ML20108A509

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Proposed Tech Specs,Updating Primary Containment Bases to Incorporate Results of plant-unique Analysis Rept for Mark I Containment long-term Program
ML20108A509
Person / Time
Site: Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date: 11/02/1984
From:
VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20108A504 List:
References
NUDOCS 8411150024
Download: ML20108A509 (3)
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1 113 e

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$o 3.7 STATION CONTAINMENT __ SYSTEMS g 4 gr A. primary Containment

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-gM The integrity of the primary containment and operation of the core standby cooling systems in combination' limit the off-site doses to values less than to those suggested in 10 CFR 160 in the event of a break in j, the primary system piping. Thus, containment integrity is specified whenever the potential for violation jp of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor

is critical, above atmospheric pressure and temperature above 2120F. An exception is sede to this l requirement during initial core loading and while a low power test program is being conducted and ready j access to the reactor vessel is required. The reactor may be taken critical during the period; however,

, restrictive operating procedures will be in effect again to minimize the probability of an accident'

{ occurring. Procedures and the Rod Worth Minimizer'would limit control worth to less'than 1.30% dolca k.

The pressure suppression pool water provides the heat sink for the coactor primary system energy release

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.. following posulated rupture of the system. The pressure suppression chamber. water volume must absorb the j associated decay and structural sensible heat released during primary system blowdown from 1000 psig.

Since all the gases in the drywell are purged into the pressure suppression chamber air space during a loss-of-coolant accident, i.he pressure resulting from isothecmal compression' plus the vapor pressure of the '
liquid must not exceed 62 psig, the allowable pressure suppression chamber pressure. .The design' volume of

! the suppression chamber (water and air) was obtained-by considering that the total volume of reactor l coolant to be condensed is discharaed.to the suppression chamber and that the drywell volume is purged to l the suppression chamber (Reference Section 5.2 FSAR).

Using the minimum or maximum water volumes given in the specification, containment pressure during the design' basis accident is approximately 44 psig, Which is below the design of 56 psig.(3) The minimum

! volume of 68,000 ft3 results in a submergency of approximately four feet. The majority of the Bodega I tests (2) were run with a submerged length of four feet and with complete condensation. Thus,.with

! respect to downconer submergence, this specification is adequate.

The maximum temperature at the end of blowdown tested during the Humbolt Bay (1) and Bodega Bay tests was 1700F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperature above 1700F.

Amen 6 ment No. 50 138

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3.7.A (Cont'd) - 1' In conjunction with the Mark I Containment Long-Term program, a plant unique analysis was performed (see:

Vermont Yankee letter, dated April 27, 1984, transmitting Teledyne Engineering Services Company P.eports- j TR-5319-1, Revision 2, dated November 30, 1983 and TR-5319-2, Revision'0) which demonstrated that all stresses in the suppression chamber structure including shell, external supports, vent system, internal-structures, and attached piping meet the structural acceptance criteria of NUREC-0661. The maintenance of a drywell-suppression chamber differential pressure of 1.7 psid and a suppression chamber water level

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! corresponding to a downcomer submergence range of 4.29 to 4.54 feet will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodymanic forces.

Using a 500F rise (section 5.2.4 FSAR) in the suppression chamber water temperature and a minimum water i

volume of 68,000 ft ,3 the 1700F temp 6rature which is used for complete condensation would be approached l only if.the suppression pool temperature is 2100F prior to the DBA-LOCA. Maintaining a' pool temperature l of 90 0 F will asstre that the 1700F limit is not approached, i

! Experimental data indicate that excessive steam condensing loads can be avoided if.the peak' temperature cf 3

the suppression pool is maintained below 1600F during any period of relief valve operation with sonic j conditions at the discharge exit. Specifications have been placed on the envelope of reactor. operating l conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially j high suppression chamber loadings.

l In addition to the limits on temperature of the suppression chamber pool water, operating procedures define-

! the action to be taken.in the event a relief valve inadvertently opens or sticks open. This action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers (3) initiate reactor shutdown, and-(4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that-of the stuck-open relief valve to assure mixing and j

uniformity of energy insertion to the pool.

j Double isolation valves are provided on lines which penetrate the primary containment and open to the free

space of the containment. ' Closure of one of.the valves in each line would be sufficient to maintain the

! integrity of the pressure suppression system. Automatic initiation is required to minimize the potential-I leakage paths from the containment in the event of a loss-of-coolant accident. Details of the isolation.

valves are discussed in Section 5.2 of the FSAR.

Amendment-No. 50 139 4

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VYNpS 3.7.A (Cont'd)- .

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 structural integrity of the containment-is maintained.

Technical Specification 3.7.A.9.c is based on the assumption that the operability testin5 of the pressure

' suppression chamber-reactor building vacuum breaker, when required, will normally be performed during the.

same four hour testing interval as the pressure suppression chamber-drywell vacuum breakers -in order to minimize operation with (1.7 psi, differential pressure.

The vacuum relief system from the pressure suppression chamber to Reactor Building consists of two 100E vacuum relief breakers (2 parallel sets of 2 valves in' series). Operation of'either system will maintain the pressure differential less than 2 psig; the external design pressure is 2 pois.

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 l

Boiler and pressure Vessel Code,Section III, Subsection B, for this vessel allows e'.ght (8) operable valves, therefore, with two (2) valves secured, containment integelty is not impaired.

Each drywell-suppression chamber vacuum breaker is fitted with a redundant pair of limit switches to i provide fail-safe signals to panel mounted indicators in the Reactor Building and alarms in the Control

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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, the . accessability of the valve lever am and position

! reference external to the valve. The fail safe feature of the alam circuits assures operator. attention if a line' fault occurs.

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(1) Robbins, C.H., " Tests on a Full Scale 1/48 Segment of the Humbolt Bay pressure Suppression Containment",

l j GEAp-3596, November 17, 1960.

1 l (2) Bodega Bay preliminary Hazards Susunary Report, Appendix 1 Docket 50-205, December 28, 1962. .

l- (3) Code Allowable' peak accident pressure is 62 psig.

139a i

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