ML20024G308

Proposed Tech Specs Re Containment Water Vol & Differential Pressure Instrumentation

ML20024G308
Person / Time
Site:	Monticello
Issue date:	04/15/1977
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G306	List: ML20024G305 ML20024G307 ML20024G308
References
NUDOCS 9102110273
Download: ML20024G308 (17)

Text

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3 0 LIMITING (DNDITIONS FOR OPERATION h.0 SURVEILIANCE REQUIREME!TS 3 7 CONTAI18EDtf SYSTDG 4.7 C0ffrAIfMElff SYSTD6 .

Applicability: Applicability: -

Applies to' the operating status of the primary Applies to the primary and secondary and secondary containment systems, containment integrity.

Ob.jective: Objective: , .

S-To assure the integrity o.T the primary and To verify the integrity of the primary eat seccedary containment systems. secondary containment.. .

Specification: Specification:

A. Primary Containment. A. Primary Containment.

1. Suppression Pool Volume and Temperature U m,o 1. Suppression Pool volume and Temperature Qg At any time that the reactor water temp-M erature exceeds 212 F or work is being

$ done which has the potential to drain the

@@ vessel, except as permitted by specification nN 3.5'G.4,

. the following requirements shall be N met:

ON a. . Water temperature during normal opera- a. The suppression chamber water eemperature shall

$o tion shall be g90 F. be checked once per day.

.g @ ,*,, b. ' Water temperature during test operation b. Whenever there is indication of relief valve Nui which adds heat to the suppression pool operation which adds heat to the suppressien i

g$ shall befl00 F and chall not be>90 F pool, the pool temperature shall be continually for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. monitored and also observed and logged every

c. If the suppression chamber water tempera- 5 minutes until the heat addition is terminated.

' ture is > 110 F, the reactor shall be c. A visual inspection of the suppression chamber scrammed immediately. Power operation interior including water line regions and the shall not be resumed until the pool temp- interior painted surfaces above the water line erature is s90 F. shall be made at each refueling outage.

3.7/4.7 139 g

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3.0 LDiITIN'; CONDITIONS FOR OPEPATION 4.0 SURVEILIANCE REQUIREMEhTS .

d. During reactor isoir . ion conditions d.

the reactor pressur e vessel shall be Whenever there is indication of relief ^

valve operation with a suppression pool depressurized to<c200 psig at normal temperature 2t160 F and the primary coolant cooldown rates if the suppression system pressure > 20,0 psig, an extended ,

pool temperature exceeds ' 120 F. visual examination of the suppression

e. The suppression chamber water volume shall be2: 68,000 and =577,970 cubic chamber shall be conducted before reseming power operation.

feet,

e. The suppression chamber water volume shall

f. Two channels of suppression chamber be checked once per day.

water volume indication shall be f. The suppression chamber water volume indi-available at all times except that cators shall be calibrated annually.

one of the two channels may be out llh of service for:G30 days or both channels may be out forsE24 hours for service and maintenance.

2. Primary Containment Integrity 2. Primary Containment Integrity Primary containment ir*.egrity, as defined The primary containment integrity shall be in Section 1, shall be maintained at all demonstrated as follows:

tbnes when the reactor is critical or when the reactor water temperature is above I a. Integrated Prbnary Containment Leak Test 212 F and fuel is in the reactor vessel (IPCLT) except while performing low power physics (1) An integrated leak rate test shall be tests at atmospheric pressure during or performed prior to initial unit opera-after refueling at power les els not to tion at an initial test pressure (Pt) exceed 5 Mw(t). of 41 psig.

(2) Subsequent leak rate tests shall be perfoemed without preliminary leak de-g' W

tection surveys or leak repairs inmediately prior to or during the test, at an initial pressure of approximately 41 psig.

(3) Leak repairs, if necessary to permit integrated leak rate testing, shall be preceded by local leak rate measurements where possible. The leak rate differ-l 3.7/4.7 1 140 Rev

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4.0 SURVEILLANCE REQUIREMENTS .

3.0 LIMITING CONDITIONS FOR OPERATION When the position o,f any dryvell-b.

d. One position alarm circuit can be inoperable suppression chamber vacuum breaker valve-providing that the redundant position alarm is indicated to be not- fully closed at a circuit is operable. Both position alarm time when such closure is required, the p p

circuits may be inoperable for a period not dryvell to suppression chamber differential  !

to er 2ed seven days provided that all vacuum pressure decay shall be demonstrated f breakers are operable. to be less than that shown on Figure 3.7.1 immediately and following any evidence of subsequent operation of g ll the inoperable valve until the inoperable valve is restored to a normal condition.

c. When both position alarm circuits are made or found to be inoperable, the contral panel indicator light status shall be recorded daily to detect changes in the l vacuum breaker position.

5. Oxygen concentration

5. Oxygen Concentration Whenever inerting is required, the primary

a. The primary containment atmosphere shall containment oxygen concentration shall be l measured and recorded on a weekly basis.

be reduced to less than 5% oxygen with nitrogen gas whenever the reactor is in the run mode, except as specified in O 3.7.A.S.b.

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b. Within the 24-hour period subsequent to placing the reactor in the run mode following shutdown, the containment atmosphere oxygen concentration shall be reduced to less than 5% by weight, and maintainel Dei ~nerting may commence 24 in this condition. a hours prior to leaving the run mode for reactor shutdown.

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e 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMELTS

6. If specifications 3.7. A.1 through 3.7. A.5 cannot .

be met, the reactor shall be placed in the cold ,l shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7. DRYWELL-SUPPRESSION CHAMBER DIFFEREhTIAL PRESSURE 7. DRYWELL-SUPPRESSION CHAMBER DIFFERD4TIAL PRESSURE

a. Drywell pressure shall be maintained 2el-0 psi a. The differential pressure between the drywell above the suppression chamber pressure except and suppression chamber shall be logged once as specified in 3.7.A.7.b and c. per shift.

'l Within the 26 hour3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> period subsequent to placing b. The differential pressure indicators shall be j b.

,j the reactor in the run mode following a shut- calibrated annually, il down, the drywell pressure must be raised to

!! 251.0 psi above the suppression chamber pres-I sure and maintained in this condition. The differential pressure need not be maintained during the 26 hour3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> period prior to leaving the run mode for a reactor shutdown.

c. The differential pressure may be decreased to

<C1.0 psi for a maximen cf 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> during re-quired operability testing of the HPCI system

)! pump, the RCIC system pump, and the drywell-pressure suppression chamber vacuum breakers.

On receipt of written permission of the Nuclear Regulatory Commission, the differential pres-sure may also be decreased to4 1.0 psi for periods necessary for planned safety / relief valve testing.

d. Two channels of drywell suppression chamber jlh differential pressure indication shall be avail-able at all times except that one of the two channels may be out of service forsG30 days or both channels may be out forg;8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for service and maintenance.

e. If specification 3.7.A.7 cannot be met, an orderly shutdown shall be initiated and the re-actor shall be in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 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 />.

147B REV 3.7/4.7

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l 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIRDENIS I

B. Standby Gas Treatment System B. Standby Gas Treatment System

1. Two separate and independent standby 1. At least once per month, initiste front gas treatment systen circuits shall be the control room 3500 cfm Q 107.) flow operable at all times when secondary through both circuits of the standby contairmnent integrity is required, gas treatment system. In addition:

I except as specified in sections 1

3.7.B.I.(a) and (b). a. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from the time that one standby gas treatment systen

a. After one of the standby gas circuit is made or found to be in-treatment system circuits is made operable for any reason and daily or found to be inoperable for any thereafter for the next succeeding reason, reactor operation and fuel seven days, initiate from the handling is permissible only during control room 3500 cfm Q 107.) flow the succeeding seven days, provided through the operable circuit of the that all active components in the- standby gas treatment system.

other standby gas treatment system shall be demonstrated to be oper-able within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and daily thereafter. Within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> follow-ing the 7 days, the reactor shall be placed in a condition for which the standby gas treatment system is not required in accordance with Specification 3.7.C.1. (a) through (d).

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Bases:

3.7 A. Primary Containment The integrity of the primary containment and operation of the emergency core cooling system in combination, limit the off-site doses to values less than 10 CFR 100 guideline values in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception is g,l made to this requirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the

l. system at this time which will greatly reduce the chances of a pipe break. The reactor may be taken critical dt ring this 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 incremental control worth to less than 1.37. ok. A drop of a 1.37. ok increment of a rod does

, 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 this time, o ffers a sufficient barrier to keep off-site doses well within 10 CFR 100 guide line values. ,

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blow-down from 1000 psig.

Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure ressulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the maximum allowable primary containment pressure.

The design volum.e of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber. Reference Section 5.2.3 FSAR.

Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accident is approximately 41 psig which is below the allowable pressure of 62 peig.

The nominal downcomer submergence for the Monticello wetwell desigp is 4 feet which is in conform-ance with most of the Bodega tests. The majority of Bodega tests t l Iwere run with..a submerged (1) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.

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l length of four feet, which resulted in complete condensation. Thus with respect to downconer l submergence, this specification is adequate.

The maximum temperature at the end of blowdown tested during the Humboldt Bay andBodegaBay(2)  ;

tests was 170 F and this is conservatively taken to be the limit.for complete condensation of.the 0  !

reactor coolant, although condensation would occur for temperatures above 170 0F.

Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 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 timely manner to avoid the regime g

of potentially high suppression chamber loadings.

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. 11 tis 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 uniformity of energy insertion to the pool. .

For an initial maximum suppression chamber water temperature of 90 0F sad assuming the normal com-plement of centsinecnt cooling pumps (2 LFCI pumps and 2 containment coolInc cervice wt ter pegs) ,

containment pressure is not required to maintain adequat.e net positive soction head (HPSH) for the cCre spray, LPCI and HPCI pumps. However, during an at proximately one-day period st arting a few hours after a loss-of-coolant accident, should one Rl!R loop be inoperabic and should the containneht pressure be reduced to atmospheric pressure through any means, adequate NPS!! would not be "availablie.

Since an extremely degraded condition must. cxist, the period of vulnerability to this event is re- h stricted by Specification 3 7.A.l.b by limiting the suppression pool initial temperature and the period of operation with one inoperable ElIR loop.

(1) Robbins, C. H. , " Tests of Full Scale 1/48 Segment of the Humboldt Bay. Pressure Suppression Containment," GEAP-35%, November 17, 1960.

(2) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.

l3.7 ' BASES 157

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If a loss.of coolant accident were to occur when the reactor water temperature is below 330*F containment pressure will not exceed the 62 psig design pressure, even if no condensation . were the . to occur.

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The maximum allowable pool temperature, _ whenever the reactor is above 212*F, shall be govern

- by this specification.

water tesperatures above 212*F provides additional margin above that available a .

The large amount of water that must be added or removed to cause a significant change in the suppression chamber water inventory is not likely to go un-noticed. With a daily check of water volume there is an extremely low prehability that a loss of coolant ac cident will occur simultaneously with water volume being outside of the specified range. Two indicators provide redundant readings for comparison _ (with no automatic action initiation). The provisions allowing one or both indicators out of service are consistent with the need for a redundant indicator and the frequency for checking the volume, respectively.

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The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppress,lon chamber and between the suppression chamber and reactor building during loss of coolant accident so that structural integrity of the containment is maintained.

1 The vacuum relief system between'the pressure suppression chamber and reactor building consist of two 100% vacuum relief breakers (2 parallel sets of 2 valves in series). Operation of either system will maintain the pressure differential less than 1 psig. The external design pressure is 2 psig. One valve may be out of service for repairs for a period of seven days. ThisIfperiodrepairsiscannot basedbeoncompleted the low ll) probability that system redundancy would be required during this time.

within seven days, the reactor coolant system is brought to a condition where vacuum relief is no longer required.

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 dryvell cooling operations to less than the design limit of 2 psi. The relief valves are sized on the basis of the Bodega Bay pressure suppression system tests. Since they are in series with the reactor building to suppression chamber vacuum relief valves pressure drop across these valves must be included in the evaluation of drywell negative pressures, even though there does not appear to be a mechanism for causing negative pressures in excess of the 2 psi design pressure. With eight of the ten valves in service, the differential With this additional pressure pressure across the valves =for maximum flow conditions would increase. Containment drop the total differential pressure would still be less than the 2 psi design valve.

integrity would therefore not be impaired.

In addition to the above considerations, postulated leakage through the vacuum breaker to the suppression chamber air space could result in a partial bypass of pressure suppression in the event of a LOCA This effect could potentially result in exceeding containment or a small or Intermediate steam leak.

design pressure. As a result of the leakageIt potential, the containment response has been analyzed was found that the maximum allowable bypass area for any for a number of postulated conditions.

posttuated break size was equivalent to a six-inch diameter opening.1 This bypass corresponds to a 1 Report on Torus to Drywell Vacuum Breaker Tests and Modifications for Monticellh Nuclear Generating Plant, dated March 12, 1973, submitted to Mr. D. J. Skovholt, AEC-DL, from Mr.fla,0.

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l One inch opening of any one valve or 0.1 inch opening for all ten valves, measured at the. bottom of ,

l the disc with the top of the disc at the seat. The position indication system is designed to detect j

closure within 1/8 inch at'the bottom of the disc. .

At each refueling. outage and following any sigificant maintenance on the vacuum breaker valves,

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positive seating of the vacuum breakers will be verified by leak test. The leak test is conservatively designed to demonstrate that leakage is less than that equivalent to leakage through a one-inch orifice which is about 3% of the maximum allowable. This test is planned to establish a baseline for valve performance at the start of each operating cycle and to ensure that vacuum breakers are maintained as nearly as possible to their design condition. This test is not planned to serve as a limiting condition for operation.

llh During reactor operation, an exercise test of the vacuum breakers will be conducted monthly. This test will verify that disc travel is unobstructed and will provide verification that the valves are closing fully through the position indication system. If one or more of the vacuum breakers do not seat fully as determined from the indicating system, a leak test will be conducted to verify that leakage is within the maximum allowable. Since the extreme lower limit of switch detection capability is approximately 1/16", the planned test is designed to strike a balance between the detection switch capability to verify closure and the maximum allowable leak rate. A special test was performed to establish the basis for this limiting condition. During the first refueling outage all ten vacuum breakers were shimmed 1/16" open at the bottom of the disc. The bypass area asaociated with the shimming corresponded to 63% of the maximum allowable.1 The results of this test are shown in Figure 3.7.1.

When a drywell-suppression chamber vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights at the remote test panels are designed to function as follows:

Full Closed 2 Green - On 2 Red - Off Intermediate Position 2 Green - Off 2 Red - Off Full Open 2 Green - Off 2 Red - On  !

The remote test panel consists of a push button to actuate the air cylinder for testing, two red lights.

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and two green lights for each or the ten valves. There are four independent lirit switches on ese'

• valve. The two switches controlling the green lights are adjust.ed to prov1de an indication of disc opening of less than 1/R" at the bottom of the disc. These switches are also used to activate the l

valve position alarm circuits. The two switches controlling the red lights l are adjusted to provide indication of tt: disc very near the full open position.  !

This assures that no simple f ailure will The control room alarm circuits are .euundant and f ail safe.

defeat alarming to the control room when a valve is open beyond allowable and when power to the switches fails. The alarm is needed to alert the operator that action must be taken to correct a malfunction or to investigate possible changes in valve position status, or both. If the alarm cannot be elesred due llg to the inability to establish indication of closure of one or more valves, additional testing is required.

The alarm system allows the operator to make this evaluation on a timely basis. The frequency of the testing of the alarms is the same as that required for the position indication system.

Operability of a vacuum breaker valve and the four associated indicating light circuits shall be established by cycling the valve. The sequence of the indicating lights will be observed to be that previously described. If both green light circuits are inoperable, the valve shall be considered inoperable and a pressure test is required immediately and upon indication of subsequent operation.

l If both red light circuits are inoperable, the valve shall be considered inoperable, however, no i pressure test is required if positive closure indication is present.

l The 5% oxygen concentration minimizes the possibility of hydrogen combustion following a loss of coolant accident. Significant quantities of hydrogen could be generated if the core cooling systems failed to sufficiently cool the core. The occurrence of p.1 mary system leakage following a major refueling outage or other scheduled shutdown is more probable than the occurrence of the loss of coolant accident upon which the specified oxygen concentration limit is based. Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety I

of fered without significantly red scing the margin of safety. Thus, to preclude the possibility (llf of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, when the primary systec is at or near rated operating tenperature and pressure. The 24-hoar period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration. The primary containoent is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the containment but ' air could not leak in to increase oxygen concentration. Once the con-i

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tainment is filled with nitrogen to the required concentration, no monitoring of oxygen concentration is neces sa ry. Hewever, at least once a week the oxygen concentration will be decermined as added assurance.

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Calculations of the forces en the suppression chamber and its support system indicate that the dynamic loads during a postulated design basis loss of coolant accident are dependent on the drywell to supprese sion chamber differential pressure and the suppression chamber water volume. The specifications require that the conditions assumed in the stress analysis be met after allowing sufficient time to first inert '

containment and then to establish the differential pressure. Similar provisions are allowed for the purpose of de-inerting.

change, Provisions are included for allowing special tests without-a Technical Specification pending special review and authorization by the Nuclear Regulatory Commission. The drywell to affected the chamber suppression differential differential pressure.pressure is checked each shift to assure that equipment failure has not With a check each shift, there is an extremely low probability that a loss ibnit.of coolant accident will occur simultaneously with the differential pressure being below the specified Two monitors provide redundant readings for comparison (with no automatic action initiation). The provisions allowing one or both indicators out of service are consistent with the need for a redundant indicator and the frequency for checking the differential pressure, respectively.

B, llg{.

Standby Gas Treatment System and C. Secondary Containment The secondary might result fromcontainment is designed to minimize any ground level release of radioactive materials which a serious accident.

operation, when the drywell is sealed and in service;The reactor building provides secondary containment during re 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 ttnes that primary containment is required except, however, for initial fuel loading prior to initial power testing.

The standby gas treatment system is designed to filter and exhaust the reactor building aboosphere to the chimney from the during reactorsecondary building tocontainment the environs. isolation conditions, with a minimmn release of radioactive materials One standby gas treatment system circuit is designed to auto-matically start upon containment isolation and to maintain the reactor building pressure at the design negative pressure so that all leakage should be in-leakage. Should one circuit fail to start, the redundant alternate standby gas treatment circuit is designed to start automatically. Each of the two circuits has 100% capacity.

' building atmosphere Onlyupon one containment of the two standby gas treatment system circuits is needed to cleanup the reactor isolation.

immediate threat to the containment system performance.If one systen is found to be inoperable, there is no may continue while repairs are being made. If neither circuit Therefore, reactor operation or refueling operation j is operable, the plant is placed in a condition that does not require a standby gas treatment system.

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While only a small amount of particulates are released from the primary containment as a result of the. loss of coolant accident, high-efficiency particulate filters before and after the charcoal

' filters-are specified to minimize potential particulate release to the environment and to prevent clogging of the charcoal adsorbers. 'The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than 1% bypass leakage for the charcoal adsorbers using Falogenated hydro-carbon and a HEPA filter efficiency of at least 99% removal of DOP particulates. Laboratory carbon sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions. Operation of the standby gas treatment circuits significantly different from the design flow will change the removal efficiracy of the HEPA filters and charcoal adsorbers.

If the performance requirements are met as specified, the calculated doses would be less than the guidelines stated in 10 CFR 100 for the accidents analyzed.

D. Primary Containment Isolation. Valves Double isolation valves are provided on lines penetrating the primary containment. Closure of one of the valves .in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiatic:. is required to minimize the potential' leakage paths from the containment in the event of a loss-of-coolant accident. Details of the isolation valves are discussed in Sections E.2 and 7.2 of the FSAR.

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Bases: t 4.7 A. Primary Contalraent -

The 1.e.,water in the suppression chamber is used only for cooling in the event of an accident; it is not used for normal operation; therefore, a weekly check of the temperature and -

volume is adequate to assure that adequate heat removal capability is present. For additional margin, these will be checked once per day. '

The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspec-tion of the paint during each major refueling outage, approximately once per year, assures the paint is intact and is not deteriorating. Experience with this type of paint indicates that the inspection interval is adequate. g' Because of the large volume and thermal capacity of the supprdssion pool, the volun;e ar.d temperature normally changes very slowly _ and monitoring these parameters daily is sufficient to establish sny temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently logged durini periods of significant heat addition, the' temperature trends will be closely followed so that appropriate action can be taken. The requirement for an er.ternal visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on st actural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

Visual inspection of the suppression chamber including water line regions each refueling outage is adequate to detect any changes in' the suppression chamber structures.

The primary containment preoperational test pressures are based upon the caiculated prinary

, containment pressure response in the event of a loss of coolant accident. The peak drywell pressure would be about 11 psig, which would rapidly reduce to 25 psig within 10 seconds follow-4 ing the pipe break. Following the pipe break, the suppresslon chamber pressure rises to 25 psig within 10 seconds, equalizes with drywell pressure and thereafter rapidly decays with the dry-g well pressure decay. See Section 5.2.3 HiAR.

The design pressure of the drywell and absorption chamber ic 56 psig. See Section 5.2 3 FSAR.

The design leak rate is 0 5'f,/ day at a pressure of 56 prig. As indicated above, the pressure resp'onse of the drywell and -suppression char:ber following an accident would be the same after about 10 seconds. Based on the calculated containment pressure response discussed above, the primary containment preoperational test pressures were choden. Alca, brced on the prirvary a

containment pressure response and the fact that the drywell and suppreselon charnber ibnction as a unit, the primary containment will be tested as a unit rather than the individual compo-nents separately.

4.7 BASES 1 61

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The design basis loss of coolant accident was evaluated at the primary containment maximum allowable accident leak rate of 1.5% day at 41.psig. The analysis showed that with this leak a .

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a Bases Continued:

• B. Standby Gas ' Treatment Syster, and C. . Secondary Containment. -

Initiating reactor building isolation and operation of the standby gas treatment system to maintain the design negative pressure within the secondary. containment provides an adequate test of the reactor building isolation valves and the standby gas treatment system. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment.

l system operational capability..

The. frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform.as evaluated. Standby gas treatment system inplace testing procedures will be established utilizing applicable sections of ANSI N510-1975 standard as a procedural guideline only. Redundant heaters in the standby gas' treatment system room prevent moisture buildup on the adsorbent. If painting, fire, or chemical release occurs .

such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis should be performed as required for operational use. Replacement adsorbent should be qualified according.to the guidelines of Regulatory Guide'1.52 Revision 1 (June 1976). The charcoal adsorber efficiency test procedures will allow for the removal of one representative sample cartridge. The sample will be at least two inches in'df ameter and a length equal to the thickness of the bed. If the iodine removal efficiency test results are unacceptable, all adsorbent in the system will be replaced. High efficiency particulate filters are installed before and after the charcoal filters to prevent clogging of the carbon adsorbers and to minimize potential release of particulates to the environment. _ An efficiency of 99% is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by inplace testing with 00P as the testing medium.' Any HEPA filters found defective will be replaced with filters qualified pursuant to regulatory guide position C.3.d of Regulatory Guide 1.52 Revision 1 (June 1976). Once per operating cycle demonstration of HEPA filter pressure drop, operability of inlet heaters at rated power, automatic ' initiation of each standby gas treatment systemLcircuit, and leakage tests after maintenance or testing which could affect leakage, is necessary to assure system perfonnance capability.

h.7 BASES 165 REV

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The containment is penetrated by a large number of small diameter instrument lines.. A program for l the periodic testing (see Specification 14.7.D) and examination of the valves in these lines has been developed and a report covering this program was submitted to the AEC on July 27, 1973.

The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability.

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