B11736, Forwards Addl Responses to Questions Raised Re Proof & Review Tech Specs.Marked-up Tech Specs Encl

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Forwards Addl Responses to Questions Raised Re Proof & Review Tech Specs.Marked-up Tech Specs Encl
ML20133F419
Person / Time
Site: Millstone Dominion icon.png
Issue date: 09/20/1985
From: Mroczka E, Opeka J
NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To: Youngblood B
Office of Nuclear Reactor Regulation
References
B11736, NUDOCS 8510110022
Download: ML20133F419 (34)
v • d • e


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P.O. BOX 270 HARTFORD. CONNECTICUT 06141-0270 L L J C',Z',OZ72'Z~. (203) 665-5000 September 20,1985 Docket No. 50-423 Bil736 Director of Nuclear Reactor Regulation Mr. B. J. Youngblood, Chief Licer sing Branch No. I Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Reference:

(1) 3. F. Opeka letter to B. J. Youngblood, Technical Specifications - Proof and Review, dated September 19,1985.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 3 Technical Specifications - Proof and Review In reference (1) Northeast Nuclear Energy Company (NNECO) submitted f information requested by the staff concerning certain draft technical specifications for Millstone Unit No. 3. Enclosed please find additional NNECO responses to questions raised.

We trust the attached will resolve the Staff's concerns. If there are additional questions, please contact our licensing representative directly.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY et.al BY NORTHEAST NUCLEAR ENERGY COMPANY Their Agent k IWN

3. F. Opeka Senior Vice President

/ 4t%

By: E. J. () j ig Vice Pres ident/firoczka 8510110022 850920 PDR ADOCK 05000423 A PDR

' STATE OF CONNECTICUT )

) ss. Berlin COUNTY OF HARTFORD )

Then personally appeared before me E. 3. Mroczka, who being duly sworn, did state that he is Vice President of Northeast Nuclear Energy Company, an Applicant herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Applicants herein and that the statements contained in said information are true and correct to the best of his knowledge and belief, b 'A 191. W Notary Public-My Commission Expires March 31,1986

ADDITIONAL REVIEW REQUIRED ltem 6: Specification 3.1.2.1 and 3.1.2.2 - Boric Acid flow path - Tank Versus Tanks (Refer NU letter, dated August 21,1985 (Bil671)).

NNECO's Response:

The proposed change of wording to specifications 3.1.2.1.a and 3.2.2.2.a are necessary to make these specifications consistent with the Millstone 3 design and the wording of the specification 3.1.2.5 and 3.1.2.6. This specification will allow the use of one boric acid tank or two. A change of " contained volumed" to

" useable volume" is requested in specifications 3.1.2.5.a and 3.1.2.6.1 to allow use of both tanks or one. The minimum volume required depends upon the useable volume in the boric acid storage tanks and the number of tanks in use. A useable volume of 4100 gallons should be required in Modes 5& 6 for specification 3.1.2.5.a. A useable volume of 21,020 gallons should be required in Modes 1,2, 3 & 4 for specification 3.1.2.6.a. The unuseable volume in each boric acid storage tank is 1300 gallons.

ADDITIONAL REVIEW REQUIRED item: Technical Specification 3.2.5 (Table 3.2-1) FSAR Table 14.0-3 Explain the difference in technical beween specification tableTavg 3 =2-1590.7 degrees-F and the vessel as presented average temperature = 537.1 utilized in the accident analysis as presented in FSAR Table 15.0-3.

NNECO's Response:

FSAR Table 15.0-3 lists the nominal value of plant parameters utilized in accident analysis. To these nominal values are added appropriate steady state the maximum analytical limit for temperature is 595 errors. In the degrees-F case of Tavkhe as presented in bases section of technical specifications (B3/4.2-6).

Table 3.2-1 of technical specifications places limits on indicated reactor coolant system Tavg. This means that allowance has to be made to account for the errors encountered from the sensor to the indicator where the temperature is read. To account for these errors a channel statistical allowance is calculated similar to the procedure utilized in the statistical setpoint study for protection systems. This allowance is then conservatively applied (subtracted) to the analytical limit (595 degrees-F) to arrive at the indicated value (590.7 degrees-F).

1

ADDITIONAL REVIEW REQUIRED ltem: 3/4.3.3.6, Accident Monitoring Instrumentation NNECO needs to provide the minimum number of channels for the

" Containment Area. Purge and Exhaust Isolation Radiation Monitor" (read Fuel Drop Monitor).

- NNECO's Response:

Tables 3.3-10 and 4.3-7 have been reviewed and are being resubmitted. The instruments included are those that have been designated as Category I variables. This listing may change as a result of the accident monitoring instrumentration SER when finally issued.

The action requirements for 3.3.3.6 have been reviewed and modified to reflect the instruments included in the tables.

It is requested the line item " Containment Area - Purge and Exhaust Isolation Radiation Monitor" be deleted. The requirements for this monitor are incorporated in specification 3/4.3.3.1, Tables 3.3-6 and 4.3-3.

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SEP 20 '85 10:23 QATEL SERVICES CORP. PAGE.03 N M & YEVTN T py AUG 161985 INSTRUMENTATION

_ ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICA8_ILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE accident monitoring instrumentation -

- channels less than the Total Number of Channels shown in Table 3.3-10,' restore the inoperable channel (s) to OPERA 8LE status within 7 days, or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the number of OPERABLE accident monitoring instrumentation channels except ? : _ i t  : i M ;' -- :- - "  ; - - - - ' ' -- ' " -*---

.'" - t " ^' -- - - - M ' ^ ^ ' : x M 1_ '

range radiation monitor, and the rS. the containment atmosphere-high ctor coolant radiation level '

monitor less than the Minimus Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERA 8LE status

)

within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With the number m- ---u-_ of OPERABLE channels for i zi ~ . u ... _ -- u . <. u ,.m . _ . /. ' t . . . . _ . ,m

..,, . o . % s

.ser Ue"clntainment. atmosphere-high range radiation monitor, or the reactor coolant radiation level monitor less than required by the Minimum Channels OPERA 8LE requirements, initiate an alternate method of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and either restore the inoperable channel (s) to OPERA 8LE status within 7 days or prepare and submit a Special Report to the Commission, pur-suant to Specification 6.9.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERA 8LE status,

d. The provisions of Specification 3.0.4 are not applicable.

1 SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated I l

OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the (

frequencies shown in Table 4.3-7. l

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> I TABLE 3.3-10 .:

I' ACCIDENT MINITORING INSTRtmENTATION f

5

! TOTAL MINIMLM N5 i INSTRIDENT NO. OF CHAfGIELS j ro !,;

CHANNELS OPERA 8LE *

1. Containment Pressure ..

e

. c. Norinal Range m ,

b. 2 Extended Range 1 2 o  !'

1 -

2.

Reactor Coolant Outlet Temperature - T  ;,

HOT I"Id' I'""} A !! ##7 ' i 1

' 3. -

Reactor Coolant inlet Temperature - Tg (Wide Range)

! '4.

///kAi 1

'km -

Reactor Coolant Pressure - Wide Range i 2 1 "

1 . 5. Pressurizer Water Level 2 1 N

m <

6. Steam Line Pressure 2/ steam generator 1

1/ steam generator 1/ steam generator

't.

Steam Generator Water Level - Wide Range g;

M

- .. - 1/ steam generator 1/ steam generator '

j S.

Refueling Water Storage Tank Water Level Afdarns a VL 2 1 t l 10. " "'"" "'UX

'-'*-- *---^' <

2 1

11. Auxillary Feedwater Flow Rate l Vsteam generator 1/ steam generator i 12. R: actor Coola' nt System Subcooling Margin Monitor P~~tA(M" 2 Rm i

.__.. . ... M-1

13. L -_;

o p 3!

ti.

2henes, lheau.

~*n :o n\

C*MnMMHYA4.dB

. _ .. m ..... . .. . .. ?dMd

. ...< AMrbrAurEntsAar 2rt,.= - f$')'

th.a E $\

, (s. h,M.WMJMKYHW?C'AN //Avrrn4 -3 o l

.a ,._..-

1A.a

, 1. Q ,

- -- == - t a ~=:: = .x $ '

g rt

'e- $

O TABLE 3.3-10 (Continued)

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ACCIDENT MONITDRING INSTRUMENTATION

  • p, m.

! n 4 TOTAL gc MINIfRM c NO. OF "INSTRt#ENT CHANNELS rm 3

CHAIMELS OPERABLE

i. u ;

T/I.

t. e Containment Water Level (Wide Range) CG 2 1 1 . E -

'7[. -inCorebsr-roocouples ...,.____an,.

4/ core quadrant 2/ core quadrant c

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n. .

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Off". Contakseent -

High Range Radiation Monitor M.A. 1

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r" x f [ Reactor Vessel Water Level 2 1 ' i e

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23. Reactor Coolant Radiation Level Monitor M.A. 1 r
< r

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5

! (

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r n1 5

1 bC I 0- ! -

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! M b ,h.

g f g h g- a

  • k f!1 8  !

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a :1 l i A

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1 TABLE 4.3-7 i f

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ACCIDENT MINITORING INSTRtBENTATION SURVEILLANCE REQUIREDENTS '

f CHANNEL CHApeIEL N5

! Ilt5TRt#ENT ',

_ CHECK CALIBRATION ^) '

1. o <

Contalment Pressure .

c. Momal Range m' en
b. Extended Range M R

! M R l

9-l2. Reacter Coolant Outlet Temperature - T I"Id* I'"9') n> l HDT M "

l i

3. ' React 6r~coolanfiklel'Tembridare - Tg (Wfde Range) M o5 R

4.

R: actor Coolant Pressure - Wide Range M 1 R

  • r
5. Pressurizer Water Level M, R N
6. m Steam Line Pressure C M R -

n i 7.. Steam Generater Water Level - Narrow Range m M R

  • j 8. Steam Generater Water Level - Wide Range n, M oi R m j 9.. R2 fueling Water- Storage Tank Water Level j

M R

$ .h*T(*".=VL.*'E_.,__,__, I

10 3 ---_._.. .

1 M A l 11. Auxfllary. Feeerster. Flow Rate M R i

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12. .R: actor, Coolant,5ysnem SubcoolIng Margin Moniter M ENN1,_- -

R y ,

y "

\ g3; c ' .am,amaXe'G. R.wer m few.ar Zwn g .

h!

14 .=: =1 m "M" f $@

- g -- % :r =

WtisN W M M R kh

= oi e

g$, --.- . n y A, -

y 3 :. n 2 :._.: e = = # F : = - - ;:: y(jn e = ) '

0

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. i TABLE 4 3-7 (Continued) i, s

ACCIDENT IWNITORING INSTRLSENTATION SLitVEILLANCE RE7JIRDENT  !'

,, IllSTRt#ENT CHANNEL CHANNEL CHECK ui CALIBRATION "'

' i /(g. Containment Water Level (Wide Range)  !

M R

Q [ i n Core euples E' M R

. .a set ,,,; _ ;;;g, _ ;; et? ,n , ,;,,, S,.

n  ;

  • a . '^n L;;.T 0;;J. ~~ 2.=-;;.;.; ;-,i wr {'

a 0

{ (8K. Containment ".jexn  :;..ne. - High Range Radiation Monitor '

N Ra c!, '

fj[. Reacter Vessel Water Level ,

i M R 4 23. Reactor Coolant Radiation Level Monitor f'_

M R

\ E, '

.! 5 n

O

'OWelEt CALIBRATION may consist of an electronic calibration of the channel, not including the1 d j

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' installed er portable gamma source.fo: range decades above 10 R/h and a one point calibration check .

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- ADDITIONAL REVIEW REQUIRED f

Item : Technical Specification 3.3.3.7 Chlorine Detection Systems Provide Justification to Delete Specification 3.3.3.7.

NNECO's Response:

1. The minimum channels operable requirement for the chlorine monitors will be covered in specification 3.3.2 Table 3.3- 3. The setpoints for the 1 chlorine monitor will be covered in specification 3.3.2 Table 3.3-4. l
2. The action statements for specification 3.3.3.7 will be included in the notes associated with Table 3.3-3.
3. The surveillance requirements for specification 4.3.3.7 will be included 4 under the surveillance requirements of specification 4.3.2.1, Table 4.3-2.

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ADDITIONAL REVIEW REQUIRED Item : T/S 3.4.1.5 and 3.4.1.6, Reactor Coolant System, Isolated Loop.

These specifications need to be revised to include boron concentration; revised action station / surveillance requirements.

NNECO's Response:

See the attached copy of the proposed specification. The modified specification 3/4.4.1.5 is requested to provide adequate assurance that the loop stop valve are shut and deenergized during operation in Modes 1 through 4. This will ensure that a positive reactivity addition will not occur due to an inadvertent opening of a loop stop valve. The modified specification 3/4.4.1.6 is requested to ensure conformance with the analysis assumptions for returning an isolated loop to service.

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SEP 19 '85 15:05 0 T E !. SEPVICES COPP. PAGE.05 NF & REVIEW COPY -

REACTOR COOLANT SYSTEM ' '

ISOLATED LOOP LIMITING CONDITION FOR OPERATION .

3. 1.5 grea The boron concentration of an isolated loop shall be maintained r than or equal to the baron concentration of the operating loops.

APPLICA8 TY: M00ES 1, 2, 3, 4, and 5.

i' ACTION:

WIth the requirene the isolated loop's of the above specification not satisfied, do not open the isolated loop to w valves; either increase the boron concentration of -

l-i STAN08Y within the next in the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least NOT to a $HUTDOWN MARGIN equiva urs with the unisolated portion of the RCS borated

nt to at least 25 Ak/k at 200*F.

=A-

$URVEILt.ANCE REQUIRFMENTS aw <

4.4.1.5 The boron concentration of an isolated loop s greater then or equal to the boron concentration of the 1 be determined to be least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and within 30 minutes prior to openrating loops either theathot I leg or cold leg stop valves of an isolated loop.

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SEP 19 '85 15:05 0ATEL SEPUICES CORP.

PAGE.07 I

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sEP is '85 15:06 QATEL SEPUICES COPP. PAGE.08 PROOF & REVEN COPY REACTOR COOL,.T SYSTEM

^

I M I.6 1985 JSOLATED Lo0P STARTUP LJMITING CON 0! TION FOR OPERATION

3. 4.1. 6 A reactor coolant loop shall remain isolated until:

)

a.

The e'-' " M '_ -_s" - ; r n' . , . . . . .- . . - ; . . : _

. _ _ ~

n_ . _ .

1- _ : - 1 temperature at the cold leg of the isolated loop is within 20*F of '

theTM$d highest W cold 1 g temperature of the operating loops, and I cg. The reactor is subcritical by at least & Ak/k. '

)

APPLICA8ILITY: Att M00Esf f d ~

ACTION:

M With

.-a- the >--,.requirements w ,.._ of the above specification not satisfied, < _ . _ _

smA.

_r g g , 2 ) J ,- l b,  % pee.rhte e5 3MdfkE I* *' N *** ** 499  % a

- - -SURVE!LLANCE REQUIREMENTS

4. 4.1. 6.1 The isolated loop cold leg temperature shall be determined to be 30 minutes prior to opening the cold leg stop valve.within 20*

e 4.J.1. 6. 2

/ Ak/k within 30 minutes prior to opening the cold leg stop valv

/.I i

THIS PAGE OPEN PENolNG RECEIPT OF lNPORMATION FROM THE APPUCANT SW 8 i>

)

Insert A

b. The boron concentration of the isolated loop is greater than or equal to the boron concentration of the operating loops, and Insert B 4.4.t.6.3 The boron concentration of an isolated loop should be determined to be greater than or equal to the boron concentration of the operating loops within 30 minutes prior to opening either the hot leg or cold Icg stop valves of an isolated loop.

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ADDITIONAL REVIEW REQUIRED ltem : Technical Specification 3.5.3, ECCS Subsystem - Tavg Less than 3500F.

NNECO's Response:

Action A of specification 3.5.3 should read as follows:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, or the containment recirculation pump, or the containment recirculation heat exchanger, or the inoperability of the flow path capable of taking suction from the containment sump, restore the inoperable ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

ADDITIONAL REVIEW REQUIRED I e;m, Technical Specification 3-9.6, Refueling Machine.

Provide justification for the values in specifiction 3.9.6 -

REFUELING MACHINE CAPABILITY.

NNECO's Response:

The refueling machine hoist used for fuel measurement has a minimum design capacity of 6000 pounds. The hoist gripper mast has a maximum assumed weight of 1200 pounds. The actual weight of the mast in air is less than !!00 pounds.

The analysis assumes that the maximum weight on the gripper is 2000 pounds (Fuel Assembly & Control Rod & Overload). Therefore, the minimum design capacity of the hoist / griper is 3200 pounds (mast weight & maximum assumed load). An overload of less than or ec,ual to 3200 pounds is required. The refueling machine vendor, Westinghouse, recommend a load test to 125% of the design capacity (4000 pounds) prior to a refueling.

The auxiliary hoist on the refueling machine has a capacity of 3000 pounds.

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i SEP 20 '95 10:30 0,ATEL SERU!CES CORP. PAGE.04 3/4.9.6 REFUELING MACHINC OPERASILITY LIMITING CONDITION FOR OPERATION 3.9.6 --

(

rods or fuel assemb'its and shall be operable with:The ref c a.

' The refueling machine used for movement of fuel assemblies

1. .

i A minimum design rated load of 3200 pounds on the hoist.

1 2.

, An overload cutoff limit less than or equal to 3200 pounds.

b.

plug handling operations having:The auxiliary hoist mused e 1.

A minimum capacity of 3000 pounds, and 2.

A 1000 loads pound for these load indicator operations. which shall be used to monitor APPLICABILITY:

reactor pressure vessel.During movement of drive rods or fuel..nin assemblies the ACTION:

l - '

i With not the reqdirements satisfied for the refueling machine and/or auxiliary hoist from ope. suspend use of any inoperable refueli within the reactor pressure vessel.

epplicable. The provisions of Specification 3.0.3 are not 1

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-_ SURVEILLANCE RE00!REMENTS 4.9,6.1 Each refueling machine used for movement of fuel assemblies reactor press 0rs vessel shall be demonstrated OPERABLE within 10 start of such operations by performing a load test of at least 1255 of design rated load and by demonstrating an automatic loadueling cutoff wh

' machine load gaceeds the setpoints of Specification 3.9.6.a.2.

4.9.6.2 ,

tach aux'iliary hoist and associated load indicator used for mov drive rods within the reactor pressure vessel shall be demonstra 1001250 least hourspounds. prior to the start of such operations by performing n a lo

. -...__m ._ ._-._____.___._,,_.m.,.__- .._ _ ,_ . , _ ~ . . . . _ _ . _ _ _ _ , - _ . _ _

ADDITIONAL REVIEW REQURIED item: Technical Specification 4.9.8.1/4.9.8.2, RHR and Coolant Circulation Provide Millstone 3 specification numbers for circulating RCS flow and bases for the above sur seillance requirements.

NNECO's Response:

'y; The correct value is 2800 gpm for both surveillance requirements. This value, based on engineering Judgement, is sufficient to erisure adequate cooling capacity is available to remove decay heat in MODE 6 and sufficient coolant circulation is maintained within the core to preclude boron stratification.

This value also appears in the Callaway Unit I and Byron Units 1 and 2 Technical Specifications.

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u-_______________._._____.__.___.

l ADDITIONAL REVIEW REQUIRED Item: Technical Specification: Bases Section 2.2.1 Bases Section 2.2.1 (Reactor Coolant Flow). NU to provide an updated value for P-8.

NNECO's Response:

The bases section should read in part as follows: "

...a power level of .

approximately 375.." This value of P-8 is consistent with the P-8 value in Table 1 2.2-1.

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es 4

ADDITIONAL REVIEW REQUIRED item 51: Table 6.2-1 should be modified to reflect the correct position title for the 2nd SRO and the two non-licensed operators as follows:

"SRO" changed to "SCO" "AO" changed to "PEO" "SRO '- Individual with a Senior Operating license on Unit 3" changed to "SCO - Supervising Control Operator with a Senior Operating license on Unit 3" "AO - Auxiliary Operator" changed to "PEO - Plant Equipment Operator a non-licensed operator"

. Table 6.2-1 "(Other than the Shif t Technical Advisor)" should be deleted.

NNECO's Response:

Millstone Unit No. 3 utilizes either the SS or SCO (Dual Role concept) as the STA ' qualified individual on shift. The above wording has the effect of prohibiting- the SS from leaving the control room when the STA Dual Role

. concept is utilized.

Item 51:

Table 6.2-1 (*) "as required by the NRC" changed to (*)"as required by Section

,- 6.2.3".

NNECO's Response:

Section 6.2.3 of Technical Specifications describe the qualifications for the STA.

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ADDITIONAL REVIEW REQUIRED Section 6.2.2.f should be modified to be consistent with Millstone Unit Nos. I and 2, Section 6.2.2.f specification:

" Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions.

These procedures should follow the general guidance of NRC Policy Statement on working hours (Generic Letter No. 32-12)."

NNECO's Response:

The modification is requested because all three Millstone Units operate under common administrative controls described in Section 6.0 of Technical Specifications.

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ADDITIONAL REVIEW REQUIRED ltem 55:

1. Section 6.3.1 modified to reflect the FSAR commitment to Regulatory Guide 1.8 us follows:

1

" Regulatory Guide 1.8, September 1975" changed to Regulatory Guide l 1.8, May 1977" NNECO's Response: (6.3.1 and 6.4.1)

FSAR Section 13.1.3 " Qualifications of Nuclear Plant Personnel" define the education and experience requirements for licensed Operators and Senior Operators to ANSI N18.1-1971, Section 4.5.1, as referenced by Regulatory Guide 1.8, 1977.

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ADDITIONAL REVIEW REQUIRED Item : Sections 6.5.3.9.b, 6.5.3.9.c and 6.5.4.7.e to correct typographical errors.

NNECO's Response: ,

I 6.5.3.9.b j

" Specification 6.5.2.6" changed to " Specification 6.5.3.6"

'6.5.3.9.c

" Specification 6.5.2.7" changed to " Specification 6.5.3.7" 6.5.4.7.c "24 months." changed to "24 months.*"

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ADDITIONAL REVIEW REQUIRED ltem 61 - Sections 6.8.4d should be deleted.

NNECO's Response:

6.8.4d FSAR Section 4.4.6.5 " Instrumentation for Detection of inadequate Core i Cooling" describes the ICC Monitoring System as a Category 1 (Class IE) system

- with redundant trains (Train A and Train B). Section 6.8.4d is only required for Units with a single channel of monitoring instrumentation. l r

o. , . . .. . . .. .. . .

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l L ADDITIONAL REVIEW REQUIRED item Section 6.10.3.i should be modified to reflect proper document title.

NNECO's Response:

^

" Operational Quality Assurance Manual" changed to

" Quality Assurance Topical Report"

ADDITIONAL REVIEW REQUIRED Item 42: Table 3.6-2, Containment Isolation Valves NNECO's Response:

1 Table 3.6-2 is attached.

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n TABLE 3.6-2 CONTAINMENT ISOLATION VALVES MAXIMUM VALVE NUMBER FUNCTION ISOLATION TIME (Seconds)

1. PHASE A ISOLATION 3SSR-CTV26 REACTOR COOLANT HOT LEG SAMPLE (INSIDE) 60 3SSR-CTV27 REACTOR COOLANT HOT LEG SAMPLE (OUTSIDE) 60 3SSR-CTV22 PZR LIQUID SAMPLE (INSIDE) 60 3SSR-CTV23 PZR LIQUID SAMPLE (OUTSIDE) 60 3SSR-CTV20 PZR VAPOR SPACE SAMPLE (INSIDE) 60 3SSR-CTV21 PZR VAPOR SPACE SAMPLE (OUTSIDE) 60 3SSR-CV8026 PRT GAS SAMPLE (INSIDE) 60 3SSR-CV8025 PRT GAS SAMPLE (OUTSIDE) 60 3SSR-CTV29 REACTOR COOLANT COLD LEG SAMPLE (INSIDE) 60 3SSR-CTV30 REACTOR COOLANT COLD LEG SAMPLE (OUTSIDE) 60 3SSR-CTV 32 S.I. ACCUMULATOR SAMPLE (INSIDE) 60 3SSR-CTV33 S.I. ACCUMULATOR SAMPLE (OUTSIDE) 60 3SIL-CV8968 NITROGEN TO S.I. ACCUMULATORS (INSIDE) 60 3SIL-CV8880 NITROGEN TO S.I. ACCUMULATORS (OUTSIDE) 60 ,

l 3PGS-CV3046 PRIMARY GRADE WATER TO PRT (INSIDE) 60 3PGS-CV8028 PRIMARY GRADE WATER TO PRT (OUTSIDE) 60 3CHS-MV8112 SEAL WATER RETURN FROM RCP's (INSIDE) 60 3CHS-MV8100 SEAL WATER RETURN FROM RCP's (OUTSIDE) 60 3CHS-CV8160 REACTOR COOLANT LETDOWN (INSIDE) 10 3CHS-CV8152 REACTOR COOLANT LETDOWN (OUTSIDE) 10

TABLE 3.6-2 (Continued)

CONTAINMENT ISOLATION VALVES q MAXIMUM '

VALVE NUMBER FUNCTION ISOLATION TIME (Seconds) l j

3DGS-CTV24 PRT AND CTMT DRAINS TRANSFER PUMPS DISCHARGE (INSIDE) 60 3DGS-CTV25 PRT AND CTMT DRAINS TRANSFER PUMPS DISCHARGE (OUTSIDE) 60 CTMT DRAINS SUMP PUMP DISCHARGE (INSIDE) 60 3DAS-CTV24 CTMT DRAINS SUMP PUMP DISCHARGE (OUTSIDE) 60 3DAS-CTV25 3VRS-CTV20 PRT AND CTMT DRAINS TRANSFER TANK VENT (INSIDE) 60 PRT AND CTMT DRAINS TRANSFER TANK VENT (OUTSIDE) 60 3VRS-CTV21 3CVS-CTV20A CTMT VACUUM PUMP SUCTION (OUTSIDE) 60 CTMT VACUUM PUMP SUCTION (OUTSIDE) 60 3CVS-CTV20B 3CVS-CTV21A CTMT VACUUM PUMP SUCTION (OUTSIDE) 60 CTMT VACUUM PUMP SUCTION (OUTSIDE) 60 3CVS-CTV21B 3C DS-CTV91 A CHILL WATER SUPPLY (INSIDE) 60 CHILL WATER SUPPLY (INSIDE) 60 3CDS-CTV91B 3CDS-CTV38A CHILL WATER SUPPLY (OUTSIDE) 60 CHILL WATER SUPPLY (OUTSIDE) 60 3CDS-CTV38B CHILL WATER RETURN (INSIDE) 60 3CDS-CTV40A CHILL WATER RETURN (INSIDE) 60 3CDS-CTV40B 3CDS-CTV39A CHILL WATER RETURN (OUTSIDE) 60 CHILL WATER RETURN (OUTSIDE) 60 3CDS-CTV39B INSTRUMENT AIR (INSIDE) 60 31AS-MOV72

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l TABLE 3.6-2 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM VALVE NUMBER FUNCTION ISOLATION TIME (Seconds) 3IAS-PV15 INSTRUMENT AIR (OUTSIDE) 60 3FPW-CTV49 FIRE PROTECTION (INSIDE) 60 3FPW-CTV48 FIRE PROTECTION (OUTSIDE) 60 3 CMS-MOV24 CTMT ATMOSPHERE MONITOR DISCHARGE (INSIDE) 60 3 CMS-CTV23 CTMT ATMOSPHERE MONITOR DISCHARGE (OUTSIDE) 60 3 CMS-CTV20 CTMT ATMOSPHERE MONITOR SUCTION (OUTSIDE) 60 3 CMS-CTV21 CTMT ATMOSPHERE MONITOR SUCTION (OUTSIDE) 60 3SIH-CV8871 S.I. TEST AND ACCUMULATOR FILL (INSIDE) 60 3SIH-CV8964 S.I. TEST AND ACCUMULATOR FILL (OUTSIDE) 60 3SIH-CV8888 S.I. TEST AND ACCUMULATOR FILL (OUTSIDE) 60 3GSN-CTV105 NITROGEN SUPPLY HEADER (INSIDE) 60 3GSN-CV8033 NITROGEN SUPPLY HEADER (OUTSIDE) 60 3 SSP-CTV7 POST ACCIDENT SAMPLE (INSIDE) 60 3 SSP-CTV3 POST ACCIDENT SAMPLE RETURN (INSIDE) 60 3SIH-CV884'3 HIGH PRESSURE BORON INJECTION TO TEST HEADER 60 (INSIDE) 3SIL-CV8890A RHR COLD LEG IN3ECTION TO TEST HEADER (INSIDE) 60 3SIL-CV8890B RHR COLD LEG IN3ECTION TO TEST HEADER (INSIDE) 60 3SIL-CV8825 RHR HOT LEG IN3ECTION TO TEST HEADER 60 3SlH-CV8881 S.I. PUMP HOT LEG INJECTION TO TEST HEADER 60 3SlH-CV8824 S.I. PUMP HOT LEG INJECTION TO TEST HEADER 60 4-

1 TABLE 3.6-2 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM l VALVE NUMBER FUNCTION ISOLATION TIME (Seconds) 3SlH-CV8823 S.I. PUMP COLD LEG INJECTION TO TEST HEADER 60 j i

2. PHASE B ISOLATION 3CCP-MOV45A RPCCW CTMT SUPPLY (OUTSIDE) 60 3CCP-MOV45B RPCCW CTMT SUPPLY (OUTSIDE) 60 3CCP-MOV48A RPCCW CTMT RETURN (INSIDE) 60 3CCP-MOV48B RPCCW CTMT RETURN (INSIDE) 60 3CCP-MOV49A RPCCW CTMT RETURN (OUTSIDE) 60 3CCP-MOV49B RPCCW CTMT RETURN (OUTSIDE) 60 l 3. POWER OPERATED VALVES 3CHS-MV8105 REACTOR COOLANT CHARGING 10 3CVS-MOV25 CTMT VACUUM PUMP DISCHARGE (INSIDE) NA 3RSS-MOV20A CTMT RECIRCULATION PUMP SUCTION (OUTSIDE) 60 3RSS-MOV20B CTMT RECIRCULATION PUMP SUCTION (OUTSIDE) 60 3RSS-MOV20C CTMT RECIRCULATION PUMP SUCTION (OUTSIDE) 60 3RSS-MOV20D CTMT RECIRCULATION PUMP SUCTION (OUTSIDE) 60 3RSS-MOV23A CTMT RECIRCULATION SPRAY HEADER SUPPLY (OUTSIDE) 60 3RSS-MOV23B CTMT RECIRCULATION SPRAY HEADER SUPPLY (OUTSIDE) 60 3RSS-MOV23C CTMT RECIRCULATION SPRAY HEADER SUPPLY (OUTSIDE) 60 3RSS-MOV23D CTMT RECIRCULATION SPRAY HEADER SUPPLY (OUTSIDE) 60

TABLE 3.6-2 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM VALVE NUMBER FUNCTION ISOLATION TIME (Seconds) 3QSS-MOV34A QUENCH SPRAY HEADER SUPPLY (OUTSIDE) 30 3QSS-MOV343 QUENCH SPRAY HEADER SUPPLY (OUTSIDE) 30

4. MANUAL VALVES

TABLE 3.6-2 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM VALVE NUMBER FUNCTION ISOLATION TIME (Seconds) 3CHS-V467 REACTOR COOLANT PUMP SEAL WATER SUPPLY (INSIDE) NA 3CHS-V501 REACTOR COOLANT PUMP SEAL WATER SUPPLY (INSIDE) NA 3CCP-V886 DEMINERALIZED WATER CTMT SUPPLY (INSIDE) NA 3CCP-V887 DEMINERALIZED WATER CTMT SUPPLY (OUTSIDE) NA 3CCP-V18 COMPONENT COOLING WATER CTMT SUPPLY (INSIDE) NA 3CCP-V60 COMPONENT COOLING WATER CTMT SUPPLY (INSIDE) NA 3CVS-V13 CONTAINMENT VACUUM PUMP DISCHARGE (OUTSIDE) NA 3SGF-V29 STEAM GENERATOR CHEMICAL FEED SUPPLY (INSIDE) NA 3SGF-V31 STEAM GENERATOR CHEMICAL FEED SUPPLY (INSIDE) NA 3SGF-V33 STEAM GENERATOR CHEMICAL FEED SUPPLY (INSIDE) NA 3SGF-V35 STEAM GENERATOR CHEMICAL FEED SUPPLY (INSIDE) NA l

  • MAY BE OPENED ON AN INTERMITTENT BASIS UNDER ADMINISTRATIVE CONTROL.

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