ML20115C114

Proposed Tech Specs Revising LCO 3.7.1.2, Auxiliary Feedwater Sys, & Associated Bases

ML20115C114
Person / Time
Site:	Beaver Valley
Issue date:	10/05/1992
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20115C096	List: ML20115C095 ML20115C101 ML20115C114
References
NUDOCS 9210190158
Download: ML20115C114 (22)

Text

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ATTACHMENT A i

Beaver Valley Power Station, Unit !!o. 1 Proposed Technical Specification Change No. 19 "d

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Revise the Technical Specification as follows:

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DPR-66 PJJd4T SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three staan generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with a.

Two feedwater

pumps, each capable of being powereu from separate emergency busses, and b.

One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY:

MODES 1, 2 and 3.

ACTIQll:

a.

With one auxiliary feedwater pump inoperable, restore at least three auxiliary feedwater pumps (two capable of being powered from separata emergency busses and one capable of being powered by an OPERABLE steam supply system) to j

OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be

!n 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 />.

b.

With the motor driven auxiliary feedwater pump supplying the redundant header inoperable, realign the two remaining auxiliary feedwater pumps to separate headers within 2 ADO u o 7t4SEU hi hours.

4-SURVEILLANCE REQUIREMENTS l

! 4.7.1.2 Each auxiliary feedwiter pump shall be demonstrated i

OPERABLE l

l a.

When tested pursuant to Specification 4.0.5:

1.

By verifying, that on recirculation flow the acove required motor driven pumps develop a

discharge pressue greater than or equal to 3.155 psig.

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2.

By verifying, that on recirculation flow the above required steam turbine driven pump develops a

discharge pressurs greater than or equal to 1155 psig When the secondary steam pressure it greater than 600 pr19 w

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BEAVER VALLEY - UNIT 1 3/4 7-5 Amendment No.

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Attachment to " Auxiliary Feoclyater Syriem" Insert "A" 4.7.1.2 Each auxiliary fecawater pump shall be demonstrated

. OPERABLE a.

When tested pursuant to Specification 4.0.5:

1.

By verifying,.'that the pump's developed head at the flow tost point is greater than or equal to the required developed -head as specified in the Inservice Testing Program.

The provisions of Specification 4.0.4 are not applicable for ent for the steam turbinedrivenpumptesting.(()intoMode3 I

Insert "B"

c.

With two auxiliary feedwater pumps inoperable, be in at least

!!OT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 11oT SilUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, d.

With three auxiliary feedwater pumps inoperable,-immediately initiato corrective action to restore at least-one auxiliary feodwater pump to OPERABLE status as soon as possible.

1 BEAVER VALLEY UNIT 1:

(Proposed Wording, we

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DPR-66 Pl. ANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days by:

1.

Verifying thet nach valva (manual, power operated or automatic) in the flow path that is not

locked, sealed, or otherwise secured in position, is in its correct position.

2.

Reverifying the requirements of Technical Specification Surveillance 4.7.1.2.b.1 by a second and independent operator.

3.

Establich and maintain const.mt communications between the control roos and the auxiliary feed pump room while any normal discharge valve is closed during surveillance testing.

4.

Verifying operability of each River Water Auxiliary supply valve by cycling each manual River Water to Auxiliary Feedwater System valve through one complete OELE7E Y 1**

c.

Tclieving-- :n e::t:nded pl:nt Out:@ Verify - Auxiliary Feedwater flow from WT-TK-10 to the $ team Generators h

the Auxiliary Feedwater Valves in their normal alignment d.

At least once per 18 months during shutdown by:

A00 1.

Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.

2.

Verifying that each automatic valve in the flow path actuates to its correct position on a test signal.

3.

Verifying that each pump starts automatically upon receipt of a test signal.

ADD v6 E d Suf\\ltib wce. sS ce red 4e be pederwed pher de e,h mho Mohe. A wkoneuce W. phA h s been gr**br ba 3 o ccabw" kg gw Medd 5 or 6 for BEAVER VALLEY - UNIT 1 3/4 7-6 Amendment No.-

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4 DPR-66 PLANT SYSTEMS BASES

_n SAFETY VALVES (Continuedt U=

maximum number of inoperable safety valves per operating steam line Power Range Neutron Flux-High Trip Satpoint for (N)

(109)

=

loop operation (W) =

71 percent of RATED T!!ERMAL POWER permissible by P-8 Setpoint for 2 loop operation with stop valves open.

=

66 percent of RATED THERMAL POWD; permissible by P-8 (W)

Setpoint for 2 loop operation with stop valves closed.

Total relieving capacity of all safety valves ner X

=

stena line in lbs/ hour (4,261,666)

Maxiuum relieving capacity of one safety valve in Y

=

lbs/ hour (873,600) 3/4.7.1.2 AUXILIgY FEEDWATER PUMPE The OPERADILITY of the auxiliary feedwater pumps ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of off-site power.

3, --DELETE hh--electeic driven-euxi+1ery--feedwater ptmp is -capable Of GreM4ering

- tota-1 feedwater-6-low-of-350 gpm-st-6 pressure Of 11;3 i

psig-to--the--entrance c,f steam generators.

Tite--steem drivan auxil-iary feedwater-pump-14-eapable-of-deMvering -:- t-et+1-feedwater )

f-low-Of 700 gpm--at-a-pressur-e-of-1-14-psig-to-the-enti$anoc c f thjl,_

stean-- ganuater c.

Thi capacitykQs sufficient to ensure that._.

adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant Systen temperature to less than 350*F when the Residual Heat 3emoval System may.be placed into operation.

3/4.7.1.3 PRIMARY PLANT DEMINERALllEP W ATER

• • ( fy The OPERABILITY of the PPDW storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 350'F in tha event of a total loss of off-site power.

The minimum water volume is sufficient to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to atmosphere.

BEAVER VALLEY - UNIT 1 B 3/4 7-2

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ATTACHMENT B Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 192 REVISION OF TECH!!ICAL SPLCLFICATIO!! 3.7.1.2

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DESCRIPTIO!i OF AMENDMENT REQUEST The propoced amendment would add two additional action statements "c"

and "d"

to Limiting Condition For Operation (LCO) 3.7.1.2.

The proposed action statement "c" deals with the condition when i

two auxiliary feudwate- (AFW) pumps are inoperable.

The proposed action statement "d"

deals with the condition when all three auxiliary feedwater pumps i s inoperable.

Surveillance Requirement (SR) 4.7.1.2.a would be revised by deletion of the specific auxiliary feedwater pump parameter for discharge pressure.

The proposed wording wnuld require that the auxiliary feedwater pumps be debonstrated operable by testing pursuant to Specification 4.0.5.

The testing will include verification that the pump's daveloped head at tho flow test point is greater than j

or equal to the required developed head as specified in the Inservice Testing (IST)- Program.

Additionally, an exception to specification 4.0.4 will be-added for the steam turbine driven auxiliary feedwater pump testing.

A note would be added to the bottom of the page which states that the steam turbine driven pump testing shall be conducted only when secondhry side steam prescure is greater than 600 psig.

Surveillance requirement 4.7.1.2.c would be revised to clarify that this curveillance requirement applies when the plant is in Modes 5 or 6 for greater than 30 continuous days and must ba performed prior.to entry into Mode 2.

The Bases for Limiting Condition for Operation 3.7.1.2 would be revised by deleting the specific pump paraueters for flow and pressure delivered to the steam generators.

B.

BAChJROUND The current SR 4.7.1.2.a wording requires that a

spccific discharge pressure be obtained-for ca sh auxiliary feedwater

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pump.

The proposed change to SR 4.7.1.2.a deletes these specific values and will allow tha required developeo head and flow raten to be established and periodical'v rec <aluated through our procedures for ASME Section XI testing, The AFW pumps are equipped with minimum flow recirculation lines which include orifices large enough to ensure that the pumps will not overheat-when run with all other flow paths closed.

This minimum flow was the basis for the current. pump dicchargo pressurs specified in SR 4.7.1.2.a.

The currcnt value ci 1155 psig discharge pressure was found to be in error and has been administratively raised to eN to that the AFW pumps will perform as assumed in the acciuent analyses.

Review of past AFW pump performance data has suown acceptable pump operation when evaluated against the new acceptance criteria.

The value for pump discharge pressure specified in the current wording for SR 4.7.1.2.a is derived by first determining the minimum operating point (MOP) for the required pump head and flow as assumed in applicable safety analysis for the AFW pumps.

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ATTACHMI:NT B, continued Prteosed Technical Specification Change No. 192 Page 2 j

the MOP is determined, a

value for purp discharge pressure is i

"j determined for the full recirculation conditions Jpecified in SR 4.7.1.2.a by following the shape of pump performance curve of record from the MOP to that specific point on the curve.

Once 4

this value for pump discharge pressure for full recirculation conditions has been determined and. incorporated into.SR 4.7.1.2.a, the pump (s) must always be run at this test point for the purpose of demonstrating pump operability.

4 i

The proposed change to SR 4.7 1.2.a would delete the reference to i

this specific tort point.

The proposed wording for SR 4.7.1.2.a allows the AFW pumps to be tested at any point from complete i

recirculation to full flew conditions.

The curve which will i

define whether the AFW pumps develop the specific head required i

to meet safety analyses will be the individual MOF curve for each AFW Pump.

Each MOP curve will be contained in the IST Program 4

and controlled in accordance with program. requirements.

A draft revision to the IST program is contained in Attachment B-1 for 4

informational purposes.

Future changes to those MOP curves, if this amendment request is

approved, will be made as necessary i

l through the 10 CFR 50.59 process and will be sent to the Nuclear j

Rc qulatory Commission (NRC) as part of revision updates to the IST Program.

l The proposed addition of action statement "d"

is for the condition where all three AFW pumps are inoperable.- The plant is-i in degraded condition with no safety related means for conducting a

cooldown, and only limited means for. conducting a cooldown with non-safety related equipment.

In such a condition, the plant should not be perturbed by any action that might result l

in a reactor trip, t

f The proposed exception to Specification 4.0.4 for entry into Mode 3

for the ateam driven AFW pump is necessary due to incufficient-amount of steam in Mode 4, 5 or 6 to perform a valid test.

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JUSTIFICATION 7ho proposed addition of action stat mont "F is consistent with-the Standard Technical Specifications (Sis) and DVPS Unit '

Technical Specifications.

The proposed action statement "c" will allow the plant to be shutdown.to Mode.4 under-the guidance.of the LCO action statement when two AFW pumps are inoperable rather than being required' to enter LCO-3,0.3.

LWithout.the proposcd' action statement "d",

the plant would enter LCO 3.0.3 which would.

require the plant to commence a shutdown within one hour ~without a

safety.

related means-for conducting theJcooldown. -Without4any-AFP

pumpc,

.the plant is in a safer condition by not attempting a plant cooldown 'than it would. be by entering' LCO '3.0.3'and 4

possibly resulting. in a

reactor trip without a safety related means for conducting the cooldown.

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' ATTACHMENT B, continued Proposed Technical Specification Change No. 192 Page 3 The proposed exception to Specification 4.0.4 for-entry into Mode 3

for the steam driven AFW pump is consistent with STS and BVPS Unit 2

Technical Specifications.

This proposed exemption is necessary since there is insufficient amount of steam in Modes 4, 5 or 6 to perform a valid test.

I The proposed addition of the note which states that-the steam turbine driven pump testing chall be conducted only when secondary side steam pressure is greater than 600 psig is consistent with existing wording and is necessary due to the 1

specific design of this turbine driven pump to ensure valid test i

l results.

The proposed revision to SR 4.7.1.2.a will' continue-to verify that each AFW pump will develop the specific head and flow necessary to demonstrate acceptable pump operation.

The specific criteria for pump head and flow will be controlled in accordance with the IST Program requirements.

Any future changes to'these pump head and flow requirements will-be made under the 10 CFR-50.59 process and will be forwarded-to the NRC as part of_the IST Program revision process.- This will reduce the need to-submit a request for a technical specification change on this Surveillance Requirement due to changes in plant _analysos or changes in pump perforrance characteristics due to pump overhaul.

This is consistent with the NRC's policy-on Technical Specification i

improvements.

The proposed change to SR 4.7.1.2.a will also allow the AFW pumps to be tested for the purposes of satisfying SR 4.7.1.2.a at full flow conditions as

lant conditions allow.

Full-flow pump testing is similar to conditions when the pump (s) are performing their safety function.

With the current wording contained in SR

~

4.7.1.2.a, the AFW pumps must be run in full recirculation conditions -since the surveillance specifies only one-flow test point.

In certain cases, the AFW pump testing must be performed more than once per specific surveillance interval'due to the specific test conditions which must be met _when performing the current SR 4.7.1.2.a.

With the proposed wording of SR 4.7.1.2.a when an AFW pump is running for any reason, SR 4.7.1.2.a can be performed ac necesscry at that flow condition.

This will result 4

in less AFW pump running time for the purpose of satisfying SR 4.7.1.2.a.

AFW pump run time and wear would be reduced due.to this added testing flexibility.

l The proposed revision to SR 4.7.1.2.c soul'd clarify that this surveillance requirement applies.when the-plant is in Modes 5 or 6

for-greater than 30 continuous days and must-be performed prior to entry. into Mode.

2.

The existing wording only states the surveillance is applicable following an extended plant outage and does not define what. length of time an_ extended plant outage i

involves.

The proposed wording will clearly specify when this surveillance must be performed, and is consistent with the'STS.

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' ATTACHMENT B, continued i

Proposed Technical Specification Change No. 192 Page 4 4

i D.

SAFETY ANALYSIS The AFW pumps will continue to be tested in a manner which will i

demonstrate that they will delivo; sufficient foodwater to the steam generators to remove decay heat from the reactor coolant j

system upon the losc of normal feodw. tor supply.

The proposed i

changes to SR 4.7.1.2 will not affect the ability of the AFW system to perform this function.

Changes to the A7W pump head i

and flow requirements will be made under the 10 CPR 50.59 process and controlled under the IST progrrm administrative requirements.

Therefore, changes to these specific pump parameters will be controlled under a process uhich will continue i

to ensure safe plant operation.

The added flexibility cf i

allowing the AFW pumps to be run at various-flow conditions for the purposes of satisfying SR 4.7.1.2.a W.ill reduce the time the l

pumps are run on recirculation flow and total pump running time.

This will reduce pump wear as a. result of these two factors.

The addition of the third action statement will address the condition i

when two AFW pumps are inoperable.

This change to.the action statements is administrative in nature.

The plant will continue to be. placed in a mode where the L.C.O.

does not apply for the i

condition when two AFW pumps are inoperable._ Therefore, l

safety will not be affected by thic proposed change.

' plant 4

The addition of a fourth action statement will ensure that when no safety related means for conducting a

plant cooldown are available, the plant is not forced to shutdown.

Therefore, plant j -

cafety will be enhanced by this proposed change.

The addition of the exemption to Specification 4.0.4 for the steam driven pump is an administrative change and does not affect the availability of the steam driven pump. to perform as required.

The additional' wording which would be added-to l:

SR 4.7.1.2.c is an-editorial change intended to. clarify when this i

surveillance is -required to be performed.

Therefore, plant j

safety will not be affected by this proposed change.

Therefore, this change is considered safe based on the fact that I

the AFW pumps _ will concinue to-perform as assumed in the safety analysis to automatically deliver the-required flow to the steam generators in' order to remove. decay heat from the reactor coolant

system, the plant-will' not be forced into a shutdown with no safety-related means for conducting:the cooldown, and the added testing flexibility will have a net result in reducing pomp wear.

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NO SIGNIFICANT'HAZ RDS. EVALUATION The no significant hazard considerations involved with -the proposed amendment have been evaluated, _ focusing on the threo standards set forth in 10 CFR 50.92(c) as quotedLbelow:

[

The Commission may make.a final determination, pursuant;to the procedures in paragraph 50.91, that a proposed = amendment-s to an-operating license for_

a.

facility licensed-under

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' ATTACHMENT B, continued Proposed Technical Specification Change No. 192 Page 5 paragraph 50.21(b) or paragraph 50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would nots (1)

Involve a

sigaificant increase in the probability or consequences of an accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The following evaluation is provided for the no significant hazards consideration standards.

1.

Does the change involve a

significant increase in the probability or consequences of an accident previously evaluated?

The probability-. of occurrence of a --previously evaluated accident is not increased because the allowable outage time for the AFW pumps remains unchanged.

The 'AFW pump performance and reliability will also not be changed by this proposed amendment.

The probability of-an accident-will be reduced by the addition of action statement "d".

By not requiring the plant to commence a

shutdown without any safety related means for conducting the

cooldown, the probability that a

reactor. trip will occur. concurrent with the loss of the safety related means for' removing decay heat is. lowered.

Due to these factors, the probability of an accident previously evaluated is not significantly increased.

The proposed. changes to SR 4.7.1.2 do not affect the ability of the AFW pumps to-perform as assumed 'in.the safety analyses.

The proposed changes will not result in any additional challenges to. the plant equipment.

Because the plant design limits will continue'to be met, the fuel and reactor coolant system' pressure boundary integrity is not challenged for the assumptions employed in *.he calculation of the offolte radiological doses.

Hence, the consequences of the-accidents considered in the BVPS Unit No. 1 licensing

-basis remain unchanged.

Therefore, the proposed changes do not involve a significant increase - in the probability or consequences of an accident previously evaluated.

2.

Does ~ the cha.1ge create the possibility of a new.or different kind of accident from any accident'previously evaluated?

There-would be no change -to system configurations, plant equipment.

or analysis as

'a result of-this proposed' amendment, i

2

'ATTACllMEliT II, continued l

Proposed Technical Specification Change llo. 192 Page 6 1

Therefore, the proposed changes

.do not create the possibi)ity of a new or different kind of_ accident from any previously evaluated.

3.

Does the change involve a significant reduction in a margin of safety?

The proposed changes will not affect the heat removal capability of the auxiliary feedwater system to a value less-i than that assumed in the safety analysis.

The proposed changes will not result in any additional challenges to the a

plant equipment including the fuel and reactor coolant l

system pressure boundary.

The plant will continue to operate within the' bounds of the various safety analyses.

Therefore, the proposed changes do not involve a significant i

j reduction in a margin of safety.

s j-F.

110 SIG!1IFICAllT 11AZARDS CollSIDERAT1011 DETERl4I!1ATIO!1 Based on the cor.siderations expressed above. it is-concluded that the activities associated with this license amendment request satisfies the no-significant hazards consideration standards of

'i 10 CFR 50.92(c)

and, accordingly, a

no significant hazards consideration finding is justified.

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1 ATTACllMENT B-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 192 i

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Proposed Revisions to the Inservice Testing Program for Pumps and Valves 1

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Beaver Valley Power Station Unit 1 NAFP INSERVICE TESTING (IST) PROGRAM FOR PUMPS AND VALVES Page 1 of 6 4.

The pump shall not be returned to service until the condition has been corrected.

I The corrective action shall be considered completed when a satisfactory inservice j

test has been conducted in accordance with IWP 3111.

t Per IWP 3500 each pump shall run at least 5 minutes under condTons as stable as the i

system permits prior to measurement of the specified parameters (when bearing temperature measurements are not required). When bearing temperature measurements are required each pump shall be run until the bearing temperatures stabilize prior to making the specified measurements. A bearing tempenture is considered stable when i

three successive readings taken at 10 minute intervals do not vary by more than 3%

l Bearing temperature measuremJnts are required annJally (normally in August).

l l

At certain times plant conditions may preclude returning a pump to the same reference condition for its normally scheduled surveillance. Since IWP 3112 permits the I

establishment of additional sets of reference values, a pump curve which is.merely a graphical representation of these reference values will be used, d

Records of the results of inservice tests and corrective actions as required by subsection IWP-6000 are trended in tabular from. Pump performance characteristics will be examined i

for trends.

"Punp Minhann Operating Point- (MDP) Curves" l

The followingene. sections of this document are the " Pump Testing Outlinesgand

• Pump j

Relief Requests' sections. The " Pump Testing Outilnes" section is a listing of all the pumps in the IST Program, their testing requirements, and their specific relief request reference numbers. The pumps are arranged according to system and pump mark number. The following abbreviatjons and designations are used on the Pump Testing Outlines and throughout the IST Program for pumps:

1.

Under Pa.ramq1gr column a.

(N)

Speed b.

(Pi)

Inlet Pressure-i

c. (AP)

Differential Pressure a

d.

(Q)

Flowrate

e. (V)

Vibration 1

f.

(Tb)

Bearing Temperature I

g.

(L)

Lubricant Level or Pressure

2. Under OST column

a. (1BVT)

Unit 1 Beaver Valley Test b.

(10ST)

Unit i Operating Surveillance Test c.-(Q)

Quarterly Test Frequency d.

(A)

Annual Test Frequency e.

(R)

Refueling Test Frequency f.

(NA) ~

Not Applicable

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Beaver Valley Power Station Unit 1 ERArr INSERVICE TESTING (IST) PROGRAM FOR PUMP 3 AND val.VES Page 2 of 6

3. Under Rea'd column a, (RR)

- Rellet Request j

b.

(X)

Meets or exceeds ASME requirements j

. c.

(E)

- Exempt d.

(NA)

- Not Applicable m

The " Pump Relief Requests ~ section contains the detailed technical description of particular I

conditions and equipment installations prohibiting the testing of some of the characteristics i

of safety related pumps. An alternate test method and the frequency of revised testing is also included to meet the intent of 10CFR50.55a.- The relief request (s) for a specific pump

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is referenced by the number (s) listed on the pump's testing outline sheet.

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'Ihe " Pump Minimum Operating Ibint (PCP) Curves":section contains a graphical-representation of the minimum allcwable pump f1cu versus head -

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which is required to meet applicable safety analyses for each pump in the Unit 1 IST Prtgram.

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ATTACHMENT C Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification change No. 192 l-

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3/4 7-5 3/4 7-6 B 3/4 7-2 i

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' PLANT SYSTEMS l

AMXJLIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3. 7.1. 2 _

At least three steam generator auxiliary feedwater puros a:

associated flow paths shall be OPERABLE with:

a.

Two feedwater

pumps, each capable or-being powered from separate emergency busses, and b.

One feedwater pump _ capable of being powered from arr OPERABLE steam supply system.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

a.

With one auxiliary feedwater pump inoperable, restore at least three auxiliary feedwater pumps (two capable of being powered from separato emergency _ busses and~one capable of being powered by an OPERABLE steam supply system) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or.be inLHOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With the motor driven auxiliary feedwater pu'p supplying the redundant' header inoperable, realign the two remaining auxiliary feedwater pumps to separate headers within 2

hours, c.

With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

4 d.

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as-possible.

SURVEILLANCE REQUIREMENTS f

4.7.1.2 Each auxiliary -feedwater pump shall be demonstrated OPERABLE:

I 1a.

When cested pursuant to ^ Specification 4.0.5:

k 1.

By verifying, that the pump's ' developed head at the-flow test point is greater than or equal to the required developed head as specified in the Inservice Testing Program.

The provisions of Specification 4.0.4 are not applicable for ent into Mode 3 for the steam turbine driven pump testing.

(1)

Secondary side steam pressure shall be greater than 600 psig when performing this surveillance requirement for the steam turbine driven pump.

BEAVER VALLEY - UNIT 1 3/4 7-5 Amendment No.

(Proposed Wording)

..'PLA'T S1FTEMS N

SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days by:

1.

Verifying that each valve (manual, power operated or automatic) in the flow path that is not

locked,

sealed, or otherwise secured in position, is in its correct position.

2.

Reverifying-the requirements of Technical Specification Surveillance 4.7.1.2.b.1 by a second and independent operator.

3.

Establish and maintain constant communications between the control room and the auxiliary feed pump room while any normal discharge valve is closed during surveillance testing.

i 4.

Verifying operability of each River Water. Auxiliary Supply valve by cycling each manual River Water to Auxiliary Feedwater System valve-through one complete cycle.

c.

Verify Auxiliary Feedwater flow from WT-TK-10 to the Steam Generators with Auxiliary Feedwater Valves in their normalalignment.(ge d.

At least once per 18 months during shutdown by:-

1.

Cycling each power operated (excluding automatic) valve in the flow path that is not testable during-plant operation ~,

through at least one complete cycle of full travel.

2.

Verifying that cach automatic valve in the tiow path actuater to its correct position on a test-signal.

3.

Verifying that each pump starts automatically upon receipt of a test signal.

(2)

This surveillance is required. to be performed prior to entry into Mode 2 whenever-the plat.5 has been in. Modes 5 or 6 for greater than 30 continuous days.

BEAVER VALLEY - UNIT 1 3/4

'-6 Amendment No.

(Proposed Wording)

. PLANT CYSTEMS BASES d

U=

maximum number of inoperable safety valves per operating steam line Power. Range' Neutron Flux-High Trip Setpoint for (N)

(109)

=

loop' operation 71 percent. of RATED THERMAL POWER permissible by P-8 (W)

=

Setpoint for 2 loop operation with stop valves;open..

66 percent of RATED THERMAL POWER permissible by P-8 (W)

=

Setpoint for 2 loop operation with stop valves closed.

Total relieving capacity of all safety valves per X

=

steam line in lbs/ hour (4,261,666)'

4 Maximum relieving capaelty of an-one safety valve in Y

=

lbs/ hour (873~,600) 3/4.7 1.2 AUXILIARY FEEDWATER PUMPS-The OPERABILITY c41 the auxiliary _feedwater pumps ensures'that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of.a total loss of off-site porsr.

The capacity of each aexiliary feedwater1 pump is sufficient to ensure that adequate feedwater flow -is available to remove decay heat and reduce the Reactor Coolant System ts.nperature to less than 350'F when the Residual Heat-Removal Syst em may be placed into i

operation.

3/4.7.1.3 PRIMARY PLANT DEMINEPALIZED WATEn-The OFERABILITY-of the PPDN storage: tank with~the minimum water volume ensures that sufficient water is available-for cooldown of' the Reactor Coolant System to less than 3CO*F in the event-of a total loss of off-site-power.

The minimum water _ volume is sufficient.to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to atmosphere.

t I

T BEAVER VALLEY - UNIT 1 B 3/4 7-2 Amendment No.

4 (Proposed Wording) s 4

e