ML17264A253

Proposed Tech Specs for Implementation of 10CFR50,App J, Option B

ML17264A253
Person / Time
Site:	Ginna
Issue date:	11/27/1995
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17264A251	List: ML17264A250 ML17264A252 ML17264A253
References
NUDOCS 9511290180
Download: ML17264A253 (38)

Text

Attachment II Marked Up Copy ofR.E. Ginna Nuclear Power Plant Technical Specifications and License Included Pages:

License, page 5 4.4-1 4.4-2 4.4-3 44 4 4.4-5 4.4-6 4.4-7 4.4-8 4.4-11a 4.4-12 4.4-13 4.4-14 4.4-17 9511290180 951127 PDR ADOCK 05000244, P..,,,,,

.,,PDR,

5 (a)

Provisior s estahlishing preventive maintenance and periodic visual inspection requirements; and (h>

Leak.test requirements tor each system at a frequency not. to exceed refueling cycle intervals.

(61 Iodine Monitorin The licensee shall implement a program which will ensure the I capability to accuratel.v determine the airborne iodine concentration in vital areas under accident. conditions.

This program shall include the following:

(a)

Training of personnel; (b)

Procedures for monitoring; and (c)

Provisions for maintenance of samplina and analysis equipment.

o.~h D.

The I"ccility require exemptions from certain requiremenis of 10 CFR 50.46(a)(1),

50.48(c)(4) and A

OCF art 5

These include:

(1) an exemption rom 50.46(a) (1), that ECCS performance be calculated in accordance with an acceptable calculational model which conforms tn the prnvisions in Appendix K (SER dated April 18, 1978).

The exemption will ex ire u on recei t and approval of revised ECCS calculations

2) certain exemptions rom poen

~x to art 50 section III.A.4.(a) maximum allowable leakage rate for reduced pressure

tests, section III.R.l acceptable technique

+or performing local (Type 8) leakage rate tests, section III.D.1 scheduling of containment integrated leakage rate tests, and section III.D.2 testinq in al cnntainment airlocks (SER dated March 28, 1978);

and an exemption to the sche u ar requiremen.s or t e alterna

~ve shutdown svstem as set forth in 10 CFR 50.48(c)(4)

(NRC letter dated Mav 10, 1984).

The exemption is effective until startup from the 1986 refueling outage.

The aforementioned exemptions are authorized by law and will not endanger life or propert i or the common defense and securitv and are otherwise in the public interest.

Therefore, the exemptions are hereby granted pursuant tn 10 CFR 50.12.'.

Physical Protection

- The 11censee shall maintain in effect and fullv implement all provisions o

the following Commission-approved documents, including amendments and changes made pursuant

.o the authority of 10 CFR 50. 4(p), which are beina withheld from public disclosure pursuant tn 10 CFR 73.21:

4.4 Containment Tests

~A1 bi 1 Applies to containment leakage and structural integrity.

~0b'ective To verify that potential leakage from the containment and the pre-stressing tendon loads are maintai'ned within specified values.

S ecification

.4.1 Inte rat Leg kg e ate Test 4.4.1 Defi itions

~Pa QQ is the 60 psig.

ntainm t vessel/'design pr sure Pt psig) is the cont inment v sel red ed test ressure for per odic tes ng.

Lt (w ight per ent/24 ho rs) is t e maximu allowabl eakage r es of th contain nt vesse test at sphere at pre ure Pt.

4. 4-1

~

4

~O IC,l~~

oMu~ ~Mxe Pli~~

laa.4;o.~~

La>(w ght p

cent/24 ours) i the max'm allow le eakage ate of e cont nment v sel test tmos-here at go~Pa,'.2 she.A Sc

~C qr<~g~ M~w&e~n3e o.~p

'4+ pkr 4.

Lam nd Ltm ('ght pere nt/24 hour are e

ota measured ontainme leakage r tes of th contain-ment ssel tes atmospher at pressur s

Pa and t re ectively 4.4.1.2 Pretest Re uirements a.

A visual examination of the accessible interior and exterior surfaces of the containment structure (Q

Q w,~Q;~

46>> Rq-ol,

+~Worn 9,+,L 0 C, L.Ll 3,

~~an C

~ 3 shall be performed to uncover any evidence of structural deterioration which oey affect either the containment.structure integrity or leak-tightness.

If there is evidence of structural deterioration, integrated leak rate testing shall not be performed until appropriate corrective action has been taken.

Except for repairs to correct structural deterioration,

however, no repairs or adjustments shall be made during the period between the initiation of the inspection and the performance of the test.

~:

Pr&ix ) ~~5 5Co.Q - hQRQ b.

Closure of containment isolation valves shall be accomplished by normal operation and without any preliminary exercising or adjustments.

4.4-2

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4.4.1.4 Acce tance Criteria b.

he leaka e rate Lt shall be 0.75 Lt Pt.

P is defin d as the ontainmen vessel r duced te pressur which is reater th n or equa to 35 ps Ltm is defined s the tot measure contai ent leaka e rate at pressure P

Lt is efined a

the max'm allowa le leakag rate at essure P

Ptl'I*

Lt shall be etermined as Lt = L

~aJ whic equals 528 pere t weight'r day 35 psi Pa is efined a the calcu ted peak ontainme intern pressure elated to esign ba s accide ts which s

greater han or e al to 60 ig.

La 's define as the m

imum all able lea age rate at Pa ich equal

.2 perce weight r day.

I ~

c.

The leakage rate at Pa-shall be <0.75 La.

am s

deMi e as t tota easur contai en le a

a at

~~~~ A ~M, ~iko~img pX~

4.4.1.5 Test Pre uenc S~&q 9<<~~ ~~iavva~

l~a ~ ~~+~~

a.

A s of thr e integ ted'eak te tests s

e pe ormed approx ately eq 41. interval during ch 10-ar serv'ce period.

The thir test of each s

shall b

conducte in the fi l year o

the 1 -year ser ice period or one ye before after the fina year of t 10-year s rvice er'od p ~

~~+ ~~ ~~Q~XC ~~VWXhh~4+~~3. ~VS e

~t..s i.

the interval, hetween any two T

e A tests

~Co'>~ll~ not exceedg:-e~ years,

(~i~ ~o.4 i~~~W

> 5 ewa~

ii.

following each in-service ins ection the containment airlocks t e )team nerato nspp tion/m ntena ce penetration an the eq c

are ea tested prior to returning the plant to operation, and-any repair, replacement, or modification of a

containment barrier resulting from the inservice inspections shall be followed by the appropriate leakage test.

Amendment No.

54

4. 4-4

b. If any test fails to meet the acceptance criteria C hd.~~a R mi~o ~n 9g ol MWsam of 4.4.1.4.a he est sched e for su equent re arly sc duled ins vice test shall be b-mitted t the Comni ion for r view and a

roval.

c.

I two consec 've tests ail to m

the accep ance criteria q 4.4.1.4.a a retest shall be performed at e ch refuel ng shutd n or a joximatel ever wa+hi~

LS

Qmonths, sc sr until two con-c~~~~~~wN ~'lB~~

)

secutive tests meet the acceptance ct iteria of~

4.4.1.4.a, after which time the retest schedule of 4.4. 1.5.a may be resumed.

4.4.1.6 Additional Re uirements p~~4, a.

A summary technical report shall be admit d ro c..u Co ss on after the conduct of each integrated f4.W,~ (+

Ni,L 9q ol w~~,~,i'Q 0 m~ (e~~ Sc. 5-A'LQ leak rate test.

Information on any valve closure malfunction or valve leakage that requires cor-rective action before the test shall be included in the report.

4.4.2 Local Leak Detection Tests 4.4.2.1 Test e

Local leakage rate tests shall be performed at intervals specified in 4.4.2.4 below and at a pressure of not less than "psi~

~~c @~a~

4.4-5

((O b.

The local leakage rate shall be measured for each.

of the following components:

i.

Containment penetrations that employ resilient

seals, gaskets, or sealant compounds, piping penetrations with expansion bellows and electrical pene'trations with flexible metal uxor. gq o)

~ A&RT/p~

seal assemblies'.

ii. Air'lock and equipment door seals.

iii. Fuel transfer tube.,

iv v

Isolation valves on the testable fluid systems lines penetrating the containment.

Other containment components, which require leak repair in order to meet the acceptance criterion for any integrated leakage rate test.

4.4.2.2 Acce tance Criterion Containment isolation boundaries are inoperable from a leakage standpoint when the demonstrated leakage of a

single boundary,or cumulative total leakage of all boundaries is greater than 0.60 La.

4.4.2.3 Corrective Action Xf at any time it is determined that the total leakage from all penetrations and isolation boundaries exceeds 0.60 La, repairs initiated immediately.

shall be Amendment No.54 4.4-6

e b.

If repairs are not completed and conformance to the acceptance criterion of 4.4.2.2 is not demonstrated A

I within ~ hour~

t:he reactor shall be shutdown and depressurized until repairs are effected and t'e local leakage meets t:he acceptance criterion.

c.

If it. is determined that the leakage through a mini-purge

'3, g.g Ah.~ye, ~~4 engi ering aluat:ion s ll be p formed and lans 4

rectiv action de eloped.

.4.2.4 Test Fre uenc M~X~~ ~MWon i~pu~ ~

i4OX~ ~a~i~W,MVA.

a.

Except, as specified in b.

and c.

below, individual penetrations and containment isolation valves shall be a6 i.\\4 tested in accordance withe 0

FR 50, Append x 0 as modified by approved exemptions.

mS o~O~~~ ~~~~~-~am~ ~~~a b.

She containment equipment

hatch, fuel transfer

tube, 4> Qq-oc,

~ l 0,'L.l.+

tfh 1 sooner.

steam generator inspection/maintenance penetration,.and g x'40~~ t.mi'iAe,~

oh',M~M Zu o shutdown purge system flanges shall be tested eW E

a Amendment.

No.

$ $,$ $,59 4.4-7

INSERT 1 Airlock acceptance criteria are:

1)

For each air lock, overall leakage rate is z 0.05 L, when tested at > Pand 2)

For each door, leakage rate is c 0.01L, when tested at a P,.

K~o c.

The containment air locks shall be ested at 30 intervals of no more than

~~months by pressurizing the space between the air lock doors.

Xn addition, following opening of the air lock door during the interval, a test shall be performed by pressurizing between the dual seals of each door

opened, within~48

~(or 3o h,~ 'i4 o~+

of the openings, unless the reactor was in the cold phutdown condition at the time of the opening or has been subsequently brought to the cold shutdown condition.

A test shall also be performed by pressurizing between the P C.iaaf dual seals of each door wxt n

48 h urs o

leaving the cold shutdown condition, unle he ors have t

een op since the ast t st perfo ed either y

pressur ing the space tween the ir lock ors or res urizing bet en the dual door seals Amendment No.

gg 5<

4.4-8

4.4.7 4.4.7.1 Containment H dro en Monitors Demonstrate that two hydrogen monitors are operable at least daily by verifying that the unit is on or in standby.

4.4.7. 2 Basis:

At least once per quarter perform a channel 'calibration using two sample gases containing'nown concentrations of hydrogen.

The yIl)ainmen is design d for an a

ident pre ure of 60 ps' rj1hi e the rea or is oper tingr the 'nternal en ironment o

the cont nment wil be air at pproximat y atmosph ic ressure.

he maxim temperatu e of the s earn-air mi use at t peak acci ent press e of 60 p g is calc ated to be 286 F.

4 lla Amendment, N0.9

Prior to initial operation, the containment was strength tested at 69 psig and then was leak tested.

The acceptance criterion for this p o

rational leakage rate test was established as O. 1 jo per 24 ho s at 60 ps This leakage rate was believed consistent with the onstruction (2) of the co tainment, which is equipped with independent ak-testable pene ration and contains channels over all containme liner welds, which were in endently leak tested during const ction.

Safety analyses ha e been performed on the bas's of a l.eakage rate of 0.20% per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a 60 psig.

With this le age rate and with minimum containment engineered feguards oper ing (i. e., either 2 filter units and no spray, or 1 filter un'nd 1 sp ay, or no filter units and 2 sprays) the public exposure would be w 1

elow 10 CFR 100 values in the event (3) of the design basis accident.

~Performance of the integra d leakage ate test provides an over-all assessment of potential akage from the ontainment in case of an accident that would p essurize the interior o the containment.

In order to provide realistic appraisal of the inte rity of the contain-ment under ac ident conditions, the test is to be pe ormed without preliminar leak detection surveys or leak repairs, an contain-ment is ation valves are to be closed in the normaL mann The test ressure of 35 psig for the integrated leakage rate test i suf-fi iently high to provide an accurate measurement of the leakage rate and it duplicates the preoperation'al leakage rate test at 35 psig.

4.4-12

he Specif'ication also allows for possible deterioration of the le kage rate between tests, by requiring that the total measure leak e rate be only 75% of the maximum allowable leakage rate The durat'on and methods for the integrated leakage r e test established AHSX N45.4-1972 provide a minimum level accuracy and allow for aily cyclic variation in temperature and thermal radiation.

The equency of the integrated leaka e rate test is keyed to the refue 'ng schedule,. for the react because these tests can best be perf rmed during refueling s utdowns.

Refueling shutdowns are scheduled t approximately o

year intervals.

The specified frequency of in egrated eakage rate tests is based on three major considerations.

Fi t is the low probability of leaks in the liner, because of (a) he use of weld channels to test the leaktightness of the welds ring erection, (b) conformance of the complete containment to 0.1%

pe day leak rate at 60 psig during preoperational test ng, and (c) ab ence of any significant stresses in the liner du ing reactor operatio Second is the more frequent testing, at e full.accident pressure of those portions of the containment nvelope that are most likely o develop leaks during reactor o eration (penetrations and isolati valves) and the low value 0.60 La) of the total leakage that is ecified as acceptable.

Third is the tendon stress surveillance prog

, which provides assurance than an important part of the str tural integr' of the containment is maintained.

Amendment No.

54 4.4-13

The b

s for specification of a total leakage of 0.60 La penetrations isolation boundaries, is that only a portion of e

allowable integrate akage rate should be from those urces in order to provide assurance t t the integrated l age rate would remain within the specified limits ri

. he intervals between integrated leakage rate tests.

ause m

leakage during an integrated leak rate test oc rs though penetration d isolation

valves, and )because most genetic'ations and isolation va es a

smaller leaka rate would result, from an integrated leak test than from a cal test, adequate assurance of maintaining the integrated leakage rate within the specified limits is provided.

he limiting leakage rates from the Recirculation Heat Removal Systems are judgement values based primarily on assuring that the components could operate without mechanical failure for a period on the order of 200 days after a design basis accident.

The test Amendment No.

54 4.4-14

The pre-stress confirmation test provides a direct measure of the load-carrying capability of the tendon.

Xf the surveillance program indicates by extensive wire breakage oz behavin tendon stress relation that the pre-stressin t,

d g

en ons are not The s e

e aving as expected, the situation will be evaluated imm d' 1

e xa e y.

the situ p cified acceptance criteria are such as to alert ation well before the tendon load-carrying capability would a

en ion to mi ht deteriorate to a point that failure during a desi b 'd a

esign asis accident could be mig be possible.

Thus the cause of the incipient d t e erzora eon shut do e evaluated and corrective action studied without d t d wn the reactor.

The contai.nment is provided with two ou nee o

readily removable tendons that might be useful to such a study.

Xn addition, t)ere are 40 tendons, each containing a removable wire which will b'e used to monitor for possible corrosion effects.

Operabxlzty of the containment isolation boundaries ensures that the containment atmosphere will be isolated from th t 'd nment in the event of a release of radioactive material to the containment atmosphere oz pressurization of the containment.

Performance of cycling tests and verification of isolation times associated with automatic containment isolation valves a ves is covered y

mp an Valve Test Program.

Compliance with Appendix J to 10 CFR 50 is addressed under local leak testing requirements.

References:

(2)

UFS Sectio 3.1.2.2.7 FSAR Se ion 6.2.6 (3)

UFS Section 1.6.4.3 N~ ~wc,h (4)

UFSAR Section 6.3.3.8 (5)

UFSAR Table 15.6-9 (6)

FSAR Page 5.1.2-28 (7)

North-American-Rockwell Report 550-x-32, Reliability Handbook, February 1963.

Autonetics (8)

FSAR Page 5.1.2-28

(

Amendment No.

54 4.4-17

Attachment III Marked Up Copy ofImproved Technical Specifications Submitted in LAR Dated May 26, 1995 to Reflect NRC Implementation Guidance Included Pages:

Attachment C 1.1-3 3.6-2 3.6-7 3.6-16 5.0-16 B 3.0-14 B 3.6-2 B 3.6-3 B 3.6-4 B 3.6-6 B 3.6-7 B 3.6-9 B 3.6-15 B 3.6-16 B 3.6-34 B 3.6-35

Definitions 1.1 II

1. 1 Definitions (continued)

K AVERAGE DISINTEGRATION ENERGY E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies (in HeV) per disintegration for non-iodine isotopes, with half lives > 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

L.

LEAKAGE The m

imum allo ble prima containm t leakage rate L., shall e 0.2% of primary co tainment ai wei t per da

. at the c culated pe containme p

ssure (P

LEAKAGE from the'CS 'shall be:

a.

Identified LEAKAGE l.

LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or return),

that is captured and conducted to collection systems or a sump or collecting tank; 2.

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 3.

Reactor Coolant System (RCS)

LEAKAGE through a steam generator (SG) to the Secondary System; b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or return) that is not identified LEAKAGE; (continued)

R.E.

Ginna Nuclear Power Plant

1. 1-3 Draft A

Containment 3.6.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.1

-NOTE SR 3.0.2 is not applicable.

Perform required visual examinations and leakage rate testing exce t fo containment air lock and containment mini-e va v testin in accordance wit 1

CFR SP, ppe ix J, s

mo se a

rovers ex ptions MoW~ RoJR. WM4~ P~~

In accordance 0

R50 p endi dif' by appr ved ex ption SR 3.6. 1.2 Verify containment structural integrity in accordance with the Containment Tendon Surveillance Program.

In accordance with the Containment Tendon Surveillance Program R.E.

Ginna Nuclear Power Plant 3.6-2 Draft A

Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1

-NOTES 1.

An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

Phsc,~~c 4O t

2.

Results shall be evaluated against acce tanc

'ria 49 SR 3.6.1.1 n

accordance wjgh 10 CFp 0,

ppendix as gkdified/by apprg4ed exemptions.

3.

SR 3 0.2 is not lic e

'cud

~ a+a.i~~~ 4uxL4a.~

wN~g '&Mo~

Perform required air lock leakage rate testin in accordance wit 10 Ap en ix J, s

mo i ie y appr ed e emptions The acce tance cri eria for ir lock testin are:

In ccordance with 0

CF 50, pp ndix as m

ified y

prove exempt ons a ~

eakage r te for e ch air loc is Z 0.05

, when te ed at P

Leak e rate f r each doo is

< 0 1 L. wh tested at P..

SR 3.6.2.2 Verify only one door in each air lock can be opened at a time.

24 months R.

E. Ginna Nuclear Power Plant 3.6-7 Draft A

Containment Isolation Barriers 3.6.3 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.6.3.4

+44 Mo~~~

\\ m~~ ('ra~~

Perform required leakage rate testing of containment mini-ur e valves with res> ient seals in accordance wit 0

CFR 0,

ppendix as mod ie y

pproved e

mptions.

The acce ance cri eria fo each min'-purge valve th resil't sea is g 0.0 L.

when ested at P..

In accordance wi 10 CF 50, Ap ndix as m dified y

pprove exempt ons.

SR 3.6.3.5 Verify each automatic containment isolation valve that is not locked,

sealed, or otherwise secured in the required position actuates to the isolation position on an actual or simulated actuation signal.

24 months R.E. Ginna Nuclear Power Plant 3.6-16 Draft A

Programs and Hanuals 5.5 5.5 Programs and Hanuals 5.5.14 SFDP (continued) c.

A required system or trains redundant to the inoperable support system(s) or train for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists.

If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.15 Secondar Water Chemistr Pro ram This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation; This program shall include:

a.

Identification of a sampling schedule for the critical variables and control points for these variables; b.

Identification of the procedures used to measure the values of the critical variables; c.

Identification of process sampling points; d.

Procedures for the recording and management of data; e.

Procedures defining corrective actions for all off control point chemistry conditions; and A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative

events, which is required to initiate corrective action.

R.E.

Ginna Nuclear Power Plant 5.0-16 Draft A

Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix I, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss ofcoolant accident, P is 60 psig.

Th) maximum allowable primary containment leakage rate, LatPshall be 0.2%

ofcontainment air weight per day.

Leakage Rate acceptance criteria are:

Containment leakage rate acceptance criterion is < 1.0 L,. During the first plant startup followingtesting in accordance with this program, the leakage rate acceptance criteria are < 0.60 L, for the Type B and Type C tests and c 0.75 L, for Type Atests; b.

Airlock testing acceptance criteria are:

1)

For each air lock, overall leakage rate is z 0.05 L, when tested at z Pand 2)

For each door, leakage rate is < 0.01L, when tested at z P,.

Mini-purge valve acceptance criteria is < 0.05 L, when tested at > P,.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

LCO Applicability 3.0 BASES SR 3.0.2 (continued}

~ tiadthn, 0 t srt~oR sS M~4 assed <~~oQ. so.~~

ar.~WS.

~ Zg. tm~

O-hJgttst t e,~V=~g~~ Zttta4ua.y,

'R 3.o.4 aalu

~" xt-~ ta eSLu~

ta~

~Mt'he 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency.

This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs.

The exceptions to SR 3.0.2 are those.Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.

These exce ~ons are stated in the i dividual S ecifications.

n ample of re SR 3.0 does n

apply is a Sur ance ith a Freq ncy of "in acco ance w' 10 CF 50 A

e dix J a

dified b a

oved e

mptions he re uirements of re ulations e

ence over e

he~S canno n

an of th rlfselves exte a test 'erva sp dified in e regulati s.

herefor there is ote in t Frequency

ating, "SR 3.

is not ap icable."

As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per..." basis.

The 25%

extension applies to each performance after the initial performance.

The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time.

One reason for not allowing the 25%

extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with Refueling intervals) or periodic Completion Time intervals beyond those specified.

R.E.

Ginna Nuclear Power Plant 8 3.0-14 (continued)

Draft A

Containment B 3.6.1 BASES BACKGROUND (continued)

The cylinder wall is connected to sandstone rock located beneath the containment by use of 160 post-tensioned rock anchors that are coupled with tendons located in the cylinder wall.

This design ensures that the rock acts as an integral part of the containment structure.

The concrete containment structure is required for structural integrity of the containment under DBA conditions.

The steel liner and its penetrations establish the leakage limiting boundary of the containment.

Maintaining the containment OPERABLE limits the leakage of fission product radioactivity from the containment to the outside environment to within the limits of 10 CFR 100 (Ref. 3).

SR 3.6. 1. 1 leakage rate requirements comply with 10 CFR 50, Appendix J (Ref.

4 as modified by approved exemptions.

0 p4%~ '6 The isolation devices for the. penetrations in the containment boundary are a part of the containment leak tight barrier.

To maintain this leak tight barrier:

a.

All penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE automatic containment isolation system, or 2.

Closed by OPERABLE containment isolation

barriers, except as provided in LCO 3.6.3, "Containment Isolation Barriers."

b.

Each air lock is OPERABLE, except as provided in LCO 3.6.2, "Containment Air Locks."

c.

All equipment and personnel hatches or doors are closed when the air lock is not being used for entry into and exit from containment.

APPLICABLE The safety design basis for the containment is that the SAFETY ANALYSES containment must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

(continued)

R.E.

Ginna Nuclear Power Plant B 3.6-2 Draft A

Containment B 3.6.1 BASES APPLICABLE SAFETY ANALYSES (continued)

~ 6L4sg2p WELSH LUVP The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA), a steam line break, and a rod ejection accident (REA) (Ref. 5).

In addition, release of significant fission product radioactivity within containment can occur from a LOCA or REA.

In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage.

The containment was originally strength tested at 69 psig (115% of design).

The acceptance criteria for this test was 0. 1% of the containment air weight per day at 60 psig which was based on the construction techniques that were used (Ref. 5).

Following successful completion of this test, the accident analyses were performed assuming a leakage rate of 0.2% of the containment air weight per day.

This leakage rate, in combination with the minimum containment engineered safeguards operating (i.e., either 2 post-accident charcoal filter trains and no containment

spray, 1 post-accident charcoal filter train and 1 containment spray tr ain, or no post-accident charcoal filter trains and 2 containment spray trains) results in offsite doses well within the limits of 10 CFR 100 (Ref. 3) in the event of a DBA.

t Qp +0vt The leakage rate of 0.2% of the contain air weight per day is defined in 10 CFR 50, Appendix J (Ref. 5),

as L.:

the maximum allowable containment leakage rate at the calculated peak containment internal pressure (P.) resulting rom e

The allowable leakage rate represented by L, forms the basis for the acceptance criteria imposed on all containment leakage rate testing..

L. is assumed to be 0.2% per day in the safety analysis at p.'

psig~.

uo Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY.

The containment satisfies Criterion 3 of the NRC Policy Statement.

R.E.

Ginna Nuclear Power Plant B 3.6-3 (continued)

Draft A

Containment 8 3.6.1 BASES (continued)

Op~~ ~

LCO Containment OPERABILITY is maintained by limiting leakage to

.0 L. except prior to entering HODE 4 for the first time ollowing performance of periodic testing erformed

'ccordance with 10 CFR 50, Appendix J.

t that time, the combined Type 8 and C leakage must be

< 0.6 L. on a maximum pathway leakage rate (HXPLR) basis, and the overall Type A leakage must be

< 0.75 L..

At all other times prior to performing as found testing, the acceptance criteria for Type 8 and C testing is < 0.6 L. on a minimum pathway leakage rate (HNPLR) basis.

Containment OPERABILITY is also defined by acceptable structural integrity following a DBA.

Compliance with this LCO will ensure a containment configuration, including personnel and equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.

Individual leakage rates specified for the containment air lock (LCO 3.6.2) and mini-purge valves with resilient seals (LCO 3.6.3) and administrative limits for individual isolation barriers are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J.

Therefore, leakage rates exceeding these individual limits only result in the containment being inoperable when the leakage results in exceeding the acceptance criteria of Appendix J for Type A, 8, and C tests.

APPLICABILITY In MODES 1, 2, 3, and 4, a

DBA could cause a release of radioactive material into containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.

Therefore, containment is not required to be OPERABLE in MODES 5 and 6 to prevent leakage of radioactive material from containment.

R.E.

Ginna Nuclear Power Plant 8 3.6-4 (continued}

Draft A

Containment B 3.6.1 BASES (continued)

SURVEILLANCE REQUIREMENTS

~ Nv+u.xw~

P~fbi'~

SR 3.6.1.1 Haintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test re uirements 1

pen ix e

as mogifi app ve exem ions (Re s.

6 and 7

ai ure o mee air loc an mini-purge va ve wi ress sent seal leakage limits specified in LCO 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of these overall leakage determinations unless their contribution to overall Type A, B, and C leakage causes these limits to be exceeded.

As left leakage prior to entering MODE 4 for the first time following performance of required 10 CFR 50, Appendix J periodic testing, is required to be

< 0.6 L. for combined Type B and C leakage on a HXPLR basis, and

< 0.75 L. for overall Type A leakage (Ref. ~

At all other times between the required leakage

tests, the acceptance criteria is based on an overall Type A leakage limit of~1.0 L..

This is maintained by limiting combined Type 8 and C leakage to < 0.6 L. on a HXPLR basis until performance of as found testing.

At~.0 L., the offsite dose consequences are bounded by the assumptions of the safet anal sis.

SR Fre uencies are as required by en i as if e roved ex@

io

hus, w ic ows requen ex ensions oes no ply.

se per o ic es ing requiremen s veri y a

e containment leakage rate does not exceed the leakage rate assumed in the safety analysis.

SR 3.6.1.2 This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program.

Testing and Frequency are generally consistent with the recommendations of Regulatory Guide 1.35 (Ref.

except that tendon material tests and inspections are not required R.E.

Ginna Nuclear Power Plant 8 3.6-6 (continued)

Draft A

Containment B 3.6.1 BASES (continued)

REFERENCES l.

Atomic Industry Forum, GOC 10 and 49, issued for comment July 10, 1967.

2.

UFSAR, Section 3.8. 1.

3.

10 CFR 100.

, op~~8.

4.

10 CFR 50, Appendix J.

5.

UFSAR, Section 6.2.

tter fro

. L. Ziemann, RC, to L. D. +Kite,

RGEE,

Subject:

Amendment No.

7 to Provisiyfal Operating Licens

'ated Harch 1978.

7.

Let er from O. H.

C utchfield, g

, to J.

E. Haier R

E,

Subject:

/ompletion o+Appendix J Review ated Hay 6, 1 81.

e ulat Guide

-10 Q)9t Regulatory Guide 1.35, Revision 2.

Letter from J. A. Zwolinski, NRC, to R.

M. Kober, RG&E,

Subject:

"Safety Evaluation Containment Vessel Tendon Surveillance Program," dated August 19, 1985.

~~~~~ 4oiMXi~R~ ~is.w m~g P~4'~~~~ -Q~h-4Pwo~

cP-le c.c~

~> " ~isis~ 0, R.E.

Ginna Nuclear Power Plant B 3.6-7 Draft A

Containment Air Locks B 3.6.2 BASES (continued)

APPLICABLE SAFETY ANALYSES The DBAs that result in a release of radioactive material within containment are a loss of coolant accident and a rod ejection accident (Ref. I).

In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage.

The containment was designed with an allowable leakage rate of 0.25 of containment air weight per day (Ref. I).

Th's

,+'8 leakage r ate is defined in 10 CFR 50, Appendix J ef. 2),

as L. - O.Bo of containment air weight per day, the maximum allowable containment leakage rate at the calculated peak containment internal pressure P. =~ psig followin gk This allowable leakage rate forms e

asis for the acceptance criteria imposed on the SRs associated with the air locks.

The containment air locks satisfy Criterion 3 of the NRC Policy Statement.

LCO The equipment hatch and personnel hatch containment air locks form part of the containment pressure boundary.

As part of containment, the air lock safety function is related to control of the containment leakage rate following a DBA.

Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

Each air lock is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the 10 CFR 50, Appendix J Type 8 air lock leakage test (i.e.,

SR 3.6.2. 1),

and both air lock doors must be OPERABLE such that they can remain closed with leakage within acceptable limits following a DBA.

The interlock allows only one door of an air lock to be opened at a time.

This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE.

Closure of a single 'door in each air lock is sufficient to provide a leak tight barrier following postulated events.

Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.

R.E.

Ginna Nuclear Power Plant B 3.6-9 (continued)

Draft A

Containment Air Locks B 3.6.2 BASES ACTIONS (continued)

D.l and D.2 If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a

HODE in which the LCO does not apply.

To achieve this status, the plant must 'be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURYEILLANCE REQUIREMENTS SR 3.6.2.1 Maintaining containment air locks OPERABLE requires corn 1'~ca wi.th& e-~k ~

st re uirements of 0

C 50 Appen x 0 R f. 2),

s modif d b a

ove xe tio Ref 3

is re ec s t e leakage rate testing requirements with regard to air lock leakage (Type B

leakage tests).

The acceptance criteria were established based on industry experience.

The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate.

The SR has been modified byg~ Notes.

Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

This is co'nsidered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA.

Note 2 requires that the 'results of this SR be evaluated against. the acceptance criteria of This ensures that air lock leakage is properly accounted for in determinin the overall containment leaka e rate.

ote a es at SR 3.0.

w ic a

ows requency ex e sons) does ot apply s' the Freque is required Appen

'R

. 2),

as ified by app ed exemption (Ref.

3

~ san%~ PA)g~.

(continued)

R.E.

Ginna Nuclear Power Plant B 3.6-15 Draft A

Containment Air Locks 8 3.6.2 BASES SURVEILLANCE REQUIREHENTS (continued)

SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock.

Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment

pressure, closure of either door will support containment OPERABILITY.

Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment.

Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur.

Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is only challenged when the containment airlock door is opened, this test is only required to be performed once every 24 months.

The 24 month Frequency is based on engineering judgment and is considered adequate in view of other indications of door and interlock mechanism status available to operations personnel.

REFERENCES 1.

UFSAR, Section 6.2.1.1.

2.

10 CFR 50, Appendix J.

C) pram E 3.

Le er from D L. Ziemann, RC, to L. D.

ite, RGE E, bject:

endment No.

to Provisi 1 Operating License," dated Harch 1978.

R.E.

Ginna Nuclear Power Plant B 3.6-16 Draft A

Containment Isolation Barriers 8 3.6.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.3.2 (continued)

The Note applies to containment isolation barriers located in, high radiation areas and allows these devices to be verified closed by use of administrative means.

Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons.

t Therefore, the probability of misalignment of these isolation barriers, once they have been verified to be in their proper position, is small.

SR 3.6.3.3 Verifying that. the isolation time of each automatic containment isolation valve is within limits is required to demonstrate OPERABILITY.

The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.

The isolation time and Frequency of this SR are in accordance with the Inservice Testing Program.

SR 3.6.3.4 Qg~~ +

For containment mini-pur e alves with resilient seals, additional leakage rate esting beyond the test requirements of 10 CFR 50, Appendix J, is required to ensure OPERABILITY.

Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types.

Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the outside environment),

a leakage acceptance criteria of < 0.05 L, when tested at Z P. is specified for each mini-ur e

isolation valve with resilient seals.

e Fre uency of esting is specs se i

pen sx as mo ie y ap ve ~empta on efs.

6 d 7).

(continued)-

R.E.

Ginna Nuclear Power Plant B 3.6-34 Draft A

Containment Isolation Barriers 8 3.6.3 BASES SURVEILLANCE REQUIREHENTS (continued)

SR 3.6.3.5 Automatic containment isolation valves close on a

containment isolation signal to prevent leakage of radioactive material from containment following a DBA.

This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a, containment isolation signal.

This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES Atomic Industry Forum GDC 53 and 57, issued for comment July 10, 1967.

2.

Branch Technical Position CSB 6-4, "Containment Purging During Normal Operation."

3.

UFSAR, Section 6.2.4 and Table 6.2-15.

4.

10 CFR 50, Appendix A, GDC 55, 56, and 57.

5.

Ginna 'Station Procedur A-3.3.

6.

Le er from D. L.

niemann, NRC, to L

. White, RG&E, bject:

"Amen( ent No.

17 to Pro sional Operating License," dat d Harch 28, 1978.

Letter frp D. H. Crutchfie NRC, to J.

E.

Ha

RGRE, S traject:

"Completi of Appendix J Rev',"

dated ay 6, 1981.

R.E.

Ginna Nuclear Power Plant B 3.6-35 Draft A