ML19351A363

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Proposed Tech Spec Tables 4.3-1 Reactor Protective Instrumentation Surveillance Requirements & 4.3-2, ESFAS Instrumentation Surveillance Requirements.
ML19351A363
Person / Time
Site: Waterford Entergy icon.png
Issue date: 10/16/1989
From:
LOUISIANA POWER & LIGHT CO.
To:
Shared Package
ML19351A361 List:
References
NUDOCS 8910200118
Download: ML19351A363 (16)
v • d • e


Text

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l ATTACHMENT A NPF-38-103 4.

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8910200118 691016 PDR ADOCK 05000382 P PDC wana 2

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e TABLE 4.3-1 *.

5 Q REACTOR PROTECTIVE INSTaipENTATION SURVEILLANCE REQUIMIENTS _

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. CNAIBIEL SWOES FOR 488I O OWWWEL CNAISEL FWICTIONAL SURVEILLAIICE j k-* FUNCT1000AL UNIT CIECR CALISAATION TEST 15 REQUIRES

" 1. Manuel Reacter Trip N.A. N.A. E and 5/U(1) 1, 2, 3*, 4*, 5*

2. Lineer Power Level - Nigh 5 D(2,4),00(3,4), M 1, 2

, Q(4)

3. Logarithmic Power Level - High 5 R(4) M and 5/U(1) 28, 3, 4, 5
4. Pressurizer Pressere - Nigh 5 R M 1, 2 j g 5. Pressurizer Pressure - Low 5 R M 1, 2 i

y 6. Containment Pressure - Nigh 5 R M 1, 2 U

7. Steam Generator Pressure - Low 5 R M 1, 2 l
8. Steam Generater Level - Low 5 R M 1, 2
9. Local Power Density - High 5 D(2,4),R(4,5) M. R(6) 1, 2
10. DISR - Low 5 5(7), D(2,4), M. R(6) 1, 2 M(8), R(4,5)

! 11. Steam Generator Level - Nigh 5 R M 1, 2 1 E l E 12. Reactor Protection System k Logic N.A. N.A. M and S/U(1) 1, 2, 3*, 4*, 5*

'i 5

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, fastt 4.3-1 (Centlemed) -

d REAC108 PSeiECTIM INSTaleENTAllSN SMRVEltLANCE REpipElENTS B

B

, OWueEL Osets FOR W1101 OtWeEL CIWW8IEL FINETISIIAL SURVEILLANCE i h

-4 FUNCTICIIAL IRIli CNECK CAL 10RATIOII TEST 15 AEGIIISE8 w 13. Reactor Trip Breakers N.A. N.A. II(14), $/U(1) 1, 2, 3*, 4*, 5*

14.- Core Protection Calculaters S O(2,4),R(4,5) II(9),R(6) ,

1, 2

15. CEA Calculators 5 R M,8(6) 1, 2
16. Reacter Coelant Flow - Low 5 R M 1, 2 Y. f

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

. l

.- . TA8LE 4.3 1 (C:ntinued) f TABLE NOTATION $ f "With the r6 actor trip breakers in the closed position, the CEA drive l system capable of CEA withdrawal, and fuel in the reactor vessel. l fThe ptovisions of Specificaties 4.0.4 are not applicable when reducing l reactor power to less than }91 of RATED THERML POWER from a reactor i power level grgater than 10 1 of RATED THERM L POWER. Upon reducing j power below 101 of RATED THERML POWER, a CHANNEL FUNCTIONAL TEST shall  ;

be performed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if not perfomed during the previous 31 days.  !

This requirement does not apply with the reactor trip breakers open. l (1) Each startup or when required with the reactor trip breakers closed  !

and the CEA drive systse capable of rod withdrawal, if not performed l in the previous 7 days. j (2) Heat balance only (CHANNEL FUNCTIONAL TEST not included), above 15%  ;

of RATED THERML POWER: adjust the Linear Power Level signals and the i CPC addressable constant multipliers to make the CPC AT power and CPC  !

nuclear power calculations agree with the calorimetric calculation if  ;

absolute difference is greater than 25. During PHYSICS TESTS, these i daily calibrations may be suspended provided these calibrations are j performed upon reaching each major test power plateau and prior to  !

proceeding to the next major test power plateau. l (3) ' Above 155 of RATED THERML POWER, verify that the linear power sub-  !

channel gains of the excore detectors are consistent with the values i used to establish the shape annealing matrix elements in the Core l Protection Calculators. l

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

{

(5) After each fuel loading and prior to exceeding 705 of RATED THER M L i POWER, the incore detectors shall be used to determine the shape i annealing matrix elements and the Core Protection Calculators shall i use these elements, j i (6) This CHANNEL FUNCTIONAL TEST shall include the injection of simulated  !

process signals into the channel as close to the sensors as practicable i to verify 0PERA41LITY including alam and/or trip functions. t (7) Above 705 of RATED THERMAL POWER, verify that the total RCS flow rate  !

as indicated by each CPC is less than or equal to the actual RCS total  ;

flow rate determined by either using the reactor coolant pump differential pressure instrumentation or by calorimetric calculations and if necessary, '

adjust the CPC addressable constant flow coefficients such that each  ;

CPC indicated flow is less than or equal to the actual flow rate. The  ;

flow measurement uncertainty is included in the SERR1 tem in the CPC and is equal to ar greater than 45. i (4) Above 705 of RATED THER E L POWER, verify that the total RCS flow rate -

as indicated by each CPC is less than or equal to the actual RCS total  :

flow rate detemined by calorimetric calculations, j (9) The monthly CHANNEL FUNCTIONAL TEST shall include verification that ,

the correct valuts of addressable constants are installed in each OPERABLE CPC. j (10) At least once per 1G months and following maintenance or adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent verification of the undervoltage trip function and the shunt trip function.

WATERFORD - UNIT 3 3/4 3-12 AMENDMENT NO.40 I . - _ _ _ _ , _ _ _ _ . _ . . _ , _ , . _ . . _ _ _ _ _ , _ . . _ _ . _ , _ _ _ _ _ _ _ . . _ _ _ _ _ _

z -

TA8tE 4.3-2 E ~

D ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTAION SURVEILLANCC REQUIRDENTS O

3 CHANNEL IWDES FOR MtICet E CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE e FUNCTIONAL UNIT CHECK CALIBRATION TEST 15 REQUIRED C

5

1. SAFETY INJECTION (SIAS)
a. ~ Manual (Trip Buttons) N.A. N.A. R 1, 2, 3, 4
b. Contafruneet Pressure - High 5 R M 1, 2, 3
c. Pressurizer Pressure - Low 5 R M 1, 2, 3
d. Automatic Actuation Logic M.A. N.A. M(1) (2) (3) 1, 2, 3
2. CONTAINNENT SPRAY (CSAS)
a. Manual (Trip Buttons) N.A. M.A. R , 1, 2, 3, 4
b. Containment Pressure --

High - High 5 R M 1, 2, 3

c. Automatic Actuation Logic M.A. M.A. M(1) (2) (3) 1, 2, 3 Y
3. CONTAINNENT ISOLATION (CIAS)

Y a. Manual CIAS (Trip Buttons) N.A. M.A. R 1, 2, 3, 4

7. b. Containment Pressure - High 5 R M 1, 2, 3
c. Pressurizer Pressure - Low 5 !I M 1,2,3
d. Automatic Actuation Logic N.A. M.A. M(1) (2) (3) 1, 2, 3
4. MAIN STEAM LINE ISOLATION
a. Manual (Trip Buttons) N.A. N.A. R 1, 2, 3
b. Steam Generator Pressure - Low 5- R M 1, 2, 3
c. Containment Pressure - High 5 R M 1, 2, 3
d. Automatic Actuation Logic M.A. N.A. M(1) (2) (3) 1, 2, 3

- - r _ _ _ _ - _ _ _ _,.______..___________-_____a_m__.______._____.___---___._m_

_____m __ _c__-.m_____________m___._ _m-_--___ --_m-_-----_-_.--____ .-- -- .s

l- 5 -

- 3:

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TABLE 4.3-2 (Continued) .

g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIMIENTS -

4 g CHANNEL S W ES FOR 4081CM o

CHAfMIEL CHAlmEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIED E

= 5. SAFETY INJECTION SYSTEN

, RECIRCULATION (RAS)

a. Manuel RAS (Trip Buttons) N.A. N.A. R 1, 2, 3, 4
b. Refueling Water Storage -

Pool - Low 5 R M 1,2,3,4

c. Automatic Actuation Logic M.A. N.A. It(1) (2) (3) 1, 2, 3, 4
6. LOSS OF POWER (LOV) i a. 4.16 kV Emergency Bus .

Undervoltage (Loss of

, Voltage) N.A. A D(4) 1, 2, 3

} b. 480 V Emergency 8vs

, Undervoltage (Loss of

! 4 Voltage) N.A. R G(4) 1, 2, 3

e c. 4.16 kV Emergency Bus
  • i Undervoltage (Degraded Voltage) N.A. R D(4) 1, 2, 3 G.

-_______._m_- _-__ __.__ ._____.__. m____ -.u-__ m_ _ _ _ _ _ _ _ - _u_.__-_ _ h- - -u+_zsau--- - - ---.- NT _ . - +__.-____-u__---_m.C __-_wmm_-.a-_ _ A_m__wn.w_m__- - - - _ -u_m-- - --*_mmm et---u- - -

s. . . . - .

4 #

TA8tE 4.3.-2 (Continued)

  • h
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRINGENTATION SURVEILLANCE REQ
g g CaesEL MODES FOR WNICM CHANNEL CHANNEL FUNCTIONAL

, FUNCTIONAL UNIT SURVEILLANCE CHECK CALIBRATION TEST _IS HEQUIRED

-4

7. EMERGENCY FEEDWATER (EFAS)
a. Manual (Trip Buttons) M.A.

w b. N.A. R 1, 2, 3 SG Level (1/2)-Low and AP (1/2) - High 5 R M '

c. SG Level (1/2) - Low and No 1, 2, 3 Pressure - Low Trip (1/2) S R M
d. Automatic Actuation Logic M.A.

1, 2, 3

e. N.A. M(1) (2) (3) 1, 2, 3
  • Control Valve Logic 5 R (Wide Range SG Level - Low) SA(5) 1, 2, 3 em Y

O TA8tE NOTATION (1) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(2)

Testing of Automatic Actuation Logic shall include energization/deenergization of each initiation relay and verification of the OPERA 8ILITY of each initiation relay.

(3)

A subgroup relay test shall be performed which shall include the energitation/deenergization of each subgroup relay and verification of the OPERASILITY of each suegroup relay. Relays K109, K114. K202, K301, K305, K308 and K313 are exempt from testing during power operation but shall be tested at least once per 18 months and during each COLD SHU1DOWN condition unless tested within the previous 62 days.

(4) Using installed test switches.

(5) To be performed during each C0iD SIRIIDOWN if not performed in the previous 6 menths.

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c, 3/4.3 INSTRUMENTATION l

BASES 3/4.3.5 and 3/4.3.3 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUAT;:0N SYSTEM 5 ;:NSTRUMENTATION The OPERASILITY of the Reactor Protective and Engineered Safety Features Actuation Systems instrumentation and bypasses ensures that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated i when the parameter monitored by each channel or combination thereof reaches l its setpoint, (2) the specified coincidence logic is maintained (3) sufficient j redundancy is maintahwd to permit a channel to be out of service for testing or mairtenance, and (4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity sssumed available in the facility design for the protection and mitigation of accident and transient conditions. The j integrated operation of each of these systems is consistent with the assumptions 1 vsed in the safety analyses, j The redundancy design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEACs become inoperable. I If one CEAC is in test or inoperable, serification of CEA position is performed i 1 at leest every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the second CEAC fails, the CPCs will use ONBR and j LPD peralty factors to restrict reactor operation to some maximum fraction of l RATED ThKRMAL POWER. If this maximum fraction is exceeded, a reactor trip '

will occuc.

The Surveillance Requirements specified for these systems ensure that the overall systte functional capability is maintained comparable to the original l design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to vemonstrate this capability.

The measurement of response time at the specified frequenciec provides I assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the safety analyses.

No credit was taken its the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, L overlapping, or total chant,el test measurements provided that such tests demonstrate the total channe) response time as defined. Sensor response time verification may be demonstrated by either (1) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTAT!ON i The OPERABILITY of the radiation monitoring channels ensures that:

l (1) the radiation levels are continually measured in the areas served by the .

1 WATERFORD - UNIT 3 8 3/4 3-1 J

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W ATTACHMENT B 3

NPF-38-103 r

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4 a- TMLE 4.3-1 ,,

D REACTOR PROTECTIVE INSTRISENTATION SURVEILLANCE REqWIWENTS -

i Q

  • o 8 OWWRIEL IR10E5 FOR %dNICN 1 CMMcEL CNAleIEL FIRICTIONAL SURVEILLAIICE

! OIECK CALIceATIGII TEST 15 IIEQIflAES

)

  • FIRICTIONAL INIIT i

Messel Reector Trip N.A. N.A. R and S/U(1) 1, 2, 3*, 4*, 5*

  • ** 1.
2. Lineer Power Level - High 5 D(2,4),9t(3,4), h
  • 1, 2 j Q(4) i j 3. Logarithmic Power Level - High 5 R(4) N and S/U(1) 29, 3, 4, 5 i

5 R M. 1, 2 j 4. Pressurizer Pressere - Nip 5 .M 1, 2 l t' 5. Pressuriner Pressure - Low R

  • l /

Containment. Pressure - Nigh 5 1, 2 j y 6. R Nl I

l o N 1, 2

7. Steam Generator Pressere - Low 5 R i l i s. Steen Genereter Level - Lew 5 R N 1, 2 1 \
9. Local Power Bensity - Nigh 5 D(2,4),R(4,5) M, R(6) 1, 2

, 10. DSR - Law 5 5(7),D(2,4), N R(6) 1, 2 M(8),R(4,5) l

11. Steam Generator Level - High 5 A N, 1, 2 E 12. Reacter Protection System /

3*, 4*, 5*

k Logic M.A. N.A. M>and 5/U(1) 1, 2,

=

g as u u.

. g .. m -

m O w gy n "Q"

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

~ m

~

,. TASLE 4.3-1 (Continued)

-5

.g REACTOR P901ECTIVE INSTE M NTATION SURVEILLANCE RE W IREM Nis i o l 8 peOES FOR 446tCN

! , ORueIEL I CHANNEL CHAleIEL FUNCTIONAL SURVEILLANCE j $

FUNCTICIIAL UNIT CHECK CALIBRATI0ff TEST 15 REQUIRED l

w 13. Reacter Trip Breakers M.A. N.A. M(10),S/U(1) 1, 2, 3*, 4*, 5*

14.- Core Protection Calculators S 3(2,4),R(4,5) M(9),R(6) 1, 2

15. CEA Calculators S R M R(6) 1, 2
16. Reacter Coolant Flow - Low 5 R M\ 1, 2 l

4 k Rchc_

w nw

  • m a

) w .id *CL" Y

M 4

l i

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.. , TABLE 4.3 1 (Continued)

TABLE NOTATIONS

system capable of CEA withdrawal, and fuel in the reactor vessel, j 4.0.4 are not applicable when reducing I

  1. The provisions reactor power to ofless Specificatipg%

than }9 of RATED THERMAL POWER from a reactor power level grgater than 10 'X of RATED THERMAL POWER. Upon reducing power below 10 1 of RATED THERMAL POWER, a CHANNEL FUNCTIONAL TEST shall be performed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if not perforised during the previous 31 days, i This requirement does not apply with the reactor trip breakers open. {

l (1) Each startup or when required with the reactor trip breakers closed i and the CEA drive system capable of rod withdrawal, if not performed  !

in the previous 7 days.  !

(2) Heat balance only (CHANNEL FUNCTIONAL TEST not included), above 15%

of RATED THERMAL POWER: adjust the Linear Power Level signals and the ,

CPC addressable constant multipliers to make the CPC AT power and CPC l nuclear power calculations agree with the calorimetric calculation if 1 absolute difference is greater than 25. During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

(3) 'Above 15% of RATED THERMAL POWER, verify that the linear power sub-channel gains of the excore detectors are consistent with the values used to establish the shape annealing matrix elements in the Core Protection Calculators.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) After each fuel loading and prior to exceeding 705 of RATED THERMAL POWER, the incore detectors shall be used to determine the shape annealing matrix elements and the Core Protection Calculators shall use these elements.

(6) This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to the sensors as practicable to verify 0PERA81LITY including alarm and/or trip functions.

(7) Above 705 of RATED THERMAL POWER, verify that the total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate detemined by either using the reactor coolant pump differential pressure instrumentation or by calorimetric calculations and if necessary, adjust the CPC addressable constant flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate. The flow measurement uncertainty is included in the SERR1 term in the CPC and is equal to or greater than 4%.

(8) Above 70K of RATED THERMAL POWER, verify that the total RCS flow rate

'N as indicated by each CPC is less than or equal to the actual RCS total P" 1N flow rate determined by calorimetric calculations.

g.jN) TheQ, CHANNEL FUNCTIONAL TEST shall include verification that the corW et values of addressable constants are installed in each OPERA 8LE CPC.

(10) At least once per 18 months and following maintenance or adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent verification of the undervoltage trip function and the shunt trip function.

WATERFORD - UNIT 3 3/4 3-12 AMEN 0 MENT NO.40 l-

~

. . .. -. 2

-TABLE 4.3-2 r ,

3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTAION SURVEILLANCE REQUIREMENTS 9

5 CHANNEL NODES FOR WHICH E CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE e FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED c

25 1. SAFETY INJECTION (SIAS) -

I

a. k nual (Trip Buttons) N.A. N.A. R 'INM C- 1,2,3,4
b. Containment Pressure - High
c. Pressurizer Pressure - Low 5

5 R

R

{ 1, 2, 3 1,2,3

d. Automatic Actuation Logic M.A. N.A. M(1) (2) (3) 1, 2, 3
2. CONTAIl99ENT SPRAY (CSAS)
a. k nual (Trip Buttons) N.A. N.A. R * *'" 1, 2, 3, 4
b. Containment Pressure -- ""3 ' A High - High 5 R e 1, 2, 3
c. Automatic Actuation Logic M.A. N.A. M(1) (2) (3) 1, 2, 3
3. CONTAINNENT ISOLATION (CIAS) gaec t.

Y a. knual CIAS (Trip Buttons) N.A. N.A. R / w y- 1,2,3,4 0 h. Containment Pressure - High 5 R w .m o" 1, 2, 3

c. Pressurizer Pressure - Low 5 R 1, 2, 3
d. Automatic Actuation Logic N.A. M.A. M(1) (2) (3) 1, 2, 3
4. MAIN STEAM LINE ISOLATION
a. m nual (Trip Buttons) N.A. N.A. -R RNE 1,2,3
b. Steam Generator Pressure - Low 5- R M 6"" ' # 1, 2, 3
c. Containment Pressure - High 5 R 1, 2, 3 1
d. Auton,atic Actaation Logic M.A. N.A. N(1) (2) (3) 1, 2, 3 4

l 9, . , . , , , ,,,, ,, , , ,

, . , ,, ,,w, , ,ec. ,,n , . , a, , .,,.g, ,,., , ,n,_a- ,,

m -

's 1 -

-c. .q -

. x ,.

TA8tE 4.3-2 (Continued) 5 y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILtANCE REQUIRDENTS 4

g CHANNEL NDDES FOR %dNICM o CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST - IS REQUIRED 4

E q 5. SAFETY INJECTION SYSTEM

, RECIRCULATION (RAS) nectie,
a. Manual RAS (Trip Buttons) N.A. N.A. R 1, 2, 3, 4
b. , q Refueling Water Storage j Pool - tow 5 R F 1, 2, 3, 4
c. Automatic Actuation Logic M. A. M.A. M(1) (2) (3) 1, 2, 3, 4
6. LOSS OF POWER (LOV)
a. 4.16 kV Emergency Bus ,

Undervoltage (Loss of l , Voltage) N. A. R D(4) 1, 2, 3

} b. 480 V Emergency Bus

, Undervoltage (Loss of 4 Voltage) M . I.. R D(4) 1, 2, 3 os c. 4.16 kV Emergency Bus -

Undervoltage (Oegraded Voltage) N.A. k D(4) 1, 2, 3 t

m__.m_ . _ -. _ '-m_E _.d

- .i-, . .4n,%,- * , + . . - _

,--_e-m__._a-_mm__ __- _ _ e__-___ ____m, m___mm ___mm_. ~___ . = - m _ . . ___.__ _ _ . _ _ _ _ . _ _ _

_ n 7

. .- - . . ,~2- '.

  • . 4 [

TA8tE 4.3.-2 (Continued) -

k ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMF E

c, g CHS.NNEL CHANNEL CHAMMEL MDDES FOR WHICH FUivCTIONAL SURVEILLANCE

, FUNCTIONAL UNIT CHECK CAllBRATION TEST .IS REQUIRED H

7. EMERGENCY FEEDWATER (EFAS)
a. Manual (Trip Buttons) N.A.

w N.A. R 1, 2, 3

b. SG Level (1/2)-Low R c "E i

and AP (1/2) - High 5 R Y G >" " '

c. 1, 2, 3 SG Level (1/2) - Low and No """T~

Pressure - Low Trip (1/2)

d. Automatic Actuation Logic S

N.A.

R N.A.

W 1, 2, 3

e. Control Vahe Logic 5 M(1) (2) (3) 1, 2, 3 R SA(5) 1, 2, 3 (Wide Range SG Level - Low) b u, .

O TABLE NOTATION (1) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(2)

Testing of Automatic Actuation Logic shall include energization/deenergization of each initiation relay and verification of the OPERASILITY of each initiation relay.

(3)

A subgroup relay test shall be performed which shall include the energization/deenergization of each subgroup relay and verification of the OPERABILITY of each subgroup relay. Relays K109. K114. K202, K301, K305, K308 and K313 are except from testing during power operation but shall be tested at least once per 18 months and during each COLD SHUTDOWN condition unless tested within the previous 62 days. i (4) Using installed test switches.

(5) To be performed during each COLD SHUTDOWN if not performed in the previous 6 months.

i

d j

U - . 1 .,  !

i i g . @b ;. Th2 q crterly fr:quzncy fcr tha ch:nnst furstisnal  ;

7 , f. , tests fcr thess cystens comes fro ths ancly:ss presented in topical report CEN-327: RPS/ESFAS Extended Test Interval Evaluation, as supplemented. _

[

].

3/4.3 INSTRUMENTATION' BASES 3/4. 3.1 and 3/4. 3. 2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES 3 ICTUATION SYSTEMS INSTRUMENTATION t

. The Or'ERASILITY of the Reactor Protective and Engineered Safety Features Actuation Systems instrumentation and bypasses ensures that (1) the associated l Engineered Safety Features Actuation action and/or reactor trip will be initiated .

when the parameter monitored by etch channel or combination tnereof reaches  !

y its setpoint, (2) the specified coincidence logic is maintained. (3) sufficient redundancy is maintained to permit a channel' to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall ,

reliability, redundancy, and diversity assumed available in the facility design '

for the protection and mitigation of accident and transient conditions. The . 1 integrated operation of each of these systems is consistent with the assumptions  !

p used in the safety analyses.  ;

. The redundancy design of the Control Elcment Assembly Calculators (CEAC) l provides reactor prctection in the event one or both CEACs become inoperable. I If one CEAC is in test or inoperable, verification of CEA position is performed 4 l

i at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the second CEAC fails, the CPCs will use DNBR and l- LPD penalty factors to restrict reactor operation to some maximum fraction of L RATED THERMAL POWER., If this maximum fraction is exceeded, a' reactor trip

(, will occur.

The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original ,

design standards. The periodic surveillance tests performed at the minimum  !

frequencies are sufficient to demonstrate this capability. / )

l The measurement of response time at the specified frequencies provides  !

assurance that the protective and ESF action function associated with each j channel is completed within the time limit assumed in the safety analyses. I do credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, l overlapping, or total channel test measurements provided that such tests I demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INS'RUMENTATION L 3/4.3.3.1 RADIATION MONITORINC INSTRUMENTATION The OPERABILITY of the radiation monitoricg channels ensures that:

(1) the radiation levels are continually measured in the areas served by the WATERFORD - UNIT 3 8 3/4 3-1

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