ML20091A889

Proposed Tech Specs Re Reactor Trip Sys Instrumentation Trip Setpoints in Ref to RTD Bypass Sys

ML20091A889
Person / Time
Site:	Seabrook
Issue date:	03/20/1992
From:	PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
Shared Package
ML20034D387	List: ML20091A880 ML20091A882 ML20091A889 ML20091A895 ML20091A900
References
NUDOCS 9203300253
Download: ML20091A889 (31)

Text

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

, TABLE 2.2-1 -

m g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINIS 8

' og g SENSOR '

co , ; TOTAL. ERROR dc do z FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SEIP0lNT ALLOWABLE VALUE gg q 1. Manual Reactor Trip N.A. N.A. N.A. N.A. M.A.

L no g N 2. Power Range, Neutron Flux o4 oo a. liigh Setpoint 7.5 4.56 0 1109% of RTP* ill1.1%.of RIP

• ou

'U O PJ ggo b. Low Setpoint 8.3 4.56 0 125% of RTP* $27.1% of P.TP*  !

u

3. Power' Range, Neutron Flux, 26 0.5 0 <5% of RTP* with <6.3% of RTP* with i High Positive Rate- 3 time constant i time constant i 12 seconds 32 seconds I

rf 4. Power Range, Neutron Flux, 1.6 0. 5 - 0 $5% of RTP* with

• High Negative Rate 16.3% of RTP* with a time constant time. constant '

32 seconds 12 seconds

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

Neutron Flux 131.1% of RTP*

6. Source Range, Neutron Flux 11.0 10.01 P $10S' cps 11.6 x 105 cps' 3.5 i

7. Overtemperature AT 6. 5 6 ,O** See i4c3e 1 See Note 2 d ,

-06s ,)

I:~f

~

g4 - y

8. Overpower AT- ) .

Sap.. Nota 3 1~

S.3e Note 4 b - (\

2.2  %'%

9. Pressurizer Pressure - Low 3.12 0.86 0.99 ?M.1 ')ig 31,931 psig 1)J

10. Pressurizer Pressure - liigh 3.12 1.00 0.99 $2385 psig <2,398 psig -

N

• RTP =' RATED T11ERMAL POWER I7 0

• • The sensor error for I is M and the sensor error for Pressuriier Pressure is[ .,4 . "As measured" -

sensor errors may be used in lieu of either or both of these values, which then must be summed to deter-mine the overtemperature AI tctal channel value for L

]D J T i

_ .'

• u=p

m TABLE 2.2-1 (Continued) .

h

n REACTOR TRIP . .IEM INSTRUMENTATION TRIP SETPOINTS h SENSOR TOTAL ERROR

[ FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT Alt 0WABLE VALUE E 11. Pressurizer Water Level - High 3.0 4.20 0.84 <92% of instrument <93.75% of instrument

12. Reactor Coolant flow - Low 2.5 IjflJj 0.6 of loop h 4; f.9 _ design flow

• design flow *

13. Steam Generator Water 14.0 12.53 0.55 >14.0% of narrow >12.6% of narrow Level Low - Low range instrument fange instrument span spac

14. Undervoltage - Reactor 15.0 1.39 0 >10,200 volts >9,822 volts Coolant Pumps ry 15. Underfrequency - Reactor 2.9 0 0 >55.5 liz >55.3 Nz 0;olant Pumps

16. Turbine Trip

a. Low Fluid Oil Pressure N.A. N.A. N. A. >500 psig >450 psig

b. Turbine Stop Valve N.A. N.A. N.A. >1% open ->1% op s Closure

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

from EST lNw h

• Loop design flow = 95,700 gpm 9

0 J'

-~

TABLE 2.2-1 (Continued)

M MBtE NOTATIONS o

NOTE 1: OVERIEMPERATURE AT AT' I

, g3 ,ag) $ AT, Mi - K 2 [I (3 ts3)

- P ] + K3 (P - P') - f,(AI))

E a

e Where: AT = MeasuredATbyRfDlHanifeldl Instrumentation; f{ = lead-lag compensator on measured AT; 13 T2

= Time constants t. ilized in lead-lag compensator for AT, t i>8s, 12 _ 3 5; 1

= ag c mpensa r n measured AT; 1+t 53 u

J 13 =

Time constants utilized in the lag compensator for AT,13 = 0 s;

= Ind:cated AI at RATED THERMAL POWER; AT, K = 1.0995; K2 = 0.0112/*F; I * **

1+1 35

= The function generated by the lead-lag compensator fcr T dynamic compensation; avg 14, is = Time constants utilized in the lead-lag compensator for T i, > 33 s ,

d'9, 15 < 4 s; T = Average temperature. *F; D

g I = f*

Lag c spensator on measured I  ;

1+t S g

\s te = Time constant utilized in the measured I lag compensator, to = 0 s; k

TABLE 2.2-1 (Continued) m 9 TABLE NOTATIONS E

8 NOTE 1: (Continued)

T' at RATED TilERMAL POWER);

c

-< 588.5'F (Nominal T avg z

Q K3 = 0.000519/psig; e

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

5 = Laplace transform operator, s ';

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

(1) For.q g between .35% and,+ 8%, f (a!) = 0, where q and o are b percent RATED TilERMAL

] t b t POWER in the top.and bottom halves of the core respectively, and q +ob is t tal. THERMAL POWER in percent of. RATED THERMAL POWER;

(/) For each percent that the magnitude of q t ~a b exceeds -:35%, the AT Trip Setpoint shall-be automatically reduced'by 1.09% of its value at RATED THELAAL POWER; and (3) For each percent that the magnitude of q t g exceeds +-8%, the AT Trip Setpoint shall

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

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

of AT. span. g Ws

-6 4

%d-6 s

h

i'

, TABLE 2.2-1 (Continued) h-

. TABLE NOTATIONS (Continued) 8 7 NOTE 3: (Continued) O*# '38M"E E K. =j 0.^0128/*] for T > 1" and Ka = 0 for T < 1",

M ~

g =  !

T As defined in Note 1 ,

T" =

Indicated T ,g at RATED THERMAL. POWER (Calibration temperature for AT instrumentation, 5 588.5'F), '

5 = As defined in Note 1, and f (at) = 0 for all AI. '

. f 7 NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 5 of AT span.

2. .o r, l

J l

lti 4 en h

9f'_ ,

g -

~

6 o

LIMIVfNG SMETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Intermediate and Source Range, Neutron Flux The Intermediate r.nd Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels, The Source Range channels.will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Raige channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active, Overtemperature aT-The Overtemperature AT trip provides core protection to prevent DNB'for.all combinations of pressure, power, coolant temperature..and axial power distribu-tion, provided that the transient is slow with respect to pipina transit delays frocn the core to the temperature d:t::t '; (about 4 seconds), and pressure.is within the range between the Pressurizer High and Low Pressure trips. The Set-point is automatically varied withi (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water and includes-dynamic compensation for piping dela

- deteches, (2) pressurizer pressure,ysand from(3)the corepower axial to. thedistribution. toep temperature With +

normal axial power distribution, this Reactor trip. limit is always-below the ,

core Safety Limit as shown in Figure 2.1-1. If-axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear _ detectors, the Reactor trip is: automatically reduced according to the notations in Table 2.2-1.

Overpower AT measurent . s'M a n d +ey rn%s in s fru M (#d ' o " ck /4 YJ The Overpower AT-trip provides assurance of fuel-4tegrity:(e;g., no-fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range:for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of waters and (2) rate of_ <

change of temperature for dynamic compensation for piping delays from the core to the kee temperature detceter:, to ensure that the allowable heat genera-tion rate (kW/ft) is not exceeded The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported 1n WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

seasurew& sysfem b SEABROOK - UNIT 1 e 2-5

. POWER DISTAL,BUTION LIMITS

% C. 6 2. - 0 (

3/4 2.5 NB DARAMETF.R5 h )O

M!7"4G ::90! ::N:00 SPERATION 3.2.5 ine f allowing DNB-relateo carameters snall oe maintainea ~1tnin :no the following iimits:

5.

Reactor C:olant System T, g, < 594.3 F

b. Pressurizer Pressure, t 2205 psigd

. Reactor Coolant System Flow, g(:01, ^00 gpm""

APPLICABILITY: MODE 1. 382,800 ACTIg:

With any of the aoove parameters exceeaing 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 reauce 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 REOUIREMENTS 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 at least once The per RC518flow rate indicators shall be subjected to CHANNEL CALIBRATION months.

4,2.5,3 9s% 3 The RCS total flow rate shall be determined by a pra s n neaf balance i

measurement to De within its limit prior to operation above 4 of RATED THERMAL POWER after eacn fuel loading. The provisions of Specification 4.0,4 are not applicable for entry into M00E 1.

1 1

' Limit not applicaule curing either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER steo in excess of 10%

i of RATED THERMAL POWER.

" d e.e : I : ?.* Y' [ = E cr N t U [: ; 5 rk

- _ _ - = _

SEABROOK - UNIT 1 3/4 2-10 YY ' /

,/cq.

TABIE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTAllON Si!RVEllt ANCE RIQUIRIMINIS "n

8 IRIP

^ ANALOG ACTUATING MODES FOR

• CilANNEL DEVICE ilitCH E CHANNEL CilANNEL OPERATIONAL OPERATIONAL ACit!AT ION SURVilllANCI Z FUNCTIONAL UNIT CilECK CALIBRATION TEST ILSI LOGIC IESI 15 REQUIRIS w -

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

1. .

2. Power Range, Neut ron Flux

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

Q(4, 6), ~

R(4, 5)

b. l.ow Setpoint 5 R(4) 5/U(l) N.A. N.A. I'**, 2 Power Range, Neutron flux, N.A. R(4) Q(16) N.A. N.A. I, 2 3.

g- High Positive Rate s,-

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

• o liigh Negative Rate Intermediate Range, 5 R(4. 5) 5/U(1) N.A. N.A. 1***, 2 5.

Neutron Flux Source Range Neutron Flux 5 R(4, 5) 5/U(I),Q(9,I6) H.A. N.A. 2*", 3, 4, 5 6.

N.A. N.A. 1, 2

(/ 7. Overtemperature AT 5 I [ Q(16) 5 R Q(16) v4. A. N A. 1, 2

8. Overpower AT 5 R Q(16,17) N.A. N.A. I

9.  : Pressurizer Pressure--Low Pressurizer Pressure--High 5 R Q(16,ll) N.A. N.A. 1, 2 A 10 Pressurizer Water Level--liigh 5 R Q(16) N.A. i4. A. I

. 11.

\ N A. N.A. t D 12. Reactor Coolant f low--Low 5 R Q(16)

]

[,

ISC f2 -nf bM TABLE 4.3 1 (Continued)

TABLE NOTATIONS (Continued) '

@)-#e+4 M "T04ypen-4*p-f4+w-estev (13) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERAB1tlTY of the undervoltage and shunt trip circuits for the Manual Renctor Trio Function.

Bypass Breaker trip circuit (s),The telt shall also verify the OPERABILITY of the (14) local manual shunt trip prior to placing breaker in service, (15) Automatic undervoltage tr p.

(16) Each BASIS,channel shall be tested at least every 92 days on a STAGGERED TEST (17) These channels n'se erovide inputs to ESFAS. Comply with the applicable MODES and surveilla..ce frequencies o,'

Specification 4.3.2.1 for any per-tion of the chantiel required to be OPERABLE by Specification 3.. 2,

/

-gl SEABROOK - UNIT 1 3/4 3-13 yg b

l

' POVER DISTRIBUTION LIMITS rsc. 4 2.- ol BASES Poqf. N c

,6_

w U4. 2. 5 CNB PARAMETERS L_ analytical limits ine limits on the ONU-relateo carameters issudi ach~oY Uie carameters is maintained within the normal steac the transient ano accident analyses. y state enveldw}of coeretton assumea in FSAR assumptions and have been analytically demonstrated aceouate to maintain Thelimitsardconsistentwiththeinitial

• a minimum DNBR of 1.30 througnout each analyzed transient. Operating nrocedures includeallowancesformeasurementandindicatinn]uncertaintysot of 594.3'F for T,y and 2205 psig for pross"We,r sre not excteced. -

y= + _w pressuzo and 382,800 gpm for total RCS flow

=-

uncertainty The measurement mg gf LN for RCS total flow rate is based upon per-forming a precision heat calInce Tind using the result to normalize th RCS flow rate indicators.

Potential fouling of the feedwater venturi which mignt not be cetected ctuid servative manner. bias the result from the precision heat balance in a noncon-feeawater venturi is applied.Therefore, a penalty of 0.1% for undetected fouling of the measurement greater than 0.L% can be detec Any fouling Mich might bias the RCS flow rate ous plant performance parameters. by monitoring and trencing vari-If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quant' Tied and compensated for in the RCS flow rate measurement or tM venturi shall be cleaned to eliminate _

m _ - -

the fouling.

The flow rate measurement is perf ormed prior to operation above 95% RTP to provide margin for flow degradation that la' masked by changos in albow tap normalization.

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

The periodic surveillance of indicated RCS flow is sufficient to detect only limit.flow degradation which could lead to operation outside the specified SEABROOK - UNIT 1 B 3/4 2-4 4

Y s

_ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - ^ - - - ' ~ ~ ~ '

111. Biirpe of Pritposed Gjtam.d b at tached retype of proposed changes to Technlent Specifications. The attachesl retype reflects the currently issued version of Technical Specifications. Pending technic 61 Specification Changes or Technical Specification changes issued subsequent. to thi6 submittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

Revision bars are provided in the rhht hand margin to indicate a revision to t he text. fio revision bara are us.111 red when the page is changed solely to acconnodate the shif ting of text due to additions or deletions.

i 7

i llI {lll 1! l l l e

h h i n

t t i i m r

E

• wtn wt U P

• n

• s e L T P

• a 2 a P p t e

A R T P+ Pt T c g g d" d V R Tr Ts R f R ns Ens 5 2 4 i i e E o f od od f 0 s s ro L o f cn fcn o 1 e e p p ut s

B  % o o o o t t o

A 1  % ec  % x o 1 8 ad e

W %ecme %me N N 3 9 O . l 1

3.i s 3.ti s2 1

6 e e 9, 3, :e c

L L

A.

N l

i s

7 2 6 t

2 6 1

3 s

1 s S e S e 1 2

2 sw As A $ sa2 5a2 "

e

.b h h 5. st tt tt 0u S T

• in wat in wat s m T

N N P *

• in I

I T P

• s

• s P g g e

O O R T Pns Pns T I 3 eh P P R Tod Tod R i i rt T T f Rcn Rcn s e e s s' u E E o f o o f o

p t o

t o

p p sh sc S o fec f ec c S

P

% ome ome N N 5 4

5 8

ei rh P . 9  % i s i s 5 I

I A.

0 5 %t %t 5 0 e e 9 3 Pw R R 1 2 5 2 5 2 2 1 e e 1 2 r ,

T T N s s sa2 sa2 s s S S 2 s es R

• N OR *

• ze O SO .

• 5 9 9 iu NR) 7 9 9 rl I

T ERS A. 7. 0 ua sv 1 A SE( N 0 0 0 0 0 0 1 + 1 0 0 s 2 E T

N ee rs Pe 2 M U h rt E R T

. 6 6 1 1 6 0 o L 5 5 5 5 4 0 5 2 8 0 ff B S A.

A N I Z

N 4 4 0 0 8 0 3 2 0 1 r o T 1 oh M rt E r o .

T ) ebS S A Y T rrr S ( ooo s f P E nr I C eee R N shu T A .

2 2 tl LW 5 3 6 6 0 0 5 9 1 1 eia R AO A. hev O TL N 7 8 1 1 7 7 '6 4 3 3 t T OL 1 1 fl C TA doe A n n E aun R ea

, , x w h

g 7.ihl c 1

x x x u o i nl u u u l L H sia l l l F i t F F F '- - do n R p

n o o n n o ,

o r

e r

e r

. E WTuT

$t i

r r re re e g

t T u u O A t t tt tt u A s s Pre T u n t ua ua n e s s obe r

e i n eR eR a N e e e Lf r N o p

i N N R r r r A yu o o e ,e , u P P Mrat t , t p ,v v ex e t T Eoma r c e e ei ei g a A r r Hr T

I a

e g

n S

S t

e gt ni gt na t u al iF n

a r

e r e z

e' z'

T rse erp N

U R

R a h g w as Ro ag Re d

en R p m

e w

i r

i r

D Erre om L

l a r i

H L o

r P

r N mo rr e

g e

t o

p u

s u

s Tort Aser A u e eh wg eh wg et r r r s s Rn e N n w o

t u ne u e e e e r

erv O a a

. . oi oi o v v r = soo I M P b PH PH IN S O O P P s T

C P'e n e Th eh N

. . . . . . . . . RTst U 1 2 3 4 5 6 7 8 9 0 *

• F 1

• M$8a , c5]

j  ?* pEgEV g- -

,i ,! f !li!!i,t lillll

$$ TABLE 2.2-1-(co,tinued)

@s REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

SE SEN50R

, TOTAi ERROR FUNCTIONAL UNIT ALI.0WANCE (TA_1 Z (51 TRIP SETPOINT ALLOWABLE VALUE g-I 11. Pressurizer Water Level - High 8.0 4.20 0.84 592% of instrument $93.75% of ir.strument

span .

span

12. Reactor Coolant Flow - Low 2.5 1.9 0.6 290% 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 tow - Low- range instrument range instrument span span

14. Undervoltage - Rcactor 15.0 1.39 0 210,200 volts 29,822 volts Coolant Pumps O' 15. Underfrequency - Reactor: 2.9 0' O 255.5 Hz. 255.3 Hz

'" Coolant Pumps

16. Turbine' Trip

a. Low Fluid Oil Pressure N.A. N.A. N.A. 2500 psig 2450 psig

b. Turbine Stop Valve N.A. N.A. N.A. 21% open 21% open Closure-

17. Safety Injection Input N.A. N.A. N.R. N,A. N.A.

=from ESF:

t 5I

@'

• Loop design flow = 95,700 gpm

?

L e n

~ '

l ,

M TABLE 2.2-1 (Continued)

E TABLE NOTATIONS 8

R NOTE 1: OVERTEMPERATURE AT s AT, @, - K2 (1 + 7,5) 7 U -(1 + r,5) ' +Y~ )~ 2IAI)f

+r) (1 75) 3 Where: AT - Measured AT by RTD Instrumentation; l 1+75- 3 Lead-lag compensator on measured AT; I+752

-7,72 1

- Time constants utilized in lead-lag compensator for AT,17 28s,

,'2 r $ 3 s; I - Lag compensator on measured AT; 7 1+753

~

73 = Time constants utilized in the lag compensator for AT, 73 - O s; ATe - Indicated AT at RATED THERMAL POWER; K1 - 1.0995; K2 - 0.0ll2/*F; I + r S .The function generated by the lead-lag compensate- for T, I+75- 3 dynamic compensation;.

r., is - Time constants utilized in lead-lag compensator for T,, 7. 2 33 s, N'_

o 73 s 4 s:

j h' T . Average temperature, *F;

! 8

" 1- = Lag compensator on measured T,;

• 1 + r,5 7 - Time constant utilized in the measured T, lag compensator, 7. - O s; i

TABLE 2.2-1 (Continued)

%2

@; TABLE NOTATIONS B

5? NOTE 1: (Continued) c: T' s 588.5*F (Nominal T,,, at RATED THERMAL POWER);

-d K3 = 0.000519/psig; P = Pressurizer pressure, psig; P' - 2235 psig (Nominal RCS operating pressure);

S - Laplace transform operator, s-*;

and fi(AI) is a function of the indicated difference 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 q, - q3 between -35% and + 8%, fi(AI) - 0, where q, and q, are percent RATED THERMAL o' - POWER in the top and bottom halves of the core respectively, and q, + g, is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of q, - q, exceeds - 35%, the AT Trip Setpoint shall be automatically reduced by 1.09% of its value at RATED THETAL POWER; and (3) For each percent that the magnitude of q. - q3 exceeds +8%, the AT Trip Setpoint shall be automatically rede ed by I.00% of its value at RATED THERMAL POWER.

t NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.5% l of AT-span.

?

8 8- .

(

a

'M TABLE 2.2-1 fContinue:f)

-@ TABLE NOTATIONS (Continued)

S R NOTE 3: (Continued) c K. = 0.001386/*F For T > T* and K, = 0 for T s T",

5

• T = As defined in Note 1, T" = Indicated T y at RATED THERMAL POWER (Calibration temperature for AT instrumentation, s 588.5'F),

S - As defined in Note 1, and i f2(AI) = 0 for all AI.

NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.0% of. AT span. l E

l a

il .

E

.??

. . . . . . . . . ~ . .....~

I LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSJEM INSTRUMENTATIONXlPOINTS (Con"md)

Intermediate and Source Ranae. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron flux channMs. Thb Source Range channels will initiate a Reactor trip at about 10' counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P 10 becomes active.

Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for.

all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature measurement system and temperature instrumentation delays (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for pipina delays froin the core to the temperature *.ieasurement system and temperature instrumentation delays, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distritsution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1 1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2 1.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel-pellet molting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip'. The Setpoint .is automatically varied with: (1) coolant-temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature _for dynamic compensation for piping delays from the core to the temperature measurement system, to ensure that the allowable heat generatinn rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigste the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response-to Excessive Secondary Steam Releases."

a SEABROOK - UNIT l' B25 Ainendment No.

POWER DISTRIBUTION LIMITS 3/4.2,5 DNB PARAMETERS UjilTING CONDITION FOR OPERATION 3.2.5 The following DND-related parameters shall be maintained within the the following limits:

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

b. Pressurizer Pressure, a 2205 psig*

c. Reactor Coolant System flow. 2 382,800 gpm l

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 />.

SURVEILLANCE 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 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 95% of RATED THERMAL P0E R 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 wcess of 5% of RATED THERMAL POWER per minute or-a THERMAL. POWER step in excess of 10%

of RATED THERMAL POWER.-

SEABROOK UNIT 1

]

3/4 2-10 Amendment No.

__j

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS '

w TRIP h ANALOG ACTUATING MODES FOR 8 CHANNEL DEVICE WHICH jil!

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REOUIRED c.

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

1.

]

2. Power Range, Neutron Flux

a. High Setpoint S D(2,4), Q(16)" N.A. N.A. 1, 2 M(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, N.A. R(4) Q(16) N.A. N.A. 1, 2 l High Positive Rate

! #* N.A. N.A. 1, 2

! 4. Power Range, Neutron Flux, N.A. R(4) Q(16)

Y High Negative Rate w

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

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

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 2j 10. Pressurizer Pressure--High 5 R Q(16,17) N.A. N.A. I, 2 f 11. Pressurizer Water Level--High S

12. Reactor Coolant flow--Low S R

R Q(16)

Q(16)

N.A.

N.A.

N.A.

N.A.

1 I

.E

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

TABLE 4.3 1 (Continued) l TABLE NOTATIONS (Continued)

I f (12) Number not used. I (13) Th TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the j OPERABillTY of the undervoltage and shunt trip circuits for the Manual Reactor Tri) function. The test shall also verify the OPERABILITY of the Bypass Brea(or trip circuit (s),

i (14) local manual shunt trip prior to placing breaker in service. l (15) Automatic undervoltage trip. l (16) Each channel shall be tested at least overy 92 days on a STAGGERED TEST- l f BASIS. l (17) These channels also provide inputs to ESFAS. Comply with the applicable l 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.

l i

i i

l l

l I .

i SEABROOK UNIT I 3/4 3 13 Amendment No.

l b

-,-,-,+-it -n wy--*--w*m--++-m6-wyn y we y-- w e,ww -v w w wp se m -- w+ e,g y g- v w ,= c y, - - w wp isyvvve mey e-- +=www ww + e sr,y-w W --*- v t- r-w e e-w- wee rw www v e+- r-ww~y ww y ew w e W www- v m ru' ve

• E0Fl!LF ',1R18UT10N L1 tilts W, . _ . _

3/4.2.5 DNB FARAMETERS The limits on the DNB-related parameters assure thdt each of the parameters is maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are analytical limits 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 594.0'F for T.,, and 2205 psig for pressurizer pressure and 382,800 gpm for total RCS flow are not exceeded. l The measurement uncertainty for RCS total flow rate is basea upon per- l 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 measurement or the venturi shall be cleaned to eliminate the fouling. The flow rate measurement is performed prior to operation above 95% RTP to provide margin for flow degradation that is masked by changes in elbow tap normalizotIon.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillanc9 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 I B 3/4 2-4 Amendment No.

IV. fiaiet y Evalunt ion of 1.icense Amendment Request 92-01 proposed Channes New Hampshire Yankee is planning to implement a design chang, (DCR 90-03) at Seabrook Station during the second refueling outage. This design change will remove the existing Resistance Temperature Detector (RTD) Bypass Syrtem and replace this hot leg and cold leg temperature measurement method with a modified system consisting of f ast response thermowell mounted RTDs in.talled in the reactor coolant loop piping. The existing RTD Bypass System and the modified hot leg and cold leg temperature measurement system ar e described below. Vestinghouse has prepared a t opical report. VCAP-13181

• RTD Bypass E1.imination Licensing Report for Seabrook Nucicar Power Station' (Proprietary) in support of the four loop operation of Seabrook station utilir.ing the new thermove11 mounted RTDs. A copy of this report is provided in Section VIII. Yankee Atomic Electric Company (YAEC) has n'so evaluated the RTD Bypass System Elimination relative to containment response, St eam Cenerat or Tube Rupture and Boron Dilution evants. The Vestinghouse and YAEC evaluation conclusions and documentation are discussed below and in Section V.

r Exist inn RTD Hvoass Synt:3 Currently, the hot leg and cold leg RTDs used for reactor control and reactor protection are inserted into manifolds in the Reactor Coolant System bypass loops. Separate bypass loops are provided for each reactor coolant loop such that individual loop temierature signals may be developed ior use in the renF or control and reactor protection systems. A bypass loop f rom the hot leg side of each steam generator to the intermediate leg is used f or the hot leg RTDs. Another bypass loop from the cold leg side of the reactor coolant pump to the intermediate leg is used for the cold leg RTDs. Both hot leg and cold leg msnif olds empty through a common header to the intermediate leg between the steam generator and reactor coolant pump. The RTDs are located within manifolds and are inserted directly into the reactor coolant bypass flow without thermowells. The bypass manifold system limits high velocity coolant flow to the RTDs and 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 o# the mixing ef fects of the reactor coolant pump, only one connection is required for bypass flow to the cold leg bypass manifold.

The output f rom the bypass loop RTDs provides the signal necessary to calculate the average loop temperature (T,yg) and the loop ~ differential temperature (Delta T). The T,y and Delta T signals are then input to the reactorprotectionsystemandkhereactorcontrolsystem, piodified Hot in and Cold Len_ Temperature Measurement System The individual loop temperature signals- required for input to the reactor control and reactor protection systems will be obtained using RTLs installed in each reactor coolant loop.

B

Th t4 hot leg temperature measurement on each loop will be accomplished using three fast response, narrow range, dual element RTDs mounted in thermowells. 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 RTD bypass manif old system and to minimize the need for additional hot leg piping penetrations. the thermovells 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. E. and D the independent boss is located in the same cross-sectional plane as the existing scoops, but of f set 30' f rom the unused location. On loop C the boss will be relocated to a position approximately 22 inches upotream of /

the existing scoops at approximately 105* f rom top dead center. The unused 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 reactor coolant loop Delta T and 7,yg.

This modification will not affect the single wide range RTD currently installed near the entrance of each steam generat *r. This RTD will continue to provide 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 RTD 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 rer.ctor coolant 1 cap Delta T and T,yg. The existing cold leg RTD bypass penetration nozzle will be modified to accept the RTD thermowell. Both elements of the cold icg 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 dischar ,e of the reactor coolant pump. This RTD will continue to provide the cold leg temperature for monitoring-and control.

The RTD bypass manifold return 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 diagram of the modified electronics. The hot leg RTD measurements (three per loop) will be electronically-averaged in the reactor protection system. The hot--leg averaging will be accomplished by additions ' to- the existing process protection equipment. The aversged Tgot signal will then be used with the Tm signal to calculate reactor coolant loop Delta T and T og 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 protection equipment. The RTDs are environmentally qualified per New Hampshire Yankee's - compliance with 10CFR50.49.

9

4 i

i .. .

indicators and alarms provide the Existingofcontrol means identifyingboard RfDDelta T and 7,,bould failures. $ the f ailure of a hot leg RTD j be diagnosed, two methods are available f or addressing the failed RTD. l

' The preferred method is to utilize the second elecent of the RTD. Since both elements of each dual element RTD rre wired to the appropriate process protection rack. ILC personnel can disconnect the failed element from the rack terminal strip and connect the other RTD element. If the spare element is not available the second method la for the 1&C personnel to

- defeat the failed hot leg RTD and rescale the electronics to average the remaining two signals and incorporate a bias based upon the hot leg streaming measured in the loop. WCAP.13181 Appendix B provides the I calculational methodology for hot leg temperature bias values. Should a f ailure of a cold leg RTD be diagnosed, the 1&C personnel would disconnect the f ailed element f rom the rack terminal strip and connect the other RTD element.

4 The ef fect of the increased instrument uncertainty on updated Final Safety Analysin Report (UFSAR) Chapter 6 and 1$ 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, Yankae 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) Baron 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 slightly due to the increase uncertain associated with the modification. However, the long term and peak contairunent pressure are still valid and the ef fects 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 offect on the Steam Generator Tube Rupture analysis which was performed by them and submitted to the NRC on April 16 ~ 1991 in NHY letter NYN-91061. Yankee Atomic Electric Company also concluded that the modification will have negligible effect on the Baron Dilution analysis to be performed by them for Cycle 3. The YAEC evaluation of the affect 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.

10 1-

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

N P

. . . . j i

V. Potennt nat ionj,,$1cni ficant llarards for Liernse Amendment Renuent 9 7 - 0],

fLoltosed Channtag  ;

llew Hampshire Yankee has determined that License Amendment Request 92 01 does not involve a significant hazard consideration pursuant to the standards of 10CFR$0.92 based on the f ollowing evaluation.

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

i

, Westinghouse has prepared WCAP-13181 *RTD Bypass Elimination Licensing Report for Seabrook Huclear Station' (Proprietary) in support of the four ' Loop operation of Seabrook Station utilizing new thermovell mounted R7D's. For the Westinghouse scope. WCAP 13181 contains a saf ety ev61untion for this modified hot leg and cold leg temperature measurement system. This significant hazards evaluation ,

addresses both the mechanic 61 modif,' cations to the' reactor coolant I system preneure boundary and the instrumentation uncertainty changes  :

associated with the modified system. j The installation of thermowells and fast response RTDs will not inctense the probability of-an acciden previously-- analyzed.-- The -

modifications.to the Reactor Coolant System pressure boundary will be performed utilizing the same ASME Sect ton III installation requirements as were used for the original . Installation. The +

installation requirements ate specified in the ASHE Section III 1977 Edition thru Winter 1977 Addenda. .

The removal of the bypass piping and valves associated with this piping wi11' enhance the ht yrity of the Reactor Coolant System.

By removing significant lengths of piping, numerous valves and .

Instrument penetrations the probability of a small break LOCA will l 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 FA term cafety analysis limit _and. conservative changes to the K6 term to assure protection for_ all power ranges. Thi ' Overpower ' AT_-' ar.d Overtemperature AT functions 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 Vestinghouse 7300 series reactor protection system technology as has been previously qualified and used in the 3 reactor protection cystem of Seabrook Station. There-is no change in the use of the temperature. signals by any reactor. protection or reactor control-system.

The compliance of Seabrook Station to IEEE 279-1971, (IEEE Standard s

' Criteria for Protection Systems for Nuclear Power- Generating

• 11  !

l l

i 1

i Stations"). applicable NRC General Design Criteria and regulatory guides has not changed.

! This modification dnes not increase the radiological consequences of any accident previously evaluated. Although the pressure boundary will be modified. proper welding techniques, penetrant testing, radiographs, and system hydrostatic tests will insure the integrity of the pressure boundary and thus not contribute to any radiological consequences.

The proposed revisions to Technical Specification 3/4.2.5 (DNB parameters) for RCS flow from a value that includes measurement uncertainty to the analysis limit has no effect on the accident analyses since the analysis limit which is based on the thermal design flow will not be- changed. Appropriate measurement uncertainties for the method used to measure RCS flow, including ,

the effect of venturi fouling, have been determined. This )

uncertainty will be added to the RCS flow requirement of Technical Specification 3/4.2.5 to establish the acceptance criteria for the measured value of RCS flow. The acceptance criterie for the measured +

value of RCS flow will be specified in appropriate procedures. _

Surveillance Requirement 4.2.5.3 for the precision heat balance  !

determination of RCS flow is changed from'being required prior to {

operation above 752 Rated Thermal Power (RTP) to being required prior to exceeding 952 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 uncertainties that are exacerbated at lower power icvels. The precis)on heat balance is performed each cycle to detect changes in the RCS flov element (elbow taps) characteristics that would atfect ,

the accuracy of the RCS flow indication. Significant changes in the characteristics of al: of the elbow tape over a. single operational cycle is not (,cedible. . 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 taps.

The effect of the increased instrument uncertainty on updated Final '

Safety Analysis Report (UFSAR) Chapter 6 and 15 LOCA and non-BOCA accident analyses within the Westinghouse scope has been evaluated as discussed in WCAP-13181. Relative to both the LOCA and non-LOCA I

safety analyses. Westinghouse has concluded in WCAP-13181 that the modification does not affect the conclusions-of the UPSAR safety analyses, i 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.-

12

-,.-_, .a.- - - . - - - - _ . - - . _ . . . - , _ . - - . . . - - . - - - - . . - - - . . , . . . , . - , - . - . . , , , ,

e 4 . .

e Relative to containment response YAEC concluded that during the limiting event (large break LOCA). the early containment pressure response during the blowdem phase may increase slightly due to the increased uncertainties associated with the modification. However.

l the long term and peak containment pressures are still valid and the eff& cts of the mellfication on the containment response is bounded by the current analysis. The YAEC evaluation of the affect of the modificat!on a containment response is enclosed in Section VIII.

Yankee Atomic Electric Company has conclud i that the increase

- uncertainties associated with the modification will have a negligible effect on the Steam Generator Tube Rupture analysis which was performed by them arid submitted to the NRC on April 16, 1991 in NHY letter NYN.91061. Yankee Atomic Electric Company also concluded that the mndification will have negligible effect on the Boron Dilution analysit 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 perf ormed -i 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 Systein will not create the possibility of a new or dif ferent kind of accident from any accident previously

• valuated. 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 wH1 continue to be supplied. Other equipment important to safety will be unaff ected and will continue to function as designed.

The removal of the Resistance Temperature Detector (RTD) bypass

. piping and the installation of a modified temperature measurement i

system does not affect the integrity of the reactor coolant system '

pressure boundary. This is due to the reactor coolant piping g (pressure boundary component) modifications adhering to the ASME Code j (Sections III, Class 1 and Section XI) and to the NRC General Design >

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 vell as a major contributor to occupational radiation exposure. Additionally, with these components removed, the probability of a malfunction

f rom

-them is eliminated. The installation of fast response thermowell mounted RTDs onthe reactor coolant loop' piping and additional processing electronica vill continue to provide the individual loop-temperature rignals for input to the reactor control and reactor protection systems using components that are environmentally and seismically qualified.

-13 1

l l

t ww-ww , < w +

The RTD Bypass System flow alarm is no longer required to warn of flow reduction that could affect instrument system response. Flow through the scoop tubes with thermovells it 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 controls that prevent the introduction 1 of objects that could block the flow path.

The modification does not aff ect the ability of the protection system to mitigate the radiological consequences of any accidens. The new RTD signala 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 failed RTD channel remain the esme.

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 ST 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 meesurement uncertainty are also affected by the modification. 'The safety evaluations of this modification which have been perf ormed by Westinghouse and YAEC referenced above conclude that sufficient margin exists such that margins to safety are not affected.

The proposed Techni::a1 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 change to the RCS flow requirement to specify analysis values provides consistency ja this Technical Specification for DNB -

limits which currently spscifies analysis values for Tavg and pressurizer pressure. This change to an- analysis value for RCS flowrate does not affect any margin of safety.

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 controls 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.

14-

. 4 , ,

VI. Proposed Schedule for Liceqse Amendment Issuance and Effectiveness 11ew Hampshire Yankee requests ICtC review of License Amendment Request 92-01 and issuance of a license amendment by July 1. 1992 (see enclosed License Amendment Request 92-01.Section VI). This schedule is proposed in support of 14HY's plans to impicment the RTD Uypass System Elimination design change during the second refueling outage which is scheduled ' o begin in September 1992.

6 15

. . _ - ________.___m._.____... . . . _ _ . .

s .e .

VII. Icnviron;nent al 1:nnact Assessment New Hampshire Yankee (fil!Y) has reviewed the proposed licenso amendment against the criteria of 10CFR$1.22 for environmental considerations. The proposed changes do not involve a nignificant hazards consideration, nor increase the types and amounts of ef fluents that may be released of f site.

I nor significantly increase individual t'r cumulative occupational radiation u posures.- Based on the foregoing. 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.

1 l '

l I

t t

1 i

16 f

i

.....,,,,..,,.,,,...._._m,,...%_..-,_.,., , , . . , , , _ _ _ , , , _ . . , , , , , , , . , , _ , , , , ,_y . . ,_,,, , _ , , _ , . , , , , _ , , , , . , , . , , , . , , __

V111. Dt her Supportina inforinathtq Westinghouse Authorination 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 Ef fects of Removal of the RTD Bypass System on containment Response of Seabrook Station Evaluation of the Ef fects of Removal of the RTD Bypass System on YAEC-1698 and the Seabrook Station Boron _ Dilution Ana. lysis Notes (1) New: Hampshire Yankee notes that WCAP-13181 does not reflect NHY's proposed inclusion of the thermal design flow analysis value in Technical Specification 3.2.5c. VCAP-13181 specifies the RCS flow rate of 392.000 gpm which is associated with a revised RCS flow calorimetric uncertainty of 2.3r attributable to the new RTD hot leg ' and cold leg measurement system.

(2) New Hampshire Yankee notes that the deletion . of the Techni,r.

Specification for_ low RCS Tavg coincident with reactor trip f eedwate.

isolaties. Functional Unit (Table 3.3-4. Functional Unit-6.b)' recommended in WCAP-13181 will be addressed in a future license amendment request.

17

_ _ _ _ . _ . _