ML18033A447

Forwards Addl Info to Resolve Certain App R Draft SER Open Items,Per 870323 Request.Main Steam Relief Valves Have Two Stated Shutdown Functions

ML18033A447
Person / Time
Site:	Browns Ferry
Issue date:	06/22/1987
From:	Gridley R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
TAC-60627, NUDOCS 8811300164
Download: ML18033A447 (18)

Text

I ACCELERATED DISIAUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8811300164 DOC.DATE: 87/06/22 NOTARIZED: NO DOCKET FACIL:50-259 Browns Ferry Nuclear Power Station, Unit 1, Tennessee 05000259 50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME AUTHOR AFFILIATION GRIDLEY,R.L. Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

R

SUBJECT:

Forwards addi info re -App R draft safety evaluation rept 1 open items,per 870323 request.

DISTRIBUTION CODE: DF01D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: Direct Flow Distribution: 50 Docket (PD Avail)

NOTES:1 Copy each to: F.McCoy,J.G.Partlow, S.Richardson,S.Black, 05000259 S B. D.Llaw.

1 Copy each to: S.Black, B. D.Llaw,F.McCoy.

J.G.Partlow, S.Richardson 1 Copy each to: S. Black, J.G.Partlow,S.Richardson 05000260 05000296 g j

B. D.Liaw,F.McCoy.

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'RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL INTERNAL: NUDOCS-ABSTRACT 1 1 ~Q~EG F~ 01 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 NOTES: 5 5

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NVZE 'lO ALL "RIDS" D, RECIPIEÃIS'IZASE HELP US IO REDUCE WASTE! CONI'ACT 'IHE DOCGMEPZ COFZfRL DESK, D ROOM Pl-37 (EXT. 20079) TO ELIHINATE YOUR KQK FKM DISTEGBUTION LISTS KlR DOCUMEHX8 YOU DON'T NEED!

tt lI TOTAL NUMBER OF COPIES REQUIRED: LTTR ~ ENCL

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i 11/28/1988 18'12 t SITE LICENSING BFN L44 8706Z2 SOS t28S '729 3111 P.82 5N l578 I.ookout. Place JUN Srm8>

U.S. Nuclear Regulatory Commission ATTN. Document Control Desk Qaghington, D.C. 20555

%mtlemen'n 4he Matter of Docket Nos. 50-259 Tennessee Valley Authority 50-260 50 296 BROWNS FEISTY NUCLEAR PLANT (BFN) 10 CFR 50, APPENDIX R l

eq ested in the NRCITVA March 23,. 987 meeting held in Bethesda, Maryland, ve are providing the. enclosed material to resolve certain Appon x r .

safety evaluation report open items.

W'is expected Chat folloving the rev ev of this material a final safety reviev evaluation for BFS vill be issued. Shou ld additiona an orma

~bisect be required, please refer any questions to James D. o co Licensing at (205) 729-2689.

very truly yours>

TENNESSEE VALLEY AUTHORITY Qdghel Nyek 8f R. I Oddly'.

Gridley, Director Nuclear Safety and Licensing

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Owg~ FERR RESPONSE TO NRC REQUEST Foi{ ADDlTZONAL ZNFORHATlON FROM MARCH 23, 1987 APPENDIX R MEETING IN BETHESDA, MARYLAND Item 3.4. 3-1 Containme t Atmos her e Dilution CAD Valve Ali nrnont

1. Pz'ovide justification for changing manual action to align CAD system valves f rom one hour to two bours.

2, Show that alignment of CAD system is needed only for non-automatic depressurkxatkan system (ADS) valves.

4 3, Show that the operatoz'an achieve manual actions to align CAD system valves after a fice.

4. Shaw that each fice area has at least ane ADS valve available during a fico event.

Show that safe shutdown can be achieved vith one 'relief valve after initial reactor depressurixatian.

MSPJMSP<

The manual action of aligning the cAD system is to provide pneumatic.

(r6trogen) supply ta operate the main steam relief valves (MSRVs) for safe shutdown after a fkze event. The MSRVs have tvo safe shutdown functions. The MSRVs are required to depressurixe the zeactoz'essel which allovs the RHR system ta operate jn the low pz'essure coolant kn)ection {LPCZ) mo4e and maintain coolant inventory. For this function, three HSRVs are required within the first, 20 minutes of the fice event. The initial pneumatic supply for these MSRVs is from the tvo receiver tanks af the drywall air central system, each with a 57 ft3 capacity. MSRVs with the ADS function have their own accumulators as backup pneumatic supply. These ADS accumulators aze sixed for five valve operations. only one valve operation (to apen valve) is required of the MSRVs {bath ADS and non-ADS). It is estimated that the receiver tanks of the dzywell air control system az'e capable of holding the MSRVs open for at least one hour. The zeactoc'ill be sufficiently depressurixed by then. Tharefare, no manual action is required to provide pneumatic supply to the MSRVs for the first safe shutdown function of the HSRVso After the initial depressurixatkon, the MSRVs are used ta provide a flaw path foz removing decay heat from the reactor vessel to the suppression pool. Only one MSRV is required for this function (See table 4.1 of HEDC-31'Xl9). For each fire area or fire xone, at least one MSRV used foz'he initial, de ressurisatkon vill be an Aps valve. This valve vill be sufficient to satisfy the second safe shutdown function since it has its awn accumulator SL'xed far five valve operations. Except for minor leakage through the '

solenoid vaLve to the MsRU pneumatic actuator, the operate.an of h MSRV open will nat axpend any pneumatic supply. It is estimated that the ADs

28S 729 3111 P.84 tem 3.4.3-) Co tainment Atmos here Dilutian cAD velya All ament (Continued)

QEEPONSE (Continued) accumulator can keep the valve open for at least 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (refer to sER dated 7/24/85, Subject NUREC-0737, Xtem l?.K.3.28, "Qualification of ADS Accumulator"). Therefore, no manual action is required to ensure the flow path for residual hest xemovsl circulation from the reactor vessel to the suppression pool within the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the blawdown.

The manual action which slijns the CAD system to pxovide nitrogen ta operate the MSRVs within ane hour is s conservative requirement which allows the flexibility to use the non-ADS valves for the second safe shutdown function The required manual, action>> involve opening two one-inch manual valves snd clasing another two-inch manual valve which can be pex'formed within two to three minutes. The equivalent fixe loads for the fire zones in question are approximately 30 to 40 minutes. The area containing the valves (Elevation 565 of such Reactor Building, Fire Zones 1-1, 1-2, 2-1, 2-2, 3>>1, snd 3-2) hss ares detection snd suppression, emexgency lighting, limited combustibles in the axes, a high ceiling (20 to 30 feet) snd large open axes, snd the presence of an open hatch about 50 feet fram the CAD location, portable smoke ejectors, self-contained breathing apparatus, and portable lanterns <ill be available for use by the operators, if necesssxy, to minimize any potential problems with smoke obscurstian. To alleviate any possible concern on this matter, the one-hour requirement in the manual action tables will be changed to two hours. Therefore, the operator will have s significant amount af time after the fire to perform the manual action requiring only twa to three minutes.

Ttem 3.4.3-2 Manual Actions

l. Show that each fire ares (with the exception of fire axes 16 af the Contxol Building) hss at least one emergency equipment cooling water (EEL) pump auto-start cspsbi.lity.

2. provide analysis to support extending the time limit for manual start of EECM pumps,

3. Commit to update correction to manual action calculation table for stax ting EECW pumps.

4. Show cost impact of performing modification to provide EEL pump auto-start feature for a Contral Building fire.

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The current post,-fire shutdown procedure for BFN has s manual action requirement to manually verify the EEcW pump availability in five minutes after the start. of a diesel generator, normally assumed to be at the initiation of the fire event.

11r28>1988 18r 14 S I TE L I CENS ING BFN 85 '729 3111 P @S

~pBVON F (Coflthtlu+4) hecent review of the plant configuration arrd the EECk'ump "tart logic indicates that the previous approach and assumptions went. beyond the design basis for Appendix. R fire event.s, The EECL pump logic includes an aut.omat.ic st aPt circuit r y based on diesel generator sr art recogni ti ori signals and a manual start. and stop circuitry fo". each pump, .he circu' y for the d esel ger>era or st a. t I ecogniti ari si gnalg, isfans jde he heac o.- buildings Diesel Generaror Buildings, and the respective shutdown board rooms. -.he c.'rcu;c."y for the manual st.art and stop signals is inside the cont.ro'uilding and t)ie respective shutdown board rooms and locally in the iotake pump "tat ion.

Analysis demonstrates that for a ire in any o. he fire areas (4 through '5 and 17 through 24), at, least two EEC4'umps would start. automatically upon a start signal wit;h its associated diesel generators, thus eliminating the need to manually verify the EEC)'ump sta. t from ou:side 'he cont:rol room for these areas.

Analysis also indicates that the auto-start circuitries associated <ith the diesel. generator run zecognition for the two divisions of t;he EECt4 pumps (A3 and C3 versus 33 and D3) are separated from each other by the wall between the unit 2 and unit 3 Reactor Buildings {figure 1). Thus, EECI'umps A3 and C3 would be available for a fire in unit 1 or 2. Fo" most loca ions in units 1 and 2, the A3 and C3 pumps wauld stazt. automatically. Howeves', for fire tones 1-3 and 2-3, because the cont, ol power feeder to 4 kV Shutdown Board 3EA is located in these two zones, the automatic start Of the A3 pump cannot be ensured. The A3 pump can be starzed by manual ac ion outside af the contz'ol room. The automatic start of the C3 pump is ensured for these two fire zones. Far fire zone 1-3, only unit 3 diesel generataz's az'e used .Or safe shutdown, The C3 yuzy provides dizect cooling water to the un't 3 diesel generators and no spurious operation of the sectionalizing val valves on the EECV headez would defeat this direct- cooling capability. Fo" fire zone 2-3, one of the unit 1 and 2 diesel generators (DG A) is used foz supply'ng power to

't unit 1 RHR pump and other required ac equ'pment. Since unit 11-is not a f're affected unit for a fire in .'re zone 2-3, the diesel generator DG A which provides power to un=t 1 w ll no. be reauireC fo. the 'r '

event. A spur ious operation o tne sec.iona i"'ng va ve 'nn un't . 2 (FCV-67-2')

would not affect the unit 3 diesel genezators w'.hach supply pawer to unit 2; however, it wzruld cut off the SEC)'low from EECM Pump C3 to "he uni s 1 anC 2 diesel generators. X this happens, the ape"atoz can tr'p the required dieso generator PG h. from the control room upon receiving the high temperature alarm. The operataz would then manually start the EEL pump before restartizrg the diesel generator at 30 ~utes, Ehould the operator ~ail ta trip diesel generator and should the diesel generator subsequently fail, the o p orator can align anathe" unit 3 diesel genewtor to supply t5e necessary ac powez'o support the safe shutdown .unc con in uni 1-3 and 2-3, M the KECM flow from EECW pump C3 does not provide adequate coaling water'o the requireC diesel genera a.s, M ope- o eratoz could close ~We

'un it 3 ssectianaJ.izivg valves from the main control room loads. The operator would then manua11y smrt the EEW pump A3 locally when i

oom to isolate unessential

iiz28ri988 i8:i5 SITE LLCENSING BFN 85 729 3iii P.86

~q gpggsz <continued) l required to opex'ate the unit and unit 2 diesel generators. Therefore, the Reactor Buildings (fire areas X through 3) will only require a manual action inside the control room'o ensure an adequate cooling water supply to the 4iesel generators.

The intake pumping station {part of fire area 25) will have a two-pump auto-start capability; however, a spurious trip of one pump would leave a single pump to supply the xequired diesel generators. To ensure adequate flow, the operatoz may be requixed to either close the EECW headex sectionalicing valves or manually start a second EECW pump at, the shutdown board. The circuits to sectionalicing valves would be unaffected by s fire in the intake pumping, station and could be opexated from the control room The Control Building (fire area l6) contains only the manual staxt and stop cix cuitry for all of the EEcW pumps. Consequently, s fire in the Contx'ol Building will not affect the auto-start circuitry of the EEL pumps and the EECW pumps would receive the automatic start, signal if the diesel generators should start. An EECW pump ~ould not. start during a Contr'ol Building fire only if the fire damages the manual stop cizcuitry causing the EECW pumps to spuriously stop. Zt would requize four simultaneous and identical spurious o ez'ations to cause a total loss of the EECW pumps, more spurious operations to cause s total loss of the E CW pump It would x'equire even fully loading up the diesel gsnezatars. These conditians are beyond the requiz'ed spurious operation assumptaons g1ven in Generic

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er4c Letter Le 86-%0.

One af the proposed fixes to ensuxe the diesel generator availability is to provide an automs ticc tri r p oof thee diesel,genexators based on high jacket water temperature. A simple trip system would have two sa feet, y gra radee temperature sensors with dual contacts at each diesel generator and connecting the logic l i l to' loca t erm na block oc for the tzip relays. The cost estimate foz this simple system is approximately 4589,770 for sll ree un s-includes all engineering, materials, and constru t ction costs. The estimated schedule is appx'oximately 30 ~eeks fox engineex'ing and 60 days for constructian. Howevez, this modification is nat desirable because he s e protect on for or thee diesel generstozs will also contribute to the unavailability of the diesel generators dur ng a a ~

high cost cannat be justified in view of the fact that the 'trip may reduce 1,'.

e reliability of the diesel genexators.

Based on the above, it is evident that the five minute manual act1on

.requirement for the EECW pumpss is unnecessary.

unnec Analysis demonstrates that, All fire zones m fme areas

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d'l for a fire in fire area locat1on s 4 throu g h 24, st less t two EKCW pumps would start automatically upon a start siign al from its assoc'.atet d 1es generator.

l-snd 3, except for cones -3 an 4 2-3 wauld auto start capability of at least two EECW pumps. Foz ha ve thee au or fire ire axes 25 and fire cones 1>>3 and 2-3 spurzous pump trip ri or valve closux'e would leave a single EECM pump w ~hich ustee until appropriate contz'ol room <or ich is aadeequs (or local) oca manual actions are taken. The manual actions ta will be changed to remove tablee w

ii~28<L988 i8: iS SITE LICENSIHG BFH 85 '729 3lla P.87 r~

t Continued) e'R~S)gag the 5-minute action accordingly making the 'soonest oper'atox ection 10 minutes, The 10-minute actions are either to vexify the scram and isolation functions ox to disable the HpCE system overfilling the reactor vessel.

if it is operating unreliably and t 3,4.6-2 << Credit for Dr all Coolers QQ~U~

1. State that drywall coolexs are not needed or taken credit for during an Appendix R event.

Z. State that drywell coolers will not be deliberately txipped unless the entxy conditions are met.

gg~pggf The drywell coolers are not requi.red for an Appendix R safe shutdown event.

However, the drywell cooling function will not be delibexately removed unless the following thx'ea entry conditions for the safe shutdown procedure are met:

1 A confirmed fire occurrence of significant severity in any plant location in which an immediate fire extinguishment cannot be achieved.

2. Loss of adequate high pressure makeup capability such that no x'easonable alternate exists but to proceed to a low pressure source of makeup.

3. Xnability to pxovide normal ox'mergency powex to vital equipment.

Since the dx'ywell. blowers are not an essential safety function, they were not A dix R xe uixements. To pxeclude the possibility of the drywall blower cables from contributing to a high impedance fault ul con condition the breakex'o the blower is tripped when its respective reactor motor i li operated valve b oar d s a gne d, For those cases in which the RHOV boaxd is

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used fax the A endix R event and in which it is essential that thee drywall cooli,ng function be defeated, either powex' to thee blowers ox source power to the board vill be removed.

.4.3-4 - Scr Verification

~ggES'f State th a t shutdown s u pxoceduxes will allow scram vex'ific e ification to be verified by txipping reactor pressure system (RPS) mo o g enerator motor (HG) set breaker from the battery board room.

11r28r1988 18:16 SITE LICENSING BFN

~85 729 3111 P.88

~ (I R~PEM Th e Appen dix R Procedure will require either opening the RFS NG set generator u p ut breaker on the battery board ar opening the feeder breaker to the RFS out HC set mator on the 480V reactor motor operated valve board, This action will ensure power is removed from the reactor protection system which causes an automatic scram on loss of power and verifies the scram will take place.

2 em 4.0,h.3 - Tome Level and Tem eratura Instrumentation

~RUEST

1. Summarize the previously made case (HEDC 31119, November 21, 1986 submittal) that the instrument ci.rcuits will survive a fire event.

2, summarize the previously made case (NEDC 31119, November 21, 1986 submittal) for not requiring the use of the instruments,

3. Provide cost impact of performing modification to ensure instrument indication availability during a fire event on elevation 593'f reactor building.

QEs~PPsE Documentation has been provided in MEDC-31119 and in the November 21, 1986 submittal which demonstrates that the suppression pool level and temperature instrumentation are not required to mitigate the design basis Appendix R avant.

Xn particular, the analysis has shown that the instrumentation will not be re to determine any operator actions to preserve paol level or renduuired re temperature. All operator actions are based on reactor con ditions o {reactor pressure and reactor ~ster level) and preanalyzed worse-case conditions in whi c h a ctions are performed regardless of the torus canditions. Operator act on inn thee Appen endix R Shutdown Procedure will not depend upon owled 8 e of o knowl torus level or temperature. All sources which are caps a ble e oof overfilling or dra n ng th e toorus have ve been determined and these sources will be isalated.

Th e prace d ur es will direct the operators to depressurize thee r reactor and establish a heat removal path {alternate shutdown coaal ing) based u p on low reactor water level or high torus temperaturee or the inability to determine torus temp erature. Analysis demonstrates that the peak tor'us temperature b 1 t rus temperature limits given the worst Hx'arne assumed {two thee&'ct to three hours) for performance of these operator actions, ons, Therefore, torus temperature and level process variables are not n necessar y to perform or control the functions given in XXX.L.2 of Appendix R-f the cable routing far the torus instrumentation was provided in the November 21, 1986 submittal. The cables or e inside the Reactor Building stay within its own ~nit. Separat on e wee

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11<28<1988 18: 16 SITE LICENSING BPN S 729 3111 P.89 R~{ SPOMSK {Continued) redundant trains exists on the 519 and 565 elevstians of each Reactor Building. However, the north area of khe 593 elevatian of the Reactor Buildings is the arse of the closest proximity for most of the torus instrumentation cables. Consequently, the situation is that for a fire in other than the north area (593 elevatjan) of each Reactor Building, the an)'rea torus instrumentation is available to the operator if hs desires ta canfirm the torus condition. Since a separate loop of the torus instrumentation is available at the backup control panel, the t.orus instrumentation is available for e fire event in the Control Building. For the north area of the 593 elevation, it is expected that sufficient instrumentation would survive a fire on this elevation at this particular lacatian where ths redundant trains of torus instrumentation are st the closest proximity. This is because the north area af 593 elevation is presently protected by an existing open head spray system and will be protected by e sprinkler system designed to NFPA-1$

requirements. The major combustible in this area is cable insulation which is either in conduit or coated liberally with Flamastic.

If the instrumentation did not survive, alternate means exist to determine ths torus conditions i.f the operator desires so. For example, the RHR eat exchanger inlet temperature can be used to determine the pool temperature.

The cables for the RHR heat exchanger inlet temperature sruti.th indication in the control raom is approximately 10 fest fram the closest proximity for the pool temperature indications. If the cables are eisa lost during the fire, thermowells are available to provide a local determination of the torus temperature. Direct torus water level indicatian can be determined by taking diffsrsnti.al pressure readings from the instrument taps at the local torus level transmitter. Therefare, various alternate means can be used to determine the torus conditions should the operator desire such information.

The approx ima teel y,,

cost to separate the tor'us instrumentation is estimated to be

$ 2 238 876 faro all three units, engineering, materials, and canstruction costs. The es This estimate includes all duration is approximately 30 weeks per unit for engineering which could be escalated by 50 percent due to Class 1B and safety-related pracurement and 30 to 60 days for construction per unit.

Xn summary~ thee high g costs of providing further protection for torus instrumentatian at the 593 elevation cannot be )ustif is{4 based on the f'allowing cansideratians.

a. The benefit of the torus instrumentation has been demonstrated ta bs minimal.

b. It is highly probable 'that the existi.ng instrumentatian would survive a fire in the critical area,

c. Alternate means tions are available to Ms of determining torus condit'on operators.

xx/85/ JWCIO JO< J I F

Item 4.0.h-4 - Thr'ee Phase Shorts on HI-Lo pressure Interface Cir cuit for Residual Heat Removal RHR Shut Down Coolin SDC Fla<

Contr al Valve PCV 74-48

~RU EST Commit to tag out the breaker far Fcv 24-48 to ensure removal o~ mat "e power.

State what steps aze necessary t,o ensure the valve is in its P>>P<<

position prior to tagging out and what controls exist to ensure the breaker is periodically verified for proper tagged aut pasition

~Rl~gSE Far HI-LO pressure interface valve FCV 74-48 motive power will be removed by opening the breaker .disconnect switch and placing a valve hold ozdex tag, on the breaker'nd the cantrol zoom switch. Before tagging out the equipment>

.... position indicating,lights are available to ensure the valve's proper position. The tagout i.s verified by a Semiannual Hold Order Audit which tagged'ut includes verification of the position of the tagged equipment.

Zte 4.0.8.4 - Hanua ctions in Fire Areas 4 8 9 X2 and 1,3

1. ConfiarL that the manual actions performed in these fire areas do nat x'equi.re caid shutdown repairs ta accomplish.

2, Clarify what manual actions are performed in these fire areas

3. Commit to correcting mistakes i.dentified in manual action calculations (Enclosure 4 of November 21, 1986 submittal to NEDC 31119) for these fire areas'nclosure 4 of the November 21, 1986 submittal provided a table of manual actions for each fire area. The table fax fi.re areas 4, 8, 9, 12, and 13 mistakenly identified an action to be performed an a potentially fiz'e damaged circuit breaker. This error was identified duri.ng development of the operating procedure end will be corrected in manual action calculations. The valves will be di.sabled by verifying that pawez'o the board is removed from its pawer source located in a different fire area. The valves are disabled to valve using the local handwheel. 'o prevent spurious operation and allow the operator to manually positian the repair or repair proeeduxe is required to obtain hot or cold shutdown conditions foz'hese areas.

12~28>1988 18r 2.7 SITE LICENSING pFN 8S 729 3111 P.ii 26-0 20 <<9 i9 I

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11/28/1988 18: 18 S I TE L I(;hN'hlNU 'l-N Item 4.0.8 5 - III.G. 1 and III.G.3 Fi t'e Areac

~R~UES 1 ~ Confirm that III,G.l fixe ax'eas identified in November 21, 1986 submittal to NEDC 31119 do not contain any redundant cables, equipmont, ox'omponents.

2. Clax'ify the withdx'awal of the zeal.G.3 exemption for the turbine building.

R SPO gg The following information is provided to clarify fire area ZZZ.G designations.

A fixe area designated solely as III.G.1 will have availablo redundant equipment, cabling, and systems for Appendix R that, is not located in that fire area.

The November 2l, 1986 submittal clarified the definition for alternate shutdown such that the turbine building (part of fix'e ax'ea 25, subject of exemption i) would now be classified as IXI.G.2 fire area. Exemption i.

includes a XXX.C.3 exemption for the turbine building which is no longer needed. Therefoxe, i that portion of Exemption will be withdrawn.

The intake pumping station fire axea classification of XZZ.G.1 as shown in the november 21, 1986 enclosure 3 table should bo IXZ.G.2.

It 4.b .6 T sti oi Safe Shutdown S stems

!KQEEZ

1. Describe the initial, postmodification, and periodic testing of remote shutdown panel 25-32.

QEJi~~S; The backup contx'ol panel was ini,tially tested in the SFN preoperational test program using General Electric test procedures. Shen modifications are performed, postmodification testing is required to ensure the affected equipment vill operate as designed, The x'emote shutdown panel and electrical distribution system are currently'ested each operating cycle. This testing includes checks to determine remoto opexati.on capability and circuit i.solation from the Control Building.

~ 11r28r1 Jtlv ill:1u hl I C L JVCIROllBQ DCI'V Xtem from Section 3.c ~ 3 oE Draft Sachet Evaluation Rc art REVEST State what manual actions are being taken to ensure the fallowing objecti<<s can be accomplished within 10 minutes.

l. Transfer to local control {if needed) and ensure closure of main steam isolation valves (HSEV).

2 ~ Close the high pressure coolant injection {HpcX) steam supply (in the fire affected unit) to prevent water intrusion into the shutoff'valve main steam lines.

Manual actions have been provided to verify that the main steam isalation valves are closed following the automatic or manual control raam isolation.

This acti. on involves accessing the remote control panel (panel 25-32),

transferring the control switches, end placing the control switch for each MSZV to the closed position. There are eight valves (four inboard and four outboard) and the action on the panel requires only seconds to accomplish, The HPCZ system can potentially operate spuri.ously requiring an action to prevent filling the reactor vessel up to the main steam lines. To prevent this occurrence, the HPCZ steam line will be isolated by closing the steam supply valve to the HpcX turbine. This action requires the operator ta transfer the breaker to emergency on the 250V reactor motor operated valve board h and then operate the control switch to the closed position. The access and actian time should only require one or two minutes.