ML20091C865

Amend 139 to License DPR-35,changing Tech Spec by Imposing New Limit of 2 Gpm Increase,Over Any 24 H Period,Of Reactor Coolant Leakage Into Primary Containment from Unidentified Sources

ML20091C865
Person / Time
Site:	Pilgrim
Issue date:	07/29/1991
From:	Shankman S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20091C868	List: ML20091C865 ML20091C873
References
NUDOCS 9108070261
Download: ML20091C865 (14)

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t UNITED STATES

?ek3,..h'%i NUCLEAR REGULATORY COMMISSION ikddiG'g?

WASHINGTON, D.C. EMS g

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j BOSTON E0150N COMPAtW DO,CKET NO. 50-293 P1LGRIM K EL,E,Ay,f M F S,T,ATjp!1 AMENDMEllT TO FACILITY OPERATillG LICENSE u..m.

c. L License No. DPR-35 1.

The Puclear Regulatory Commission (the Comission or the NRC) has found that:

A.

The aprlication for amendment filed by the Boston Edison Conpany (the licensee) dated February 4,1985, and revised on April 10, 1901 and June 13, 1991, complies with the standards and requirements of the Atomic Energy Act of 19E4, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; F..

The facility will opcrote in conformity with the application, the provisions of the Act, and the rules and regulations of the Coranission; C.

There is reescr,able assurance: (i) that the activities authorized by this anendtent can be corducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in cerpliarct with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this atendment will r.ct be inimical to the corron defense and security or to the health e.,c safety of the public; and E.

The issvence of this an.cndrent is ir ectordance with 10 CFR Part 51 of the Commission's regelations ar.d all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technic 81 Specifica-tions as indicated in the attachrent to this license amendment, ar.d paragraph 3.B of Facility Operating License No. DPR-35 is ! ereby amended to read as follows:

Technical Specificatiors The Technical Specifications contained in Appendix A, as revised threuch Ar1endmer-t I:o.139, cre hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Tect r.ical Specifications.

9108070261 910729 PDR ADOCK 05000293 PDR p

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3.

This litense an,endntnt is effective as of its date of issuance and l

shall be irrpieniented witbin 30 days, i

FOR THE ttVCLEAR REGULATORY COMM1$510tl l

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Project Directorate 1 3 Divi $1on of Reactor Projects. 1/11 Office of Iluclear Reetter Regulation

Attachment:

Changes to tie Technic 61 Spe cificaticris Date of-Issuerte:

July 29, 1991 l

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4 ATT ACHt4E,NT.,TO,,L,1 Celi,SE, At E!!p[gi,T,!%,l}9, FACitITY OPEPATl!!C LICEllSE 110. DPR-35 DOCKFT 110. 50-293 Rep 1rce the following ) ages of the Appendix A Technical Specification $ with the 4ttached pages.

Tie revised pages are identified by Amendment nunbte and contaiti vertical lines indicating the area of change.

Rernove Insert 5b Sb 44 44 E7 57 0!

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126b' 176 126 S

143 143 144 144 h

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1.0 DU 141TIONS (Continued)

Z.

Offsite Dose Calculation a nual (ODCH) - An offsite dose calculation manual (ODCH) shall'be a manual containing the current methodology and parameters to be used for the calculation of offsite doses due to radioactive gaseous and liquid effluents, the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints, and the conductofthefadiologicalEnv#,ronmentalMonitoringProgram.

AA.

A: tion - Action shall be that part of a specification which prescribes emedial measures required under designated conditions.

i f dt?OLof_the Public - Member (s) of the public shall include all BB.

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category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the site.

CC.

Site hadAtyl - The site boundary is shown in figure 1.6-1 in the ISAR.

DD.

Radwaste Treatment System 1.

Gnienys Radwatie Treatment System - The gaseous radwaste treatment system is that system identified in figure 4.8-2.

2.

Liauid Radwaste Treatment Systt!D - The liquid radwaste treatment system is that system identified in Figure 4.8-1.

EE.

bMiomatic Primary Containment Isolation Valves - Are primary containment isolation valves which receive an automatic primary containment group isolation signal.

FF.

Pressure Boundary LeAtast - Pressure boundary leakage shall be leakage through a non-isolable fault in a reactor coolant system component body, pipewall or vessel wall.

GG, identified Leakaoe - Identified leakage shall be:

1.

Reactor coolant leakage into drywell collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or 2.

Reactor coolant leakage into the drywell atmosphere from sources which are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be Pressure Boundary leakage.

HH.

9ttid1ntified Leakaae - Unidentified leakage shall be all reactor coolant leakage which is not Identified Leakage.

See FSAR figure 1.6-1 Amendment No. 29, 113. 728, 139 Sb

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LIMITING CONDITION FOR OPERATION SURVE1LLANCE REQQlREMENT l

E.

Drvwell Leak Detection E,

Drvve11 Leak Detection The limiting conditions of Instrumentation shall be i

operation for the instrumentation functionally tested, celibrated i

that monitors drywell leak and checked as indicated in detection are given in Section Section 4.6.C.

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3.6.C.

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Stirveillance Information Readouts F.

Surveillance-Information Read;y u The limiting conditions for the Instrumentation shall be

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instrumentat W t'**

calibrated and c!...

surveillance information rencouts adicated in Table 4.2.F.

are given in Table 3.2.F.

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Amendment No. 139 57

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l Amendment No.139 65 l

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3.2 M ili (Cont'd)

The flow comparator and scram discharge volume high level components have only one logic channel and are not required for safety.

The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time.

The automatic pressure relief functi m is provided as a bactor to +Se e in the event the HPCI does nat operate.

ihe arrangement of trit tripping contacts is such as to provide this function when necessary and minimize spurious operation.

The trip settings given in the specification are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, testing or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.

Four radiaticn monitors are provided which initiate the Reactor Building Isolation and Control System and operation of the standby gas treatment system.

The instrument channels monitor the radiation from the refueling area ventilation exhaust ducts.

Tour instrument channels are arranged in a 1 out of 2 twice trip logic.

Trip settings of < 100 mr/hr for the monitors in the refueling area ventilation exhaust ducts are based upon initiating normal ventilation isolation-and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal ventilation path but rather all the activity is processed by the standby gas treatment system.

o Amendment No. 29, 133,I39 72

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LIMITING CONDITIONS FOR OPERATION S1LRyEILLANCE REOUIREMENTS 3.6.8 Coolant Chemistry (Cont'd) 4.6.8 Coolant chemistry (Cont'd) 3.

For reactor startups and for 3.

a. Hith steaming rates of 100,000 i

the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after pounds per hour or greater, a placing the reactor in the rer.ctor coolant sample shall oower operating condition, the

'oe taken at least every 96 following limits shall not be hours and analyzed for exceeded:-

chloride ton content.

Conductivity.

10 pmho/cm

b. When all continuous Chloride ion.

0.1 ppm conductivity monitors are inoperable, a rtactor coolard 4.

Except as specified in 3.6.8.3 sample shall be taken at least above, the reactor coolant daily and analyzed for water shall not exceed the conductivity and chloride ton following limits when operating content.

with steaming rates greater i

than or equal to 100,000 pounds per hour:

Conductivity.

10 pmbo/cm Chloride. ton.

1.0 ppm 5.

If Specification 3.6.B cannot be met. an orderly shutdown shall be initiated and the reactor shall be in Hot I

Shutdown within 24 hrs, and Cold Shutdown within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

3.6.C. Coolant igAMgt-4.6.C Coolant Leakaat Any time irradiated fuel is in Any time irradiated fuel is in the reactor vessel and coolant the reactor vessel and coolant temperature is above 212'f. the temperature is above 212'f.

following limits shall be the following surveillances observed:

shall be performed:

1. Operational Leakaae

1. Oggutional Leakagg

.a.-

Reactor. coolant. system Demonstrate drywell leakage is leakage shall be limited to:

within the limits specified in 3.6.C.1 by monitoring the

)..No Pressure Boundary coolant leakage detection Leakage systems required to be

-2.

15 gpm Unidentified operable by 3.6.C.2 at least Leakage; once every 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> $.

3.

125'gpm Total Leakage l

averaged over any 24 l

hour period.

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. Amendent No.- 42.139 125

UMillRG10hD.U10'iS T0.LOHE110N SEE1LLANCE RE0WREHENTS 3.6.C.1 Opnttiqul_Letkaag (Cont'd) 4.

52 gpm increase in Unidentified Leakage within any 24 he r period when in RUN mo e.

b.

With any reactor coolant system leakage greater than

'he limits of 2. and/or ',

above, reduce the leakage to within acceptable limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

With any reactor coolant system leakage greater than the limits of 4. above, identify the source of leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, d.

When any Pressure Boundary Leakage is detected be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and be in Cold Shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. Leakace Detection Systems 2.

Le3Lagg_QttnLinn Systemt a.

The following reactor The following reactor coolant

. coolant system leakage leakage detection systems shall detection systems shall be be demonstrated Operabic:

Operable:

a.

For each required drywell 1.

One drywell sump sump monitoring system monitoring system, and perform:

.either 1.

An instrument functional test at least once per 31 days, and Am(ndment No. 139 I?Sa n

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LIHfTING CONDITIONS FOR OPERAf10N SURVEILLANCE REOUIREMENTS 3.6.C.2 udnt_ Detection Systen 4.6.C.2 dakaoe Detection Systems (Cont'd)

(Cont'd) 2.

One channel of a drywell 2.

An instrument channel atmospheric particulate calibration at least once radioactivity monitoring per 18 rnonths.

system, or b.

For each required drywell 3.

One channel of a drywell atmospheric radioactivity atmospheric gaseous monitoring system perform:

radioactivity monitoring system.

1.

An instrument che:k at least once per day, b.

1.

At least one drywell sump monitoring system shall 2.

An instrument functional be Operable; otherwise, test at least once per 31 be in Hot Shutdown within days, and the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the 3.

An instrument channel following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, calibration at least once per 18 months.

2.

At least one gaseous or particulate radioactivity monitoring channel must be Operable; otherwise, reactor operation may continue for up to 31 days provided grab sargles are c,btained and analyzed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

With no required leakage detection systems Operable, be in Cold Shutdown within 24

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hours.

Amendment No.139 125b

U NITING (Q@ITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.6.0. Safety and Relief valves-4.6,0.

111tfyandReliefValves

1. During reactor power operating 1.

At least one safety valve and conditions and prior to reactor two relief / safety valves shall startup from a Ccid Condition, be checked or replaced with l

or whenever reactor coolant bench checked valves once per pressure is greater than 104 operating cycle. All valves psig and temperature greater will be tested every two cycles.

than 340'F both safety valves and the safety modes of all 2.

At least one of the relief valves-shall be operable, relief / safety valves shall be The nominal setpoint for the disassembled and inspected each relief / safety valves shall be refueling outage.

selected between 1095 and 1115 psig. All relief / safety valves 3.

Whenever the safety relief shall be set at this nominal valves are required to be setpoint 1-11 psi.

The safety operable, the discharge pipe valves shall be set at 1240 temperature of each safety psig i 13 psi.

relief valve shall be logged daily.

2. If Spe:ification 3.6.D.1 is not met an orderly shutdown shall 4.-

Instrumentation shall be beInitiatedandthereactor calibrated and checked as coolant pressure shall-be below indicated in Table 4.2.F.

104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Note: Technical Specifications 5.

Notwithstanding the above, as a 3.6,0.2 - 3.6.0.5 apply only minimum, safety relief valves when two Stage Target Rock SRVs that have been in service shall are installed.

be tested in the as-found condition during both Cycle 6

3. If the temperature of any and Cycle 7.

safety relief discharge pipe exceeds 212'r during normal reactor power operation for a period of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an engineering evaluation shall be performed justifying continued operation for the corresponding temperature increases.

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- Amendment No. 42. 56. EE,133,139 126

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4 3.6.C and 4.6.:

(pole-t Leakagg Allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to makeup coolant system leakage in the eveht of loss of offsite a-c power. The ncrmally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were also considered in establishing the limits.

The behavior of cracks in piping systems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary Coolant System Rupture Study (the Pipe Rupture Study). Hork utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth. This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly.

Hmver, the establishment of allowable unidentified leakage greater than that gi m in 3.6.C on the basis of the data presently available would be premature because of uncertainties associated with the data.

For leakage of the order of 5 gpm, as specified in 3.6.C. the experitnantal and analytical data $Uggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propogation.

Leakage less than the magnitude specified a.a be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time the plant should be shut down to allow further investigation and corrective action.

Verification of the integrity of the reactor coolent system (3.6.C.I.a.1.: No Pressure Boundary leakage) is provided during RPV Class I system hydrostatic

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and leak tests conducted to meet section 3/4.6.G: Structural Integrity (ASHE Code,Section XI, IHA 5000, and IHB 5000.)

1 Two leakage collection sumps are provided inside primary containment.

i Identified leakage is piped from pump seal leakoffs, reactor vessel head flange seal leakoff, selected valve stem leakoff including recirculation loop and maia steam isolation valves, and other equipment drains to the drywell equipment drain sump.

The second sump, the drywell floor drain collection sump receives leakage from the drywell coolers, control rod drives, other i

valve stems and flanges, floor drains, and closed cooling water system drains. Drainage into the drywell floor drain sump is generally considered Unidentified Leakage.

Both sumps are equipped with level and flow monitoring equipment to alert operators if allowable leak rates are approached.

l A drywell sump monitoring system, as required in 3.6.C.2, consists of one equipment sump pump and one floor drain sump pump, plus associated i

instrumentation, flow integrators, one for the equipment drain sump snd Amendment No. IP 143

MELS:

3.6.C and 4 6tC Cnt'"1t_ LeALMtt (Continued) another for the floor sump, comprise the basic instrument system, and are used to record the flow of liquid from the drywell sumps. A manual system whereby the time interval between sump pump starts is utilized to provide a back-up to the flow integrators if the instrumentation is found to be inoperable.

This time interval determines the leakage flow because the capacitv of the pump is known.

The capacity of each of the two drywell floor sump pumos is 50 gpm and the capacity of each of the two drywell equipment sump pumps is also 50 gpm.

Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.

In addition to the sump monitoring of coolant leakage, airborne radioattivity levels of the drymell atmosphere is monitored by the Reactor Pressure Boundary Leak Detection System.

This system consists of two panels capable of monitoring the primary containment atmorphere for particulate and gaseous radioactivity as a result of coolant 15.'ks.

The 2 gpm limit for coolant leakage rate increase within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is a limit specified by the NRC in Generic Letter 88-01:

"NRC Position on IGSCC in L dt Austenitic Stainless Steel Piping".

This limit applies only curing the RUN mode to accommodate the expected coolant leakage increase during pressurization.

The total leakage rate consists of all leakage, which flows to the drywell equipment drain sump (Identified leakage) and floor drain sump (Unidentified leakage).

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Arendment No. 139 144 l

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