ML20024G347

Proposed Tech Specs Re Drywell to Suppression Chamber Differential Pressure

ML20024G347
Person / Time
Site:	Monticello
Issue date:	11/05/1976
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G337	List: ML20024G336 ML20024G338 ML20024G347
References
NUDOCS 9102110348
Download: ML20024G347 (9)

Text

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CD p 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIRDD:'ES co 00 Ob M 6. If specifications 3.7. A.1 through 3. 7. A.5 cannot o9 be met, the reactor shall be placed 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 />.

o8o 7.

t3 fu t.n DRYWELL-SUPPRESSION CHAMBER DIFFERENTIAL PRESSURE 7. DRYWELL-SUPPRESSION CHAMBER DIFFERENTIAL PRESSURE

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a. Drywell pressure shall be maintained 2 1.0 psi The differential pressure between the dryvell and above the suppression chamber pressure e apt supprersion chamber shall be logged once per shift. h as specified in 3. 7 A. 7.b and c.

b. 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 the reactor in the run mode following a shutdown, the drywell pressure must be raised to 21.0 psi above the suppression chamber pressure 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

< 1.0 psi for a maximum of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> during required operability testing of the HPCI system pump, the RCIC syst e pump, and the drywell- .

pressure suppression chamber vacuum breakers.

d. If specification 3.7.A.7 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the hot shutdown condi-tion 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 3.7/4.7 REV

. - _ - . ..~ _. _ a 2 .. 4 3.0 LIMITING CONDITIONS FOR OPERITION 4.0 SURVEILIANCE REQUIRDHDrrS I

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

1. Two separate and independent standby 1. At least once per month, initiate from gas treatment system circuits-shall be the control room 3500 cfm Q 107.) flow

• operable at all times when secondary through both circuits of the standby contairunent integrity is required, gas treatment systest. In addition: .

except as specified in sections

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 system

a. After one of the standby gas circuit is made or l'ound to be in-treatnent 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 M permissible only during control room 3500 cfm Q 107.) flow the sue;eedi::g seven days, provided through the operable circuit of the that all active components .in the standby gas treatment system.

other standby gas treatsnent 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).

3.7/4.7 148 REV w

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,n BASES CONTINUED:

system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure. n e 24-hour period to provide inerting is judged to be sufficient to j

perform the leak inspection and establish the requimd oxygen concentration. %e primary containment is l normally slightly pressurized during per'ods of reactor cperation. Nitrogen used for inerting could leak out of the containment but air could not leak in to increase oxygen concentration. Once the containment is .

l filled with nitrogen to the required concentration, no monitoring of oxygen concentration is necessary.

However, at least once a week the oxygen concentration will be determined as added assurance.

Calculations of the forces on the suppression chamber and its support system indicate that the dynamic loads 4 during a postulated design basis loss of coolant accident are dependent on the drywell to suppression chamber differential pressure and the suppression chamber water volme. n e specifications require that the conditions assumed in the stress analysis be met after allocing sufficient time tc first inert containment and then to establish the differential pressure. Similar provisions are all - ed for the purpose of de-inerting.

B. Standby Gas Treatment System and C. Seconnary Containment The secondary containment is designed to minimtre any ground level release of radioactive materials which might result from a serioua eccMent. He reactor building provides secondary containment during reactor operation, when the drvve is sealed and in service; the reactat building provides primary containment when Ae rt actor is shutdown and the drywell is open, as during refue1N 3. Because the secondary (c

• dinment is an integral part of the complete con-tainment system, secondary containment is required .t all times that primary contaiment is required except, however, for initial fuel loading prior to initial power testing.

The standby gas treatment system is desigu J f.o filter and exhaust the reactor building atmo-sphere to the chimney during secondary conr: araeut isolation conditions, with a minimum release h

of radioactive materials from the reactor b sildinL to the environs. One standby gas treatment system circuit is designed to automatically s . art r upon containment isolation and to maintain the reactor building pressure at the design aegative pressure so that all leakage .should be in-leakage. Should one circuit fail to start, the ceduncat alternate stardby ; gas treatment circuit is designed to start automatically. Each of the t m M 'cuits has 1007. capacity. Only one of the two standby gas treatment system circuits is needed to cleanup the reactor building atmo-sphere upon containment isolation. If one system is found to be inoperable, there is no immediate t$reat to the containment system performance. Therefo: e, reactor operation or refueling operation may continue while repairs are being made. If neither circuit is oper-able, the plant is placed in a condition that does not require a standby gas treatment system.

159 3.7 BASES REY

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Bases Continued:

The acceptable values for local leak rate tests have been specified in terms of standard i cubic feet per hour (sef/hr) for purposes of clarity. Following is the list of equivalent values given in terre of an allowable percentage of the allowable operational leak rate -

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. i 17.2 sef/hr = 5% L to

@ 41 psig _

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l 34.4 scf/hr = 10% L to G 41 psig 103.2 scf/hr = 30% L to

@ 41 psig where Lt o = .75 Lt (the maximum allowable leak rate) i and L t = 1.2 weight percent of the contained air at the test pressure of 41 psig.

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Results of loss of coolant accident analyses indicate that fission products would not be

' released directly to dhe environs 'vecause of leakage through the main line isolation valves l

due to holdup in the steam gysten complex. Although this effect shows that an adequate l margin exists with regard to release of fission products, the results of leak tests on the main steam line isolation valves will be closely followed in order to deteentne the adequacy of these valves to perform their intended function. ll)

Monitoring the nitrogen makeup requirenents of the f nerting system provides a method of observing .ak rate trends and would detect gross leaks in a very short time. This equip-ment must be periodically removed from service for test and maintenance, but this out-of-service tLne will be kept to a practical minimum.

I Drywell and suppression chamber pressure is continually recorded. Logging the drywell to suppt ssion chamber differential pressure each shift provides a method of chaerving slow trenui. Abrupt reductions would only result fran equipment other than the nitrogen recirculats on systen which is provided to maintain the differential.

164 REV 4.7 BASES -

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EMIIBIT C Specific Responses to October 4, 1976 Letter from D L Ziemann to L 0 Mayer i

1. Drywell - Torus Differential Pressure I.A. Technical Specification Changes Proposed Technical Specification pertaining to the dry-well-torus differential pressure, along with proposed Bases, is included as Exhibits A and B of this submittal.

l 1 I.B. Operating Procedures. System Changes and Valve Lineup

• 5 i Changes to Laplement Differential Pressure Requirement This subject is addressed in the April 14, 1976 License Amendment No 18 to DPR-22 which issued the Technical

-Specification Changes necessary to Unplement the Monticello nitrogen recirculation equipment. The attached Figure C-1 shows the nitrogen recirculation systen identifying the newly added- equipment. Containment isolation valves are identified by their unique numbers. The License Amendment acknowledged the addition of a normally open-isolation valve, CV-7440. It also acknowledged that three isola-tion valves, the torus vent bypass valve, CV-2384, and the drywell purge inlet valves, A0-2381.and A0-2377, were 4 changed from normally closed to normally open.

The nitrogen recirculation system removes nitrogen from 1

.tbe torus via CV-2384 and CV-7440 using one of two 100%

capacity blowers. The blower will discharge nitrogen through the drywell purge inlet line to the drywell via A0-2381 and A0-2377. Recirculation-flow is manually ad-justed to maintain the drywell pressure greater than or equal to 1.0 psi above the torus pressure. The system up j, to the second containment isolation valve in the suction and discharge lines satisfies all design, quality assurance,

))l, and testability requirements set forth in 10 CFR 50.

The added isolation valve, CV-7440, receives motive power from instrument air, failing closed on loss of air. The a

circuitry used is the same as that used througaout the plant on outboard air operated isolation valves. A solenoid valve-is.cnergized.to open the ve~ve. The solenoid is de-energized by an isolation signal or loss of the instrument ,

bus associated with outboard valvae. A detailed discussion of the scheme used'can be found in Monticello FSMR Section 7.

There is no potential for high energy line break pipe whip or missiles in the vicinity of CV-7440.

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EXHIBIT C 2

I.C. Differential Pressure Instrumentation The attached Figure C-2 shows the instrumentation available to measure the drywell to torus differential pressure.

The pressure sensing lines to the main transmitters are l -

supplied with local indicators PI-3050 and PI-3051 which have a 0 to 100 psig range. The pressure signals used to h monitor the 1 pet drywell to torus differential originate l at pressure transmitters PT-2994 A and B which have i % accuracy over the 13 to 17 psia range. These sig-nals are displayed in the control room on a dual pen 4" recorder, PR-2994, which also has an accuracy of j i % of span. (The drywell pressure pen can be switched to a wide range scale in the event of a drywell pres-

surization; pressure transmitter PI-7348 provides a 0 1 to 30 psig input to the recorder.)

The operator maintains the 1 psi differential by main-

! taining at least one inch separation between pens. The ultimate accuracy in maintaining the 1 pai lies in the

! ability to read the chart. The chart-reading error is I within 1/16 inch which corresponds to 0.06 pai.

, The single channel of instrumentation available to monitor the drywell-torus differential pressure is independent of the pressure switches used for ECCS initiation signals.

, II. Torus Water Level II.A. Allowable Range of Torus Level On October 12, 1976 a report was subciitted to Mr V Stello

(USNRC) by Mr L 0 Mayer (NSP) entitled, "Monticello Nuclear Generating Plant Supplement to Short Term Program, Plant Unique Torus Support and Attached Piping Analysis".

The report concludes that, "...the Short Term Program

{; Criteria is satisfied for all structural components of the torus support structure and the piping systems

attached to the torus, for all water levels within the current Technical Specification limits." Therefore, there are no new requirements imposed on torus water l 1evel beyond those addressed in the initial plant licensing process and the Technical Specifications.

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l EKHIBIT C II.B. Torus Level Instrtamentation The torus level instrumentation available is shown in the attached Figure C-2. This is a single channel system initiated by level transmitter LT-2996. Level indication in the main control room is provided by LI-2996; high and low level alarms in the main control room are initiated by level switches LS-2996 A and B.

Technical Specification 3.7.A.1 requires torus water volume to be a minimte of 68,000 cubic feet and a maximum of 77,970 cubic feet. For the convenience of f i operators, these volumes are converted to levels about j a nceinal target level of 70,950 cubic feet. (At I this volume, one inch of level is equivalent to approxi- e mately 740 cubic feet.) The Technical Specification i- limits can be expressed as approximately -3.0" to I

+ 10,5" about the nominal level when the 1 psi dif-

ferential between drywell and torus is established and -4.0" to + 9.5" when the pressures are equalized.

Level indicator LI-2996 has an accuracy of 27. across

, its 4 " display of the range of -15" to + 15". Level g switches LS-2996 A and B provide high and low torus

level alarms at the levels of -1" and -2 3/4", re-

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spectively. (This range is selected to keep the torus

, level, and therefore the dynamic load resulting from _

a reactor blowdown, to a minimum.) Both the level indicator and the switches are fed by level transmitter LT-2996. The level transmitter and switches lire 1 \7.

devices over the 30" span. The combined error of the alarm sutpoint is therefore 1 0.2". Through the indicator and the alarms, the operator has available contint'ous monitoring of torus water level.

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, NRC DISTRIBUTION ron PART 50 DOCKET MATERIAL ROM: oATE oF ooCuu T Mr Stello Northenn States Pwr Co 76 Minneapolis, Mn L U Mayer oATc nectiveo ,

GLgTren WNoronizco enoe Neut r onne Nu.use n or coeits stetiveo GoMiolNAL S uNC LAS$lelE o Ocoey 3 signed DESCRIPitoN ENCLOGuRC Ltr notarized 11-10-76....*trans the following Amde to 01/ Change to Tech Specs:[ Consisting of revisions with _ regard to drywell to sup-pression chamer differential pressure re our 10-4-76......

(40 cys enc 1 ree'd) 1;h N.C .* - - - -

PLANT NAME: Monticello -

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SAFETY FOR ACTION /INFCRM ATION ENYTRn 11-11-76 chf ASSIGhT.D AD: McTee n AD.

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