Text

Amends 112 & 116 to Licenses DPR-44 & DPR-56,respectively, Revising RCPB Leak Detection Tech Specs Re Measuring of Airborne Radioactivity in Primary Containment
	ML20137Y221
Person / Time
Site:	Peach Bottom
Issue date:	11/19/1985
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Philadelphia Electric Co, Public Service Electric & Gas Co, Delmarva Power & Light Co, Atlantic City Electric Co
Shared Package
ML20137Y224	List: ML20137Y221; ML20137Y235;
References
	DPR-44-A-112, DPR-56-A-116 NUDOCS 8512100753
	Download: ML20137Y221 (22)
	v • d • e

- _ _ _ _ _ _.

[

'o UNITED STATES g

8 NUCLEAR REGULATORY COMMISSION o

r,

.E WASHINGTON, D. C. 20555

s.,...../

PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.112 License No. DPR-44 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Philadelphia Electric Company, et al. (the licensee) dated May 4, 1983, as amended by letters dated November 10, 1983, and November 29, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operatino License l'o. ')PR-44 is hereby amended to read as follows.

1 i

I h

l P

2-Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.112, are hereby incorporated in the license. PECO shall operate the facility in accordance with the Technical Specifications.

3.

This ifcense amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

[

John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 19, 1985 i

l i

ATTACHMENT TO LICENSE AMENDMENT NO. ll2 FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.

Remove Insert v

v 76 76 85 85 93 93 146 146 146a 146a 1

156 156 156a 156a s

i

~

t o

PBAPS Unit 2 e.

LIST OF TAET "

l N

M ERES 3.1.1 Reactor Protection System (Scram) 37 Instrumentation Requirement 4.1.1 Reactor Protection System (Scram) 41 Instrument Punctional Tests 4.1.2

~ Reactor Protection System (Scram) 44-Instrument Calibration-3.2.A Instrumentation That Initiates Primary 61 Containment Isolation

3. 2. B Instrumentation That Initiates or-controls 64

.the Core and Containment Cooling Systems

.3. 2. C Instrumentation That Initiates control 73 Rod Blocks 3.2.D aadiation Monitoring Systems That Initiate 75 and/or Isolates Systems

. 3.2. F Survelllance Instrumentation 77

~ 3. 2. G Instrumentation That Initiates Recirculation

~

79 Pumo-Trip 4.2.A

, Jtinimum Test and Calibration Frequency for PCIS 80 i

l l

~

Amendment No. M.)f,112 O

h.

M e

.wem

,_-,,,--wv---.--'--"'"'

i Table 3.2.E Deleted Amendment No. 112

-76

g 2

3 TABLE 4.2.E

=

0 MINIMUM TEST AND CALIBRATION FREQUENCY FOR DRYWELL LEAK DETECTION ro Instrument Channel Instrument Functional Calibration Instrument Test Frequency Check l

1)

Equipment Drain Sump Flow Integrator (1)

Once/3 months Once/ day h

2)

Floor Drain Sump Flow Integrator (1)

Once/3 months once/ day 1

3)

Drywell Atmosphere Radioactivity (1)

Once/3 months Once/ day Monitor i

1 i

O l

i

1 '

~

PBAPS 3.2 RASES (Cont'd) j Reactor Building Isolation function and operation of theF a e the treatment system.

the refueling area ventilation exhaust ducts and four instPour ins standby gas Each set of the instrument channels is arranged in a i

rument i

trip logic.

i twice 4

ventilation exhaust ducts are based upon initiating norm area isolatism and standby gas treatment system operation so th t entilation' I

the activity released during the refueling accident leav a none of Ruilding via the normal ventilation path but rather all th es the Reactor is.psossesed by the standby gas treatment syeten e activi ty I

Flow integrators.are used to record the integrated flow of I

liquid from the drywell sumps.

The integrated flow is 1

indicative of reactor coolant' leakage.

Radioactivity Monitor is provided to give supportingA Drywell Atmosphere monitoring system.information to that supplied by the reactor coolant leakaije 4

(See Bases for 3.6.C and 4.6.C)

For.each paraIn~eter acnitored, as listed in Table 3.2.F, therg are two (2) channels of instroentation.

performance is available.two (2) channels, a near continuous surveillanc Any deviation in readings will initiate an early recalibratica, thereby maintaining the quality of the instrument readings.

l The recirculatica pump trip has been added at the suggestion of ACRS as a means of limiting the consecuences of the unlikely occurrence of a failure to scram during an anticipated transient.

i j

events given in, General Electric Company Topical Repo The response of i

j dated March,1971.

i 4

i i

Amendment No. Jef.112 93-

t 4

a,

--. _., - -. - -,,.. ~, -,.. -,

,-.,~,---.--.--,-_...-.---._----.v,-,.~-,,,,,v.

,,---,,,,n-

l PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREHENTS 3.6.B Coolant Chemistry (Cont'd) 4.6.B Coolant Chemistry (Cont'd) b)

Chloride Concentration Time above 2 weeks / year 0.2 ppm Maximum limit 1.0 ppm c) pH During operations, if the conductivity exceeds 1.0 umho/cm, pH shall be measured and brought within the 5.6 to 8.6 range within 2' hours. If the pH cannot be corrected, or if the pH is outside a range of 4 to 10, the unit shall be placed in Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown with 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

C.

Coolant Leakage C.

Coolant Leakage 1.

Any time irradiated fuel is 1.

Reactor coolant system leakage in the reactor vessel and reactor shall be determined by the conlant temperature is above primary containment (Drywell) 212 degrees F, the rate of sump collection and flow reactor coolant leakage to the monitoring system and recorded

-primary containment f rom unidenti-every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or less.

fled sources shall not exceed 5 gallons per minute.

The rate 2.

Drywell atmosphere radioactivity of change of unidentified leakage levels shall be monitored and shall not exceed 2 gallons per recorded at least once per day.

~

minute per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance period when the reactor is operated in the "Run" mode. In addition, the total reactor coolant system leakage into the primary contain-ment shall not exceed 25 gpm averaged over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance period."

Amendment No. 77,77),112

-146-

l PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.C. Coolant Leakage 2.

The primary containment (Drywell) sump collection and flow monitoring system shall be operable during reactor power operation.

From and after the time that this system is made or found to be inoperable for any reason, reactor power operation is permissible only during the succeeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the system is made operable sooner.

For purposes of this paragraph, the primary containment (Drywell) sump collection and flow monitoring system operability is defined as the ability to measure reactor coolant leakage.

3.

The Drywell Atmosphere Radioactivity Monitor shall be operable during reactor power operation as a supple-ment to the reactor coolant leakage monitoring system.

From and after the time that this system is made or found to be inoperable for any reason, reactor power operation is permissible for up to 30 days provided grab samples of the containment atmo-sphere are obtained and analyzed at least once per ~24 hours.

~

4.

If the conditions in 1, 2, or 3 cannot be met, an orderly shutdown shall be initiated and the reactor shall 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 Condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

t Amendment No. 7H.112

-146a-l

l PDAPS l

3.6.C & 4.6.C BASES i

1 Coolant Leakage a

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 1

The j

system leakage in the event of loss of offsite ac power.

normally expected background leakage due to equipment design and i

1 the detection capability for determining coolant system leakage were also considered in establishing the limits.

The behavior of i

j cracks in piping systems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary j

Coolant System Rupture Study (the Pipe Rupture Study).

Work utilizing the data obtained in this study indicates that leakage i

i 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 i

i unidentified leakage, the probability is small that imperfections I

or cracks associated with such leakage would grow rapidly.

l l

However, the establishment of allowable unidentified leakage greater than that given in 3.0.C on the basis of the' data j

presently available would be premature because of uncertainties associated with the data.

For leakage of the order of 5 gpm, as l

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 propagation.

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 shutdown to allow further investigation and corrective action.

)

A rate of change limit of 2 gpm per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance period 'is specified to provide additional conservatism.

This limit.is applicable to reactor operations in the "Run" mode, during which 1

time there is little variation in primary coolant system pressure.

The limit does not apply to the "Startup" mode since this period j

is characterised'by large variations in system pressure and consequently, changes in measured leakage would not be indicative l

of system degradation.

During the limited duration of the startup i

phase, the 5 gpm limit will ens'ure the integrity of the primary coolant system.

4 f

I The total leakage rate consists of all leakage, unidentified and i

identified, which flows to the drywell floor drain and equipment drain sumps, respectively.

Both the Drywell floor drain and the equipment drain sumps have pump-out capacitites of 50 gpm per pump.

Any one pump can therefore handle in excess of the maximum allowable total leakage of 25 gym.

If the ability to measure pump-out flow from either of these sumps is-lost, the inoperable Amendment No, Jpp, 112

-156 -

-w.

mMr mypr-++g-<<a%=y-->94.wggr - +sy y w.e

,,pw m,,e m---9 w m gin um4syigp eg m+

- tm e fewa-9 =,gg w, w eg9.me

-,ge wm_

w q

-.ywyy.

wyy,9-wr p wig,.m&+<ep W g.-

99 Td'FN-'Wt'T iN'.

.MW'T--"

k sump will overflow into the r'emaining operable sump.

The remaining operable sump pump-out flow will then represent the total leakage rate.

During the time when one sump is overflowing, any increase in total flow will be assumed to be from an unidentified source.

This primary containment (Drywell) sump collection and flow monitoring system can provide viable measurement of reactor coolant system leakage so long as one pump and its associated flow meter are operable.

The Drywell Atmosphere Radioactivity Monitor provides supporting information to that provided by the reactor coolant leakage monitoring system.

There is no direct correlation between the radioactivity monitor indication and the leakage rate because of the uncertainties regarding coolant activity levels, source of leakage, and background radiation levels.

While the radioactivity monitors will not quantify primary coolant leakage, they would provide an early warning of a major leak especially if there is a significant difference in the radioactivity level between the leakage source and drywell background.

i a

I i

s 4

i l

Amendment No. Jp),112

-156a-f

/

o UNITED STATES

^

NUCLEAR REGULATORY COMMISSION li E

WASHINGTON, D. C. 20555

\\.....}

PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-278 PEACH BOTTOM ATOMIC POWER STATICh, UtlTT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.ll6 License No. DPR-56 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Philadelphia Electric Company, et al. (the licensee) dated May 4, 1983, as amended by letters dated November 10, 1983, and November 29, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-56 is hereby amended to read as follows:

l

. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.116, are hereby incorporated in the license. PEC0 shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION b

O John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

November 19, 1985 4

m-.

y...

y-

a ATTACHMENT TO LICENSE AMENDMENT NO. 116 FACILITY OPERATING LICENSE NO. DPR-56 DOCKET NO. 50-278 4

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.

Remove Insert y

y 76 76 85 85 93 93 146 146 146a 146a 156 156 156a 156a 4

4 1

l 1

I i

n 1

1

}

l l

_ -. - _ _ _ -._~ _ _. _.

PBhPS Unit 3 i

LIST OF Thut*5 t

IRhlt Illh IAE2 3.1.1 Reactor Protection system (scram) 37 Instrumentation Requirement 4.1.1 Peactor Protection System (scram) 41 Instrument Functional Testa 4.1.2 Reactor Protection system (scram)

Instrument Calibration 44

3. 2. A Instrumentation That Initiates Primary 61 Containment Isolation 3.2.5 Instrumentation That Initiates or controla the core and Containment Cooling systems 64 3.2.C Instrumentation That Initiates Control 73 -

Rod Blocks 3.2.D Padiation Monitoring Systems That Initiate 75 and/or Isolates Systems Surveillance Instrumentation b

3.2.F 77 3.2.G Instrumentation That Initiates Recirculation Pump Trip 79

4. 2. A Minimum Test and Calibration Frequency for PCIS 80 j

I l

i O

e 9

4 Amendment No.M fr.116 s

-v

+

= m +-eene -

e w-G

.N e

use esp

9 9

e,

-e- - - -, - - - - - - - -

-,---m-

1 i$

4 j

i 4

3 i

4 4

-l 4

i i

5

.i i

a Table 3.2.E Deleted

.I i

l i

i 4

I i

Amendment No. 116

-76 4

4 f

~ - _ _ _ _. _ _ _ _. - -, _.._.___ -__-._ _____._,-_.--_--. _______... _. _.._-, _.

k TABLE 4.2.E 3

MINIMUM TEST AND CALIBRATION FREQUENCY FOR DRYWELL LEAK DETECTION k'

Instrument Channel Instrument Functional Calibration Instrument Test Frequency Check e

3 1)

Equipment Drain Sump Flow Integrator (1)

Once/3 months Once/ day i

2)

Floor Drain Sump Flow Integrator (1)

Once/3 months Once/ day 3)

Drywell Atmosphere Radioactivity (1)

Once/3 months Once/ day Monitor i

l e

O s

?

PBAPS 3.2 BASIS (Cont'd)

Four sets of two radiatica aceiters are provided which initiate the Reactor Building Isolation function and operation of the standby gas treatment system. Four instrument channels monitor the radiatica from the refueling area ventilation exhaust ducts and four instrumert channels monitor the building ~ ventilation below the refueling floor.

Each set of the instrument chan==1e is arranged in a 1 out of 2 twice trip logic.

Trip settings of <16 ar/hr for the sonitors in the refueling area e

ventilation exhaust ducts are based upon initiating normal ventilation i

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 1

is processed by the standby gas treatment system.

j

~

Flow integrators are used to record the integrated flow of liquid from the drywell sumps.

The integrated flow is indicative of reactor coolant' leakage.

A Drywell Atmosphere Radioactivity Monitor is provided to give supporting Information to that supplied by the reactor coolant leakage monitoring system.

(See Bases for 3.6.C and 4.6.C)

's For.each parameter scnitored, as listed in Table 3.2.F, there are two (2)' channals of instrumentation.

By comparing readings betwe'en the two (2) channels, a near continuous surveillance of instrument performance is available.

Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.

The recirculatica pump trip has been added at the suggesticri of ACRS as a means of limiting the consecuences of the unlikely occurrence of a failure to scram during an anticipated transient.

j The response of the plant to this postulated event fall within the envelope of study events given in, General Z1ectric Company Topical Report, NEDD-10439, i

dated March,1971.

I l

l i

I Am udsent No. 361f, 116 !

I.

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.B Coolant Chemistry (Cont',d1 4.6.B Coolant Chemistry (Cont'd) b)

Chloride Concentration Time above 2 weeks / year 0.2 ppm Maximum limit 1.0 ppm c) pH During operations, if the conductivity exceeds 1.0 umho/cm, pH shall be measured and brought within the 5.6 to 8.6 range within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the pH cannot be corrected, or if the pH is outside a range of 4 to 10, the unit shall be placed in Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown with 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

C.

Coolant Leakage C.

Coolant Leakage 1.

Any time irradiated fuel is 1.

Reactor coolant system leakage in the reactor vessel and reactor shall be determined by the coolant temperature is above primary containment (Drywell) 212 degrees F, the rate of sump collection and flow reactor coolant leakage to the monitoring system and recorded

. primary containment from unidenti-every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or less.

fied sources shall not exceed 5 gallons per minute.

The rate 2.

Drywell atmosphere radioactivity of change of unidentified leakage levels shall be monitored and shall not exceed 2 gallons per recorded at least once per day,

^

minute per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance period when the reactor is operated in the "Run" mode. In addition, the total reactor coolant system leakage into the primary contain-ment shall not exceed 25 gpm averaged over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance period.-

Amendment No Q, J D, 116

-146-

.. - = _.

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i

3.6.C. Coolant Leakage 2.

The primary containment (Drywell) sump collection and flow monitoring system shall be operable during reactor power operation.

From and after the time that this system is made or found to be inoperable for any reason, reactor power operation is permissible only during the.

succeeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the system j

is made operable sooner.

For purposes of this paragraph, the primary containment (Drywell) sump collection and flow monitoring system operability is defined as the ability to measure reactor coolant leakag,e.

3.

The Drywell Atmosphere Radioactivity Monitor shall be operable during reactor power operation as a supple-ment to the reactor coolant leakage monitoring system.

From and after i

the time that this system is made or found to be inoperable for any reason, reactor power operation is pe,rmissible for up to 30 days provided i

grab samples of the containment atmo-1 sphere are obtained and analyzed at least once per -24 hours.

4.

If the conditions in 1, 2, or 3 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in at least Hot Shutdown within i

the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in Cold Shutdown Condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l l

Amendment No. Jpp,1116

-146a-

..., _ -.. -..,_ m _._

m.-

--..-,m, m.---,-

,_- _,- _._..-..___..,.,-. _ _,..-.~..,.... _.., -,

i 2

PDAPS i

3.6.C & 4.6.C BASES Coolant Leakage 1,

I Allowable leakage rates of coolant from the reactor coolant system I

have been based on the predicted and experimentally observed I

behavior of cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite ac power.

The normally expected background leakage due to equipment design and j

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 a

Coolant System Rupture Study (the Pipe Rupture Study).

Work l

l utilizing the data obtained in this study indicates that leakage j

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 l

that for leakage somewhat greater than the limit specified for i

unidentified leakage, the probability is small that imperfections i

or cracks associated with such leakage would grow rapidly.

However, the establishment of allowable unidentified leakage j

greater than that given in 3.6.C on the basis of the data i

presently available would be premature because of uncertainties i

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 j

reasonable margin of safety that such leakage magnitude would not i

result from a crack approaching the critical size for rapid i

propagation.

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 t

determinr3 in a reasonaoly short time, the plant should be j

j shutdown to allow further investigation.and corrective action.

i

+

A rate of change limit of 2. gym per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance period is i

specified to provide additional conservatism.

This limit is j

applicable to reactor operations in the "Run" mode durine; which time there is little variation in primarg coolant system pressure.

l The limit does not apply to the 'Startup mode since this period is characterised -by large variations in system pressure and consequently, changes in measured leekage would not be indicative of system degradation.

During the limited duration of the startup phase, the 5,gpa limit will ensure the integrity of the primary i

coolant, system.

4 l

The total leakage. rate consists of all leakage, unidentified and i

identified, which flows to the drywell floor drain and equipment

,p l

drain sumps, respectively.

Both the Drywell floor drain and the equipment drain sumps have pump-out capacitites of 50 gym per pump.

Any one pump can therefore handle in escess of the maximum allowable total leakage of 25 gym.

If the ability to measure l

pump-out flow from either of these sumps is lost, the inoperable l

Amendment No. 19),

116

-156 -

i 1

--g--

-.e-.%, -

,-w.-w-

-w ww,.

-e-3

,-v,

,wr,,-%--_%.v.-.mw._

y.

e -m.

.,_em-w,

sump will overflow into the remaining operable sump.

The remaining operable sump pump-out flow will then represent the total leakage rate.

During the time when one sump is overflowing, any increase in total flow will be assumed to be from an unidentified source.

This primary containment (Drywell) sump collection and flow monitoring system can provide viable measurement of reactor coolant system leakage so long as one pump and its associated flow meter are operable.

The Drywell Atmosphere Radioactivity Monitor provides supporting information to that provided by the reactor coolant leakage monitoring system.

There is no direct correlation between the radioactivity monitor indication and the leakage rate because of the uncertainties regarding coolant activity levels, source of leakage, and background radiation levels.

While the radioactivity monitors will not quantify primary coolant leakage, they would provide an early warning of a major leak especially if there is a significant difference in the radioactivity level between the leakage source a.nd drywell background.

l 1

d Amendment No, ypp,116

-156a-

.

ML20137Y221

Text

Attachment:

Attachment:

Navigation menu

Search