ML20069Q480

Proposed Tech Spec Revisions Changing Duration of Containment Integrated Leak Rate Test from 24-h to 8-h. Justification for Proposed Revisions Encl

ML20069Q480
Person / Time
Site:	Browns Ferry
Issue date:	12/03/1982
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20069Q469	List: ML20069Q467 ML20069Q480
References
NUDOCS 8212090088
Download: ML20069Q480 (36)

Text

. .

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS TVA BFNP TS 146 SUPPLEIGINT 1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 8212090088 821203 PDR ADOCK 05000259 P pop

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O LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS 2.a. Brimary containment integrity 2. Integrated Leak Rate Testing L

shall be maintained at all Primary containment nitrogen times when the reactor is critical or when the reactor ~ consumption shall be monitored

~ - ' ~

water temperature is above

~ to 212 F and fuel is in the determine the average daily reactor vessel except while nitrogen consumption for the performing "open vessel" last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Excessive leakage physics tests at power levels is indicated by a N2 consumption not to exceed SMW(t). rate of)2% of the primary con '

tainment free volume per 24 hoars

b. Primary containment integrity (corrected for drywell temperature, is confirmed if the maximum pressure, and venting operations) allowable futegrated leaki~ge at 49.6 psig. Corrected to normal rate, La, does not exceed drywell operating pressure of 1.1 the equivalent of 2 percent psig, this value is542 SCFE. If of the primary containment this value is exceeded, the action volu'e m per'24 h'ours at'the specified in 3.7.A.2.C shall,be 49.6 psig design basis taken.

accident pressure, Pa.

The containment leakage rates

c. If N2 makeup to the primary shall be demonstrated at the containacht averaged over following test schedule and shall 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for be determined in conformance with pressure, temperature, and the criteefe specified in Appendix J venting operations) exceeds to 10 CFR 50 using the methods and 542 SCFH, it must be reduced provisions of ANSI N45.4(1972).

to<542 SCFH within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the ' reactor shall be placed a. Three type A tests (overall in hot shutdown within the integrated containment laakage I next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, rate) shall be conducted at 40110-month intervels during shutdown at either P , 49.6 ppig, or at J g ,25 psig, during each -

10-year plant inservice inspection.

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b. If any periodic type A test faile to meet egeher 0.75 L or 0.75 Le the test schedule !or sub-sequent type A tests shall be reviewed and approved by the Commission.

If two consequtive type A tests fail to meet either 0.75L or 0.75 L a type A test shall beperfo,rmedatleastevery 18 months until two consecutive 229 4

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• cLIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINNENT SYSTEMS 4.7 CONTAINNENT SYSTEMS _

type A tests meet either 0.75 L, or 0.75Lg, at which time the above test sbhedule may be resumed.

c. 1. Test duration shall be at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

2. A %-hour stabilization period will be required and the containment i atmosphere will be considered stabilized when the change in weighted I

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i 229a 9

LIMTING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRMENTS

-- -- -3 .-7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS -

average air temperature averaged over an hour does not deviate by more than

- 0.5*R/ hour from the average rate of change of temperature measured from the previous '

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

- - . . d. 1. At least 20 sets of data

. points at approximately equal time intervals and

. in no case at intervals greater than one hour

- shall be provided for proper statistical analysis.

2. The figure of merit for the instrumentation system shall never exceed 0.25 L,.

e. The test shall not be concluded with an increasing calculated leak rate.

f. The accuracy of each type A test shall be verified by a supplemental test which:

1

1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the type A test data is within

- 0.25 L,or 0.25 L't.

2. Has duration sufficient to establish accurately the

, change in leakage rate between the type A' test and the supplemental test.

i

3. Requires the quantity of gas injected into the

! containment or bled from l the containment during the l

supplemental test to be equivalent to at least 25 percent of the total measured leakage at Pa (49.6 psig), or Pt (25 psig).

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS, 4.7 CONTAINMENT SYSTEMS

g. Local L'ak e rate tests (LLRT's) shall be performed on the primary containment testable penetrations and isolation. valves, which are not part of a water-sealed system, at not less than 49.6 psig (except for the main steam isolation valves, see 4.7.A.2.1) and not less than 54.6 psig for water-sealed valves each operating cycle. Bolted double-gasketed seals shall be tested whenever the seal is cicsed after being opened and at least once per operating cycle. Acceptable methods of testing are halide gas detection, soap bubbles, pressure decay, hydrostatically pressurized fluid flow or equivalent.

The personnel air lock shall be tested at a pressure of 49.6 psig during each operating cycle. ,In addition, following each opening, the personnel air lock shall be leak tested at a pressure of22.5 psig within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the l

first of each series of openings whenever contain-ment integrity is required.

The personnel air lock shall be leak tested at a pressure ofk2.5 pois at least once every 6 months from the

- first of each series of openings to verify thd 231

, . - , - - - - - - , , - , , - , . , - - - - . , - - - - . , - -~ -n . - , - --_ - - - -__- - --

LIMITING CONDITIONS FOR OPERATION SURVEIIJ.ANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 ColfrAINMElfr SYSTEMS condition of the air lock assembly whenever containsant integrity is required. The

• total leakage from all penetrations and isolation valves shall not exceed 60 percent of L, per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Leakage from containment i

isolation valves that

! , terminate below suppression pool water level may be excluded from the total leakage provided a sufficient fluid inventory is available to ensure the sealing function for at least 30 days at a pressure of 54.6 psig. Leak-age from containment isolation valves that are in closed-loop, seismic class I lines that will be water sealed during a DBA will be measured but will be excluded when computing the total leakage.

Penetrations and isolation valves are identified as follows:

(1) Testable penetrations with double 0-ring seals - Table 3.7.B, (2) Testable penetrations i with testable bellows Table 3.7.C, (3) Isolation valves with-

, out fluid seal - Table 3.7.D, (4) Testable electrical penetrations - Table 3.7.H. and (5) Isolation valves sealed with fluid -

Tables 3.7.E, and 3.7.F.

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4 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS

h. (1) If at any time it is determined that the criterion of 4.7.A.2.g is exceeded, repairs shall be initiated iczediately.

~

• (2) Ifcriterion conformance to the of 4.7.A.2.g is not demonstrated 1

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TABLE 3.7.D AIR TESTED ISOLATION VALVES

~

Valve Valve Identification 1-14 Main Steam 1-15 Main Steam 1-26 Main Steam 1-27 Main Steam 1-37 Main Steam '

1-38 Main Steam 1-51 Main Steam J 1-52 -

Main Steam 1-55 Main Steam Drain 1-56 Main Steam Drain 2-1192 Service Water 2-1383 Service Water 3-554 Feedwater 3-558 Feedwater 3-568 Feedwater 3-572 Feedwater 32-62 Drywell Compressor Suction 32-63 Drywell Compressor Suction'

• 32-336 Drywell Compressor Return 32-2163 Drywell Compressor Return 33-1070 Service Air 33-785 Service Air 43-13 Reactor Water Sample Lines 43-14 Reactor Water Sample Lines, .63-525 Standby Liquid Control Discharge 63-526 Standby Liquid Control Discharge 64-17 Drywell and Suppression Chamber Air Purge Inlet 64-18 Drywell Air Purge Inlet 64-19 Suppression Chamber Air Putge Inlet 64-20 Suppression Chamber vacuum Relief 64-c.v. Suppression Chamber Vacuum Relief 64-21 Suppression Chamber Vacuum Relief 64-c.v. Suppression Chamber Vacuum Relief 64-29 Drywell Main Exhaust 64-30 Dryvell Main Exhaust 64-32 Suppression Chamber Main Exhaust 64-33 Suppression Chamber Main Exhaust 64-31 Drywell exhaust to Standby Gas Treatment 64-34 Suppression Chamber to Standby Gas Treatment 64-139 Drywell pressurization, Compressor Suction 64-140 Drywell pressurization, Compressor Discharge 68-508 CRD to RC Pump Seals68-523 CRD to RC Pump Seals68-550 CRD to RC Pump Seals68-555 , CRD to RC Pump Seals 286

TABLE 3 7.D (Continued) g Valve Identification RWCU Supply 69-1 RWCU Supply ,

69-2 RWCU Return 69-579 RCIC Steam Supply 71-2 - RCIC Steam Supply 71- 3 RCIC Pump Discharge '

71-39 RCIC Pump Discharge 71-40 RCIC Steam Supply g) 73-2 RCIC Steam Supply 73-3 HPCI Pump Discharge 73-44 HPCI Pump Discharge 73-45 -

HPCI Steam Supply Bypass 73 P,1 RHR Shutdown Suction 74-47 RHR Shutdown Suotion 74-48 RHR Shutdown Suction 74-661 RHR Shutdown Suction 74-662 Drywell/ Suppression Chamber Nitrogen Purge 76-17 Drywell Nitrogen Purge Inlet 76-18 Suppression Chamber Purge Inlet 76-19 Drywell/ Suppression Chamber Nitrogen Purge 76-24 Containeent Atmospheric Monitor 76-49 Containment Atmospheric Monitor 76-50 Containment Atmospheric Monitor 76-51 Containment Atmospheric Monitor 76-52 Containment Atmospheric Monitor 76-53 Containment Atmospheric Monitor 76-54 Containment Atmospheric Monitor 76-55 Containment Atmospheric Monitor 76-56 Containment Atmospheric Monitor 76-57 Containment Atmospheric Monitor 76-58 Containment Atmospheric Monitor 76-59 Containment Atmospheric Monitor 76-60 Containment Atmospheric Monitor 76-61 Containment Atmospheric Monitor 76-62 Containment Atmospheric Monitor 76-63 Containment Atmospheric Monitor 76-64 - Containment Atmospheric Monitor 76-65 Containment Atmospheric Monitor 76-66 Containment Atmospheric Monitor 76-67 Containment Atmospheric Monitor ,-

76-68 Drywell Floordrain Sumo 77-2A Drywell Flocrdrain Sump 77-2B Drywell Equipment Drain Sump 77-15A Drywell Equipment Drain Sump 77-15D Containment Atmospheric Dilution 84-8A Containment Atmospheric Dilution 84-8B Containment Atmospheric Dilution 84-8C Containment Atmospheric Dilution 84-80 Containment Atmospheric Dilution 84-19 Main Exhaust to Standby Gas Treatment 84-20 Main Exhaust to Standby Gas Treatment 84-600 84-601 Main Exhaust to Standby Gas Treatment 84-602 _ __

Main Exhaust to Standby Gas Treatment 259

TABLE 3 7.D (Continued)

Valve Valve Identification 84-603 Main Exhaust to Standby Cas Treatment 85-576 CRD Hydraulic Return 90-254A '

Radiation Monitor Suction 90-254B Radiation Monitor Suction 90-255 Radiation Monitor Suction 90-257A Radiation Monitor Discharge 90-257B Radiation Monitor Discharge ,

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TABLE 3.7.E FRD'ARY COICAIIOEIE ISOLATICIT VALVES WIIICII TERMIIIATE DELCN TIE SUPPRESSICII POOL UATER LEVEL Valve Valve Identification 12-738 Auxiliary Boiler to RCIC 12-7h1 Auxiliary Boiler to RCIC 43-28A RIE Suppression Chamber Scmple Lines h3-28B RIE Suppression Chamber Sample Lines h3-29A - RIE Suppression Chamber Sample Lines43-29B RIE Suppression Chamber Sample Lines 2-11h3 Dwineralized Water 71-14 RCIC Turbine Exhaust 71-32 RCIC Vacuum Pump Discherge 71-580 RCIC Turbine Exhaust 71-592 RCIC Vacuum Pump Discharge 73-23 HEI Turbine Exhaust 73-24 IIPCI Turbine Exhaust Drain 73-603 HKI Turbine Exhaust ,73-609 HFCI Exhaust Drain 7h-722 RER 75-57 Core Spray to Auxiliary Boiler 75-58 Core Sprof to Auxiliary Boiler Core Spray to Auxiliary Boiler e

0 262

TABLE 3 7.F ,

PRIMARY CofffAIIIMENT ISOIATION VALVES LOCATED IN WATER SEALED SEISMIC CLASS 1 L UES Valve Valve Identification 74-53 RIE LICI Discharge 74-54 RIE 7h-57 RHR Suppression Chamber Spray 74-58 RIE Suppression Chamber Spray 74-60 RIE Drywell Spray Th-61 RHR Drywell Spray 74-67 RIE LICI Discharge 74-68 RHR LICI Discharge 74-71 RHR Suppression Chamber Spray 74-72 RHR Suppression Chamber Spray 7h-74 RIE Drywell Spray 74 ~75 RHR Drywell Spray 74-77 RHR Head Spray '

74-78 RHR Head Spray 75-25 Core Spray Discharge 75-25 Core Spray Discharge 75-53 Core Spray Discharge 75-54 Core Spray Discharge i

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TABLE 3,7,G (This table intentionally left blank) 9 e

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-BASES j .

3 7.A & 4.7.A Primary Containment i The integrity of the primary containment and operation of the core standby cooling system in combination,-ensure that the release of radioactive materials from the j containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the >

leakage rate limitation, will limit the site boundary radiation doses to within the i

limits of 10 CFR Part 100 during accident conditions.

During initial oore loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will be no pressure on the system thus greatly reducing the chegRes of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability of an accident occuring.

] The limitations on primary containment leakage rates ensure that the total l containment leakage volume will not exceed the value assumed in the accident analyses

{ at the peak accident pressure of 49.6 psig, P . As an added conservatism, the measured overall integrated leakage rate is further limited to 0 75 L during a performance of the periodic tests to account for possible degradation of the ,

containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50 (type A, B, and C tests).

l The pressure suppression pool water provides the heat sink for the reactor primary i system energy release following a postulated rupture of the system. The pressure l suppression chamber water volume must absorb the associated decay and structural i sensible heat release during primary system blowdown from 1,035 psig. Since all of

, the gases in the drywell are purged into the pressure suppression chamber. air space during a loss of coolant accident, the pressure resulting form isothermal compression

( plus the vapor pressure of the liquid must not es.ceed 62 psig, the suppression j chamber maximum pressure. The design volume of the suppression. chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is ,

purged to the suppression chamber.

Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 psig, which is below the maximum of 62 psig. The maximum water level indications of -1 inch c'orresponds to a downcomer submergence of 3 feet 7 inches and a water volume of 127,800 cubic feet with or 128,700 cubic feet without the drywell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with differential pressure control and -7.25 inches without differential pressure control corresponds to a downcomer submergence of approximately 3 feet and water volume of approximately 123,000 cubic feet. Maintaining the water level between these levels will ensure that the torus water volume and downcomer submergence are within the aforementioned limits during normal plant operation. Alarms, adjusted for instrument error, will notify the operator when the linits of the torus water level are approached. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with response to downcomer submergence, l this specification is adequate. The maximum temperature at the end of blowdown tested during the Humboldt Bay and Bodega Bay tests was 1700F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 1700 F.

267 l

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t L'IMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 _ CONTAINMENT SYSTEMS 2.a. Primary containment integrity 2. Integrated Leak Rate Testing shall be maintained at all l times when the reactor is Primary containment nitrogen

_ consumption _shall be monitored critical or when the reactor _

water temperature is above to 212 F and fuel is in the determine the average daily reactor vessel except while ni'trogen consumption for the performing "open vessel" last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Excessive leakage physics tests at power levels is indicated by a N2 consumption not to exceed 5MW(t). rate of>2% of the primary con-tainment free volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

b. Primary containment integrity (corrected for drywell temperature, is confirmed if the maximum pressure, and venting operations)

' at 49.6 psig. Corrected to normal allowable integrated leakage rate, La, does not exceed drywell operating pressure of 1.1 the equivalent of 2 percent psig, this value is 542 SCFH. If of the primary containment this value is exceeded, the action volume per'24 h'ours at the specified in 3.7.A.2.C shall be 49.6 poig design basis taken.

accident pressure, Pa.

The containment leakage rates

c. If N2 makeup to the primary shall be demonstrated at the containment averaged over following test schedule and shall 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for be determined in conformance with pressure, temperature, and e the criteria specified in Appendix J venting operationa)' exceeds to 10'CFR'50 usi'ng the methods and 542 SCFH, it must be reduced provisions of ANSI N45.4(1972).

to <542 SCFH within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be placed a. Three type A tests (overall in hot shutdown within the integrated containment leakage next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, rate) shall be conducted at 40110-month intervals during shutdown at either P , 49.6 psig, or at.P g,25 psig, during each 10-year plant inservice inspection.

b. If any periodic type A test fails to meet either 0.75 L or 0.75 Le thetestschedule!orsub-sequent type A tests shall be reviewed and approved by the l

Commission.

If two consecutive type A tests fail to meet either 0.75L or

~

0.75L'atypeAtestsh!11 t

be perrormed at least every 18 months until two consecutive 229 e

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5 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTFMS 4.7 CONTAINMEfff SYSTEMS type A tests meet either 0.75 L, or 0.75Lg, at which time the above test schedule may be resumed.

c. 1. Test duration shall be at

. least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

2. A 4-hour stabilization period will be required and the containment sta sphere will be considered stabilized when the change in weighted e

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- _ ~_ , . --

LIMTING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRMENTS

. -3 .-7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS .

average air temperature averaged over an hour does

not deviate by more than

. 0.5'R/ hour from the average rate of change of temperature measured from the previous 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

. . . . . d. 1. At least 20 sets of data

,. points at approximately equal time intervals and in no case at intervals greater than one hour

- shall be provided for proper statistical analysis.

2. The figure of merit for the-l instrumentation system shall never exceed 0.25 L,.

a. The test shall not be concluded

. with an increasing calculated leak rate.

f. The accuracy of each type A test shall be verified by a l supplemental test which:

le Con. firms the accuracy of the test by verifying that-l the difference between the l supplemental data and the

[

type A test data is within 0.25 L,or 0.25 L't.

2. Has duration sufficient to establish accurately the change in leakage rate between the type A' test and the supplemental test.

3. Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at P a

, .- (49.6 psig), or Pg (25 psig).

1 230

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTOLS

g. Local Lehk rate tests (LLRT's) shall be performed on the primary containment testable penetrations and isolation. valves, which are not part of-a water-sealed system, at not less than 49.6 poig (except for the main steam isolation valves, see 4.7.A.2.1) and not less than 54.6 pois for water-sealed valves each operating cycle. Bolted double-gasketed seals shall be tested whenever the seal is closed after being opened and at least once per operating cycle. Acceptable methods of testing are halide ga? detection, soap bubbles, pressure decay, hydrostatically pressurized fluid, flow

~

~cr equivalent.

i The personnel air lock l

shall be tested at a l pressure of 40.6 psig during each operating cycle. ,In addition, following each opening, the personnel air lock shall be leak tested at a pressure of22.5 peig within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the l

first of each series of openings whenever contain-ment integrity is required.

l The personnel air lock shall l ~ be leak tested at a pressure of22.5 psig at least once every 6 months from the

' first of each series of openings to verify thd' 231 i

LIMITING CONDITIONS 70R OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS condition of the air lock assembly whenever containment integrity is required. The total leakage from all penetrations and isolation valves shall not exceed 60 percent of L,per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Leakage from containment isolation valves that terminate below suppression pool water level may be excluded from the total leakage provided a sufficient fluid inventory is available to ensure the sealing function for at least 30 days at a pressure of 54.6 psig. Leak-age from containment isolation valves that are in closed-loop, seismic class I lines that will be water sealed during a DBA will be measured but will be excluded when computing the total leakage.

Penetrations add isolation valves are identified as follows:

(1) Testable penetrations with double 0-ring seals - Table 3.7.B, (2) Testable peretrations with testable bellew:

rable 3.7.C.

(3) Isolation valves with-i

, out fluid seal - Table 3.7.D, (4) Testable electrical penetrations - Table 3.7.H, and (5) Isolation valves  !

sealed with fluid -

i Tables 3.7.E, and 3.7.F.

l l

232

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMEhTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS

h. (1) If at any time it is determined that the criterion of 4.7.A.2.g is exceeded, repairs shall be initiated immediately.

-(2) If conformance to'the'~

criterion of 4.7.A.2 3

, is not demonstrated e

I l

l 232a e

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BASES .

j ;3.7.A & 4.7.A Primary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, ensure that the release of radioactive materials from the

, containment atmosphere will be restricted to those leakage paths and associated leak' rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the j limits of 10 CFR Part 1q0 during accident conditions.

During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will be no pressure on the i system thus greatly reducing the changes of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability of an accident occuring.

The limitations on brimary containment leakage rates ensure that.the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 49.6 psig, P g. As an added conservatism, the measured overall integrated leakage rate is further limited to 0 75 La during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

i The surveillance testing for measuring leakage rates are consistent 'tb the requirementsofAppendixJof10CPRPart50(typeA,B,andCtest0}.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the systen. The pressure l suppression chamber water volume must absorb the associated decay and structural

sensible heat release during primary system blowdown from 1,035 psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space .

l during a loss of coolant accident, the pressure resulting form isothermal compression

! plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be cor.densed is discharged to the suppression chamber and that the drywell volume _is ,

l purged to the suppression chamber. .

Using the nihimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 psig, which is below i the maximum of 62 psig. The maximum water level indications of -1 inch corresponds to a douncomer submergence of 4 feet 7 inches and a water volume of 129,000 cubic feet with or without the drywell-suppression chamber differential pressure control.

The minimum water level indication of -7 inches with differential pressure control ,

l and -8 inches without differential pressure control corresponds to a downcomer ,"

l submergence of approximately 3 feet and water volume of approximately 123,000 cubic feet. Maintaining the water level between these levels will ensure that'the torus

! water volume and downcomer submergence are within the aforementioned limits during

, normal plant operation. Alarms, adjusted for instrument error, will notify the operator when the limits of the torus water level are approached. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with response to downcomsr submergence, this specification is l

adequate. The maximum temperature at the eni of blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 0 F and this is conservatively taken to be the limit for complete condensation of the reactor coolant,'although condensation would occur for' temperatures above 170 0F.

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L'IMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS 2.a. Primary containment integrity 2. Integrated Leak Rate Testing shall be maintained at all times when the reactor is Primary containment nitrogan critical or when the reactor consumption shall be monits_14 water temperature is above to 212 F and fuel is in the determine the average daily reactor vessel except while ni'trogen consumption for the performing "open vessel" last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Excessive leakage physics tests at power levels is indicated by a N2 consumption not to exceed SMW(t), rate of>2% of the primary con-toinment free volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

b. Primary contaf,nment integrity (corrected for drywell temperature, is confirmed if the maximum pressure. and venting operations) allowable integrated Icakage at 49.6 psig. Corrected to normal rate, La, does not exceed drywell operating pressure of 1.1 the equivalent of 2 percent peig, this value is 542SCFH. If of the primary containment this valu2 is exceeded, the action volu'e m per 24 h'ours at the specified in 3.7. A.2.C shall be

~

49.6 psig design basis taken.

accident pressure, Pa.

The containment leakage rates

c. If N2 makeup to the primary shall be demonstrated at the containment averaged over following test schedule ~and shall 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for be determined in conformance with pressure, temperature, and the criteria specified in Appendix J venting operations) exceeds to 10 CFR 50 using the methods and 542 SCFH, it must be reduced provisions of ANSI N45.4(1972).

to<542 SCFH within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be placed a. Three type A teste (overall in hot shutdown within the integrated containment leakage next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, rate) shall be conducted at 40210-month intervals during l

shutdown at either P,, 49.6 psig, or at .P t.25 psig, during each 10-year plant inservice inspection.

i I b. If any periodic type A test fails to meet either 0.75 L or 0.75 Lt thetestscheduleforsub-sequent type A tests shall be reviewed and approved by the

(

' Commission.

If two consecutive type A tests l fail to meet either 0.75L or O.75 L atypeAtestsh!11 l

beperEo,rmedatleastevery 16 months until two consecutive l 233

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYST1!MS 4.7 CONTAINMENT SYSTEMS type A tests meet either 0.75 L,or 0.75L g, at which time the above test schedule may i be resumed. ,

c. 1. Test duration shall be at

. least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

2. A 4-hour stabilization period will be required and the containment atmosphere will be considered stabilized l when the change in weighted f

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LIMTING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRMENTS

-3 .-7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS .

average air temperature averaged over an hour does not deviate by more than

. 0.5*R/ hour from the average rate of change of temperature measured from the previous 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

. . . . J. 1. At least 20 sets of data

. points at approximately equal time intervals and in no case at intervals greater than one hour

. shall be provided for proper statistical analysis.

2. The figure of merit for the instrumentation system shall never exceed 0.25 L,.

e. The test shall not be concluded with an increasing calculated leak rate.

f. The accuracy of e. n type A test shall be verified by a supplemental test which:

1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the type A test data is within

- 0.25 L aor 0.25 L t.

2. Has duration sufficient to establish accurately the change in leakage rata between the type A~ test and the supplemental test.

3. Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at Pa l

- (49.6 psig), or Pg (25 psig).

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS

g. Local Leak rate tests (LLRT's) shall be performed on the primary containment testable penetrations and isolation. valves, which are not part of a water-sealed system, at not less than 49.6 psig (except for the main steam isolation valves, see 4.7.A.2.1) {

and not less then 54.6 psig for water-sealed valves each operating cycle. Bolted double-gasketed seals shall be tested whenever the seal is closed after being opened and at least once per operating cycle. Acceptable methods of testing are halide gra detection, soap bubbles, pressure decay, hydrostatically .

pressurized fluid flow or equivalent.

The personnel air lock shall be tested at

,6-month intervals at an internal pressure of not less than 49.6 psig. In addition, if the personnel air lock is opened during ,

periods when' containment l integrity is not =

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required, a test at the end of such a period will l

be conducted at not less l

than 49.6 psig. If the personnel air lock is opened during a period when containment integrity is r3 quired, a test atr2 5 l

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, psig shall be conducted

' within 3 days after being opened. If the air lock is opened more frequently than once every 3 days, the air lock shall be tested at least once every l RM

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS 3 days during the period of frequent openings.

The total leakage from all penetrations and isolation valves shall not exceed 60 percent of L per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Leakage from, containment isolation valves that terminate below suppression pool water level may be excluded from the total leakage provided a sufficient fluid inventory is available to ensure the sealing function for at least 30 days at a pressure of 54.6 psig. Leak-age from containment isolation valves that arn in closed-loop, seismic class I lines that will be water sealed during a DBA will be measured but will be excluded when computing the total leakage.

Penetrations and isolation valves are identified as follows:

(1) Testable penetrations with double 0-ring seals - Table 3.7.B, (2) Testable penetrations with testable bellows Table 3.7.C.

(3) Isolation valves with-

. out fluid seal - Table 3.7.D, l (4) Testable electrical penetrations - Table 3.7.H. and (5) Isolation valves sealed with fluid -

Tables 3.7.E, and 3.7.F.

237

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. LIMITING CONDITIONS FGR OPERATION SURVEILLANCE REQUIREMENTS

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3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS

.h.

(1) If at any time it is

, determined that the criterion of 4.7.A.2.g

.- is exceeded, repairs

^' -

shall be initiated 3

immediately.

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• (2) If conformance to'the criterion of 4.7.A.2.g is not demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following detection of excessive local leakage,

, the reactor shall be shut down and depressurized until repairs arc effected

) ,

snd the local leakage meets the acceptance

criterion as demonstrated by re-test.

i. -The main steamline isolation valves shall be tested at a pressure of~25 psig for leakage -

during each refueling outage. If the leakage rate of 11.5 scf/hr for l any one main steamline isolation valve is ex-ceeded, repairs and

- retest shall be performed to correct the condition.

. j. Continuous Leak Rate Monitoring

/

When the primary contain-j < , ment is inerted, the

/ -,

containment shall be

'/ - '

. continuously monitored

- for gross leakage by j review of the inerting

- < system makeup requirements.

This monitoring l ,- ,

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BASES , ,

3.7.A & 4.7.A Primary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, ensure that the release of radioactive naterials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

During initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required, there will be no pressure on the system thus greatly reducing the changes of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect to minimize the probability or an accident occuring.

~

The linitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 49.6 psig, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to 0.75 La during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50 (type A, B, and C tests).

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat release during primary system blowdown from 1,035 psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting form isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and j

air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 psig, which is below l

the maximum of 62 psig. The maximum water level indications of -1 inch corresponds

( to a downcomer subosrgence of 3 feet 7 inches and a water volume of 127,800 cubic, feet with or 128,700 cubic feet without the drywell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with ,

I differential pressure control and -7.25 inches without differential pressure control corresponds to a downcomer submergence of approximately 3 feet and water volume of approximately 123,000 cubic feet. Maintaining the water level between these levels

~

will ensure that the torus water volume and downcomer submergence are within the l aforementioned limits during normal plant operation. Alarms, adjusted for instrument l error, will notify the operator when the limits of the torus water level are l approached. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with response to downcomer submergence, i this specification is adequate. The maximum temperature at the end of blowdown

tested during the Humboldt Bay and Bodega Bay tests was 1700F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, al.though condensation would occur for temperatures above-170 F.-

285

9 ENCLOSURE 2 JUSTIFICATION FOR PROPOSED REVISIONS

BACKGROUND The original request for shortened Containment Integrated Leak Rate Test (CILRT) from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration to only 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, was made by TVA BFNP TS 134, submitted by TVA letter from L. M. Mills to H. R. Denton dated December 17, 1979. Class III and Class I fees totalling $4,800 were forwarded at that time. The next request was made by TVA BFNP TS 137, submitted by L. M. Mills' letter to H. R. Denton dated April 15, 1980. TS 137 proposed to bring the specifications more into line with the BWR Standard Technical Specifications while proposing the shortened CILRT. TS 137 superceded TS 134 in its entirety. The next request was made by TVA BFNP TS 146 submitted by L. M. Mills' letter to H. R. Denton dated August 19, 1980. That request supplemented and replaced TS 137 in its entirety. TVA BFNP TS 146 Supplement 1 updates and replaces the original TS 146.

DESCRIPTION AND JUSTIFICATION T3146 Supplement 1 updates TS 146 to reflect modifications made to the torus. It reflects the addition of new hydrogen-oxygen i

l analyzers and the removal of the old hydrogen-oxygen analyzer system.

The change to page 229 is to change the drywell operating pressure from 15 psig to 1.1 psig and the allowable leakage from 549 SCFH to 542 SCFH. This adds conservatism to the allowable leakage rate.

i l

1 This is done to ensure that the test method does not produce nonconservative results, i

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Tables 3 7.D through 3 7.G for unit 1 submitted in TS 146 have been updated to reflect changes made to those tables since submittal of TS 146. The tables for units 2 and 3 submitted in TS 146 have been withdrawn. These tables for unit 3 were updated as necessary to reflect plant configuration in the unit 3 reload 4 license amendment. These tables for unit 2 have been updated as necessary and included in the. unit 2 reload 4 request, TVA BFNP TS 179, currently under review by the NRC staff.

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