ML20101P138

Proposed Tech Specs Re Secondary Containment Integrity

ML20101P138
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/04/1996
From:	VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20101P128	List: ML20101P126 ML20101P138
References
NUDOCS 9604090287
Download: ML20101P138 (16)

Text

-

United States Nuclear Regulatory Commission April 4,1996 Pronosed Change 184 Marked-Up Technical Specification Pages 9604090287 960404 PDR ADOCK 05000271 P

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G VYt!PS 3.7 LIMITIt1C CC!!OITICtlS FCR 4.7 OPCFATICN SURVEILLAt:CE REOUIRD4ENTS AP is reduced to

<l.7) during required operability testing of the HPCI system pump, the RCIC system pump, the drywell-suppression chamber vacuum breakers, and the suppression chamber-reactor building vacuum l

breakers, and SGTS testing.

d.

If the specifications of 3.7.A.9.a cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours.

B.

Standbv Cas Treatment Svstem B.

Standbv Gas Treatment Svstem l

1.

a.

Except as specified 1.

At least once per in Specification operating cycle, not to 3.7.B.3.a below, exceed 18 months, the whenever the following conditions reactor is in Run shall be demonstrated.

Mode or Startup p got.

Mode A both circuits a.

Pressure drop m%3 of the Standby Gas across the combined a,st:wo Treatment System HEPA and charcoal shall be operable filter banks is at all times when less than 6 inches secondary of water at centainment 1500 cfm 210%.

integrity is required.

b.

Inlet heater input is at least 9 kW.

b.

Except as specified in Specification 3.7.B.3.b below, whenever the reactor is in or Cold Refuel Mode,4both shwhb>o circuits of the Cond;Wn Standby Gas Amsndment No, M, 49, M- 'C 152

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VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION C.

Secondary Cont a i runen t Svstem C.

Secondary Containment System 1.

I egrity of he 1.

Surveillance of econdary ntainment secondary containment system sh 1 be shall be performed as maintain d during al follows:

modes plant oper ion excep when all of the a.

A preoperational foll wing condit ns are secondary me containment capability test The rea or is shall be conducted subcri ical and after isolating the Spec ication 3.3.

Reactor Building is et and and placing either Standby Gas b.

e reactor wa er Treatment System temperature filter train in below 2120F and the operation.

Such reactor c lant tests shall system i vented.

demonstrate the capability to c.

No ac vity is maintain a beir performed 0.15 inch of water wh' h can reduce vacuum under calm

.e shutdown gin wind elow that (2 < u < 5 mph) opecified i condition with a Hpecificat*on filter train flow

'.3.A.

rate of not more than 1500 cfm.

d.

The f 1 cask or irr lated fuel b.

Additional tests no being moved n shall be performed l

e Reactor during the first Building.

operating cycle under an adequate number of different environmental wind conditions to enable valid extrapolation of the test results.

c.

Secondary i

containment In ser { 3.7.C. l capability to maintain a 4hrogh 3.7.c.af 0.15 inch of water vacuum under calm wind (2<G<5 mph) conditions with a filter train flow rate of not more than 1,500 cfm, shall be demonstrated at least quarterly and at each refueling i

outage prior to refueling.

.Ty) Sert Amendment No. -itt-SC '4.7'C i. d 155

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVE.'.LLANCE REQUIREMENTS OPERATION

-m

)

d.

Operabili testing 1.

Secondary Containment Integrity shall of the Reactor be maintained during the following Building Automatic

)

" SY'D*"

Modes or conditions:

isolation valves shall be performed

a. Whenever the reactor is in the Run in accordance with Mode, Startup Mode, or Hot Shutdown Specification condition; or 4.6.E.

b. During movement of irradiated fuel assemblies or the fuel cask in j

secondary containment; or

c. During Alteration of the Reactor Core; or

d. During operations with the potential for draining the reactor vessel..

2.

With Secondary Containment Integrity 2.

Intentionally blank.

not maintained with the reactor in 3.

Intentionally blank.

the Run Mode, Startup Mode, or Hot 4

Shutdown condition, restore Secondary 4.

Intentionally blank.

Containment Integrity within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3.

If Specification 3.7.C.2 cannot be met, place the reactor in the Hot Shutdown j

l condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in the Cold Shutdown condition within the QQ following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.

With Secondary Containment Integrity not maintained during movement of irradiated fuel assemblies or the fuel cask in secondary containment, during Alteration of the Reactor Core, or during operations with the potential for draining the reactor vessel, immediately perform the following actions:

a. Suspend movement of irradiated fuel

~

assemblies and the fuel cask in secondary containment; and

b. Suspend Alteration of the Reactor Core; and

c. Initiate action to suspend operations with the potential for draining the reactor vessel.

~

k 155a

VYNPS 3.7 LIMITING CONDITICNS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION Core spray and LPCI pump The core spray and LPCI lower compartment door lower compartment g,

openings shall be closed 6,

openings shall be at all times except checked closed daily.

during passage or when reactor coolant temperature is less than 212*F.

D.

Primary Containment D.

Primary Containment Isolation Valves Isolation Valves l

1.

During reactor power 1.

surveillance of the operating conditions all primary containment isolatien valves listed isolation valves should in Table 4.7.2 and all be performed as follows:

i instrument line flow 1

check valves shall be a.

The operable operable except as isolation valves specified in that are power Specification 3.7.D.2.

operated and 1

automatically I

initiated shall be tested for automatic initiation and the closure times specified in 3

Table 4.7.2 at least once per operating cycle, b.

operability testing of the primary containment isolation valves shall be performed in accordance with Specification 4.6.E.

1 I

c.

At least once per quarter, with the reactor power less than 75 percent of rated, trip all main steam isolation valves (one at a time) and verify closure time.

d.

At least twice per week, the main steam line isolation valves shall be exercised by partial closure and subsequent reopening.

i Amendment No. %, M, H e, + M-156 4

..,m.+

e,..,.,

I 9

VYNPS r

BASES:

3.7 (Cont'd)

The vacuum relief system from the pressure suppression chamber to Reactor Building consists of two 1001 vacuum relief breakers (2 parallel sets of 2 valves in series). Operation of either syntem will maintain the pressure differential less than 2 psig; the external design pressure is 2 psig. With one vacuum breaker out of service there is no immediate threat to accident mitigation or primary containment and, therefore, reactor operation can be continued for 7 days while repairs are being made.

j The capacity of the ten (10) drywell vacuum relief valves is si=ed to i

limit the pressure differential between the suppression chamber and drywell during pest-accident drywell cooling operations to the design limit of 2 psig. They are si:ed on the basis of the Bodega Bay pressure suppression tests. The ASME Boiler and Pressure Vassel Code,Section III, Subsection B, for this vessel allows elcht (8) operable valves, therefore, with two (2) valves secured, containment integrity is not impaired.

Each Jrywell-suppressicn :hambar vacuum breaker is fitted with a redundant pair of limit switeht e provide fail-safe signals to panel mounted indicators in the. actor Building and alarms in the Control Room when the disks are open more than 0.050* at all points along the seal surface of the disk. These switches are capable of transmitting the disk closed to open signal with 0.01 movement of the switch plunger. Continued reactor operation with failed components is justified because of the redundance of components and circuits and, most importantly, tha accessibility of the valve lever arm and position reference externa; to the valve. The fail safe feature of the alarm circuits assures operator attention if a line fault occurs.

The requirement to inert the containment is based on the recommendation of the Advisory Committee on Reactor Safeguards.

This recommendation, in turn, is based on the assumption that several I

percent of the irconium in the core will undergo a reaction with steam during the loss-of-coolant accident. This reaction would release sufficient hydrogen to result in a flammable con.:entration in the primary containment building. The oxygen concentration is therefore kept below 4% to minimize the pcssibility of hydrogen combustion.

1 General Electric has estimated that less than 0.1% of the zirconium would react with steam following a loss-of-coolant due to operation of emergency core cooling equipment. This quantity of zirconium would not liberate enough hydrogen to form a combustible mixture.

B. and C. Standbv Gas Tres_tment System and Secendarv Containment Svstem The secondary containment is designed to minimize any ground level release of radioactive materials which might result from a serious accident. The Reactor Building provides secondary conta. ament during reactor operation, when the drywell is sealed and in service; the Reactor Building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling.

Because the secondary containment is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required except, however, for initial fuel loading and low power physics testing.

14Strb 6a515 3.].C Amendment No. 44, i ::: che ;:

lis-165 m

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the In the Cold Shutdown condition or the Refuel Mode,arobabil in these conditions.

p-essure and temperature limitation 4-

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-w a_ _ a_ s - ve,

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.t = 4 n_4 ".g.C e c c"~d a ry C.. a.4..~...a...

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4 required in the Cold Shutdcwn condition or the Refuel Mode, t

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.b. s4 c. 4 ' 4 " =...

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such as during cperations

1. c-ot..a-w.

e.< - a_c t radicactive material can be postulated,

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w_ 5. a o, c t =."..* 4 a_'

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of irradiated 3

4 Alteration of the Reactor Core, or during movement fuel assemblies or the fuel cask in the secondary containment.

With the reactor in the Run Mode, the Startup Mode, or the Hot Integrity is not Shutdown condition if Secondary ContainmentIntegrity must be res:

Ored Sec0ndary Containment The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> orovides a c.eriod of time to correct maintained, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

is commensurate with the importance of the problem that maintaining seconds y containment during the Run Mode, theThis time period an6 t e Hot shutdown condition.

Startup Mode, (recuiring also ensures that the probability of an accidentoccurring during periods where Integrity) s r....4 ~...2 _'

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Secondary Containmen:

i

~. ~.~. a_ c. d ev..i s ~^ t m.. a 4....=. 4 ".a d Sec-.~d.=.~f C^.* =4...a..*.

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~a_ 4cd, t ".a_

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condition in which the LCO does not apply.

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Alteration of the.eactor Core, and secondary containment, Or vesse' can coerations with the potential for d.11.inc_ the rea : release to the secondary be postulated to cause fission produ:In such cases, the secodary contairment containment.

" 4. s _d an c_--d".c. s

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assemblies and the fuel cask must be immediately suspended ifc ".e "y

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a.

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"..o v S a~- ~.~d a f C -*.. = #.m.e.~.t preclude ccmpleting an action that these activities shall no to a safe position.

Also, actica involves moving a ccmponentbe immediately initiated to suspend cperations with potential for draining the reacter vessel to minimi::e th mus:

fission product release.

Actions must d d.

with the potential for draining the reacter vessel are suspen e th uk

VYNPS BASES:

4.7 (Cont'd)

At the end of each refueling cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by at least 1 psi with respect to the suppression l

chamber pressure and held constant. The 2 psig set point will not be exceeded.

The subsequent suppression chamber pressure transient (if l

any) will be monitored with a sensitive pressure guage.

If the drywell pressure cannot be increased by 1 psi over the suppression chamber pressure it would be because a significant leakage path 1

l exists; in this event the leakage source will be identified and j

l eliminated before power operation is resumed.

If the drywell j

pressure can be increased by 1 psi over the suppression chamber the i

l rate of change of the suppression chamber pressure must not exceed a rate equivalent to the rate of leakage from the drywell through a q

i 1-inch orifice.

In the event the rate of change exceeds this value then the source of leakage will be identified and eliminated before power operation is resumed.

The drywell-suppression chamber vacuum breakers are exercised in accordance with Specification 4.6.E and immediately following termination of discharge of steam into the suppression chamber. This monitoring of valve operability is intended to assure that valve operability and position indication system performance does not degrade between refueling inspections. When a vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights are designed to function as follows:

t Full Closed 2 White - On l

(closed to 1 050' open) 0 Open 2 White - Off l

(>0.050' open to full open)

[

During each refueling outage, two drywell-suppression chamber vacuum l

breakers will be inspected to assure sealing surfaces and components l

have not deteriorated.

Since valve internals are designed for a 40-year lifetime, an inspection program which cycles through all l

valves in one-eighth of the design lifetime is extremely conservative.

Experience has shown that a weekly measurement of the oxygen concentration in the primary containment assures adequate surveillance of the primary containment atmosphere.

B. and C.

Standby Gas Treatment System and Secondary Containment System Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 0.15 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leakage tightness of the reactor buil

'g, and performance of the standby gas treatment l

system.d Functiona testing of initiating sensors and associated l

trip channels demon ates the capabilit/ for automatic actuation.

Periodic testing gives sufficient confidence of reactor building

(

integrity and standby gas treatment system performance capabilitv.

l v

nsyt.

~^ v v

i The testing of reactor building automatic ventilation system isolation valves tn accordance with TS 4.6.E demonstrates the operability of these valves.

j i

A

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i i

j Amendment No. -t44-169

)

' United States Nuclear Regulatory Commission April 4,1996 Procosed Change 184 New Technical Specification Pages

1 VYNPS I

3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION AP is reduced to

<l.7) during required operabi-lity testing of the HPCI system pump, the RCIC system pump, the drywell-suppression chamber vacuum breakers, and the suppression chamber-reactor building vacuum breakers, and SGTS testing.

d.

If the specifica-tions of 3.7.A.9.a cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following eighteen (18) hours.

B.

Standby Gas Treatment System B.

Standby Gas Treatment System 1.

a.

Except as specified 1.

At least once per in Specification operating cycle, not to 3.7.B.3.a below, exceed 18 months, the whenever the following conditions reactor is in Run shall be demonstrated.

Mode or Startup Mode or Hot Shut-a.

Pressure drop down condition, across the combined both circuits of HEPA and charcoal the Standby Gas filter banks is Treatment System less than 6 inches shall be o erable of water at v

at all times when 1500 cfm *10%.

secondary contain-ment integrity is b.

Inlet heater input required.

is at least 9 kW.

b.

Except as specified in Specification 3.7.B.3.b below, whenever the reactor is in Refuel Mode or Cold Shutdown condition, both circuics of the Standby Gas Amendment No. 14 49, 50, 144, 152

~

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION C.

Secondary containment System C.

Secondary containment System 1.

Secondary Containment 1.

Surveillance of Integrity shall be secondary containment maintained during the shall be performed as following modes or follows:

conditions:

a.

Whenever the a.

A preoperational reactor is in the secondary i

Run Mode, Startup containment

)

Mode, or Hot capability test Shutdown condition; shall be conducted or after isolating the Reactor Building

)

b.

During movement of and placing either irradiated fuel Standby Gas assemblies or the

'"r :tm:nt System fuel cask in filter train in secondary operation.

Such containment; or tests shall demonstrate the c.

During alteration capability to of the Reactor maintain a Core; or 0.15 inch of water j

vacuum under calm d.

During operations wind with the potential (2 < u < 5 mph) for draining the condition with a reactor vessel.

filter train flow rate of not more than 1500 cfm.

b.

Additional tests

]

shall be performed 1

during the first operating cycle under an adequate number of different environmental wind conditions to enable valid extrapolation of the test results, c.

Secondary containment capability to maintain a 0.15 inch of water vacuum under calm wind (2<G<5 mph) conditions with a filter train flow rate of not more than 1,500 cfm, shall be demonstrated at least quarterly and at each refueling outage prior to refueling.

Amend.T.ent No. M4, 155

l l

1 l

VYNPS 1

3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS j

OPERATION d.

Operability testing of the Reactor Building Automatic Ventilation System isolation valves shall be performed

)

in accordance with Specification i

4.6.E.

l 2.

With Secondary 2.

Intentionally blank.

Containment Integrity i

not maintained with the reactor in the Run Mode, I

Startup Mode, or Hot Shutdown condition, restore Secondary l

Containment Integrity within four (4) hours.

i l

l 3.

If Specification 3.7.C.2 3.

Intentionally blank.

l cannot be met, place the reactor in the Hot Shutdown condition 3

l within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in l

the Cold Shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.

With Secondary 4.

Intentionally blank, i

l Containment Integrity not maintained during movement of irradiated fuel assemblies or the fuel cask in secondary containment, during alteration of the Reactor Core, or during l

operations with the potential for draining the reactor vessel, i

immediately perform the following actions:

a.

Suspend movement of irradiated fuel assemblies and the i

fuel cask in secondary containment; and l

b.

Suspend alteration i

of the Reactor Core; and c.

Initiate action to j

suspend operations with the potential for draining the reactor vessel.

Amendment No.

155a

,+.-

yv,,>+,

l VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION l

5.

Core spray and LPCI pump 5.

The core spray and LPCI l

lower compartment door lower compartment openings shall be closed openings shall be at all times except checked closed daily, during passage or when reactor coolant temperature is less than 2120F.

D.

Primary Containment D.

Primary Containment Isolation Valves Isolation Valves 1.

During reactor power 1.

Surveillance of the operating conditions all primary containment isolation valves listed isolation valves should in Table 4.7.2 and all be performed as follows:

instrument line flow check valves shall be a.

The operable operable except as isolation valves specified in that are power Specification 3.7.D.2.

operated and automatically initiated shall be tested for automatic initiation and the closure times specified in Table 4.7.2 at least once per operating cycle, b.

Operability testing of the primary containment isolation valves shall be performed in accordance with Specification 4.6.E.

c.

At least once per quarter, with the reactor power less than 75 percent of rated, trip all main steam isolation valves (one at a time) and verify closure time.

d.

At least twice per week, the main steam line isolation valves shall be exercised by partial closure and subsequent reopening.

Amendment No. SG, e6, lee, 444, 156

1 l

VYNPS i

l l

BASES:

3.7 (Cont'd)

The vacuum relief system from the pressure suppression chamber to

.)

Reactor Building consists of two 100% vacuum relief breakers 1

(2 parallel sets of 2 valves in series).

Operation of either system i

will naintain the pressure dif ferential less than 2 psig; the external design pressure is 2 psig.

With one vacuum breaker out of service there is no immediate threat to accident mitigation or primary containment and, therefore, reactor operation can be continued for 7 days while repairs are being made.

The capacity of the ten (10) drywell vacuum relief valves is sized to limit the pressure differential between the suppression chamber and drywell during post-accident drywell cooling operations to the design limit of 2 psig. They are sized on the basis of the Bodega Bay pressure suppression tests.

The ASME Boiler and Pressure Vessel Code,Section III, Subsection B, for this vessel allows eight (8) operable valves, therefore, with two (2) valves secured, containment integrity is not impaired.

Each drywell-suppression chamber vacuum breaker is fitted with a redundant pair of limit switches to provide fail-safe signals to l

panel mounted indicators in the Reactor Building and alarms in the Control Room when the disks are open more than 0.050" at all points along the seal surface of the disk. These switches are capable of transmitting the disk closed to open signal with 0.01" movement of the switch plunger. Continued reactor operation with failed components is justified because of the redundance of components and l

circuits and, most importantly, the accessibility of the valve lever arm and position reference external to the valve. The fail safe feature of the alarm circuits assures operator attention if a line fault occurs.

The requirement to inert the containment is based on the l

recommendation of the Advisory Committee on Reactor Safeguards. This recommendation, in turn, is based on the assumption that several percent of the zirconium in the core will undergo a reaction with steam during the loss-of-coolant accident. This reaction would release sufficient hydrogen to result in a flammable concentration in the primary containment building.

The oxygen concentration is therefore kept below 4% to minimize the possibility of hydrogen combustion.

General Electric has estimated that less than 0.1% of the zirconium would react with steam following a loss-of-coolant due to operation of emergency core cooling equipment. This quantity of zirconium would not liberate enough hydrogen to form a combustible mixture.

I l

B. and C.

Standby Gas Treatment System and Secondary Containment System The secondary containment is designed to minimize any ground level release of radioactive materials which might result from a serious accident. The Reactor Building provides secondary containment during reactor operation, when the drywell is sealed and in service; the Reactor Building provides primary containment when the reactar is shutdown and the drywell is open, as during refueling.

Because the

{

secondary containment is an integral part of the complete containment 3

system, secondary containment is required at all times that primary j

containment is required except, however, for initial fuel loading and i

low power physics testing.

I j

In the Cold Shutdown condition or the Refuel Mode,the probability and j

consequences of the LOCA are reduced due to the pressur e and temperature limitations in these conditions. Therefore, maintaining i

Secondary Containment Integrity is not required in the Cold Shutdown condition or the Refuel Mode, except for other situations for which j

Amendment No. 49, Scac: Changt, 119, 165

1 a

)

VYNPS i

BASES:

3.7 (Cont'd) significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel, during alteration of the Reactor Core, or during movement of irradiated fuel assemblies or the fuel cask in the secondary containment.

With the reactor in the Run Mode, the Startup Mode, or the Hot Shutdown condition, if Secondary Containment Integrity is not maintained, Secondary Containment Integrity must be restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during the Run Mode, the Startup Mode, and the Hot Shutdown condition. This time period also ensures that the probability of an accident (requiring Secondary Containment Integrity) occurring during periods where Secondary Containment Integrity is not maintained, is minimal.

If Secondary Containment Integrity cannot be restored within the required time period, the plant must be brought to a mode or condition in which the LCO does not apply.

Movement of irradiated fuel assemblies or the fuel cask in the secondary containment, alteration of the Reactor Core, and operations with the potential for draining the reactor vessel can be postulated to cause fission product release to the secondary containment.

In such cases, the secondary containment is the only barrier to release of fission products to the environment.

Alteration ot the Reactor Core and movement of irradiated fuel assemblies and the tuel cask must be immediately suspended if Secondary Containment Integrity is not maintained.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position.

Also, action must be immediately initiated to suspend operations with the potential for draining the reactor vessel to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must contir.1e until operations with the potential for draining the reactor vessel are suspended.

Amendment No.

165a

o VYNPS BASES:

4.7 (Cont'd)

At the end of each refueling cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by at least 1 psi with respect to the suppression chamber pressure and held constant.

The 2 psig set point will not be exceeded. The subsequent suppression chamber pressure transient (if l

any) will be monitored with a sensitive pressure gauge.

If the

)

drywell pressure cannot be increased by 1 psi over the suppression chamber pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed.

If the drywell pressure can be increased by 1 psi cver the suppression chamber the rate of change of the suppression chamber pressure must not exceed a rate equivalent to the rate of leakage from the drywell through a l-inch orifice.

In the event the rate of change exceeds this value then the source of leakage will be identified and eliminated before power operation is resumed.

The drywell-suppression chamber vacuum breakers are exercised in accordance with Specification 4.6.E and immediately following termination of discharge of steam into the suppression chamber. This monitoring of valve operability is intended to assure that valve operability and position indication system performance does not degrade between refueling inspections.

When a vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights are designed to function as follows:

Full Closed 2 White - On (Closed to 10.050" open)

Open 2 White - Off

(>0.050" open to full open)

During each refueling outage, two drywell-suppression chamber vacuum breakers will be inspected to assure sealing surfaces and components have not deteriorated.

Since valve internals are designed for a 40-year lifetime, an inspection program which cycles through all valves in one-eighth of the design lifetime is extreraely conservative.

Experience has shown that a weekly measurunent of the oxygen concentration in the primary containment assures adequate surveillance of the primary containment atmosphere.

B. and C.

Standby Gas Treatment System and Secondary Containment System Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 0.15 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leakage tightness of the reactor building, and performance of the standby gas treatment system.

The testing of reactor building automatic ventilation system isolation valves in accordance with Technical Specification 4.6.E demonstrates the operability of these valves.

In addition, functional testing of initiating sensors and associated trip channels demonstrates the capability for aucomatic actuation.

Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system performance capability.

Amendment No. H 4, 169

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