Text

Revised Proposed Tech Spec Re Reactor Coolant Leakage Detection Sys
	ML20072T540
Person / Time
Site:	Pilgrim
Issue date:	04/10/1991
From:	; BOSTON EDISON CO.
To:	;
Shared Package
ML20072T537	List: BECO-91-052, Revised Application for Amend to License DPR-35,changing Tech Spec Re Reactor Coolant Leakage Detection Sys at Plant; ML20072T540;
References
	NUDOCS 9104180169
	Download: ML20072T540 (27)
	v • d • e

_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - __ ___ _

Revised Reauest for a ProoostLChgge to the Reactor Coolant Leakaae Detection Systems Technical Soecifications e

Attachment B Revised Pages:

Sb 44 57 65 72 125 126 143 144 Added Pages:

125a 125b f

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9104180169 910410 PDR ADOCK 05000293 p PDR

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1.0 DEFINITIONS (Continued)

Z. O_ffsite Dose Calculation Manual (00fM - An offsite dose calculation manual (0DCM) shall be a manual containing thri 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 conduct of the Radiological Environmental Monitoring Program.

AA. Action - Action shall be that part of a specification which prescribes remedial measures required under designated conditions.

BB. Member (s) of the Publici - Member (s) of the public shall include all persons who are not occupationally associated with the plant. This 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 Boundarvi - The site boundary is shown in Figure 1.6-1 in the FSAR.

DD, Radwaste Treatment System

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

2. Liauid Radwaste Treatment System - The liquid radwaste treatment system is that system identified in figure 4.8-1.

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

FF. Pressure Boundary leakaae - Pressure boundary leakage shall be leakage i through a non-isolable fault in a reactor coolant system component body, l pipewall or vessel wall. 1 I

GG. Identified Leakaae - Identified leakage shall be: 1 I

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

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

HH. Unidentified Leakagg - Unidentified leakage shall be all reactor coolant l leakage which is not Identified leakage. I l

See FSAR Figure 1.6-1 Amendment No. 89, 773, 128, 5b

l

. . j l

l LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT E. Drvwell Leak Detection E. Drywell Leak Detection The limiting conditions of Instrumentation shall be operation for the instrumentation functionally tested, calibrated that monitors drywell leak and checked as indicated in detection are given in Section l Section 4.6.C. l 3.6.C. l F. Surveillance Information Readouts F. Surveillance Information Readouts The limiting conditions for the Instrumentation shall be instrumentation that provides calibrated and checked as surveillance information readouts indicated in Table 4.2.f.

are given in Tabl' 3.2.F.

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l Amendment No. 89, 44 l l l L l

4 l

i This page is intentionally left blank, l I

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i

' Amendment No. 57

l This page is intentionally left blank. l i

Amendment No. 65

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3.2 MSIS (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 L.PCI to operate in time. The automatic pressure relief function is provided as a backup to-the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as-to provide this function when necessary and minimize j spurious operation. The trip settings given in the specification are l j- 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.

1 Four radiation 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 1 refueling; area ventilation exhaust ducts. j four instrument channels are arrangtd in a 1 out of 2 twice trip logic.

Trip settings'of < 100 mr/hr for the monito'rs 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 i

Building via the normal ventilation path but rather all the activity is processed by the standby gas treatment system,

_l_

l 7

l Amendment No. 89. 133, 72 ,

LIMITING CONDITIONS FOR OPERATION SURVfLLLANCE REOUIREMENTS 3.6.8 Coolant Chemistry (Cont'd) 4.6.B Coolant Chemistry (Cont'd)

3. For reactor startups and for 3. a. With steaming rates of 100.000 the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after pounds per hour or greater, a placing the reactor in the reactor coolant sample shall power operating condition, the be taken at least every 96 following limits shall not be hours and analyzed for exceeded: chloride ion content.

Conductivity . 10 pmho/cm b. When all continuous Chloride ion . 0.1 ppm conductivity monitors are inoperable, a reactor coolant

4. Except as specified in 3.6.B.3 sample shall be taken at least above, the reactor coolant daily and analyzed for water shall not exceed the conductivity and chloride ion following limits when operating content, with steaming rates greater than or equal to 100,000 pounds per hour:

Conductivity . 10 pmho/cm Chloride ion . 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 Shutdown within 24 hrs, and Cold Shutdown within the next B hours.

3.6.C Coolant Le.c!;An 4.6.C Coolant Leakage Any time irradiated fuel is in l Any time irradiated fuel is in I the reactor vessel and coolant I the reactor vessel and coolant I temperature is above 212'F, the l temperature is above 212*F, I following limits shall be I the following surveillances l observed: l shall be performed: 1 I I

1. Ooerational Leakaae l 1. Ooerational Leakaae l I I

a. Reactor coolant system l Demonstrate drywell leakage is l 1eakage shall be limited to:l within the limits specified in l l 3.6.C.1 by monitoring the 1

1. No Pressure Boundary I coolant leakage detection l Leakage I systems required to be l

2. 15 gpm Unidentified I operable by 3.6.C.2 at least l l Leakage I once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . l

3. 125 gpm Total Leakage l averaged over any 24 l hour period. l Amendent No. 42, 125

LIMITING CONDITIONS FOR OPERATION , SURVEILLANCE REOUIREMENTS 3.6.C.1 Ooerational Lul m (Cont'd) l 1

4. 12 gpm increase in i Unidentified Liakage I averaged over a.sy 24 I hour period when in RUN l mode. l I

b. With any reactor coolant I system leakage greater than l the limits of 2. and/or 3., l above, reduce the leakage to l within acceptable limits l within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at I least Hot Shutdown within l the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in I Cold Shut' lown within the I following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l I

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

d. When any Pressure Boundary l Leakage is detected be in at I least Hot Shutdown within l the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and be in I Cold Shutdown within the l next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 1 I

2. Leakaae Detection Systems l 2. Leakaae Detection Systemt I

a. The following reactor i The following reactor coolant coolant system leakage I leakage detection systems shall detection systems shall be i be demonstrated Operable:

Operable: l l a. For each required drywell

1. One drywell sump l sump monitoring system monitoring system, and I perform:

either l

1. An instrument functional test at least once per 31 days and Amendment No. 125a

f LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.6.C.2 Leakaae Detection Systems l4.6.C.2 Leakaae Detection Systems l (Cont'd) l (Cont'd) I l l l

2. One channel of a drywell l 2. An instrument channel l I

atmospheric particulate l calibration at least once l radioactivity monitoring l per 18 months. I system, or l l l b. For each required drywell l

3. One channel of a drywell I atmospheric radioactivity I atmospheric gaseous I mmitoring system perform: l radioactivity monitoring l l system. I 1. An instrument check at l l least once per day, I

b. 1. At least one drywell sump l l monitoring system shall l 2. An instrument functional l be Operable; otherwise, I test at least once per 31 l be in Hot Shutdown within l days, and I the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in l l Cold Shutdown within the l 3. An instrument channel l following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . I calibration at least once l l per 18 months, l

2. At least one gaseous or 1 particulate radioactivity l monitoring channel must I be Operable; otherwise, l reactor operation may I continue for up to 31 l days provided grab l samples are obtained and I analyzed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , l or be in Hot Shutdown I within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I and in Cold Shutdown l within the following 24 l l hours. l l l l c. With no required leakage l l

detection systems Operable. l be in Cold Shutdown within 24 l l hours. l l

l l

l Amendment No. 125b l

LIEITING CONDITIONS FOR OPERATION SURV ELLANCE REOUIREMENTS I I 3.6,0. Safety and Relief Valvei 4.6.D. Safety and Relief Valves

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 Cold Condition, be checked or replaced with or whenever reactor coolant bench checked valves once per pressure is greater than 104 operating cycle. All vilves 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 i 11 psi. The safety operable, the discharge pipe valves shall be set at 1240 temperature of each safety psig 13 psi, relief valve shall be logged daily.

2. If Specification 3.6.D.1 is not met, an orderly shutdown shall 4. Instrumentation shall be be initiated and the reactor calibrated and checked as coolant pressure shall be below indicated in Table 4.2.F.

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

Note: Technical Specifications 5. Notwithstanding the above, as a 3.6,0.2 - 3.6.D.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'F during normal reactor power operation for a period of greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , an engineering evaluation shall be performed justifying continued operation for the corresponding temperature increases.

Amendment No. 42, 56, 88, 123, 126

BASES:

3.6.C and 4'6.C.

Coolant Leakaae Allowable leakage rates of coolant from the reactor coolant system have been based on--the pred_icted and experimentally observed behavior of cracks in pipes and on the ability _ to makeup coolant system leakage in the event of loss of offsite a-c power. _The normally 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 (th6 Pipa Rupture Study). Work 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.

However, the establishment of allowable unidentified leakage greaier than that l

given 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 experimental and analytical- data suggest 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 can 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- I to allow further investigation and corrective action.-

-Verification of the integrity of the reactor coolant system (3.6.C.1.a.l.: No ' I Pressure' Boundary Leakage) is provided during RPV Class I system hydrostatic l and leak tests conducted to meet section 3/4.6.G: Structural Integrity-(ASHE l Code,Section XI, IHA 5000, and IHB {000.)- l I

Two: leakage collection sumps are provided inside primary containment. I Identified-leakage is piped from pamp seal leakoffs, reacter vessel head I

flange seal leakoff, selected valve stem leakoff including recirculation loop- l and main steam isolation. valves, and other equipment drains to the drywell i

equipment drain sump. The second sump, the drywell floor drain collection I sump receives ~1eakage from-the drywell coolers, control rod drives, other 1

= valve stems and flanges, -floor drains, and closed cooling water system l drains. Drainage into the drywell _ floor drain sump is generally considered 1

- Unidentified Leakage. Both sumps are equipped with level and flow monitoring -l equipment .to alert operators if allowable leak rates are approached. I I

A drywell sump. monitoring system, as required in 3.6.C.2, consists of one i equipment sump pump and one floor _ drain sump pump, plus associated l-instrumentation. Flow integrators, one for the equipment' drain sump and i Amendment No. 143 l

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3.6.C and 4.6.C Coolant Leakaae (Continued) another for the floor sump, comprise the basic instrument system, and are used I l to record the flow of liquid from the drywell sumps. A manual system whereby l the tin,e interval between sump pump starts is utilized to provide a back-up to I the flaw integrators if the instrumentation is found to be inoperable. This l time interval determines the leakage flow because the capacity of the pump is l know'i. l Th9 capacity of each of the two drywell floor sump pumps is 50 gpm and the l ct.pacity of each of the two drywell equipment sump pumps is also 50 gpm. I Rimoval of 25 gpm from either of these sumps can be accomplished with considerable margin.

In addition to the sump monitoring of coolant leakage, airborne radioactivity I levels of the drywell atmosphere is monitored by the Reactor Pressure Boundary I Leak Detection System. This system consists of two panels capable of I monitoring the primary containment atmosphere for particulate and gaseous I radioactivity as a result of coolant leaks. I I

The 2 gpm limit for cooltnt leakage rate increase over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is a 1 limit specified by the NkC in Generic Letter 88-01: "NRC Position on IGSCC in l BHR Austenitic Stainless Steel Piping". This limit applies only during the l RUN mode to accommodate the expected coolant leakage increase during l pressurization. l l

The total leakage rate consists of all leakage, which flows to the drywell l equipment drain sump-(Identified leakage) and floor drtin surrp (Unidentified I leakage). I Amendment No. 144

i

- !l Revised Reauest for a ProDosed Chanae to the i Reactor Coolant Leakaae Detection Systems j Technicai Soecifications i I

J

. Attachment C ;

i Marked-up Pages: l-

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Sb 44 57 65 72 125 126 143 144 Inserts "A" through "D"

E 9 e

. l .0 ' DEFINITIONS (Continued)

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^ 2. Offsite Dose Calculation Manual (ODCH) - An of fsite dose calculation \

manual (0DCM) shall be a inanual containing the current methodology and N parameters to be used for the calculation of offsite doses due to N i

, radioactive gaseous and liquid effluents, the calculation of gaseous and x liquid effluent monitoring instrumentation alarm / trip setpoints, and the -

3 conduct of the Radiological Environmental Monitoring Program. N AA. Action - Action shall be that part of a specification which prescribes remedial measures required under designated conditions.

BB, Member (s) of the PublicI - Member (s) of the public shall include all persons who are not occupationally associated with the plant. This 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 BoundarvI - The site boundary is shown in Figure 1.6-1 in the FSAR.

DD, RLdwaste Treatment System

1. Gaseous Radwaste Treatment System - The gaseous radwaste treatment system is that system identified in Figure 4.8-2.

s

2. l.iauid Radwaste Treatment System - The liquid radwaste treatment system is that system identified in Figure 4.8-1.

EE. Automatic Primary Containment Isolation Valves - @ M imary containment isolation valves which receive an automatic primary containment group isolation signal.

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U i See FSAR Figure 1.6-1 RwkN>n MO Amendment No. 89, il31M- Sb

' LIMITING CONDITION FOR OPERATION SURVEILLANCE REQ'JIREMENT

k. Dry = ell Leak Detection b

I s, E. Drywell teat Detection E.

', . ./ The limiting conditions of Instrumentation'shaU be ,

q operation for the instrumentation functionally tested,lalibrated '

that monitors drywell-leat wj n , and checked as indicated in ht46

,---detection are given in +etM I i / tri -

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7. Survd+ra'nclNJ nformatTBR Readopt , F. Surveillance Information Readouts The limiting conditions for the' Instrumentation snall be instrumentation that provides calibrated and checked as surveillance information readouts indicated in Table 4.2.F.

are given in Table 3.2.F.

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. MINIMJM TEST AND CALIBRATION F1TERUENCY FOR DRWELL IZAK Dt.nx110N N N) : .. ,

.j .Instrueent Channel Instrument htnetional Test Calibration Frequency Instrument Check ' . \) .- .!

1) Equipment Drain Sump Flow Integrator (1) Once/3 months Once/ day. .

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2) Floor Drain Sump Flow Integrator. (1) .

'Once/3 months Once/ day:

3) . Air Sampling System' (1)' Once/3 ennthe Once/ day

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INSTRUMENTATION THAT PONITORS DRYWELL LEAK DETECTION N '\ ,

N u .timum 8 of . J

]l J Operable Instrument Action Instrument /

) [ Channele Equipment Drain Sump A ,

- 1 j Flow Integrator g .

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Floor Drain Sump ] ,

." Flow Integrator --

\ l Air Sampling System A ) ,

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The twe (2) flow integrators, one for the equiree@in sump end the other for the i'

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floor t rain sump, comprises the basic instrument system.

N An alternate system to determine the lea' age flow is manaal system whereby the time f N -

- betwmn sump pump starts is monitored. This tine int rval will determine the leakage ! l

( - firve because the volume of the sucp will be known. ,t

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2. Action -

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y,\ A. Whenever the reactor coolant leskage systent is required to be operable, there shall be

\ one operable system, or the reactor shall be placed in a Cold Shutdown Condition vii.h-in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Refer to Specifiestion 3.6.C.

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t 3.2 BASES (Cont'd)

The flow comparator and scram discharge volume high level components have only one logic r.hannel 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 function is providea as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the 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 radiation 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.

Four instrument channels are 'tranged 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 norraal 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 r as,tt_eJLtment sys.itm Flow integrators are used to record the integrated flow of liquid from )

the drywell sumps. The alarm unit in each integrator is set to ,

annunciate before the values specified in Specification 3.6.C are /

exceeded. A system whereby the time interval to fill a known volume will be utilized to provide a back-up to the flow integrators. An air [)

sampling system is also provided to detect leakage inside the primary /

containment.

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Revi4iM4b Amendment No. 89, 9 3 72

3.2 ILA1LS (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 function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the 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 radiation 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.

Four 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.

.fN/' '

Flow integrators are used to record the integrated flow of liquid from the drywell sumps. The alarm unit in each integrator is set to annunciate before the values specified in Specification 3.6.C are exceeded. A system whereby the time interval to fill a known volume will be utilized to provide a back-up to the flow integrators. An air sampling system is also provided to detect leakage inside the primar.

contd nment.

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h.6s kn 44L Amendment No. 29,133 72

1:Mb71NCCOND1710NTOROPIRATION SURVIILLAN0! Plo"!P_NIN*S 3.6.B Coelant Cherts:rv (Cont'd) 4.6.3 Coelan: Chemistrv (Cont'd) s

3. For reactor startups and for the 3. a. Vith steaming rates of 100,000 first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter placing the pounds per hour or greater, a b reactor in the power operating reactor coolant sample shall be coodicion, the f onoving limits taken at least ever/ 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months

, shan not be exceeded, and analyzed for chloride ion content.

Conductivity.. 10 u=ho/c=

b. When all. continuous coeductivity Chloride ion.. 0.1 pym monitors are inope:1ble a reactor coolant sacple shall be taken at least daily and analysed for

. conductivity and chieride icn content..

4 I.xcept as specified in 3.6.B.3 above, the reacter coolant water shan not exceed the fe noving .

limits vnen operating with steam-6 ing rates greater than or equal to 100,000 pounds per bour.

Conductivity.. 10 u=he/cm chloride ion.. 1.0 ppm

5. If Specification 3.6.B cannot be net, an orderly shutdown shan be initiated and the reactor shan be in Ect Shutdovn vithin 24 hrs.

and Cold Shutdovn within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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'1. Any time irradiated fuel is in the f l1. Reactor coolant system leakage shall reactor vessel and reactor coolant N s be checked by the sump ans air .

) j sa=pli:g system and recorded at

/ te=perature is above 2120T, reactor ,

coolant leakage into the primary / _

least once per day, cent.ainment from unidentified (

/

sources shall not exceed 5 gp=. In - 'N -

I addition, the total reactor coolant # ,, .,

system leakage into the pri=ary Ie) o/ C containment shall not exceed 25 gym. 4,

2. Both the su=p and air sampling sys- /

te=s shall be operable during reac- /

ter power operation. from and after the date that one of these k syster.sismadeorfoundtobeiso/

etable for any reason, reactor '

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Anendnent No. 42 , .

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125

L1HITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS

^ 4.6 3.6.C CoolantLeakaae(Cont'd(

power operation is p'ermissible >

only during the succeeding seven days, f 3. If the conditions in 1 or 2 above cannot be met, an orderly shutdown

(

('

shall be initiated and the reactor *//OpO i

/ shall be in a Cold Shutdown 4 ondition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C pjn s

\

Safety and Relief Valves Safety and Relief Valves D. D.

1. During reactor power operating 1. At least one safety valve and two conditions and prior to reactor relief / safety valves shall be startup from a Cold Condition, or checked or replaced with bench whenever reactor coolent pressure checked valves once per operating is greater than 104 psig and cycle. All valves will be tested temperature greater than 340*F, every two cycles, both safety valves and the safety modes of all relief valves shall 2. At least one of the relief / safety be' operable, valves shall be disassembled and inspected each refueling outage.

The nominal setpoint for the

/ relief / safety valves shall be 3. Whanever the safety relief valves

,f' selected between 1095 and 1115 are required to be operable, the

, psig, All relief / safety valves discharge pipe temperature of each

/ shall be set at this nominal safety relief valve shall be logged setpoint 11 psi. The safety daily.

/, valves shall be set at 1240 psig 2

^

13 psi, 4. Instrumentation shall be calibrated and checked as indicated in Table

2. If Specification 3.6,0.1 is not 4.2.F.

met, an orderly shutdown shall be initiated and the reactor coolant 5. Notwithstanding the above, as a pressure shall be below 104 psig minimum, safety relief valves that within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Note: Technical have been in service shall be Specifications 3.6,0.2 - 3.6,0.5 tested in the as-found condition apply only when two Stage Target during both Cycle 6 and Cycle 7.

Rock SRVs are installed.

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

2 H , 22, d[ 126

. . ~~ .- - .-. - . ~ - ~ . --- .-

. . . .. i x

BASEST M C and l. 6.C

-Conlant tenkare

'. l l . ' , Allovabic leakape rates of coolant from the reactor cont- i ant system have been based on the predicted and export. l mentally. c,bservet behavicr of cracks in pipes and on the I ability to makeup coolant system leakape in the event of .

loss of offnite 4 c power. The normally expected back- I p.round leakage due to equirment design and the detection capability for determining coolant system Icakage vote

also considered in-establishing the limits. The behavior of cracks in piping systems has been experimentally and.

l

-analytically investigated as part of the 1;SAEC sponserce Reactor Primary Cool Ant SystCm Eupture $1udy (the pjpc Rupture Study). flork utilising the data obtained in this

.stutiy indicates that leakage from a crack can be detected I before the crack grows to a dangerous or critical size by sechar.ically or thermally induced cyclic los$ on, or 1

-stress corrosion crecling or some other mech..n r ;haracter- 1 ized by. I;radual crack gred This evidene. s.,n n ts that for leakage. somewha t greater than the Mat t t.per.ified for unidentified leakage, the probability is small that im-

. perfections or cracks associated with such leakage vould

- --grow rapidly, llovever, the establishment or allowable

unidentified Icahage greater than that given in 3.6.C on .

/. .y the basis of the data presently-available vould be prca mature because of uncertainties associated with the data.: Tor leakage nf the order of 5 ppm,as spceified in'3.6.C, the experimental and analytical data suggest

a. reasonable mars.in of. safety that such leakage man-a nitude would not result from a crcck approaching.the critical sfac for rapid propagation. Leakage less than

=the magnitude specified can be detected reasonably in a matter of a few hours utilizing the availrble leakage detection schemes.: and if the origin cannot be deter-mined in.a reasonably short timt the_ plant should be shut down to allow turther investigation and entrective action.

/'N f '[The total leakage rate consists-of all leakt ge, identified Mj) y and a; unidentified, u and $ r t drainwhich sumps.flows to the drywell H,arTr71n e'p / -~ q,,f /

7 d

- eto,, e ,. ;

The city of theg ryvel1 floor sump pumps is d gpm and

. i spacityofty)cryvellequipment sump pumps is also

,fg 100.g m. Removal o 25 r.pm from either of these sumps con ccomplished with considert.ble margin, m .

<f /$ s) d g43

.i -

i-l

>=r

m.- ,

P AETS:

4

. _ . 3.6.c ne.d h.6.c '

coolu,t !.vaktt e ( Co.a.t ' 6)

,. 3 ,/ ' ' w -ofQ ,leakec.e detc tion '

[* [O f The . rfore.ar. reactor o(olar.t *

/ syste s vill be t.luated dtu 1 - the fity yent o.

( s tatier. peratior, c.. conclu a of ti?t y g lua.

"'$1on vill be' reported to the A W.

6

\

l

--.d*

h

~

144

Insert "A" FF. Pressure Boundary Lg3]gLqe - Pressure boundary leakage shall be leakage through a non-isolable fault in a reactor coolant system component body, pipewall or vessel wall.

GG Identified Leakage - 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. Recctor 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 Unidentified Leakaae - Unidentified leakage shall be all reactor coolant leakage which is not Identified Leakage.

Insert "B" 3.6.C Coolant Leakaae Any time irradiated fuel is in the reactor vessel and coolant temperature is above 212*F, the following limits shall be observed:

1.00erational Leakage

a. Reactor coolant system leakage shall be limited to:

1. No Pressure Boundary Leakage

2. 15 gpm Unidentified Leakage

3. 125 gpm Total Leakage averaged over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

4. 12 gpm increase in Unidentified Leakage averaged over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period when in RUN mode,

b. With any reactor coolant system leakage greater than the limits of 2 and/or 3., above, reduce the leakage to within acceptable limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

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

Insert "B" (Cont'd)

2. Leakage Detection Systems

- The following reactor coolant system leakage detection systems shall be Operable:

1. One drywell sump monitoring system, and either

2. One channel of a drywell atmospheric particulate radioactivity monitoring system, or

3. One channel of a drywell atmospheric gaseous radioactivity monitoring syster,

b. 1. At least one drywell sump monitoring system shall be Operable; otherwise, be in Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

c. With no required leakage detection systems Operable, be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Insert "C" 4.6.C Coolant Leakaag Any time irradiated fuel is in the reactor vessel and coolant temperature is above 212*F, the following surveillances shall be performed:

1. Querational Leakasa Demonstrate drywell leakage is within the limits specified in 3.6.C.1. by monitoring the coolant leakage detection systems required to be operable by 3.6.C.2. at least once every B hours.

2. Leakaae Detection Systems The following reactor coolant leakage detection systems shall be demonstrated Operable:

a. For each required drywell sump monitoring system perform:

1. An instrument functional test at least once per 31 days, and

2. An instrument channel calibration at least once per 18 months.

_.._._.._.___._m

v .:

Insert "C" (Cont'd)

b. For each required drywell atmospheric radioactivity monitoring system perform:

1. An instrument check at least once per day,

2. An instrument functional test _at least once per 31 days, and

3. An instrument channel calibration at least once per 18 months.-

Insert "D" '

Vsrification,0f-- the integrity of the reactor coolant system (3.6.C.1.a.1. : No Pressure Boundary Leakage) is- provided during RPV Class I system hydrostatic and leak tests conducted to meet section 3/4.6.G: _ Structural Integrity (ASHE p . Code,Section XI,.IHA:5000,-and.IHB 5000.)

Two leakage collection sumps are provided-inside primary containment. _

Identified leakage is piped from pump seal leakoffs, reactor vessel head

flange 1 seal leakoff,! selected valve stem leakoff including recirculation loop and' main 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 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. .

[

I -

A drywell: sump; monitoring' system,.as required in 3.6.C.2,_ consists of one' _

equipment sump l pump-and one floor drain sump ~ pump, plus associated !'

instrumentation.- Flow integrators, one for the. equipment drain _ sump and._

anotherLfor.the floor sump, comprise the basic instrument system, and are used

.to record the1 flow of liquid from the drywell-sumps. A. manual-system whereby the time: interval between sump pump starts istutilized 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 capacity of the pump is known, I- The capacity = of each of the two drywell floor sump pumps is_50 gpm and the capacity.of-each of the two drywell equipment sump pumps is also 50 gpm.

" Removal- of 25 gpm fro'n either of these sumps can be accomplished with considerable margin.

E _In addition to the sump monitoring of coolant leakage, airborne radioactivity-levels of the drywell atmosphere is monitored by the Reactor Pressure Boundary Leak Detection System. _This ! system consists.of two panels capable of monitoring-the primary-containment atmosphere for particulate-and gaseous radioactivity as a result of coolant leaks.

y -

l --

1' , - -

Insert."D" (Cont'd)

The 2 gpm limit for coolant leakage rate increase over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is a limit specified by the NRC in Generic Letter 88-01: "NRC Position on IGSCC in BHR Austenitic Stainless Steel Piping". This limit applies only during 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).

- - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

ML20072T540

Text

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