ML20214F519

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Proposed Tech Specs,Extending Applicable Surveillance Intervals
ML20214F519
Person / Time
Site: River Bend Entergy icon.png
Issue date: 05/15/1987
From:
GULF STATES UTILITIES CO.
To:
Shared Package
ML20214F508 List:
References
TAC-65396, NUDOCS 8705260126
Download: ML20214F519 (34)
v • d • e


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- ^ 3/4.3 INSTRUMENTATION' -

.3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION

LIMITING CONDITION FOR OPERATION

' 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERA 8LE with the REACTOR PROTE

~ RESPONSE: TIME as shown in Table 3.3.1-2.

$. . APPLICABILITY: As shown in Table 3.3.1-1.

I , ACTION:

a. With the number of OPERABLE channels less than required by the Mini-mum OPERA 8LE Channels per Trip System requirement for one trip system, 1

place the inoperable channel (s) and/or that trip system in the tripped.

condition

  • within one hour. The provisions of Specification 3.0.4

, are not applicable, j

b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requiremint for both trip systems, 4

place at least one trip' system ** in the tripped condition within one l

hour and take the ACTION required by Table 3.3.1-1. ,

SURVEILLANCE REQUIREMENTS 9

i 4.3.1.1 Each reactor protection system instrumentation channel shall be demon-i strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL 1

TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS-and at the frequencies shown in Table 4.3.1.1-1.

! 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and. simulated automatic operation of

' all channels shall be performed at least once per 18 months."*

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip f functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its p limit at least once per 18 months. Each test shall include at least one channel per trip system 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 g trip system. ,

is8a. *An inoperable channel need not be placedIninthese the tripped oo would cause the Trip Function to occur. cases, condition the inoperable wherechannel this So shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by

,o g l

i nro Table 3.3.1-1 for that Trip Function shall be taken.

j q"8 **The trip system need not be placed in the trippejl condition if this would cause the Trip Function to occur. When a trip syste.n can be placed in the f

n tripped condition without causing the Trip Functicq to occur, place the trip i M systemwiththemostinoperablechannelsinthet7h,$pedcondition;ifboth ma.a. systems have the same number of inoperable channels, place either trip system In the tripped condition. The requirement to place a trip system in the tripped condition does not apply to Functional Units 6 and 10 of Table 3.3.1-1.

"*togic System functional Test period may be extended as identified by note 'p' l

on table 4.3.1.1-1.

3 RIVER BEND - UNIT 1 3/4 3-1

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i TABLE 4.3.1.1-1 b

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

<= CHANNEL OPERATIONAL r l .

J E CHANNEL FUNCTIONAL .

CHANNEL CONDITIONS IN WHICH j FUNCTIt " 1 MIL CHECK TEST CALIBRATION (a). SURVEILLANCE REQUIRED 4

7 i E 1. Intermediate Range Monitors: ,

a a. Neutron Flux - High S/U,5,(b) S/U(c) ,W R 2  :

I w S W R- 3,4,5 l

j b. Inoperative NA W NA 2, 3, 4, S'

2. Average Power Range Monitor: III
  • i a. Neutron Flux - High, S/U.S,(b) S/U(C),W SA 2 Setdown S W SA 3,4,5 i b. Flow Biased Simulated Thermal Power - High S,D(h) S/U(c) ,y y(d)(e) , 3g(o). g(i) 7 l {

S/U(c) ,y g(d) , 3g y l

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{ c. Neutron Flux - High. S

d. Inoperative NA W NA .1, 2, 3,'4, 5
3. Reactor Vessel Steam Dome ~

R IU) IP) 1, 2 0) 4 '

i Pressure - High -S M l

4. Reactor Vessel Water Level -

Low, Level 3 S M R IU) 1, 2

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'5. Reactor Vessel Water Level -

R I8) 1 l- k High, Level 8 S M '

j i 6. Main Steam Line. Isolation 2 Valve - Closure NA M R l1

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7. Main Steam Line Radiation -

2 0) 2 High S M R '1, j - .o

- 8. Drywell Pressure'- High 5 M R I8) 1,2(I) b 4 , ,

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

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TABLE 4.3.1.1-1 (Continued) 1

$ REAC10R PROTECTION SYSTEN INSTRUNENTATION SURVEILLANCE REQUIREBENTS E

-a

) $ (f) The LPRMs shall be calibrated at least once per 1000 effective full pouer hours (EFPH) .. .

j usin W IP system.

(g) Calibrate Rosemount trip unit setpoint at least once per 31 days.

1 c 2

3 (h) Verify measured drive flow to be less than or equal to established drive flow at tne existing flow ..-

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control valve position. -

j H (i) This calibration shall consist of verifying the simulated thermal pouer time constant to be less  !

than 6.6 seconds.

l (j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Soecification 3.10.1.

(k) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or

~3.9.10.2.

(1) This function is not required to be OPERA 8LE when DRYWELL INfEGRITY is not required per Specifica-- '

tion 3.10.1 i '

(2) Verify the Turbine Bypass Valves are closed when THERMAL POWER is greater than or equal to 405 RATED d s THERMAL POWER.

(n) The CHANNEL FUNCTIONAL TEif and CHANNEL CALIBRATION shall include the turbine ~ first stage pressure l [ instruments.

t.

a

  • (o) The CHANNEL CALIBRATION shall exclude the flow reference transmitters; these transmitters sha11 be

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calibrated at least once per 18 months.

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' (p) 'Ihis period may be extended to the first refueling outase, not to exceed 9-15-87. f 1

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INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.* -

4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per trip system 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 isolation trip system. >

  • Isolation System Response Time and Logic System Functional testing period may be extended as identified by notes C and D on Table 4.3.2.1-1.

RIVER BEND - UNIT 1 3/4 3-11

4 TA8LE 4.3.2.1-1 (Continued) 2 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL E CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHiCH 5 CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 4

. TRIP FUNCTION E 6. RH2 SYSTEM ISOLATION Z a. RHR Equipment Area Ambient R 1,2,3 S M

- Temperature - High

b. RHR Equipment Area R 1,2,3 a Temperature - High 5 M
c. Reactor Vessel Water Level - M R(b) 1, 2, 3 S

Low Level 3

d. Reactor Vessel Water Level - R II 1,2,3 S M Low Low Low Level 1
e. Reactor Vessel (RHR Cut-in 1,2,3 .

w Permissive) Pressure - High S M M

R((b)(c)

R bI 1 , 2 ,. 3 S

1 f. Drywell Pressure - High MA 1, 2, 3 NA M 5

m

7. MANUAL INITIATION
  • When handling irradiated fuel in the Fuel Building.
    • When the reactor mode switch is in Run and/or any turbine stop valve is open.

(a) Each train or logic channel shall be tested at least every other 31 days (b) Calibrate trip unit setpoint at least once per 31 days.

4 (c) May be extended to the first refueling outage, not to exceed 9-15-87.

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ATTACHMENT 2 GULF STATES UTILITIES COMPANY RIVER BEND STATION DOCKET 50-458/ LICENSE NO. NPF-47 MAIN STEAM LINE FLOW-HIGH LICENSING DOCUMENT INVOLVED: TECHNICAL SPECIFICATIONS ITEMS: 4.3.2.2 PAGE: 3/4 3-11 Table 4.3.2.1-1 3/4 3-26 3/4 3-29 REASON FOR REQUEST The River Bend Station (RBS) Technical Specifications require many surveillance tests be performed every eighteen (18) months (plus a maximum extension defined by Specification 4.0.2). This proposed change is a request to extend the interval for the subject Surveillance Requirement to the scheduled refueling outage (09-15-87).

The proposed change provided here in accordance with 10CFR50.90 extends the Channel Calibration and Logic System Functional Test surveillance intervals for the Main Steam Line Flow - High instrument loops from 18 months to the first refueling outage. The purpose for this change is to allow the above surveillance to be delayed from its scheduled due date of August 22, 1987 until the first refueling outage scheduled to begin September 15, 1987. This change shows that a surveillance interval up to 24 months (plus a maximum extension defined by Technical Specification 4.0.2) would have no impact on safe operation of the plant. This extension to the first refueling outage will allow for performing these surveillances while the plant is in cold shutdown. GSU has determined, based on the group isolation logic of the isolation actuation instrumentation, that these surveillances should be performed while in cold shutdown due to the high risk of placing the plant in a scram condition. For the plant to shutdown solely to perform surveillances would cause an unnecessary thermal transient on the plant.

GSU requests to amend the subject Technical Specifications contained in Appendix A to the River Bend Station (RBS) Operating License, as discussed below, to perform the subject test during a scheduled refueling outage. Should these proposed changes not be granted in a timely manner, GSU may be forced to implement an unnecessary outage during the first cycle.

DESCRIPTION Technical Specification 4.3.2.1 requires each isolation actuation instrumentation channel be demonstrated operable by the performance of the Channel Check, channel functional test and channel calibration operations of the operational conditions and at the frequencies shown in Table 4.3.2.1-1. For the Main Steam Line Flow -

High, the channel calibration is required every 18 months and a calibration of the trip

x. y unit setpoint'is r.equired at' least- once per '31 ' days. 'A -channel functional test is required every month and a channel check every shif t.

By definition a channel calibration .shall Lbe ~ the ' adjustment, 'as '

necessary, of the channel output such that it- responds with the: '

necessary range and accuracy to known values of the parameter which the-channel monitors. The channel calibration shall . encompass:'the~ entire' ,

channel' including.the sensor and alarm and/or trip. functions, and shall  ;

include the channel functional' test which for analog channels is defined

-as the injection of a simulared signal into the channel as close to the sensor as practicable to verify operability including alarm and/or trip i functions and channel failure trips.

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Technical Specification 4.3.2.2 requires a logic 'systen functional test and simulated automatic operation of all channels be performed at least once per.18 months.

By definition the logic system functional test shall be a' test of all logic components (i.e., all relays and contacts, all trip units, solid ,

state logic elements, etc.) of a logic circuit, from sensor through and  ?

including the actuated device, to verify operability.

The flow in each of the four main steam ' lines is monitored by. four-  ;

differential pressure transmitters, which measure 'the pressure drop r across a venturi type flow element located between the reactor pressure -

vessel and the inboard Main Steam Isolation Valves (MSIVs). . The flow-

. element serves a dual purpose. . It serves as the primary. element for.the measurement of flow, and' is also specifically designed to limit the maximum flow in one steam line, resulting from a postulated guillotine-break of a main steam line, to a rate of flow 170 percent or less of rated mass flow. This choked flow rate is the design. safety , limit.

Main steam line high flow signals are also input into the MS1V and main ateam line drain valve isolation logic. An analytical limit --of 140 percent of rated mass flow is specified for the main steam line flow-high setpoint.

Current instrument setpoint calculations follow the- General ' Electric i (GE) standard setpoint methodology (ref. GE Topical Report NEDC-31336).

Flow element calibration data are used to convert the analytical limit '

of 140 percent flow to differential pressure units -(psid). The technical specification allowable value is calculated from the-analytical limit with allowances for channel. instrument .-accuracy, channel calibration accuracy, and primary measurement variables. These

_ parameters are unaffected by surveillance frequency. The technical specification trip setpoint also includes allowances for channel instrument drift.. The current technical specification trip setpoint of 173 paid has an allowance for 5 psid of drift.- The calculated drift for a 30 month surveillance interval, based on the method outlined in NEDC-31336 is 1.41 paid. Because this calculated drift'is' well within the allowable value there is no required change to the current instrument -

settings, technical specification trip setpoint or allowable value for an extended channel calibration surveillance interval on the main steam line flow - high instrumentation.

As additional justification, note (b), to Technical Specification Table 4.3.2.1-1, requires calibration of the trip unit setpoint at least once per 31 days. Therefore, this requested change only involves calibration of the transmitters, calibration of the trip unit analog meters, gross fail setpoints, and loop setpoints. The channel functional test also required by the channel calibration is performed monthly.

With respect to the requested extension of the surveillance interval for the logic system functional test, an analysis was performed using a fault tree model of the MSIV isolation logic. The basic event failure rates were taken from the BWR individual plant evaluation methodology and are based upon conservative WASH-1400 data. The basic event failure probabilities were determined for surveillance intervals of 18 months and 24 months ThefailureoftheMSIVgeolationlogicwascalculatedto be 1.41-x 10-3/ demand and 1.06 x 10~ / demand for 18 and 24 month intervals respectively. This does not represent an appreciable increase in the probability for failure of the MSIV isolation logic in a high main steam line flow event.

The River Bend Station Final Safety Analysis Report (FSAR) (Section 15.6.4) does take credit for the main steam line isolation function initiated by these instruments during a main steam line pipe break outside containment. However, the function of these instruments is not being altered. Credit is not taken for these instruments in any other FSAR analysis.

NO SIGNIFICANT HAZARDS CONSIDERATION The action (s) specified by this Technical Specification change involves no Significant Hazards Considerations (as defined in 10CFR50.92) as specified below:

1. No significant increase in the probability or the consequences of an accident previously evaluated results from this change because:

The existing Technical Specification Trip Setpoint and Allowable Value can accommodate an additional calculated drif t for a 30 month channel calibration inte rval. Furthermore, the increased LSFT surveillance interval results in no significant probability of an MSIV isolation logic failure.

2. This change would not create the possibility of a new or different kind of accident from any accident previously evaluated because:

This change does not delete or reduce the functional capability of the MSL Flow -

High instrumentation. Therefore, no new kind of accident can result from this change and the response to an event will be as analyzed.

3. This change would not involve a significant reduction in the margin of safety because:

The instrument setpoint is not changed nor should it change as a result of this surveillance extension. The current setpoint has an allowance for 5 psid of drif t. The calculated drift for a 30 month (24 months plus twenty-five percent) surveillance interval is 1.41 paid. Therefore, the calculated drift is well within the allowance for drift as assumed in the analysis. The intent of the Technical Specification basis is met because no significant reduction in the margin of safety or effectiveness of the MSL Flow -

High instrumentation in mitigating the consequences of an MSL break outside containment is involved.

The proposed amendment, as discussed above, has not changed the system l design, function and operation contained in the FSAR and therefore will not increase the probability or the consequences of a previously evaluated event and will not create a new or different event. Also the results of the change are clearly within all acceptable criteria with respect to system components and design requirements, as a result the ability to perform as described in the FSAR is maintained and therefore, the proposed change does not result in a significant reduction in the l

margin of safety. GSU believes that no significant hazards considerations are involved.

l REVISED TECHNICAL SPECIFICATION As indicated above, River Bend Station is currently in compliance with the applicable Technical Specification. This Technical Specification revision is required prior to August 22, 1987 to avoid a unit outage to conduct the required surveillance test discussed above.

NOTIFICATION OF STATE PERSONNEL A copy of this amendment application has been provided to the State of Louisiana, Department of Environmental Quality - Nuclear Energy Division.

ENVIRONMENTAL IMPACT APPRAISAL Revision of this Technical Specification does not result in an environmental. impact beyond that previously analyzed. Therefore, an approval of this amendment does not result in a significant environmental impact nor does it change any previous environtental impact statements for River Bend Station.

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ENCLOSURE' t

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INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERA 8LE by the performani:e of the CHANEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION opera; ions for the OPERATIONAL CONDITIONS and at the frequencies shown in Tablu 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.*

4.3.2.3 The ISOLATION SY. ITEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2-3 Each shall testbeshall demonstrated to beone include at least within its limitper channel at trip least once per 18 months.

system such that all chan:lels are tested at least once every N times 18 months,

,where N is the total numbar of redundant channels in a specific isolation trip Jsystem.

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  • Isolation System Response Time and Logic System Functional testing period may be extended as identified by notes C and D on Table 4.3.2.1-1.

1 RIVER BEND - UNIT 1 3/4 3-11

s~ m TA8LE 4.3.2.1-1 5

. ISOLATION ACTUf. TION INSTRUNENTATION SURVEILLANCE REQUIRE 9ENTS g CHANNEL OPERATIONAL g CHANNEL FUNCTIONAL CHANNEL C00eITIONS IN WHICH ,

, T_ RIP FUliCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

--a

-PRIMARY CONTAINMENT ISOLATION

, g a. Reactor Vessel Water Level -

,. Low Low Level 2 S M R(b) 1, 2, 3 .

R II

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' m b, _Drywell Pressure - High S M 1, 2, 3 -

c. Containment Purge Isolation Radiation - High 5 M R .1, 2, 3
2. MAIN STEAM LINE ISOLATION -

w a. Reactor Vessel Water Level ~- g)

) Low Low Low Level 1 S M R 1, 2, 3 w b. Main Steam Line Radiation - -

y , High S M R 1,2,3 1
c. Main Steam Line Pressure -

Low S M R(b) y i d. Main Steam Line Flow - High S M R II 1, 2, 3 i e. Condenser Vacuum - Low S M R(b) 1, 2**, 3**

l f. Main Steam Line Tunnel Temperature - High S M R 1,2,3

g. Main Steam Line Tunnel A Temperature - High S M R 1,2,3

. h. Main Steam Line Area 3 M R } 1,2,3 i Temperature-High '

(Turbine Bui1 ding) i i

TABLE 4.3.2.1-1 (Continued) .

ISOLATION ACTUATION INSTRUNENTATION SURVEILLANCE REQUIREENTS y '

OPERATIONAL CHANNEL E CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH

$ CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

. TRIP FUNCTION E 6. RHR SYSTEM ISOLATION Z a. RHR Equipment Area Ambient R 1,2,3 5 M i

- Temperature - High

b. RHR Equipment Area R 1, 2, 3 S M a Temperature - High i
c. Reactor Vessel Water Level - M R(b) 1, 2. 3 5

Low Level 3 -

d. Reactor Vessel Water Level - M R(b) 1.2,3 5

Low Low Low Level 1

e. Reactor Vessel (RHR Cut-in M 1, 2, 3 w

Permissive) Pressure - High 5 M

R((b)

R b) 1. 2. 3 5

) f. Drywell Pressure - High MA M M4 1. 2. 3

7. MANUAL INITIATION
  • When handling irradiated fuel in the Fuel Building.
    • When the reactor mode switch is in Run and/or any turbine stop valve is open. s (a) Each train or logic channel shall be tested at least revery other 31 days ,

(b) Calibrate trip unit setpoint at least once per 31 days.

(c) May be extended to the first refueling outage, not to exceed 9-15-87.

ATTACHMENT 3 GULF STATES UTILITIES COMPANY RIVER BEND STATION DOCKET 50-458/ LICENSE NO. NPF-47 LOW REACTOR VESSEL WATER LEVEL LICENSING DOCUMENT INVOLVED: TECHNICAL SPECIFICATIONS ITEMS: 4.3.2.2 Page 3/4 3-11 Table 4.3.2.1-1 3/4 3-26 3/4 3-27 3/4 3-29 REASON FOR REQUEST The RBS Technical Specifications require many surveillance tests be performed every eighteen (18) months (plus a maximum extension defined by Specification 4.0.2). This proposed change is a request to extend the interval for the subject Surveillance Requirement to the scheduled refueling outage (09-15-87).

The proposed change provided here in accordance with 10CFR50.90 extends the interval for channel calibration and logic system functional testing (LSFT) for the Primary Containment, Secondary Containment, and Reactor Water Cleanup (RWCU) Level 2 and Main Steam Line Level 1 isolation actuation instrumentation from 18 months to the first refueling outage.

GSU has and will make a good faith effort to conduct the above listed surveillances on the current frequency if an outage of sufficient duration occurs. The purpose for this change is to allow the above surveillances to be delayed from their scheduled due date of August 16, 1987 until the first refueling outage scheduled to begin September 15, 1987. GSU has determined that these surveillances should be performed while in cold shutdown due to the high risk of placing-the plant in a scram condition. For the plant to shutdown solely to perform surveillances would cause an unnecessary thermal transient on the plant.

GSU requests to amend the subject Technical Specifications contained in Appendix A to the River Bend Station (RBS) Operating License, .n s discussed below, to perform the subject test during a scheduled refueling outage. Should these proposed changes not be granted in a timely manner, GSU may be forced to implement an unnecessary outage during the first cycle.

DESCRIPTION The Technical Specifications require channel calibration and logic-system functional tests for the primary containment, secondary containment, and RWCU Level 2 and Main Steam Line Level 1 isolation actuation instrumentation An evaluation by GSU, with vendor concurrence, has shown that an increase of the calibration interval from 18 up to 24 months does not

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i significantly ;effect' setpoint drift. , Current isetpoints'-allow Jfor-channel drift up to 4 inches of. water,' while.the calculated-drift-for 24:

months is 3.76 inches. .Therefore, the current setpoints and allowable values,'as shown in the. applicable Technical: Specifications, : remain conservative., In addition, the analytical limit. based on the current e FSAR Section.6.3 analysis,jfor a LOCA inside containment, is still valid'

. for the extended surveillance interval.

i-Because.the current design allows for instrument drift greater than that-calculated for ,a.24 month period, no change to the setpoint is required.

Also, because the . surveillance period does.~not .effect instrument

  • accuracy, no change in allowable value is-required, Jnor is there. any.

effect'on the analytical limit or the safety analysis.-

A- review of the FSAR, SER, SRP, design documents and regulatory guides has.not identified any specific requirement to perform;the_LSFT forn the t

! . subject instrumentation every 18 months. Based on this information, Lit.

is concluded that the LSFT was. intended to be conducted during ~ a'-plant l'

outage. A review of plant operating history has found that the logic

- for Level 2 operated successfully-during a loss of feedwater event which
caused . Level 2 to' be reached. This event occurred on 3/1/86 and is.

, reported in LER 86-021. All Level 2 isolations E occurred during a the -

event. This event provides additional assurance that the system is-capable of performing its design function and is reliable.

l. To provide additional consideration to extend the surveillance lutarval,'

j an analysis was performed by GSU using a fault tree model of the Main i Steam Isolation Valve (MSIV) isolation logic. The conclusion. of this.

analysis is -that the extension does not l result:in any appreciable increase in failure probability.. In addition to the low probability of; failure, a review of the Nuclear Plant Reliability Data System (NPRDS) l was performed to gather data on manual switches and auxiliary l relays.

This review found no reports of manual switch failures and a mean time between failure (MTBF) of 239,278 hours0.00322 days <br />0.0772 hours <br />4.596561e-4 weeks <br />1.05779e-4 months <br /> for auxiliary relays. This data 3

and analysis shows that the basic. design 1of the, logic (one-out-of-two-taken twice) is highly reliable and is unlikely to fail.due to the delay in testing.

f Based on the' review of the subject documentation, CSU concludes'that the intent of these technicni specifications will allow for testing during a refueling outage. Review of documentation confirms that an extension to-the surveillance period is acceptable within the.. present requirements.

The performance of the system and its. components remains consistent with

) the design bases, Technical Specification and.FSAR..

SIGNIFICANT HAZARDS CONSIDERATION In accordance with-10CFR50.92, the-following discussions are provided to-support the NRC Staff in its review of. "no~ significant hazards .

considerations".

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1. No significant increase in the probability-or the consequences of an accident previously evaluated results from this . change because:

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'It will'not result in a.significant: reduction in system reliability-nor effect-the ability.of the system to perform its design function.

4 The. -increased _ calibration ' interval' does.-not effectL current-instrument setpoints due to existing design margin. The. system will continue to function within the existing design bases and analysis.

The change in LSFT surveillance interval is supported by successful

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operation of:the instrumentation during startup testing and initial-operation. .In addition, NPRDS: reports no manual switch failures and

-a NTBF 'of 239,278' hours for auxiliary relays;.thus, showing high reliability. ,

{ 2. This change would not create the possibility of.a~new' or-.different-

' kind of accident from any accident previously evaluated because:.

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There is no change to system configuration or analysis. The - change in surveillance interval does' not create any new' types of accidents' - .

- 3. This change would not involve a;significant reduction.in the margin ,

of safety b'ecause:

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The change in calibration test interval does not-impact instrument setpoints. The calculated drift is within the. allowable value' as given in the: Technical Specifications._ Current setpoints allow:for-channel drift of up'to 4 inches of_ water. The calculated drift including this extension is' 3.'76 inches. There is no' change to.

analytical limit used in any analysis. Delay in .th e -l og i c system i functional test does .not significantly effect.the probability of-system failure. Therefore, this - change- does~ not - significantly -i reduce the margin of safety.

This change- is not considered, as stated 3above,- to increase the probability or consequences of a previously analyzed accident or- reduce safety' margins. Further, the results of the change are. clearly within:

all acceptable . criteria with respect to _- the-' system or components specified. The basis for this conclusion is the requested frequency =

, will not affect the ability of the system to ' perform as discussed in the l justification. Therefore, the criteria _for . system performance as discussed in the FSAR have not-been affected. ,

Since the proposed amendment does not change any previously" reviewed and i'

approved description or analysis described in the FSAR,1the proposed i amendment does not . create the possibility of a new or_ different type of accident, 'auul the. proposed change'.does not involve' a significant reduction in a margin of safety.. GSU. believes. that -no significant

- hazards are involved.

. REVISED TECHNICAL SPECIFICATION 1 l

As -indicated above, River Bend Station.is currently in compliance with j the applicable Technical Specification. This -Technical Specification revision is required-prior to August 16, 1987- to avoid a unit outage to-conduct the required surveillance test'as discussed above.

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NOTIFICATION OF STATE PERSONNE_L.

A copy of the amendment application and this submittal has been provided

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to the State of Louisiana, Department of Environmental Quality . Nuclear Energy Division.

ENVIRONMENTAL IMPACT APPRAISAL

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Revision' of this Technical; Specification does not result in an environmental impact beyond that previously analyzed. Therefore, an approval off this amendment does not result. - in . a significant environmental impact nor does: it change any previous environmental impact statements for River Bend Station.-

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. ENCLOSURE f

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INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNrL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

  • 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2.3Eachshall testbe demonstrated shall to beone include at least within its limit channel at trip per least once per 18 months.

system 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 isolation trip system.

  • Isolation System Response Time and Logic System Functional testing period may be extended as identified by notes C and D on Table 4.3.2.1-1.

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RIVER BEND - UNIT 1 3/4 3-11

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TABLE 4.3.2.1-1 5

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS g CHANNEL OPERATIONAL g CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH

, TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED h

-4

1. PRIMARY CONTAINMENT ISOLATION g a. Reactor Vessel Water Level -

Low Low Level 2 S M R(b)(c) 1,2,3 . l

b. Drywell Pressure - High S M R(b) 1, 2, 3
c. Containment Purge Isolation 4

Radiation - High S M R .1, 2, 3

2. MAIN STEAM LINE ISOLATION
a. Reactor Vessel Water Level -

Low Low Low Level 1 S M R(b)(c) 1, 2, 3 y b. Main Steam Line Radiation - * '

g , High S M R 1, 2, 3

c. Main Steam Line Pressure -

Low S M R IU) 1

d. Main Steam Line Flow - High S M R( } 1,2,3
e. Condenser Vacuum - Low S M R I) 1, 2**, 3**
f. Main Steam Line Tunnel Temperature - High S M R 1,2,3 2 g. Main Steam Line Tunnel

! A Temperature - High S M R 1, 2, 3

h. Main Steam Line Area S M. R II 1,2,3 j Temperature High *

(Turbine Building) 4 4

-)

i__________________________ _ _ _ _ _ _ _ _ . _ _ _ _ . . - - - .- -

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TABLE 4.3.2.1-1 (Continued) -

y ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E CHANNEL OPERATIONAL

.$ CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH e TRIP FUNCTION CHECK TEST CALIBRATION SudVEILLANCE REQUIRED E 3. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water -l

" }

Level - Low Low Level 2 5 M R 1,2,3

b. Drywell Pressure - High S M R 1,2,3 -
c. Fuel Building Ventilation Exhaust Radiation - High S M R *-
d. Reactor Building Annulus Ventilation Exhaust '

Radiation - High S M R 1,2,3 t

4. REACTOR WATER CLEANUP SYSTEM ISOLATION

} ,

, y a. A Flow - High S M R 1. , 2. 3 -

U b. A Flow Timer NA M ,

Q 1,2,3

c. Equipment Area Temperature -

High S M R 1,2,3 l d. Equipment Area j

A Temperature - High S M R 1,2,3

~

e. Reactor Vessel Water l
Level - Lcw Low Level 2 S M R IE) 1, 2, 3 ,
f. Main Steam Line Tunnel Ambient Temperature - High S M R 1,2,3

, g. Main Steam Line Tunnel '

A Temperature - High S M R 1,2,3 l

h. SLCS Initiation NA MI *} ' NA 1,2,3 I

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e TARLE 4.3.2.1-1 (Continued)

B ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIRDENTS OPERATIONAL CHANNEL E CHANNEL CONDITIONS IN WHICH CHANNEL FUNCTIONAL SURVEILLANCE REQUIRED 5 CHECK, TEST _

CALIBRATION

. TRIP FUNCTION E 6. RHR SYSTEM ISOLATION

a. RHR Equipment Area Ambient R 1,2,3 Q S M Temperature - High RHR Equipment Area 1, 2, 3
b. 5 M R A Temperature - High
c. Reactor Vessel Water Level - R II 1,2,3 S M Low Level 3
d. Reactor Vessel Water Level - M RII 1,2,3 Low Low Low Level 1 S

b)

)

e. Reactor Vessel (RilR Cut-in 1,2,3 Permissive) Pressure - High S M M

R(IbI R ' 1, 2, 3 i w 5 1 f. Drywell Pressure - High NA 1, 2, 3 NA M

7. MANUAL INITIATION .

4

  • When handling irradiated fuel in t.he Fuel Evilding.
    • When the reactor mode switch is in Run and/or any turbine stop valve is open.

4 (a) Each train or logic channel shall be tested at least every other 31 days; (b) Calibrate trip unit setpoint at least once per 31_ days.

(c) May be extended to the first refueling outage, not to exceed 9-15-87.

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a ATTACHMENT 14

~ CULF STATES UTILITIES COMPANY RIVER BEND STATION DOCKET 50-458/ LICENSE NO. NPF-47 l- REACTOR PROTECTION INSTRUMENTATION SY' STEM i

. LICENSING DOCUMENT. INVOLVED: . TECHNICAL-SPECIFICATIONS ITEMS: 4.3.3.2' PAGE: . 3/4-3-30 C Table 4.3.3.1 3/4~3-41L _

3/4 3-42 3/4 3-43 ,

4.3.5.2 3/4 3-54 Table'4.3.5.1 3/4 3-58~ ,

t 4. 5 .1 ' .3/4 5-5 l REASON FOR REQUEST i The- RBS Technical Specifications. require .many surveillance tests be-performed every eighteen (18) months (plus a maximum' extension defined by Specification 4.0.2). This proposed chan8eJis a request to extend

the interval for the subject Surveillance Requirement to.~ the scheduled e refueling outage (09-15-87).

1 .

i The proposed change.provided here in accordance with 10CFR50.90 extends

the Automatic Depressurization System (ADS) Trip System ; Reactor: Vessell Water ' Level - Low Level 3, Reactor Core Isolation. Cooling (RCIC) System j Reactor Vessel Water Level - High.-Level 8 actuation ' instrumentation-l surveillance frequency from -18 months to the _ first Lrefueling outage.

The purpose of this change is to allow the .above surveillance; to 'be-i delayed from the scheduled due date of August 25, 1987 until the first i refueling outage scheduled to begin-September 15,; 1987. : GSU has and -

, will make a good faith' effort to_ condu'et this surveillance on_the current frequency _if an outage of :sufficienti duration ' occurs. - :- The-nuclear boiler instrument design is sensitive to disturbances-of'the.

instrument sensing lines associated with level transmitters 1B21*LTN095A and 1B could cause interactions between Reactor Protection System (RPS),.

Nuclear Steam Supply System (NSSS), Low Pressure. Core _ Spray (LPCS) -and= .

Residual Heat Removal System (RHR) _ trip' systems.' _ During previous

i calibrations of instruments that. share common sensing legs,1 RivervBend- j Station has experienced spurious level / pressure transients which have- 1 resulted in ECCS initiations and reactor - scrams. Therefore, GSU has determined -that these surveillances should be performed while in cold- '

shutdown due to' the high risk of placing;the plant'in a scram condition. >

For- the plant to shutdown solely to perform surveillances,'would cause.

an unnecessary thermal transient on the-plant. GSU_-requests to amend' l the subject Technical Specifications contained in Appendix A to the- 7 River Bend Station (RBS) Operating License, as discussed; below, 'to perform the subject test during a scheduled refueling outage.--Should these proposed changes not be granted in a timely manner, _ GSU may be forced to implement an unnecessary outage during the first cycle.-

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L1 3 DESCRIPTION' l' The ADS relieves reactor ' pressure. steam to'the suppression pool by..

,.~ activating seven of the -sixteen, safety l relief ~ valves -(SRV)'.

Level i' transmitters 1B21*LTN095A and B.are associated with the ADS Channel A of~

-SRV Division I and Channel B of SRV:DivisionLII and- provide a' low: level confirmatory signal upon reaching the reactor vessel low water (Level 3)-

~

g -setpoint. The low water level m confirmatory; signal- prevents spurious

=

l initiations due.,to testing or malfunctions; -The ADS is,not initiated'

i. until other permissives areL satisfied .(i.e. , low-low-low - reactor water
level'(Level 1) and high drywell pressure).

3 The'same level transmitters which monitor reactor vessel low water level also actuate on reactor vessel;high water-level (Level 8) to- provide a -

L

'RCIC trip' by. closing the RCIC turbine steam supply valve. .The RCIC-

- turbine is tripped when the reactor vessel high water level setpoint _ is _

reached' to protect against moisture carry-over and subsequent turbine-damage.

s The' River Bend Station Final Safety Analysis, Report (FSAR),. Section '

15.6.4 and Section' 6.2.1.1.3.1.6,- take credit for the ADS confirmatory

(Level 3) ' function performed by these _ instruments _during a . main steam. -

i line break 1outside containment and.an intermediate size break accident.. '

i Credit is not taken in any other FSAR analysis. : Credit.for the Leve1~ 8 i RCIC trip is not assumed in any FSAR analysis;,

Technical Specification Table 4.3.3.1-1, "ECCS Actuation Instrument.

1, Surveillance Requirements", and Table 4.3.5.1-1, "RCIC System Actuation' i

- Instrumentation -Surveillance Requirements". specify that the Emergency- .

Core Cooling System (ECCS)/RCIC actuation ADS trip system reactor-. vessel

t. water level low (Level 3) instrumentation--channel calibration be;

! performed at least- once per'_18 months.. The ' current _ setpoint and I allowable value allowance of maximum- instrument: drift is 1-inch of- '

wa ter. : GSU, with vendor concurrence..has reviewed-the; system design and found the calculated drift for a 24 month ~ surveillance _ period to be 0.86 inches of water. Since the 24 month period -envelopes. the _ requested change period and the calculated drift is within present system. design j

allowances, the present technical specification'setpoint and allowable- I values are not changed. In addition, the current FSAR analysis remains unchanged for the extended surveillance period.

A logic system functional test is required .by Technical Specification ~

i Surveillance Requirements Sections 4.3.3.2 and-4.3.5.2.- The ADS trip i system logic. system functional test (LSFT) applies a -known. pneumatic 7

test pressure 'to the sensor and ensures proper. operation in the ADS logic to enable-the opening of the SRV's.

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The system is designed so-that a single active or passive component failure does not disable the1 1

[ ADS. . In conjunction with redundant transmitter design, - the low 'leveli j j (3)- ADS permissive signal -(one-out-of-two- design logic)' assures l continued function of these channels. :The non-ECCS reactor vessel high.

water level 8 RCIC. LSFT decreases the Epressure at.the sensor and -i verifies that relays energize to close- the- RCIC ' steam supply _ valve.

This trip system design employs a two-out-of-two!1ogic. Credit-for the- '

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clevel.8 trip;is not assumed in any FSAR analysis. The_ performance of a cmonthly; functional: test, as required-by Technical. Specifications,nwilli

~

continue to provide testing from_the trip-units to relay contacts. "The

sensor.- operability. . is Lverified . daily . by ' channel- checks.- GSU has-

-reviewed component reliability data via.the . Nuclear Plant Reliability Data System (NPRDS)'.and found the LSFT and auxiliary' relays to be highly' f reliable and not sensitive :to the test-interval.

> Based on the review.of the' subject documentation, GSU fconcludes that the -

intent c of . these Technical' Specifications is .not 'affected by a 21 day-

. extension and the required-testing may be performed- during the ifirst; 4

refueling. ou: age. A review of the RBS FSAR, and_ applicable Regulatory:

Guides and IEEE Standards confirms that an extension oflthe surveillance- i period is acceptable within the present guidelines. The. instruments.and.

system design .is not adversely affected_-byt the extension. of 'the f- surveillance interval and the performance of, the : system and_.its f  ; components' remain consistent with the current- design bases,_-Technical Specifications and FSAR.-

NO SIGNIFICANT HAZARDS CONSIDERATION' ,

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, The revision requested by this LTechnical Specification. change involves

no significant hazards as defined in 10CFR50.92 and discussed below:
1. No significant increase in the probability or the consequences of'an
accident previously evaluated results from this change 1because

1- As discussed above, the requested l change. to' thef. surveillance) interval has been found to be within the present design value for

q. the setpoint drift and will remain :within . technical. specification 1 allowable values for the requested extension. The change is;also

! found to have no significant effect.on the_-system logic . function .

because of the system design and reliability of the components. In "

i addition, NPRDS reports show the LSFT and-_ auxiliary relays to .be highly reliable and not sensitive to test interval.- Therefore, 4

there is no change to the current safety analysis required ~ .

j- 2) This change .would not create the possibility of- a new or different i kind of accident from any accident previously evaluated because:

  • 1 s

1- The increase in the reactor water level instrumentation. surveillance '

interval does not increase the possibility of an accident or a malfunction of a different type than .previously evaluated ~since

.there is no change in function or hardware.-

1

3) This change would not involve a significant: reduction'in the margin

~

l of safety because:

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The change in this reactor vessel water: level instrumentation

surveillance interval does not involve a reduction in the margin. of safety since the instruments setpoints and allowable values are not 3

changed and.the calculated drifts are well- within. the allowable

, values. Since the change maintains the present safety analysis, there is no significant reduction to the margin of safety.

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_ _ ._ ,, . ._.a,. _ , _ a ,. . . :;; . -. ._-. _, _

This change is not considered, as stated above, to increase. the probability or consequences of a previously analyze'd accident or reduce safety margins; further the results of the change are clearly within all acceptable criteria with respect to'the system or components specified.

The basis for this conclusion is the requested frequency will not affect the ability of the system to perform as discussed in the justification.

.Therefore, the criteria for system performance as discussed in the FSAR.

have not been affected.

Since the proposed amendment does not change any. previously reviewed and approved analysis described in the FSAR, the proposed amendment'does not create the possibility of a new or different type of accident, and the proposed change does-not involve a significant reduction in a margin of safety. GSU believes that no significant hazards are involved.

REVISED TECHNICAL SPECIFICATION As indicated above, River Bend Station is currently in compliance with.

the applicable Technical Specification. .This Technical. Specification revision is required prior.to August 25, 1987 to avoid a unit outage.to.

, conduct the required surveillance test as discussed above.

NOTIFICATION OF STATE PERSONNEL ~

A copy of the amendment application and this submittal has been provided-to the State of Louisiana, Department of Environmental Quality - Nuclear Energy Division.

ENVIRONMENTAL IMPACT APPRAISAL l Revision of this Technical Specification does not result. in an environmental impact beyond that previously analyzed. Therefore, an approval of this amendment does not result in a significant environmental impact .nor does it change any previous environmental impact statements for River Bend Station.

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ENCLOSURE 1

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I INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown-in Table 3.3.3-3.

APPLICABILITY: As shown in Table 3.3.3-1.

ACTION:

a. With an ECCS actuation instrumentation channel trip setpoint less ,

conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is

. restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. '

b

b. With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.
c. With eitner ADS trip system "A" or "B" inoperable, restore the inoperable trip system to OPERABLE status:
1. Within 7 days, provided that the HPCS and RCIC systems are OPERABLE, or
2. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, provided either the HPCS or the RCIC system is inoperable.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />'and reduce reactor steam dome pressure to less than or equal to 100 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and .

CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3.1-1. )

l 4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of i all channels shall be performed at least once per 18 months. 88 1

4.3.3.3 At least once per 18 months",'.the ECCS RESPONSE TIME of each ECCS trip l function shown in Table 3.3.3-3 shall be demonstrated to be within the limit.

Each test shall include at least one channel per trip system such that all j channels are tested at least once every N times 18 monthsf#where N is the total l number of redundant channels in a specific ECCS trip system.

88 Logic System Functional and EC,CS Response time testing period may be extended as identified by note d b and c on Table 4,3.3.1-1.

RIVER BEND - UNIT 1 3/4 3-30

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TABLE 4.3.3.1-1 g EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUNENTATION SURVEILLANCE REQUIREMENTS

, m

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$ CHANNEL OPERATIONAL

$ CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH

, TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

-k A.* DIVISION I TRIP SVrTEM

-4 1. RHR-A (LPCI MV I) AND LPCS SYSTEM H a. Reactor Vessel Water Level -

Low Low Low Level 1 S M R 1,2,3,4*,5*

b. Drywell Pressure - High S M R 1, 2, 3
c. LPCS Pump Discharge Flow-Low g)

S M R 1, 2, 3, 4*, 5*

d. Reactor Vessel Pressure-Low S M Rg ,) 1, 2, 3, 4*, 5*

(LPCS/LPCI Injection Valve Permissive)

~

e. LPCI Pump A Start Time Delay w

Relay NA M Q g) 1, 2, 3, 4 * , 5*

g f. LPCI Pump A Discharge Flow-Low S M R 1,2,3,4^,5*

l m g. LPCS Pump Start Time Delay NA M Q 1, 2, 3 , 4 * , 5*

i Relay H h. Manual Initiation NA- R NA 1, 2, 3, 4* , 5*

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"# ,
a. Reactor Vessel Water Level -

1 Low Low Low Level 1 S M R 1, 2, 3 i

b. Drywell Pressure-High S M R 1, 2, 3 4' -
c. ADS Timer NA M Q 1,2,3 j d. Reactor Vessel Water Level - -

) Low Level 3 S M RI "} ) 1, 2, 3 i e. LPCS Pump Discharge l , Pressure-High S M R(*) 1, 2, 3

f. LPCI Pump A Discharge
  • j Pressure-High 'S M R(* 1,2,3
g. ADS Drywell 9ressure Bypass NA M Q 1,2,3

. Timer

h. ADS Manual Inhibit Switch NA M NA 1, 2, 3

{_ i. Manual Initiation NA R . NA .

1, 2, 3 r

. . = _ _ _ _ _ .

, TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

' E CHANNEL OPERATIONAL 5 CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH ,

a TRIP FUNCTION CHECK TEST CALIBRATION ' SURVEILLANCE REQUIRED B. DIVISION II TRIP SYSTEM

1. RHR 8 AND C (LPCI MODE)
a. Reactor Vessel Water Level -

Low Low Low Level 1 5 M 1, 2, 3, 4*,.5*

.b. Drywell Pressure - High S M Rf"

.R(,) 1, 2, 3 i

c. Reactor Vessel Pressure-Low S M R, 1, 2, 3, 4*, 5*

(LPCI Injection Valve Permissive)

d. 'LPCI Pump B Start Time Delay '

w Relay ~ NA M 1, 2, 3, 4 * , 5*

} e. LPCI Pump Discharge Flow-Low Qg ,)

S M R 1, 2, 3, 4*, 5*

w f. trCI Pump C Start Time Delay NA M Q 1, 2, 3, 4*, 5*

1 Relay

g. Manual Initiation NA R NA 1, 2, 3, 4*, 5*

i 2. AUTOMATIC DEPRESSURIZATION SYSTEM '

i TRIP SYSTEM "B"#

' a.

' " Reactor Vessel Water Level - I Low Low Low Level 1. S M R 1,2,3 l b .' Drywell Pressure-High S M R 1,2,3  :

j c. ADS Timer NA. M Q 1, 2, 3  ;

4

'd. Reactor Vessel Water--Level - "

i Low Level 3 S M R(a)(b) 1,2,3  ;

e. LPCI Pump 8 and C Discharge .

Pressure-High .

S M R(,) 1,2,3

f. ADS Drywell Pressure Bypass Timer NA M Q 1, 2, 3
g. ADS Manual Inhibit Switch NA M NA 1,2,3
h. Manual Initiation NA R NA 1,'2, 3

}

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

)

+'

TABLE 4.3.3.1-1 (Continued) h EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS S CHANNEL OPERATIONAL.

g CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH g TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED' h

C.

DIVISION III TRIP SYSTEM ~

p 1. HPCS SYSTEM

>" a. Reactor Vessel Water Level -

Low Low Level 2 S M 1, 2, 3, 4 ^ 5* -i

b. Drywell Pressure-High S M Rf*.

R 1,2,3

' c. Reactor Vessel Water Level-High Level 8 5 M - RI ") 1, 2, 3, 4*, 5*

d. Condensate Storage Tank Level -

Low S M 'RI ") 1, 2, 3, 4*, 5*

e. Suppression Pool Water

' ~ '

. . Level - High S. M R 1, 2, 3, 4 * , 5*

g f. Pump Discharge Pressure-High S M R,,) 1, 2, 3, 4*, 5" w g. HPCS System Flow Rate-Low S M E- 1,2,3,4*,5*

4 h. Manual Initiation NA R NA 1, 2, 3, 4 * , 5"

, D. LOSS'0F POWER

. 1. Divisions I and II

a. 4.16 kw Standby Bus Under- S M R 1, 2, 3, 4^*, 5**'

voltage (Sustained Under-4 voltage) .

b. 4.16 kw Standby Bus Under- .5 M R 1, 2, 3, 4**, 5**-

+

voltage (Degraded Voltage) .

4

}- 2. Division III

a. 4.16 kv Standby Bus Under- S NA R 1, 2, 3, 4**, 5**

voltage (Sustained Under- .

voltage) ,

j 'b. 4.16 kv Standby Bus Under-voltage (Degraded Voltage) S M R 1, 2, 3, 4 * * , 5*

  • l p # Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

When the system is required to be OPERABLE per Specification 3.5.2.

+ ** Required when ESF equipment is required to be OPERABLE.

i (a) Calibrat? trip unit setpoint at least once per 31 days.

l (b) . Channel calibration and LSFT may be extended to the first refueling outage, not to exceed 9-15-87

m INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION s LIMITING CONDITION FOR OPERATION 3.3.5. The reactor core isolation cooling (RCIC) system actuation instrumenta-tion enannels shown in Table 3.3.5-1 shall be OPERABLE with their trip set- _

points set consistent with the values shown in the Trip Setpoint column of  :

Table 3.3.5-2. l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 with reactor steam dome pressure greater than 150 psig.

?

ACTION:  !

l

a. With a RCIC_ system actuation instrume'ntation channel, trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.5-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. .
b. With one or more RCIC system actuation instrumentation channels inoperable, take the ACTION-required by Table 3.3.5-1.

SURVEILLANCE REQUIREMENTS 4.3.5.1 Each RCIC system actuation instrumentation channel shall be demon-strated OPERABLE by the performance of the CHANNEL CHECK. CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown.in Table 4.3.5.1-1. i 4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.# '{

)

1

  1. Channel calibration and LSFT period may be extended as identified by note b - .l on Table 4.3.5.1-1;  !

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._ )

RIVER BEND - UNIT 1 3/4 3-54 s

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TABLE 4.3.5.1-1 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

  • $ CHANNEL 4' $ CHANNEL ' FUNCTIONAL CHANNEL

, FUNCTIONAL UNITS CHECK TEST CALIBRATION i C l $ '

a. Reactor Vessel Water Level -

Rg ,)

H Low Low Level 2 S M

  • w .
b. Reactor Vessel Water S M RI ")

Level - High Level 8

c. Condensate Storage Tank -

Level - Low S M R(*)

Suppression Pool Water Level.-

d.  ?

w High S M- R(,)

- N i

w e. Manual Initiation NA R NA j (a) Calibrate trip unit setpoint at least once per 31 days.

(b) Channel calibration and LSFT may be extended to the first refueling outage to begin 9-15-87.-

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- - _ _ _ _ _ __L_____:__-_-- _

"~~

EMERGENCY CORE COOLING SYSTEMS'

~

SURVEILLANCE REQUIREMENTS (Continued) e .- At least once per 18 months for the ADS by:

1. Performing a system. functional test which includes simulated automatic actuation of the system throughout_its emergency.

operating sequence, but excluding actual valve actuation.**

2. Manually opening each. ADS valve when the reactor steam dome pressure is greater _than or equal;to 100 psig* and observing l that:

a) The control valve or' bypass valve position responds accordingly, or-

b)~ There is a corresponding change in the measured _ steam i

. flow, or c) The acoustic; monitoring system indicates the valve'is open.

i i

f i

"The provisions of Specification 4.0.4 are not applicable provided the surveillance f:s performed within'12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

    • May be extendad to the completion of the first refueling outage.

RIVER BEND - UNIT 1 , 3/4 5-5

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