ML20078A473

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Application for Amend to License NPF-30 for Callaway Plant. Amend Would Relocate TS 3.3-2, Reactor Trip Sys Instrumentation Response Times
ML20078A473
Person / Time
Site: Callaway Ameren icon.png
Issue date: 01/13/1995
From: Schnell D
UNION ELECTRIC CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ULNRC-3129, NUDOCS 9501250008
Download: ML20078A473 (35)
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N' I .1901 Chouteau Avenue

  • 1 Ptat 0mce Bax 149 '

St. tovis, Missouri G3166

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' #d UNION January 13, 1995 s no$[PYe [ent Etscnuc -,

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U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, DC 20555 ULNRC-3129 Gentlemen: ,

DOCKET NUMBER 50-483 CALLAWAY PLANT RELOCATION OF INSTRUMENTATION

, RESPONSE TIME LIMITS

Reference:

NRC Generic Letter 93-08 dated December 29, 1993 Union Electric Company herewith transmits an application for amendment to Facility Operating License No. NPF-30 for the Callaway Plant.

This amendment application would relocate Technical Specification (TS) Tables 3.3-2, " Reactor Trip System Instrumentation Response Times," and 3.3-5,

" Engineered Safety Features Response Times," to FSAR ,

Chapter 16, Section 16.3. The NRC has already '

implemented this line-item TS improvement in the new Standard Technical Specifications (NUREG-1431.for Westinghouse plants). This amendment application is taken directly from NRC Generic Letter 93-08. The Bases discussion specific to Table 3.3-5 would also be relocated to FSAR Section 16.3.

The Callaway Plant Onsite Review Committee and the Nuclear Safety Review Board have reviewed this amendment application. Attachments 1-through 4 provide the Se.fety Evaluation, Significant Hazards Evaluation, Environmental Consideration, and proposed Technical Specification revisions, respectively, in support of this amendment request. It has been determined that this amendment application does not involve an unreviewed safety question as determined per 10CFR50.59 nor a significant hazard consideration as determined '

per 10CFR50.92. , Pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

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9501250008 950113 0 PDR ADOCK 050004B3 At 4

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U. S. Nuclear Regulatory Commicsion Page 2 If you have any quotions on this amendment application, please contact us.

Very truly yours,

/ Donald F. Schnell GGY/jdg Attachments: 1 - Safety Evaluation 2 - Significant Hazards Evaluation 3 - Environmental Ccnsideration 4 - Proposed Technical Specification Revisions i

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STATE OF MISSOURI )

) SS CITY OF ST. LOUIS )

Garry L. Randolph, of lawful age, being first duly sworn upon oath says that he is Vice President-Nuclear Operations

- and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are trus and correct to the best of his knowledge, information and belief.

By 4 A Randolph Garrf L.

Vice President Nuclear Operations SUBSCRIBED and sworn to before me this ~/ / M /<t f d -day of ( biturup , 1995.

V' /

}L4 / h. f$AX4 BARBARI J. PFAfh, NOTARY PUBLIC-STATE OF MISSOURI MY COMMISSION EXPIRES APRIL 22,199Z EL LOUIS COUNDf.

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t cc: T. A. Baxter, Esq. "

Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W.

Washington, D.C. 20037

, M. H. Fletcher Professional Nuclear Consulting, Inc.

18225-A Flower Hill Way Gaithersburg, MD 20879-5334 L. Robert Greger Chief, Reactor Project Branch 1 U.S. Nuclear Regulatory Commission Region III 801 Warrenville Road Lisle, IL 60532-4351 Bruce Bartlett Callaway Resident Office U.S. Regulatory Commission RR#1 Steedman, MO 65077 L. R. Wharton (2)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E21 11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department Missouri Public Service Commission ,

P.O. Box 360 Jefferson City, MO 65102 Ron Kucera Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 T

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L ULNRC-3129 ATTACHMENT ONE SAFETY EVALUATION l

Attachment 1 Page 1 of 3 1

SAFETY EVALUATIQE INTRODUCTION This amendment application would relocate Technical Specification (TS) Tables 3.3-2, " Reactor Trip System Instrumentation Response Times," and 3.3-5, " Engineered Safety Features Response Times," to FSAR Chapter 16, Section 16.3. The NRC has already implemented this line-item TS improvement in the new Standard Technical Specifications (NUREG-1431 for Westinghouse plants). This amendment application is taken directly from NRC Generic Letter 93-08.

The Bases discussion specific to Table 3.3-5 would also be relocated to FSAR Section 16.3.

BACKGROUND The periodic measurement of response times provides assurance that the reactor trip and ESF actuation associated with a specific analog channel is completed within the time limit assumed in the safety analyses. A listing of assumed response times is given in FSAR Table 15.0-4.

The limiting conditions for operation (LCOs) for Reactor Trip System (RTS) and Engineered Safety Features Actuation System (ESFAS) instruments currently require that these systems be operable with response times as specified in TS tables for each of these systems. The surveillance requirements specify that we test these systems and verify that the response time of each function is within its limits. Relocating the tables of the RTS and ESFAS instrument response time limits from the TS to the updated FSAR will not alter these surveillance requirements. The updated FSAR will now address the response time limits for the RTS and ESFAS instruments, including those channels for which the response time limit is indicated as not applicable. The updated FSAR will also clarify response time limits where footnotes are included in the tables that '

describe how those limits are applied. This TS change will allow administrative control of changes to the response time limits for the RTS and ESFAS instruments in accordance with the provisions of 10CFR50.59 without the need to process a license amendment request.

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Attachment 1 Page 2 of 3 DETERMINATION OF NO UNREVIEWED SAFETY OUESTION The proposed changes do not involve an unreviewed safety question because operation of Callaway Plant in accordance with these changes would not:

1) Involve an increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.

Overall protection system performance will remain within the bounds of the accident analyses documented in FSAR Chapter 15, WCAP-10961-P, and WCAP-11883 since no changes to the response times or measurement interval are proposed.

The RTS and ESFAS will continue to function in a manner consistent with the above analysis assumptions and the plant design basis. As such, there will be no degradation in the performance of nor an increase in the number of challenges to equipment assumed to function during an accident situation.

These Technical Specification revisions do not involve any hardware changes nor do they affect the probability of any event initiators. There will be no change to normal plant operating parameters or accident mitigation capabilities.

Therefore, there will be no increase in the probability or consequences of any accident or safety-related equipment malfunction occurring due to these changes.

2) Create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR.

The proposed Technical Specification changes do not involve any design changes nor are there any changes in the method by which any safety-related plant system performs its safety function. The normal manner of plant operation is unaffected.

i No new accident scenarios, transient precursors, failure )

mechanisms, or limiting single failures are introduced '

as a result of these changes. There will be no adverse l effect or challenges imposed on any safety-related  !

system as a result of these changes. Therefore, the <

possibility of a new or different kind of accident is I not created. )

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Attachm:nt 1 Dage 3 of 3 There are no changes which would cause the malfunction of safety-related equipment, assumed to be operable in the accident analyses, as a result of the proposed Technical Specification changes. No new mode of failure has been created and no new equipment performance burdens are imposed. Therefore, this amendment will not create the possibility of a new or different malfunction of safety-related equipment.

Plant procedures for response time testing include acceptance criteria that reflect the limits in the tables being relocated from the TS to FSAR Section 16.3.

After approval of this amendment request, an FSAR Change Notice will be written to reflect the limits currently contained in TS Tables 3.3-2 and 3.3-5. Thereafter, these tables and any changes thereto will be reflected in the FSAR, updated as required by 10CFR50.71(e).

Further, following approval of this amendment request, changes to the response time limits in the FSAR will be made under 10CFR50.59. Future safety analyses could justify response time increases or indicate that a response time limit is not applicable if no credit were taken for a given instrument channel's trip function in the new or revised analysis.

3) Involve a reduction in the margin of safety as defined in the basis for any Technical Specification.

No response time changes are proposed in this amendment application; only the document where these limits are listed will be changed. Future changes will be processed under 10CFR50.59. There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of prottction functions. There will be no impact on DNBR limits, Fg, F-delta-H, LOCA PCT, peak local power density, or any other margin of safety.

Based on the information presented above, the proposed amendment does not involve an unreviewed safety question and will not adversely affect or endanger the health or safety of the general public.

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ULNRC-3129 ATTACHMENT TWO SIGNIFICANT HAZARDS EVALUATION l

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Attachment 2 Page 1 of 2 SIGNIFICANT ELZARDS EVALUATION This amendment application would relocate Technical Specification (TS) Tables 3.3-2, " Reactor Trip System Instrumentation Response Times," and 3.3-5, " Engineered Safety Features Response Times, "to FSAR Chapter 16, Section 16.3. the NRC has already implemented.this line-item TS improvement in the new Standard Technical Specifications (NUREG-1431 for Westinghouse plants). This amendment application is taken directly from NRC Generic Letter 93-08.

The Bases discussion specific to Table 3.3-5 would also be relocated to FSAR Section 16.3.

The proposed changes to the Technical Specifications do not involve a significant hazards consideration because operation of Callaway Plant in accordance with these changes would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

Overall protection system performance will remain within the bounds of the accident analyses documented in FSAR Chapter 15, WCAP-10961-P, and WCAP-11883 since no changes to the response times or measurement interval are proposed.

The RTS and ESFAS will continue to function in a manner consistent with the above analysis assumptions and the plant design basis. As such, there will be no degradation in the performance of nor an increase in the number of challenges to equipment assumed to function during an accident situation.

These Technical Specification revisions do not involve any hardware changes nor do they affect the probability of any event initiators. There will be no change to normal plant operating parameters or accident mitigation capabilities. Therefore, there will be no increase in the probability or consequences of any accident occurring due to these changes.

2) Create the possibility of a new or different kind of accident from any previously evaluated.

As discussed above, there are no hardware changes associated with these Technical Specification revisions nor are there any changes in the method by which any safety-related plant system performs its safety function. The normal manner of plant operation is unaffected.

AttOchment 2 Page 2 of 2 2 '

No new accident scenarios, transient precursors, failure i

mechanisma, or limiting _ single failures are introduced e as-a result of these changes. There will be no adverse effect or challenges imposed on any safety-related

' system as a result of these changes. Therefore, the possibility of a new or different type of accident is ,

not created.  !

3) Involve a significant reduction in a margin of safety. (

No response time changes are proposed in this amendment application; only the document where these limits are  :

~

listed will be changed. _There will be no effect.on the manner in which safety limits or limiting. safety system t settings are determined nor will there be any effect on

-those plant systems necessary_to assure the accomplish-ment of protection functions. There will be no impact on DNBR limits, Fg, F-delta-H, LOCA PCT, peak local  ;

power density, or any other margin of safety.  ;

Based upon the preceding information, it has been determined that.the proposed changes to the Technical Specifications do l not+ involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant ,

reduction in a margin of safety. Therefore, it is concluded l that the proposed changes meet-the requirements of i 10CFR50. 92 (C) and do not involve a significant hazards r consideration.

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A ULNRC-3129 ATTACHMENT THREE ENVIRONMENTAL CONSIDERATION

Attachmtnt 3 Page 1 of 1 ENVIRONMENTAL, CONSIDERATION This amendment application would relocate Technical Specification (TS) Tables 3.3-2, " Reactor Trip System Instrumentation Response Times," and 3.3-5, " Engineered Safety Features Response Times," to FSAR Chapter 16, Section 16.3. The NRC has already implemented this line-item TS improvement in the new Standard Technical Specifications (NUREG-1431 for Westinghouse plants). This amendment application is taken directly from NRC Generic Letter 93-08.

The Bases discussion specific to Table 3.3-5 would also be relocated to FSAR Section 16.3.

The proposed amendment involves changes with respect to the use of facility components located within the restricted area, as defined in 10CFR20. Union Electric has determined that the proposed amendment does not involve:

1) A significant hazards consideration, as discussed in Attachment 2 of this amendment application;
2) A significant change in the types or significant increase in the amounts of any effluents that may be released offsite;
3) A significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10CFR51.22 (c) (9) . Pursuant to 10CFR51.22 (b) , no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

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ULNRC- 3129 ATTACHMENT FOUR PROPOSED TECHNICAL SPECIFICATION REVISIONS e

INDEX .

3 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS .

SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... 3/4 2-1 DELETED 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fg(Z)...................... 3/4 2-4 DELETED 3/4 2-5 ]

N 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F3H............ 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-10 3/4.2.5 DN B P ARAMET E R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 2- 13 TABLE 3.2-1 DNB PARAMETERS..... ................................... 3/4 2-14 3/4.3 INSTRUf1ENTATION i

3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................... 3/4 3-1

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TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION.................... 3/4 3-2 ,

TABLE 3.3-2 RCA" TOR TRIP SYSTCM INSTRUM;NTATIGN REST 0NSE TIMES..... 3/4 3 ~-

bet.s reb TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................................... 3/4 3-9  ;

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION......................................... 3/4 3-13 TABLE 3.3-1 .NGINEERED SAFETY FEATURES ACTUATION SYSTEM I NSTRUMENTAT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 3- 14 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0!NTS....................... 3/4 3-22 ,

TABLE 3.3-5 ' "CTY FEATURES RC;rCNSE TIM 5.............. 3/4 3- G CN0"'

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. TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............ 3/4 3-33 CALLAWAY - UNIT 1 V Amendment No. 28.58

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3/4.3 INSTRUMENTATION,,

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^ 3/a.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDIT?ON FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of . Table 3.3-1 shall be OPERA 8 leg ^2 . --^r::: T .%: ;; :'.n n ' n APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

- I SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM' RESPONSE TIME of each Reactor trip function.

I shall be demonstrated to be within its limit at least once per 18 months.!ZX/fEA7"A Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total nuncer of redundant channels in a specific Reactor trip function as shown in the

" Total No. of Channels" column of Table 3.3-1.

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- CALIAWAY - UNIT 1 3/4 3-1

INSERT A Neutron detectors are exempt from response time testing.

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Call.AWAY - UNIT 1 3/4 3-8 Amendment No. 43

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INSTRUMENTATION ~ i 3/4.3.2 ENGINEERED SAFETY FE&,TURES ACTUATION SYSTEM INSTRUMENTATION

.llMIT!NG CONDITION FOR OPERATiCN  !

3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shail be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip 5etpoint column of

'" ^"'"^"'" " """ - "

Table 3.3-4, " '

APPLICABILITY: As shown in Table 3.3-3.

ACl10N:

a. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Trip Setpoint column I but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip l Setpoint value.
b. With an ESFAS Instrumentation or Interlock Trip Setpoint less l l

conservative than the value shown in the Allowable Valuas column of T able 3.3-4, either:

1. Adjust the Setpo- , consistent with.the Trip Setpoint value of Table 3.3-4 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 i was satisfied for the affected channel, or l
2. Declare the channel incoerable and apply the applicable ACTION statement reouirements of Table 3.3.3 until the channel is estored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 2

  • R + S 5,TA

.W Where:

2 = The value from Column Z of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent span) of rack error j for the affected channel,  ;

S = Either the "as measured" value (in percent span) of the '

sensor error, or the value from Column 5 (Sensor Error) cf Table 3.3-4 for the affected channel, and ,

TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.  ;

c. With an ESFAS instrumentation channel or interlock inoperaDie, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance l

ut the ESFA5 Instrumentation Surveillance Requirements speccified in Table 4.3-2.

4.3.2.? The LNGINEERED SAFETY FEATURES RESPONSE TIME of each E5FAS function shall be demonstrated to be within the limit at least once per 18 months. Each Le t shall include at least one train such that t:oth trains are tested at least l once per 36 munths and one channel per function such that all channels are tested nt least once per N -imes 18 months wnere N is the total numoer of redundant chan-nels in a specific ESFAS f unction as shown in the " Total No. of Channels" Column .

of Table 3.1-3.

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CALLAWAY . UNIT 1 3/4 3-32]a) Amendment No. 22,43 i

INSTRUMENTATION 3

' BASES -

PEACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM lh5IAUMENTAT.'Oh (Cont 1nueo)

Engi eeH Wety F;:: rc; rc;;;r:: t':: ;p;;i'kd in T:ble 2.3 5 d kh 4 n:hde : querti:1 Operatier c' t": -

'"?S' :.nd "CT-v:! v e: (2::: ? :n: ') cc be;ed en scise; :::vmcd in th; n:r-LOCA ;iftty n:1y;;;. 'h;;; On:1y; : .:ke--

cr:dit for inj:: tion of bcr:: d _;t:r 'rcr th; "UST. 'njection of bort:cd

te- R : emed not t: : ur u" tt: VC7 chegi;puptu::ierv:h : :re

+4es. red fellcuing ::ning ' it: "UST :h rgim; pump ;;;thr ::h::. Uh:n th:

ce;ue-th! Oper:thr of the ""ST :nd VCT v:1ve: E n:t in:ht:d ir th; ec;;ca; -

44r:: (St M , t5: v:h:: :p::i' icd are b;;;d on :he LOCA :nciy;c;. The LOCA

!yte tche credi: 'er'nje::hr 'hu re;:rdk:: ef th: :: gree. Verh:t he 4 :::ur: t*:: the 2::r ptMe:

Of :t: re: pen:: :' :peci' icd '- T:b k 2.3 5

d f0r th
LOCA
nd non-LOCf en
1y;;; with rc;p;;; ;; cper icn of :he "CT and "UCT ,el.si e,e ,elid.

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded.

If they are, the signels are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the required logic combination is completed, the syst:m sends actuation signals to

..,.L those Engineered Safety Features components whose aggregate function best -

serves the requirements of the condition. As an example, the following actions J may te initiated by the Enginn red Safety Features Actuation System to mitigate the consequences of a steam iOne break or loss-of-coolant accident: (1) Safety In.iection petis start and automatic valves position, (2) Reactor trips, (3) Feedwa W System isolates, (4) the emergency diesel generators start, (5) containment spray pumps start and automatic valves position, (6) contain-ment isolates, (7) steam lines isolate, (8) Turbine trips, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic velves position, (1)) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency Ventilation System.

CALLAWAY - UNIT 1 B 3/a 3-2(a) Amenoment No. 77,22,E4

r ' \

INDEX '

L-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE -

2L4.2 POWER DISTRIBUTION LIMITS 3 /4.2.1 ' A Xl A L FLU X DI FF E R EN C E . . . . . . . . . .. . . . . . .. . . . . . .. . . . .. . . . . . .. . . . . . . . .. . .. . . .. . . 3 /4 2- 1 DELETED 3/4 2-3

, l 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR -0F (Z) ........................... 3/4 2-4 DELETED 3/4 2-5 N

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F3H......... 3/4 2-8 3/4.2.4 O UA DR ANT POWE R TILT R ATIO . .. . . . . . . . . . ... . . . . .. . .... .. . . . . .. . . . . . . . . ... 3 /4 2 10 i

3/4.2.5 DN B PA R A M ETE R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 2 - 13 TABLE 3.2-1 DN B PA R A METERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2- 14 3/4.3 INSTRUMENTATION l I

3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION .......................... 3/4 3 1 I TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION .................. 3/4 3-2 TABLE 3.3 2 DELETED i

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE l R E Q U I R EM ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3 - 9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM IN STR UM ENTATlO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3- 13 TABLE 3.3 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM j I N STR UM ENTATl ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 3 - 14 j 1

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM l lNSTRUMENTATION TRIP SETPOINTS ............................. 3/4 3 22 l TABLE 3.3-5 DELETED TABLE 4.3 2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS....... 3/4 3-33 CALLAWAY - UNIT 1 - V Amendment No. 28,58 i

. . . . .__ _ . __ _ . - - - . - -- --- . --- I

W 3/4.3 INSTRUMENTATION

, 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION i

LIMITING CONDITION FOR OPERATION  !

)

3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION: i As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automa:ic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table )

4.3-1. .

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function  ;

as shown in the " Total No. of Channels" column of Table 3.3-1.

CALLAWAY - UNIT 1 3/4 3-1

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CALLAWAY - UNIT 1 3/4 3-8

1 Tr r )

A INSTRUMENTATION

. JLil ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

~

- I.lMITING CONDITION FOR OPERATION

-. 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.  ;

APPLICABILITY: As shown in Table 3.3-3. j ACTION:

I

a. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value.
b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than  !

the value shown in the Allowable Values column of Table 3.3-4, either:

1. Adjust the Setpoint consistant with the Trip Setpoint value of Table 3.3 4 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or ,
2. Declare the channelinoperable and apply the applicable ACTION statement -

requirements of Table 3.3.3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + S s; TA Where:

Z = The value from Column Z of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the *as measured" value (in percent span) of the sensor i error, or the value from Column S (Sensor Error) of Table '

3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.

c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION

. shown in Table 3.3-3.

MRVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2. '

4.3.2.2 The ENGINEERED SAF ETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS funct:on as shown in the " Total No. of Channels" Column of Table 3.3-3.

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INSTRUMENTATION  :

2.

BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYST EM INSTRUMENTATION (Continued)-

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are c,ombined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once 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 the consequences of a steam line break or loss-of-coolant accident: (1) Safety injection pumps star; and automatic valves position, (2) Reactor trips, (3) Feedwater System isolates, (4) the emergency diesel generators start, (5) containment spray pumps start and automatic valves position, (6) containment isolates, (7) steam lines isolaf.e, (8) Turbine trips, (9) auxiliary feedwater '

pumps start and automatic valves position, (10) containment cooling fans start and l automatic valves position, (11) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency _

Ventilation System.

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CALLAWAY - UNIT 1 B 3/4 3-2(a) Amendment No. U, 22, 64 L

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