ML20032E556

Forwards Comments on NRC 800703 Draft Evaluation & 810814 Safety Evaluation of SEP Topic V-11.A, Isolation of High & Low Pressure Sys

ML20032E556
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/12/1981
From:	Vincent R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
TASK-05-11.A, TASK-5-11.A, TASK-RR NUDOCS 8111200686
Download: ML20032E556 (17)

Text

e O Consumers Power Company Generet offices: 212 West Michigan Avenue. Jackson, MI 49201 * ($17) 784-0650 November 12, 1981 iM

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Att Mr Dennis M Crutchfield, Chief 1.

h Director, Nuclear Reactor Regulation Operating Reactors Branch No 5 g

US Nuclear Regulatory Commission D

Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - SEP TOPIC V-11.A, ISOLATION OF HIGH AND LOW PRESSURE SYSTEMS By letters dated July 3, 1980 and August 14, 1981, the NRC issued a draft evaluation and a safety evaluation for SEP Topic V-11. A (Isolation of High and Low Pressure Systems) for Big Rock Point. Consumers Power Company has reviewed these documents and provides the attached comments for your consideration.

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i Robert A Vincent Staffl Licensing Engineer i

CC Director, Region III, USNRC NRC Resident Inspector-Big Rock Point Attachments 1o35-s

/N 8121200686 811112 DR ADOCK 05000155 PDR

TCNR Th-81 ENCLOSURE FW4 0F SEP TCPIC V-11. A "ISOLATICN CF HIGH AND LOW PRESSURE SYSTEMS" BIG ROCK POIN" PLANT A.

Summary of NRC Evaluation The purpose of the NRC evaluation was to determine if the electrical, instrumentation and control (EI&C) features, used to iscAate syste=s with a lover pressure rating than the reactor coolant primary syster, are in ec=pliance with current licensing criteria. To make such a determination, the NRC evaluated the EILC features carrently installed at the plant to isclate the Shutdown Cooling System (SDCS) and the Core Spray System (CSS).

As current licensing criteria, the NRC cited the Standard Reviev Plan and 3 ranch Technical Positions as described below.

SDCS Criteria Isolation requirements for F3R systems contained in BTP RS3-5-1 are:

(1) Se suction side =ust be provided with.the following isolation features :

a) Two pover-operated valves in series with position indicated in the control roo=.

b) The valves =ust have independent and diverse interlocks to prevent opening if the reactor coolant syste= (RCS) pressure is above the design pressure of the RER syste=.

c) The valves r2st have independent and diverse interlocks to ensure at least one valve closes upon an increase in RCS pressure above the design pressure of the PRR syste=.

(2) Se discharge side =ust be pr vided with one of the following features:

a) The valves, p4'm m indicators, and interlocks described in (1) r.) through Li c) above.

b) One or = ore check valves in series with a norm 11y-closed pover-operated va'.ve which has its position indicated in the control room. If this valve is used for an Emergency Core Cooling Syste=

(ECCS) funeen, the valve =ust open upon receipt of a safety injection signal (SI') when RCS pressure has decreased below FER syste= design pressure.

c) Three check valves in series.

d) Two cheek valves in series, provided that both =ay be period-ically checked for leak tightness and are checked at least annually.

2 CSS Criteria Isolation require =ents for the ECCS are given in SRP 6.3. Isolation of ECCS to prevent overpressurization must =ett one of the following features:

(1) One or more check valves in series with a normally-closed motor-operated valve (MOV) which is to be opened upon receipt of a SIS when RCS pressure is less than the ECCS design pressure.

(2) Three check valves in series.

(3) Two check valves in series, provided that both may be periodically checked for leak tightness and are checked at least annually.

In addition.o ORP 6.3, other applicable isolation criteria are contained in BTP EICSB-3 vhich states:

(1) At least two valves in series =ust be provided to isolate the syste= vhen RCS pressure is above the system design pressure and valve position should be provided in the control room.

(2) For syste=s with two MGVs, each MOV should have independent and diverse interlocks to prevent opening until RCS pressure is below the-syste= design pressure and should automatically close wh..

RCS pressure increases above syste= design pressure.

(3) For sycte=s with one check valve and a MOV, the MOV should be inter-locked to prevent opening if RCS pressure is above syate= design pressure and should automatically close whenever RCS pressure exceeds syste= design pressure.

According to the NRC evaluation, the SDCS and the CSS syste=s do not meet current licensing criteria contained in SRP 6.3 and RSB-5-1 for isolation of high and 1cv pressure syte=s as listed below.

(1) The SDCS isolation valves C7 not have diverse and independent interlocks to prevent operation when RCS pressure exceeds SDCS design pressure.

(2) The CS syste= has no interlocks to prevent opening the isolation valves from the local control station when RCS pressure exceeds CS syste= design pressure.

B.

Detailed Review of NRC Evaluation B.1 SDCS Attachment #1 (Dwg M-107) shows the SDCS and the motor-operated isolation valves (MOVs) used to isolate the SDCS frc= the high-pressure reactor system.

As can be seen, two M07s (MO-7056 and MO-7057) are on the suction side of the SDCS and two MOVs (MO-7058 and MO-7059) are on the discharge side.

3 Attachment #1 also shows that these valves are electrically operated by pressure switches PS/619 and PS/620. The system has been designed such that the pressure switches operate the MOVs to ensure that the SDCS is not overpressurized by the reactor system.

Attach =ent #2 (Dvg E-112, Sh 1) describes the electrical interlocks that have been designed into the valves' control circuits to prevent the valves from being open should the reactor system pressure exceed the SDCS design pressure. Attachments #2 and #3 (instru=ent data sheet, Page 75) show that pressure switches PS/619 and PS/620 are both set to open their contacts upon a high reactor syste= pressure setpoint of 300 psig. Therefore, if the reactor system pressu2e is 300 psig, the pressure switch contacts vill be open and auxiliary reltys PSX619 and PSX 620 vill be deenergized.

Deenergizing these auxiliary relay 1 vill result in opening Contact 1-2 of PSX619 (Valve MO-7056 is typical) thereby preventing Valve MO-7056 fro =

being opened.

In addition to preventing the valves from opening if the reactor system pressul'e is. greater than the SDCS pressure, the interlocks also serve to auto =atically close the valves, if open initially, should the SDCS pressure be exceeded. As can be seen in Attach =ent #2, should the reactor system pressure exceed 300 psig the PS/619 and PS/620 contacts vill open to deen-i ergize relays PSX619 and PSX620 resulting in closing contact 3-h of PSX619 (MO-7056 typical) and applying a close signal to the valve.

The pressure switch setpoint of 300 psig prevents the SDCS fro = being overpressurized since the design rating of the SDCS is 315 psia (see Attach =ent #L which is actually a portion of Faragraph h.l.2 of the plant's Technical Specifications).

Although interlocks exist to prevent the SDCS from being overpressurized, the existing interlocks are not diverse as required by BTP RSB-5-1.

Credit should be taken, however, for the fact that either of the two aforementioned pressure switches PS/619 or P7/620 vill result in protecting the SDCS against overpressurization.

Contrary to that chown in Attachment #1 and as described in Paragraph 31 or che subject NP.C ev'aluation, either pressure sviten vill serve to operate all fosr valves. As shown in Attach =ent #2, both pressure switch contacts (PS/619 and PS/620) are wired in series to operate two parallel auxiliary relays PSX619 and PSX620.

(The sche =atic representation has been verified as accurate ut 113 zing the viring diagrams. ) The auxiliary i

relays co=plement each other to act. ate all four MOVs. Therefore, should one of the in-containment pressure switches (PS/619 or P' '620) fail to operate and open its contact when the reactor pressure exceeds 300 psig, the other pressure switch will initiate the safety function. The operation of the functional pressure switch vill result in the closure of the four MOVs and the prevention of reopening the valves once closed.

It also should be noted that the system can also tolerate the failure of an auxiliary relay to energize. As shown in Attachment #2, relays PSX619 and PSX620 are wired in parallel. The contacts of PSX619 are wired into the control circuits of both SDCS outboard valves (MO-7056 in the suction line and MO-7058 in the discharge line). The contacts of PSX620 are wired into the control circuits of both SDCS inboard valves (MO-7057 in the suction line and MO-7059 in the discharge line). As a result, a failure of either relay (PSX619 or PSX620) to energize vill still allow one valve in each SDCS line to isolate the'SDCS.

k In conclusion, it also should be noted that position indication for each valve is located in the control roc =. This is shown on Attachment #2 vhere the main centrol panel designatien " Col" appears on either side of the red and green positi on lights.

B.2 CSS Attach =ent #5 (Dwg M-123) shevs the CSS and the two MOVs on either side of a check 'alve in both the no::le and ring spray header piping. These valves serve tc ist late the CSS frc= the reactor syste= vhenever the reactor syste=

pressure exceeds that of the CSS design rating. As can be seen in the Attach =ent, isolation of the no:nle header is afforded by MOVs MO-7070 and MO-TC T1 and by check valve VPI-303 (located between each MOV). Isolation of the spray header is afforded by MOVs MO-7051 and MO-Tc61 and by check valve VpI-30L. As can be seen in Attact=ents #6 and #7 (Dvgs E-103 and E-11h Sh 1, respectively) each valves' position is irdicated in the centrol roc =.

Chis is shcvn in the Attach =ents where the =ain centrol panel designation

" Col" er "CO2" appears on either side of the red and green position lights.

As indicated in Attach =ents #6 and #7, all four MOVs vill aute=atically open to provide ECCS jection whenever the reactor pressure drops to 200 psig and reacter level rops to 2'-9" above the core. Attach =ents #8 and

1. 9 (instru=ent data sheets pages 8' and 39, respactively) give the pressure and level setpoints. Technical Specificatien Tables 11.3.1.ha and 11.L.l.La also identify the setpoin*

si an uld be noted that the MCVs automatically open at a pressure tscew! +t gret.er than the design rating of the CSS.

(The design rating of the d55 is actually the rating of the fire syste=

which is 150 psig at 80 F.)

paragraph 3.2 of the NRC evaluation erreneously indicates that the MOVs autc=atically open when the reactor syste= pressure decreases belev the CSS syste= design pressure. Although the four MOVs open at a reactor syste= pressure screwhat greater than the design pressure of the CSS, the CSS will re=ain isolated frc= the reactor syste=.

Isolation of the CSS vill continue until the reactor syste= pressure decreases belev the operating pressure of the CSS "hich v 11 unceat check Valves 7PI-303 flecated between MC-7070 an:1 MO-7071 i and '?I-30h (located between MO-7051 and MO-7061) and allow core spr n water te.nter the core.

The basis for selecting a reacter syste= pressure of 200 psig as the setpoint at which valves MO-7051, MC-7061, MO-7070 and MO-7071 aute=atically open deals with reacter vessel water inventory during a =ajor 1 css-of-coolant accident (LOCA). Selecting 200 psig as the setpoint, ensures that the opening cycle for these MOVs is initiated early enough in the LOCA depres-surization transient such that each of the valves is open and ready to pass flev vhenever the reactor pressure decreases below the operating pressure of the CSS.

As stated in the Big Rock Point Flant FHSR Paragraph 13.k.2.h, ".

. the principal purpose of the core spray syste= is to pro-tect against fuel damage.

.". The FHSR paragraph goes on to state:

" Operation of the core spray syste= vould be autonatically initiated when the reactor syste= pressure dropped to 200 psig, which would occur about 15 seconds after the syste= rupture." It should be noted that this statenent provides an indicatien of hev rapid the reactor syste= depres-surization is during a " Maxi =u= Credible Accident".

5 Relying er the afore=entioned check valves to isolate the reactor syste=

frc= the CSS does not jeopardize plant safety for tvo reascus. The first reason is that these check valves are required to perfor= this isolation during only a a all percentage of the total depressurization transient.

During reactor syste= depressurization, the check valves (by the=selves) are expected to isolate the CSS fro: the reactor system for a period of ti=e at which the reactor syste= pressure drops for= 200 psig to the operating pressure of the fire system which is approximately 110 psig.

(It should be noted that the CSS pressure could be as high as ik3 psig if the fire pu=ps are runnirg without a discharge flovpath.)

The second reason is that a high degree of ecnfidence in the ability of these valves to isolate exists due to frequent leakage testing and a rel-a atively repair-free - a tenance history. Each check valve is leakage checked conthly per surveillance precedure T30-22 "Energency Core Cooling Gyste= Valve Tests".

During the verification, the ability of the check valve to isolate the reactor syste= h: gh pressure frc= the CSS operating pressure is conducted while at power

'.th the electric fire pt. p running.

The check valve's ability to isolate 1. evaluated by alternately opening each MCV cn either side of the check valve and then observing the fire syste= pressure for an increasing indication indicative of check valve leakage. This test is perfor=ed =enthly or both check valves; VPI-303 and VPI-30h.

Regarding the valves' =aintenance history, plant =aintenance records were reviewed over the last five year =. This review indicated that since 1976 each check valve has required repair for internal leakage only once. VPI-303 was repaired during th? 1978 refueling cutage an:1 ViT-30L vas repaired during the 1979 refueling outu;e. It should be noted that as part of the plant's preventive =aintenance progra=, the valves are disasse= bled every refueling cutage f:r inspection. ~his disasse=bly and inspection is perfor:ed every outage regardless of whether or not valve problens are suspect.

Regarding the response of the four MOVs whenever the core spray initiation signal is cleared, the NEC evaluation is incorrect. The NEC evaluation er-reneously states that the MCVs vill autc=atically close upon clearing the initiation signal or increasing ECS pressure above GS syste= design pressure.

As shown in Attach =ents #6 and #7,neither of the afere=entioned MCVs (MO-7051, MG-7061, MG-7070 and MO-7071) vill autc=atically close. The attachments shev that no autc=atic c1cse features exist in the valves' control circuits.

In additien, it shculd be noted that neither of the four MCVs feature interlocks to prevent opening the valves frc= the control roo= when reactor syste= pressure exceeds the CSS design pressure.

It should also be noted that Valves MC-7070 and MO-7071 can be also opened frc= the le al control station when the reacter pressure is greater than the CSS design pressure.

Althugh neither of the four nor ally-closed (Plant start-up check sheet

1. A-8 " Post-Incident Syste=s" calls for all four of these valves to be closed prior to start-up. ) MOVs feature electrical interlocks that would prevent the: fre= being opened while the reactor syste= pressure exceeds the CSS design pressure, the probability of overpressurizing the CSS is very lov since the check valve vill ulti=ately provide isolation capability.

Assu=ing a vorst-case, highly unlikely condition that these four valves vould be cpened during full power operation, the Censumers P0ver Co=pany would expect that each of the tvc check valves (VPI-303 and VPI-30L) would serve to isolate the two syste=s.

"his expectation is based en the folleving facters:

1 i

a

1) As previously described, each check valve is tested =onthly.

During the test, the check valve's isolation capability is de=onstrated by opening the check valve's downstream MOV thereby pressurizing the valve with the reactor syste=.

2) As also described, the check valves' =aintenance history is relatively repair-free.

3) The check valves have been designed to withsta=d such conditions.

The design pressure and te=perature ratings for check valve VPI-303 is 1500 psig at 850 F.

Although the design ratings for VPI-30k are only 900 psig at 3500F, analysis has shown (and Ite: 1 above provides substutiating evidence) that the valve is capable of withstanding reactor syste= pressure at expected te=peratures should the MOVs be inadvertently opened. Dtter DABixel to DKDavis (USNRC) dated 8/2'+/77 provides this analysis.

C.

Corrections To The NRC Evaluation C.1 SDCS Paragraph 3.1 of the subject NRC evaluation incorrectly state 9 that the

" interlocks for these valves are not diverse since only one pressure indicator is used for each valve".

There are no pressure indicators in the interlock syste-

• eality, the interlocks are not diverse since only cne interlock contact frc= only one pressure =easuring system is incorper-ated into the valves' opening and closing circuitry.

Paragraph 3.1 of the evaluation also incorrectly states that "The inboard Lesest to RCS) isolation valves both use the same pressure sensor. 2e outboard valves also use a co=nen sensor." This statement is corrected in Section 3.1 of this repert.

C.2 CSS Paragraph 3 2 of the subject NRC evaluation incorrectly states that the MOVs vill open upcn a CSS initiation signal when the reactor system pressure has decreased belev the CSS syste= design pressure. As described in Section 3.2 cf this report, the MOVs open upon a CSS initiation signal which occurs at a reactor pressure of 200 psig. As described in 3.2, this pressure exceeds the CSS design pressure of 150 psig.

In addition, Paragraph 3 2 also incorrectly states that the MOV vill auto =atically close upon clearing the initiation signal or increasing reacto:-

syste= pressure above CSS syste= design pressure. As described in Section B.2 of this review, the MOVs have no autc=atic closing circuitry.

D.

Non-Co=cliances and Justification D.1 SDCS The existing interlocks for the SDCS isolation valves (MO-7056, MO-7057, MO-7058 and MO-7059) at the Big Rock Point Plant are not diverse. Therefore, the isolation provisions of the SDCS do not =eet the current licensing criteria of BTP RSB-5-1.

Although the interlocks for the SDCS are not diverse,

T the isolation syste= can tolerate a failure of one of the in-containment pressure switches and/or the failure of one of the pressure switch auxil-iary relays and still te capable of closing and keeping closed one MOV in the SDCS suction line and one-MOV in the SDCS discharge line.

In an atte=pt to ensure that these valves will not allow the SDCS to be overpressurized no matter hev small the probability may be, the valves' control circuits are maintained deenergized during power operation.

Syste= Cperating Procedure SOP 5 " Reactor Shutdown Syste=" states :

'To not place the shutdown syste= in service at a reactor pressure greater than 300 psig. Place the syste= in opeation after reactor pressure has been love ed to approxi=ately 200 psig."

The procedure also states:

"The shutAovn syste= isolat. ion valves are interlocked so that they cannot open at reactor pressure greater than 300 psig." Paragraph 4.1.2(b) of the plcnt's Technical Specifications states: "The shutdown cooling system shall be ready for service during power operations with the LSO volt circuit breakers for isolation valves MO-7056, MO-7057, Mo-7058 and MO-7059 checked "cpen" when reactor pruture is above 300 psig." This requirement e aures that the valves' control circuits are deenergized above reactor system pressures which exceed the design rating of the SDCS. This elimates the possibility of energiz:.ng the valves' opening coil and opening the valve whenever the reactor pressure exceeds 300 psig (or the design rating of the SDCS).

D.2 CSS Contrary to the require =ents of BTP EICSB-3, the MOVs M0-7051, MO-7061, MC-TC70 and MO-7071 automatically open upon a reactor syste= pressure thr.t is so=evhat greater than the design pressure of the CSS. le described in Section B.2, however, a check valve exists in each core

. cay injection line which provides adequate isolation between the reaeter syste= and the CSS until the reactor syste= pressure decreases below that of the CSS l

design rating.

Also contrary to BTP IICSB-3, the MC7s do not feature aatonatic close circuitry to drive the valves closed should the initiating signal clear or should the retetor pressure increase above the CSS design pressure.

In addition, the MOVs do not feature interlocks to prevent cpening the valves when reactor pressure exceeds the design pressure rating of the CSS.

Al-i though the Big Rock Point Plant CSS MOVs do not confor= to current licensing criteria as stated above, the unique syste= design and operation as described in Section B.2 of this reviev meets the intent of the current criteria, i

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Reactor Power Operation Cooling:

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Page w

INSTRUMENT Sy PtilD LOC ELECT ELECTRICAL RANCE AND OR llHER-PM INST I S

DATA Y

NUMBER r

SERVICE DWG DWG DWG SWITCN PolNT SETP0lNT CONNECTIONS PROC SHEET P

gg E

(S-HtC5 CS CRu 4ecum.t.eek Detection C40122 'Il0030 1107 )4

?O2-l*"dll f*l, 2.*3, M

. Iso set Point (Flued)

Local finge_

L t.3-pte$ C6 Catu Accus.Lemk lktection G40622 410u 30 agjk, cop l*(le % 28( 2% [I No Set Point. (Fised)

Imcel Nolase p

I.G-8t105 bl CitD Accum. Leak Detection G4012? 430010 307}4 OO2.l-4)tt'/gl,'/E,,77 No Set rotat. (Fised)

Racal Nveces O

la-kIGS D2 Cit D Accine. Leak Duteetton G40122 i130030 ag4_ CO 2.l.q)e n /*l,7Q 2 7 No Get Point. (Ftmed)

Imcal ta.wicea p

LS-latc$ P)

OtD Accus. Leek Detection G40122 6130030

30734 C02.l 316) Oaj,2 6, J )

No tiet Point - (rtaed) toen! Devices O

LS-tiL25 L'e CbD Accusa. Leak betection cfiolM 110010 ;)i73 0 m.g. ptg.g y.rg g No ces Posnt. (6: sed) socet w vicae I

LS-htf5 DS CRD Accum.Lenk Detection G40122 430u)0 3Gijk (*p g.l. g gp tjel '2 6,7 f No Set Potat - (rtmed) local Devicae D

LS-h!C) b6 CHD Accum. Leak Detection G40122.3300)O 10734

(*pp.g.,g g g j4 7,g y J N3 Cet Point - (Fised)

Local Ibericais

[

1A-h12$ Et CHD Accus. leak Detection G40122 130010

)dT14 s o 2*l- 08) '.8 61 '/5 / #

no Set Point. (Famed) tocol tiensaes

^

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p I.S-P.Ic5 E2 CHD Accwn.Lenk Detection G40122 130030 130734 N3 2-l-elll /476,M. taa Set Point. (Flued)

Local swet.es O

LS-B125 El ChD stec T.. Leak Detection c40122

';)0030 30734 CO2.l.s)s) Q*l, ]/d 2") No Set Point. (rtsed) 8.ocal h vises o

LS-til25 E4 CitD 4ects. Leak leeteetten 140122

30030

30734 Co,'Al (M 942dy No Set Polot. (rtsed) local tweseen 0

IS-RIDS E$

ChD Accum.Lenk Detection d40122

)0030 G)G73'l 0 0 2 *l-lie)aff,2 4 2 ]

No Set Point. (Flued)

Lacet 1.tvecce 0

LS-ft!2$ 16 CRD Accues. Leak Detection C40122 G100]0 1307}4 { gg 7-l. h g ) 'f.l, 7*d / No Set Point. (FlaeJ) tacel hvises D_

LS-htc5 F2 CHp Accum.leck Detection C40122 G300)O G30T)4 C(yJ.-l-Pq l ).l$ 2 8] ') J 14u Set Point. (FineJ) tacal Ovvites O

LS-Dic5 F3 CHD Accum. Leak Detection C40122 G30010 130734 Co*pt l)f) :/4ll 2Sj f ido Set Potat. (Fired) facel Devices D

LS-k!US F4 Citu Accue. Leak Detection C40122 G)0010 030lfe CG2.-l.i tO laj,26g Me set Point. (Flued) 1 meet inevice.

D LS-ItIM) F5 CitD Accues.14ak Detection C40122 C)o030 130734 (to2. l-O s t 2 1,7.*; / / i30 Set Point (blue 4)

IAcal Devitee D

LS-EE06A hP3 rest Sta.Drue 60*

C40121 G)l006 X)OT4)

Cte-5, CatT. i:CtB tewTrtP:lB.5*setowg IAs-isro n thP-4 P

LS-IlkT4B ltrs East Sta.Drue 60*

C40121 G)l(X)6 43074)

CO2-5. EB2), lib 24 law Trip d 8.$*below C I2f-pin us -

Thr 4 P

LS-REu9Al PTS Heactor Water tavel C40121 G100k2 110114 72 32 law Triptf.2'9'Above Core Ho-Tu'a t FP-2 P

L3-R}DJA2 BP3 Henctor %%ter level C40121 G)0042 K3074)

CO2 5 CCl). CGth low Trip I 2'9'Above Core IAI R Ts)A INP-2 P_

14-bE0981 P!3 heactor W ter h vel C40121 G)MA2 5 13 72 31 Low Trsp:5 2'9 ALove Core MO 7o6 rar-2 r

E5ittx6u2 ItP3 iteactor W ater Level C4012t 310u42 13014~)

002-5 DDl9, it:20 tov Trip:1 2*8/Above Core Ill HrJ13 ipr-2 P

1.5-ItE0W1 FIS Reactor W ter level C40121 G30042 130114 32-32 Isw Trip:12'9"Above core MO-To$1 InP-2 P

to-ittV)C2 kPa Reactor Veter Level C40121 Gluuh2 43oT4)

(X)2-5, Cat 5, Cat 6, now Tite:I 2'9"Above Core tal-Re 03A ISP-2 P

IS-RE0')D1 PIS lteactor Water leval G:80124 010042 130114 72 31 few Trip:E 2'9"ALove Core HO-Td8 g2 P

tS-hh.09r2 L*G peector W4er Level C40121 C)0042 i33 6 E5 Esat. I!D22 tow Trie:E2'9"Above care tal-DEJ ts litP-2 TS iiTSA T iS'~-~ naaetor Wter tevet chot2a 63iki

solo) 52-2a4) tow trip

g 2'9"above Core MG-N(O Il4P-2 P

T II.-tiL16F PIS inemator W ter Lev.!

c40128 a10042.:3010)

$2-2D42 Low TrtP E 2'9"Above Core s0-7071 IHP-2 P

LS hW)0 PIS steactor Wter Leve

• j t4-hE09tt PIS Reactor Wter level _

C40621 G0042

1010) 52-2h41

.fow TriptE 2' R have Core HD-7070 JUP-P P

C40121 430042 3010)

S2-PD42 saw TRIP:I 2'9"Above Core MO-7071 ISP 2 P

i 13

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