' -w C_. --

a '. -/ POWER O!STRIBUTION-LIMITS 3/4.2,5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION

3. 2. 3 The following DNB-related parameters shall be maintained within the the following limits:-

a. Reactor Coolant System T,yg, 1 594.3'F ,

b. Pressurizer Pressure, 1 2205 psig*

c. Reactor Coolant System Flow, 1 001,000 gpm**

APPLICABILITY: MODE 1. 3 D /000 ACTION:

With any of the above parameters exceeding its limit, restore-the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE' REQUIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The RCS. flow-rate indicators shall be subjected to CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by a precision heat balance measurement to be within its limit prior to operation above of RATED. THERMAL POWEk after each fuel loading. .The provisions of Specificatio 4.0.4 are not

-applicable for entry into MODE 1. ,

f$$*

• Limit not applicable.during either a THERMAL POWER ramp in excess of 5% of

- RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%

of RATED THERMAL POWER.

• • Includes a 4:45 flow measurement uncertainty, h fQ.

SEABROOK - UNIT -1 3/4 2-10 L

IABLE 4.3-1 REACIOR IRIP SYSIIM INSIRUN[NIAll0N SURVElltANC[ RindIRLMtNIS 8

• IRIP

• ANALOG ACIUAIING MOOL5 IOtt CHANII[L D(VICE WHICH E CHANNEL CHANNEL OPfRAll0NAL DhRAIIONAL ACIUAIION SURVilt1 ANLL U fuMCIIONAL UNil CMCK CALIBRAII004 IEST ItST 10 Git itSI 15 RigulRin v -

1. Manual Reactor Trip N.A. M.A. N.A. R(13) N.A. 1, 2, 3a,4*,

Power Range, Neutron flux 2.

a. High Setpoint '

5 D(2, 4), Q(16) N.A. N.A. 1, 2 M(3, 4),

Q(4, 6),

R(4, 5)

b. Low Setpoint 5 R(4) $/U(1) N.A. N.A. 1"" , 2

3. Power Range, Neutron flux, M.A. R(4) Q(16) N.A. M.A. 1, 2 y High Positive Rate s

y 4. Power Range, Neutron flux, M.A. R(4) Q(16) N.A. N.A. 1, 2

.o High Negative Rate

5. Intermediate Range, 5 R(4, 5) 5/U(1) N.A. M.A. 1"= , 2 Neutron fIum

6. Source Range, Neutron flux 5 R(4, 5) 5/U(!),Q(9,16) N.A. N.A. 2 " , 3 , 4 , *,

7. Overtemperature AI 5 Q(16) N.A. N.A. I, 2

^

8. Overpower al 5 R Q(16) M.A. N.A. 1, 2

9. Pressurizer Pressure--tow 5 R Q(16,lI) N.A. N.A. I

10. Pressurizer Pressure--liigh 5 R Q(16,ll) N.A. N.A. 1, 2

11. Pressuriser Water level--liiejh 5 R Q(16) NA N.A. I 1/. Iteottor Coularit ilesw--low '

, R Q( Ile) NA N A. I

.g faltE 4.3 1 (continued)

TABLE NOTATIONS (Continued)

!n2} ": ", tr: "O L;;:: ':::: :1 r:t:.

(13) De TRIP ACTUATING OEVICE OPERATIONAL. TEST ohall independen tne OPERABILITY of the undervoltage and snunt trio circuits for the Reactor Trio Function. wasua Bypass Breaker trip circuit (s).The test shall also verify the OPERABILITY o (14) local manual shunt trip prior to placing breaker in service.

(15) Autematic undervoltage trip.

(16) Each BASIS.

channel shall be tested at least every 92 days on a $TAGGEREO TEST (17) These channels aise orovide inputs to ESFAS.

Comply with the accli: : e W) DES and surveillance frecuencies of Specification 4.3.2.1 for any cor-tion of the channel required to be OPERA 8LE by Soecification 3.3.2.

/

SEABROOK UNIT 1 3/4 3-13

,,m 9,ww--=-*N

l . ..-

e, .'

POWER DISTRIBUTION LIMITS

/%

( f b,. ;

BASES 3/4.2.5 DNB PARAMETERS The limits on the ONB-related parameter; assure that each of the parameters is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. Operating procedures include allowances for measurement and indication uncertainty so that the limits of 594.3'F for T,yg and 2205 ps or pressurizer are not exceeded.

The measurement error of e-t% for RCS total flow rate is based upon per-forming a precision heat balance and using the result to normalize the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a noncon-servative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is applied. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending vari-ous plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measuremont or the venturi shall be cleaned to eliminate the fouling, The 12-hour periodic surveillance of these parameters through instrument [r.A readout is sufficient to ensure that the parameters are restored within their '

limits following load changes and other expected transient operation.

The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could 1 cad to operation outside the specified limit, d

k.3/

SEABROOK - UNIT 1 B 3/4 2-4

,,-n _ . _

~ 7 -- -_

_ 37

Ill. Retype of Proposed Chanyn See attached retype of proposed changes to Technical Specifications. The attached tetype reflects the currently issued version et Technical Specifications. Pending Technical Specification Changes or Technical Specification changes issued subsequent to this submit t al are not r ef lec t ed in the enclosed retype. The enclosed retype should be chec ked for continuity with Technical specifications prior to i s s.u a n c e .

Revision bars are provided in the right hand margin to indicate a revision to the text. No revision bars are utilized when the page is changed solely to accommodate the shifting of text due to additions or deletions.

Y 6

2

, ~.-

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M- TABLE 2.2-1 g

REAC10R TRIP SYSTEN INSTRUNENTATION 1 RIP SEIPOINTS 8

u g SENSOR i , TOTAL .

ERROR FUNCTIONAL UNII. Att0WANCE (TA) 1 (S) TRIP SETPOINT Att0WABtI VAtUE-

1. . Manual Reactor Trip N.A. N.A. M.A. 'N.A. M.A.

2. Power Range, Neutron Flux

a. High Setpoint 7.5 4.56 0 s109% of RTP* ' $111.1% of RTP*

b. Low Setpoint- 8.3 4.56 0 $25% of RTP* s27.1% of RTP*

) '3. Power Range, Neutron Flux, 1.6 0.5 0 $5% of'RTP* with

$6.3% of RTP* with High Positive Rate a time constant a time constant 22 seconds 22 seconds 4

,  ? 4. Power Range, Neutron Flux, 1.6 0.5 0 55% of'RTP* with $6.3% of RTP* with

!

• High Negative Rate a time constant a time constant

. 22 seconds 22 seconds

5. Intermediate Range, 17.0 8.41 0 $25% of RTP* s31.1% of RTP*

Neutron Flux

6. Sourge Range, Neutron Flux 17.0 10.01 0 $105 cps $1.6 x 10' cps

)

7. Overtemperature AT 6.5- 3.5 1.7** See Note 1 See Note 2

+0.5**

8. Overpower AT 4.9 2.2 1.7 See Note 3 See Note 4 l p 9. Pressurizer Pressure - Low 3.12 0.86 0.99 21945 psig 21,931 psig c .

g 10. .ressurizer Pressure - High 3.12 1.00 0.99 .s2385 psig s2,398 psig 3

a 2

• HIP .RAIE0 1HEMAL POWER

• *lhe tensor error for I,,, is 1.7 and the sensor error for Pressurizer Pressure is 0.5. "As measured

• l

. senso errors may be used in lieu of either or both of these values, which then must be summed to determine i the usertemperature AY total (hannel value for 5.

i

X; TABLE 2.2-1 (continued) 3; REACTOR TRIP SYSTON INSTRUMENTATION TRIP SETPOINTS  !

E R . . SENSOR

, TOTAL .

ERROR e FUNCTIONAL UNIT Att0MANCE (TA) 1. (S) TRIP SETPOINT Att0WABLE VALUE 5

j 11. Pressurizer Water Level - High_ 8.0 4.20 0.84 592% of instrument span s93.75% of instrument - .

span  !

12. Reactor Coolant flow - Low 2.5 1.9 0.6 290s of loop 289.3% of loop l design flow

• design flow * ,

13. Steam Generator Water- 14.0 12.53 0.55 214.0% of narrow 212.6% of narrow Level Low - Low range instrument' range instrument '

span - span ,

14. Undervoltage - Reactor 15.0 1.39- 0 210,200 volts 29,822 volts l Coolant Pumps t

0* 15. Underfrequency - Reactor 2.9 0 0 255.5 Hz 255.3 Hz l Coolant Pumps i

16. Turbine Trip .

t

a. Low Fluid Oil Pressure N.A. M.A. N.A. 2500 psi 9 2450 psig l i

b. Turbine Stop Valve N.A. N.A. N.A. 21% OPen 21% open i Closure "

17._ Safety injection input N.A. N.A. M.A. N.A. N.A. f f' rom ESF i ar i 3

• Loop design flow - 95,700 gpm E

r r

l t

f t

ih

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

y

[ .

.r i .,

v>.

- g' TABLE 2.2-1'(Continued!-

[ E TABLE NOTATIONS

g i l x NOTE 1: '0VERIEMPERATURE AT b

H i+ (I 75)3 SAT,'@,-(II{# (l r 5) 3 (1 + r,5)

~

• 2(P - P') - f,(AI)}

- i

f. l. '

l- idhere: AT. - Measured AT by RTD Instrumentattoa;

)' l .

1+rS - Lead-lag compensator on measured AT;.  !

.1 + 7,S -

i 7 3 , r, -

Time constants utilized in lead-lag compensator for AT, ra 2.8 s, i

~

72 5 3 Si '

j 1 .- Lag compensator on measured AT; j- l + 7sS 7

~

73 -

Time constants utilized in the lag compensator for AT, 37 - O s;

[

, AT, -

Indicated AT at RATED THEIMAL PJL TR;

}i K3 - 1.0995; K, = 0.Oll2/*F; l

I + r.S -

The function generated by the lead-lag compensator for T.,

i -1+7S3 dynamic compensation; i 7., r3 - - Time constants utilized in lead-lag compensator for T. , 7. 2 33 s, i R 73 s 4 s; i E j &

. T -- Average temperature. *F;.

} 5 Lag compensatcr on measured-T 1 - -

j g 1 + 7,5 4

f 7,, -

lime constant utilized in the measured T.., lag compensator, 7. - O s; i-j _.

.. . - . , , . . . , , . . , -- ., , . _ - ,_ _ . . _ . _ _ _ _ _ _._,__._-__-_-._,_______s

M TABLE 2.2-1 (Continued) 3; TA8tE NOTATIONS i'

8 jii! NOTE 1: (Continued)

T' s 588.5 F (Nominal T. ,at RATED THERMAL POWER);

4 .

K3 - 0.000519/psig;-

.: P -

Pressurizer pressure, psig; i

i P' =

2235 psig (Nominal RCS operating pressure);

! +

S - Laplace transform operator, s-2; i

j and f,(AI) is a function of the indicated difference between top and bottom detectors of the j power-range neutron ion chambers; with gains to be selected based on measured instrument i response during plant startup' tests so that:

4

?

• (1) for q, - g between -35% and + 81, f,(AI) = 0, where q, and g are percent RATED THERMAL i

POWER in the top and bottom halves of the core respectively, and- g + g is total THERMAL

POWER in percent of RATED THERMAL POWER; f-l (2) for each percent that the magnitude of q, - g exceeds - 35%, the AT Trip Setpoint shall be automatically reduced by 1.09% of'its value at RATED THERMAL POWER; and 4

j (3) for each percent that the magnitude of q, - q, exceeds +8%, the AT Trip Setpoint shall

be automatically reduced by 1.00% of its value at RATED THEIMAL POWER.

1

NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.5%

j of AT span. l 4

i it

~

! ~

j $

i 4

. _ . ""~

+g .

en + - y - --. - - . , _ m +.,. - -_, -m- -_w v.- w

M TABtf 2.2-1 (Continued)

E TABLE NOTATIONS (Continued)

E R NOTE 3: (Continued)

K. - 0.001386/*F for T > I" and K. - O for I s T",

h I - As defined in Note 1 I"  ;-

Indicated T ,, at RATED IliERMAL POWER (Calibration temperature for AT instrumentation, s 588.5'F), 1 S - As defined in Note 5, and f2 (AI) = 0 for all AI.

N0!E 4: The channel's maximum Trip Setpoint shalil not exceed its computed Trip Setpoint by more than 2.0% of AT span. l c;

if O

Ilr# M

m

.'P09fR DISTRIBUTION tlMITS 3/4.2.5 DNR PARAMETERS LlHITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the following limits:

a. Reactor Coolant System T.,,, s 594.3'f

b. Pressurizer Pressure, t 2205 psig*

c. Reactor Coolant System flow. 2 392,000 gpm**

APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVElll AN[JL RE0VIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within -its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CAllBRAT10;4 at-least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by a precision heat balance measurement _to be within its limit prior to operation above 95% of RATED 1HERMAL '

POWER after each fuel loading. The provisions of Specification 4.0.4 are not applicable for entry into MODE 1.

• Limit not applicable during either a THERMAL POWER ramp in excess of b% of

~ RATED THERMAL POWER per minute or a THERMAL POWER step in excesc of 10%

of RATED THERMAL POWER.

• • Includes a 2.4% flow measurement uncertainty.

SEABROOK - UNIT 1 3/4 2-10 Amendment No.

TABLE 4.3-1 REACTOR TRIP SYSTEN INSTRUMENTATION SURVEILLANCE REQUIRENENTS M

g" TRIP 8 ANALOG ACTUATING MODES FOR jR CHANNEL DEVICE WHICH

. CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE c: FUNCTIONAL UiJIT CHECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED 5

] 1. Manual Reactor Trip N.A. N.A. N.A. R(13) N.A. 1, 2, 3*, 4*

2. Power Range, Neutron Flux

a. High Setpoint S D(2,4), Q(16) N.A. N.A. 1, 2 N(3,4),

Q(4,6),

R(4,5)

b. Low Setpoint S R(4) S/U(1) N.A. N.A. 1***,2

3. Power Range, Neutron Flux, M.A. R(4) Q(16) N.A. N.A. 1, 2 High Positive Rate 40

4. Power Range, Neutron Flux, N.A. R(4) Q(16) N.A. N.A. 1, 2 Y

w tilgh Negative Rate

5. Intermediate Range, S R(4,5) S/U(1) N.A. N.A. 1***, 2 Neutron Flux

6. Source Range, Meutron Flux 5 R(4, 5) S/U(1),Q(9,16) N.A. N.A. 2**, 3, 4, 5

7. Overtemperature AT S R Q(16) N.A. N.A. I, 2 l

8. Overpower AT S R Q(16) N.A. N.A. 1, 2

9. Pressurizer Pressure--Low S R Q(16,17) N.A. N.A. I j" 10. Pressurizer Pressure--High 5 R Q(16,17) N.A. N.A. 1, 2 Si 11. Pressurizer Water Level--High 5 R Q(16) N.A. N.A. I 3

12. Reactor Coolant flow--Low S R Q(16) N.A. N.A. 1 O

L = -*_4m m.M

l TABLE 4.3 1 Icontinued)

VABLE NOTATIONS _(Continued)

(12) Number not used.

(13) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the j OPERABILITY of the undervoltage and shunt trip circuits for the Manual l Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).

(14) local manual shunt trip prior to placing breaker in service. l (15) Automatic undervoltage trip.

l )

(16) Each channel shall be testea at least every 92 days on a STAGGERED TEST l BASIS.

(17) These channels also provide inputs to ESFAS. Comply with the applicable [

MODES and surveillance frequencies of Specification 4.3.2.1 for any por-tion of the channel required to be OPERABLE by Specification 3.3.2.

st SEABROOK - UNIT 1 3/4 3 13 Amendment No.

~

.' POWER DISTRIBUTION tlMITS BASES 3/4.2.5 DNR PARAMETERS The limits on the DNB related parameters assure that each of the parameters is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. Operating procedures include allowances for measurement and indication uncertainty so that the limits of 594.3*F for T,y and 2205 psig for pressurizer are not exceeded.

The measurement. error of 2.4% for RCS total flow rate is based upon per-forming a precision heat balance end using the result to normalize the RCS flow rate indicators. Potential fouling _of the feedwater venturi which might not be -

detected could bias the result from the precision heat balance in a noncon-servative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is- applied. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending vari-

- ous plant - performance -parameters. If detected, action' shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the specified limit.

SEABROOK - UNIT 1 B ?/4 2-4 Amendment No.

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2 IV. , Safety Evaluat ion of Sunplement 1 to hirenne Amendment _ Request 9?-01 New Hampshire Yankee is planning to implement a design change (DCR 90-03) at Seabrook Station during the second refueling outage. This design change i will remove the existing Resistance Temperature Detector (RTD) Bypass system and replace this hot leg and cold leg temperature measurement method I with a modified sy stem consisting of f ast-response thermowell mounted RTDs installed in the reactor coolant loop piping. The existing RTD Bypass system and the modified hot leg and cold leg temperature measurement system are d escribed below. Westinghouse has prepared a topical report VCAP-13181 1

• RTD Bypass Elimination Licensing Report for Seabrook Nuclear Power Station (Proprie ry) in support of the four loop operation of Seabrook Station utilizing the new thermovell mounted RTDs. A copy of-this report is provided in Section VIII. Yankee Atomic Electric Company (YAEC) has also evaluated the RTD Bypass System Elimination relative to containment rtsponse. Steam Generator Tube Rupture and Boron Dilution event s. The Vestinghouse and YAEC evaluation conclusions and documentation are i discussed below and in Section V.

Existany RTD Bypass !;ystem Currently, the hot leg and cold leg RTDs used for reactor control and reacter protection are inst rted into manifolds in the Reactor Coolant System bypass loops. Separa :e bypass loops are provided for each reactor coolant loop such that individual loop temperature signals may be developed for use in the reactor contrcl and reactor protection systemn. A bypass loop f rom the hot J ag side of e.ach steam generator to the intermediate leg is used for the hot leg RTDs. Another bypass loop f om the cold leg side of the reactor coolant pump to the intermediate leg is used for the cold leg RfDs. Both hot leg and cold leg manifolds empty through a common header to the intermediata leg between the steam generator and reactor coolant pump. The RTDs are located within manif olds and are inserted directly into the reactor coolant bypass flow without thermowells. The i byp.iss manifold system limits high velocity coolant flow to che RTDs and l compensates for the temperature streaming effects present_ in the hot leg piping. For each hot leg bypass loop, flow is provided by three scoop tubes located at 120 degree intervals around the hot leg. Because of the mixing ef fects of the reactor coolant pump. only one connection is required for bypass flow to the cold leg bypass manifold. 1 The output from the bypass loop RTDs provides the signal necessary to calculate the average loop temperature (Tyg) and the loop ditferential temperature (Delta T). The T.y, and Delta T signals a.e then input to the reactor protection system and the reactor cont rol sy9 tem.

Modified Hot Len and Cold Len Temneraturefgargjpent SysteJg l The individual loop temperature signals requised fo; input to the reactor control and reactor protection systems will be obta med using RTDs l

installed in each reactor coolant loop.

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The hot leg temperature messurement on each loop will be accomplished using three fast response, narrow range, dual element RTDs mounted in thermove11s. Both elements of each hot leg RTD are wired to the appropriate process protection rack where the second RTD input is a spare.

To accomplish the sampling f unction of the RfD bypass manif old system and to minimize the need for additional hot leg pipin, penetrations. the thermowells will be located within two of the three existing hot leg RTD

- bypass manifold scoops. Due to a structural interference. the third RTD will be located in an independent boss. On loops A. B. and D the ,

independent boss is located in the same cross-sectional plane as the ,

existing scoops but offset 30' from the unused location. On loop C. the boss will be relocated to a position approximately 12 inches upst team of the existing scoops at approximately 105' f rom top dead center. The unused 1 scoops (the 120* location on loops A & C and the 240" location in loops B & D*, will be capped. These 3 RTDs will be used to obtain the hot leg temperature used for generation of teactor coolant loop Delta T and T,y.

This modification will not affect the single - wide range RTD currently  !

installed near the entrance of each steam generator. This RTD will continue to prov!de. the hot leg temperature used for monitoring and control.

The cold leg temperature measurement on_each loop will be accomplished _

using one fast response, narrow range, dual-element R1D located in each cold leg at the discharge of the reactor coolant pump (as replacements f or the cold leg RTDs located in the bypass manifold). This RTD will measure

- the cold leg temperature which is used to calculate reactor coolant loop Delta T and T,y. The existing cold leg RTD bypass penetration nozt.le will be modifiec to accept the RTD thermowell. Both elements of the cold leg RTDs will be wired to the appropriate process protection rack where the '

second RTD input is a spare.

This modification will not affect the single wide range RTD in each cold leg currently installed at the discharge of the reactor coolant pump. This RTD will continue to provide the cold leg temperature for monitoring and control.

The RTD bypass manifold returri line to the RCS crossover leg will be capped

- at the connection to the crossover leg.

WCAP-13181, Figure 1.3-1 provides a block disgram of the modified electronics. The hot leg RTD measurements (three per loop) will be 1 electronically -averaged in the. reactor protectier, system. The hot. leg l

averaging. will he accomplished by additions to the existing process protection equipment. The averaged T u signal will then be used with the T ut, signal to calculate reactor coolant loop relta T and T,y which are

. used in the reactor control and reactor protection system. _

- The process protection equipment modifications will be qualified to the same level as the existing process protectivn equipment. The RTDs are environmentally ' qualified per New Hampshire Yankee's ccmpliance with 10CFR$0.49.

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s Existing control board Delta T and T indicators and a?.:rms provide the means of identifying RTD failures. obouldthe S failure of a hot leg RTD be diagnosed, two methods are available for acdressing the failed RTD.

The preferred method is to utilize the second element of the RTD. Since both elements of each dual element RTD are wired to the appropriate procen protection rack, Instrumentation and Control (I&C) personnel can disconnect ,

the failed element f rom the rack terminal strip and conner the other RTD j element. It the spare element is not available, the secor., method it. for the IEC personnel to defeat the failed hot leg RTD and rescale the electronics to average t.he remaining two signals and incorporate a bias based upon the hot leg streaming measured in the loop. VCAP 13181 Appendix B provides the calculational methodology for hot leg temperature bias values. Ghould a failure of a cold leg RTD be diagnosed, the ILC personnel would disconnect the f ailed element irom the rack terminal strip and connect the other RTD element.

The ef fect of the increased instrument uncertainty on updated Final Saf ety Analysis Report (UFSAR) Chapter 6 and 15 LOCA and non-LOCA accident '

analyses within the Westinghouse scope has been evaluated as discussed in WCAP 13181. Relative to both the LOCA and non-LOCA safety analyses, Vestinghouse has concluded in WCAP-13181 that the modification does not affect the conclusions of the UFSAR safety analynes.

Additionally, Yankee Atomic Electric Company (YAEC) has evaluated the affect of the modified system for hot leg and cold leg temperature measurement on (1) containment response, (2) Boron Dilution events and (3)

Steam Generator Tube Rupt.are design basis events.

Relative to centainment response, YAEC concluded that during the liaiting event (large break LOCA), the early containment pressure response during the blowdown phase may increase slightly due to the increase uncertain associated with the modification. However, the long term and peak ccatainment pressure are still valid and the effects of the modification on the containment response is bounded by the current analysis. The YAEC evaluation of the affect of the modification on containment response is  ;

enclosed in Section VIII.

Yankee Atomic Electric Company has concluded that the increased uncertainty associated with the modification will have a negligible ef fect on the Steam Generator Tube Rupture analysis which was perf ormed by them and submitted to the URC on April 16, 1991 in NHY letter NYN-91061. Yankee Atomic Electric Company also concluded that the modification will have negligible effect on the Boron Dilution analysis to be performed by them for Cycle

3. The YAEC evaluation of the af f ect of tne modification on the Steam Generator Tube Rupture analysis and on the Baron Dilution analysis which is to be performed for Cycle 3 is enclosed in section VIII.

New Hampshire Yankee has also proposed to increase the Reactor Coolant System flow rate requirement of Technical Specification Limiting Condition for Operation (LCO) 3.2.5 from the current value of 391,000 gpm to a new 9

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., . I d.* j value of 392.000 gpm to reflect the increase in flow measurement uncertainty as documented in WCAP-13181 The proposed 2.42 flow measurement uncertainty value includes an additional penalty of 0,12 flow to account for undetected feedwater ventur;i f ouling as stated in the Bases for Technical Specification 3/4.2.5. ,

The proposed revisions to Technical Specification 3/4.2.5 (DNB Parameters) for RCS flow from a value that includes 2,12 rnessurement uncertainty to a value that includes 2.42 measurement uncertainty has no effect on the accident analyses since the analyvin limit which is based on the ,ermal l

design flow will not be changed, i I

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v. petermination of sinnificant Hazards for sunnlement I to hicense AnmnoD wnt Request 92-01 New Hampshire Yankee has determined that License Amendment Request 92-01 and Supplement 1 thereto do not involve a significant hazard consideration pursuant to the standards of 10CFR50.0 : based on the f ollowing evaluation.

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

Westinghouse has prepared WCAP-13181 "RTD Bypass Elimination Licensing Report for Seabrook Nuclear Station" (Proprietary) in support of the four loop operation of Seabrook Station utilizing new thermowell mounted RTD's. For the Westinghouse scope. WCAP-13181 contains a safety evaluation for this modified hot leg and cold leg temperature measurement system. This significant her.ards evaluation i addresses both the mechanical modifications to the reactor coolant system pressure boundary and the instrumentation uncertainty changes associated with tne modified system.

The installation of thermowells and fast responso RTDs will not increase the probability of an accident previously analyzed. The modifications to the-Reactor Coolant System pressure boundary will be performed utilizing the same ASME Section III installation requirements as were used for the original installation. The installation requirements are specified in the ASME Section 11I 1977 Edition thru Winter 1977 Addenda.

The removal of the bypass piping and valves associated with this piping will enhance the integrity of the Reactor Coolant System.

By removing significant lengths of piping, numerous valves and instrument panetrations the probability ut a small break LOCA will be reduced.

The new thermowell mounted RTDs have a total response time equivalent to the existing system as discussed in WCAP-13181. The increased instrumentation uncertainty associated with the new thermowell mounted RTDs necessitated an increase in the overpower AT K4 term safety aralysis limit and conservative changes to the K6 term to assur, protection for all power ranges. The Overpower AT and Overtemperatuta AT funct' ;ns thus continue to provide an equivalent degree of reactor protection. RTD signal processing and the added circuitry to the reactor protection system racks will be accomplished using the same type of Westinghouse 7300 series reactor protection system _t ;hnology as has been previously qualified and used in the reactor protection system of Seabrook Station. There is no change i in the use of the temperature signals by any reactor protect; or reactor control system.  ;

The compliance of Seabrook Station to IEEE 279-1971 ('IEEE Standard:

Criteria for Protection Systems for Nuclear Power Generating 11

Stations'), applicable NRC General Design Criteria and regulatory guides has not changed.

This modification does not increase the radiological ce 'uences of any accident previously evaluated. Alth c;h the pressi ndary will be modified, proper welding techniques, penetra ting, radiographs, and system hydrostatic tests will insure ths egrity of the pressure boundary and thus not contribute to any radiological consequences.

The proposed revisions to Technical Specification 3/4.2.5 (DNB Parameters) f or RCS flow f rom a value that includes 2.1% measurtment uncertainty to a value that includes 2.42 measurement uncertainty has no ef fect n the accident analyses since the analysis limit which is based on the thermal design flow will not be changed. The effect of undetected venturi fouling, has been included in the RCS flow -

requirement of Technical Specification 3/4.2.5.

Surveillance Requirement 4.2.5.3 for the precision heat balance L determination of RCS flow is changed from being required prior to opetation above 75Z Rated Thermal Power (RTP) to being required prior to exceeding 95% RTP. Performance of the precision heat balance above 90% RTP was recommended by Westinghouse in association with the RTD bypass elimination to minimize flow rate measurement uncertaintii that are exacerbated at lower power levels. The precision hen balance is performed each cycle to detect changes in the RCS flow element (elbow taps) characteristics that would affect the accuracy of the RCS flow indication. Significant changes in the characteristics of all of the elbow taps over a single operational cycle is not credible. Performing the flow rate measurement prior to exceeding 952 RTP provides adequate margin to DNB in the highly improbable event that there is a degradation in RCS flow rate that is masked by a simultaneous non-conservative change in all elbow p taps, w

The effect of the increased instrument uncertainty on updated Final Safety Analysic Report (UFSAR) Chapter 6 and 15 LOCA and non-LOCA accident analyses within the Westinghouse scope has been evaluated as discussed in WCAP-13181. Relative to both the LOCA and non-LOCA safety analyses, Westinghouse has concluded in WCAP-13181 that the modification does not affect the conclusions of the UFSAR safety analyses.

Additionally, Yankee Atomic Electric Company (YAEC) has evaluated the affect of the modified system for hot leg and cold leg temperature measurement on (1) containment response, (2) Boron Dilution events and (3) Steam Generator Tube Rupture design basis events.

Relative to containment response YAEC concluded that during the limiting event (large break LOCA), the early containment pressure response during the blowdown phase may increase slia,htly due to the 12

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increased uncertainties associated with the modification. However, the long term and peak containment pressures are still valid and the effects of the modification on the contaiament response is bounded by the current analysis. The YAEC evaluation of the affect of the modification on containment response is enclosed in Section VIII.

Yankee Atomic Electric Company has concluded that the increased uncertainties associated with the modification will have a negligible effect on the Steam Generator Tube Rupture analysis which was performed by them and submitted to the NRC on April 16, 1991 in NHY le tter NYN-91061. Y ' ! Atomic Electric Company also concluc;ed that the modification will nave negligible effect on the Boron Dilution analysis to be performed by them for Cycle 3. The YAEC evaluation of the af f ect of the modification on the Steam Generator Tube Rupture analysis and on the Boron Dilution analysis which is to be performed for Cycle 3 is enclosed in Section VIII. -

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

The removal of the RTD Bypass System will not create the pm =ibility of a new or different kind of accident from any accident previously evaluated. The reactor coolant pressure boundary modifications design and installation will be equivalent to the original RCS design and installation. Reactor coolant loop temperature inputs for reactor control and reactor protection functions will continue to be supplied. Other equipment important to saf ety will be unaf f ected and will continue to function as designed.

The removal of the Resistance Temperature Detector (RTD) bypass piping and the installation of a mcdified temperature measurement system does not affect the integrity of the reactor coolant system pressure boundary. This is due to the reactor coolant piping (pressure boundary component) modifications adhering to the ASME Code (Sections III, Class 1 and Section XI) and to the NRC General Design -

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Criteria. Installation requirements will be equivalent to the original RCS installation pursuant to ASME Section III,1977 Edition thru Winter 1977 Addenda.

The removal of the RTD Bypass System eliminates components that have been a major cause of plant outages in the industry as well as a major contributor to oc~upational radiation exposure. Additionally, with these components removed, the probability of a malf unction f rom them is eliminated. The installation of fast response thermowell mount- RTDs on the reactor coolant loop piping and additional proces=ing electronics will continue to provide the individual loop temperature signals for input to the reactor control and reactor protection systems using components that are environmentally and seismically qualified.

The RTD Bypass System flow alarm is no longer required to warn of flow reduction that could affect instrument system response. Flow 13

through the scoo> tumes with thermowells is not monitored because blockage of the tiow path is not credible. Blockage is not credible because of the nultiple scoop tube holes , the size of tha holec, and administrative and chemistry controls that prevent the introduction of objects that could block the flow path.

The modificr tion does not af fect the ability of the protection system to mitigate the radiological consequences of any accident. The new RTD signals are processed to provide equivalent signals to those provided by the original direct immersion RTDs. Since three RTDs will be used to provide an average hot leg temperature as opposed to the original use of one RTD, the consequences from a failed RTD are unchanged. Manual actions to bypass a f ailed RTD channel remain the same.

3. The proposed changes do not result in a significant reduction in the -

margin of safety.

The instrumentation uncertainty analysis associated with this modification has resulted in proposed Technical Specification changes to the uncertainty terms associated with overpower AT and Overtemperature AT and low Reactor Coolant System (RCS) Flow reactor trip functions. Additionally RCS average temperature measurements used for control board indication and input to the rod control system, and the value of the RCS flow measurement uncertainty are also affected by the modification. The safety evaluations of this modification which have been performed by Westinghouse and YAEC referenced above conclude that sufficient margin exists such thet margins to safety are not affected.

The proposed Technical Specification changes also include the elimination of the bypass piping loop low flow alarms and the revision to the Technical Specification requirement for RCS flow.

The proposed increase in the RCS flow requirement reflects the RCS -

flow measurement uncertainty increase associated with the new RCS temperature measurement system. The proposed RCS flow limit will ensure that RCS flow is greater than or equal to the thermal design flow analysis value.

The RTD Bypass System flow alarm is no longer required to warn of flow reduction that would affect instrument system response. Flow through the scoop tubes with thermowells is not monitored because blockage of the flow path is not credible. Blockage is not credible because of the multiple scoop tube holes , the size of the holes , and administrative and chemistry centrols that prevent the introduction of objects that could block the flow path. The removal of this alarm does not result in a reduction in the margin of safety.

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VI. Royosed Schedule for License Amendment Issuance and Effectiveness New Hampshire Yankee requests NRC review of Supplement 1 to License Amendment Request 92-01 and issuance of a license amendment by August 15, 1992. This schedule is proposed in support of NHY's plans to implement the RTD Bypass System Elimination design change during the second refueling outage which is scheduled to begin on September 7, 1992.

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VII. Environmental Impact Assessment New Hampshire Yankee (NHY) has reviewed the proposed license amendment against the criteria of 10CFh51.22 for environmental considerations. The proposed changes do not involve a significant hazards consideration, nor increase the types and amounts of ef fluents that may be released of f site, nor significantly increase individual or cumulative occupational radiation exposures. Based on the f oregoing, NHY concludes that the proposed change meets the criteria delineated in 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement.

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VIII. Other Supportinn Information The following documents were enclosed in License Amendment Request 92-01. They are not affected by the alternative Technical Specification changes proposed in Supplement 1 to License Amendment Request 92-01. Refer to License Amendment Recuest 92-01 for a copy of these documents.

Westinghouse Authorization Letter CAW-92-255 and accompanying affidavit Proprietary Information Notice Copyright Notice Westinghouse WCAP-13181 (Proprietary). "RTD Bypass Elimination Licensing Report for Seabrook Nuclear Station" --

Westinghouse WCAP-13193 (Non-Proprietary), "RTD Bypass Elimination Licensing Report for Seabrook Nuclear Station" Evaluation of the Effects of Removal of the RTD Bypass System on Containment Response of Seabrook Station Evaluation of the Effects of Removal of the RTD Bypass System on YAEC-1698 and the Seabrook Station Boron Dilution Analysis Note: (1) New Hampshire Yankee notes that the deletion of the Technical Specification for low RCS Tavg coincident with reactor trip feedwater isolation Functional Unit (Table 3.3-4, Functional Unit 6.b) recommended in WCAP-13181 will be addressed in a future license amendment request.

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