Text

Proposed Tech Specs Modifying List of Valves That Can Be Opened in Modes 1 Through 4,removing Footnote on Type a Testing & Rewording TSs & Bases to Provide Clarity & Consistency
	ML20211M908
Person / Time
Site:	Millstone
Issue date:	10/07/1997
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20211M901	List: B16758, Application for Amend to License NPF-49,modifying List of Valves That Can Be Opened in Modes 1 Through 4,removing Footnote on Type a Testing & Rewording TS & Bases to Provide Clarity & Consistency.Withdrawal Requested of 970630 LAR; ML20211M908;
References
	NUDOCS 9710150175
	Download: ML20211M908 (30)
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[ Docket No. 50-423 i i

j B16758 i

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i Attachment 2 9

Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification j Containment Systems i (PTSCR 3-34-97)

Marked Up Paoes i

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October 1997

- 9710150175 971007 PDF ADOCK 05000423 l P PDR

U.S. Nucirr R:gulatory Commission B16758%ttachment 2\Page1 MARKUP OF PROPOSED REVISION Refer to the attached markup of the proposed revision to the Technical Specificat'ons.

The attached markup reflects the currorcJy issued version of the Technical Specifications listed below. Pending Technical Specification revisions or Technical Specification revisions istued subsequent to this submittal are not reflected in the enclosed markup.

Proposed revisions to Technical Specification letters B15028, dated December 14, 1994, and B15193, dated April 28,1995, are not reflected in the enclosed markup.

The following Technical Specifications changes are included in the attached markup:

. The surveillance administrative control footnote ""' is modified to change the valve list and descriptions.

4.6.1.1.a 1

. The surveillance wording is changed from 'not less than Pa. 53.27 psia (38.57 psig)" to "a pressure greater than or equal to Pa. 38.57 psig".

4.6.1.1.c 4.6.1.2.a 4.6.1.3.b e The wording is changed from 'Pa, 53.27 psia (38.57 psig)' to *a pressure greater than or equal to Pa,38.57 psig'.

4.6.1.2.d

. The wording is changed from 'Pa, 53.27 psia (38.57 psig)" to " greater than or equal I to Pa. 38.57 psig".

4.6.1.2.e e The wording is changed from *Pa,53.27 psla (38.57 psig)" to *Pa, 38.57 psig'.

3.6.1.2.a 3.6.1.3.b 4.6.1.3.a e The wording is changed to provide clarity.

B3/4.6.1.1 B3/4.6.1.2 B3/4.6.1.3 e The footnote is deleted, 4.6.1.2.a

. . --- ~ - . . - . - -

3/A,6 CONTsINWENT SYSTEWS Januar/ 23, 199*$

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s V k^k' jj.I 3/t.6.1 ocinsnv CCNTATHMENT

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ii f0NTAINHENT INTEGRTTY 11Mf7f NG CONDITION FOR OPERATION 3 ., .

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$ 3.6.1.1 PrimaryCONTAINMENTINTEGRITYshallbemaintained.

APPticABilfTY: MODES 1, 2, 3, and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINHE 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STAN08Y within the next SHUTDOWN,within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .-

l L SURVEf ttsNCE REOUTREMEVTS - -

4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demenstrated:

a.

At least'once per 31 days 'b[ verifying that all penetrations' not capable of being closed by OPERA 3LE containment autcmatic isolation valves or operater action during perieds when containment isolation

([L ~ -

vaivas closedare cpened during underesnditions accident administrative contro1,** and recuired to be g,, ! are closed by valves, blind flanges, or deactivated automatic. valves secured in their posittens.

I J I. c'--': b.

&Of :r 7 By verifying that each containment air lock is in compliance with the requirnments of Specification 3.6.1. ' d

.g

~~ Q t.d).o), c. .

Qrmter de er e t After excepteach closing of each the centainment air locks penetration nblictRVlypi%qp*lesting.

3 Ti d ,i if opened following a Type A or 8 pR tast, by leak rate testing the se,al with gas at a pressure --t

%ee P "' " --' 138.57 psigT and verifying that when ha E,si p measurth leakagt rate for these seals is added to the leakage rate the

]

j> datamined pursuant to Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L

g ., +

a 5 i _:

C $ '5 -! located inside the containment' and areExcept valves. -blind flang I

secured in the closed position. locked, sealed, ior otherwise These penetrations shall be verified closed during each COLD SHVTDOWN except that such verification ne'ed not

be pertened acre often than once'per 92~ days.

! ** . -M9WALJ VALVE S i The following eam administrative control. *4. valve may be opened on an intermittent basis under i

3HCS6V2, 3HC3W3, OTTL7:0;, -:TPS;;;, 335PAV13, 353P/W14, 3HC5W9, 3HC5W10, 3NCSW6,

t.: V:0, 3tH57tV371, 3HCSW13, :OVOS.

N@" C ::P-Y:00, 200?-Y 07, : VLV;L 3mssw V894 3m55'O i

3ms6* V667. l'Erncrre nwJuAL vA(.ves 3R.asx t+N 8701 A;3RASM i

mv87ot Bj 5R45* mv 9702A 6Rhtst 7 rnV6?CY'B -

1 L _ _

Sju G-l 4

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3 ~ /S ~ 9 7 CONTAINMENT SYSTEMS CONTAINMENT LEAKAnt February 5,1996 l LIMITING CONDITION FCR OPERATION

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3 l 3.6.1.2 containment leakage rates shall be limited to:

a. An overall integrated leakage rate of less than or equal to L.,

0.3% by weight of the co tainment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P.,

l 00.07;;tp8.57psig, l b. A combined leakage rate of less than 0.60 L for all penetrations

and valves subject to Type B and C tests, wban pressurized to P.;

i and l

l c. A cambined leakage rate of itss than or equal to 0.042 L, for all i

penetrations that are Secondary Containment bypass leakage paths when pressurized to P. l l APPLICABIL11(: MODES 1, 2, 3, and 4.

j EI1Dfi:

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With the measured overall integrated containment leakage rato axceeding 0.75

! L., er the measured combined leakage rate for all penetrations and valves subject to Type B and C tests exceeding 0.60 L., or the combined bypass i leakage rate exceeding 0.042 L., restore the overall integrated leakage rate

to less than 0.75 L., the combined leakage rate for all penetrations subject I

! to Type B and C tests to less than 0.60 L., and the combined bypass leakage j rate to less than 0.042 L, prior to increasing the Reactor Coolant System

, temperature above 200'F.

SURVEILLANCE REQUIREMENTS i

4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria

! specified in Appendix J of 10 CFR Part 50 using methods and provisicns of ANSI

! N45.4-1972 (Total Time Method) and/or ANSI /ANS 56.8-1981 (Mass Point Method):

! a. Three Type A tests (Overall Integrated Containment Leakage Rate) i shall be conducted at ap)roximately equal intervals during shutdown i - ~ % at a pressure - '

n g{Q'th10-year servTS:t 1:::h: P., 50.07 ;;t:g38.57 psigW curing each period J

b. If any periodic Type A test fails to meet 0.75 L., the test schedule i

for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L., a

Type A test shall be performed at least every 18 months until two

! consecutive Type A tests meet 0.75 L, at which time the above test 4

scheault may be resumed; i

1 *Me th+rd Typ: A test will h andat;d durir.g th; :ixth r:falia; =t:;:. A:

i --e-etiult, th: d:r ti= cf th: fir:t 10 ynr nrvia p;ri:d will M ::t=d:d i:

th: ed Of th: 9 th - 9 tP r; r t ;:_

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l 3/4 6-2 Amendment No. pp, 77, pp, JJJ,126

_ . _ _ _ NILL 5T W E - WIT 3_ - . _

3 3-/r- 7 7

January 3,1995 j CONTAllMENT SYSTDt3 I SURVEILLANCE REQU!REMENTS (Continued) i

c. The accuracy of each Type A test shall be verified by a supplemental

> test which:

1) Confins the accuracy of the test by verifying that the supple-1 mental test results, L., minus the sum of the Type A and the superisposed leak, L., is equal to or less than 0.25 L,1 4 2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and .

3) Requires that the rate at which gas is injected into the containment or bled from the containment during the supplemental test is betwee and 1.25 L. -_

j ressuie~ ce446-% er

d. Type 8 and C tests shall be con ps atvP., "... ;;i:

p(38.57 psigM at intervals no greater than 24 months except for l tests involving:

1) Air locks

e. The combined bypass leakage rate shall be determined to be less than or equal to 0.042 L, by applicable Type B and C tests at least once per 24 months except for penetrations which are not individually l testable penetrations not individually testable shall be detemined to have no detectable leakage when tested with soap bubbles while the containment is pressurized gto P., !!.07 ;:tQ38.57 psig)?

during each Type A test; r r 6 oT W

f. Air locks shall be tested and amonsratN E by the requirements of Specification 4.6.1.3; l

g. Purge supply and exhaust isolation valves shall be demonstrated OPERABLE by the requirements of Specifications 4.6.3.2.c and 4.g.g.

h. The provisions of Specification 4.0.2 are not applicable.

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l 3/4 6-3 Amendment No. pp. 77100 MIL.LSTONE - Lat!T 3

3-/C-97 8

Corrected: July 12,1995 Original: January 25,1991 CONTAINMENT FftTDt1 4

i CorrAINMENT Aft LOCK 5 -

LIMITINS CONDITION FOR OPERATION 3.5.1.3 The containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for nonna) transit entry and exit through the containment, then at least one air lock door shall be closed, and "

b. An overall air lock leakage rate of less than or equal to 0.05 L at P.,50.07;i/3R.57psig)f" APPLICARfL17Y: MODES 1, 2, 3, and 4.

i EIElti: *

n. With one containment air lock door inoperable 1.

3 Maintain at least the OPERA 3LE air lock door closed

and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed,

2. Operation may then contirve until performance of the next required overall air lock leakage test provided . that . the OPERABLE once air lock door is verified to be locked closed at least per 31 days,

' 3. Othervise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , and

4. Entry into an DPERATIONAL WDE is permitted while subject to these ACTI.0N requirements,

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1

Except during entry to repair an inoperable inner door, for a cumulative time not to exceed I hour per year.

MI esL.LSTONE - 1MIT 3 3/46-5 Amendment No p. 59

3~/r- 9 7 Ja wary 25, 1991 CONTAINMENT SYSTDis

$URVffLLANCE RFOUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated CPERABLE:

a. 1) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying no detectable seal leakage by pressure decay when the volume between the door seals is pressurized to greater than or equal to P,, "!.!? ;m p(38.57 psigW for at least 15 minutest or

2) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air -

lock is being used for multiple entries, then at least once par 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying that the seal leakage is less than 0.01 L as determined by precision flow measurements when measured f$r at least constant 30 seconds prqsure of greaterwith thanthe or volume equal tobetween the scsis at a g18.57ps<gf; P,, ' .2? ;&

or

3) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being used for multiple entries, then at least once per 72 h,ours, by completing an overall air lock leaka e test per 4.6.1.3.b.

a, pee,ss(Ga]reatev- ,;

b. By . conducting overall r lock leakage tests at :t .r: "r P '

$3.*7 ;-A+A38.57 psig and ve;ifying the overall air lock leaka$e, ~

rate is within its lie t:

1) At least once per 6 months.* and

2) Prior to establishing CONTAINMENT INTEGRITY when maintenance has been perfonned on the air lock that could affect the air lock sealing capability **

c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

The provisions of Specification 4.0.2 are not applicable.

- Thisre>resentsanexemptiontoAppendixJ,paragraphIII.D.2.(b;,ii),of 10 CFR Jart 50.

MILLSTONE - UNIT 3 3/4 6 6 hht so. 59

3-lN7 May 8,1995 3/4.s c0NTAINMENT SYSTEMS BASES 3/4.6.1 PRIMRY CONTAINMENT 3/4.s.1.1 t0NTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restric.

tion, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guidelines of 10 CFR Part 100

. during accident conditions and the control room operators dose to within the guide ines of GDC 19.

A nm pare 3raph on ddhmn oP x/m A omneh. . .. .5MM c-3/4.6.1.2 CONTAINMENT LEAKAGE

~

The limitations on containment leakage rates ensure that the total

. containment leakage volume will not exceed the value assumed in the safety

) analyses at the peak accident pressure, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than +o-l :;d t 0.75 L during performance of the periodic test to account l for possible degradation of the containment leakage barriers between leakage tests, gg 7 g The surveillance. testing for measuring leakage rates are consistent with the recuirements of Appendix J of 10 CFR Part 50. A partial exemption has been grantec: from the requirements of 30CFR50 Appendix J,Section III.D.1(a). The exemption removes the requirement that the third Type A test for each 10 year period be conducted when the plant is shut down for the 10 year plant inservice inspection (Reference License Amendment No.1,,1,1.).

The enclosure building bypass leakage paths are listed in Operating Procedure 3273, ' Technical Requirements - Supplementary Technice.1 Specifica-tions.' The addition or deletion of the enclosure building bypass leakage

, paths shall be made in accordance with section 50.5g of 10CFR50 and approved t>y the Plant Operation Review Conrnittee. .

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment laak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock lettage tests, y INIC#7 6 0 3/4.6.1.4 and 3/4.6.1.5 AIR PRESSURE and AIR TENPERATURE The limitations on containment pressure and average air temperature ensure that:

design negativepressure (1) the containment of 8 psia, and structure is prevented (2) the containment peakfrom exceeding pressure does its h!LLSTONE - UNIT 3 3 3/4 6-1 Amendment No. 77,77,111 ease s

u

Bases Document Change 3/4.6.1.2 CONTAINMENT LEAKAGE INSERT A The Limiting Conditionfor Operation defines the limitations on containment leakage ratesfor compliance with 10CFR$0, Appendit J. The leakase rates are ver@cd by sun cillance testing in accordance with the requirements ofAppendit J. Although the LCO spec @es the leakage rates at accidentpressure, P,. It is notfeasible toperform a test at such an exact valueforpressure.

Consequently, the sun elliance testing is performed at a pressure greater than or equal to P, to -

accountfor test instrument uncertainties and stabilisation changes. This conservative testpressure ensures that the measured leakage rates are representative ofthose which uvuld occur at accident pressure while meeting the interr ofthe LCO. This test methodology is consistent with the guidanceprovided in ANSUANS 36.81931for meeting the requirements setforth in Appendit J.

3/4.6.1.3 CONTAINMENT AIR LOCKS INSERT B ll7 tile the leakage rate limitation is spec $cd at accident pressure, P,. the actual sur veillance testing is performed by applying a pressure greater than or equal to P,. This higherpressure accountsfor test instrument uncertainties and test volume stabili:ation changes which occurs under actual test conditions. This method ofperforming surveillance te.iting is consistent with the guidanceprovided in ANSI $6.81981 and ensures that the leakage rate measured meets the intent ofthe LCO and Appendit J.

Bases Document Change 3/4.6.1.1 CONTAINMENT INTEGRlW INSERT C The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrctive control includes the following considerations: (1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing this operatar to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

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Docket No. 50323 11.11 151 1

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Attachment 3 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Containment Systems (PTSCR 3-34-97)

Retvoed Paaes October 1997 l;

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,v,,.,.. . - . . , , . . . . _ , , , .. ,- - _ . . _ , _ - - - . - . _ . , _ _ - _ . ~ ._

U.S. Nuclect Regulatory Commission .

B16758%ttachment 3\Page1 RETYPE OF PROPOSED REVISION RU;- *o the attached retype of the proposed revision to the Technical Specifications.

I t

The Ntached retype reflects the currently issued version of the Technical Specifications. Pending Technica? Specification revisions or Technical Specification reilslons, issued subsequent to this submittal are not reflected in the enclosed retype.

The enclosed retype should be checked for continuity with T3chnical Specifications prior to issuance.

{

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, 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABillTY: MODES 1, 2, 3, and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within E

I hour or be in at 10ist HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the fallowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS s 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations

not capable of being closed by OPERABLE containment automatic isolation valves or operator action during periods when containment isolation valves are opened under admir.lstrative control,** and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions,

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and

c. After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at a pressure greater than or equal to P , 38.57 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L.,

T ~ Except valves, blind flanges, and dcactivated automatic valves which are located inside the containment and are iocked, sealed, or otherwise secured in the closed position. These per. 3trations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

The following valves may be opened on an intermittent basis under administrative control. Manual valves 3 SSP *V13, 3 SSP *V14, 3HCS*V2, 3HCS*V3, 3HCS*V9, 3HCS*V10, 3HCS*V6, 3HCS*V13, 3CHS*V371, 3 MSS *V885, 3 MSS *V886, 3 MSS *V887. Remote manual valves 3RHS*MV8701 A, 3RHS*MV8701B, 3RHS*MV8702A, 3RHS*MV87028.

MILLSTONE - UNIT 3 3/4 6-1 Amendment No. JJ.

0540 l

C e CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LINITING CONDITION FOR OPERATION 1

3.6.1.2 Containment leakage rates shall be limited to:

a. Ar overall integrated leakage rate of less than or equal to L.,

0.3% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P., 38.57 psig; I

b. A combined leakar a rate of less than 0.60 L, for all penetrations and valves subject t . Type B and C tests, when pressurized to P,; and

c. A combined leakage rate of less than or equal to 0.042 L, for all penetrations that are Secondary Containment bypass leakage paths when pressurized to P,.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the measured overall integrated containment leakage rate exceeding C,75 L., or the n.easured combined leakage rate for all penetrstions and valves subject to Type B and C tests exceeding 0.60 L , or the combined bypass leakage rate exceeding 0.042 L., restore the overall integrated leakage rate

, to less than 0.75 L., the combined leakage rate for all penetrations subject l to Type B and C tests to less than 0.60 L., and the combined bypass leakage rate to less than 0.042 L, prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the fcllowing

test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using methods and provisions of ANSI N45.4-1972 (Total Time Method) and/or ANSI /ANS 56.8-1981 (Mass Point Method):

a. Three Type A tests (Overal' Integrated containment Leakaoe Rate) shall be conducted at approximately equal intervals during shutdown at a pressure greater than or equal to P., 38.57 psig, during each 10-year service period,

b. If any periodic lype A test fails to meet 0.75 L., the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L., a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L, at which time the above test schedule may be resumed; MILLSTONE - U'"T 3 3/4 6-2 Amendment No. 57, 0540

CONTAlttIENT SYSTEMS SURVEILLANCEREQUIREMENTS(Continued)

c. The accuracy of each Type A test shall be verified by a supplemental test which:

1) ' Confirms the accuracy of the test by verifying that the supple-mental test results, L , minus the sum of the Type A and the superimposed leak, L., Is equal to or less than 0.25 L ;

2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and

3) Requires that the rate at which gas is injected into the containment or bled from the containment during- the supplemental test is between 0.75 L, and 1.25 L.,

d. Type B and C tests shall be conduc+.ed with gas at a pressure greater than or equal to P -38.57 psig, at intervals no greater than 24 months except for tests I,nvolving:

1) Air locks

e. The commined bypass leakage rate shall be determined to be less than or equal to 0.042 L, by applicable Type B and C tests at least once per 24 months except for penetrations which are not individually l testable; panetrations not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the containment is pressurized to greater than or equal to P.,

38.57 psig, during each Type A test;

f. Air locks shall be tested and demonstrated OPERABLE by the requirments of Specification 4.6.1.3; i

g. I' urge supply and exhaust isolation valves shall be demonstrated '

OPERABLE by the requirements of Specifications 4.6.3.2.c and 4.9.9.

h. The provisions of Specification 4.0.2 are not applicable.

MILLSTONE UNIT 3 3/4 6-3 Amendment No. 57, /), Jpp, ono

________.______.a

CONTAINMENT SYSTEMS CONTAIMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 The containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to 0.05 L, at P., 38.57 psig. l APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed

and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed,

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door 13 verified to be locked closed at least once per 31 days,

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , and

4. Entry into an OPERATIONAL N00E is permitted while subject to these ACTION requirements,

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the ir.aperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Except during entry to repair an inoperable inner door, for a cumulative time not to exceed I hour per year.

MILLSTONE - UNIT 3 3/4 6-5 Amendment No. J7, JJ.

oue

0 CONTAINMENT SYSJD1) 3RVEILLANCEREQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. 1) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying no detertable seal leakage by pressure decay when the volume between the door seals is pressurized to greater than or equal to P,, 38.57 psig, for at least 15 minutes; l or

2) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying that the seal leakage is less than 0.01 L, as determined by precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of greater than or equal to P,, 38.57 psig; l l or

3) Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closii.g, except when the air lock is bein hours, gbyusod for multiple completing an entries, then overall air at least lock oncetest leakage perper 72 4.6.1.3.b.

b. By conducting overall air lock leakage tests at a pressure greater than or equal to P., 38.57 psig, and verifying the overall air lock leakage rate is within its limit:

1) At least once per 6 months,* and

2) Prior to establishing CONTAINMENT INTEGRITY sen nintenance has beer, performed on the air lock that could affect the air lock sealing capability.**

c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

The provisions of Specification 4.0.2 are not applicable.

- This represents an exemption to Appendix J, paragraph III.D.2.(b)(ii), of 10 CFR Part 50.

MILLSTONE - UNIT 3 3/4 6-6 Amendment No. JJ, ouo

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and assodated leak rates assumed in tha safety analyses. This restric-tion, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guidelines of 10 CFR Part 100 during accident conditions and the control room operators dose to within the guidelines of CDC 19.

The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuririg that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

The environmental conditions will not preclude access to close the containment isolation valves listed in the footnote and that valve closure will control the release of radioactivity outside the containment.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total

., containment leakage volume will not exceed th value assumed in the safety analyses at the peak act ident pressure, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than 0.75 L, l during performance of the periodic test to account for possible degradation of the cor.tainment leakage barriers between leakage tests.

The Limiting Condition for Operation defines the limitations on containment leakage rates for compliance with 10CFR50, Appendix J. The leakage rates are verified by surveillance testing in accordance with the requirements of Appendix J. Although the LC0 specifies the leakage rates at accident pressure, P., it is not feasible to perform a test at such an exact value for pressure.

Consequently, the surveillance testing is performed at a pressure greater than or equal to P, to account for test instrument uncertainties and stabilization changes. This conservative test pressure ensures that the measured leakage rates are representative of those which would occur at accident pressure while meeting the intent of the LCO. This test methodology is consistent with the guidance provided in ANSI /ANS 56.8-1981 for meeting the requirements set forth in Appendix J.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50. A partial exemption has been granted from the requirements c' 10CFR50, Appendix J, Section III.D.l(a). The exemption removes the requirement that the third Type A test for each 10-year period be conducted when the plant is shut down for the 10-year plant inservice inspection (Reference License Amendment No. 111).

. MILLSTONE - UNIT 3 B 3/4 6-1 Amendment No. 57, 77, JJJ, 0641 E

, 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1.2 CONTAINMENT LEAKAGE (continued)

The enclosure building bypass leakage paths are listed in Operating Procedure 3273, " Technical Requirements - Supplementary Technical Specifica-tions." The addition or deletion of the enclosure building bypass leakage paths shall be made in accordance with Section 50.59 of 10CFR50 and approved

[by the Plant Operation Review Committee.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONIAINHEiiT INTEGRITY and centrument leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. While the leakage rate limitation is specified at accident pressure, P., the actual surveillance testing is performed by applying a pressure greater than or equal to P,. This higher pressure accounts for test instrument uncertainties and test volume stabilization changes which occurs under actual test conditions. This method of performing surveillance testing is consistent with the guidance provided in ANSI 56.8-1981 and ensures that the leakage rate measured meets the intent of the LCO and Appendix J.

3/4.6.1.4 and 3/4.6.1.5 AIR PRESSVRE and AIR TEMPERATURE The limitations on containment pressure and average air temperature ensure that: (1) the containment structure is prevented from exceeding its design negative pressure of 8 psia, and (2) the contain:nent peak pressure does not exceed the design pressure of 60 psia during LOCA conditions. Measure-ments shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature. The limits on the pressure and average air temperature are consistent with the assumptions of the safety analysis. The minimum total containment pressure of 10.6 psia is determined by summing the minimum permissible air partial pressure of 8.9 psia and the maximum expected vapor pressure of 1.7 psia (occurring at the ire:imum permissible containment initial temperature of 120*F).

MILLSTONE - UNIT 3 8 3/4 6-la Amendment No. J7, 77, 0641

__________o

e o

Docket No. 50-423 B16758 Attachment 4 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Containment Systems (PTSCR 3-34-97)

Backaround and Safety Assessment October 1997

.s U.S. Nucl:ar.Regul: tory Commission B16758\ Attachment 4\Pagei Backarourld Technical Specifications 4.6.1.1, 3/4.6.1.2 and 3/4.6.1.3 require the testing of the containment to verify that leakage limits at a specified test pressure are consistent with accident assumptions. The proposed changes will modify the list of valves that can be opened in modes 1 through 4, remove a footnote and reword the Technical Specifications and Bases to provide clarity and consistency.

Safety Assessment The proposed revision to Technical Specification Survevance 4.6.1.1.a modifies footnote

""" The Footnote is being modified to move the word " manual" to before the list of manual valves, to delete valves "3FPW-V661, 3FPW-V666, 3SAS-V875, 3SAS-VSO, 3CCP-V886, 3CCP-V887, 3CVS-V13", to add valves "3MSSN885, 3MSSN886, 3MSSN887" to the list of manual valves, to add "Remoto-manual valves 3RHSWV8701A, 3RHSWVSN '

3RHS"MV87028, 3RHSWV8702A", and to replace the "" in the designation of the remaining valves with an "".

The following shows the wording of the existing footnote followed by the wording of the

- footnote after the changes desenbed above.

Existing footnote wording:

"The following manual valves may be opened-on an intermittent basis under administrative control. 3FPW V661, 3FPV/ 666, 3 SSP-V13, 3 SSP-V14, 3HCS-V2, L 3HCS-V3, 3HCS-V9, 3HCS-V10, 3HCS-V6, 3HCS-Vi3, 3SAS-V875, 3SAS-VSO, 3CHS-V371, 3CCP-V886, 3CCP-V887, 3CVS-V13."

Revised footnote wording:

"The following valves may be opened on an intermittent basis under administrative control. Manual valves 3SSPN13, 3SSPN14, 3HCST2, 3HCSN3, 3HCSN9, 3HCSN10, 3HCS"V6, 3HCSN13, 3CHS"V371, 3MSSN885, 3MSSN886, 3MSSN887. Remote manual valves 3RHSWV8701A, 3RHSWV8701B, 3RHSWV8702A,3RHSWV87028."

The bases section for this specification is being modified to add the following:

"The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with control room, at the valve controls, (2) instructing this operator to close these valves ir an accident situation, and (3) assuring that environmental l

6

U.S. NuclIar Regul: tory Commission

- B16758\ Attachment 4\Page 2 conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment."

Basically, these modifications add and delete containment iwlation valves to the list of those that may be opened intermittently under administrative cor' trol. The change to the bases section defines the considerations used for administrative control.

The replacing of the "" with an "" is an administrative change These valves are containment isolation valves and are Quality Assurance Category 1 valves. Millstone Unit 3 uses an "" in the designation of coripenents to designate that the component is Quality Assurance Category 1. This administrative change makes the component designation in the Technical Specification identical to the designation used elsewhere for Millstone Unit 3. '

These changes are being made to add the appropriate considerations of administrative control, in accordance with GL 91-08. The valvas being deleted are in lines that penetrate l containment where either the inside containment isolation valve is a manual valve, or the l line communicates with containment atmusphere, or has minimal operational need to be opened under administrative control. Therefore, the valves in these penetrations are being deleted. The addition of the RHR containment isolation valves is a reflection of the fact that these valves are containment isolation valves and are open in Mode 4.

There are no changes in the operability requirements of the containment isolation valves.

The valves that are being deleted are in lines that penetrate containment where either the inside containment = isolation valve is a manual valve, or the line communicates with containment atmosphere, or the valve has minimal operational need to be opened Deleting these valves from the list of valves that may be opened under administrative control means that these valves will remain closed and, therefore, this can not affect the failure probability of a containment isolation valve to close.

The valves being added, 3MSSN885, 3MSSN886, and 3MSSN887, are in the steam <

-lines to the steam-driven auxiliary feedwater pump These valves can be opened to warm the steam lines prior to testing the steam-driven auxiliary feedwater pump. These valves were recently reclassified as containment isolation valves, which resulted in the need to add them to the list of valves allowed to be opened under administrative control. The

- addition of these valves places their opening under ' administrative control which is consistent with Generic Letter (GL) 91-08, Reference 1. 'me appropriate administrative control considerations are being added to the bases section of this specification, as recommended by GL 91-08. The opening of these valves under appropriate administrative control does not affect the failure probability of a containment isolation valve to close.

- The Residual Heat Removal System valves, RHSW8701 A/B and 3RHSW8702A/Bi are opened during normal cooldown and heatup in Mode 4. These valves are the normal Residual Heat Removal (RHR) System A and System B valves that are open when the RHR System A or B is in service. Nse valves are containment isolation valves that are remote manual valves controlled from the control room. The use of administrative controls

U.S. Nucl:ar Regulttory Com.nission B16758\ Attachment 4\Page 3 for the opening of these valves in Mode 4 is considered appropriate. Allowing these containment isolation valves to be opened is consistent with Technical Specification 3.4.1.3, Reactor Coolant System - Hot Shutdown, which allows the RHR system to be used in Mode 4. Allowing these valves to be open in Mode 4, under administrative control, to use the RHR system, does not affect the failure probability of a containment isolation valve to close.

The change is to the surveillance requirements of the containment isolation valves. The change does not modify the failure mode of any of the containment isolation valves. _ The -

change adds to the bases of the Technical Specification the appropriate considerations for administrative controls, consistent with GL 91-08. These administrative controls provide

! assurance that containment isolation will be accomplished, if required. Therefore, the changes do not affect the consequences of a previously evaluated malfunction of equipment important to safety.

l The surveillance requirements currently allow designated containment isolation valves to i l be open under administrative control. The change does not modify the containment l isolation valves. The change does add valves that can be open under administrative l

controls. Allowing additional valves to ba open under administrative control can not introduce the possibility of a malfunction of a different type.

The change deletes valves from the list of containment isolation valves that may be opened under administrative control. Deleting the valves, which means that they are not allowed to be opened under the Limiting Condition of Operation, can not cause an accident. The valves being added in the steam lines to the steam-driven auxiliary feedwater pump can be used to heat the steam lines prior to testing the steama:iriven auxiliary feed water pump.

Heating the steam lines prior to testing the steam-driven auxiliary feedwater pump does not increase the likelihood of a steam line break.

The addition of the RHR system containment isolation valves reflects the fact that these valves are opened during Mode 4. These valves are opened to allow plant heatup and cooldown in Mode 4. Plant heatup and cooldown in accordance with the Technical Specifications does not increase the likelihood of the above accidents.

The administrative controls include the apprcpriate considerations that containment integrity will be established, when required. By establishing containment integrity, the assumptions in the design basis analyses are essured. This means that for LOCA, steam line break and feed line break accidents inside containment, there is no effect on their consequences.

Valves in the steam lines to the steam drive, auxiliary feedwater pump are being added to the list of valves allowed to be open under administrative control. This means that these '

could be open at the initiation of an accident. As stated above, the administrative controls ui.jer which these valves are opened prov. ides assurance that containment integrity will be established, if required. Similarly, for an SGTR, Locked Rotor or Control Rod Ejection

I U.S. Nuclur Regulatory Commission B16758\ Attachment 4\Page 4 event, the administrative controls provides assurance that these valves will be closed and,

. therefore, allowing them to be opened will not adversely impact the consequences of these events. If failure to close is postulated as a single failure for these events, the results would be bounded by the analyses described in the FSAR. For example, the Locked Roter accident assumes a stuck open steam generator power-operated pressure relief valve (SG PORV). The steam released by the assumed single failure of the SG PORV, for the twenty minutes until the valve is isolated, would exceed the expected releases as a result of failure to close valve 3 MSS *V885, 3 MSS *V886, or 3 MSS *V887, which are in 1/4 inch lines.

Therefore, allowing these valves to be opened under administrative control does not affect the consequences of the previously evaluated accidents.

l The FSAR, Section 15.1.5, provides the assumptions on steam releases for the l consequences of the steam line break accident. The steam generator with the broken l

steam line is assumed to be open to the atmosphere for the duration of the event and, therefore, these valves being open would not impact that assumption. For the unaffected steem generators, steam is assumed released to the atmosphere to remove decay heat.

These valves are in 1/4 inch lines which means that any steam released via this path would only be a small fraction of decay heat and will not adversely affect control of decay heat removal. Therefore, whether these valves are open or not will not affect the consequences of a steam line break outside containment.

Allowing the RHR system containment isolation valves to be open, under administrative contr.r in Mode.4, does not change the way the system is operated. This change to the i

footnote does not change the operators response to an accident in Mode 4. Therefore, the addition of these valves does not affect the consequences of the previously evaluated accidents.

Deleting containment bolation valves from the list of those that are allowed to be opened under administrative control can not modify plant response to an accident. As di=m==ad above, the changes can not cause an accident. The changes do not modify the mitigation strategy for any accioent. Therefore, the changes do not introduce the possibility of a accident of a different type.

The considerations associated with administrative control are being added to the bases of the technical specification. These considerations are identical to those provided in GL 9108. This means that the changes will maintain the margin of safety. The valves that are p allowed to be open in the steam lines to the steam-driven auxiliary feedwater do not impact the accident analyses and therefore do not reduce the margin of safe *y The addition of the RHR system containment isolation valves reflects the fact that these valves are opened for heatup and cooldown in Mode 4 when containment integrity is required.l The change adds the requirements of administrative control to these valves in Mode 4, but does not modify the use of these valves. Therefore, the changes do not reduce the margin of safety.

The proposed change to Technical Specification Surveillance 4.6.1.2.a. will delete footnote "" which referred to an exemption granted by the NRC by letter dated 5/8/95,

Y U.S. Nucirr Regulttory Commission

- B16758%ttachment 4\Page 5 to permit the Type A test to be delayed until RFO6. However, the current extended shutdown has significantly delayed RFO6 and NNECO intends to perform the Type A test during this midcycle shutdown. Therefore, the footnote

to surveillance 4.6.1.2.a

- will be deleted.

The proposed changes to reword the Technical Specifications to provide clarity and consistency include:

Changing the wording from "not less than Pa, 53.?? psia (38.57 psig)" to " greater than or equal to Pa. 38.57 psig" in surveillances 4.6.1.1.c,4.6.1.2.a and 4.6.1.3.b.

This proposed change will word the requirements associated with these surveillances to be more consistent with other surveillances.

Changing the wording from "Pa,53.27 psia (38.57 psig)* to "a pressure greater than or equal to Pa, 38.57 psig" in surveillances 4.6.1.2.d. This proposed change will _,

word the requirements associated with these surveillances to be more consistent with other surveillances.

l

Changing the worrling from "Pa,53.27 psia (36.5/ psig)" to " greater than or equal to

! Pa. 38.57 psig" in surveillances 4.6.1.2.e. This proposed change will word the-requirements associated with these surveillances to be more consistent with other surveillanc.es.

Changing the wording from "Pa. 53.27 psia (38.57 psig)" to "Pa, 38.57 psig" in 3.6.1.2.a,3.6.1.3.b and 4.3.1.3.a. This proposed change will word the requirements associated with these specifications to be more consistent with other specifications.

Changing the wording contained in Bases Sections 3/4.6.1.1, 3/4.6.1.2 and-3/4.6.1.3. The proposed changes will provide further clarification for the surveillances.

The proposed changes do not alter the design, maintenance or function of the containment or containment airlocks, alter the testing of the. containment or containment airlocks, or alter any assumption used in the accident analyses. Based on the above, the proposed changes are not an unreviewed safety question, would not present undue risk to health and safety of the public and are safe.

=

Docket No. 50-423 B16758 Attachment 5 l

Millstone Nuclear Power Statior. Unit No 3 Proposed Revision to Technical Specification Containment Systems (PTSCR 3-34-97)

Slanificant Hazards Consideration and Environmental Gopiderations October 1997 m _ ._

,r a.

U S. Nuclear Regulitory Commission B16758%ttachment 6\Page 1 Sionificant Hazards Consideration -

NNECO has reviewed the proposed revision in accordan:e with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC).

The basis for this conclusion is that the three criterin of 10CFR50.92(c) are not satisfied. The proposed revision does not involve a SHC because the revision would not:

1. Involve a significant increase in the probability of consequence of an accident previously evaluated.

The proposed change to Technical Specification Surveillance 4.6.1.1 deletes valves from the list of containment isolation valves that may be opened under administrative con'rol Deleting the valves, which means that they are not allowed to be opened under the Limiting Condition of Operation, can not cause an accident.

The valves being added in the steam lines to the steam-driven auxiliary feedwater pump can be used to heat the steam lines prior to testing the steam-driven auxiliary feed water pump. Heating the steam lines prior to testing the steam-driven auxiliary feedwater pump does not increase the likelihood of a steam line break.

The administrative change of replacing the "" with an "" in the valve designation can neither cause and accident nor affect the consequences of any accident.

The addition of the RHR system containment isolation valves reflects the fact that these valves can be opened during Mode 4 to allow plant heatup and cooldown Plant heatup and cooldown, in accordance with normal plant operation and the Technical Specifications, does not increase the likelihood of the above accidents.

The administrative controls include the appropriate considerations that containment integrity will be established, when required. By establishing containment integrity, the assump'. ions in the design basis analyses are assured. This means that for LOCA, steam line break and feed line break accidents inside containment, there is no effect on their consequences.

Valves in the steam lines to the steam-driven auxiliary feedwater pump are being added to the list of valves allowed to be open under administrative control. This means that these could be open at the initiation of an accident. The administrative controls under which these valves are opened provides assurance that containment integrity will be established, when required. Similarly, for an SGTR, Locked Rotor or Control Rod' Ejection event, the administrative controls provides assurance that these valves will be closed and, therefore, allowing them to be opened will not adversely impact the consequences of these events, if failure to close is postulated as a single failure for these events, the results would be bounded by the analyses described in the FSAR. For example, the Locked Rotor accident assumes a stuck open steam generator power-operated pressure relief valve (SG PORV). The steam

.r-U.S. Nucl:ar Regulatory Commission B16758\ Attachment 5\Page 2 released by the assumed single failure of the SG PORV, for the twenty minutes until the valve is isolated, would exceed the expected releases as a result of failure to -

close valve 3MSSN885, 3MSSNSd6,_or 3MSSN887, which are in 1/4 inch lines.

Therefore, ellowing these valves to be opened under administrative control does not i affed the consequences of the previously evaluated accidents. l The FSAR, Section 15.1.5, provides the assumptions on steam releases for the consequences of the steam line break accident. The steam generator with the broken steam line is assumed to be open to the aiunosphere for the duration of the event and, therefore, these valves being open would not impact that assumption.

For the unaffected steam generators, steam is assumed released to the atmosphere to remove decay heat. These valves are in 1/4 inch lines which means that any l steam released via this path would only be a small fraction of decay heat and will not adversely affect control of decay heat removal. Therefore, whether these valves are open or not will not affect the consequences of a steam line break outside containment.

Allowing the RHR system containment isolation valves to oe open, under administrative control in Mode 4, does not change the way the system is operated.

This proposed change to the footnote does not change the operators response to an accident in Mode 4. Therefore, the addition of these valves does not affect the consequences of the previously evaluated accidents.

The proposed change to Technical Specification Surveillance 4.6.1.2.a. will delete footnote "" which referred to an exemption granted by the NRC to permit the Type A test to be delayed until RFO6. However, the current extended shutdown has significrditly delayed RFO6 and NNECO intends to perform the Type A test during this midcycle shutdown. The deletion of the footnote does not alter the operation of any system or the containment or containment airlocks, as assumed for accident analyses.

Additionally, Technical Specifications 4.6.1.1, 3/4.6.1.2 and 3/4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/4.6.1.3 are reworded to provide clarity and consistency. These proposed changes do not alter the operation of any system or the containment or containment airlocks during accident analyses, i

Therefore, the proposed revision does not involve a significant increase in the probability or consequence of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident

- previously evaluated.

(

.r

(

U.S. Nucl:ar Rsgul: tory Commission B16758\ Attachment 5\Page 3 The proposed - changes - to Technical Specifications 4.6.1.1, 3/4.6.1.2 - and 3/4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/4.6.1.3 do not alter the operation of any system or the containment or containment airlocks, during normal operation or as assumed in accident analyses.

Deleting containment isolation valves from the list of those that are allowed to be opened under administrative control can nut modify plant response to an accident Adding administrative control when the RHR system containment isolation valves are opened in Mode 4 for normal plant cooldown and heatup can not create a new or different accident. Allowing va!ves to be opened to heat he steam lines to the steam-driven auxiliary feedwater pump prior to testing does not create the possibility of a new or different accident. The administrative change to the valve designation can not modify plant response Therefore, the proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed char.ges to Technical Specifications 4.6.1.1, 3/4.6.1.2 and 3/4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/4.6.1.3 do not alter the design, maintenance or function of any system or the containment or the containment airlocks. Additionally, the proposed changes do not alter the testing of any system or the containment or containment airlocks, or alter any assumption used in the accident analyses.

The considerations associated with administrative control are being added to the bases of the technical specification. These considerations are identical to those

- provided in GL 9108. This means that the changes will maintain the margin of safety. The valves that are allowed to be open in the steam lines to the steam-driven auxiliary feedwater do not impact the accident analyses and therefore do not reduce the margin of safety. The addition of the RHR system containment isolation valves reflects the fact that these valves are opened for heatup and cooldown in Mode 4. The change adds the requirements of administrative controls to these RHR system valves in Mcde 4, but does not modify the use of these valves. The administrative change to tne valve designation can not affect the margin of safety.

Theref re, the proposed revision does not involve a significant redudion in a margin of safety.

In conciusion, based on the information provided it is determined that the proposed revision does not involve an SHC.

t-U.S. Nucl:ar Regulitory Commission B16758%ttachment 5\Page 4 Environmental Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for onvironmental considerations. The proposed revision does not involve a SHC, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusion from the requirements for environmental review.

. .__m

ML20211M908

Text

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