Text

Proposed Tech Specs 3/4.6 Re Containment Sys,Per App J, Option B
	ML20107G351
Person / Time
Site:	Summer
Issue date:	04/16/1996
From:	; SOUTH CAROLINA ELECTRIC & GAS CO.
To:	;
Shared Package
ML20107G344	List: ML20107G351; RC-96-0109, TS Change TSP 950011 to License NPF-12,revising TS 3/4.6, Containment Sys Bases 3/4.6 & TS 6.8, Administrative Controls - Programs & Procedures, to Adopt Implementation Requirements of 10CFR50,App J,Option B;
References
	NUDOCS 9604230196
	Download: ML20107G351 (29)
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_

Document Control Desk

AttachmentI TSP 950011 RC 96-0109

' Page 1 of 2 SCE&G - Explanation of Changes Summary AII" PAGE Description of Change Reason for Change g cti i-2 1.7.d 1

Reworded "d." to reference the These limits have been relocated to TS Containment Leakage Rate Testing 6.8.4 under new segment "g." which has

]

Program.

been added to describe the VCSNS Containment Leakage Rate Testing Program in accordance with the prescribed guidelines of NEI and WOG, 3/4 6-1 4.6.1.1.c 1

Removed from TS and relocated to the This SR is addressed in the VCSNS VCSNS Containment Leakage Rate Containment Leakage Rate Testing Testing Program.

Program in accordance with NEI 94-01 and WOG guidelines for implementation t

of Option B of Appendix J.

3/462 3.6.1.2 i

Removed "a." and "b." and reworded Limits have been relocated to TS 6.8.4 3.6.1.2 to state " Containment leakage under new segment "g." which has been rates shall be limited in accordance with added to describe the VCSNS the Containment Leakage Rate Testing Containment Leakage Rate Testing Program."

Program in cccordance with the prescribed guidelines of NEI and WOG.

3.6.1.2 2

Reworded ACTION to address Mode Based on WOG guidelines for W PWRs, restraints and completion times when NEI and NRC mooel technical leakage limits are exceeded.

specification to implement Option B.

J 4.6.1.2 3

Removed "a." and reworded 4.6.1.2 to This SR is addressed in the VCSNS 4.6.1.2.a state "The containment leakage rates Containment Leakage Rate Testing shall be demonstrated at the intervals Program which implements the derived through the Containment requirements of Reg. Guide 1.163 and Leakage Rate Testing Program and shall ANSI /ANS-56.81994.

be determined per the program criteria.".

3/4 6-3 4.6.1.2.b 1

Removed from TS and relocated to the Relocation is based on model technical 4.6.1.2.c VCSNS Containment Leakage Rate specifications to implement Option B.

4.6.1.2.d Testing Program.

These SRs are addressed in the VCSNS 4.6.1.2.0 Containment Leakage Rate Testing 4.6.1.2.f Program.

4.6.1.2.g 3/4 6-4 3.6.1.3 1

Removed "a." and "b." and reworded Limits have been relocated to TS 6.8.4 3.6.1.3 to state "Each reactor building air under new segment "g." which has been lock shall be OPERABLE with both added to describe the VCSNS doors closed except when the air lock is Containment Leakage Rate Testing being used for normal transit entry and Program in accordance with the exit through the containment,then at prescribed guidelines of NEI and WOG.

least one air lock door shall be closed."

3/4 6-5 4.6.1.3.a 1

Reworded "a." to reference the Rewording is based on model technical Containment Leakage Rate Testing specifications to implement Option B.

Program, 4.6.1.3.b 2

Removed "b." from TS.

This SR is addressed in the VCSNS Containment Leakage Rate Testing Program and limits have been relocated 9604230196 960416 to TS 6.8.4.g.

PDR ADOCK 05000395 P

PDR

Document Control Desk

AttachmentI TSP 950011 RC-96-0109

Page 2 of 2 SCE&G -- Explanation of Changes Summary Affeted U

PAGE Description of Change Reason for Change gg 9n 3/4 6 5 4.6.1.3.d 3

Removed "d" from TS and relocated to Limits have been relocated to TS 6.8.4.g TS 6.8.4.g and the VCSNS Containment and SR relocated to the VCSNS Leakage Rate Testing Program.

Containment Leakage Rate Testing Program in accordance with the prescribed guidelines of NEI and WOG.

Footnote 4

Removed # and

Footnotes.

Footnotes have been incorporated into TS 6.8.4.g and the VCSNS Containment Leakage Rate Testing Program.

3/4 6-10 4.6.1.6.3 1

Reworded to reference the Containment Inspection frequencies are addressed in Leakage Rate Testing Program.

the VCSNS Containment Leakage Rate Testing Program which implements the requirements of Reg. Guide 1.163 and ANSI /ANS-56.8-1994.

3/4611 4.6.1.7.3 1

Reworded to reference the Containment Testing frequencies are addressed in the Leakage Rate Testing Program and VCSNS Containment Leakage Rate incorporate the frequency allowed by Testing Prog:' im which implements the Reg. Guide 1.163.

requirements of Reg. Guide 1.163 and ANSI /ANS-56.8-1994.

B 3/4 6-1 B

1 Reworded to reference the Containment Limits have been relocated to TS 6.8.4.

3/4.6.1.2 Leakage Rate Testing Program and under new segment"g." which has been incorporate changes presented in model added todescribe the VCSNS technical specification to implement Containment Leakage Rate Testing Option B.

Program in accordance with the prescribed guidelines of NEI and WOG B

2 Reworded to remove reference to leakage Leakage rate limits have been relocated l

3/4.6.1.3

limits, to TS 6.8.4 under now segment "g."

which has been added to describe the l

VCSNS Containment Leakage Rate Testing Program in accordance with the prescribed guidelines of NEI and WOG.

j B 3/4 6-3 B

1 Reworded to reference the Containment Testing frequencies are addressed in the 3/4.6.1.7 Leakage Rate Testing Program.

VCSNS Containment Leakage Rate Testing Program which implements the requirements of Reg. Guide 1.163 and ANSI /ANS 56.8-1994.

6-12b 6.8.4.g 1

Added"g."under PROCEDURES AND New segment which has been added to PROGRAMS.

describe the VCSNS Containment Leakage Rate Testing Program in accordance with the prescribed guidelines ofNEI and WOG for implementation of Appendix J, Option B.

612c 6.8.4.g i

New page.

New page to accommodate new 6.8.4.g.

l

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

All penetrations required to be closed during accident conditions a.

are either:

1)

Capable of being closed by an OPERA 8LE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for

~

4:2 valves that are open under administrative control as permitted

7 by Specification 3.6.4, I

b.

All equipment hatches are closed and sealed, Each air lock is in compliance with the requirements of c.

Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 2.5.1.2,-and

(. 8.'l.3 The sealing mechanism associated with each penetration (e.g.,

e.

welds, bellows, or 0-rings) is OPERASLE.

t CONTROLLEO LEAKAGE i

1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of

)

movement of a component to a safe conservative position.

j CORE OPERATING LIMITS REPORT 1.9a The CORE OPERATING LIMITS REPORT (COLR) is the unit specific document that provides core operating limits for the current operating reload cycle.

The cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.11.

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

-00SE E0VIVALENT I-131

)

1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131 I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance factors for Power and Test Reactor Sites."

SUPMER - UNIT 1 1-2 Amendment No. Eg,110

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when; All penetrations required to be closed during accident conditions a.

are either:

1)

Ca 3able ofbeing closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.4, b.

All equipment hatches are closed and sealed, Each air lock is in compliance with the requirements of c.

Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 6.8.4.g, and The sealing mechanism associated with each penetration (e.g.,

e.

welds, bellows, or 0 rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 4

1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.9a The CORE OPERATING LIMITS REPORT (COLR)is the unit specific document that provides core operating limits for the current operating reload cycle.

The evele specific core operating limits shall be determined for each reload cycly.a accordance with Specification 6.9.1.11. Plant operation within the.( $0ratire f.imits is addressed in individual specifications.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration ofI 131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture ofI-131, I 132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

SUMMER -UNIT 1 12 Amendment No. 88,110,

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT

{

CONTAINMENT INTEGRITY l

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 one hour 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 .

SURVEILLANCE REQUIREMENTS 1

t 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 OPERAB LE containment automatic isolation valves and required to be closed during accident conditions are cbsed by valves, blind flanges or deactivated automatic valves secured in tb eir positions, exce,pt for valves that are open under administrative control as permitted by Specification 3.6.4.

b.

By verifpg that each containment air lock is in compliance with the requirements ofSpecification 3.6.1.3.

c'

^"----^^---#--'---^'--"^--kt^^^^"'"--

Mp

'" '~Eo'n't airIIx52s,"fo~pe~n7 folio'wiYg~a~ '~ 't i

~

verifvm,vleak rate e seal with gas

".o psig) and test,b g that when the meas age rate for these seals is De LETE added to the leaka etermined pur Soecification 4.6.1.

o er Type B and C penetrations, the c

._. age rate is lere er OMO L,.

1

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

SUMMER - UNIT 1 3/461 Amendment No.M119

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 one hour or be in atleast 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 .

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 DPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valvos, blind flanges or deactivated automatic valves secured in their positions, exce,pt for valves that are open under cdministrative control as permitted by Specification 3.6.4.

b.

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.2.3.

c.

Deleted.

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

SUMMER - UNIT 1 3/4 6-1 Amendment No.110,119,

i 2 -.

j 1

CONTAINMENTSYSTEMS r

CONTAINMENT LEAKAGE

,{

LIMITING CONDITION FOR OPERATION i

3.6.1.2. Containmentleakage rates shall belimited4ei In ac'or'd'ac* du th* C'*4'. "*#

P Leakage Rd f*M% rogram j

A- ^"erell i"te7*'ed !^* lay Pete ef:

=

i 1.

Les r equal to La,0.20 percentby weight o ontainment air per 24 o

,53.5 psig, or i

2.

Less than or equal to v we' ht of the containment air per 24 ho re uced pressure o sig.

e, b.

med leakage rate ofless than 0.60 L v&ee subject te _ /;e E rd C te&,-h::;a for all penetrations 1

r:::==:d t: P,.

APPLICABILITY: MODES 1,2,3 and 4.

^ g 3 gopsy

}

W;%;n j hour iOId6 d

. ce s uro *M ACTION:

j w; n. n.vt G h*gry h. ors.

g +s m tw. Otosias i

With either(2) the measure exceedingM6L ar 0.75 L,,4overallintegrated containmentleaka,ge rate T

r e p!!ceble, er 'H6 S: mer=: :=hin:d f,o le-' e e re fer el enetr2tiare;end v&er erh'eet te T ' rate toles n - a A o '--'-

i L-~ ^ 80 Efestore the overallinteg5 L

he bl i

ateileakag~e ereer equal to 6.75 La er !?"" " " " "r ^'4 " *' t^ ^ -

L e e, and the

}

combined leakage rate for all penetrations subject toYype B and C tests to le ss than.0.60 La prior to increasing the Reactor Coolant System temperature 4o %y ~

a >ove 2007.

4 2

i SURVEILLANCE REQUIREMENTS J

4.6.1.2 The containment leakage rates shall be demonstrated at the f:'JI-test eched"le --d hel!be deter-: edin e^ #crre: id Se:rit -ic:p::i i

in AppendL J ef10 01". ",0.

3 4

5

- inierw ah d.Aved %<e@ tb. cont ='i nm nt Lc.a k. 3e g %.

ddr*rm la*d pu W pegran c.r'itar,$ 4 -

Teshg Pre 3 ram W shalt h*

d SUMMER ~ UNIT 1 3/462 Amendment No./A,///,

126

i CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2. Containment lenkage rates shall be limited in accordance with the Containment Leakage Rate Testing Program.

APPLICABILITY: MODES 1,2,3 and 4.

)

i ACTION:

With the measured overall inte gated containment leakage rate exceeding 1.0 La, within I hour initiate action to be in 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 . Restore the overall integrated leakage rate to less than or equal to 0.75 La and the combined leakage rate for all penetrations subject to Type B and C tests to less than or equal to 0.60 La 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 intervals 4

derived through the Containment Leakage Rate Testing Program and shall be j

determined per the program criteria.

4 l

SUMMER - UNIT 1 3/46-2 Amendment No. 97,119,126,

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (continued) b.

Deleted 1

c.

e accuracy of each Type A test shall be verified by a suppleme tes ~ accordance with Appendix J.

l d.

Type B an ests shallbe conducted with gas at P

.5psig)in accordance wi pendix J except for tests invo g:

gg4l3 1.

Air locks.

in 2.

Purge sup lv and exhaust ation valves with resilient g

material a'Is.

D e.

Purge supply and e ust isolation valves wi resilient material seals shall be te and demonstrated OPERAB er Surveillance Requireme

.6.1.7.3.

f.

Air s shall be tested and demonstrated OPERABLE per eillance uirement 4.6.1.3.

g.

Se prc'risiens of SpecL9 cation 4.0.2 re net 2pplicab!c.

Y e

Amendment No.g,126 SUMMER - UNIT 1 3/4 6 3

i 1

i i

i

. THIS PAGE INTENTIONALLY LEFT BLANK Page 3/4 6-3 has been deleted.

1 4

i i

i a

b 4

1 SUMMER -UNIT 1 3/46-3 Amendment No. 119,126,

=

or p

.-= a

-r..,

,, -... + -. - -

c

. - >,.u-..

-,w e

y y

u-c -

1

_ CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each reactor building air lock shall be OPERABLE with)(

-Eh Lgoth doors closed escebt when the air lock is beinken a used for normal transit entry and exit ough the containment,t airlock door shall be closedg k

A" ^"^-a" ab !eh !ea.hage rate ef!cc than er cque.1 tc 0.10 La at Pa,53.5 p:ng.

APPLICABILITY: MODES 1,2,3 and 4.

t ACTION:

With one reactor building air lock door inoperable:

a.

1.

Maintain at least the OPERABLE air lock door closed and either restore the inoaerable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or loc s the OPERABLE air lock door closed.

2.

Operation may then continue until peiformance of the next required overall air lock leakage test provided that the OPERABLE air lock door is venfied to be locked closed at least once per 31 days.

3.

Otherwise, be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , i

4.

The provisions of Specification 3.0.4 are not applicable.

b.

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

SUMMER. UNIT 1 3/4 6-4 Amendment No.119

- = - -.

4 4

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each reactor building air lock shall be OPERABLE with both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed.

1 9

APPLICABILITY: MODES 1,2,3 and 4.

ACTION:

a.

With one reactor building air lock door inoperable:

l.

Maintain at least the OPERABLE air lock door closed and either restore the ino aerable air lock door to OPERABLE status within i

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or locLc the OPERABLE air lock door closed.

i 2.

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

3.

Otherwise, be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4.

The provisions of Specification 3.0.4 are not applicable, b.

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

b 4

SUMMER - UNIT 1 3/4 6-4 Amendment No.119,

4 CONTAINMENTSYSTEMS SURVEILLANCE REQUTREMENTS 1

4.6.1.3 Each reactor building air lock shall be demonstrated OPERABLE:

W: 72 her fellev ing ::ch elecing, leas::::pt ;-hen de ci '- '- b;

{

a.M Ey ver;fq/$

being usea entries, then at

ours, g

byverifyingthat the se feco), m, 4;,,

ess than or equalto e

e J th 0

p" ^r M"2 t0 3 0 C's r Ot.^0 t, CCuSC -,% Q e etween thedoorse-

.His Coatala et P

o---

L.cak.43. Rd*

f b.

1-keh k

= p t :t; at net k :: S an "a, y g +M.,

Mconduct'-[ vaa!!

verifying the overall air lock leakage rat

p,3 withinitslinu.

A 1.

Atleast once per 6 mon Dela.fcd i

2.

Prior to s ing CONTAINhENT IN r.

'when enance has been performed on the air lock a

d affect the air keh caling pability.*

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

airlock can be opened at a time.

d.

At kact :::: per S men &:#, by v -ifying ^lume hetvc :: Sclanhch::1

-*^^-^~11^,t.-~

.a, is kr Sa; a equal t: 0.01 h vehen S vc a

Daldg

^-a

-^-1 : prescrized te genter the er eq--! te 8.0 peig fer p

e + 1 - ~ + o ~ : m a --

- f 5 prerbien cfSp::ifkatic; 4.0.2 re ::t appli=bk.

_. v _ _

u...__.A.

A.y y _-- R _ T.c 1. A.r9 D.D.a n..

y I

. v.

m 1

SUMMER UNIT 1 3/46-5 Amendment No.119

l 1

i 4

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

By verifying leakage rates in accordance with the Containment l

a.

Leakage Rate Testmg Program.

l b.

Deleted.

l c.

At least once per six months by verifying that only one door in each airlock can be opened at a time.

d.

Deleted.

l 1

l l

I l

SUMMER - UNIT 1 3/46-5 Amendment No.119,

l CONTAINMENT SYSTEMS l

~

^

SURVEILLANCE REQUIREMENTS (Continued)

If the inspections performed at 1, 3, and 5 years indicate no abnormal degradation of the tendon system, the number of sample tendons may be reduced to 3 dome, 3 vertical and 3 hoop for subsequent inspections.

Upon the comple-tion of the five year, inspection, the results of the first three inspections shall be evaluated to determine if an abnormal condition is evident fo tendon system.

with their Base Values and Normalizing Factors will be specif subsequent inspections.

4.6.1.6.2 At the same inspection frequency as the tendons, the structural integrity of the end anchorages of all tendons inspected pursuant to Specifica-tion 4.6.1.6.1 and the adjacent concrete surfaces shall be determined by a visual inspection and verifying that no abnormal material or structural behavior is evident u acco. d wee. w.% the Cc^b"at l O. Me Tesb S

3 Pr*5 cm 4.6.1.6.3 8.t the 3 e 4aspectic" 'aeque cy es +*e Tyra 6

-an*=4

-a

+

+ete tae','the structural integrity of the exposed accessible interior and1==k=ce exterior surfaces of the containment shall be determined p-fr-to each j

Typa a cont 2ir eat leak 2;e eta tett (Specirat4 r d.S.I.2) by a visual inspection of these surfaces and verifying that no abnormal material or structural behavior is evident.

i SUMMER - UNIT 1 3/4 6-10

_ CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

If the inspections performed at 1,3, and 5 years indicate no abnormal degradation of the tendon system, the number of sample tendons may be reduced to 3 dome,3 vertical, and 3 hoop for subsequent inspections. Upon the comple-tion of the five year inspection, the results of the first three inspections shall be evaluated to determine if an abnormal condition is evident for the tendon system. Based on the conclusions of this evaluation, the sample tendons with their Base Values and Normalizing Factors will be specified for all subsequentlocations.

4.6.1.6.2 At the same inspection frequency as the tendons, the structural integrity of the end anchorages of all tendons inspected pursuant to Specifica.

tion 4.6.1.6.1 and the adjacent concrete surfaces shall be determined by a visual inspection and verifying that no abnormal material or structural behavior is evident.

4.6.1.6.3 In accordance with the Containment Leakage Rate Testing Program, the structural integrity of the exposed accessible interior and exterior surfaces of the containment shall be determined by a visual inspection of these surfaces and verifying that no abnormal material or structural behavior is evident.

1 SUMMER-UNIT 1 3/4 6-10 Amendment No.

CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERAIION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:

Each 36-inch containment purge supply and exhaust isolation valve a.

shall be sealed closed.

b.

The 6-inch containment purge supply and exhaust isolation valves may be open for less than or equal to 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months per 365 days.

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

ACTION:

With a 36-inch containment purge supply and/or exhaust isolation a.

valve (s)openornotsealedclose close and/or seal close the open valve (s)orisolatethepenetration(s)within4 hours,otherwisebe in at least HOI 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 .

b.

With a 6-inch containment purge supply and/or exhaust isolation valve (s) open for more than 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months per 365 days, close the open 6-inch valve (s) or isolate the penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 .

c.'

With a containment purge supply'and/or exhaust isolation valve (s) l having a measured leakage rate exceeding the limits of Surveillance Requirements 4.6.1.7.3, restore the inoperable valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise be in at least HOI STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUIDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENIS 4.6.1.7.1 Each 36-inch containment purge supply and exhaust isolation valve shall be verified to be:

Closed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a.

b.

Sealed closed at least once per 31 days.

4.6.1.7.2 Ihe cumulative time that the 6-inch purge supply and exhaust isolation l

vilves have been open during the past 365 days shall be determined at least once per 7 days.

30

4. 6.1. 7. 3 At least once per-6-months cr. : STf.CCERED TEST S?.5!! each containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by '/cr fying that th: ::::ured !:: Rage r:t i: h:: +5Le i

er cau 1 t: 0.05 L, ' ten prcruhtd t: P,.buwbud.%A Waked N Refe. "Tob S r-.

3 3

SUMMER - UNII 1 3/4 6-11

~.

i CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM T

LIMITING CONDITION FOR OPERATION 1

3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:

i a.

Each 36 inch containment purge supply and exhaust isolation valve i

shall be sealed closed.

b.

The 6-inch containment purge supply and exhaust isolation valves may be open for less than or equal to 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months per 365 days.

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

3 ACTION:

With a 36-inch containment pur a.

valve (s) open or not sealed close,ge supply and/or exhaust isolation close and/orseal close the open valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 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 i

SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With a 6-inch containment purge supply and/or exhaust isolation i

valve (s) open for more than 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months per 365 days, close the open 6-inch valve (s) or isolate the penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months otherwise t

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 t

SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With a containment purge supply and/or exhaust isolation valve (s) c.

having a measured leakage rate exceeding the limits of Surveillance Requirements 4.6.1.7.3, restore the inoperable valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; 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 .

SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each 36-inch containment purge supply and exhaust isolation valve shall be verified to be:

a.

Closed atleast once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

Sealed closed at least once per 31 days.

t 4.6.1.7.2 The cumulative time that the 6-inch purge sup 31y and exhaust isolation valves have been open during the past 365 days shall be tetermined at least once per 7 days.

4.6.1.7.3 At least once per 30 months each containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE in accordance with the Containment Leakage Rate Testing Program.

4 SUMMER -UNIT 1 3/4 6-11 Amendment No.

i

..m

-..m.

k 1

3/4.6 CONTAINMENTSYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY a

i i

Primary CONTAINMENTINTEGRITY ensures that the release ofradioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restnc.

tion in conjunction with the leakage rate limitation, willlimit the site boun, dary radiation doses to within the limits of10 CFR 100 during acciden conditions.

3/4.6.1.2 CONTAINMENT LE AKAGE The limitations on containment leakage rates (including those used in demonstrating a 30 day water seal) ensure that the total containment leakage i

volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P,, As an added conservatism, the measured overallinte-p% Lated leakage rate is further limited to less than or equal to 0.75 La+F-possible' u r.pp!!chh, during performance of the $eriodic tes S

ss

\\

The surveillance testing for measurin the req.:ir:=:d of Appen L "J" of10 CF,..gleakage rates-are-consistent with d

50. c.d.'nm.a Ltak=3. Ret Tutfr~

i 3/4.6.1.3 RE ACTOR BUILDING AIR LOCKS The limitations on closure -> 1-t- ~+e for the reactor building air locks are required to meet the restrictions on CONTAINMENT INTEGRI'IT and containment leak rate. Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the interwls between air lock leakage tests.

t i

a

l SUMMER - UNIT 1 B 3/4 61 Amendment No. g,j /,126 d

9 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release ofradioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restric-tion, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of10 CFR 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE

\\

The limitations on containment leakage rates (including those used in demonstrating a 30 day water seal) ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P. As an added conservatism, the measured overallinte-p grated leakage rate is further limited to less than or equal to 0,75 La during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

i The surveillance testing for measuring leakage rates is consistent with the Containment Leakage Rate Testing Program.

3/4.6.1.3 REACTOR BUILDING AIR LOCKS The limitations on closure for the reactor building air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between airlock leakage tests.

SUMMER-UNIT 1 B 3/4 6-1 Amendment No, 44,97,126,

i CONTAINMENT SYSTEMS 1

BASES 3/4.6.1.7 REACTOR BUILDING VENTILATION SYSTEM The 36-inch containment purge supply and exhaust isolation valves are required to be closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident.

Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will nnt be released via the containment purge system. To provide assurance that the 36-inch valves cannot be inadvert-ently opened, they are sealed closed in accordance with the Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevent power from being spplied to the valve operator.

The use of the containment purge lines is restricted to the 6 inch purge supply and exhaust isolation valves since unlike the 36 inch valves the 6 inch valves will close during a LOCA or steam line break accident and therefore the site boundary dose guidelines of 10 CFR 100 would not be exceeded in the event of an accident during purging operations.

Periodic leakage integrity tests uith a m 4-"- 2?'c"25'e 'eab?ge " ate for be rebmadprevide ear'y 4"dic?+4aapurge supply and exhaust isolation valves with resilient m a# sea' de;reditier 2nd "4 allcw the oppertun!ty f -

..Wm repci-befer gre:: 12kage f ai!Uaer deva'ep. The 0.50 L 10:k:g

'4-it th H N,ob ret bc exceeded 'haa +he 'e=kega *=+ae da+=" 4aed by tha aikage 4ateg*ity-W W E d 8-te:t: cf th::: V:lv:: Or: 2dded to the previe":1y deter-fr.ed tet:1 f P-a44-m i:ag

. u....,.

a e,..,e+.,+4me,

, m s 4,,.+ --

sr- - -- - - ------

+,+,-o o,ma e

+e,*,

r-3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS i

3/4.6.2.1 REACTOR BUILDING SPRAY SYSTEM The OPERABILITY of the reactor building spray system ensures that reactor building depressurization and cooling capability will be available in the event of a steam line break.

The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.

The reactor building spray system and the reactor building cooling system are redundant to each other in providing post accident cooling of the reactor building atmosphere.

However, the reactor building spray system also provides a mechanism for removing iodine from the reactor building atmosphere and therefore the time requirements for restoring an inoperable spray system to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

SUMMER - UNIT 1 B 3/4 6-3

l CONTAINMENT SYSTEMS a

BASES

.y _

3/4.6.1.7 REACTOR BUILDING VENTILATION SYSTEM i

The 36-inch containment purge supply and exhaust isolation valves are required to be closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident.

Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system. To provide assurance that the 36-inch valves cannot be inadvert-ently opened, they are sealed closed in accordance with the Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevent power from being supplied to the valve operator.

The use of the containment purge lines is restricted to the 6 inch purge i

supply and exhaust isolation valves since unlike the 36 inch valves the 6 inch valves will close during a LOCA or steam line break accident and therefore the site boundary dose guidelines of10 CFR 100 would not be exceeded in the event i

of an accident during purging operations.

Periodic leakage integrity tests for purge supply and exhaust isolation 1

valves with resilient material seals will be performed in accordance with the Containment Leakage Rate Testing Program.

i.

3/4.6.2. DEPRESSURIZATION AND COOLING SYSTEMS 3'4.6 2.1 RE ACTOR BUILDING SPRAY SYSTEM The OPERABILITY of the reactor building spray system ensures that reactor building depressurization and cooling capability will be available in the event of a steam line break. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.

The reactor building spray system and the reactor building cooling system are redundant to each other in providing post accident cooling of the reactor building atmosphere. However, the reactor building spray system also provides a mechanism for removing iodine from the reactor building atmosphere and therefore the time requirements for restoring an inoperable spray system to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

SUMMER-UNIT 1 B 3/4 6-3 Amendment No.

5

a ACMENTSTRATTVE CONTROL 5 f.

Radiercaical Environmental Monitorina Procram (Continue 3

2)

A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that codifications to the monitoring program are made if required by the results of the census; and' 3)

Participation in an Inter-laboratory Ce:parison Program to ensure that independent checks on the precision and accuracy of measurements of radioactive materials in environmental sample matrices are performed as part of the quality' assurance program,

for environmental monitoring..

M ^-

g.

Containment Leakace Rate Testina Procram A program shall be established to implement leakage rate testing of the containment system as required by 10 CFR 50.54 (o) and 10 CFR 50, A modified by approved exemptions. ppendix J, Option B, as This program shall be in

}

accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test ggf Program", dated September 1995; NEI 94-01, " Industry l'yvl [g)

Guideline for Performance-Based Option of 10CFR50, Appendix J", Revision 0; ANSI /ANS-56.8-1994, "Contain-( g?O ment System Leakage Testing Requirements"; as modified by (i

approved exceptions.

N O

k The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is,45.1 psig.

The maximum allowable containment leakage rate, La, at Pa, is 0.20 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Leakage rate acceptance criteria are:

1)

Containment overall leakage rate acceptance criterion is 5 1.0 La. During*the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and Type C tests, and s 0.75 La for Type A tests; 2)

Air lock testing acceptance criteria are:

Overall air lock leakage rate is s 0.10 La when a.

tested at 2 Pa.

b.

For each door, leakage rate is s 0.01 La when pressurized to 2 8.0 psig for at least 3 minutes.

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.are applicable to the containment Leakage Rate Testing Program.

SUMMER - UNIT 1 6-12b Amendment tio. JM,117

4 ADMINISTRATIVE CONTROLS 1

f.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides 4

i in the environs of the plant. The program shall provide (1) representative measures of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1)

Monitoring, sampling, analysis, and reporting of radiation and radionuclides m the environment in accordance with the methodology and parameters in the ODCM; A Land Use Census to ensure that changes in the use of areas 2) at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by i

. the results of the census; and l

3)

Participation in an Inter-laboratory Comparison Program to ensure that independent checks on the precision and accuracy of measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

l g.

Containment Leakage Rate Testing Program i

.A program shall be established to implement leakage rate testing of i

the containment s tem as required by 10 CFR 50.54(o) and 10 CFR j

50, AppendixJ, tion B, as modified by approved exemptions. This j

arogram shall be in accordance with the guidelines contained in 3egulator Program",y Guide 1.163, " Performance-Based Containment Leak-Tes dated September 1995; NEI 94-01, " Industry Guideline for Performance-Based Option of10 CFR 50, Appendix J", Revision 0; ANSI /ANS-56.8-1994, " Containment System Leakage Testing Requirements"; as modified by approved exceptions.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 45.1 psig.

1 The maximum allowable containment leakage rate, La, at Pa, is 0.20 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , l

Leakage rate acceptance criteria are:

1)

Containment overallleakage rate acceptance criterion is s 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and Type C tests, and s 0.75 La for Type A tests; SUMMER -UNIT 1 6-12b Amendment No.104,117,130, 1

~

ADMINISTRATIVE CONTROLS g.

Containment Leakage Rate Testing Program (Continued) 2)

Air lock testing acceptance criteria are:

a.

Overall air lock leakage rate is s 0.10 La when tested at 2 Pa.

b.

For each door, leakage rate is s 0.01 La when pressurized to 2 8.0 psig for at least 3 minutes.

The provisions of Specification 4.0.2 do not a? ply to the test frequencies specified in the Containment Leakage Rate '?esting Program.

The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

SUMMER-UNIT 1 6-12c Amendment No.

Document Control Desk j

Attcchm:nt II TSP 950011

)

RC-96 0109 Page 1 of 2 i

SAFETY EVALUATION FOR REVISING THE CONTAINMENT LEAKAGE RATE 1

TESTING SPECIFICATION IN THE VIRGIL C. SUMMER NUCLEAR STATION TECHNICAL SPECIFICATIONS i

4 4

DESCRIPTION OF AMENDMENT REQUEST South Carolina Electric & Gas Company (SCE&G) proposes to revise the Virgil C.

Summer Nuclear Station (VCSNS) Technical Specifications (TS) to adopt the implementation of100FR50, Appendix J, Option B for containmentleakage rate testing. TS 3/4.6.1, through changes to Limiting Conditions for Operations (LCO),

ACTION statements, and Surveillance Requirements (SR), is being revised to remove the existing prescriptive requirements for containment leakage rate testing and replace them with reference to the VCSNS Containment Leakage Rate Testing Program (CLRTP). TS 6.8.4.g has been added to the ADMINISTRATIVE CONTROLS section of TS to describe the CLRTP. The associated B ASES for the j

specification is also affected.

The proposed request is submitted to support the implementation of Option B of 10CFR50, Appendix J for containment leakage rate testing. The request includes Type A, B and C testing.

In accordance with Appendix J, either Option A or Option B may be implemented; however, licensees must submit a TS revision in accordance with 10CFR50, Appendix J, Option B, V.B for the adoption of Option B.

SAFETY EVALUATION This proposed change will remove the prescriptive TS requirements for the performance of containment leakage testing and allow leakage testing to be conducted as determined appropriate through the performance-based or risk-based alternatives described in the VCSNS CLRTP The CLRTP is developed in accordance with Regulatory Guide 1.163 and NEI 94-01. Since the requirements of Appendix J to 100FR50 will continue to apply, the type of testing will not change.

.e'-

a e,-

s

-e a--

--a w

u.-

Documtnt Control D:sk Attachment II i

TSP 950011 RC-96-0109 Page 2 of 2

- The proposed request does not modify any plant equipment or systems. It provides a mechanism within the VCSNS TS for implementing an NRC approved regulation.

The proposed request is necessary to identify that VCSNS will adopt the performance-based option of 10CFR50, Appendix J and structure its program for implementation in accordance with Regulatory Guide 1.163 and NEI 94-01.

The changes proposed do not involve any significant safety risk for the following reasons:

The requirements of Appendix J will continue to govern the type of test, testing methodology, and acceptance criteria for Type A, B and C testing.

~,

The NRC has concluded, prior to approving Option B, that performance-based testing would eliminate or modify prescriptive regulatory l) requirements for which the burden is marginal-to-safety. Reviews and analyses considered by the NRC are presented in NUREG-1493, J

PERFORM ANCE-B ASED CONTAINMENT LE AK-TEST PROGRAM, Final 2

Report, September 1995.

\\

Previous leakage testing performed at VCSNS has demonstrated low i

i overall containment leakage and supports the implementation of Option B.

The combination of the NRC review and analyses results presented in NUREG-1493, the low overall containment leakage exhibited by VCSNS, and the requirements of Appendix J controlled through Regulatory Guide 1.163 and NEI 94-01 and implemented by ANSI /ANS-56.8-1994 shows that the proposed change to the VCSNS TS has insignificant impact on the health and safety of the public.

i i

Document Control D:sk Attachm:nt III TSP 950011 RC-96-0109 Page 1 of 2 NO SIGNIFICANT HAZARDS DETERMINATION FOR REVISING THE CONTAINMENT LEAKAGE RATE TESTING SPECIFICATION IN THE VIRGIL C. SUMMER NUCLEAR STATION TECHNICAL SPECIFICATIONS Description of Amendment Reauest South Carolina Electric & Gas Company (SCE&G) proposes to revise the Virgil C.

Summer Nuclear Station (VCSNS) Technical Specifications (TS) to adopt the implementation of10CFR50, Appendix J, Option B for containmentleakage rate testing. TS 3/4.6.1, through changes to Limiting Conditions for Operations (LCO),

ACTION statements, and Surveillance Requirements (SR), is being revised to remove the existing prescriptive requirements for containment leakage rate testing and replace them with reference to the VCSNS Containment Leakage Rate Testing Program (CLRTP). TS 6.8.4.g has been added to the ADMINISTRATIVE CONTROLS section ofTS to describe the CLRTP. The associated BASES for the specification is also affected.

The proposed request is submitted to support the implementation of Option B of 10CFR50, Appendix J for containment leakage rate testing. The request includes Type A, B and C testing.

In accordance with Appendix J, either Option A or Option B may be implemented; however, licensees must submit a TS revision in accordance with 10CFR50, Appendix J, Option B, V.B for the adoption of Option B.

Basis for No Sismificant Hazards Consideration Determination SCE&G has evaluated the proposed changes to the VCSNS TS described above j

against the Significant Hazards Criteria of10 CFR 50.92 and determined that the changes do not involve any significant hazard for the following reasons.

1 1

1.

The probability or consequences of an accident previously evaluated is not significantly increased.

1 There is no increase in the probability of an accident since there is no work that would affect containment integrity. The testing of containment isolation valves 4

(CIVs) and other containment penetration sealing devices is not postulated as an accident precursor or initiating event.

Type A testing is capable ofdetermining the totalleakage from both local leakage paths and gross containment leakage paths. Our Type B and C testing has consistently provided accurate leakage rates for valves and penetrations.

L i

i Document Control Desk 1

(

AttachmentIII TSP 950011 RC-96-0109 Page 2 of 2 Administrative controls govern maintenance and testing such that there is very low probability that unacceptable maintenance or alignments can occur. Prior to and following maintenance on CIVs and penetrations, a local leak rate test (LLRT) is required to be performed. As a result, Type A testing is not required to accurately quantify the leakage through containment penetrations.

Any specific exemptions to the requirements of Appendix J will require approval by the NRC before implementation.

Therefore, this proposed change does not involve a significant increase in the possibility or consequences of an accident previously evaluated.

2.

The possibility of an accident or a malfunction of a different type than any previously evaluated is not created.

The proposed request does not involve any physical changes to the plant, affect the operation of the plant, or change testing methods or acceptance criteria. The history ofcontainment testing verifies that containment integrity has been maintained, l

The frequency changes allowed by implementation of Option B will not significantly decrease the level of confidence in the ability of the reactor building to limit offsite doses to allowable values. No accident or malfunction can be the result of the allowed changes to test schedule or frequency.

Since the proposed request will not directly impact equipment, procedures or operations, the changes will not create the possibility of any new or different kind of accident from any previously evaluated.

i 3.

The margin of safety has not been significantly reduced.

The reason for performing containment leakage rate testing is to assure that the 3

leakage paths are identified, and that any accident release will be restricted to those paths assumed in the safety analysis. The purpose for the schedule is to assure that containment integrity is verified on a periodic basis.

Implementation of Option B to provide flexibility in the scheduled requirements does not mean that containment integrity will be compromised. The historical leakage rate test results for VCSNS and for the nuclear industry support extension of testing frequencies and demonstrate that structural integrity has been maintained.

Therefore, the margin of safety has not been significantly reduced.

I-

.

ML20107G351

Text

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