ML18032A225

Proposed Tech Specs,Deleting Optional Reduced Pressure Test Method for Containment Integrated Leak Rate Test & Correcting Error in Acceptable Leakage Rate

ML18032A225
Person / Time
Site:	Browns Ferry
Issue date:	05/15/1987
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18032A224	List: ML18032A223 ML18032A225
References
TAC-00028, TAC-00029, TAC-00030, TAC-28, TAC-29, TAC-30, TAC-R00028, TAC-R00029, TAC-R00030, TAC-R28, TAC-R29, TAC-R30, NUDOCS 8705260452
Download: ML18032A225 (36)

Text

ENCLOSURE 1

PROPOSED TECHNICAL SPECIFICATIONS REVISIONS BROMNS FERRY NUCLEAR PLANT

'NITS 1, 2, AND 3 (TVA BFN TS 230) 8705260452 870515 PDR ADOCK 05000259 P

PDR

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.

Primar Containment 4.7.A.

Primar Containment 2.a.

Primary containment integrity shall be maintained at all times

'hen the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the reactor vessel except while performing "open vessel" physics tests at power levels not to exceed 5 MM(t).

b. Primary containment integrity is confirmed if the maximum allowable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the 49.6 psig design basis accident pressure, Pa.

c. If N2 makeup to the primary containment averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for

pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced 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 in Hot Shutdown within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

2. Inte rated Leak Rate Testin Primary containment nitrogen consumption shall be.

monitored to determine the average daily'nitrogen consumption for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Excessive leakage is indicated by a N2 consumption rate of > 2T. of the primary containment free volume per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-(corrected for drywell temperature,

pressure, and venting operations) at 49.6 psig.

Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCFH. If this value is exceeded, the action specified in 3.7.A.2.C shall be taken.

The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J to 10 CFR 50 using the methods and provisions of ANSI N45.4 (1972).

a.

Three type A tests (overall integrated containment leakage rate) shall be conducted at 40

+ 10-month intervals during shutdown at Pa, 49.6 psig, during each 10-year plant inservice inspection.

BFN Unit 1 3.7/4.7-3

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primar Containment 4.7.A.2.

(Cont'd)

b. If any periodic type A test fails to meet 0.75 La, the test schedule for subsequent type A tests shall be reviewed and approved by the Commission.

If two consecutive type A tests fail to meet 0.75La, a type A test shall be performed at least every 18 months until two consecutive type A tests meet 0.75 La, 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 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 />.

BFN Unit 1 3.7/4.7-4

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primar Containment-4.7.A.2. (Cont'd)

d. l. 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 La.

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.

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 Pa (49.6 psig).

BFN Unit 1 3.7.4.7-5

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Primar Containment 4.7.A Primar Containment 3.7.A.4.b (Cont'd) as it is determined to be not more than 3'pen as indicated by'he position lights.

4.7.A.4.b (Cont'd) time when operability is required all other vacuum breaker valves shall be exercised immediately and every 15 days thereafter until the INOPERABLE valve has been returned to normal service.

c.

Two drywell-suppression chamber vacuum breakers may be determined to be INOPERABLE for opening.

c ~ Once each operating cycle each vacuum breaker valve shall be inspected for proper operation of the valve and limit switches.

d. If Specifications 3.7.A.4.a, 3.7.A.4.b, or 3.7.A.4.c.

cannot be met, the unit shall be placed in a Cold Shutdown condition in an orderly manner within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d.

A leak test of the drywell to suppression chamber structure shall be conducted during each operating cycle.

Acceptable leak rate is 0.09 lb/sec of primary containment atmosphere with 1 psi differential.

5.

Ox en Concentration 5.

Ox en Concentration a.

After completion of the fire-related startup retesting

program, containment atmosphere shall be reduced to less than 4V. oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.S.b.

a.

The primary containment oxygen concentration shall be measured and recorded daily.

The oxygen measurement shall be adjusted to account for the uncertainty of the method used by adding a predetermined error function.

b.

Within the 24-hour period subsequent to placing the reactor in the RUN mode following a shut-down, the containment atmosphere oxygen concentration shall be reduced to less than 4% by volume and maintained in this condition.

Deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown.

b.

The methods used to measure the primary containment oxygen concentration shall be calibrated once every refueling cycle.

c. If Specification 3.7.A.5.a and 3.7.A.S.b cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

BFN Unit 1 3.7/4.7-11

3.7/4 7

BASES (Cont'd) be inoperable or the valve disc is stuck.

For this case, a check light on and red light off confirms the disc is in a nearly closed position even if one of the indications is in error.

Although the valve may be inoperable for full closure, it does not constitute a safety threat.

If the red light circuit alone is inoperable, the valve shall still be considered fully operable.

If the green and red or the green light circuit alone is inoperable the valve shall be considered inoperable for opening. If the check and green or check light circuit alone is inoperable, the valve shall be considered inoperable for full closure. If the red and check light circuits are inoperable the valve shall be considered inoperable and open greater than 3'.

For a light circuit to be considered operable the light must go on and off in proper sequence during the opening-closing cycle. If none of the lights change indication during the cycle, the valve shall be considered inoperable and open unless the check light stays on and the red light stays off in which case the valve shall be considered inoperable for opening.

The 12 drywell vacuum breaker valves which connect the suppression chamber and drywell are sized on the basis of the Bodega pressure suppression system tests.

Ten operable to open vacuum breaker valves (18-inch) selected on this test basis and confirmed by the green lights are adequate to limit the pressure differential between the suppression chamber and drywell during postaccident drywell cooling operations to a value which is within suppression system design values.

The containment design has been examined to determine that a leakage equivalent to one drywell vacuum breaker opened to no more than a nominal 3's confirmed by the red light is acceptable.

On this basis an indefinite allowable repair time for an inoperable red light circuit on any valve or an inoperable check and green or check light circuit alone or a malfunction of the operator or disc (if nearly closed) on one valve, or an inoperable green and red or green light circuit alone on two valves is justified.

During each operating cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber.

The drywell pressure will be increased by at least one psi with respect to the suppression chamber pressure and held constant.

The two psig setpoint will not be exceeded.

The subsequent suppression chamber pressure transient (if any) will be monitored with a sensitive pressure gauge.

If the drywell pressure cannot be increased by one psi over the suppression chamber pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed.

With a differential pressure of greater than one psig, the rate of change of the suppression chamber pressure must not exceed 0.25 inches of water per minute as measured over a 10-minute period, which corresponds to about 0.09 lb/sec of containment air.

In the event the rate of change exceeds this value then the source of leakage will.be identified and eliminated before power operation is resumed.

BFN Unit 1 3.7/4.7-45

3.7/4.7 BASES (Cont'd)

The water in the suppression chamber is used for cooling in the event of an accident; i.e., it is not used for normal operation; therefore, a daily check of the temperature and volume is adequate to assure that adequate heat removal capability is present.

4 The interior surfaces of the drywell and suppression chamber are coated as necessary to provide corrosion protection and to provide a more easily decontaminable surface.

The surveillance inspection of the internal surfaces each operating cycle assures timely detection of corrosion.

Dropping the torus water level to one foot below the normal operating level enables an inspection of the suppression chamber where problems would first begin to show.

The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a LOCA.

The peak drywell pressure would be about 49 psig which would rapidly reduce to less than 30 psig within 20 seconds following the pipe break.

Following the pipe break, the suppression chamber pressure rises to 27 psig within 25 seconds, equalizes with drywell pressure, and decays with the drywell pressure decay.

The design pressure of the drywell and suppression chamber is 56 psig.

The design leak rate is 0.5-percent per day at the pressure of 56 psig.

As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after about 25 seconds.

Based on the calculated containment pressure response discussed

above, the primary containment preoperational test pressures were chosen.

Also based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.

The calculated'radiological doses given in Section 14.9 of the FSAR were based on an assumed leakage rate of 0.635-percent at the maximum calculated pressure of 49.6 psig.

The doses calculated by the NRC using this bases are 0.14 rem, whole body passing cloud gamma dose, and 15.0 rem, thyroid dose, which are respectively only 5 x 10

~ and 10 1 times the 10 CFR 100 reference doses.

Increasing the assumed leakage rate at 49.6 psig to 2.0 percent as indicated in the specifications would increase these doses approximately a factor of three, still leaving a margin between the calculated dose and the 10 CFR 100 reference values.

Establishing the test limit of 2.0-percent/day provides an adequate margin of safety to assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime.

Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak rate.

The allowable operational leak rate is derived by multiplying the maximum allowable leak rate by 0.75 thereby providing a 25-percent margin to allow for leakage deterioration which may occur during the period between leak rate tests.

P BFN Unit 1 3.7/4.7-46

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.

Primar Containment 4.7.A.

Primar Containment 2.a.

Primary containment integrity shall be maintained at all times

'hen the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the reactor vessel except while performing "open vessel" physics tests at power levels not to exceed 5 MW(t) ~

b. Primary containment integrity is confirmed if the maximum allowable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the 49.6 psig design basis accident pressure, Pa.

c. If N2 makeup to the primary containment averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for

pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced 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 in Hot Shutdown

, within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

2. Inte rated Leak Rate Testin Primary containment nitrogen consumption shall be monitored to determine the average daily'nitrogen consumption for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Excessive leakage is indicated by a N2 consumption rate of > 2T. of the primary containment free volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for drywell temperature,

pressure, and venting operations) at 49.6 psig.

Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCFH. If this value is exceeded, the action specified in 3.7.A.2.C shall be taken.

The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J to 10 CFR 50 using the methods and provisions of ANSI N45.4 (1972).

a.

Three type A tests (overall integrated containment leakage rate) shall be conducted at 40

+ 10-month intervals during shutdown at Pa, 49.6 psig, during each 10-year plant inservice inspection.

BFN Unit 2 3.7/4.7-3

0

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primar Containment 4.7.A.2.

(Cont'd)

b. If any periodic type A test fails to meet 0.75 Lag the test schedule for subsequent type A tests shall be reviewed and approved by the Commission.

If two consecutive type A tests fail to meet 0.75Laf a type A test shall be performed at least every 18 months until two consecutive type A tests meet 0.75 La, 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 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 />.

BFN Unit 2 3.7/4.7-4

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primar Containment 4.7.A.2.

(Cont'd)

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 La.

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.

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 Pa (49.6 psig).

BFN Unit 2 3.7/4.7-5

i 3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Primar Containment 3.7.A.4.b (Cont'd) 4.7.A Primar Containment 4.7.A.4.b (Cont'd) as it is determined to be not more than 3

open as indicated by the position lights.

time when operability is required all other vacuum breaker valves shall be exercised immediately and every 15 days thereafter until the INOPERABLE valve has been returned to normal service.

c.

Two drywell-suppression chamber vacuum breakers may be determined to be INOPERABLE for opening.

c ~ Once each operating cycle each vacuum breaker valve shall be inspected 'for proper operation of the valve and limit switches.

d. If Specifications 3.7.A.4.a,

.b, or.c cannot be met, the unit shall be placed in a Cold Shutdown condition in an orderly manner within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d.

A leak test of the drywell to suppression chamber structure shall be conducted during each operating cycle.

Acceptable leak rate is 0.09 lb/sec of primary containment atmosphere with 1 psi differential.

5.

Ox en Concentration 5.

Ox en Concentration a.

After completion of the fire-related startup retesting

program, containment atmosphere shall be reduced to less than 4'L oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.5.b.

a.

The primary containment oxygen concentration shall be measured and recorded daily.

The oxygen measurement shall be adjusted to account for the uncertainty of the method used by adding a predetermined error function.

b.

Mithin the 24-hour period subsequent to placing the reactor in the RUN mode following a shut-down, the containment atmosphere oxygen concentration shall be reduced to less than 4V. by volume and maintained in this condition.

Deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown,

b. The methods used to measure the primary containment oxygen concentration shall be calibrated once every refueling cycle.

c. If Specification 3.7.A.5.a and 3.7.A.5.b cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

BFN Unit 2 3.7/4.7-11

3.7/A.7 BASES (Cont'd) be inoperable or the valve disc is stuck.

For this case, a check light on and red light off confirms the disc is in a nearly closed position even if one of the indications is in error.

Although the valve may be inoperable for full closure, it does not constitute a safety threat.

If the red light circuit alone is inoperable, the valve shall still be considered fully operable.

If the green and red or the green light circuit alone is inoperable the valve shall be considered inoperable for opening. If the check and green or check light circuit alone is inoperable,- the valve shall be considered inoperable for full closure.

If the red and. check light circuits are inoperable the valve shall be considered inoperable and open greater than 3'.

For a light circuit to be considered operable the light must go on and off in proper sequence during the opening-closing cycle. If none of the lights change indication during the cycle, the valve shall be considered inoperable and open unless the check light stays on and the red light stays off in which case the valve shall be considered inoperable for opening.

The 12 drywell vacuum breaker valves which connect the suppression chamber and drywell are sized on the basis of the Bodega pressure suppression system tests.

Ten operable to open vacuum breaker valves (18-inch) selected on this test basis and confirmed by the green lights are adequate to limit the pressure differential between the suppression chamber and drywell during postaccident drywell cooling operations to a value which is within suppression system design values.

The containment design has been examined to determine that a leakage equivalent to one drywell vacuum breaker opened to no more than a nominal 3's confirmed by the red light is acceptable.

On this basis an indefinite allowable repair time for an inoperable red light circuit on any valve or an inoperable check and green or check light circuit alone or a malfunction of the operator or disc (if nearly closed) on one valve, or an inoperable green and red or green light circuit alone on two valves is justified.

During each operating cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber.

The drywell pressure will be increased by at least one psi with respect to the suppression chamber pressure and held constant.

The two psig setpoint will not be exceeded.

The subsequent suppression chamber pressure transient (if any) will be monitored with a sensitive pressure gauge. If the drywell pressure cannot be increased by one psi over the suppression chamber pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed.

With a differential pressure of greater than one psig, the rate of change of the suppression chamber pressure must not exceed 0.25 inches of water per minute as measured over a 10-minute period, which corresponds to about 0.09 lb/sec of containment air.

In the event the rate of change exceeds this value then the source of leakage will.be identified and eliminated before power operation is resumed.

BFN Unit 2 3.7/4.7-45

3.7/4.7 BASES (Cont'd)

The water in the suppression chamber is used for cooling in the event of an accident; i.e., it is not used for normal operation; therefore, a daily check of the temperature and volume is adequate to assure that adequate heat removal capability is present.

nn The interior surfaces of the drywell and suppression chamber are coated as necessary to provide corrosion protection and to provide a more easily decontaminable surface.

The surveillance inspection of the internal surfaces each operating cycle assures timely detection of corrosion.

Dropping-the torus water level to one foot below the normal operating level enables an inspection of the suppression chamber where problems would first begin to show.

The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a LOCA.

The peak drywell pressure would be about 49 psig which would rapidly reduce to less than 30 psig within 20 seconds following the pipe break.

Following the pipe break, the suppression chamber pressure rises to 27 psig within 25 seconds, equalizes with drywell pressure, and decays with the drywell pressure decay.

The design pressure of the drywell and suppression chamber is 56 psig.

The design leak rate is 0.5-percent per day at the pressure of 56 psig.

As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after about 25 seconds.

Based on the calculated containment pressure response discussed

above, the primary containment preoperational test pressures were chosen.

Also based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.

The calculated radiological doses given in Section 14.9 of the FSAR were based on an assumed leakage rate of 0.635-percent at the maximum calculated pressure of 49.6 psig.

The doses calculated by the NRC using this bases are 0.14 rem, whole body passing cloud gamma dose, and 15.0 rem, thyroid dose, which are respectively only 5 x 10 3 and 10 times the 10 CFR 100 reference doses.

Increasing the assumed leakage rate at 49.6 psig to 2.0 percent as indicated in the specifications would increase these doses approximately a factor of three, still leaving a margin between the calculated dose and the 10 CFR 100 reference values.

Establishing the test limit of 2.0-percent/day provides an adequate margin of safety to assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime.

Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak rate.

The allowable operational leak rate is derived by multiplying the maximum allowable leak rate by 0.75 thereby providing a 25-percent margin to allow for leakage deterioration which may occur during the period between leak rate tests.

BFN Unit 2 3.7/4.7-46

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A.

Primar Containment 4.7.A.

Primar Containment 2.a.

Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212'F and fuel is in the reactor vessel except while performing "open vessel" physics tests at power levels not to exceed 5 MW(t).

b. Primary containment integrity is confirmed if the maximum allowable integrated leakage rate, La, does not exceed the equivalent of 2 percent of the primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the 49.6 psig design basis accident pressure, Pa.

c. If N2 makeup to the primary containment averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (corrected for

pressure, temperature, and venting operations) exceeds 542 SCFH, it must be reduced 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 in Hot Shutdown within the next 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

2. Inte rated Leak Rate Testin Primary containment nitrogen consumption shall be monitored to determine the average daily'nitrogen consumption for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Excessive leakage is indicated by a N2 consumption rate of > 2T. of the primary containment free volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

'corrected for drywell temperature,

pressure, and venting operations) at 49.6 psig.

Corrected to normal drywell operating pressure of 1.1 psig, this value is 542 SCFH. If this value is exceeded, the action specified in 3.7.A.2.c shall be taken.

The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J to 10 CFR 50 using the methods and provisions of ANSI N45.4 (1972).

a.

Three type A tests (overall integrated containment leakage rate) shall be conducted at 40

+ 10-month intervals during shutdown at Pa, 49.6 psig, during each 10-year plant inservice inspection.

BFN Unit 3 3.7/4.7-3

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primar Containment 4.7.A.2. (Cont'd)

b. If any periodic type A test fails to meet 0.75 Lag the test schedule for subsequent type A tests shall be reviewed and approved by the Commission.

If two consecutive type A tests fail to meet 0.75 La, a type A test shall be performed at least every 18 months until two consecutive type A tests meet 0.75 Laf 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 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

~ hours.

BFN Unit 3 3.7/4.7-4

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.7.A.

Primar Containment 4.7.A.2.

(Cont'd)

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 La.

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.

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 Pa (49.6 psig).

BFN Unit 3 3.7/4.7-5

3.7/4.7 CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.A Primar Containment 4.7.A Primar Containment 3.7.A.4.b (Cont'd) 4.7.A.4.b (Cont'd) as it is determined to be not more than 3'pen as indicated by the position lights.

time when operability is required, all other vacuum breaker valves shall be exercised immediately and every 15 days thereafter until the INOPERABLE valve has been returned to normal service.

c.

Two drywell-suppression chamber vacuum breakers may be determined to be INOPERABLE for opening.

co Once each operating cycle, each vacuum breaker valve shall be inspected for proper operation of the valve and limit switches.

d. If Specifications 3.7.A.4.a, 3.7.A.4.b, or 3.7.A.4.c, cannot be met, the unit shall be placed in a Cold Shutdown condition in an orderly manner within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d. A leak test of the drywell to suppression chamber structure shall be conducted during each operating cycle.

Acceptable leak rate is 0.09 lb/sec of primary containment atmosphere with 1 psi differential.

5.

Ox en Concentration 5.

Ox en Concentration a.

After completion of'he 300-hour warranty run, containment atmosphere shall be reduced to lass than 4'L oxygen with nitrogen gas during reactor power operation with reactor coolant pressure above 100 psig, except as specified in 3.7.A.5.b.

a.

The primary containment oxygen concentration shall be measured and recorded daily.

The oxygen measurement shall be adjusted to account for the uncertainty of the method used by adding a predetermined error function.

b.

Mithin the 24-hour period subsequent to placing, the reactor in the RUN mode following a shut-down, the containment atmosphere oxygen concentration shall be reduced to less than

47. by volume and maintained in this condition.

Deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown.

b.

The methods used to measure the primary containment oxygen concentration shall be calibrated once every refueling cycle.

c. If the specifications of 3.7.A.5.a through 3.7.A.S.b cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

BFN Unit 3 3.7/4.7-11

3.7/4.7 BASES (Cont'd) be inoperable or the valve disc is stuck.

For this case, a check light on and red light off confirms the disc is in a nearly closed position even if one of the indications is in error.

Although a valve may be inoperable for full closure, it does not constitute a safety threat.

If the red light circuit alone is inoperable, the valve shall still be considered fully operable.

If the green and red or the green light circuit alone is inoperable the valve shall be considered inoperable for opening. If the check and green or check light circuit alone is inoperable,. the valve shall be considered inoperable for full closure. If the red and check light circuits are inoperable the valve shall be considered inoperable and'open greater than 3'.

For a light circuit to be considered operable the light must go on and off in proper sequence during the opening-closing cycle. If none of the lights change indication during the cycle, the valve shall be considered inoperable and open unless the check light stays on and the red light stays off in which case the valve shall be considered inoperable for opening.

The 12 drywell vacuum breaker valves which connect the suppression chamber and drywell are sized on the basis of the Bodega pressure suppression system tests.

Ten operable to open vacuum breaker valves (18-inch) selected on this test basis and confirmed by the green lights are adequate to limit the pressure differential between the suppression chamber and drywell during postaccident drywell cooling operations to a value which is within suppression system design values.

The containment design has been examined to determine that a leakage equivalent to one drywell vacuum breaker opened to no more than a nominal 3

as confirmed by the red light is acceptable.

On this basis an indefinite allowable repair time for an inoperable red light circuit on any valve or an inoperable check and green or check light circuit alone or a malfunction of the operator or disc (if nearly closed) on one valve, or an inoperable green and red or green light circuit alone on two valves is justified.

During each operating

cyc1e, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber.

The drywell pressure will be increased by at least one psi with respect to the suppression chamber pressure and held constant.

The two psig setpoint will not be exceeded.

The subsequent suppression chamber pressure transient (if any) will be monitored with a sensitive pressure gaure.

If the drywell pressure cannot be increased by one psi over the suppression chamber pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed.

With a differentia1 pressure of greater than one psig, the rate of change of the suppression chamber pressure must not exceed 0.25 inches of water per minute as measured over a 10-minute period, which corresponds to about 0.09 lb/sec of containment air.

In the event the rate of change exceeds this value then the source of leakage will Pe identified and eliminated before power operation is resumed.

BFN Unit 3 3.7/4.7-43

3.7/4.7 BASES (Cont'd)

The water in the suppression chamber is used for cooling in the event of an accident; i.e., it is not used for normal operation; therefore, a daily check of the temperature and volume is adequate to assure that adequate heat removal capability is present.

The interior surfaces of the drywell and suppression chamber are coated as necessary to provide corrosion protection and to provide a more easily decontaminable surface.

The surveillance inspection of the internal surfaces each operating cycle assures timely detection of corrosion.

Dropping the torus water level to one foot below the normal operating level enables an inspection of the suppression chamber where problems would first begin to show.

The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident.

The peak drywell pressure would be about 49 psig which would rapidly reduce to less than 30 psig within 20 seconds following the pipe break.

Following the pipe break, the suppression chamber pressure rises to 27 psig within 25 seconds, equalizes with drywell pressure, and decays with the drywell pressure decay.

The design pressure of the drywall and suppression chamber is 56 psig.

The design leak rate is 0.5-percent per day at the pressure of 56 psig.

As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after about 25 seconds.

Based on the calculated containment pressure response discussed

above, the primary containment preoperational test pressures were chosen.

Also based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.

The calculated radiological doses given in Section 14.9 of the FSAR were based on an assumed leakage rate of 0.635-percent at the maximum calculated pressure of 49.6 psig.

The doses calculated by the NRC using this Bases are 0.14 rem, whole body passing cloud gamma dose, and 15.0 rem, thyroid dose, which are respectively only 5 x 10

~ and 10 1 times the 10 CFR 100 reference doses.

Increasing the assumed leakage rate at 49.6'sig to 2.0 percent as indicated in the specifications would increase these doses approximately a factor of three, still leaving a,margin between the calculated dose and the 10 CFR 100 reference values.

HstablishinF tho tost limit of 2,0-percent/day provides an adequate margin of safety to assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime.

Additional margin to maintain the containment in the "as built" condition is achieved by establishing the allowable operational leak rate.

The allowable operational leak rate is derived by multiplying the maximum allowable leak rate by 0.75 thereby providinF a 25-percent margin to allow for leakage deterioration which may occur during, the period between leak rate tests.

BFN Unit 3 3.7/4.7-44

ENCLOSURE 2

DESCRIPTION AND JUSTIFICATION BROMNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 Descri tion of Chan e

The Browns Ferry Nuclear Plant units 1, 2, and 3 Technical Specifications are being revised to delete references to an optional reduced pressure test method for the integrated leak rate test.

The full pressure test requirements will not be affected.

This revision is made in Surveillance Requirements (SR) 4.7.A.2.a, 4.7.A.2.b, 4.7.A.2.f.l, and 4.7.2.f.3. It is also reflected in the bases for section 4.7.A.

A second change is being made to SR 4.7.A.4.d and the bases to correct the acceptable leak rate of drywell atmosphere to the suppression chamber with a one psi differential pressure.

The limits currently stated in the technical specifications are 0.38 inches of water per minute pressure change in the suppression pool, which corresponds to the 0.14 pound per second of containment air leakage specified in SR 4.7.A.4.d.

The correct limits should be 0.25 inches of water per minute and 0.09 pounds per second of containment air.

Reason for Chan e

It was formerly believed that a correlation existed between data obtained from the performance of a reduced pressure test (25 psig) and a full pressure test (49.6 psig), allowing a prediction of the full pressure leakage by the performance of a reduced pressure test only.

However, experience has shown that no correlation exists between the results of each test.

Therefore, BFN proposed to delete the low pressure test for clarity and conciseness.

The change to the acceptable leakage rate is to correct an error in accordance with TVA's response, (U5.1-5, Section 3.3) to the Atomic Energy Commission (AEC), question 5.1 dated December 6,

1971.

Justification for Chan e The controlling factor for BFN leak rate testing is the requirement to meet 10 CFR 50 Appendix J.

Although Appendix J currently allows periodic integrated leakage rate test (ILRT) to be conducted at a reduced test

pressure, section III.A.2 of the proposed new Appendix J will eliminate all reduced pressure testing as an option for the ILRT and require that ILRTs only be done at full pressure.

TVA currently conducts all periodic containment ILRTs at full test pressure in compliance with this proposed change to Appendix J to 10 CFR 50.

TVA has conducted preoperational ILRTs at both pressures with the purpose of establishing a relationship between the measured leakage at half pressure and full pressure.

However, these test results do not show a clear correlation between the reduced pressure and full pressure leakage.

In fact, two cases showed increased leakage occurring due to the inception of leaks brought on by increasing pressure, demonstrating, that leakage paths closed at lower pressure may open at full pressure.

Justification for Chan e (Cont'd)

For these reasons, deleting the option to perform an ILRT at reduced pressure is a conservative change that will not reduce the margin of safety.

The original technical specifications listed the acceptable drywell to suppression chamber leakage as 0.25 inches of water per minute pressure change and 0.14 pounds per second.

These numbers were supposed to correspond to each other, but they do not. It was realized that an inconsistency existed and therefore, the technical specifications were changed by amendment Nos.

114, 108, and 82, respectively.

However, this change was based on the assumption that the limit of 0.14 pounds per second given in technical specification 4.7.A.4.d was correct and the 0.25 inches of water per minute stated in the bases was wrong.

The opposite of that previous assumption is actually true.

This amendment is based on TVA's response to the AEC question 5.1 which states a limit of 0.25 inches of water per minute in the acceptability section (U5.1-5, Section 3.3). It should also be noted that the proposed change is in the conservative direction in that it will allow less leakage.

Therefore, this change will not reduce the margin of safety.

ENCLOSURE 3 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROMNS FERRY NUCLEAR PLANT Descri tion of Amendment Re uest The proposed amendment would change the Browns Ferry Nuclear Plant (BFN)

Technical Specifications for units 1, 2, and 3 by deleting an option to perform. the containment integrated leak rate test (ILRT) at a reduced-test pressure and by correcting an error in the acceptable leakage rate.

The first requested change will simply remove an option that exists. in the current technical specifications, and thereby require the BFN conduct the ILRT at full pressure.

This is the current BFN practice since the optional reduced pressure test may not provide conservative results.

The second requested change will correct an error in the limit listed as acceptable for atmospheric leakage from the drywell to suppression chamber with a one psi pressure difference.

The current limit of 0.14 pound per second of drywell atmosphere leakage to the suppression chamber will be changed to 0.09 pound per second.

Basis for Pro osed No Si nificant Hazards Consideration Determination The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 59.92(c).

A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not:

(1) involve a significant increase in the probability or consequences of accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction, in a margin of safety.

1.

Involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.

Neither of the proposed amendments would affect the probability of an accident since one only deletes an optional test method and the other only affects a surveillance test acceptability limit.

Likewise, the first change has no affect on the consequences of an accident since the limiting conditions will still be verified by the remaining surveillance requirements.

The second change will actually reduce the allowable leakage from the drywell to the suppression chamber and could therefore further mitigate the consequences of an accident.

2.

Create the ossibilit of a new or different kind of accident from an accident reviousl evaluated.

These technical specification changes will not eliminate or modify any protective functions nor permit any new operational conditions.

Deletion of an appropriate reduced pressure leak test and the reduction in the allowable leakage limit will therefore not create the possibility of a new or different kind of accident.

Basis for Pro osed No Si nificant Hazards Consideration Determination (Cont'd) 3.

Involve a si nificant reduction in a mar in of safet These changes do not modify the intent of the technical specifications.'eleting the optional test method will only result in the test having to be performed by the approved method that remains in the technical specifications.

Correcting the leakage limit is consi:stent. with the originally intended margin of safety.

Since the application for amendment involves proposed changes that are encompassed by the criteria for which no significant hazards consideration

exists, TVA has made a proposed determination that the application involves no significant hazards consideration.

S Ihl I

l

ENCLOSURE 3

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROWNS FERRY NUCLEAR PLANT Descri tion of Amendment Re uest The proposed amendment would change the Browns Ferry Nuclear Plant (BFN)

Technical Specifications for units 1, 2, and 3 by deleting an option to perform the containment integrated leak rate test (ILRT) at a reduced test pressure and by correcting an error in the acceptable leakage rate.

The first requested change will simply remove an option that exists in the current technical specifications, and thereby require the BFN conduct the ILRT at full pressure.

This is the current BFN practice since the optional reduced pressure test may not provide conservative results.

The second requested change will correct an error in the limit. listed as acceptable for atmospheric leakage from the drywell to suppression chamber with a one psi pressure difference.

The current limit of 0.14 pound per second of drywell atmosphere leakage to the suppression chamber will be changed to 0.09 pound per second.

Basis for Pro osed No Si nificant Hazards Consideration Determination The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 59.92(c).

A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not:

(1) involve a significant increase in the probability or consequences of accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

1.

Involve a si nificant increase in the robabilit or conse uences of an accident reviousl evaluated.

Neither of the proposed amendments would affect the probability of an accident since one only deletes an optional test method and the other only affects a surveillance test acceptability limit.

Likewise, the first change has no affect, on the consequences of an accident since the limiting conditions will still be verified by the remaining surveillance requirements.

The second change will actually reduce the allowable leakage from the drywell to the suppression chamber and could therefore further mitigate the consequences of an accident.

2.

Crea"

. the ossibilit of a new or different kind of accident from an accident reviousl evaluated.

These technical specification changes will not eliminate or modify any protective functions nor permit any new operational conditions.

Deletion of an appropriate reduced pressure leak test and the reduction in the allowable leakage limit will therefore not create the possibility of a new or different kind of accident.