ML20137N718

Submits Addl Info for plant-specific Review of Implementation of TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pump, Per Request Re Generic Ltr 85-12. Info Covers Items Common to Snupps Plants

ML20137N718
Person / Time
Site:	Callaway
Issue date:	11/27/1985
From:	Schnell D UNION ELECTRIC CO.
To:	Thompson H Office of Nuclear Reactor Regulation
References
TASK-2.K.3.05, TASK-TM GL-85-12, ULNRC-1215, NUDOCS 8512040212
Download: ML20137N718 (24)

Text

7

' Union Etscraic a

NOVOMbGE 27* I900 1901 Gratiot Street. St. Louis Mr. Hugh L.

Thompson, Jr.

Director, Division'of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555 ULNRC-1215

Dear Mr. Thompson:

CALLAWAY PLANT, DOCKET NUMBER 50-483 RESPONSE TO GENERIC LETTER 85-12 IMPLEMENTATION OF THE TMI ACTION ITEM II.K.3.5

" AUTOMATIC TRIP OF REACTOR COOLANT PUMPS" Union Electric Company received the subject NRC Generic Letter 85-12 in July, 1985.

The Generic Letter finds acceptable the Westinghouse Owner's Group generic response to the concerns of TMI Action Item II.K.3.5, " Automatic Trip of Reactor Coolant Pumps", and requests additional information to complete the callaway Plant specific review.

The Generic Letter requested r

that licensees furnish the plant specific information listed in Sections A, B, and C of the Generic Letter attachment entitled, IV IMPLEMENTATION.

This submittal provides the requested information.

The responses to Section A and B address those items common to the SNUPPS plants.

The response to Section C is Callaway site specific.

If additional information is required, please let us know.

Very truly yours, i

Donald F.

Schnell DJW/1jr Enclosure hE 0232 851127 DOCK os 3

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STATE OF MISSOURI )

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i-CITY OF ST. LOUIS )

l t-Donald F.

Schnell, of lawful age, being first duly sworn upon oath says that_ he is Vice President-Nuclear and an officer of l

Union Electric' Company; that he has read the foregoing document and i

knows the content thereof; that he has executed the same for and on l

behalf of.said company with full power and authority to do so; and l

that the facts therein stated are true and correct to the best of his knowledge, information and belief.

i' By e

D6dald ~F.

Scfinell Vice President l

Nuclear SUBSCRIBED and sworn to before me this / 7 day of 1986~

AA44-BARBARA bAF[

NOTARY PUBUC, STATE Of MfS00URI MY COMMIS l0N EXrtRES APRIL 22,1989 ST. Louts COUNTY i

1 cc:

Gerald Charnoff, Esq.

Shaw, Pittman, Potts & Trowbridge 1800 M.

Street, N.W.

Washington, D.C.

20036 Nicholas A. Petrick Executive Director

.SNUPPS 5 Choke Cherry Road Rockville, Maryland 20050 C. W. Hehl Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road L

-Glen Ellyn, Illinois 60137 Bruce Little-Callaway Resident Office U.S. Nuclear Regulatory Commission RRil Steedman, Missouri 65077 Tom Alexion Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop P-316 7920 Norfolk Avenue Bethesda, MD 20014

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Encloauro ULNRC - 1215 November 27, 1985 Page 1 of 18 RESPONSE.CONCERNING IMPLEMENTATION OF THE REACTOR COOLANT PUMP (RCP) TRIP CRITERIA A.

Determination of RCP Trip Criteria 1.

NRC Request

,i Identify the instrumentation to be used to determine the RCP trip setpoint, including the degree of redundancy of each' parameter signal needed for the criterion chosen.

SNUPPS Response to A.1 In reference 4, SNUPPS notified the NRC that reactor coolant system (RCS) pressure has been chosen as the trip s

. parameter.for the SNUPPS plants.

There are three wide-range pressure indicators that are available to the operator, two that receive their signals from the nuclear incore instrumentation guide tubes at the seal table (BBPI 403 and 405), and one that receives its signal from the top of the reactor vessel (BBPI 406).

These redundant, Class lE transmitters are located outside of the containment.

Each of the transmitters is associated i

with a.different Separation Group of SNUPPS plant instrumentation- (Ref. FSAR Section 7.1).

The design features of this pressure instrumentation (consistent with the function, location, and environmental conditions) have been reviewed and availability is adequately assured for accident mitigation.

2.

NRC Request Identify the instrumentation uncertainties for both normal and adverse containment parameters.

Describe the basis.for the selection of the adverse containment parameters.

Address, as appropriate, local conditions such as fluid jets or pipe whip which might influence the instrumentation reliability.

SNUPPS Response to A.2 Instrumentation uncertainty for wide-range pressure indicators PI-403, PI-405, and PI-406 was determined using the statistical methodology, as described in reference 3, previously approved by the NRC.

An operator reading error of one-half the smallest instrument scale dimension was statistien11y included.

The resulting 4

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Enclosure ULNRC - 1215.

' November 27, 1985 c

Page 2 of 18 i

u error for' normal environmental conditions is 3.5%, which

' gives -an. instrument uncertainty of 105' psi.

ThefRCS wide-range pressure" transmitters in the SNUPPS plants are supplied by Westinghouse and are located

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outside containment in areas where they will not be

-adversely affected by accident conditions inside

,c containment other than radiation, or accident conditions 1

outside containment other than a potential local auxiliary steam line break which could. adversely affect i

one of the three transmitters.

However, implementation of the RCP, trip criteria is not required for an auxiliary steam line break event..The total integrated radiation dose-for a-six-month. post-LOCA period in the vicinity of these transmitters does not exceed the Westinghouse threshold criteria for a harsh. radiation environment.

Therefore, the instrument uncertainty for normal' environmental conditions applies'for all cases when the RCP trip criteria may be invoked.

3.-

NRC Request

'In addressing the selection of the criterion, consideration to uncertainties associated with the WOG i

supplied analyses values must be provided.

These uncertainties include both uncertainties in the computer program results and uncertainties resulting from plant-specific features not. representative of the generic data group.

If a licensee determines that the WOG alternative I

, criteria are marginal for preventing unneeded RCP trip, it is recommended that a more discriminating plant-~

specific procedure be developed.

For example, use of the NRC-required inadequate-core-cooling instrumentation may be useful to indicate.the need for RCP trip.

Licensees should take credit for all equipment - (instrumentation) available to the operators for which the licensee has sufficient confidence that it will be operable during the expected conditions.

SNUPPS Response to A.3-j The LOFTRAN Computer code was used to perform the alternate RCP trip criteria' analyses.

Both Steam Generator Tube Rupture (SGTR) and non-LOCA events were simulated in these analyses.

Results from the SGTR analyses were used to obtain all of the trip parameters.

LOFTRAN.is 'a Westinghouse licensed code used for FSAR i

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SGTR and non-LOCA analyses.

The code has1 been validated.

againstfthe-January, 1982JSGTR event at the.Ginna plant.

Results of this validation show that LOFTRAN can accurately predict.RCS pressure, ARCS temperatures and

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secondary pressures, especially in the first ten minutes of the transient.

This is the critical time period when

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minimum' pressure and subcooling is experienced.-

The: major causes ofLuncertainties and conservatism in the

-computer program results, assuming no changes in the

' initial. plant conditions (i.e., full power, pressurizer

. level,'all SI and AFW pumps run) are due to either calculated models.or inputs to LOFTRAN.

The following are considered to have the most impact on the determination of the RCP trip criteria:

t 1.

Break flow 2.

-SI flow 3.

Decay heat 4..

Auxiliary feedwater flow The following sections provide an evaluation of the>

uncertainties associated with each of these items.

k.

To conservatively simulate a double ended' tube rupture in safety analyses, the break flow ~model used in LOFTRAN includes a substantial. amount of conservatism (i.e, predicts higher' break flow than actually expected).-

Westinghouse has performed nnalyses and developed a more realistic break flow model that has been validated against the-Ginna SGTR tube rupture data.

The break flow model used~in'the WOG analyses has been'shown'to be approximately 30% conservative when the effect of the higher predicted break flow is compared to the more realistic model.

The consequence of' the higher predicted

. break flow'is a lower than expected predicted minimum pressure.

The SI flow inputs used were derived -from best estimate

. calculations, assuming all SI trains operating.

An-evaluation of the calculational methodology shows that these inputs have a maximum uncertainty of 4 10%.

The decay heat model used -in the WOG analyses was based on. the 1971 ANS 5.1 standard.

When compared with the more recent 1979 ANS 5.1 decay heat inputs, the values used in the WOG analyses are higher by about St.

To determine the effect of the uncertainty due to the decay heat model, a sensitivity study was' conducted for the

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Enclosure-ULNRC

,1215 Novembdt)\\27[1985 Page 4 o ~10-q.-.

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SGTR.- The results ofthis study show that a 20%. decrease

_ in decay heat resulted in only a 1% ' decrease in RCS pressure for-the first-10 minutes of the. transient.

Since;RCS temperature is controlled by the steam; dump,-it is not-affected by the' decay heat'model. uncertainty.

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The AFW flow rate input.used in the WOG analyses are best estimate. values, assuming that all auxiliary feed' pumps are running, minimum pump start delay, and no throttling.

To evaluate the uncertainties with AFW flow rate, a sensitivity study was performed.- Results from the two '

loop plant study-show that a 64% increase'in AFW flow

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resulted in only an 8% decrease in minimum RCS pressure.

Results from the.3 loop plant study show that a.27%i

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increase in AFW flow resulted in only a 3% decrease in

minimum RCS pressure. f$

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-8 The effects of-all these uncertaint.ics in the models a'ad

-% input parameters were evaiuated, and'It.was concludsd that the contributions fros'the break. flow conservatism

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,and the SI uncertainty dominate.

The calculated overall uncertainty in the WOG analysis.for the'SNUPPS plants, as *

.a result of'these considerations, is a -150 psig to +150 s

psi 4^for the minimn's RCS piessure RCP trip setpoint.

Due to the minimal ef fectsifrom.the decay heat model and AFW -

input uncertainties on Ch,e RCS pressure uncertainty, the calculational uncertainty tesult includes only the ef fects of the uncertainties due to the break flow.model

' and-SI flow inputs.

j The're are no uncertainties 'resulting from plant-specific

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features ' not representative. of the ' gdneric data group.,'

-RCP trip on RCS pressure has been selected as the

.. / appropriate trip parameter.' RCS pressure provides asple margin (in excess %300 psig) to the trip setpoint for the non-LOCE"acciden'ts that were evaluated in the WOG analyses.'

B.

Potential Reactor Coolant Pump Problems' s

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NRC Request k,

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Assure that containment isolation, including inadvertent p

isolation, will not cause problems if it occurs for non-LOCA. transients and accidents.

a.

Demonstrate that, if water services needed for RCP

. operations are terminated, they can be restored ~(ant

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' Enclosure i

.7 ULNRC - 1215 November -- 27, 1985

.l Page 5 of118 h-F + :- s

'd enough once a'non-LOCA situation is confirmed to prevent seal damagelor failure.

b. : Confirm that containment isolation with continued

. pump operation will not lead to seal ~ or pump' damage

or failure.

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.SNUPPS1 Response to'B.l.a The automatic (Hi-3) containment isolation signal isolates component cooling water to and from each RCP motor and' thermal barrier.

Automatic (Hi-1) or manual

- containment =lsolation isolates ROP seal water return to the Chemical and' volume Control. System (CVCS) but1does_

.not isolate seal water injection to the RCP. 'Although the RCP. seal system can operate for some ' time with seal water injection.only, the RCP motor bearings are more limiting and are qualified for 10 minutes operation

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.without component cooling water.with no resultant damage (re ference.10). - Ten minutes provides adequate time-for' the operator:to_ determine *either that a.non-LOCA accident has occuredland to restore 1 component cooling water flow to the RCP orlto determine that a'LOCA-has occured'and to

trip the'RCPs.

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SNUPPS Response to B.l.b

- Containment isolation '(Hi-3) does not isolate seal' water injection, but-does' isolate component cooling water to the reactor-. coolant pump and. seal water return to the CVCS.

The limiting; components associated with the:RCP

'under=these conditions.are the motor-bearings /which are

qualified ? for 110 minutes operation without resultant og

. i sdamage.

As; described in the response tolB.l.a.,-10 e

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minutes provides adequate time to: restore component cooling water flow orJto trip the-RCPsfas appropriate.

2.

NRC Request g

. Identify the components --available to trip the-RCPs,'

L including relays,' power supplies, and breakers.

Assure L

.that'RCP, trips.when determined to be necessary, will l

occur.; If necessary, as a result of the location of any critical component, include the effects of adverse L

containment conditions on RCP trip' reliability. ' Describe

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the basis for the adverse' containment parameters K

selected.

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Page,6 of 18 s

SNUPPS Response to B.2-j 1

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'The components available to manually ~ trip the RCPs are.

~11sted on Attachment A.

R) trip one RCP, three active. devices must function:

the hand-switch on the MCB; the trip coil mechanism; and the breaker.

The Control and' Turbine Bu'ildings are.not

. subject to4 harsh environmenta1' conditions resulting'from

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'a'LOCA in the' containment.-

There are no components required'for the. function of' 3

tripping the RCPs located inside containment.

The-only

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equipment.inside' containment associated with. interrupting

. power 1to the RCP motors are the RCP motors, electrical 4

F power cable to the motors, the in-containment portion of.

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the. electrical penetration assembliesicarrying pow'er to theipump; motors, the differential current transformers

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.on the power cables)yand associated electrical! cable and-

. electrical penetrat' ion assemblies for the differential relay current circuits.

^It is unlikely that-adverse environmental conditions could affect the above equipment' s

" prior to; initiation of.a' manual trip; however,

-degradation of either power cables or differential relay

. current' circuits would most likely result in an RCP trip

signal" generated.by the differential relay.- Onceithe RCP breakersiare open, the RCPs receive no power and further 2

degradation of the equipment inside containment will not-t

. result in.the breakers'reclosing.

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Adverse containment parameters' assumed in this evaluation are those post-LOCA conditions-of temperature, pressure,'

humidity, radiation,1 chemical spray, and potential l submergence provided in the SNUPPS NUREG-0588 Submittal.

(reference 5).

1 C.--Operator Training and' Procedures (RCP Trip) 1..

NIK: Request-I f

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Describe the Loperator; training Lprogram,for RCP trip.

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. Include the general philosophy regarding the need to trip

pumps versus the desire to keep pumps running.

CALLAWAY Response to C.1

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1The Callaway Training Department.provides in-depth training on Plant Emergency procedures and their bases to students-in initial License Training and. Licensed

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Enclosure ULNRC - 1215 November 27, 1985

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Page 7 of 18 il s

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. Operator Requalification Training.

Part of their

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-training includes a detailed. study of Reactor Coolant

-Pump trip criteria.

Subjects discussed in this area are as follows:

.Large break versus small' break LOCA concerns This section 11ncludes information on the effects of?

large break _and'small break.LOCA's, time frames for i

.several site lLOCA's, and' the possibility of core uncovery.

L Integrated _ mass loss versus time of RCP trip This section-includes info'rmation on the effects of RCS mass loss with RCP's running during I,0CA's.

Continuous iRCP -oper ation :

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This section includes information.on the effects of RCS voiding.

-1 Bases for RCP trip criteria-

.The bases for.RCP trip criteria are discussed inL

, detail.,Information on this-topic is derived from the Westinghouse Owner's Group analyses of RCP-trip

, criteria.

L-Applicability.ofiRCP:tripssteps

_ The applicability of. RCP trip st'eps 'during controlled-cooldownsland other' plant evolutions is discussed..

>The bases for the-Callaway Plant Emergency procedures RCP n trip criterialis discussed inL detail in.both -initial training-and retraining.- The significance of running RCP's during -Loss of Coolant Accidents is discussed, with the effects of RCP~ operation during small break Loss of

Coolant' Accidents emphasized.. Voiding in the Reactor (Coolant' System during' pump on and pump off conditi~ons is reviewed-in,the Callaway Mitigation Core Damage' course.1

.In-addition to this training, the-bases _for specific

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_ steps 'in' allimajor Emergency ' Procedures are discussed.

These steps include RCP trip criteria for theJvarious'

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. procedures.

Classroom training ~on these topics is

-reinforced through theLuse of simulator exercises.

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Loverall, approximately: 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of classroom material is presented'on these topics.

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The' philosophy _regarding-RCP' trip criteria presented to students during this training is consistent with that presented in the Westinghouse owner's Group Background-Documents !for the -Emergency. Procedures. _ Students are trained to evaluate plant conditions during emergencies and follow'the guidance provided in the Plant Emergency.

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Procedures.-

-While the' des'irability of maintaining RCS forced flow for

. specific accidents. (i.e., S/G. Tube Ruptures) is a

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' discussed, the training reinforces.the concept that actions performed during emergency conditions shallibe-in

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accordance with plant procedures.

2 '. - NRC' Regtjest

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.1 Identify.those procedures'which include RCP' trip related operations:

f(a) -RCP trip using WOG alternate criteria (b)

RCP restart

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(c)

Decay heat removal by natural circulation L(d)

Primary system void removal 1

-(e )

Use of steam generators with and without RCPs

.. Loperating (f)

RCP: trip for other. reasons-

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Callaway Response to C.2 The responses to-~this section~ reference numerous Operation Department procedures.' Therefore, Attachment B f.

provides a listing of applicable procedures and their 1 -

titles for easy reference.

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Enclosure ULNRC - 1215 November 27, 1985 Page 9 of 18 (2.a)

The procedures identified below generically direct operators to trip all Reactor Coolant Pumps (RCP) when both of the following conditions exist:

a)

Charging pumps'or Safety Injection pumps - AT LEAST ONE RUNNING and b)

Reactor Coolant Syste,m (RCS) pressure is less than-1400 psig.

This is an incorporation of the WOG alternate criteria.

E-0 ES-0.4 2.

ES-0.0 E-1 ES-0.1 E-3 ES-0.2 ECA-2.1 (2.b)

Callaway procedures incorporate the following generic instructions and cautions for restarting the reactor coolant pumps:

Cautions

- After any attempt to start where the motor has failed to achieve full speed before it is stopped, a restart'should not be attempted until the motor'has been allowed to cool by standing idle for a' period of-not less than thirty minutes.

- Two successive RCP starts are permitted, provided the motor is allowed to coast to a

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stop between starts.

- A third RCP' start may be made when the winding and the core have cooled by running for a period of 20 minutes or by standing idle for a period of 45 minutes.

- When three starts or attempted starts have been made within a two-hour period,'then a fourth start should not be made until the motor has been allowed to cool, by standing idle for at least one hour.

- Start only one RCP at a time.

Instructions

Enclosure ULNRC - 1215 November-27,-1985 Page 10 of 18 Ensure the following:

a.

RCS pressure > 325 PSIG b.

13.8 kV buses energized (PA01 and PA02) c.

il seal dP > 200 PSID d.

VCT pressure > 15 PSIG CVCS in operation with seal injection e.

6-13 gpm per pump f.

il seal leak off > 0.2 GPM g.

CCW in operation, supplying cooling water to the following:

(1)

Thermal barrierJheat exchangers (2)

Upper and lower bearing oil coolers (3)

Motor air coolers h.

All annunciator alarms clear for RCP operation Start the RCP Oil Lift Pump-NOTE A pressure interlock prevents starting the RCP unless a minimum oil.

pressure of 700 PSIG is available to the motor thrust bearing oil lift system.

This interlock is satisfied when the white light on the respective oil lift pump control switch is lit.

After.the Oil Lift Pump has been' running

. for two minutes, start the RCP.

. After the RCP has been running:for one-minute, stop the Oil Lift Pump.

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ULNRC - 1215 November,. 27,-1985 Page 11 of 18 c.

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Subsequent steps include.; directions to go to other steps of_the. procedure if either an RCP is running or if:an RCP.

cannot be started.

The following list of procedures are

^ identified as containing-the above instructions and cautions:

ES-0.1 ECA-3.1 ES-0.2 ECA-3.2 ES-0.4 ECA-3.3 ES-1.l~

FR-C.1 1,

ES-1.2 FR-I.3 E-3 FR-P.1

-ECA-2.1 In addition, one additional procedure, OTN-BB-00003, also.

contains the cbove instructions.and cautions while, providing detail on equipment lineups, required valve

. positions, and stopping the RCPs at designated loads.

(2.c)~ The following proceduresjinclude RCP trip related operations for decay heat removal by

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natural circulation:.

1 ES-0.1 ECA-2.1 ES-0.2 ECA-3.-l ES-0.4 ECA-3.2 ES-1.1-ECA-3.3

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ES-1.2-FR-P.1 E-3 Leach of these procedures =h'as a checklist with which to verify Natural Circulation.

Checklist Items include:

- RCS.subcooling -'MORE SUBCOOLED THAN INSTRUMENT ERROR Two methods for determination of subcooling are provided.

One method issthe RCS SUBCOOLING METER ERROR CORRECTION calculation.

The otherciscuse of'RCS

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SUBCOOLING CURVES.

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- Steam pressure - STABLE.

- RCS hot leg temperature - STABLE OR SLOWLY

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. DECREASING.

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Enclosure ULNRC - 1215 November 27, 1985 Page 12 of 18

- RCS cold' leg temperature - NEAR SATURATION TEMPERATURE FOR STEAM PRESSURE.

NOTE (1)

Approximately 20 minutes'will be required-to establish stable 1 natural circulation conditions.

(2)- Hot leg temperatures are expected to initially respond by increasing to 575 deg. F and then.

stabilizing.

- Core exit thermocouples (TC) - STABLE OR SLOWLY DECREASING.

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- IF natural circulation is verified, THEN return to the appropriate step of the procedure.

- IF natural circulation is NOT verified, T5EN increase dumping steam and verify

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natural; circulation from trended values.

Specifically, Procedure ES-0.2 provides actions to perform a natural circulation RCS cooldown and depressurization to cold shutdown, with no accident-in progress under requirements that will. preclude any upper head void formation.

This procedure is entered from ES-0.1 and ECA-0.1 when it has been determined that a natural circulation cooldown is required.

Specifically, Procedure ES-0.4 provides actions to continue plant cooldown and depressurization to cold shutdown, with no accident in progress, under conditions that allow for the potential formation of a void in the upper head region without a vessel level system available to monitor void growth.

This procedure is entered from ES-0.2.

Specifically, Procedure ECA-0.1 provides the following list of steps to ensure natural circulation, and if the proper response is not obtained, further procedural steps

. increase the dumping of steam from intact steam generators.

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ULNRC - 1215 November 27, 1985 Page-13 of.18

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.- RCS subcooling - PRESSURE AND TEMPERATURE _

W_ ITHIN PERMISSIBLE RANGE.

Use'RCS subcooling curves to' ensure adequate subcooling.

For normal containment, use core.exi t.TC 's'..

For adverse' containment,

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use RCS wide range RTD's..

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Steam. Generator (S/G) pressures - STABLE OR DECREASING.

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- RCS hot leg temperatures - STABLE'OR DECREASING.

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- Core exit TC's - STABLE OR DECREASING.

- RCS cold leg temperatures - AT SATURATION

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TEMPERATURE FOR S/G PRESSURE.

I The procedures OTN-BB-00003, OTG-ZZ-00001 and.

.OTG-ZZ-00006 address natural circulation in-x

.the PRECAUTIONS AND LIMITATIONS SECTION.

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They recommend. if one - RCP cannot be started, that natural circulation using the steam dump to condenser or atmosphere be utilized and

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. verified.via the following-indications::

- RCS subcooling - based on ~ core Jexit TC's -

GREATER.THAN-ll-DEG. F.

S/G pressures:-' STABLE OR DECREASING.

- RCS hot.legLtemperatures - STABLE OR DECREASING.

RCS cold leg temperatures - AT SATURATION TEMPERATURE-for S/G PRESSURES.

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- Core outlet temperatures -LAT LEAST 10 DEG.

F.'BELOW SATURATION TEMPERATURE.

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(2.d)

Primary system void removal,las'related-to RCP trip operations,-is accomplished in the procedures listed below.

These procedures are-entered from procedure ~CSF-1.

l-I, ES-0.4 FR-I.3 a.-..

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'ULNRC - 1215 November 27, 1985 Page 14 of 18 Description of ES-0.4 was given in the response to 2.c.

FR-I.3 provides. actions to respond to voids in the_ reactor vessel and is entered from-

-CSF-1 when the pressurizer level is at or

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above normal and the reactor vessel is less than full.

CSF-1 is a procedure th'at provides two

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functions:

1) provides general surveillance under all sets of unusual or abnormal ~

-conditions that can, lead to or result from initiation of safety injection,_2) and it directs operator guidance in those' rare events.that'go-beyond the design basis'of the Engineered Safeguards Systems and the Emergency Operating Procedures and Emergency

Contingency Actions.,

(2.e)- ik> specific procedures address use of the-S/G's-with_the.RCP's operating.

When the 1

RCP's are not operating, natural circulation

,is required and is described in ES-0.2 and ES-0.4.-

In both cases, S/G parameters:are used to verify natural circulation.-

(2.f)

The following : discussion addresses RCP _ trip _

procedures for other reasons.-

The:following_

procedures generically direct operators to

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_ trip any~RCP if component cooling waterEto.

e that' pump is-lost to-the RCP motor for greater than,2 minutes or if the upper or lower bearing temperatures reach 195 deg. F:

E-O' ES-0.2

-ES-0.0 ES-0.4 ES-0.1 ECA-2.1 s

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Enclosure ULNRC - 1215 5-November 27,s1985 Page 15 of-18 I

The next set of procedures generically l directs. operators to trip the.affected RCP's m

when the number one seal differential pressure is.less than.215 PSID OR if the-number < one seal leakof f flow is less than l.25 GPM:

ES-1.' 2 ECA-1.1

~ES-3.1 ECA-3.1 ES-3.2-ECA-3.2 ES-3.3 ECA-3.3

~

The~next three procedures'are entered from procedure CSF-1 and are' described-separately since the RCP trip criteria differs for each..

}

.FR-C.'l - In response to a degraded core, if at least two RCS hot leg temperatures are=

"less than 350 deg. F, the operators are 4

n

. directed to trip all RCP's since they:are no longer.needed for core cooling. 'Also, this procedure has operators trip all RCP's t

upon anticipation of the losstof ~ number one seal requirements with which further-L operation would' damage the RCP's.

Loss of.

.the. seal is anticipated with' ldepressurization'of~the S/G's to atmospheric pressure.

- FR-C.2 - One step in the procedure directs 1 operators'to trip'all RCP's upon anticipated loss of number one seal requirements and a later step directs'them to trip the pumps af ter SI flow has been established'and verified to_ provide core

. cooling.

' FR-H.1 - This procedure has operators trip all RCP's in order.to extend the' time to-restore' feed flow to the S/G's since.RCP operation results in heat addition to the RCS water.

Procedure, OTN-BB-00001, contains RCP trip.

~related criteria in the section titled

" Dynamic Venting of the RCS."

Within this e

section each pump is required to run'for two minutes after-reaching full speed or until seal DP1 reaches 200 PSID and is then stopped.

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Enclosure ULNRC - 1215 November 27, 1985 Page 16 of 18 The procedure _ directs operators to trip the affected pump if.RCP seal differential pressure decreases rapidly or approaches 200 PSID.

Procedure OTN-BB-00003 directs _ operators to trip the affected RCP.immediately if vibration exceeds 5 MILS.

Finally Procedure, OTO-BB-00002, describes symptoms, probable causes, and the operator actions required for RCP off-normal conditions.

It directs operators to immediately trip the affected RCP if any of the following conditions exist:

If vibration exceeds 5 MILS on the frame, 20 MILS on-the shaft.

- If on a loss of seal injection, any of the following occur:

a.

Number one seal outlet temp. > 235 deg.

F.

b ~.

Seal injection temp. > 150 deg.

F.

c.

Thermal barrier-cooling water inlet temp. > 105 deg.

F.

d.

Pump lower bearing > 225 deg.

F.

e.

Motor bearing cooling water temp. > 120 deg. F.

f.

Motor bearing temp. > 195 deg. F.

g.

If' reactor power < 48% and a loss of CCW to one RCP motor exists for > 2-minutes or if upper or lower bearing temperature reaches 195 deg. F.

- If reactor power is > 48% and a loss of CCN to one RCP motor exists for > 2 minutes, or if uppor or lower bearing temperature reaches 195 deg.

F.,

the operator is directed to trip the reactor and turbine,-

then trip the affected RCP(s).

- It further directs operators to reduce power to < 48% and then trip the_affected RCP(s) if the-seal 2kP is < 200 PSID or if frame vibration is above 3 MILS and increasing at a rate of 2 MILS /HR or if

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1 Enclosure p,

ULNRC - 1215 November 27, 1985 m'

~

Page 17 of 18

~

shaft vibration is above 15 MILS and increasing at a rate of l~ MIL /HR.

'~

Subsequent operator actions give additional guidance'for the following off-normal.

. conditions:-

I a.' Number one seal - Leak-off flow high i

~

.b.' Number ; one ~ seal - Leak-of f flow low

c. Number two seal - Leak-off flow'high.

i.

d. Numbec three seal - Standpipe High/ Low Level eJ Loss of Seal Injection' i--

f. Loss of Component; Cooling Water

-References l'.'

NRC Generic Letter 85-12, dated 6/28/85 2.

.SLNRC 83-0021, dated 4/22/83,-Response'to NRC Generic

-Letter No.83-10c.

.3.

~

lSLNRC 84-0050, dated 3/23/84, Response to NRC Questions on Setpoint Methodology for SNUPPS.

~

4.-

SLNRC 84-66, -~ dated-4/13/84, Final Responce to NRC Generic Letter No.83-10c.

5.

SNUPPS Report' of Independent Review of Environmental Qualification. Programs-to-NUREG-0588,.SLNRC 83-15, dated 4

3/10/83 as' revised by SLNRC 83-30, dated 5/27/83 and SLNRC 84-13, dated 2/1/84.

s 15.

-WOG Emergency-Response Guidelines Executive Volume,

= dated 3/21/84.

7.-

WOG Evaluation of Computer' Code' Uncertainties, dated T/16/85.

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Enclosure ULNRC - 1215 November 27, 1985 Page 18 of 18 s

8.

.The following. Bechtel Drawing No.s:

E-03PA02 Rev. 12 E-03PA05 Rev. 11

'E-03BB01 Rev. 13

E-03PA14 Rev. 8 E-OlPK01 Rev. 13 M-02BB02 (Q) Rev. 18 M-02BB04 (Q) Rev. 6 M-0G063-07 9.

- SNUPPS FSAR. Table 3. ll(B)-3.

1.10.

SNUPPS. FSAR Section 5. 4.1.

11.

BLUE 1985,- dated 9/5/85, Circuits Penetrating Containment Excluded from T/S Table 3.8-1.-

t 12.-

Instruction Manual for 13.8 kV Switchgear, E-009-0223-05, E-01001, Rev. 7.

k.

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A

Enclosure ULNRC -1215

' November 27, 1985 Page.1 of 1 Attachment A Equipment Available to Manually Trip Reactor Coolant Pumps Equipment Location / Room i s

A.

Main Control Board RLO21 Control Building /

3601 1.

Hand ~ Indicating Switch (one per fpump)

B.

Switchgear (PA01, PA02)

Turbine Building /

.2033 ft. el. NW

,1.

Breakers (one per' pump)

^

2.

Trip Coil (one per pump)

(energize-to-trip) 3.

Fuses (2) for breaker control power 4.

Breaker "a" contacts' (two per. breaker)

15. ' Wiring / terminal blocks 12.

125 VDC Distribution Panel-Turbine Building /

(PK41, PK62) 2033 f t. el. mi (control power to breakers)-

1.

Switches (43, #4)

^

2.

Wiring i:-

D.

125 VDC Switchgear (PK01, Turbine Building /

PK 02) 2033 ft.-el. NW i.

Fuses, wiring E.. ~ 125 VDC Batteries '(PKll, PK12)

. Turbine Building /

(power to 125 VDC switchgear) 2033 ft. el. NW

4 Enclosuro ULNRC - 1215 November 27, 1985 Page 1 of 2 1

ATTACHMENT B LIST OF APPLICABLE PROCEDURES AND THEIR TITLES Emergency Response Procedures

- E-0:

Reactor Trip or Safety Injection ES-0.0:

Rediagnosis ES-0.1:

Reactor 1 Trip Response

-ES-0.2:

-Natural Circulation Cooldown ES-0.4:

Natural Circulation ~Cooldown With Steam Void

'in Vessel (Wi thout ' RVLIS)

E-1:

Loss of Reactor or Secondary Coolant

.ES-1.1:

SI Termination ES-1.2:

Post LOCA Cooldown and-Depressurization E-3:

Steam Generator Tube Rupture ES-3.1:

Post-SGTR Cooldown Using Backfill ES-3.2:

Post-SGTR Cooldown Using Blowdown ES-3.3:

Post-SGTR Cooldown Using. Steam Dump ECA-0.1:

Loss of All AC Power Recovery Without SI Required ECA-1.1:-

Loss of Emergency Coolant Recirculation ECA-2.1:

Uncontrolled Depressurization of All Steam Generators v

ECA-3.1:

SGTR With Loss of Reactor Coolant-Subcooled Recovery Desired ECA-3.2:

SGTR With Loss of Reactor Coolant-Saturated Recovery Desired ECA-3.3:

SGTR Without Pressurizer Pressure Control

p Enclosure ULNRC -1215 November 27, 1985 Page 2 of 2

.r.

0; Function Restoration Procedures

. FR-C.1:. Response.to Inadequate Core Cooling FR-C.2:

Response to Degraded Core Cooling FR-H.1:

Response to Loss of Secondary Heat Sink

~

FR-I.3:

Response to Voids in Reactor Vessel FR-P.1:

Response to Imminent Pressurized Thermal Shock Condition

'CSF-1:-

Critical Safety Function Status Trees

, Other Procedures OTG-ZZ-00001:. Plant Heatup' Cold Shutdown to Hot Standby OTG-ZZ-00006: Plant Cooldown Hot Standby to Cold Shutdown OTO-BB-00002: Reactor-Coolant Pump Off-Normal OTN-BB-00001:' Reactor Coolant System OTN-BB-00003: Reactor Coolant Pumps a

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