ML19331D639

Proposed Tech Specs 3.7 & 4.7 Re Limiting Conditions for Operation & Surveillance Requirements for Containment Sys

ML19331D639
Person / Time
Site:	Browns Ferry
Issue date:	08/19/1980
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML19331D636	List: ML19331D635 ML19331D639
References
NUDOCS 8009030411
Download: ML19331D639 (47)

Text

'.

m ENCLOSURE PROPOSED TECllNICAL SPECIFICATIONS BROWNS FERRY NUCLEAR PLANT (TVA BFNP TS 146) 80090307//

LIMITING 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 nitrogen critical or when the reactor consumption shall be monitored water#

temperature is above once each 8-hour shift 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 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 pressurn, and venting operations) allowable integrated leakage at 49.6 psig. Corrected to normal rate, La, does not exceed drywell operating pressure of 1.5 the equivalent of 2 percent psig, this value is 549SCFH. If of the primary containment this value is exceeded, the action volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 549 SCFH, it must be reduced provisions of ANSI N45.4(1972).

to 549 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

! 40!10-month intervals during shutdown at either P a, 49.6 psig, or at.P g ,25 psig, during each

(

10-year plant inservice inspection.

l l

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

If two consecutive type A tests fail to meet either 0.75L, or 0.75 L a type A test shall be perfo,rmed at least every 18 months until two consecutive

, 229 l

L

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINNENT SYSTEMS 4.7 CONTAINMENT SYSTEMS type A tests meet either 0.75 L, or 0.75Lg, at which time tras 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 />.

t

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

l i

n I

[

229a

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 s.ne hour

- shall be prosided for proper statistical analysis.

2. The figure of merit for the instrumentation system shall never exceed 0.25 La*

e. The test shall not be concluded i

with an increasing calculated

leak rate.

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

1. Coafirms the accuracy of the test by verifying that the difference between the supplemental data and the type A test data is within 0.25 La or 0.25 L t.

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

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

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

230 t .

. . . .- . . . _ . .s , . . . . . . _ . . _ ,. - . - _ _ _ __.... ._.___. ,.

LIMITING CONDIT:.ONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTLMS 4.7 CONTAINMENT SYSTEMS

g. Local L'eak rate tests (LLRT's) shall be performed on the primary contairment 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 closed 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 of 2.5 psig within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the first of each series of openings whenever contain-ment integrity is required.

The personnel air lock shall be leak tested at a pressure of 2.5 psig at least once every 6 months from the

- first of each series of openings to verify the' 231 w - - ,-.

LIMITING CONDITIONS FOR OPERATION SURVEI'. LANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 .T,NTAINMENT SYST ES cendition of the air lock assembly whenever containnent 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 ps.g. 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 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.

232

4 l

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) Ifcriterion conformance to'the of 4.7.A.2.g is not demonstrated l

l e

232a

, , . , - , - r , s s-. , , , , , - - - - , -

TABLE 3.7.D AIR TESTED ISOIATION 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 Stea=

l-51 Main Steam 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 i 32-62 Drywell Compressor Suction '

32-63 Drywell Compressor Suction - 32-336 Drywell Co: pressor Return 32-2163 Drywall Compressor Return 33-1070 Service Air 33-785 Service Air 43-13 Feactor Wat.'.r Sample Lines 43-14 Reactor Water Sample Lines i 63-525 Standby Liquid Control Discharge 63-526 Standby Liquid Control Discharge 64-17 Dryon1 and Suppression Chamber Air Purge Inlet ,

64-18 Dry. ell Air Purge Inlet 64-19 Suppression Cha=ber Air Purge inlet 64-20 Suppression Chamber Vacuum Relief 64-c.v. Suppression Cha=ber Vacuu:2 Relief

64-21 Suppression Chamber Vacuu= Relief ,

64-c.v.

Suppression Chamber Vacuu= Relief

, 64-29 Drywell Main Exhaust 64-30 Dryvell Main Exhaust 64-32 '

Suppression Chamber Main Exhaust 64-33 Suppression Chamber M_ tin Exhaust 64-31 Drywell exhaust to Standby Cas Treatment t,4-34 Suppression Cha:aber to Standby Cas Treatment 64-139 Drywell pressurization Compressor Suction-64-140 Drywell pressurization, Compressor Discharge '68-508 CRD to RC Pu=p Seals l 6S-523 CRD to RC Pu=p Seals

.68-550 CRD to RC Pump Seals68-555 , CRD to RC Pump Seals  ;

4 258 e

h

J',J.. 6, l'  ;

4

[' TABLE 3.7.D (continued) 1/

Valve Valve Identification 69'-l RWCU Supply 69-2 RWCU Supply 69-579 RL'CU Return 71-2 RCIC Steam Supply 71-3 RCIC Steam Supply 71-39 RCIC Pump Discharge 71-40 RCIC Pump Discharge 73-2 RCIC Steam Supply 73-3 RCIC Steam Supply 73-44 HPCI Pump Discharge 73-45 HPCI Pump Discharge 74-47 RHR Shutdown Suctioh 7A-48 RHR Shutdown Suction

/4-661 -

RHR Shutdown Suction 1 74-662 RHR Shutdown Sucticn 76-17 Drywell/ Suppression Chamber Nitrogen Purge Inlet 76-18 Drywell Nitrogen Purge Inlet 76-19 Suppression Chamber Purge Inlet ,

76-24 Drywell/ Suppression Chamber Nitrogen Purge Inlet 76-215 Containment Atmospheric Monitor i 76-217 Containment Atmospheric Monitor 76-220 Containment Atmospheric Fonitor 76-222 Containment Atmospheric Monitor 76-225 Containment Atmospheric Monitor - 76-226 Containment Atmospheric Monitor 76-229 Containment Atmospheric Monitor 76-230 Containment Atmospheric Monitor 76-237 Containment Atmospheric Monitor 76-239 Containment Atmospheric Monitor 76-242 Containment Atmospheric Monitor 7o-243 Containment Atmospheric Monitor 76-248 Containment Atmospheric Monitor 76-250 Containment Atmospheric Monitor 76-253 Containment Atmospheric Monitor 76-254 Containment Atmospheric Monitor 77-2A Drywell Floordrain Sump 77-2B Drywell Floordrain Sump 77-15A Drywell Equipment Drain Sump 77-15B Drywell Equipment Drain Sump 84-8A Containcent Atmospheric Dilution 84-8B Containn.. nt Atmospheric Dilution 84-8C Containme t Atmospheric Dilution 84-8D Containmt.it Atmospheric Dilution 84-19 Contlinnant Atmospheric Dilution 84-20 Main Ext.2ust to Standby Gas Treatment 84-600 Main Exhaust to Standby Gas Treatment 84-601 Main Exhaust to Standby Gas Treatment Y 84-602 Main Exhaust to Standby Gas Treatment 84-603 Main Exhaust to Standby Gas 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 GO-257B Radiation Monitor Discharge ,

259

m J M

0 0

0 6

e (DELETED) e O

260-261

TABLE 3.7.E FRE'tJtY CO:CAII"'EIE ISCLATION VALVES WHICH TIEC! ATE

- 3ELGT TIG SUPHIESSIO:I POOL UATER LEVEL l

Valve Vt.1ve Identification 12-733 Auxiliary Boiler to ~1CIC 12-7h1 Auxiliary Boiler to RCIC h3-28A Rim Suppression Chamber Scmple Lines43-283 RHR Suppression Cha2er Sa :ple Lines h3-29A . R*E Suppression Chamber Sample Lines43-293 RHR Suppression Chr.mber Sa=ple Lines 2-11h3 Demineralized Water 71-14 RCIC Turbine Exhaust 71-32 RCIC Vecuum Pmp Dischcrge 71-580 RCIC Turbine Exhaust 71-592 RCIC Vacuum Pap Discharge 73-23 HIcI Turbir.e Exhaust 73-2h HICI Turbine Exhaust Drein 73-603 nIcI Turbine Exhoust ,

73-60) HIcI Exhaust Drain

?h-722 RHR 75-57 Core Spray to Auxiliary 3 oiler 75-53 Core Sprar to Auxiliary Boiler Core Spray to Auxiliary Boiler 262 e

S

' ~ -

w n r---- ,_ _

TABLE 3.7.F ,

PRD&RY CorffAI!3mNr ISOLATION VALVES LOCATED ET WATER SEALED SEISMIC CIASS 1 LETES Valve Valve Identification 74-53 RER LICI Discharge 74-54 R!m 74-57 RIG Suppression Chamber Spray 74-58 RIG Suppression Chamber Spray 74-60 RIE Drywell Spray 74-61 RIG Drywell Spray 74-67 RIE LICI Discharge 74-68 KIR LICI Discharge 74-71 RIG Suppression Chamber Sproy "h-72 RIG Suppression Chamber Spray 74 "4 RIIR Drywell Spray 7h ~5 Rim Drywell Spray 74-77 RIE Head Spray

• 7h-78 RIG Head Spray 75-25 Core Spray Discharge 75-25 Core Spray Dischorge 75-53 Core Spray Discharge 75-54 Core Spray Discharge 263

TABLE 3.7.C (This table intentionally left blank) 264 e + --- e------ -

b 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 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 . 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 of an accident occuring.

The limitations 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 L during performance of the periodic tests to account for possible degradati"n 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 released 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 from 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 ccadensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

267

Using the minimum or maximum water levels given in the specification, containment pressure during the design basis accident is approximately 49 psip, which is below the maximum of 62 psig. The maximum water level 5 %Jication of -1 inch corresponds to a downcocer sutmergence 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 and -8 inches without differential pressure control corresponds to a downcomer submergence of approximately 4 feet and water volume of approximately 123,000 cubic eet.

Maintaining the water level between these levels will assure that the torus water volume and downcomer submergency are with 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, wita response to downcomer submergence, this specification is adequate. The maximum temperature at the end gf blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 F and this is conservatively taken to be the limit for complete condensation of the# reactor cools it, although condensation would occur for temperatures above 170 F.

267a

O O

UNIT 2 PROPOSED CHANGES 1

l

)

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 nitrogen critical or when the reactor consumption shall be monitored water temperature is above once each 8-hour shift to

1. determine the average daily 212 F and fuel is in the 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 bv a N2 consumption not to exceed SMW(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) allowable integrated leakage ' at 49.6 psig. Corrected to normal rate, La, does not exceed drywell operating pressure of 1.5 the equivalent of 2 percent psig, this value is 549SCFH. If of the primary containment this value is exceeded, the action volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 demotstrated 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 usi'ng the methods and 549 SCFH, it must be reduced provisions of ANSI N45.4(1972).

to 549 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 a, 49.6 psig, or at.P t,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 Lt the test schedule for sub-sequent type A tests shall be reviewed and approved by the Commission.

If two consecutive type A tests fail to meet either 0.75L,or 0.75 L a type A test shall

~ beperfo,rmedatleastevery 18 months until two conscrutive 229 .

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYST MS 4.7 CONTAINMENT SYSTEMS type A tests meet either 0.75 L, or 0.75Le, 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 atmosphere will be considered stabilized when the change in weighted i

O e

I 229a

+

-c -r, - - -.v - - , - - - , , , - - - , -, , . - , , ~. , ,,, -- ,---~ - , -p - -e- e ,-- -

SURVEILLANCE LEQUIRMENTS LIMTING CONDITIONS FOR OPERATION

--- - -3 r7 CONTAINMENT SYSTDLS 4.7 CONTAINMENT SYSTDiS -

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 *ieast 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.-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 La or 0.25 L t-

2. Mas duration sufficient to

~

establish accurately the change in leakage r' ate 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 teat to be equivalent to at least 25 percent of the total meacured leakage at Pa

• (49.6 psig), or Pt (25 psig).

230 pg _,g ,

% 4'*' " %s'*  ?-4 y. e . =j .e*,,p,,g g , , . __

e

,,- -. . _ . . - , . ,_ . _ , , . . .. __ ._._ .._..,_s_ . _ . , , . . - _

i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTDIS 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 than 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 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 of 2.5 psig within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the first of each series of openinga whenever contain-ment integrity is required.

The personnel air lock shall

~ be leak tested at a pressure of 2.5 psig at least once every 6 months from the first of each series of openings to verify thd 231

LIMITING CONDITIONS FOR 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 penetrations 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.

232

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 immediately.

-(2) Ifcriterion conformance to'the of 4.7.A.2.g is not demonstrated

{

{

l I

f - .

232a t

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

TABLE 3.7.D

- AIR TESTED ISOLATION VALVES Valve Valve Identification 1-14 Main Steam 1-15 M.iin Steam 1-26 Main Steam i 1-27 Main Steam

• 1-37 Main Steam l-18 thin Steam 1-51 -

Main Steam 1-52 tuin Steam 1-55 Main Steam Drain 1-56 thin Steam Drain 2-1192 Service Water 2-1383 Service Water 3-554 Fcedwater 3-558 Feedwater 3-568 Feedwater 3-572 Feedwater

• 32-62 Dryus11 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 Lines63-525 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 Purge Inlet 64-20 Suppression Chamber Vacuum Relief 64-c.v. Suppression Chamber Vacuum Relief

' 64-21 Suppression Chanber Vacuum Relief 64-c.v. Suppression Chamber Vacuum Relief 64-29 Drywell Main Exhaust 64-30 Drywell Main Exhaust 64-32 Suppression Chamber Main Exhaust 64-33 Suppression Chamber Main Exhaust 64-31 Drywell exhaust to Standby Cas Treatment 64-34 Suppression Chamber to Standby Gas Treatment 64-13') Drywell pressurizattun, Compressor Suction 64-140 Drywell pressurization, compressor Discharge

. I 258 .

, , , - , . - - . . . _ , . . , . , , . - - - - . - . --n ,gm ,g.,, , , . , , ., - ,, -.-

, . . - , , -,y --

Table 3.7.D (continued)

Valve Valve Identifica* tion 69-1 RWCU Supply 69-2 RWCU Supply 69-379 RWCU Return 71-2 RCIC Steam Supply 71-3 RCIC Steam Supply 71-39 RCIC Pump Discharge 71-40 RCIC Pump Discharge 73-2 RCIC Steam Supply 73-3 RCIC Steam Supply 73-44 HPCI Pump Discharge 73-45 llPCI Pump Discharge 74-47 RilR Shutdown Suction 74-48 RHR Shutdown Suction 74-661 RHR Shutdcwn Suction 74-662 RHR Shutdown Suction 76-17 Drywell/ Suppression Chamber Nitrogen Purge Inlet 76-18 Drywell Nitrogen Purge Inlet 76-19 Suppression Chamber Purge Inlet 76-24 Drywell/ Suppression Chamber Nitrogen Purge I..let 76-215 Containment Atmospheric Monitor 76-217 Containment Atmospheric Monitor 76-220 Containment Atmospheric Monitor 76-222 Containment Atmospheric Monitor 76-225 Containment Atmospheric Monitor

• 76-226 Contair. ment Atmospheric Monitor 76-229 Containment Atmospheric Monitor 76-230 Containment Atmospheric Monitor 76-237 Cnntainment Atmospheric Monitor 76-239 Containment Atmospheric Monitor 76-242 Containment Atmospheric Monitor 76-243 Containment Atmospheric Monitor 76-248 Containnent Atmospheric Monitor 76-250 Containment Atmospheric Monitor 76-253 Containment Atmospheric Monitor 76-254 Containment 14 xetpheric Monitor 77-2A Drywell F>so irain Sump 77-2B Drywell Vv.isirald Sump 77-15A Dryval' t^u. aent Drain Sump 77-15B Dryw d. :p ;.aent Drain Sump 84-8A Contc a nment At: aspheric Dilution 84-8B Containment Atmocpheric Dilution 84-8C Containment Atmospheric Dilution 84-8D Conta inment Atmospheric Dilution 84-19 Containment Atmospheric Dilution 84-20 Main E..haust to. Standby Cas Trratment 84-600 Main Exhaust to Standby Cas tr eatment 84-601 Main Exhaust to Standby Cas Treatment 84-602 Main Exhaust to Standby Cas Treatment 84-603 Main Exhaust to Standby Cas Treatment 259

TABLE 3,7.D (continued)

Valve Valve Identification 85-576 CRD Hydraulic Return 90-254A Radiation Monitor Suction 90-254B Radiation Monitor Suction 1 90-255 Radiation Monitor Suction

. 90-257A Radiation Monitor Discharge j 257B Radiation Monitor Discharge J

'h e

t

'260 .

w - v-w e-O O

e f

(DELETED) 1 i

i f

261

(

TA3*.I 3.7.E

- FED'A2r CCT.Ar!'I'C ISC'.A IZ VA;* Es *n7*..g mcc 32IJi CEE E'Jtcr.iSIO5 ?>JL *iATER IZ".E Valve Valve Identification 12-733 Auriliary sailer to ?mC 12-TL1 I,uxiliary Boiler to ECIC L3-22A 2 : 03 ?;ppression Chsder S:sple Lines 133 233

- 3 2 Cuppression Cha ':er Sa : pie Lines h3-29A ?2R Cupp ession Chader Gs71e ~1.e?

L3-2:3 EEE Suppression Chr.:ber Sa:ple Lines 2-11L3 resinarslized'Jater "1-1k EO!C Sarbbe Ixherst 1 .,;_ ..uiu

• ecuu: P=p .,.schcrge 71-520 2CIC Turbine Ixhaust

~1-592 20!C Vac r.= P:-; Cischarge T3-23 EICI Srbine Ixtrast 3-24 I~ICI TurbLee Ixhaust Orain 3-603 E= I ?:rbine Ixhaust .

73 'kX) I~IC Exhaust Drain

~L 22 ?E

~5-57 Core Spray to A;1L115Yi Eoiler 75-53 Core spref to k.utiliary Boiler '

Core Sprny to k.ucL11 sty EcL1er 262 e

. .r TA3LE 3.~.F ,

PRE!ARY CCITfAII:IGIC ISOLATI0IT VALVES LOCATED III WATER SEALED SEISMIC CLASS 1 LIES Val:e Valve Identification 74-53 RER LICI Discharge 74-54 R'E 7h-57 RIE Suppression Cha:nber Sprey 74-58 RIG Suppression Chamber Spray 74-60 RIE Drywell Spray 74-61 RHR Drywell Spray 74-67 RIS LICI Discharge 74-68 RIG LICI Discharge 74-71 RHR Suppression Chamber Spray 714-72 RHR Suppression Cha:nber Spray 74 ~h RIE Drywell Spray 74-75 RHR Drywell Spray 74-77 RIIR Head Spray 74-78 RIE Head Spray 75-25 Core Sprav Discharge 75-25 Core Spray Dischorge 75-53 Core Spray Discharge 75-54 Core Spray Discharge 263 9

,,m,- -- - - - - ,

TABLE 3.7.C (This table intentionally left blank)  !

l l

264

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 materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. Inis restriction, in conjunction with the leakage rate limitation, will limit the site bcundary 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 of an accident occuring.

The limitations 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 L during performance of the periodic tests to account for possible degradati$n 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 released 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 from 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 airi 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.

267 t

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

l I

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 indication of -1 inch corresponds to a downcomer 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 and -8 inches without differential pressure control corresponds to a downcomer submergence of approximately 4 feet and water volume of approximately 123,000 cubic eet.

Maintaining the water level between these levels will assure that the torus water volume and downcomer submergency are with 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 downcomer submergence, this specification is adequate. The maximum temperature at the end 0of blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation we.ald occur for temperatures above 170 F.

267a

L'IMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS UNIT 3 PROPOSED CHANGES 1

r- -- c -

- .-,e - , - - . r_e-s. . - - - , - - - - - - - .r., -.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 _ CONTAINMENT SYSTEJM type A testa meet either 0.75 or 0.75Lg , t which time L,he t 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 atmosphere will be considered stabilized when the change in weighted 8

m 234 e

..w,, - - . _ y _- ,,% . y

B SURVEILLANCE REQUIRMENTS LIMTING CONDITIONS FOR OPERATION

-3 .-7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS -

average air temperature averaged over an hour does not deviate by more enan 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 e.ccuracy of each 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 r' ate between the type A~ test and the supplemental test.

3. Requires the quantity of gas injected into the containment or bled from 4

' 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 P t(25 psig).

235

• %** M* - :ce_.... .y,_e,-:,p.='. . . . p.ye; . m v mmmm ,__

+-

- * * 'w- . .. . . * ,s . .

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYST1'MS 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 10 closed after being opened and at least once per operating cycle. Acceptable methods of testing are halide gas detection,

' soap bubbles, pressure >

decay, hydrost.atically pressurized fluid flow er equivalent.

The personnel air lock shall be tested at a pressure of 49.6 psig t

I during each operating I cycle. ,In addition, following each opening, the personnel air lock-

• shall be leak tested at a pressure of 2.5 psig g within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the first of each series of openings whenever contain-l ment integrity is required.

l l The nersonnel air lock shall be laak tested at a pressure t

of 2.5 psig at least once every 6 months from the first of each series of openings to verify the' 236

l i

l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT.S 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 ensura the sealing function for at least 30 days at a pressute 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 with testable bellows Table 3.7.C, (3) Isolation valves with-out fluid seal - Table 3.7.D, k (4) Testable electrical penetrations - Table 3.7.H. and l

(5) Isolation valves sealed with fluid -

Tables 3.7.E, and 3.7.F.

237

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

.,- -,- -- , ~ . - - - , . ~ - - , - - - , - ,, ,-m ,--,a-,. -

~- - -- - - - ---g - , - - - , - , -r, - ,n---

0 0

e 9

0 e

(DELETED)

O 238-240

. 'Y I

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 immediately.

-(2) Ifcriterion conformance to'the 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 are effected and 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 any one main steamline l

isolation valve is ex-

' ceeded, repairs and retest shall be performed l

to correct the condition, l

l I j. Continuous Leak Rate Monitoring When the primary contain-

' ment is inerted, the containment shall be continuously monitored for gross leakage by i

review of the inerting system makeup requirements.

This monitoring 241 L

l

TABLE 3.7.D

• AIR TESTED ISOLATION VALVES yalve Valve Identification 1-14 Main Steam 1-15 Main Steam 1-26 Main Steam 1-27 Main Steam

l-37 Hain Steam -

1-38 thin Steam 1-51 . thin Steam 1-52 Main Steam 1-55 Main Steam Drain 1-56 thin 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 Lines63-525 Standby Liquid Control Discharge 63-526 Standby Liquid Control Discharge 64-17 Drywell and Suppression Chamber Air Purge inlet 64-18 Drywell Air Puege Inlet 4-19 Suppression Chamber Air Purge Inlet 64-20 Suppression Chamber Vacuum Relief 64-c.v. Supprersion Chamber Vacuum Relief 64-21 Suppr ssion Chamber Vacuum Relief.

64-c.v. Suptcession Chamber Vacuum Relief 64-29 Drywell Main Exhaust 64-30 Drywell Main Exhaust 64-32 Suppression Chamber Main Exhaust 64-33 Suppression Chamber Main Exhaust 64-31 Drywell exhaust to Standby Cas Treatment 04-34 Suppression Chamber to Standby Cas Treatment 64-139 Drywell pressurization, Compre,2 ;r Suction 64-140 Drywell pressurization, Comper- aor 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 270

TABLE 3.7.D (continud)

Valv3 Valve Identification 69-1 RWCU Supply 69-2 RWCU Supply 69-579 RWCU Return 69-624 RWCU Return 7I-2 RCIC Steam Supply 71-3 RCIC Steam Supply 71-39 RCIC Pump Discharge 71-40 RCIC Pump Discharge 73-2 RCIC Steam Supply 73-3 RCIC Steam Supply 73-44 IIPCI Pump Discharge 73-45 ilPCI Pump Discharge 74-47 RilR Shutdown Suction 74-48 RilR Shutdown Suction 74-661 RilR Shutdown Suction 74-662 RiiR Shutdown Suction 76-17 Drywell/ Suppression Chamber Nitrogen Purge Inlet 76-18 Drywell Nitrogen Purge Inlet 76-19 Suppression Chamber Purge Inlet 76-24 Drywell/ Suppression Chamber Nitrogen Purge Inlet 76-49 Containment Inerting 76-50 Containment Inerting 76-51 Containment Inerting 76-52 Containment Inerting 76-59 Containment Inerting 76-60 Containment Inerting 76-61 Containment Inerting 76-62 Containment Inerting 76-63 containment Inerting 76-64 Containment Inerting 76-65 Cbut ainment Inerting 76-66 Containment Inerting 76-67 Containment Inerting 76-68 Containment Inerting 76-215 Containment Atmospheric Monitor 76-217 Containment Atmospheric Monitor 76-220 Containment Atmospheric Monitor 76-222 Containment Atmospheric Monitor 76-225 Containment Atmospheric Monitor 76-226 Containment Atmospheric lbnitor 76-229 Containment Atmospheric Monitor 76-230 Containment Atmospheric Monitor 76-237 Containment Atmospheric Monitor 76-239 Containment Atmospheric Monitor 271

~

TABLE 3.7.D (continued)

Valve Valve Identif'eation 76-242 Containment Atmospheric Monitor 76-243 Containment Arsospheric Monitor 76-248 Containment Atmospheric Monitor 76-250 Containment Atmospheric Monitor 76-253 Containment Atmospheric Monitor 76-254 Containment Atmospheric Monitor 77-2A Drywell r'loordrain Sump 77-2B Drywell Floordrain Sump 77-15A Drywell Equipment Drain Sump 77-15B Drywell Equipment Drain Sump 84-8A Containment Atmospheric Dilution 84-8B Containment Atmospheric Dilution 84-3C Containment Atmospheric Dilution 84-8D Containment Atmospheric Dilution 84-19 Containment Atmospheric Dilution 84-20 Main Exhaust to Standby Cas Treatment 84-600 Main Exhaust to Standby Cas Treatment 84-601 Main Exhaust to Standby Gas Treatment 84-602 Main Exhaust to Standby Gas Treatment 84-603 thin Exhaust to Standby Gas Treatment 85-576 CRD Hydraul'ic 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 I

I l 272 i

9 8

a e

W G

e (DELETED)

I e

O 273-278

TABLE 3 7.E PRI:*ARY COTTAI 0*E'E ISOLATI";;! VALVES WIIICII TEICCIATE BELGI TEE SUFFRESSIO i POOL UATER IEVEL Valve Val.e Identification 12-733 Auxiliary Boiler to RCIC 12-7h1 Auxiliary Boiler to RCIC h3-28A RIIR Suppression Chaser Scmple Lines h3-233

- RIIR Suppression Chafer Sagle Lines h3-29A R*iR Suppression Chat er Saeple Lines h3-293 RiiR Suppression Cht:ber Sa iple Lines 2-1143 Demineralized Water 71-lh  ?.CIC Turbine Exhaust 71-32 RCIC Vacuus Pump Discherge 71-530 RCIC Turbine Exhaust 1-592 RCIC Vacuum Pump Discharge 73-23 hTCI Turbine Exhaust

~3-24 HICI Turbine ExhousL Drain 73-603  :!PCI Turbine Exhaust ,73-609 HICI Exhaust Drain

?h-722 RER 75-57 Core Spray to Auxiliary Boiler 75-59 Core Sprar to Auxiliary Boiler Core Spray to Auxiliary Boiler 279

' ~

T!.BLE 3 7.F ,

PRIMARY COfffAD22:E ISOLATION VALVES LOCATED El WATER SEALED SEISMIC CIASS 1 LEES valve Valve Identification 74-53 RHR LICI Discharge 74-54 RER 7h-57 RHR Suppression Chamber Spray 74-58 RER Suppression Chamber Spray 7h-60 RER Drywell Spray Th-61 RHR Drywell Spray 74-67 RIE LICI Discharge 7h-68 RHR LICI Discharge "h-71 RIE Suppression Chamber Spray

~ 4-72 RHR Suppression Chamber Spray 7h ~4 RIS Drywell Spray

?le ~5 RHR Drywell Spray "h-77 RHR Head Spray '

7h-78 RIE Head Spray 75-25 Core Spray Discharge 75-25 Core Spray Dischcrge 75-53 Core Spray Discherge 75-54 Core Spray Discharge O

280 l

4 .

TABLE 3.7.G (This table intentionally left blank) 4 o

l l

l 281-282 s

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 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 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 of an accident occuring.

The limitations 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, Pa. As an added conservatism, the measured overall integrated leakage ate is further limited to 0.75 L during performance of the periodic tests to account for possible degradati8n 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 released during primary system blowdown from 1,035 psig. Since all of the gases in the drywell are purged into the pressure suppre -lon chamber air space during a loss of coolant accident, the pressure res._cing from 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 condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

285

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 indication of -1 inch corresponds to a downcomer 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 and -8 inches without differential pressure control corresponds to a downcomer submergence of approximately 4 feet and water volume of approximately 123,000 cubic eet.

Maintaining the water 1svel between these levels will assure that the torus water volume and downcomer submergency are with 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 downcomer submergence, this specification is adequate. The maximum temperature at the end of blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 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 F.

285a

_--