Text

Proposed Tech Specs,Revising TS Section 3/4.6.1, Primary Containment, to Incorporate Requirements of Revised 10CFR50,App J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Effective on 951026
	ML20095D627
Person / Time
Site:	Byron, Braidwood
Issue date:	12/06/1995
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20095D622	List: ML20095D619; ML20095D627;
References
	NUDOCS 9512130184
	Download: ML20095D627 (33)
	v • d • e

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IMEK I

i DEFINITIONS e

t i

SECTION PAGE i

1 10 DEFINITIONS 4

1.1 ACTI0N........................................................

1-1 i.

1.2 ACTUATION LOGIC TEST..........................................

1-1 I

1. 3 ANALOG CHANNEL OPERATIONAL TEST...............................

1-1 1.4 AXIAL FLN DIFFERENCE.........................................

1-1 1.5 CHAMEL CALIBRATION...........................................

1-1 1.6 CHAMEL CECK.................................................

1-1 l

1.7 CONTAll8ENT INTEGRITY.........................................

1-2 1.8 CONTROLLED LEAKAGE............................................

1-2 t

i 1.9 CORE ALTERATION...............................................

1-2 1.9.e CRITICALITY ANALYSIS OF BYRON AE BRAIDWD00 STATION FUEL STORAGE RACKS................................................

1-2 1.10 DIGITAL CHAMEL OPERATIONAL TEST.............................

1-2 1.11 00SE EQUIVALENT I-131........................................

1-2a l

1.12 1-AVERAGE DISINTEGRATION EERGY..............................

1-3 1.D ENEINEERED SAFETY FEATURES RESPONSE TIME.....................

1-3 1.14 FREQUENCY NOTATION...........................................

1-3 gL 1.15 IDENTIFIED LEAKAGE...........................................

(4 -+1-3 1.16 MASTER RELAY TEST..................

1-3 1.17 DEMER(S) 0F TE PMLIC......................................

1-3 1.2s 0FFSITE DOSE CALCULATION MANUAL..............................

1-4 1.19 OPERA 8LE - OPERA 81 LITY.......................................

1-4 s.2.04 pc.

1.19.a OPERATING LIMITS REP 0RT.....................................

1-4 l.

1.20 OPERATIONAL M DE - M DE......................................

1-4 7

1.21 PHYSICS TESTS..

1-4 1.22 PRESSURE 8002ARY LEAKAGE....................................

1-4 1.23 PROCESS CONTROL PR0 GRAM......................................

1-5 1.24 PURGE - PUREING..............................................

1-5 1.25 QUADRANT POWER TILT RATI0....................................

1-5 1.26 RATED THERMAL P0WER..........................................

1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIE............................

1-5 1.28 REPORTA8LE EVENT.............................................

1-5 BYRON - UNITS 1 & 2 I

AENDENT ND. M\\

~ 9512130184 951206 PDR ADCCK 05000454 P

PDR

1.15.a The muimum odlow:n pedey containmed leak.w3t cA L.g, skll be 0.107, of Ne Wag coo}%ewt op eff pv day d St Cal Aabed pu t todaMmM passA ( fg ),

C DEFINITIONS E - AVERAGE DISINTEGRATION ENERGY

1. 12 I shall be the average (weighted in proportion to the concentration of each radionuclide in the samole) of the sum of the average beta and gama energies per disintegration (MeV/d) for the radionuclides in the sample.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the maaitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

FRE00ENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the irtervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as a.

pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or Reactor Coolant System leakage through a steam generator to the c.

Secondary Coolant System.

MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER (S) 0F THE PUBLIC 1.17 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the licensee, its contractors or vendors and persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

BYRON - UNITS 1 & 2 1-3

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

J DEFINITIONS 4

0FFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL. (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-i active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alare/ Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The ODCM shall also contain (1) the 1

Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Sections 6.8.4e and f, and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Specifications i

6.9.1.6 and 6.9.1.7.

I OPERABLE - OPERABILITY h

i 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendent instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxilir y equipment that are required for the system, subsystem, train, componut, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATING LIMITS REPORT 1.19.a The OPERATING LIMITS REPORT is the unit-specific document that provides operating limits for the current operating reload cycle. These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.

Plant operation within these operating limits is addressed in individual specifications.

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, 50.59, or (3) otherwise approv(ed by the Commission.2) authorized under the provision i

i l

j PRESSURE BOUNDARY LEAKAGE 4

1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component i

i body, pipe wall, or vessel wall.

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M be 16. meh cMM W

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7rasm w..+ p g) Ac h de % 6, w d col =4 a"'d d BYRON - UNITS 1 & 2 1-4 AMENDMENT NO. #9

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l

'y /4. 6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION i

3. 6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

A??LICABILITY: MODES 1, 2, 3, and 4.

i ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 .

l SURVEILLANCE REQUIREMENTS

4. 6.1.1 Primary CONTAINMENT INTEGRIT( shall be demonstrated:

At least once per 31 days by verifying that all penetrations".not a.

capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of 1

Specification 3.6.3% oc for conWeaewt 'sel*kaa valves IbA Uc

.p ondv admish conWty b.

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

]

ATter-aa th closing of each penetration subject to Type B testing (

c.

test, by leak ra (te test Qhe seal withment air locks, if opened foll except the conta a pressure not less than P, 44.4 psig, and verifyin when the measured leakage rate forth$sesealsisadd e leakage rat termined pursuant to Specificati

. for all other Type B and ene ations, the akage rate is less than 0.60 L,.

"Except valves, clind flanges, and deactivated automatic valveiwhich are located inside the containment and are locked, sealed or otherwise secured in the closed position.

These penetrations shall be verified closed during each COLD SHUTOOWN except that such verification need not be performed more often than once per 92 days.

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BYRON - UNITS 1 & 2 3/4 6-1

y 3

m CONTAIMENT SYSTDtS CONTAIMENT LEAKAGE i

LINITING CONDITION FOR DPERATION

\\

3.C 1.2 Containerat leakt.ge rates shall be limited to.

An overall integrated leakage rate oft, les.s b or egud L L d P.

a.

4) tr : thr er :; rl t-L

^_" by eight f the re-tain..t zeir par 24 i::r; t ",,,??.4 p:igr-*P-4)

L?ss th-er l%,-4.SM by might-of-the-containment-sir ;:r 24 i::= for W.t 1 (a "M by =i;5t of th? entainment-ele-por 24 h.. for hit 2) :t P y-22,3-psig, t

b.

A combined leakage rate of less than 0.60 L. for all penetrations and valves subject to Type 8 and C tests, when pressurized to P,.

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

EIIM:

With either the measured overall integrated containment leakage. rate exceeding 0.75 L er 0.75 L, : Oplicable,. or the measured combined leakage rate for all pe,etrations,and valves subject to Types B and C tests exceeding 0.60 L,,

n restore the overall integrated leakage rate to less than 0.75 L er less t'r

=0.75 L, es-appHeabler and the combined leakage rate for all pe,netrations subjeci to Type B and C tests to less than 0.60 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 f: Mewing.

. test sch+1e ed ch M h d:terwir.d ir, cer.ferr..ce witti ti,e criterie syn.i-(6ed ir. ?;;::ft:: J cf 10 CR-part 50 :::ing-the-sethed: ::f prr;4sier.: O f ?."i :

N4h4-1972e b acc.cAte wit Rykh Gde 1.10, Redsw o.

a.

Type A (Overall Integrated Containment Leakage Rate) testing shall be conducted in accordance with '.h r;;;;irements-specified-in-Appendix-+

te 10 CFR-50r-as-modified-by-approved-exempt 4enst-Reykb7 6de 11g Revish 0; 1

i BYRON - UNITS 1 & 2 3/4 6-2'"

ARENDMENT NO.'62,

. __ m

_ _ _. _ _. _ _ _.. _ _ _ _ _. _ _ _ _ _. ~. _ _ _. _. _ _.. _... _. _ _ _ _ _ _ _

i CONTAINMENT SYSTEMS j

SURVEILLANCE REQUIREMENTS (Continued) b.

If'a'nY perioditType A test fails to meet either 0.75 r-4:75$

the test schedul's for-subse t Type A tart 6 1 reviewed and approved by the Commission.

'- Mutive Type A tests fall to meet 0.75 L., a Ades shall be perfiNeed et-laast every 18 gonths-until consecutive Type A tests meet 0.75 L, N i

c.

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

}

test dich;. todscha t-acu A% w;A Regv e ry Gvids :.55, Rests,.,o, Aconfirms the accuracy of the test by verifying that the /

supplemental test result, L,, minus the sum of the TypeT and the s'upert sed leak, L, is equal to or less en 0.25 L, or l

0.25 L,;

2)

Has a duration sufficient 40*e sh accurately the change in leakage rate between the A4 st and the supplemental test; and 3)

Requires tha the rate at which gas is inject ato the con-talament or bled from the containment during the supplemental ydest is between 0.75 L, and 1.25 L,.

N d.

Type B and C tests shall be conducted rit' ; : :t : p:::ure-not 1:::

.than-p

peig, et inte ; :t-r +'r. 24 :-th: er--t for

-test: h, =lWgs.

i, accoc Auce wi ReyWy Guide 1.ib3, Rw;3%g, 4)--AirW, -"

4)

"=ge-supply-and-exhaust-isolation-valves-with-cosilient asteria! --els,.

a.

Air locks shall be tested and demonstrated DPERABLE by the require-monts of Specification 4.6.1.3; j

f.

Purge supply and exhaust isolation valves with resilient material i

seals shall be tested and demonstrated OPERABLE by the requirements i

of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable; and g.

The provisions of Specification 4.0.2 are not applicable.

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y 9

BYRON - UNITS 1 & 2 3/4 6-3 AMENDMENT NO. '62,

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Y CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS I

2 LIMITING CONDITION FOR OPERATION

.'em 3.6.1.3 Each containment air lock shall be OPERABLE with:

.a.

Both doors closed except when the air lock is being used for normal transit entry and exits 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*

i at P,, ??. 4 ;;.ig.

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

ACTION:

a.

Wi!.h 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 'osed; i

^

2.

Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE 1

air lock door is verified to be locked closed at least once per j

31 days-l 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 -

l 4.

The provisions of Specification 3.0.4 are not applicable, 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 STANOBY 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 .

9 BYRON - UNITS 1 & 2 3/4 6-4

1 s

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shc11 be demonstrated OPERABLE:

\\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,isd Dei used for multiple entries, then at least once per 72

~

by g

A (1) Verify at the door seal leakage is less-0.0024La(1.11 SCFH) when uma between the doo als is pressurized to greater than or equ 3 psi ans of a permanently installed continuous pres ation and leakage monitoring sys-tem, or (2) Verifying 4 hat the door seal leakage is le 0.01La (4.63 SCEH)'a's determined by precision flow measurements q n measured or at least 30 seconds with the volume between the set) a constant pressure of greater than or equal to 10 psig; b.

By conducting overall air lock leakage tests t =t 1re +Aam-8.,

44.4 psig, r4 v:rifyir.; th: =:r:11 a(M 1::t rate-is-witMe-4ts-limit:-

Insut B

-1)

At 1::st-once-per S = W,* rf

-2)

rier to establisMg MA120ENT INTEGRITV h-eai=+=naara

-h:: 5 r performed cr. t5: cir !=k-that-could-affect-the-air-4eck-seeling-capabHity '**-

At least once per 6 months by verifying that only one door in each c.

air lock can be opened at a time.

D t-leas e per 6 months by verifying that the s@ flow measurements ka 1

hs'd than 0.0 D 3-SCEH) as determined b on when measured for at leas t with the volume betweer. the seals C

at a constant greater %

n n

,*The provisions-of-Specification-4r0r2-are-not-applicablei

*This-repre sents-an-exempt 4 on-to-Appendi x-J-of-10-CFR-Pa rt-50 ;-Paragraph-Hi-

-0,4(b)(44-).-

BYRON - UNITS 1 & 2 3/4 6-5 Amendment No. 4f

._.._...c.-..

l Insert A l

l a.

By conducting air lock door seal leakage tests following each closing at a l

constant pressure of greater than or equal to 10 psig for at least 30 seconds,.

and in accordance with Regulatory Guide 1.163, Revision 0, or by pressurizing

. the volume between the door seals to greater than or equal to 3 psig by means of a permanently installed continuous pressurization and leakage monitoring l

- system.

l i

1 i

Insert B l

I i

m accordance with Regulatory Guide 1.163, Revision 0.

l~

i l

l I

L t

Insert C j

l d.

By conducting air lock door seal leakage tests at a constant pressure of greater than or equal to 10 psig in accordance with Regulatory Guide 1.163, Revision l

0; l

l l

i

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(

r,

(

/'

CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS a

4.6.1.7.1 Each 48-inch containment purge supply and exhaust isolation valve (s) i shall be verified closed and power removed at least once per 31 days.

4.6.1.7.2 Each 8-inch containment purge supply and exhaust isolation valve shall be verified to be positioned in accordance with Specification 3.6.1.7b at least once per 3?. days.

L6oy hsbj Sh\\\\ he Co^duchd on 4.5.1.7.3 st-bst-ence-per-6-months cr. a STAGGERED-TEST-BASI &r the inboard and outboard valves with resilient material seals in each closed 48-inch containment purge supply and exhaust penetration :h:ll be d::enstrated-OpIRABLE by-ver-ifying-that-the-seasured-leakage-rate-4s-less-than4r05-L, when

.pressur-ized-to a+ least P, i'.i psik-on = STAGc.cre Test g43u in a ccorduc e w% ReyWy Gai 1.11,3 t. tesio-O.

4. 6.1. 7. 43 At le=st ence per 3 :::ths, each 8-inch containment purge supply and exnaust isolation valve with resilient material seals shel' be de=eastratM 4PEPASLE by ver4fy4ng-that-the-aeasured4eakage-rateds-less than 0.01 L, when

.pressucized to at. least P,, AA. A prig.

'.4 acco chwa wm Reykh G Ads l.163; Redsion O.

1 Leok e bsb3 sLell be co,Jaclea em 1

W A

BYRON - UNITS I & 2 3/4 6-12 AMENDMENT NO. JI

_ _... _ _. ~. _ _. _ _ _ _ _ _ _. _ _ _ _ _. _ _ _. _ _ _ _ _ _... _. _

3/4.6 CONTAIMENT SYSTEMS f

BASES 3/4.6.1 PRIMARY CONTAIMENT 3/4.6.1.1 CONTAIMENT INTEGRITY Primary CONTAIMENT 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 restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 CONTAIMENT LEAKAGE The limitations on containment leakale rates ensure that the total containment leakage volume will not exceec the value assumed in the accident analyses at the peak accident pressure, P,.

As an added conservatism, the measured overall integrated leakago rate is further limited to less than or equal to 0.75 L, er 0.75 (, es Wic91e, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50,o be 5 Rey *6.y-194y Gode 1.

r 3/. As% 0 Nude 6*1NT AIR LOCKSlash docm=*t NEi 14-i, u Acil Aw3 4 R

4.6.1.8 CONTAIME The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAIMENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become exce< ;ive due to seal damage during the intervals between air lock leakage tests.

^: re of p.ec444on-44ew-measurements of Sp;ificat4:n 4.5.1.3..(2)==t h red d=:=r tM centinec=

nit: ring r;dility != 'le cente! -- !: 15t.

3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1)the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 0.1 psig, and (2) the' containment peak pressure does not exceed the design pressure of 50 psig during steam line break conditions.

The maximum increase in peak pressure ex>ected to be obtained from a cold leg double-ended break event is 44.4 psig-Tw limit of 1.0 psig for initial positive containment pressure will limit the total ressure to 44.4 psig, which is higher than the FSM CheI er accident anal sis calculated peak pres-

-t sure assuming a limit of 0.3 psig for initial posit ye containment pressure, but is considerably less than the, design pressure of 50 psig.

/

BYRON - UNITS 1 & 2 B 3/4 6-1 Amendment No. K

~

I ATTACHMENT B-2 PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77, BRAIDWOOD NUCLEAR POWER STATION, UNITS 1 & 2 Revised Pares:

I l-3 1-4 3/4 6-1 3/4 6-2 3/4 6-3 3/4 6-4 3/4 6-5 3/4 6-12 B 3/4 6-1 1

l c

+

INDEX DEFINITIONS SECTION PAGE'

1. 0 DEFINITIONS 1.1 ACTI0N........................................................

1-1 1.2 ACTUATION LOGIC TEST..........................................

1 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................

1-1 1.4 AXIAL FLUX DIFFERENCE.........................................

1-1 1.5 CHANNEL CALIBRATION...........................................

1-1

1. 6 CHANNEL CHECK.................................................

1-1 1.7 CONTAINMENT INTEGRITY.........................................

1-2

1. 8 CONTROLLED LEAKAGE............................................

1-2

1. 9 CORE ALTERATION...............................................

1-2 1.9.a CRITICALITY ANALYSIS OF 8YRON AND BRAIDWOOD STATION FUEL g

STORAGE RACKS................................................

1-2

/ i 1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................

1-2 1.11 DOSE EQUIVALENT I-131........................................

1-2 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................

1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................

1-3 1.14 FREQUENCY N0TATION...........................................

1-3

'l. lf..a la 1.15 IDENTIFIED LEAKAGE...........................................

w-1-3 1.16 MASTER RELAY TEST............................................

1-3 1.17 ME)SER(S) 0F THE PU8LIC......................................

1-3 1.18 0FFSITE DOSE CALCULATION MANUAL..............................-

1-4 1.19 OPERABLE - 0PERASILITY.......................................

1-4 1.19.a OPERATING LIMITS REP 0RT.....................................

1-4 L 7. 0. a Pa 1.20 OPERATIONAL MODE - M0DE......................................

1-4 i

W 1.21 PHYSICS TESTS................................................

1-4 1.22 PRESSURE 8OUNDARY LEAKAGE....................................

'l-4 1.23 PROCESS CONTROL PR0 GRAM......................................

1-5

))

1.24 P U RGE - PURG I NG..............................................

1-5 1.25 QUADRANT POWER TILT RATI0....................................

1-5 1.26 RATED THERMAL P0WER..........................................

1-5 j

1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................

1-5 1.28 REPORTABLE EVENT.............................................

1-5 AMENDMENT NO. g BRAIDWOOD - UNITS 1 & 2 I

m l.IS; a Ibt My;m allowc b4 piwarf tous hlw ed ledqe tc h, [a, Sf a st be 0.10'7o cf +k< primary ca le:nn,e d a,'y ar et ylt pe r d'y at t-Aa DEFINITIONS 'dc S 'A fe CO M A > "

Pr*n u r* IAb 1

I -' AVERAGE DISINTEGRATIUN ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.14 The FREQUENCY NDTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as

Spa, a.

pump seal or valve packing leaks that are captured and conducted to add a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE 800NDARY LEAKAGE, or Reactor Coolant System leakage through a steam generator to the c.

Secondary Coolant System.

MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER (S) 0F THE PUBLIC 1.17 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the licensee, its contractors or vendors and persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupatfor,al, or other purposes not associated with the plant.

BRAIDWOOD UNITS 1 & 2 1-3 f gy g jQ

c-..... - -... - - _. -. -.. - - - _. - -. -... -

A l

DEFINITIONS f

0FFSITE DOSE CALCULATION MANUAL j

1.18 The 0FFSITE 00SE CALCULATIDN MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses re::ulting from radio-a L

active gaseous and liquid effluents, in the calculation of gaseous and liquid j

effluent monitoring alare/ trip setpoints, and in the conduct of the Envirer.-

. mental Radiological Monitoring Program..The ODCM shall also contain (1) the

{

Radioactive Effluent Controls and Radiological Environmental Monitoring l

Programs required by Sectisns 6.8.4.e and f, and (2) descriptions of the information that should be included in the Annual Radiological Environmental l

Operating and Radioactive Effluent Release Reports required by Specification h

j 6.9.1.6 and 6.9.1.7.

/

i j

OPERABLE - OPERABILITY I

i 1.19 A system, subsystem, train, component or device shall be OPERABLE or have l

OPERABILITY when it is capable of performing its specified function (s), and j

when all necessary attendant instrumentation, controls, electrical power,

{

cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its l

function (s) are also capable of performing their related support function (s).

OPERATING LIMITS REPORT 1.19.a The OPERATING LIMITS REPORT is the unit-specific document that provides operating limits for the current operating reload cycle. These cycle-specific operating limits shall be determined for each reload cycle in accordance with i

Specification 6.9.1.9.

Plant Operation within these operating limits is addressed in individual specifications.

OPERATIONAL MODE - MDDE 1.20 An OPERATIONAL MODE (i.e., MODE). hall correspond to any one inclusive combination of core reactivity condition, power level, and averaga reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE B0UNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

f Dl6

), y, tL h s L. tl b e lh < miM ~ calewle.h) fn cca lc,;w n t N"W ['/U nid 6 & A 9Qjj AMENDMENT No.,59' BRAIDWOOD UNITS ' & 2 1-4

t f

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.

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

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 bour or be in at least HOT STANDBY withih 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 .

i SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

At least once per 31 days by verifying that all penetrations

not a.

capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivatet. automatic valves secured in their positions, except as provided in Table 3.6-1 of Specif 3 cation 3.6.3l trJin cmocr i u-mI-c.;n e.a.& in taJrw vs.la.s Lw c e m us Ar a d es au n

By verifying that each containment air lock is in cosipliance with the -

b.

requirements of Specification 3.6.1.3; and Aft:r :::5 :le:in; ef errk penetratier h5t te Tm 9 tert 4 0.

c.

err--t the cent:f rt

!- 1erhr, i' :;:::d fell =in; e Type ^ cr 9 te: ting th: er! rith ;:: :t : pr:::ere net 1:5

' :t, hy 1::S ret:

than P, ".' p:f;, :nd ;:rifyf t; that 2:n th: = :er:d 1::k ;! rate-

+a ferth!!e:::!: !: dded te the ler' ;e -ster d=+-

iaed pe-sera +

Speci'ic:ti:n '.S.I.2d. f r :l' ether Type 9 trd C pe--tratiaas, +""

th:n 0. 50 L,. 8 y p a r /*re in y c c n 6in,,,, I

.ce-bined 1::k:;; r:t; i: 1:::

ha y +uh us in o.cu,Jm ui % Reptuory Gv;A t. K 3, llo,;g, a., w "Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during j

each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

BRAIDWOOD - UNITS 1 & 2 3/4 6-1 h,,y f g a / /10

~ - -.. -.,

C s

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE f

l LIMITING CONDITION FOR OPERATION f

Containment leakage rates shall be limited to:

An overall integrated leakage rate of[ /sn /4m cre pe 3.6.1.2 a.

1 t: L, 0.1-by :ight :f Th; :::t:.i :::t-i th:2 er :; :

Lee:

2) ti per 2' 5:er: :t P., ".' ;:ig, er 0.0 5 by ::ight :f th: :::t:i;;::t

t

i. ;;t f

1 t: L

-2 b L::: th:2 cr :;::

2r: f:r U:!,t I (0.0 5 by.;:ight ;f th:

ti ; r 2' 5:

21 ; r 2' 5:er: f:r Unit 2) :t P,, 22.2 ;:!;c for all penetrations A combined leakage rate of less than 0.60 Land valves s b.

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

ACTION:

With either the measured overall integrated containment leakage r

li

ble, or the measured combined leakage rate for L,,

all penetrations and valves subject to Types B and C tests exceed 0.75 L,er 0.75 L,,

the restore the overall integrated leakage rate to less than 0.75 L, er 1 ::

lic

ble, and the combined leakage rate for all penetrations subjecf. to Type B and C tests to less than 0.60 L, prior to increas J.75 L,: :

Reactor. Coolant System temperature above 200*F.

SURVEILLANCE REQUIREMENTS The containment leakage rates shall be demonstrated et the fel'o i

r: 1-t::t schedule and :h l' 5: d:t:=ined it :: f:=: :: eith the criter 4.6.1.2 the --thed: : d previrica c ' ""

f t:d ir ^.;; di: 2 ef 10 CF" P:rt 50 erie"Guib t,1Gs, La en,ms C.

in warJs,c e a) n, deyu hu.y

'."5.'-19??:

[

Type A (Overall Integrated Containment Leakage Rate) testi

^.;;::di: 2--

r conducted in accordance with the r:; fr::::t:10 CFP. 50, ::

(

a.

t:

1.1U,Rava,m D3 AMENDMENT NO.

3/4 6-2 BRAIDWOOD - UNITS 1 & 2

CONTAINMENT SYSTEMS e

SURVEILLANCE REQUIREMENTS (Continued) e, go; <<e.m t,,~t fraganay & Typ A halt chait k ;n

~

Th

,e p,ne, If =; ;:rf:dt: T;;;

te t fe!!! te ---+ ei+'-- 0_?5 L, er 0.?! L,d b.

t :t: d:11 5: - etx:d u the t::t ::hrd 1: f:r : 5: :::t T;;:

-

: d by the C--

f::ter.

!f tr er :: tir: T;;;.* t::t: f:f! t:

r,.rt either 0.75 L er 0.75 a,_ __T;;:

t::t dil.l 5: ;; '_:.

d,.:m_

t

_ _ __..g $.. 7 _

... _..__.. s.

A.... 4s 0.75 L, r 0.?5 L,; u.ccrJonc< w; f4 4 pu Wy s u, /c /. /'J, p

flevntcw O.

The accuracy of each Type A test shall be verified by a supplemental test uk4etu cuJacle) in suerJmee tai th deyuldoy GuMe 1./L] Aa&c.* d.

c.

Confi ms the accuracy of the test by verifying that the upplemental test result, L, is in accordance with the riate following equatlon:

ap l L, - (L,, +

$ 0.25 L, or lL, - (L. +

5 0.25 L, or L is t asured T test liakage and L, is where L,,impos.d leak; the super e

2)

Hasadurationsuffjc nt to esta h accurately the change in leakage rate between the Type A test the supplemental test; and res that the rate at which gas is injected in he con-3) ainment or bled from the containment during the suppl al test is between 0.75 L, and 1.25 L, or 0.75 L, and 1.25 L,.

Type B and C tests shall be conducted "!th --r it s ---rre-- -at '-cr d.

the.,, ".' ;:i;, :t t-t:re:1: r.: grnter th:. r :th: Or --t fer Guide t.It3, flaane O t::t:. ::1 ' ::: in uurdwcc wHk R., pts.q 1)

",ir la', ad s

n o..

.....so

..a. 6.... +

4..i.+4..

m. i.... m4+6 454..+- - - -

gr r

.r Air locks shall be tested and demonstrated OPERABLE by the require-e.

ments of Specification 4.6.1.3; Purge supply and exhaust isolation valves with resilient material i

f.

1 seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable; and The provisions of Specification 4.0.2 are not applicable.

g.

AMENDMENT NO.

z BRAIDWOOD - UNITS 1 & 2 3/4 6-3

I CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERA 8LE'with:

Both doors closed except when the air lock is being used for normal a.

transit entry.and exits 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 Pg,". ' ;;;;'iii.

APPLICA8ILITY: M00ES 1, 2, 3, and 4.

ACTION:

With one containment air lock door inoperable:

a.

~

1.

Maintain at.least the. 0FERA6LE air locx coor closed and either restore the inoperable air lock door to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERA 8LE air lock door closed; 2.

Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERA 8LE air lock door is verified to be locked closed at least once per 31 days; 3.

Otherwise, be in at least HOT STAN08Y 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.

The provisions of Specification 3.0/4 are not applicable.

b.

With the containment air lock inoperable, except as the result of an inoperable att lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months er be in at least NOT STAN08Y within the nex) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

9 BRAIDWOOD - UNITS 1 & 2 3/4 6-4 MrJ 6

~

,.y39

-e L.

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

ithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock i used for multiple entries, then at least once per 72 ho

, y 7hjer Y A

(1) Verify n at the door seal leakage is less 0.0024La(1.11 SCFH) when th Quaebetweenthedoor s is pressurized to y

greater than or equT 3 psig ans of a permanently ion and leakage monitoring sys-installed continuous pre tem, or (2) Verifyi pn the door seal leakage is han0.01La(4.63 SCF )1s determined by precision flow measurene en measured or at least 30 seconds with the volume between the s at a constant presstre of greater than or equal to 10 psig; j

b.

By conducting overall air lock leakage tests

t le:S thea ?,,

44A-p:i;, ed verifyia; the evere!! ei-1 9

/ N e 'e+= is wi+hia-it: li=it:

Inforf 8 1) et 10::t =:: per S :=S:,* =d 2-)

8-icr b ::t911:hi ;; C0"AI:^^2T IMGRIN esa efateneace h:: h= p:rf:;;;d = th: :f r 10:k S:t ce!d :ffet the air 4eek :=11a; :--dility **

At least once per 6 months by verifying that only one door in each c.

air lock can be opened at a time.

6 months by verifying that the seal 1 s

/

d.

At ea than 0.01La (4.63 reined ow measurements L

e volume between the seals when measured for at m.w,= -

pas t_.

ssure of greater than or equa l-niorf Cl

The previsions of Speciff retiaa 1.0.2 ere aat epplic 9!e.

- TM: represent: e e_M 7 tion 6 App edix J ef 10 CF" ":rt 50, P:r:;reph III-S.2(b)(ii).

i BRAIDWOOD - IINITS 1 & 2 3/4 6-5 Amendment No.

_. _. _. _ - _ _ _ ~ _ _. _ _ _ _

Insert A a.

By conducting air lock door seal leakage tests following each closing at a constant pressure of greater than or equal to 10 psig for at least 30 seconds, and in accordance with Regulatory Guide 1.163, Revision 0, or by pressurizing the volume between the door seals to greater than or equal to 3 psig by means of a permanently installed continuous pressurization and leakage monitoring system.

Insert B in accordance with Regulatory Guide 1.163, Revision 0.

I l

l Insert C I

i d.

By conducting air lock door seal leakage tests at a constant pressure of greater than or equal to 10 psig in accordance with Regulatory Guide 1.163, Revision 0;

. -.. = -

(

e.

I CfMITAl q SVRf B 1 j

mangryLLs afeggargerfg 6

4.8.1.7.1 'Esch 48-tech containment purge supply and exhavst solation valve (s) shall be verified glosed and power removed at least esce per l il d4ys.

i 4.8.1.7.2 Each E-inch contalaeont surge supply and exhaust fielettes' valve

[ 4 shall be vyifted to be posittened la accordance with speciftration 3.4.1.76 at

/

Teast once per 31 days, j

4.8.1.7.3 et ? ret xx r. x ' cake.J an Lu an +u n g sl it l>c s

g e - e " :tM trat ***1 8-the 1)ibsard and outboard valves with restilent material sosis la each closed48-p1 containmen ly and exhaust ty.. 3^ t p pres ^ 'f;'g sp" r:-"M !*"^;penetratten 9e ! '- fx:

nt--t__-ep5 MIL 4-

...__t..

Mte is less tH 9.M ("9

" "' '"^^^

St.t 7:'=

on c. SiAGGERED TEir 81\\ SIS in 4.a.f.Jc.uk 4;Ik Reguic,Hq Guile,UO'h 4-inch containment acu ae vis,an O.

7.4+ ^t.n:t :x: T _ x^ke eac qurge sepE x t. _@_-.

and exhaust Isolation valve with resillent seter141 seals th?? 5 gxx:1g:::t:=:P.,et.e;:t:.sc:.,;;; "-t S r r;d 1-"- : - t: 1: !re "r e.eS '., ^r

^= "- ^-'?

^

m ou,rn,, a 4

g. j, 1.10,12evisidn O.

7 bedaye hs H q Sls J be cewA&e.0 cM S

e IRAIDWOOO - UNITS 1 & 2 3/46-11 J MDf!PtDIT't NO.

t j"bhht,';6d"'

tit t U t t W-Li -111 GE D

'9 g,j' a

@ 6 '95 10:05 7086637155 PAE.0EA

~

1 1

l 3/4.6 CONTAINMENT SYSTEMS 1

BASES

\\

i l

3/4.6.1 PRIMARY CONTAINMENT t

3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive i

l materials from the containment atmosphere will be restricted to those leakage i

This i

paths and associated leak rates assumed in the safety analyses.will limit the

)

restriction, in conjunction with the leakage rate limitation, lues of 10 CFR SITE BOUNDARY radiation doses to within the dose guideline va Part 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total centainment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P,.

As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, er 0.75 L;, es --a!!ceble, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50 st, nd Akb/A N

% rim o, ha n e~mvhsnk Ac& NEZ n-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 CONTAINM6NT INTEGRITY and containment Surveillance testing of the air lock seals provides assurance that leak rate.

the overall air lock leakage will not become excessive due to seal dama during the intervals between air lock leakage tests. TM = ef preis ea y/

fle -aesurrats of Specificatien AM t =(?) =<t ha ee M

eaavec th=

ta centml rea= is ! :t.-

h caatiaeaus =aite*4~; c=rebi'ity ia 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 0.1 psig, and (2) the containment' peak pressure does not exceed the design pressure of 50 psig during steam line break conditions.

The maximum increase in peak pressure expected to be obtained from a cold The limit of 1.0 psig for initial leg double-ended break event is 44.4 psig.

positive containment pressure will limit the total pressure to 44.4 psig, which is higher than the.T" Ch;ter accident analysis calculated peak pres-sure assuming a limit of 0.3 sig for initial positive containment pressure, but is considerably less tha the design pressure of 50 psig.

UFMe Chyh r IS' BRAIDWOOD - UNITS 1 & 2 8 3/4 6-1 Amendment No

~

a ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Commonwealth Edison Company (Comed) has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10, Code of Federal Regulations, Part 50, Section 92, Paragraph c [10 CFR 50.92(c)], a proposed amendment to an operating license involves no significant i

hazards if operation of the facility in accordance with the proposed amendment would not:

1.

Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.

Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3.

Involve a significant reduction in a margin of safety.

Comed proposes to revise Byron Nuclear Power Station, Units I and 2 (Byron), and Braidwood Nuclear Power Station, Units 1 and 2 (Braidwood) Technical Specification l

(TS) Section 3/4.6.1, " Primary Containment," and the associated Bases to reflect mcent changes to Appendix J to 10 CFR 50, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." The proposed revisions include:

l 1.

Adding TS Definitions 1.15.a for the maximum allowable primary l'

containment leakage rate (L,) and 1.20.a for the maximum calculated primary containment pressure (P,). The redundant definitions i

throughout TS Section 3/4.6.1 are deleted, 2.

Adding numerous statements throughout TS Section 3/4.6.1 that leak l

rate testing is performed in accordance with Regulatory Guide (RG) l 1.163, Revision 0, " Performance-Based Containment Leak-Test Program," and its referenced documents, 3.

Deleting TS requirements that are taken verbatim from 10 CFR 50, Appendix J. The specific requirements will be placed in the containment leakage rate test program in accordance with RG 1.163, and its referenced documents, and s

4.

Clarifying Technical Specification Surveillance Requirement (TSSR) 4.6.1.1.a for consistency with NUREG-1431, Revision 1 Standard Technical Specifications for Westinghouse Plants."

A.

The proposed changes do not involve a significant increase in the

- probability or consequences of an accident previously evaluated.

10 CFR 50, Appendix J, has been amended to include provisions regarding performance-based leakage testing requirements (Option B). Option B allows plants with satisfactory Integrated Leak Rate Testing (ILRT) performance history to reduce the Type A testing frequency from three tests in tan years to one test in ten years. For Type B and Type C tests, Option B allows plants to reduce testing frequency based on the leak rate test history of each component.

In addition Option B establishes contmis to ensure continued satisfactory

- performance of the affected penetrations during the extended testing interval.

To be consistent with the requirements of Option B to 10 CFR 50, Appendix J.

Comed proposes to include appropriate changes to the TSs that incorporate the necessary revisions.

Some of the proposed changes represent minor curtailments to current TS requirements, but are based on the requirements specified by Option B to 10 CFR 50, Appendix J. Any such changes are consistent with the current plant safety analyses and have been determined to represent sufficient requirements for the assurance of the reliability of equipment assumed to operate in the safety analyses, or provide continued assurance that specified parameters associated with containment integrity remain within their acceptance limits. The other proposed changes maintain consistency with those requirements specified by Option B to 10 CFR 50, Appendix J and are consistent with the current plant safety analyses. Implementation of these changes.will provide conti9ued assurance that specified parameters associated with containment integrity will remain within their acceptance limits, and as such, will not significantly increase the probability or consequences of a previously evaluated accident.

The associated systems affecting the leak rate integrity are not assumed in any safety analyses to initiate any accident sequence; therefore, the probability of

. occurrence of any accident previously evaluated is not increased. In addition, the proposed changes to the limiting conditions for operation and surveillance requirements for such systems are consistent with the current 10 CFR 50, Appendix J, requirements. The proposed changes maintain an equivalent level of reliability and availability for all affected systems.

2-

'+1m=w-w-

-d-M-i v

-w

=amy

-+y+5W

---+m-r--g+

=rt'---

w-e-

--m

,em-s'w-re'

=

Maintaining allowable leakage within the analyzed limit assumed for the accident analyses does not adversely affect either the onsite or offsite dose consequences. Furthermom, containment leakage is not an accident initiator.

As such, there is no adverse impact on the probability of accident initiators.

Thus, there is no significant incmase in the probability or occurrence of any previously analyzed accident, or increase the consequences of any previously analyzed accident.

1 B.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Option B of 10 CFR 50, Appendix J, specifies, in part, that a Type A test may be conducted at a periodic interval based on the performance of the overall containment system. Type A tests measure both the containment system overall integrated leakage rate at the containment pressure boundary and system alignments assumed during a large break loss-of-coolant accident (LOCA), and demonstrate the capability of the primary containment to withstand an internal pressure load. The acceptable leakage rates are specified in the TSs. For Type B and C tests, intervals am proposed for establishment based on the performance history of each component. Acceptance criteria for each component are based upon demonstration that the leakage rates at design basis pressure conditions for applicable penetrations are within the limits specified in the TSs.

The proposed changes reflect the requirements specified in the amended 10 CFR 50, Appendix J, and are consistent with the current plant safety analyses. Some minor curtailments of current TS requirements are based on generic guidance or similarly approved pmvisions for other plants. These changes do not involve revisions to the design of the plant. Some of the changes may involve revision in the testing of components at the plant; however, these are in accordance with the current plant safety analyses and provide for appropriate testing or surveillance that is consistent with Option B to 10 CFR 50, Appendix J. The proposed changes will not introduce new failure mechanisms beyond those already considered in the current plant safety analyses.

No new modes of operation are introduced by the proposed changes.

Surveillance requirements are changed to reflect corresponding changes associated with Option B to 10 CFR 50, Appendix J. The proposed changes maintain at least the present level of operability of any such system that affects plant containment integrity. Therefore, the proposed changes do not create the l

possibility of a new or different kind of accident from any pmviously evaluated.

3-

7 0

i

)

l The associated systems that affect plant leak rate integrity related to the proposed amendment are not assumed to initiate any accident sequence. In addition, the proposed surveillance requirements for any such affected systems are consistent with the current requirements specified within the TSs and are consistent with the requirements of Option B to 10 CFR 50, Appendix J. The pmposed surveillance requirements maintain an equivalent level of reliability and availability of all affected systems and, therefore, do not affect the i

consequences of any previously evaluated accident. As such, the probability j

of systems associated with leak rate test integrity failing to perform their intended function is unaffected by the pmposed limiting conditions for operation and surveillance requirements.

f I

C.

The proposed changes do not involve a significant reduction in a margin of safety.

}

The provisions specified in Option B to 10 CFR 50 Appendix J, allows changes to Type A, B, and C test intervals based upon the performance of past i

leak rate tests. 'Ihe effect of extending containment leak rate test intervals is a corresponding increase in the likelihood of containment leakage. The degree to which intervals can be extended has a direct impact on the potential effect on j

existing plant safety margins and the public health and safety that can occur

)

due to an increased likelihood of containment leakage.

Changing Type A, B, and C test intervals from those currently provided in the TS to those provided for in 10 CFR 50, Appendix J, Option B, slightly

)

increases the rist associated with Type A, B, and C specific accident sequences. Historical data suggest that increasing the Type C test interval can slightly increase the associated risk; however, this is compensated by the corresponding risk reduction benefits associated with reduction in component cycling, stress, and wear associated with incirased test intervals. In addition, when considering the total integrated risk, which includes all analyzed accident sequences, the additional risk associated with increasing test intervals is negligible.

The proposed changes are consistent with those provisions specified in Option B of 10 CFR 50, Appendix J, and are consistent with current plant safety analyses. In addition, these proposed changes do not involve revisions to the design of the plant. As such, the proposed individual changes will maintain the same level of reliability of the equipment associated with containment integrity, assumed to operate in the plant safety analysis, or provide continued assurance that specified parameters affecting plant leak rate integrity, will remain within their acceptance limits. Therefore, the proposed changes provide continued assurance of the leakage integrity of the l L_

+

)

1 l

containment without adversely affecting the public health and safety and, as such, will not significantly reduce existing plant safety margins.

The proposed changes are based on United States Nuclear Regulatory Commission (USNRC) accepted provisions and maintain necessary levels of system or i

component reliability affecting plant containment integrity. The performance.-

based approach to leakage rate testing concludes that the impact on public health and safety due to revised testing intervals is negligible. The proposed changes will not reduce the availability of systems associated with containment integrity when they are required to mitigate accident conditions; therefore, the proposed changes do not involve a significant reduction in the margin of safety.

Guidance for the application of standards to license change requests for determination of the existence of significant hazards considerations has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule,51 FR 7744. This document provides examples of amendments which are and are not considered likely to involve significant hazards considerations. Thc adoption of the requirements for the revised 10 CFR 50, Appendix J, most closely fits the example of a change which may either result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin. However, the proposed amendment results in a change which is clearly within all acceptable criteria with respect to the system or component specified in NUREG-0800, Standard Review Plan, Section 6.2.6, Containment Leakage Testing. The proposed changes retain the current specification leak rate limits and acceptance criteria, thus preserving the safety mar;;in, and will not significantly increase the consequences of an accident.

This proposed amendment does not involve a significant relaxation of the criteria used to -

establish safety limits, a significant relaxation of the bases for the limiting safety system settings, or a significant relaxation of the bases for the limiting conditions for operations. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), Commonwealth Edison has concluded that these changes involve no significant hazards considerations..

l ATTACHMENT D l

ENVIRONMENTAL ASSESSMENT Commonwealth Edison Company (Comed) has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring envimnmental assessment in accordance with Title 10, Code of Federal Regulations, Part 50, Section 51 (10 CFR 51.21).

Comed has determined that the pmposed change meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within a restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

(i) the amendment involves no significant hazards considerations, As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards considerations.

(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and As documented in Attachment C, there will be no change in the types or significant increase in the amounts of any effluents mleased offsite.

i (iii) there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed change will not result in changes in the operation or configuration of the facility. There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste; nor will the proposal result in any change in the normal radiation levels within the plant. Therefore them will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

i 1

ATTACHMENT E IMPLEMENTATION PLAN FOR 10 CFR 50, APPENDIX J, OPTION B Commonwealth Edison Company's (Comed's) Byron Nuclear Power Station, Units 1 and 2 (Byron), and Braidwood Nuclear Power Station, Units 1 and 2 (Braidwood) will incorporate the performance oriented and risk-based approaches included in the following documents into their i

containment leakage rate testing programs:

Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," Option B, Regulatory Guide (RG) 1.163, Revision 0, " Performance-Based Containment Leak-Test a

Program,"

Nuclear Energy Institute (NEI) 94-01, Revision 0, " Nuclear Energy Institute Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"

and ANSI /ANS-56.8-1994, "American National Standard for Containment System Leakage Testing Requirements."

10 CFR 50, Appendix J, Option B provides a performance based option for Type A, B, and C leakage rate testing of primary containment. This option improves the focus of the regulation by i

eliminating prescriptive requirements that have been determined to be marginal to safety. The j

new rule allows for test intervals to be based on system.and component performance and

)

l provides for greater flexibility for cost effective implementation methods for regulatory safety l

objectives.

j Comed has formed an Appendix J Implementation Task Force to implement and interpret the new 10 CFR 50, Appendix J in a consistent manner throughout Comed. Each Comed nuclear station (including Byron and Braidwood) is represented in the group. The task force will provide generic guidelines for all Comed nuclear stations for the implementation of 10 CFR 50, Appendix J, Option B.

e i

I COMPONENT. LEAKAGE LIMITS i

Byron and Braidwood will use the administrative limits set by the Comed Appendix J l

Implementation Task Force for each component mquiring Types B and C leakage rate testmg.

l To determine whether an as-found local leak rate test (LLRT) passed or failed, a component's l

measured leakage is compared against its administrative limit. The task force carefully j

evaluated the administrative leak rate limits to determine the pioper limits, which are extremely important under the performance-based rule. These new administrative limits will be used to determine whether future or previous tests passed or failed, Thus, the limits chosen will affect each component's Type B or C testing frequency.

Two limits will be specified for each component, a warning limit and an alarm limit. When the component's leakage rate is above the warning limit and below the alarm limit, then the l

component should be evaluated for repair. This is not counted as a performance failure. When the component's leakage rate is above the alarm limit, then the compoaent must be repaired, except as noted below. This is counted as a performance failure.

l i

i Although administrative limits are used to maintain the containment in good condition, it should

.l l

be noted that the sum of the as-left maximum pathway leakage rates for all Appendix J barriers i

I must be less than 0.6 L, per plant Technical Specifications, where L, is defined as the maximum

. allowable primary containment leakage rate. In the past, there have been instances where the leakage from one or more components has exceeded the alarm limits. To bring the leakage rate

{

below the limit prior to start-up would have been very difficult and/or costly. For those special cases, a safety evaluation was performed. If this evaluation concluded that there was no significant safety impact, then the component (s) was(were) allowed to continue to leak in excess l'

of the individual valve leakage limit until it could be repaired, provided that the Technical Specification limit of 0.6 L, was not exceeded. It must be noted though, that the test was still i

considered to be a failure in spite of the safety evaluation. Byron and Braidwood reserve the i

option to continue to use this provision only on a critical, as needed basis.

l I

J I,

o i

l BUILDING PERFORMANCE BASELINES /ESTABLISIIING TEST FREOUENCIES i

Type A Test In accordance with the new mquimments associated with 10 CFR 50, Appendix J, Option B, i

Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 10 years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where calculated as-found performance leakage rate was less than 1.0 La. Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to detemline performance shall be the normal Bymn and Braidwood refuel interval. NEI 94-01 states that this interval shall be at least 24 months, i

however, the normal Byron and Braidwood refuel interval of 18 months is a more appropriate minimum interval between Type A tests.

The new rules allow for reviewing past performance history with several options to determine if past Type A tests were satisfactory:

a.

As-Found Type A test results can be compamd to 1.0 L, rather than the previous 0.75 L, criteria.

b.

Lcakage savings (repairs / adjustments) from Type B and C testable pathways which were added as penalties to the As-Found Type A test can be subtracted when reviewing previous Type A test results.

c.

The Type A test upper confidence limit from previous Type A tests may be recalculated using the Mass Point Methodology described in ANS 56.8-1994.

Byron has reviewed Type A test results as compared to the current requirements and criteria to establish a test frequency for the primary containment integrated leak rate test (ILRT). In reviewing Byron Type A history, it has been determined that the two most recent as-found Type A tests for Unit I have been below the 1.0 L, criteria. Therefore, Byron, Unit 1, will implement the 10 year Type A test frequency based on the criteria set forth in the new rule during the next refuel outage, Byron, Unit 1 Cycle 7, Refuel Outage (BIR07). Byron Unit 2, and Braidwood data will be evaluated to determine applicable future test frequency requirements, I

based on the Type A test performance history. Braidwood is pursuing resolution of comments on previousILRTs with the United States Nuclear Regulatory Commission (USNRC). If this effort is successful, Braidwood may implement the 10 year Type A test frequency of Option B to Appendix J.

3-

O t

l Tyne B and C Tests l

Byron and Braidwood will formulate administrative procedures for documenting Type B and C testing performance. A performance evaluation will be used to ensure that consistent criteria were applied to establish component baseline performance and their subsequent testing frequencies.

Byron and Braidwood have developed a computer database to compile all the required leak rate historical data to be used in the evaluation process. This database will continue to be updated with the most current as-found leak rate data acquired during the most recent refuel outages.

The performance history of each component will be evaluated against the administrative limit to rate component performance over the last three refuel outages. In addition to a performance history evaluation, considerations such as service life, environment, design, system application, special service conditions, and safety impact / risk from failure will be Irviewed and evaluated, and will be used to determine test frequency.

TECHNICAL CRITERI A & TESTING METHODOLOGY INTERPRETATION The containment leakage rate testing program will follow the guidance in RG 1.163, NEI 94-01, ANSI /ANS-56.8-1994, and 10 CFR 50 Appendix J, Option B. The administrative procedure (s) for the containment leakage rate testing program will follow the requirements and contain the performance criteria for the Types A, B, and C testing. The administrative procedure (s) will also contain the description of the record keeping and methodology to establish test intervals for equipment and components in the containment leakage rate testing program. The equipment and component test procedures will contain information on the proper techniques and methods for performing the Type A, B, and C tests.

ML20095D627

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