ML20024G773

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Tables 3.3-1,3.3-2,4.3-2 & Bases 3/4.3.1 & 3/4.3.2 Re Engineered Safety & Reactor Trip Functions
ML20024G773
Person / Time
Site: Comanche Peak Luminant icon.png
Issue date: 04/26/1991
From:
TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20024G772 List:
References
NUDOCS 9104300175
Download: ML20024G773 (32)


Text

_ _ _ _ _

Attachment 2 to TXX 91084 page 1 of 10 CPSES UNIT 1 NRC DOCKET N0. 50-445 LICENSE AMENDHENT REQUEST 91-003 TABLE OF CONTENTS

1. SIGNIFICANT HAZARDS CONSIDERATION
2. PROPOSED INSTRUCTIONS FOR INCORPORATION 9104300175 910426 PDR ADOCK 05000445 P PDR

s t

Attachment 2 to TXX 91084 page 2 of 10 Significant Hazards Consideration Propcsed Unit 1 Technical Specification Change Various RPS and ESF Action and Surveillance Requirements

1. DESCRIPTION This change proposes to modify the Comanche Peak Steam Elect,ric Station Unit 1 Technical Specifications by relaxing the Allowed Outage Times (A0T) and the Surveillance Test Intervals (STI) for Analog Channels shared by both the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS).
11. BACKGROUND in response to the high number of reactor trips and Engineered Safety Features Actuation System (ESFAS) actuations experienced by plants as a result of instrument testing and surveillance activities, the Westinghouse Owners Group (WOG) initiated a program to develop generic justification for revising generic and plant specific instrumentation technical specifications to reduce the test and maintenance requirements.

As a result of the WOG submittals, the NRC, in the following correspondence, approved almost all the relaxations suggested by the WOG.

  • Letter C. O. Thomas (NRC) to J. J. Shepperd (WOG) dated February 21,1985 (NRC Saf ety Evaluation f or WCAP-10271)
  • Letter Charles E. Rossi (NRC) to Roger A. Newton (WOG) dated February 22,1989 (HRC Safety Evaluation for WCAP-10271 Supplement 2 and Supplement 2 Revision 1)
  • Letter Charles E. Rossi (NRC) to Gerald T. Goering (WOG) dated April 30,1990 (NRC Supplemental Safety Evaluation for WCAP-10271 Supplement 2 and Supplement 2 Revision 1)

CPSES incorporated most of the suggested STI and A0T relaxations in the original Technical Specifications issued on February 1990; however, in the months prior to license issuance and during Technical Specification development, all review and acceptance of the Westinghouse Owners Group submittal of WCAP-10271 (and supplements) was not completed. As a result, most, but not all provisions of WCAP-10271 (and supplements) were incorporated. This change proposes to incorporate the remainder of the changes. This request is being made under the generic justification provided by Westinghouse Electric Corporation during approval of WCAP-10271 Supplement 2, and Supplement 2 Revision 1.

i

. Attachment-2 to TXX 91084-page 3 of 10 III. JUST1FICATION An increase in the allowed outage time for maintenance will allow better more deliberate testing and repair, tnus reducing the potential for human error- and ' reducing vulnerability of CPSES to the high trip rate experienced by other operating plants.

The changes A,- B, C, and D, listed in Part V of this enclosure, result directly from the completion of a Westinghouse Owners Group (WO6) evaluation of Surveillance Test Intervals (STI) and Allowed ,0utage Times (A0T) and their effect on nuclear safety. Changes E and F listed in Part V of this enclosure result from a supplemental study of the Reactor Water Storage Tank (RWST) level unavailability performed specifically for CPSES (WCAP-10271 Supplement 3), WCAP 10271 Supplement 3 was transmitted to the NRC from TV Electric via letter logged TXX-91069 and dated March 5, 1991.

Several administrative changes -are proposed, as listed below:

1. Update the BASES section at page B 3/4 3-1 to reflect the NRCs supplemental safety evaluation of WCAP-10271 and its supplements, 2, Reword action statements requiring the plant to be in HOT STANDBY in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to allow 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to restore the inoperable channel, then 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be in HOT STANDBY (same total length of time).

This will avoid the incorrect perception of being in a shutdown action statement which has reportability and emergency

. classification connotations.

3. Renumber action statements to prevent human error. This change avoids two
  • Action 12's" and. renumbers th( Action Statements of

' Table 3.3 2.

l- 4. Delete reference to STARTUP and/or POWER OPERxTION-(HODES 2 and/or H

1) in Action'17 and 23 and replace with " Operation *, similar to

-the wording used in Action:14 The LCO to which these Actions apply include Applicable Hodes beyond H0 DES 1 'and P (1, 2; 1, 2, 3: and 1, 2, 3, 4). This change avoids confusion wnen the plant is in H0 DES 3 or 4 and the Action is entered.

l S. Deletes note 'e' of table 3.3 1. The note is no longer applicable since as a result of the WCAP 10271 and its supplements the A0T

, instruments common to RPS and ESFt.S de not have differing RPS and L ESF requirements.

l:

! )

l h

t

Attachment 2-to TXX*91084 page'4 of 10-IV.. SAFETY EVALUATION in WCAP 10271 including supplements 1, and 2, the WOG evaluated the impact of the proposed STI and A0T changes on core damage frequency and public_ risk. The NRC staff concludes in its evaluation of the WOG submittal that an overall upper bound increase of the-core damage frequency due to the proposed STl/A0T changes is less than 6 percent for )

Westinghouse Pressurized Water Reactor (PWR) plants. The NRC Staff also concluded that actual core damage frequency increases for individual plants are expected to be substantially less than 6 percent., The NRC l Staff considered this core damage frequency increase to be small compared to the range of uncertainty in the core damage frequency ]

. analyses and therefore acceptable.

In WCAP-10271 Supplement 3 the Reactor Water Storage Tank (RWST) level channels-were evaluated and found to be identical in configuration to the Steam Generator level channel and therefore the RWST level channels would be bounded by the unavailability analysis for the Steam Generator level channel.

The proposed changes are consistent with the NRC Staff's letters dated February 22,-1989. and April 30, 1990, to the WOG regarding evaluation of WCAP-10271, WCAP-10271 Supplement 1 WCAP 10271 Supplement 2 and

. WCAP-10271 Supplement 2, Revision 1. The Staff has stated that approval-of these changes at a particular plant will be contingent upon confirmation that certain conditions are met. CPSES compliance with these conditions is provided below:

1. Common cause evaluation of clant orotection system failures.

CPSES has previously implemented quarterly surveillance intervals except for RWST level. Quarterly Surveillance of the RWST level is included in this proposed change to the technical specifications. Plant programs and procedures are in place and being used to evaluate failures, trend failures and perform-common mode failure evaluation when necessary. Evaluation and reporting under the Nuclear Plant-Reliability Data System (NPRDS) are included in these programs.

2. Testina of analoo-channels'in bvoass. Testing of analog channels, described in FSAR 7.2.2.2.3 and 7.3.2.2.S. is done with the channels in the trip condition except for containment spray actuation which is an energize to actuate' channel and therefore designed to be tested in bypass. Jumpers and lifted leads are not4

, used.for this testing.

I '

Attachment 2 to TXX-91084 l page 5 of 10

3. Setooint drift. CPSES implemented quarterly analog testing upon receipt of the operating license in February 1990, except for RWST level as noted above. The setpoint methodology contains adequate allowance to bound anticipated drift over a three month period.

Additionally, setpoint drift data has been trended since prior to licensing to confirm this allewance. No excessive drift has been noted over this period.

4. Anolicability of Generic Analysis to CPS 11 CPSES is a 4 loop Nestingnouse PWR with a Solid State Protection System, As described in HCAP-10271 and its supplements, all chang'es proposed in this amendment are addressed by the generic analysis except for RWST level, RWST level was separately evaluated on a plant specific basis in WCAP-10271 Supplement 3. This amplified the analysis presented in the other supplements to encompass RWST level. This analysis concluded that RWST level is identical in configuration to the steam generator level channel; therefore, the system unavailabilities resulting from relaxed STis and A0Ts were essentially the same and will have no imnact on plant safety.

V. DETAILED DISCUSSION The requested amendment revises several aspects of the Reactor Protection System (RPS) and Engineered Safety Features (ESF) of Technical Specifications 3.3.1 and 3.3.2. These changes include the following:

A. Added new Action 13 (Table 3.3 1) applicable to safety injection input f rom ESFAS and automatic trip and interlock logic allowing for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> maintenance A0T and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> surveillance A0T.

B, Changed Actions 12 (Table 3.3-1) and Actions 19, 22 and 26 (original Action 12) (Table 3.3-2) to increase the Allowed Outage Time (A0T) for Surveillance Test to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. Replaced original Action 13 with Action 17 for 3 channel systems (Taole 3.3 2', to allow the-same provis;ons as 4 channel systems.

D. Revised note 'e' of Table 3.3 2 to make the time allowed to place the Steam Generator High Level channel into trip and the Surveillance A0T consistent with the provisions for inoperable channels (Action 17).

E. Revised test frequency for the RWST Low Low level to Quarterly (Table 4.3-2).

F. Revised Action for RWST Low Low Level from 26 to 17 (Table 3.3-2).

i

1 Attachment 2 to TXX 91084 page 6 of 10 m

i G. Administrative changes as follows:

1. Updated BASES reference to include the latest SERs.
2. Actions 13 (new) (Table 3.3 1), and Actions 19, 22, and 26 (new) (Table 3.3 2) have been revised to break the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shutdown time requirement into the more standard 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to restore or be it, at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3. Deleted Actions 13 and 26 (Table 3.3-2) that were no longer used, and re-numbered original ESF Action 12 to new Action
26. This avoids the two " Action 12's" that previously existed. Renumbered Actions on Table 3.3-2 accordingly.
4. Revised Actions 17 and 23 to delete specific reference to STARTUP and/or POWER OPERATION (MODES 1 and/or 2) and apply broader term of " operation'.
5. Deleted Note 'e' on Table 3.3-1 for Pressurizer Pressure Low and Steam Generator Water Level Low Low.

VI. PRECEDEtiTS The changes proposed by this License Amendment Request (with the exception of the RWST changes) have been accepted by the NRC in the correspondence listed in part 11 above. The proposed changes to the RWST, although not specifically accepted, have been evaluated and found to be identical and thus bounded by the Steam Generator Level unavailability analysis which has Lsen accepted by the NRC. Other changes also accepted in that listed correspondence were incorporated into the CPSES Technical Specifications prior to their issuance in February, 1990.

Vll. NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATIOL' PER 10CFR50.92 The standards used to arrive at a proposed determination that the changes described involve no significant hazards consideration are included in 10CFR50.92. The regulations state that if operation of the facility in accordance with the proposed amendment would not; (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibil?ty of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety then a no significant hazards consideration determination can be made.

CPSES has reviewed the requirements of 10CFR50.92 as they relate to the proposed RPS and ESFAS Technical Specification changes for CPSES and determined that a significant hazards consideration is not involved. In i support of this conclusion.the following analyses are provided.

Attachment 2 to TXX-91084 page 7 of 10 Criterion 1 -

The determination that the results of the proposed changes are within all acceptable criteria was established in the SER(s) prepared for WCAP 10271 Supplement 2 and WCAP 10271 Supplement 2. Revision 1 issued by letters dated February 22, 1989 and April 30, 1990, implementation of the proposed changes is expected to result in an acceptable increase in total Reactor Protection System and Engineered Safety features Actuation System unavailability. This increase results in a small increase in core damage frequency (CDF) and public health rist. The values determined by the WOG and presented in the WCAp for the increase in CDF were verified by Broo6 haven National Laboratory (BNL) as part of an audit and sensitivity analyses for the NRC Staff. Based on the small value for the increase compared to the range of uncertainty in the CDF, the increase is considered acceptable. The extension of the WOG relaxations to the RWST level has been separately shown to be bounded by the increased CDF resulting from relaxation of the Steam Generator level channel and therefore should be acceptable on the same basis.

The proposed changes do not result in an increase in the severity or consequences of an accident previously evoluted. Implementation of the proposed changes affe:ts the probability of failure of the RPS or ESF but does not alter the manner in which protection is afforded nor the manner in which limiting criteria are established.

Operation of CPSES in accordance with the proposed license amer.dment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2 The proposed changes do not involve hardware changes and do not result in a change in the manner in which the Protection System provides plant protection. No change is being made which alters the functioning of the protection System. Rather the likelihood or probability of the protection System functioning properly is affected as described above.

1herefore the proposed changes do not create the pos'sibility of a new or different kind of accident.

Attachment 2 to TXX 91084 page 8 of 10 Crite 'on 3 -

The proposed changes do not alter the manner in which safety limits, limiting safety system setpoints or limiting conditions for operation are determined. The impact of reduced testing other than addressed above is to allow a longer time interval over which instrument  :

uncertainties (e.g., drift) may act. Experience has shown that the initial uncertainty assumptions are valid for reduced testing.

l implementation of the proposed changes is expected to result, in an I overall improvement in safety due tot

a. Less f requent testing which will result in fewer inadvertent )

reactor trips and actuations of the Engineered Safety j Features Actuation System components,

b. Improvements in the effectiveness of the operating staff in monitoring and controlling plant operation. This is aue to l less frequent distraction of the operator and shift '

supervisor to attend to instrumentation testing.

The foregoing analysis demonstrates that the proposed amendment to CPSES technical 1 specifications does not involve a significant increase in the probability or consequences of a previously evaluated accident, does not ,

create the possibility of a new or different Lind of accident and does '

not involve a significant reduction in a margin of safety.

Vill. NO SIGNIFICANT HAZARDS CONSIDERATION DETERHINATION The Commission-has provided guidance concerning the application of the standards for determining whether a.significant hazards. consideration l

exists by providing certain examples (51 FR 7751) of amendments that are considered not- likely to involve significant hazards' consideration.

  • Example-(i) relates to a purely administrative change to Technical Specifications: for example, a-change to' achieve consistency throughout

~the Technical Specifications, correction of an error, or a change in nomenclature. Example (vi) relates to a change which either may result in some increase to the probability.or consequences of a  !

c previously analyzed accident or may reduce in some way a safety margin.-

but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in-the Standard Review Plans-for example, a change resulting from the application of~ a small refinement of a previously used calculational model or design method.

i l

l f'

\

l l'

O - - ,-,,,

J,.,,-, ,- . - . . ._._a.

u L '..

l e

Attachment 2 to TXX 91084 page 9 of 10 in this case, the change request described above is similar to Example ,

51) in that it is partially an administrative change to achieve i consistency throughout the Technical Specifications.

Additionally it is similar to Example (vi) in that portions may result )

in some increase to the probability of a previously analyzed accident, but the increase is not significant compared to the range of uncertainty of the analysis and therefore is considered acceptable. Based upon the preceding analysis, CPSES concludes that the proposed amendment does not involve a significant hazards consideration. ,

1X. ENV,lRONMENTAL EVALVATION 10 Electric has evaluated the proposed changes and has determined that

[

the changes do not involve (i') a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significent increase-in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10CFR$1.22(c)(9), Therefore, pursuant-to 10CFR51.22(b), on environmental assessment of the proposed 1 changes is not required. l 1

X. REFERENCES

1. CPSES Unit 1 Technical Specifications
2. WCAp 10271 Supplement 1, Supplement 2, supplement 2 Revision 1 I and Supplement 3.

4 i

. , um 9 p.- - ..,,w-o ,w-e-.. + w -- e + . - ++-e-,--+-- -- = = - - = = = = - - + - - - * * * ' ' - + - ' -

Attachment 2 to TXX 91064 page 10 of 10 INSTRUC110HS FOR INCORE0 RAT 10tl RTJj211 INSERT page 3/4 3 1 thru 3/4 3 8 page 3/4 3 1 thru 3/4 3 8 page 3/4 3 15 thru 3/4 3 24 page 3/4 3 15 thru 3/4 3 24 page 3/4 3 35 thru 3/4 3-36 page 3/4 3-35 thru 3/4 3 36 page B 3/4 3 1 thru B 3/4 3 2 page B 3/4 3 1 thru B 3/4 3 ?

'I

Attachment 3 to TXX-91084 Page 1 of 22

3. a 3 :N;*: 9EN'aT::N 3 ut 3.1 staCIOR TRID SYS*!u :45*4LMENTa':CN

. !v: :3G ::NDIT!!N C*R OPE a* ION 3.3.1 as ' .-

  • e :es:t:r *rio System 'rst*w entat'on e casa e:s tac 4atericc(s f *s::e 3.3-; s s'i ce CCERaBLE.

4:cLI:a9:'.: v: as sno.n 4m " sole 3.3 1, ACT:0N:

As sne.n in Tacle 3.3 1.

SURVEILLA NCE DE0VIR!u!NTS 4,3.1.1 Each Reactcc Trio System lastrumentation channel and interlock and the automatic trio logic small os demonstrateo OPERABLE oy the cerformance of tse Reactor Trio System Instrumentation Surveillance Requirements specifiec in Tacle 4,3 1.

4,3.1.2 The REACTCR TRIP 5f5?EM REiPONSE TIME of eac 2eactor trio functi:n snell ce cemonstrated to ce .itnin its' limit at least nce per 19 montns.

Each test snail incluce at least one train suen tnat :.tn trains are tested it least once per 36 montns a c one enannel per function .sch that all channels are testec at least once every N times 16 months nere N is tPe total aumoer ,

of recuncant charae's in a specific-Reactor trip funct'on as snown in tne

" Total No, of Ca.anael s" c:1umn of Table 3.-31.

t CCMANCnE FEAK UNIT 1 3/4 3-1

O m c.. IAHIE 3.3-1 2e#-

9 E EC n g 3: LAC 10N IRIP SY5ilM INilRUMINIAll0N m :r o4

~

MINIMtM 3

% 10IAt NO. EllANNItS OlANNti5 APPI ICAllt I m" ee R IIINCI10NAt UNII- Of CHANNtl$ 10 IRIP OPlRABIE _ MODI '. ACIIGN , g

1. Maeusal Reactor Trip 2 I  ? I. 2 1 N x

z Q 2 I 7 d

3.4

  1. '," 9 h o

e

2. Power Range. Neutron Flux I
a. liigh Setpoint. 4 2 3 1. 2 2 le. Low Setpoint 4 2 3 l'. 7 2
3. Power Range. Neutron flux 4 / 3 1. 2  ?

Iligh Positive Rate t'

4. Power Range Neutron FInx. 4 2 3 1. 2 2 Y High. Negative Rate u
5. Intermediate Range, Neutron Ilux 2 1 2 l'. 2 3
6. Source Range,fleutron Ilux
a. Reactor Trip and Indicatiun -

b 2 2 2 4

1) Startup 1
2) Sinaldown 2 1 2 3. 4. 5 5 h
h. Boron Dilution Ilux Doubling 2 1 2 3 .4. 5 5 Overtemperature M-16 4 2 3 I. 2 12 7.

2 1, 2 j?

8. Overpower M-16 4 3 ,
9. Pressurizer Pressure--low 4 2 1 l'8 s.
10. Pressuriter Pressure--Iligh 4 2 3 1. 2 8.

ll

. l.

?

~

l .i g I AIII II ,I I'.I30""li*"'II' . . .

o.

. c w2,

.,,E. . RI AC IIW I.R IP.M_*.l_l M I_N_Si_l_:tME_N__I Al__lO__N_

r g, Aq'

_ "'_ MINimm ~

i CilANNIIN AITI IL Altii ww j ,1 1OiAt NO. ' CinANINI$

AC. I__10ft. o

$; I li.feC_l IONAL' 18611 01 Cl_lA_NN.I_1_5

_. . _IO_ I N I.P

_ O__P_I R_Att._i _i_ ._. M. M M...N..

2 1 6 .U

c 11. Pressurizer Water Level--Hiedi -

.1 2.. y

z
12. Reactor Coolant iIow--tow
a. Siewjle Loop- t/ loop 2/It=qi in 2/ loop l' es j l

4

.ney leng.

1 i // loop in 2/ lex p I'8 (.

' li. -Iwo Ioops. - :t/leug any two luogn -

4/sta. ep-n . 2/sta. apn. 3/sta. apn. 1. 2 i w 11. Steam Generator' W ter ~

in any .fm.

E level--law-low -

ep n.

w 1 w 2 3 l , h I I4. : thwh rvoitage--Reactor CUolant ' 4- 1/tma.'

h agis 3 4.

15. thuk rf requency--Reactor ' Coolasit . ' 4- I /letes 2 3 l - '

j- p,q,3 ..

V .

i I

14.. luilsine trip'

.s . Iow Iluitt Oii Pressuse t 2 7 l' t.

4 4 l' to li.' Itnisiew Stop Valve Closuse- 4 I 1/. Saf e ty inje tt ion leignet 7 1 2 1, 7 hb i lie.m I*IAs I

)

i i

i- .

.a .

m>

$,0

- c,

. --.i-1 I AlliI 3. .

(Continued)

C IUR __I_R_IP 3YSit M_ _IN'2iRIMI NI ALIGN

"[

+n HI A.

kx HINIMLM re "

'" Al'I'l I Call! I IDIAl NO. CllANN! t 5 CilANNIIS "

5 A Ol'1RAHIi ACiICN .

{Ji 0,lANNII5 if}_JRIP _Pt101 s N IUNCIIONAl UNIi Y y

18. Ih.ector trip Systra Interlotks Interamliate Range 2 I  ? /' / 3 g

5 a. a

-* Neutrun flux, P-fi t*

h. tow Power Reactor Irips Block, P-7 .

4 2 3 1.7 /

1) P-10 Input I /

2 1 2

2) P-13 lieput 2 3 i /

Power Range Neutron '4

c. .

I lux, P-8 w

2 4 2 3 l' /

m d. Power Range Neutron Flux, P-9 4

4 2 3 1. 2 /

e. Power Range Neutron l Flux, P-10 2 1, 2 8 Il 2 I
19. Reactor Trip Breakers  !^ . 4 ^ . */ 9 1 7 2

Atitomatic Irip and Interlock 2 1 2 1. 2 [ 33

20. 2 3^ . 4 . ^ 4 2 1 tonic

% e e

.y-Page 5 of 22 l

. TABLE !.3 1 (Centiaveo)

  • !aBLE NOTA 7 IONS 8

0nly if the reactor triD Dreakers n8Doen to De in the closeo positi:n sac tre Control Rod Drive System is cacaole of red withora a1.

DBelcw tne 8-6 (:- e meciate Range Neutron Flu 4 Interlock) 5et:;e* at.

C Below the P 10 (Low 5etootnt Power Range Neutron Flua Interlocx) Set;oint.

Abg g e P 7 (At Power) Setpoint W he appi,ssb.is i nvw o and ACTION statements for these channels noteg 1}

able 3.3+2 are more restrictive and therefore, acolicaolef

'IAbove the P*S (3 loep flow permissive) Setpoint.

9Above the P 7 and below the P B Setpoints, hThe boron dilution flux coubling signals may be blocked during reactor startup.

IAbove the P 9 (Reactor trip on Turbine trip Interlock) Setpoint ACTION STATEMENTS ACTION 1 With the numeer of OPERABLE channels ore less than the Min %um Channels OPERABLE reautrement, restore me inoperable enanre!

to OPERABLE status with,in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or t- in HOT STANOBY witnin the neat 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTICN 2 - Witn the number of OPERABLE enannels or, less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceeo proviced the following conditions are satisfied:

a. T5e inoperable channel is placed in the tripped concitten

-itnin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, ,

b, _ The Minimum Channels OPERABLE requirement is met; newever, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification

( 4.3.1.1, and Either, THERMAL POWER is restricted to less than or soual cs to 75% of RATED THERMAL POWER and the Power Range Neutron Fium Trip Setpoint is recuced to less than or toual to-

$5*. of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once cer

-12 hours per Specification 4.2.4,2.

CCMANCHE PEAK-- UNIT 1 3/4 3 5

"~ #

Att'achment 3 to TXX 91084 ' '

Page 6 of 22 *: ELE 2.3*.("ent N e)

  • i AC'tCN STATEMENT! (Continued) '

ACTION 3 ' With the numcer of channels OPERABLE one less than the Minine Channels OPERABLE recuirement and with the THERMAL DOWER 1evel:

a. Below the 8-6 (Intermeciate Range Neutron Flux Interlock)

Setooint, aestore the inope.aele enannel to OPERABL[ ggatyg prior to iscreasing THERMAL POWER above the P 6 Setcoint.

c. Above tne D 6 (Interteciate Range Neutron Flua Interlock)

Setcoint but below 10% of RATED THERMAL POWER, restore the inoperable enannel to OPERABLE status orier to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

AC' 0N 4 With the numeer of OPERABLE channels one less than the-Minimum Channels OPERABLE recuirement, suspend all operations involving positive reactivity changes, ACTION $ With the number of CPERABLE channels one less than the Mini s m

-Channels OPERABLE reovirement, restore the inoperable enannel to OPERABLE status witnin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or within the next hour ocen  !

tne reactor trip breakers, suspend all operations involvirq positive reactivity changes anc verify either valve -105 !455 or valves 1C5-8560, FCV 1118,1C5 8439,1C5 8441. ano 1C$ 8453 are closed and secured in position, and verify this cosition at least once per 14 cays thereafter. Wi no channels OPERABLE complete all the above actions within - nours and verify tne positions of tne above valves at least :nce per 14 cays tnereafter..

ACTION 6 - With the number of OPERABLE channels'o , less than the Total Numeer of Channels, $74RTUP and/or_POWEA OPERATION may preceto proviced'the following conditions are satisfied: ,

4. The inoperable channel is placed in the tripp'ed conoition L within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and- u D. The Minimum Channels OPERA 8LE requirement is met: me.ever, the inoperable channel may be bypassed for up to 4 neurs  ;

for surveillance testing of other channels per Specification 4.3.1.1, n ACTION 7 With less than the Minimum Number of Channels OPERABLE, within.

I hour determine by observation of the associated permissive annunciator window (s) that the interlock-is in its reQuireo state for the_ existing plant condition, or apply Specificaticn 3.0.3, L

!~

l i

o L

COMANCHE PEAK UNIT 1 3/4 3 6 L

Attachment 3 to TXX-91084 P6ge 7 of ??

l

  • l TABl_E 3.3 1 (Cont *nued)

ACTICN STATEMENTS (Continueo) vi nism ACTICN 8 - With the number of OPERABLE channels one less tnan tre Chanrels CPERABLE recuirement. te in at least HOT 54C55 se to

.itnin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may te oypassec f:

2 nours f or surveillance testing per Specification 4. 3. '.1 or '

maintenance, provided the other channel is CPERABLE.

ACTION 9 With the numoer of OPERABLE channels one less tnan toe Minimum Channels OPERABLE requirement, restore the inoperaole enannel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.

ACTION 10 With the number of OPERABLE cnannels less than the Total Numoer of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 11 with one of the diverse trip features (undervoltage or thunt trip attacnment) inoperable, restore it to OPERABLE s',atus within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and spoly ACTION 8. The breaksr snali not be bypassed while ore of the diverse trip features is inoperable except for the time recuired for performing maintenance to restore the breaker to OPERAB'.E' status, during which time ACTION 8 applien.

ACTION 12Number - Withof the number of OPERABLE channels'o . less than the Total Channels. STA,RTUP and/or P0ki? OPERATION may proceed provided the following conditions are :ltisfied:

The inoperable channt) is placed the tripped condition a.

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and r.owever, D. The Minimum Channels OPERABLE requirement is met:

theinoperablechannelmayoebypassedforuptognours for surveillance testing per Specifications 4.3.4.1 or 4.2.5.4 OME 'U#

igg D E hk NVM M hhkbbt.b (k W S Qw & pte, hedi OVG RAM Mb.M *"b

}' wstky JY trW hh hvMi va kI"Alwt he NOT[f&

kQfa

" , M N " A ^^^Y BY& L m.%

6 MMM

( gwd g(j ( b^

N N'3'vhb l O*' h *"db yM u'i b

N 09GRM- l

. COMANCHE PEAK - UNIT 1 3/4 3-7 ,

Y 'A M : % 3 r % h il d g e n s *i, W , , w :, % ;.,. , y y; ,

gj,f- jfy

o>.

TABLE 4.3-1 kh 3 .

ce 9

REAC10R_1 RIP SYS11M IN51RUME NTAT10N SURVE Ill ANCE REQUIREMENis o,4 g

,E 1 RIP h j; # ANA10G AC10AilNG MODES IOR ,

O

. CHANNIL CEVICE WHICH c- Cl4 ANNI L CllANNF I OPtRAIIONAL OPE RAIIONAL- ACTUAIION SURVF11 L ANCE j 5,. IUNCil0NAL UNIT CHECK - cal lBRAT ION IISI 1E5I 10GI C __I E_M_ _ 15 REQUIRED 7

1. Manual Reactor irip N.A. N.A. N.A. R(l4)' N.A. 1, 2, 3 , 4*,  !
2. Power Raswee, Neutron flux ^
a. High Setpoint 5* D(2, 4 ), O' N A. M.A. 1, 2 M( 3. 4 ),

Q(4, 6),

R(4, 5)

b. Luw Setpoint 5 R(4) 5/U(I) N.A. N.A. 1 , 2 y ,
3. Power Range Neutron F lux, N. A. R(4) Q N.A. N.A. 1, 2 (o High Positive. Rate
4. Power Range, Neutron Flux, M.A. R(4) Q N.A. M.A. I, 2 High Negative Rate C
5. Intermediate Range, 5 R(4, 5) ' 5/U( 1) N.A. M.A. I,2 Neutron flux
6. Source Range, Neutron Flux 5 R(4. 13) 5/U(1), Q(9) R(12) N.A. 2',3,4,5 Overtemperature M-16 5 D(2, 4 ) Q M.A. M A. I, 2 7.

M( 3, 4 )

Q(4. 6) '

R(4. 5)

Overpower N-16 U(2, 4 ) N.A. N.A. l. 2

8. 5' Q R(4, 5)

N.A. N A. I

9. Pressurizer Pressure--Low 5 H Q(8) , ,

N.A. N.A_ 1, 2

10. Pressurizer Pressure--High 5 R Q o_ m. _. .m -...:...._ .w__m. ,._,_, ___ ,,__m _ _ _ , _ , , ___ _ _ _ , , _ ,_ _ _ _ _

I Altl i 3.3-2 y3

~

iNGINIIRID SAII IY li AIllHI S ACIUATION SYSilM IN51RIMI NI AIION s n %3 I

MINIMiM  %

,1 0%NNII S APPIICA*111 ,

y 10lAl NO. CilANNI 15

  • O Of O!ANNII*. 10 1 rip _ 011 RA!!II Me101, *> _ Ati10N

, IllNCIIONAL UNII tk ,

Y

?

! k

1. Safety Injection (ECCS, Reactor Irip,feedwater

~ Isolation Control Room 8"

[mergency Recirculation, leergency Diesel Ge wrator Operation Containment Vent isolation, Station Service Water, Phase A isolation. Auniiiary Ieed-water-Motor Driven Pump, R *

    • Itu bine Trip, Component Y Cooling Water, Essential O Ventilation Systesas, and Contaisunent Spray Pump.

2 1, 2, 3, 4 f f>

a. Manual Initiation 2 1
b. Automatic Actuation 2 1 2 1, 2, 3, 4 [

logic and At_tuatinn Relays 1, 2. 3 Il 3 2 2 c Containment Pressure--High-1 4 2 3 I, 2, 3" Il

d. Pressuriier Pressure--low il
e. Sican tine Pressuse--low 1/stcan Iine 2/sIeam Iine 2/3icam Iine I , 2 f' in any steam line 1

m>

o r+

1 Allt i 3. 3-2 ( Cont siiticil) *N

  • " r 3

x *$

$ INGINIi HI U

  • AlI iY II AIllRI S AClilAlION SYS11M IN'siRilMINi AIION %5 "x ru r.s HINiM!!H CilANNII5 APPI ICAllt i *

.% 10IAl IMI. CilANNII5 ACIION GI't RAllt t MODI S A 32 " 01 CilANN!Is 10 iRIP

!. . IIINCi10NAl IINII e 7

E 2. Containment Spray

? pais- 1. 7 l. 4 f f. g 2 leair i pa i s- g

" .. Nnu.il Initiation operat e d s imi i L..neotis I y

?  !  ? 1. 7 1. 4 )( N

h. Automatic Attuation .

logic arid Actisation Relays I I4 2 I I. 7 4

c. Containment Pressure--

11igle-3 '

t'.'

s

'I 3. Containment Isolatiott

.a

  • a. Phase "A" Isolation 1. 2, 3. 4 It.

1 7

2

1) Manual Initi.ition 7 1. 7. 1. 4 ff 2 I -
2) Automatic Actuation logic and Actuation Relays :ii t iat siwa t iivu t iensis asi.I l

Salety Injection 'a c i t em 1. .ilmw les: a l l Sa f et y I sije-t t i..si

3) s enta i remen t <.

Is. Phase "B" Isolatinn Phase "B" isol.*- 1. 7 l. 4 18-N nual Initi.ition See item 7.a al.ove. lien

1) 1 inn i s mains.n l ly in i t s.it e.1

. ..nt .ii nment ,peay linn I i ai : . m.inee e l ly initsaleI. 3. l. I. 4 [ ( .

?

Atit om.st ic As l anat ion 7 7)

I oqit. aiul Attnation ' '

Ile l ay*, 8. /. I l '1 .

/ 3 4

3) Ensitainment Pressus e--liinh- 3 .

E

-':r >

c .

I A!!! I -1.1-/

-- (Continen

-- l) ~ * * *

~y 3

i. Nr.lNI i 1:1 li sal i I ( l i A. f ilEl_.* A_C.; f lA .I. _10_N _ .Y_ '. l. l H__I N _. I Rt_HI _N i. .A_ I_I I)N_

oo e

E *"

r mw ese e

111NIMt H l.ilANNI i *. Al'I'l!IAllii '

"r 10IAl NO. I ilANNI i '.

'S Hilll! *. Al I t.ilN. 4

  • Of tilANNI I. '. 10 1 R18' UI'._l RAlti i -

x

. I ilNC_1_ ION.A..L IIN.I I.

..  ?

c font s ensu sit Vent o a e 7 l.nletinn I. /, I. 1 l'.

2

-. 1 I em / ., .n I t... l .il.nve I) It.n.naI initiatinn I isent .a s sue:esit vesil s .e.l.it s ein a -..

m. inia.n l l y i nt e l :..l eil win se I'li.a .e "A"

..s l.it s osi I.nas t inn ur e e.eil .s inment

,geray Inew I a..n :, m.nweaIiy i n i t s . t e.1.

1. /. 1. 4 l '.

I /

Automatoc u tuatins  ?

/) -

" Ingic aint A4.tuatinei u Relay

';' ins t e at niq t eno I enn . .n I 5.stety injettinn

  • ee item 1. .itenve inr .sil *elety injetison
3) requiscoent .
4. Steam Iine isolatica

... M.uiu.it Initiatsa.a I/ngierat isey I . /' . E' /I

1) 1 plividn.it St eam 1/ te.im lisp- t/ste.im line

~

.I c.im I int-

!ine

1. /* . S' /"

! /

/

?) System l. / . I l'8 I /

/

l.. Antoni.it it fu.timat inn I nqis .isul As t n..t inia Re l .sys I.*- 1 I/

/ /

,,n e-- i 4 font.iniment l*:e li ng!r /

k

'~ .. n .I A.ltil.-' t.1-2 (Cont ima il)-

g'o > -g

. o e, C. ,

,', - E-n ..I _NGINI_I_Ri.l.l NAl l_ .l_Y ! !__A__ll*l_'.' AC lllA110N S__Y_S11 M I NSI RIMI NI A l l0N*'

- . _ . .. _ _. g

, x on

-*e &

$ MINIMIM mw 101Al NO. CilANNtiS CllANNIt5 APPI ICAllii r 3l AdiIGN "C

. IUNCIIONAL INilI "- Of CilANNII % 10 iRIP OPtRAlli[ MODI S

-e E 4. $leam Line Isolation (Continue.1) .

Q,

' el. Steam Line Pressure--tow . 3/ steam line 2/ steams line 2/ steam line 1, 2' . 3d ' t/ $

" in any steam 8 line-

e. Steam Line Pressure 3/stease line  ?/ steam line 2/steais line 3b.c .

gN Negative Rate--Hiiali - - in any steam iine w 5. Iurlsine Irip aml

. 2 f eeenater Isolation- .

2 1, 2 72

  • f a. Automatic Actuation 2 1 M togic asul Actuation i Relays Is. Steam Generator 3/sta. gen." 2/sta. gen 2/sta. gen. I, 2 /

Water level-- . in any stm.

High-liigh gen.

c. Safety leijectiam *ee i t em I . .a-..

. . . all Safety injection iniitiating lusation,.in.1 n gui rements.

I 6. Auxiliary f eedwater 2 1, 2, 3 I*J

.a. Automatic Actisatime logic 2 1 amt Actination Relays le.~ Sta. Gen: Water ievel--

Iow-Ieat 4/.tm. ge n :  ?/st a. gese. 1/s(m. age : . 3. 7 I i/

1) Start Mutus- - in e n li Driven Pasips in'any eper-i atisa; sim. gest. opes.itinq sim. gene . , ,

f d d um -p.# -wa.'" w.,- - - - g- _,, , , , . , -_,_u.___ 4 __ _ _ _ _ _ _ , _ _ _ _ _ _ , , , , ,

m>

c, ,+

'g 7

j I AEd !_13_-L!Et$1._!"""'l} ,

o x

f 14Gitll i Ri D *.All ! r II AlllRi s AtillA110ft SYsit M ltrolROMi ttl Allori C5 c U O3

~t

'" MittlMIIM m t.a N

101Al flO . CliAtJtiliS l'IIAf4til i S Al'I'l ICAllt i

.I. Mulli S Ar ! !Ord y 10 l4IP Ol'l HAl!I I Ilit1Ci10tJAl liill 1 01 i ilAtafil I *,

a x

x i.. Auxilia y l eentwate r (Continneil) 3/stm gen. I , i* , (I 1/ 3 b 2) Start lurls i ne- 4/stm qr ti .  ?/,tm. gen.

in assy in v.u to h

~ Deivesi Pump oper,etinq

/ oper.itary sim. gen. stat gen.

Sately 1:*j et. t ion *ee item 1. .ileuv e- to . ell '..itely inje(tion in e t s,et s nq t enu t iner,

.e n.1 5tas t flotor-Driven Pumps equin ments.

I /t s .t in I/t e a ss 1. 7, i 18.

al. loss of-Offsite Power I/ train

(., Start Hot or-Ils iwen

i. l'usnps and lurtaine-Driven Pietp Irip of All Main 7/Af W Inep  ?/AlW piimp I/AlW Inimp 1. ?  ?!

G e.

Icedwater Pamps Start Motor-Driven l'teps

/. Automatic initiation of LCC5 Sw i t t.liov es to Cont.einment Simp

a. Automatic Attuation 2 1 2 I. 2 f. 4 A fh logic asuf Actuatior Relays

Athchment 3 to TXX.91084 Page H of 22 N N N c. N N e

v

  • w w w w w 7.- .s

=t e

-i

. - N N N N N N -

t g' *

c. , . .

E w - -

= si . -

<, m

=t u= .-

or

-s -p .

.=

s N -

t . =

m,

-. ~

& v y

x .-

s s

s s s s e - s s

s

s. - .s. s s

- g x - - -

m -

ri E =<=. s e.

l -zm'. -

~ -. = .g i  :

~: 1

. s

l I v c:

2 3

. - c>

=s a

~g  : s s s +

v1 <t ~  :  :  :  :

+

+

a i -

^ . .

u

.i s . s s s s s

s N,1' . _ - > > N N N e. N N

<< == N : -

.I z- .: : -

ei < >  : ;

= l

=e' v-+ y

, =j -w

=: <t e:

e

_. i.

.s s s s .- s s

- - . t. . .. .

.a _..

a' s: s s s

<A

<=j w A 1 N N s.

N N s

N s

N N vi or -. =. .

.- ~

w v

>= , c z! .~ > .e. v  :.

=~ .- .-

= = v t.:: ,

y o m- e e,  : -: =

z! =

~ w: - - , -

2 m > - e -e - - -

i - o m .e 6 o=m c= s cI= .,s >:

e > >

x =

.c .m a >

Q@

2 .=

o s : r- ~s v = a

=

D wO T C' *: . L *

  • l 4 ...6

=

O w>  : > 0 U .- = 7

.s . 3 .

va ww = T,  :: wo - -

me- i - o oe w e=

.- v :: - e. E >: .., # =

m a c a

.a u o> : = s u= -= = z = 2 .e.,

-o: e 6x :. = <= =  :-

-i

-s

.: u , ..

.o vm 3 >z >e cs

> > a

.a o

> a: .:

= -

ve: - = vv : # e v sv a s a > >a -

: ->: a . ,c : ::a e

u

== e e e e =i

.s ou 3 e:

=u-

-e . o- .o =. .

= m= -: 9 m .:

av - oa- =

ea eo e e w vs  : v r

< .s o ww.-

z .ca .e :. m-

.c. . <xs  : 2 .g .

oe =v

.: s  :- .c w

~:

y

- vu u s

=

-u s. x 1

OCMANCHE PEAK - UNIT 1 3/4 3-20 l

l 1

1

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

=1

~

I Allii 3.1-2 (Cont inum d)

~

Nl S. . TEe, "r!

n iNGINIl_ kill SAII IY II AltlRI 5 AClllAl10N *.YSilM INSIMilNI NIAll0N ~

N

'E- MINIMiiM E5 !

10lAl NO CilANNI1S CilANNI I '. Al'l*l ICAlti i Nw:

,z ACll0N y

y filNCll(MtAt UNil Of CllANNLi s 10 INIP UP! NABil MIMil 5___ ,, gl

-o; li. Salety injection See item 1. above for all *,alety linjet. tion initiatisu.I lana t ions asul gi z respai rements g

~'

3-

~* o;

!' - 10. Lngineered Safety Features g Attwation.$y5 tem Interlacks 7 2 I. 2. 3 It!

a Pressurizer Pressure.- , 3 P 2- 7 2 1, 2 I 78
b. Reactor trip, P-4 w- 11. Solid State Salequards ,
  • Sequencer (5555)

! w

a. Safety injection 1/ t e .s in I/t ra s e: 1/ train 1. 7. 1. 4 / N h Sequence j
b. Blackout. Sequeence I/ train 1/ train 1/ train I 7. 3. 4 75 e

e

Attachment 3 to TXX 9108f -' N Page 16 of 22 .' ,

TABLE 3.3 2 (Continvec) .

l

  • TABLE NOTATICNS

' Trio functio may ee bloctea in tnis M00E telow the P ll (Pressurizer Dressure Interlock) Setpoint.

DTrip function automatically blocked above P-ll and may et unblocked celo. P ll

  • oy ele: ming the Safet/ Injection on low steam line pressure.

CNet soolicable if eacn affected main steam isolation valve and its associated upstreem crain pot isolation valve per steam line is closed.

UThe provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

'The channel which provices a steam generator -ater level control signal (if one of three specific trip channels is selected to provide input into steam ggnerator water level control) must be placed in the tripped condition within

(, I hour and maintained in the tripped condition with the exception that the channel may te taken out of tne tripped condition for up to hours to allow testing of reduncant enannels.

  1. N ot applicable if Preferred Offsite Source Breaker is open.

ACTION STATEMENTS ACTION 12 - with the number of OPERABLE channels one less than the Minimus Channels OPERABLE requirement, be in at least HOT STANOBv within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within t ' following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; hcwever, one channel may be bypassed.fc. up to 2 nours for surveillance testing per Specification .3.2.1, proviced the other channel is OPERABLE, l ACTION 13 'aith the number of OPERABLE channels one less than tr,e Total k Numcer of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provideo the inoperable channel is placed in the tripped condition witnin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. _- L ACTION 14 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoper-able channel is placed in the bypassed condition and the Minimum Channels OPERABt.E req'Jirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per i Specification 4.3.2.1.

a ACTION 15 - With less than the Minimum Channels OPERABLE reoutrement, opera-tion may continue provided the containment pressure relief valves are closed witnin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and maintained closeJ.

ACTION 16

  • With the numbar of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANOB) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN *within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

COMANCHE PEAK - UNIT 1 3/4 3-22

Attachment 3 to TXX-91084 Page 18 of 2?

, *a2LE 3.3-2 (Continued)

( ,-

ACTION STATEMENTS (Continued)

ACTION 24 - With the numeer of CPERABLE channels one less than tre Min mum

  • Channels OPERABLE reeuirement, restore the inoperacle enannel to OPERABLE status .ithin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or initiate ano maintain operation of the Control Room Emergency Recirculation System.

ACTICN ;5 - .ith ne number of OPERABLE channels on one or more trains less than tre Minimum Channels OPERABLE reovirement, declare tne ciesel generator (s) associated with the affectec traints) inoperaole ano apply the appropriate ACTION for Spec 1fication

3. 8.1.1.

hTION26'WiththenumberofOPERABLEchannelsonelessthantheTotal 1 umber of Channels, STARTUP and/or POWER OPERATION may proceed proviced the following conditions are satisfied:

a. The inoperable channel is placed in a tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
b. The Minimum Channels OPERABLE requirement is met; newever, one accitional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of other enannels per j

, , < ecification 4.3.2.1. f dc{lt" c% - &lh& meks 0fG yn NG {} l,y O L % M ia;.- v c k dt O FGP n3lE ,fa s a m a.s

                                                                                          'd Aubat & Mjwo$b obw&b OtWAl)l$ s k t hwn m he r dw w nobo y was 9 4 6co ;bons  6wn,  sfo~e ra Colb dau 4THu1h4
                                                                       -y 4< Qpdce  mf,4 l,Q) op h q heus e r u ve & d ib .. p q u .p n .o o a   L  oA    cla    d;n oM HolG.

y,3,3.i; p COMANCHE PEAK - UNIT 1 3/4 3-24

Attachment 3 to TXX-91084

  • Page 17 of 22 ,

l I TABLE 3. 3-2 (Conti*uec) , ACTION STATEMENTS (Continuea) I ACTICN 17 With the numoer of OPERABLE enannel e less than tne Total Number of Channels,(TIARTUP ane/er POWTMTaTO may croceed  ! provicec the following concitions are satisfieo: ferafiu k

a. The inoceraole enannel is placed in the trippea conciticn gg within 6 hours, anc D. The Minimum Channels OPERABLE recuirement is met; newever, one additional channel may be bypassea for up to 4 hours v" I for surveillance testing of other channels per Soecifica- g\1 tion 4.3.2.1.

g 3 c. ,% ACTION 18 - With less than the Minimum Number of Channels OPERAdLE, within 1 hour determine by observation of the associated cermissive

                                                                                                                            '"Di'  p'a annunciator window (s) that the interlock is in its required                                       a state for the existing plant condition, or apply Specification                       i g,~

3.0.3. j g@-}4 ACTION 19 With the number of OPERABLE ch nnels one less than the Minimum D. .i ' Channels OPERABLE requirement be in at least HOT STAN0BY within #! g4 -6t M* hours and 9 " 1am HOT SHUTOOWN within the following 6 hours; however, one channel may be bypassed for up to.X,, hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. . ACTION 20 With the number of OPERABLE channels'o a less than the Total d ';p I Number of Channels, restore the inopers:le channel to OPERABL 2 '

                                                                                                                                      '4 status within 48 hours or be in at least HOT $TANOBY within 6 hours and in at least HOT SHUTOOWN within the following                                gj e hours.                                                                                 p 3.-

ACTION 21 - With the number of OPERABLE channels one less than the Total f. Numoer of Channels, restore the inoperaole enannel ' ttoI OPE J f.o, bg able and take the ACTION eequired by Specification 3.7.1.5. MG ACTION 22-WiththenumberofOPERABLEch6nelsonelessthantheMinimum Channels OPERABLE requirement, Ace in at least HOT STANOB p M b gforhours; however, one channel may surveillance testing per Specification 4.3.2.1 provided the be bypasse

                                                                                                                 / hours other channel is OPERABLE.

1 ACTION 23 - With the number of OPERABLE channels one less than the Total ' Num:er cf Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: L l ) a. The inoperable channel is placed in the tripped Condition witnin 6 nours, and .

b. The Minimum Channels OPERABLE requirement is met.

COMANCHE PEAK - UNIT 1 3/4 3-23 i

8 I Allii 4.1-2 (Camt iluu*si) L NGINI I I:1 II *.Al i I Y I I AIHRI S ACIUAI IlWI SY*.ll M IN*.lkilMI NI Al ION - S- . .E -

                                                                                                                                                                                                                                                  *5 m                                                                               wist lij KNU~Nilpillit NiliF, 5                                                                                                                                                                                                                                           *5  .
     -,;                                                                                                                         ' l k li-                                                                                                        p      *
       *-                                                                                                   ANALOG                ACillAllNt.                                                                         ^ MilHi *.

j ( c CHAfNel i I IIANNI I (llANNII CllANNI I til VICl Di'i RAI IIW4At .Ol*1RAIIlW6Al ACtilAlION IIA .ll R St AVI.

  • l OR Willt ll I:t i At R1iAY SlH,'vllliANsIg l.Al till:Al llW4 l l *. I I.O.GI C 11 *.I il *.I II *. I _l'_. 11_ f pi t ki ll e
      .=. 1IINCIl006.A.l__IIMII.

( lit f K . _ i l *> l _ _.

      -.                   .-                                                                                                                                                                                                                            o p

I. Aul ..smal is: lesit s.sl ion of n l 1 CC*. *.w i t s.h..ver t o Cowil .s ines sil ! *ausi. (Caentisuis el) le. NWSI level-l aw-l ew S ' *R

                                                                                  .                      Y                        N.A.                       N.A.                     N.A                  NA                   1 /,       5, 4 Cesieu ielent Witti                                                                                                                                                              .

t l S.ilely injet.t ion See llem 1. almve f or all S.elety leijet.l son Surve s t i.nu e Requis ew of . i 's

      ,   11 loss of l'ower (6.9 kV &

j  ; 480 V Saleges.ards

u. System li,wlervoltage)
a. b.9 kV l'ref erreel Ol 8 site N A. N.A
                     $ource fleulervoltage                        N.A.            R                         N.A.                   (3, 2)                    N A.                                                               I, /,      f, 4
1. 6.9 kV Alternate Oltsite 91 A. N.A. N.A. I , 7, i. 4 Soure.e lieulervoliage M.A. R 14 n . (3,2)

L c. , b.9 kV lius Usuler- N.A. N A. 1, 2, J. 4 voltage M.A. K k.A. (3, 2) M .. A . i

el. h.9 kV 1)cgraile<! NA NA I, 4 Voltags- N.A. R N.A. ( ), 2) N.A. I , 2, l e. 4110 V Degradeel N A. N.A. I, 4 N.A.
                                                                                                                                                 ~

(3,7) I , 2, Voltage . N.A. R N.A. h I. 4t10 V t ow Crial HA N A. I , 7, t, 4 Deufervo l t age N. A. R N A. (3,/) N.A. l l

               ,               .m.- -      . . , , . . , y  .                              .--  v   -.r..                  ..e.,                       ...,.....-y     ,._,,__m    mm   . , _ _ ___._.__m.        _  _ _ _ _ _              _
                                                                                                                                                                                                         ,y I Atil I 4.1-7 ( C. nt isoscel)                                                                                      ae ce c.
                                                                                                                      *.Y 5 t l H_ _ I N_5_1_R_UN_I_N.

__ _ . _ . _ _ I A l _l 0N 5 _I N_G_ I_N_i_.l _Hi. l.)_ *.A_l l_l Y_ l l A lllRI *. AC_. illa l l 0H y E5

 '"                                                                                                                 i R11 m                                                                                           ANAL (N;              AClIIAI IMl;                                                     l(1111 *.           M"
 ',"                                                                                                                                                      iW. i t R
  • I AVI IliR Willili E tilANNI I Di Vli'l .

RIIAY SilRVI III AN:I u

 "                                                   ., o                                                                               ACillAi10N        RIIni CilANNI I    tllANNII
                                                                         ~

Ol'l RAI IOflAl Ul*lRAIIONAl _15_ Niijill Ri tt U,, CilANN) l I_I._N;i_C ll *.I .II *. I I I. .*_. I cal lllR_A_l. lOli 1 t *. I . I I *. I. __ _ _ c: l_ilN_ C_ _1 1. _0.NAl llN i.l. C_i.ll C.K -

                                                                                                                                                                                                            ~

_E a gf Control Renm Incrgency w

 ~       *l.

lies:: roulat n,un N A. fl. A. N.A All N.A. k

               .i. Manis.ii luitiation                 N.A.         ft. A.
                                                            *> ee !! em 1. al.ove-     Ior alI *alely inIc Iinn surve ilaine Re.gone.m.ut .

le. ..ilety injet.t inn m 10. Ingineereil Salety -

  ;_               l cat sin es in."-aal ion m                *>ystem Int erlocks                                                                                                                                                           I N.A.              N A.          N A.          l. 7 Q                    N.A.
                   ..        Pressuriler                    N.A.         R l'rev. sire, P-il Il                  N.A.              N.A.          NA            i. ?,     i Reas. Lor irip, P-4           N.A.         N.A.                   N.A.

I,. ll. Nul ial St at e % legis.si el<.

  • equenter ( SSSS)

N A. f4 A f4 A I. /. 5. 4 it A. N N.A. H( l . l.4 )

                    ..      .% lely in jet:1 inn Seyneau e N A.               HA            NA           i. /,     t.  ~4 N. A.                H(l,3,4) le        111.et tout '*.cgiocau e       N A.        R

y . _ _ -

                   ,    Attachment 3 to TXX-91084 Page 21 of 22
                        ,j 1 1 35 cLMEN' O ELSE5 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETi FEATURES ATUT:CN sv5 TEM INS N ML M TION
                ,~

The OFERASIL;TV of the Reactor Trio System and the Engineerea safety Featres Actuation 5,, stem instrumentation and interlocks ensures that: (1) the

                !'          associated ACTICN anc/or Reactor trip will be initiated wnen the carameter gg

[ vonitorec ey eacn enanrel or comoination thereof reacnes its Setooint (2) the specifiec coincieente logic anc sufficient recuncancy is maintainec tc permit

        ..                   a channel to be out-of-service for testing or maintenance consistent                      ith main-jh                  taining an appropriate le;e1 of reliability of the reactor protection and engi-neered safety features inst umentation, and (3) sufficient system functional g [.ll               capacility is available frca diverse parameters.

i a The OPERA 81LITY o' trese systems is required to provide the overall

             -P              reliatility, redunJan'.y, and diversity assumed available in the facility design for the protection Ar.a # iigation of accident and transient conditions.                     The i         ,

31

              }               integrated operation of      :ch of these systems is consistent with the assumptions O                                                      The Surveillance Requirements specified for these 4,lusedinthesafetyanalyc"s.

systems ensure that the overall system functional capability is maintained ccm-

       ] {V g

N w t aracle to the original design standarcs. The periodic surveillance tests per-ormed at the minimum frequencies are sufficient to demonstrate this capability. s g$, pe-ified survaillance intervals and surveillance and maintenance outage times

i. ave been cetermined in accordance with WCAP-10271, " Evaluation of Surveillance
          ?3 "s                 recuencies aM Out of ServicLTimes for the Reactor Protection Instrumentation
       'i. M System" and supplements to tA f reportfas approved by the NRC and documented in a- + ' tne * (letters to 1 J Wsn ' cm CW' O. TM a dated February 21,19BSf, n The Engineered     f         tN es A $ a N Sy e :'strumentation Trip c'-
  • Setpoints specified in Table 3.3-3 are the nominal va es at which the bistables Q],9j are set for each functional unit. A Setpoint is cons ;ered to be adjusted consistent with tne nominal value when the "as measureo" Setpoint is within f
     !g7 g g@e the cana allowed for calibration accuracy.

6* Tc accommodate the instrument drift assumed to occur between operational

                            . tests and the accuracy to which Setpoints can be measured and cslibrated,                                  [

3gk" d Allowable values for tne Setpoints have been specified in Table 3.3-3. Opera- P tion wi'.h Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in tne safety [ 4 D. j analysis to accommodate this error. An optional provision has been included a

                          ' for eetermining the OPERABILITY of a channel when its Trip Setroint is found                                 ?

to exceea the Allowable Value. The methodology of this option utilizes the :E "as measured" deviation from the specified calibration point for rack and 9 sensor components in conjunction with a statistical combination of the other U uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. in Equation 2.2-1, I + R + 5 < TA, the intersctive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 3.3-3, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack crift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, R or Rack Error is the das measured" ceviatior., in the percent span, for the af fected channel f rom the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of COMANCHE PEAK UNIT 1 B 3/4 3-1

Attachment 3 to TXX-91084 , Page 22 of 22 INSTRUMENTATION e[ ,* BASES REACTOR TR!p SYSTEM and LNGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRLMENTATION (Continues) . the sensor from its calibration point or the value specified in Tab 1rr 3.3-3, in percent sean, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor crif t f actor, an increased rack drif t f actor, and provides a threshold value for REPORTABLE EVENTS. l The methodology to derive the Trip Setpoints is based upon comoining all i' of the uncertainties in the channels. Inherent to the determination of the T Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that.there is a small statistical chance that this 'i will happen, an infrequent excessive orift is expected. Rack or sensor drift, ~ in excess of the allowance that is more than occasional, may be indicative of , more serious problems and should warrant further investigation. The measurement of response time specified in the Technical Requirements < Manual at the specified frequencies provides assurance that the Reactor trip-and the Engineered Safety Features actuation associated with each channel is . '-. completed within the time limit assumed in the safety analyses, No credit was; ' taken in the analyses for those channels with" response times indicate'd as not ' L

             , applicable. Response time may be demonstrated by any series of sequential,                               '"
           ?'  overlapping, or total channel. test, measurements provf:ed'that such tests                 "-                  '   *
       ,       demonstrate the total channel response time as definea. Sensor response time verification may be demonstrated by either: (1) in p' ace, onsite, or offsite ',                       ,

test measurements, or (2) utilizing replacement sensors with certified respons9-

    .I time.                                                                                         .

o The Engineered Safety Features Actuation System senses selected plant q parameters and determines whether or not predetermined limits are being exceeced. If they are, the signals are combined into logic matrices sensitive! "* - ' 1 to combinations indicative of various accidents events, and transients. Once J-the required logic combination is completed, the system sends actuation signals

 ?

to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate

g the consequences of a steam line break or loss-of-coolant accident
(1) ECCS s pumps start and automatic valves position, (2) Reactor trip, (3) feedwater isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position (6) containment isolation, (7) steam line isolation, (8) turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) station service water pumps start and automatic valves position, (11) Control Room Emergency Recirculation starts, and (12) essential ventilation systems (safety chilled water, electrical area fans, primary plant ventilation ESF exhaust fans, battery r.oom exhaust fans, i

and UPS ventilation) start. l COMANCHE PEAK - UNIT 1 B 3/4 3-2

                                                                  - - - - - - -                - - -                      .}}