ML20059A323

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Proposed Tech Specs,Clarifying Requirements for Maintaining Secondary Containment Integrity When One or More Reactor Bldg Ventilation Supply & Exhaust Valves Declared Inoperable
ML20059A323
Person / Time
Site: Oyster Creek
Issue date: 12/16/1993
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20059A303 List:
References
NUDOCS 9312300106
Download: ML20059A323 (8)


Text

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1.14 SECONDARY CONTAINMENT INTEGRITY  :

P Secondary containment integrity means that the reactor building is closed and the following conditions are met: I t

A. At least one door at each access opening is closed. 1 B. The standby treatment system is operable.

C. All automatic secondary containment isolation valves are operable or i are secured in the closed position.

1.15 (DELETED) {

t 1.16 RATED FLUX  ;

Rated flux is the neutron flux that corresponds to a steady state power level of 1930 NW(t). Use of the term 100 percent also refers to the 1930  !

thermal megawatt power level. -

1.17 REACTOR THERMAL POWER-TO-WATER -

i Reactor thermal power-to-water is the sum of (1) the instantaneous integral over the entire fuel clad outer surface of the product of heat t transfer area increment and position dependent heat flux and (2) the ,

instantaneous rate of energy deposition by neutron and gamma reactions in all the water and core components except fuel rods in the -

cylindrical volume defined by the active core height and the inner ,

surface of the core shroud. ,

1.18 PROTECTIVE INSTRUMENTATION LOGIC DEFINITIONS  :

A. Instrument Channel f An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip l system a single trip signal related to the plant parameter '

monitored by that instrument channel. '

1 B. Trio System A trip system means an arrangement of instrument channel trip  !

signals and auxiliary equipment required to initiate action to  ;

accomplish a protective trip function. A trip system may require  ;

one or more instrument channel trip signals related to one or more l plant parameters in order to initiate trip system action. I Initiation of protective action may require the tripping of a single trip system (e.g., initiation of a core spray loop, automatic depressurization, isolation of an isolation condenser, J offgas system isolation, reactor building isolation, standby gas treatment and rod block) or the coincident tripping of two trip systems (e.g., initiation of scram, isolation condenser, reactor isolation, and primary containment isolation).

0YSTER CREEK 1.0-3 Amendment No.: 10 Change 7 I

9312300106 931216 ((;

PDR ADOCK 05000219 P PDR h

2. Upon the accidental loss of SECONDARY CONTAINMENT  :

INTEGRITY, restore, SECONDARY CONTAINMENT INTEGRITY ~

within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, except as previded in specification  :

3.5.B.3. i

3. With one or more of the automatic secondary i containment isolation valves inoperable: >
a. Maintain at least one automatic secondary  !

containment isolation valve in each affected -l penetration OPERABLE.

b. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> restore the inoperable automatic secondary containment isolation valve (s) to OPERABLE status or isolate each affected ~ ,

penetration with at least one valve secured in j the closed position.  :

4. If Specifications 3.5.B.2 or 3.5.B.3 cannot be met:
a. During Power Operation:

(1) Have the reactor mode switch in the shutdown i mode position within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(2) Cease all work on the reactor or its connected systems in the reactor building which could result in inadvertent releases of radioactive materials.

(3) Cease all operations in, above or around the  ;

Spent Fuel Storage Pool that could cause release of radioactive materials.

b. During refueling:  ;

(1) Cease fuel handling operations or activities which could reduce the shutdown margin  ;

(excluding reactor coolant temperature  ;

changes).  ;

(2) Cease all work on the reactor or its connected systems in the reactor building which could result in' inadvertent releases of '

radioactive materials.

(3) Cease all operations in, above or around the Spent Fuel Storage Pool that could cause release of radioactive materials.

5. Two separate and independent standby gas treatment I system circuits shall be operable when secondary  ;

containment is required except as specified by ,

Specification 3.5.B.6. l 0YSTER CREEK 3.5-6 Amendment No.: 14,18,32,74,103 l l

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6. With one standby gas treatment system circuit I  ;

inoperable, j i

a. During Power Operation:  ;

i (1) Demonstrate the operability of the other j standby gas treatment system circuit within 2 {

hours unless significant painting, fire, or _;

chemical release has taken place in the 3 reactor building within.the previous.12 i _

hours. In this event, demonstration of operability shall take place within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of-  ;

the expiration of the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period, and'  !

1 (2) Continue to demonstrate the operability of the standby gas treatment system circuit once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable standby gas  ;

treatment circuit is returned to operable  !

status. l 1

(3) Restore the inoperable standby gas treatment i circuit to operable status within 7 days or 1 be subtritical with reactor coolant temperature less than 212 F within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.  !

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b. During Refueling:  !

(1) Demonstrate the operability of the redundant standby gas treatment system within 2-hours unless significant painting, fire, or chemical release has taken place in the reactor building within the previous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In this event, demonstration of operability shall take place within I hour of  ;

the expiration of the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period, and I (2) Continue to demonstrate the operability of the redundant standby gas treatment system once per 7 days until the inoperable system  ;

is returned to operable _ status.

(3) Restore the inoperable standby gas treatment system to operable status within 30 days or cease all spent fuel handling, core alterations or operation that could reduce the shutdown margin (excluding reactor coolant temperature changes).

7. If Specifications 3.5.B.5 and 3.5.B.6 are not met, I~

reactor shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the condition of Specification 3.5.8.1 shall be met.

0YSTER CREEK 3.5-7 Amendment No.: 14,18,32,74,103 l

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Bases:

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. Specifications are placed on the operating status of the containment systems  !

to assure.their availability to control the release of any radioactive  !

materials from irradiated fuel in the event of an accident condition. The primary containment system (1) provides a barrier against uncontrolled release of fission products to the environs in the event of a break in the reactor coolant systems. 1 l

Whenever the reactor coolant water temperature is above 212*F, failure of  !

the reactor coolant system would cause rapid expulsion of the coolant from i the reactor with an associated pressure rise in the primary containment. l Primary containment is required, therefore, to.contain the thermal energy of [

the expelled coolant and fission products which could be released from any i fuel failures resulting from the accident. If the reactor coolant is not l above 212*F, there would be no pressure rise in the containment. In .;

addition, the coolant cannot be expelled at a rate which could cause fuel j failure to occur before the core spray system restores cooling to the  ;

core. Primary containment is not needed while performing low power '

physics tests since procedures and the Rod Worth Minimizer would limit rod worth such that a rod drop would not result in any fuel damage. In addition, in the unlikely event that an excursion did occur, the reactor  ;

building and standby gas treatment system, which shall be operational during l this time, offer a sufficient barrier to keep off-site doses well below j 10 CFR 100 limits.  ;

i The absorption chamber water volume provides the heat sink for the reactor I coolant system energy released following the loss-of-coolant accident. The l core spray pumps and containment spray pumps are located in the' corner rooms l and due to their proximity to the torus, the ambient temperature in those .l rooms could rise during the design basis accident. Calculations (7) made, assuming an initial torus water temperature of 100*F.and a minimum water  !

volume of 82,000 ft.', indicate that the corner room ambient temperature l would not exceed the core spray and containment spray pump motor operating l temperature limits and, therefore, would not adversely affect the long-term i core cooling capability. The maximum water volume limit allows for an operating range without significantly affecting accident analyses with respect to free air volume in the absorption chamber. For example, the  ;

containment capability (8) with a maximum water volume of 92,000 ft.' is '

reduced by not more than 5.5% metal-water reaction below the capability with 4 82,000 ft.8 .

Experimental data indicate that excessive steam condensing loads can be  ;

avoided if the peak temperature of the suppression pool is maintained below i 160'F during any period of relief valve operation with sonic conditions at  !

the discharge exit. Specifications have been placed on-the envelope of reactor operating conditions so that the reactor can be depressurized in a .l timely manner to avoid the regime of potentially high suppression chamber l loadings. l 1

0YSTER CREEK 3.5-8 Amendment No.: 14,74,75,103,165 l

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The technical specifications allow for torus repair work or inspections that i might require draining of the suppression pool when all irradiated fuel is removed or when the potential for draining the reactor vessel has been  !

minimized. This specification also provides assurance that the irradiated' [

fuel has an adequate cooling water supply for normal and emergency  !

conditions with the reactor mode switch in shutdown or refuel whenever the .

suppression pool is drained for inspection or repair. l The purpose of the vacuum relief valves is to equalize the pressure between  !

the drywell and suppression chamber, and suppression chamber and reactor building so that the containment external design pressure limits are not-  :

exceeded. _;

The vacuum relief system from-the reactor building to the pressure i suppression chamber consists of two 100% vacuum relief breaker subsystems (2  !

parallel sets of 2 valves in series). Operation of either. subsystem will maintain the containment external pressure less than the 2 psi external ,

design pressure of the drywell; the external design pressure of the .j suppression chamber is 1 psi (FDSAR Amendment 15, Section lli The capacity of the 14 suppression chamber to drywell vacuum relief valves ,

is sized to limit the external pressure of the drywell during post-accident '

drywell cooling operations to the design limit of 2 psi. They are sized on the basis of the Bodega Bay pressure suppression tests (9)(10). In Amendment 15 of the Oyster Creek FDSAR,Section II, the area of 2920 sq. in. is stated.

as the minimum area for flow of non-condensible gases from the suppression chamber to the drywell. To achieve this requirement, at least 12 of the 14  !

vacuum breaker valves (18" diameter) must be operable. -4 Each suppression chamber drywell vacuum breaker is fitted with a redundant ,

pair of limit switches to provide fail safe signals to panel mounted indicators in the Reactor Building and alarms in the Control Room when the disks are open more than 0.1" at any point 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 redundancy of components and circuits and, most importantly, the accessibility of the i 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.

Conservative estimates of the hydrogen produced, consistent with the core cooling system provided, show that the hydrogen air mixture resulting from a loss-of-coolant accident is considerably below the flammability limit and hence it cannot burn, and inerting would not be needed. However,.inerting of the primary containment was included in the proposed design and operation. The 5% oxygen limit is the oxygen concentration limit stated by the American Gas Association for hydrogen-oxygen mixtures below which combustion will not occur.(4) The 4% oxygen limit was established by analysis of the Generation and Mitigation of Combustible Gas Mixtures in Inerted BWR Mark I Containments.(12)

OYSTER CREEK 3.5-9 Amendment No.: 75,86,100,165

To preclude the possibility of starting up the reactor and operating a long period of time with a significant leak in the primary system, leak checks must be made when the system is at or near rated temperature and pressure. 1 It has been shown(9)(10) that an acceptable margin with respect to  ;

flammability exists without containment inerting. Inerting the primary containment provides additional margin to that already considered acceptable. Therefore, permitting access to the drywell for the purpose of leak checking would not reduce the margin of safety below that considered adequate and is judged prudent in terms of the added plant safety offered by .

the opportunity for leak inspection. The 24-hour time to provide inerting-  :'

is judged to be a reasonable time to perform the operation and establish the required 0, limit.  ;

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads. It is, therefore, required that all snubbers required to protect the primary coolant system or any other safety system or component be operable whenever the systems they protect are required to be operable.

The purpose of an engineering evaluation is to determine if the components protected by the snubber were adversely affected by the inoperability of ,

the snubber. This ensures that the protected component remains capable of meeting the designed service. A documented visual inspection will usually be sufficient to determine system operability.

Because snubber protection is required only during low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replacements. .

l Secondary containment (5) 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 oraration i when the drywell is sealed and in service and provides primary containment i when the reactor is shutdown and the drywell is open, as during refueling. i Because the secondary containment is an integral part of the overall I containment system, it is required at all times that primary containment l is required. Moreover, secondary containment is required during fuel handling operations and whenever work is being performed on the reactor or its connected systems in the reactor building since their operation ,

could result in inadvertent release of radioactive material. l When secondary containment is not maintained, the additional restrictions on operation and maintenance give assurance that the probability of inadvertent releases of radioactive material will be minimized. Maintenance l will not be performed on systems which connect to the reactor vessel lower than the top of the active fuel unless the system is isolated by at least one locked closed isolation valve.

The standby gas treatment system (6) filters and exhausts the reactor building atmosphere to the stack during secondary containment isolation conditions, with a minimum release of radioactive materials from the reactor building to the environs.

0YSTER CREEK 3.5-10 Amendment No.: 14,18,74,75,86,100 Corrected 12/24/84

4 Two separate filter trains are provided, each having 100% capacity (6).  !

If one filter train becomes inoperable, there is:no immediate threat to '

secondary containment and reactor operation may continue while repairs - l are being made. _Since the test interval for this system is one month

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(Specification 4.5), the time out-of-service allowance of 7 days is based q on considerations presented in the Bases in Specification 3.2 for a i one-out-of-two system.

Two automatic secondary containment isolation valves are installed in each Reactor Building ventilation system supply and exhaust duct penetration.  ;

Both isolation valves for each supply. duct penetration are: located'inside1  !

the secondary containment boundary, and the two exhaust duct penetration isolation valves are located outside'of the secondary containment boundary.

Removal of an inboard supply or exhaust valve (closest to the boundary) is  !

permitted only when secondary containment is not required. The outboard  ;

isolation supply or exhaust valve can be removed when_ secondary containment is required as long as the inboard valve is secured'in the closed position.

References:

(1) FDS'R, Volume I, Section V-1 (2) FDSAR, Volume I, Section V-1.4.1 (3) FDS)R, Volume I, Section V-1.7 i (4) Licensing Application, Amendment 11, Question III-25 ti (5) FDSAR, Volume I, Section V-2 i (6) FDSAR, Volume I, Section V-2.4 (7) Licensing Application, Amendment 42 l (8) Licensing Application, Amendment 32, Question 3  :

(9) Robbins, C. H., " Tests on a Full Scale 1/48 Segment  ;

of the Humboldt Bay Pressure Suppression Containment," i GEAP-3596, November 17, 1960.

(10) Bodega Bay Preliminary Hazards Summary Report, Appendix.I, Docket 50-205, December 28, 1962.

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(11) Report H. R. Erickson, Bergen-Paterson To K. R. Goller,  :

NRC, October 7, 1974.

Subject:

Hydraulic Shock Sway ,

Arrestors.  !

(12) General Electric NEDO-22155 " Generation and Mitigation l of Combustible Gas Mixtures in Inerted BWR Mark I. .i Containment" June 1982. J (13) Oyster Creek Nuclear Generating Station, Mark I Containment 1 Long-Term Program, Plant Uniques Analysis Report,.  :

Suppression Chamber and Vent System, MPR-733; August, 1982.

(14) Oyster Creek Nuclear Generating-Station, Mark I l Containment Long-Term Program, Plant Unique Analysis  !

Report, Torus Attached Piping, MPR-734; August, 1982.  !

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l OYSTER CREEK 3.5-11 Amendment No.: 14,18,74,75,86,87,100 Corrected: 12/24/84

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e. Mechanical Snubbers Functional' Test Acceptance' Criteria

_The mechanical snubber functional test shall verify that:

l. The force that initiated free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.
2. Activation (restraining action) is achieved within the f specified range of velocity or acceleration in both tension ard compression. '
3. Snubbe" release rate, where required, is within the l specificd range in compression or tension. For ,

snubbers specifically required not to displace. under.

continuous load, the ability of the srubber to .

withstand load without displacement shall be verified.-

f. Snubber Service Life Monitorina A record of the service life of each snubber, the date at .

which the designated service life commences and.the [

installation and maintenance records on which the designated i service life is based shall be maintained as required by 1 Specification 6.10.2.1. 1 Concurrent with the first inservice visual ~ inspection and at i least once per 24 months thereafter, the installation and i maintenance records for each snubber shall be reviewed to verify that the indicated service life has not been exceeded  :

or will not be exceeded prior to the next-scheduled snubber service life review. If the indicataJ service life will be exceeded prior to the next scheduled snubber service life-  :

review, the snubber service life shall be re-evaluated or the  !

snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled i service life review. ~ This re-evaluation, replacement or l reconditioning shall be indicated in the records. Service  ;

life shall not at any. time affect reactor operations. l 1

R. Secondary Containment Isolation Valves -

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1. Each secondary containment isolation' valve shall be demonstrated t

operable prior to returning the valve to service after i maintenance, repair or replacement' work is performed on the- ,

valve or its associated actuator by cycling the valve through at l least one complete cycle of full _ travel. Following maintenance, j repair or replacement work on the control or power circuit for ;j the valves, the affected component shall .be tested to assure it j will perform its intended function in the circuit.

2. At least once per refueling outage, all valves shall be tested for automatic closure by an isolation signal. g OYSTER CREEK 4.5-11 Amendment No.: 100, 132, 141 i

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