ML20097B447

Proposed Tech Specs,Supplementing Amend Request Supporting Implementation of performance-based Containment Lrt Requirements

ML20097B447
Person / Time
Site:	Brunswick
Issue date:	01/29/1996
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20097B432	List: BSEP-96-0041, Submits Suppl to Amends to Licenses DPR-71 & DPR-62, Implementing performance-based Containment Lrt Requirements ML20097B447
References
NUDOCS 9602070065
Download: ML20097B447 (45)

Text

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CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS  !

l 4.6.1.3 Each primary containment air lock shall be demonstrated OPERABLE:

i

a. By verifying the seal leakage rate to be less than or equal to 5 scf '

when the gap between the door seals is pressurized to 10

1. Within 7 days following each closing, except when the air lock is being used for multiple entries. then at least once per 30 i days, and l l

2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air i lock has been used and no maintenance has been performed on the I air lock and l

3. When the air lock seal has been replaced.

b. By conducti6g an overall air loc's leakage test at P . 49 psig, and by verifying that the overall air lock leakage is with,in its limit: I

1. At least once per 30 months, and l

2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when maintenance (except for seal replacement) has been performed on y the air lock that would affect the air lock sealing capability.&

c. By verification of air lock interlock OPERABILITY:

1. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air lock has been used and

2. prior to and following a drywell entry when PRIMARY CONTAINMENT INTEGRITY is required, and

3. Following the performance of maintenance affecting the air lock interlock.

9602070065 960129 PDR ADOCK 05000324 p PDR r

mxcmption tc Appendiv J of 10 CFR L.

I BRUNSWICK - UNIT 1 3/4 6-5 Amendment No. I

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures 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.

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The safety design basis for the primary containment is that it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

~

The DBA that postulates the 'naximum release of radioactive material within primary containment is a LOCA. In analysis of this accident.1; is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary coltainnent leakage.

Analytical methods and assumptions involving the primar/ crntainment are presented in 9eferences 6 and 7.

The maximum allowable leakage rate for the primary containment (L.) is 0.5 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the maximum peak containment pressure (P,) of 49 psig.

A Primary Containment Leakage Rate Testing Program has been established a accordance with 10 CFR 50.54(o) to implement the requirements of 1.0 CFF Part 50. Appendix J. Option B (Reference 1). The Primary Contairaient Leaka Rate Testing Program conforms with NRC Regulatory Guide 1.163.@=cn@ge -

dated September 1995. ' Performance-Based Containment Leak-Rate Testing Program" (Reference 2) and Nuclear Energy Institute (NEI) 94-01. Revision 0, dated July 26.1995. " Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J' (Reference 3) with the exception of:

1. NEI 94-01. Section 8.0. " Testing Methodologies for Type A. B and C Tests" states that " Type A. Type B and Type C tests should be performed using the technical methods and techniques specified in ANSI /ANS 56.8-1994, or other alternative testing methods that have been approved by the NRC." The Brunswick Plant takes exception to ANSI 56.8 flowmeter accuracy requirements based upon compensation of instrument inaccuracies applied to the containment leakage total per the previous revision of the standard.

Brunswick Plant administrative procedures and databases already effectively address instrument error. Brunswick Plant uses standard glass tube and ball type flowmeters with a 5 percent of full scale accuracy.

Readings are compensated for back pressure, temperature, and test medium variables. To overcome the less accurate flowmeter use an equipment error is applied to the results of each test. The square root of the sum of the squares of the equipment errors for the tests is also added to the cumulative containment leakage total. This method is consistent with ANSI 56.8-1987 Appendix E and provides conservative assurance that the BRUNSWICK - UNIT 1 B 3/4 6-1 Amendment No. l

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

9 .

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT l l

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE (Continued) i cumulative containment leakage total accounts for instrument inaccuracy.

No such instrument error analysis or accounting is required per ANSI /ANS 56.8-1994.

2. INsteT The leakage rate acceptance criteria of 5 0.60 L, for the combined Type B and C tests and s 0.75 L for the Type A test ensures a primary containment configuration. including, equipment hatches. that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

Primary containment operability is maintained by limiting leakage to s 1.0 L,.

Individual leakage rates specified for the primary containment air lock are addressed in Specification 3.6.1.3.

OperatingexperienceYiththemainsteamlineisolationvalveshasindicated that degradation has occasionally occurred in the leak tightness of the valves; therefore, the special requirement for testing these valves. i Exemptions from the requirements of 10 CFR Part 50 have been granted for the main steam isolation valve leak testing and leakage calculations. I NRC Regulatory Guide 1.163. 6/%Meference 2) endorses NEI 94-01 (Reference 3) which in turn identifies ANSI /ANS 56.8-1994. " Containment System Leakage Testing Requirements" (Reference 4) as an acceptable standard regarding leakage-rate test methods, procedures, and analyses. Reduced ~

duration Type A tests may be performed using the criteria and Total Time Method specified in Bechtel Topical Report BN-TOP-1, Revision l. November 1.

1972 (References 5 and 6).

References:

1. 10 CFR Part 50. Appendix J, op%n %

2. NRCRegulatoryGuide1.163.SEEII ed September 1995.

" Performance-Based Containment Leck-Rate Testing Program."

3. Nuclear Energy Institute Guideline 94-01. Revision 0, dated July 26. 1995.

" Industry Guideline for Implementirg Performance-Based Option of 10 CFR 50 Appendix J."

4. ANSI /ANS 56.8-1994. " Containment System Leakage Testing Requirements"

5. CP&L Letter to Mr. D. B. Vassallo " Integrated Leak Rate Test."

October 20. 1983.

6. - NRC Letter from Mr. D. B. Vassallo to Mr. E. E. Utley. December 9,1983.

7. Updated FSAR. Section 6.2.

8. Updated FSAR. Section 15.6.4.

BRUNSWICK - UNIT 1 8 3/4 6-la Amendment No. l

4 BASES INSERT:

2. NEl 94-01, Section 10.2.2.2, " Repairs or Adjustments of Airlocks" states that

following maintenance on an air lock pressure retaining boundary, one of the i following tests shall be completed

a. The air lock shall be tested at a pressure of not less than P , or

b. Leakage rate testing at P, shall be performed on the affected area or

]

component.

i A previously approved exemption to 10 CFR 50, Appendix J thr. allows the 4 performance of air lock door seal leakage rate testing at a pressure less than P,

. following door seal replacement instead of air lock testing at P, has been

, retained and is listed as an exception in Technical Specification 6.8.3.4.

1 4

l l

CONTAINMENT SYSTEMS

. . BASES 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The primary containment air lock forms part of the primary containment pressure boundary. As such, air lock integrity and leak tightness are essential for maintaining primary containment leakage rate to within limits in the event of a DBA. Not maintaining air lock integrity or leak tightness may

, result in a leakage rate in excess of t'nat assumed in unit safety analysis.

The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage. In the analysis of this accident. it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary -t containment is designed with a maximum allowable leakage rate (L.) of ,

0.5 percent by weight _of the containment air ]er 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' at the maximum peak containment pressure (P,) of 49 psig. This a'lowable leakage rate forms the basis for the acceptance criteria imposed on the surveillance requirements associated with the air lock.

The primary containment air lock is required to be OPERABLE. For the air lock to be considered OPERABLE. the air lock interlock mechanism n.ust be OPERABLE.

the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be 4

OPERABLE. Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry and exit from primary containment.

Maintaining primary containment air locks OPERABLE requires compliance with '

the leakage rate test requirements of 10 CFR 50. Appendix J as established in the Primary Containment Leakage Rate Testing Program. The Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50. Appendix J.

Option B (Reference 1). The Primary Containment Leakage Rate Testing Program conforms with NRC Regulatory Guide 1.163. e e cn O M ted September 1995.

" Performance-Based Containment Leak-Rate Testing Program" and Nuclear Energy c Institute (NEI) 94-01, Revision 0. dated July 26, 1995. " Industry Guideline fog Implementing Perfo mance-Based Option of 10 CFR 50 Appendix J' as modified by - - e,xceptions (References 2 and 3).

4he. kstxd in SpeemcoMen G 8 3.4.

An inoperable air lock door does not invalidate the ]revious successful performance of the overall air lock leakage test. T1is is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA.

Only one closed door in each air lock is required to maintain the integrity of the containment. In the event of an inoperable door interlock, locking shut the inner door will ensure containment integrity while permitting access to the lock for maintenance and surveillance testing.

BRUNSWICK - UNIT 1 B 3/4 6-2 Amendment No. l

CONTAINMENT SYSTEMS BASES 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS (Continued)

References:

1. 10 CFR Part 50. Appendix J) Op%n 3,

2. NRC Regulatory Guide 1.163, m9vich-fG dated September 1995.

" Performance-Based Containment Leak-Rate Testing Program."

3. Nuclear Energy Institute Guideline 94-01. Revision 0, dated July 26, 1995.

" Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J." l 3/4.6.1.4 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the primary  !

containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 49 psig in the event of a LOCA. A visual inspection in conjunction with the Primary -

Containment Leakage Rate Testing Program is sufficient to demonstrate this capability.

References:

1. 10 CFR Part 50. Appendix J. Option B.Section III.A. I

2. NRC Regulatory Guide 1.163.6c11:icn ted September 1995.

" Performance-Based Containment Leak-Rate Testing Program."

3/4.6.1.5 PRIMARY CONTAINMENT INTERNAL PRESSURE The limitations of primary containment internal pressure ensure that the '

containment peak pressure of 49 psig does not exceed the design pressure of 62 psig during LOCA conditions. The limit of 1.75 psig, for initial positive containment pressure will limit the total pressure to 49 psig, which is less than the design pressure and is consistent with the accident analyses.

3/4.6.1.6 PRIMARY CONTAINMENT AVERAGE AIR TEMPERATURE The limitation in containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 300*F during LOCA conditions and is consistent with the accident analyses.

1 BRUNSWICK - UNIT 1 B 3/4 6-2a Amendment No. I )

I J

' ADMINISTRATIVE CONTROLS PROCEDURES PROGRAMS. AND MANUALS (Continued) I p -1.

Preventive maintenance and periodic visual inspection requirements.-and 2.

Integrated cycle leak or intervals testless. requirements for each system at refueling 6.8.3.2 In Plant Radiation Monitoring ~

l l

L A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1. Training of personnel.

L

2. Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

6.8.3.3 Post-Accident Sampling i

A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and gaseous effluents, and containment atmos)particulates in plant here samples under accident conditions. The program shall include t1e following:

_ 1. Training of personnel.

2. Procedures for sampling and analysis, and

3. .

Provisions for maintenance of sampling and analysis equipment.

6.8.3.4 Primary Containment Leakage Rate Testing Program Deson 15 A program si all be established to implement the leakage rate testing of the cont 11nment as required by 10 CFR 50.54(o) and 10.CFR 50.

AppendixJ.,dancewiththeas be in accor modified by approved exemptions. This pro '

Guide 1.163. " Performance guidelines contained in Regulatory Based Containment Leak-Test Program." dated September 1995 as modified by the following exceptions:

1. Compensation of instrument inaccuracies applied to the containment leakage total per ANSI /ANS 56.8-1987 instdad of ANSI /ANS 56.8-1994. .

2. - N SERT ------

The peak calculated containment internal pressure for the design basis loss of coolant accident. P, is 49 psig.

The maximum allowable primary containment leakage rate. L,. shall be 0.5% of primary containment air weight per day at P .

BRI.lNSWICK - UNIT 1 6-17 Amendment No. l

• o

)

i.

TECHNICAL SPECIFICATION 6.8.3.4 INSERT:

. 2. Following air lock door seat replacement, performance of door seal leakage rate

, testing with the gap between the door seals pressurized to 10 psig instead of air j tock testing at P, as specified in Nuclear Energy Institute Guideline 94-01,  !

Revision O.

h a

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t .-

ENCLOSURE 4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 SUPPLEMENT TO REQUESTS FOR LICENSE AMENDMENTS CONTAINMENT LEAKAGE RATE TESTING l

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l MARK-UPS OF TYPED TECHNICAL SPECIFICATION PAGES - UNIT 2 i

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l-l l

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CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each primary containment air lock chall be demonstrated OPERABLE:

a. By verifying the seal leakage rate to be less than or caual to 5 scf when the gap between the door seals is pressurized to 10 l

1. Within 7 days following each closing, except when the air lock is being used for multiple entries, then at least once per 30 days, and

2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air lock has been used and no maintenance has been performed on the l air lock, and l

3. When the air lock seal has been replaced.

b. By conducting an overall air lock leakage test at P., 49 psig and by verifying that the overall air, lock leakage is within its limit:

1. At least once per 30 months, and I l

, 2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when maintenance (except for seal replacement) has been performed on the air lock that would affect the air lock sealing capability.

c. By verification of air lock interlock OPERABILITY:

1. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air lock has been used, and

2. prior to and following a drywell entry when PRIMARY CONTAINMENT INTEGRITY is required. and

3. Following the performance of maintenance affecting the air lock interlock.

mxemptinn tn Appendiv j cf 10 CFn gg, l

BRUNSWICK - UNIT 2 3/4 6-5 Amendment No. I

3/4.6 CONTAINMENT SYSTEMS i

BASES 3/4.6.1 PRIMARY CONTAINMENT l

3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures 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.

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The safety design basis for the primary containment is that .. c.st withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

The DBA that postula[es the maximum release of radioactive material within l primary containment is a LOCA. In analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to l the environment is controlled by the rate of primary containment leakage. l Analytical methods and assumptions involving the primary containment are presented in References 6 and 7.

The maximum allowable leakage rate for the primary containment (L.) is 0.5 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the maximum peak containment pressure (P,) of 49 psig.

A Primary Containment Leakage Rate Testing Program has been established in i

accordance with 10 CFP,50.54(o) to implement the requirements of 10 CFR l Part 50. Appendix J. Option B (Reference 1). The Primary Containment Leakage Rate Testing Program conforms with NRC Regulatory Guide 1.163. eevisier' St.-

dated September 1995. " Performance-Based Containment Leak, Rate Testing Program" (Reference 2) and Nuclear Energy Institute (NEI) 94-01. Revision 0. ,

l dated July 26.1995. " Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J" (Reference 3) with the exception of:

1 1.

NEI 94-01. Section 8.0. " Testing Methodologies for Type A. B and C Tests" states that " Type A. Type B and Type C tests should be performed using the technical methods and techniques specified in ANSI /ANS 56.8-1994, or other alternative testing methods that have been approved by the NRC." The Brunswick Plant takes exception to ANSI 56.8 flowmeter accuracy requirements based upon compensation of instrument inaccuracies applied to the containment leakage total per the previous revision of the standard.

Brunswick Plant administrative procedures and databases already effectively address instrument error. Brunswick Plant uses standard glass tube and ball type flowmeters with a 5 percent of full scale accuracy.

Readings are compensated for back pressure, temperature, and test medium variables. To overcome the less accurate flowmeter use, an equipment error is applied to the results of each test. The square root of the sum of the squares of the equiament errors for the tests is also added to the cumulative containment leacage total. This method is cc:.sistent with ANSI N56.8-1987 Appendix E and provides conservative assurance that the BRUNSWICK - UNIT 2 B 3/4 6-1 Amendment No.

I

e

• 3/4.6 CONTAfNMENT SYSTEMS BASES 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE (Continued) i cumulative containment leakage total accounts for instrument inaccuracy.  !

No such instrument error analysis or accounting is required per  :

ANSI /ANS 56.8-1994.

2 - INSERT -

The leakage rate acceptance criteria of s 0.60 L, for the combined Type B and C tests and s 0.75 L for the Type A test ensures a primary containment configuration, including, equipment hatches. that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

Primary containment operability is maintained by limiting leakage to s 1.0 La.

Individual leakage rates specified for the primary containment air lock are addressed in Specification 3.6.1.3. ,

03erating experience with the main steam line isolation valves has indicated t1at degradation has occasionally occurred in the leak tightness of the valves: therefore. th~e special requirement for testing these valves.

Exemptions from the requirements of 10 CFR Part 50 have been granted-for the main steam isolation valve leak testing and leakage calculations. I NRC Regulatory Guide 1.163. @evi:1= Reference 2) endorses NEI 94-01 (Reference 3) which in turn identifies ANSI /ANS 56.8-1994. " Containment System Leakage Testing Requirements" (Reference 4) as an acceptable standard regarding leakage-rate test methods, procedures, and analyses. Reduced duration Type A tests may be performed using the criteria and Total Time Method specified in Bechtel Topical Report BN-TOP-1. Revision 1. November 1.

1972 (References 5 and 6).

References:

1. 10 CFR Part 50. Appendix J, op% L

2. NRC Regulatory Guicie 1.163. EF/ici= ted September 1995.

" Performance-Based Containment Leak-Rate Testing Program."

3. Nuclear Energy Institute Guideline 94-01. Revision 0 dated July 26. 1995.

" Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J."

4. ANSI /ANS 56.8-1994. " Containment System Leakage Testing Requirements"

5. CP&L Letter to Mr. D. B. Vassallo. " Integrated Leak Rate Test."

October 20. 1983.

6. NRC Letter from Mr. D. B. Vassallo to Mr. E. E. Utley. December 9. 1983.

7. Updated FSAR. Section 6.2.

8. Updated FSAR. Section 15.6.4.

BRUNSWICK - UNIT 2 8 3/4 6-la Amendment No. I

BASES INSERT:

2. NEl 94-01, Section 10.2.2.2, " Repairs or Adjustments of Airlocks" states that following maintenance on an air lock pressure retaining boundary, one of the following . tests shall be completea:

a. The air lock shall be tested at a pressure of not less than P , or

b. Leakage rate testing at P, shall be performed on the affected area or component.

A previously approved exemption to 10 CFR 50, Appendix J that allows the performance of air lock door seal leakage rate testing at a pressure less than P, following door seal replacement instead of air lock testing at P, has been retained and is listed as an exception in Technical Specification 6.8.3.4.

l 1

l

~ '

CONTAINMENT SYSTEMS 1

BASES j 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS

' i The primary containment air lock forms part of the primary containment 1

)

pressure boundary. As such, air lock integrity and leak tightness are. j i essential for maintaining primary containment leakage rate to within limits in '

the event of a DBA. Not maintaining air lock integrity or leak tightness may  !

result in a leakage rate in excess of that assumed in unit safety analysis. I

' The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In analysis of this accident it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage. In i the analysis of this accident. it is assumed that primary containment is  !

OPERABLE, such that release of fission products to the environment is I controlled by the rate of primary containment leakage. The primary i containment is designed with a maximum allowable leakage rate (L,) of 0.5 percent by weight of the containment air Jer 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the maximum peak containment pressure (P ) of 49 psig. This a'lowable leakage rate forms the basis for the acceptance criteria imposed on the surveillance requirements associated with the air lock. 1 The ]rimary containment air lock is required to be OPERABLE. For the air lock to ]e considered OPERABLE. the air lock interlock mechanism must be OPERABLE. the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross  !

breach of primary containment does not exist when primary containment is required to be OPERABLE. Closure of a single door in each air lock is sufficient to 3rovide a leak tight barrier following postulated events.

Nevertheless. ]cth doors are kept closed when the air lock is not being used I

for normal entry and exit from primary containment.

Maintaining primary containment air locks OPERABLE requires compliance l with the leakage rate test requirements of 10 CFR 50. Appendix J as l established in the Primary Containment Leakage Rate Testing Program. The Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50. Appendix J. Option B (Reference 1). The Primary Containment Leakage Rate Testing Program conforms with NRC Regulatory Guide 1.163. O vi:1= OF '

dated September 1995. " Performance-Based Containment Leak-Rate Testing Program" and Nuclear Energy Institute (NEI) 94-01. Revision 0, dated July 26, 1995. " Industry Guideline for Implementing _Perf rmance-Based Option of V

10 CFR 50 Appendix J' as modified by ena exceptions [(References 2and3).

4.he. hsW W Spedhh G.8. sA.

An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA.

Only one closed door in each air lock is required to maintain the ,

integrity of the containment. In the event of an inoperable door interlock, locking shut the inner door will ensure containment integrity while permitting access to the lock for maintenance and surveillance testing.

BRUNSWICK - UNIT 2 B 3/4 6-2 Amendment No. I

l l

CONTAINMENT SYSTEMS '

BASES l 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS (Continued)

References:

1. 10 CFR Part 50. Appendix J., Opwn B. i

2. NRC Regulatory Guide 1.163. Esuis ted September 1995.

" Performance-Based Containment Leak-Rate Testing Program." l

3. Nuclear Energy Institute Guideline 94-01. Revision 0, dated July 26, 1995.

" Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J."

3/4.6.1.4 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the primary containment steel vessel will be maintained comparable to the eriginal design standards for the life of the facility. Structural integrity is required to ,

ensure that the vessel will with' stand the maximum pressure of 49 psig in the I event of a LOCA. A visual inspection in conjunction with the Primary Containment Leakage Rate Testing Program is sufficient to demonstrate this capability.

References:

1. 10 CFR Part 50. Appendix J. Option B.Section III.A.

2. NRC Regulatory Guide 1.163. Eevicic ated September 1995.

~ Performance-Based Containment Leak-Rate Testing Program."

1 3/4 6.1.5 PRIMARY CONTAINMENT INTERNAL PRESSURE The limitations of. primary containment internal pressure ensure that the containment Jeak pressure of 49 psig does not exceed the design pressure of 62 psig during _0CA conditions. The limit of 1.75 psig, for initial positive l containment pressure will limit the total pressure to 49 psig which is less I than the design pressure and is consistent with the accident analyses.

3/4.6.1.6 PRIMARY CONTAINMENT AVERAGE AIR TEMPERATURE l

The limitation in containment average air temperature ensures that the l containment peak air temperature does not exceed the design temperature of 300*F during LOCA conditions and is consistent with the accident analyses. ,

I i

l l

l BRUNSWICK - UNIT 2 B 3/4 6-2a Amendment No. l

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

ADMINISTRATIVE CONTROLS

. PROCEDURES. PROGRAMS. AND MANUALS (Continued) l 1.

. Preventive maintenance and periodic visual inspection requirements and.

2.

a Integrated cycle leak or intervals testless.

requirements for each system at refueling 6.8.3.2 In-Plant Radiation Monitoring l

A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the followinc.

1. Training of persmnel.

2. . Procedures fo: monitoring. and 3.

Provision', for maintenance of sampling and analysis equipment.

6.8.3.3 Post-Accidert Sampling i

A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and 3 articulates in plant gaseous efflucnts and containment atmos)1ere samples under accident-conditions. The program shall include tie following:

1. Training of personnel.

2. Procedures for sampling and analysis, and

3. .

. Provisions for maintenance of sampling and analysis equipment.

6.8.3.4 Primary Containment Leakage Rate ~ Testing Program r Op h B A program slfall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50.

AppendixJ.9ancewiththeas be in accor This program modified by approved exemptions. shall Guide 1.163. " Performance guidelines contained in Regulatory Based Containment Leak-Test Program." dated September 1995 as modified by the following exceptions:

1. Compensation of instrument inaccuracies applied to the containment . leakage total per ANSI /ANS 56.8-1987 instdad of ANSI /ANS 56.8-1994. -

2. wserr-The peak calculated containment internal pressure for the design

. basis loss of coolant accident. P,. is 49 psig.

The maximum allowable primary cor.tainment leakage rate. L,. shall be 0.5% of primary containment air weight per day at P .

BR!aSWICK - UNIT 2 6-17 Amendment No. I

TECHNICAL SPECIFICATION 6.8.3.4 INSERT:

2. Following air lock door seal replacement, performance of door seal leakage rate testing with the gap between the door seals pressurized to 10 psig instead of air lock testing at P, as specified in Nuclear Energy Institute Guideline 94-01, Revision 0.

1

d a

, ENCLOSURES BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 i NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 i SUPPLEMENT TO REQUESTS FOR LICENSE AMENDMENTS i

CONTAINMENT LEAKAGE RATE TESTING I

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TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1

't INDEX

, ADMINISTRATIVE CONTROLS SECTION PAGE 6.5 REVIEW AND AUDIT (Continued) 6.5.4 NUCLEAR ASSESSMENT SECTION INDEPENDENT REVIEW PROGRAM Function........ ... ...... .... ..... ...... .... . .. 6-10 Organization.... . ...... ............ . ... ..... 6-10 4

Review.. . ... . ..... . . ............. ..... .. .. . 6-11 l l

Records.. .... ...... . .............. ....... . ...... 6-12 6.5.5 NUCLEAR ASSESSMENT SECTION ASSESSMENT PROGRAM.... .... 6-13 l

6.S.6 0UTSIDE AGENCY INSPECTION AND AUDIT PROGRAM... ..... 6-15 l l

6.6 REPORTABLE EVENT ACTION. ..... .... ....... .. ........ 6-15 l 6.7 SAFETY LIMIT VIOLATION... .... .. ... . .... ....... . 6-15 l

6.8 PROCEDURES. PROGRAMS. AND MANUALS. . .... .... . .... . 6-16 l l l

6.9 REPORTING RE0VIREMENTS Routine Reports...... .... .................. ....... . 6-17a Startup Reports.... ...... . ....... ... .. .. ..... 6-17a Annual Reports..... .... ..... ... .............. . ..... 6-18  ;

1 Personnel Exposure and Monitoring Report. . . . . . . . . . .. 6-18 Annual Radiological Environmental Operating Report..... . 6-19 Semiannual Radioactive Effluent Release Report. . . . . . . . . 6-20 Monthly Operating Reports.. . .. . ..... .. . . 6-21 Special Reports. . ... . ... . ... ... . 6-22 Core Operating Limits Report.. ....... .. ... .. ... 6-22 BRUNSWICK - UNIT 1 XV Amendment No. I

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• CONTAINMENT SYSTEMS

. PRIMARY CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:

a. An overall integrated leakage rate of:

1. Less than or equal to L- 0 containment air per 24 b,our.5 s at P percent

, 49 psig.by weight of the I

2. Deleted. I

b. A combined leakage rate of less than or equal to 0.60 L for penetrationsandvalvessubjecttoTypeBandCtestswhen 3ressurized to P, in accordance with the Primary Containment Leakage late Testing Program described in S main steam line isolation valves *. pecification 6.8.3.4. except for

c. *Less than or equal to 11.5 scf per hour for any one main steam line isolation valve when tested at 25 psig.

APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1.

ACTION:

With:

a. The measured overall integrated primary containment leakage rate exceeding 0.75 L , or I 4

b. The measured combined leakage rate for penetrations and valves subject to Type B and C tests in accordance with the Primary Containment Leakage Rate Testing Program, except for main steam line isolation valves *, exceeding 0.60 L , or

c. The measured leakage rate exceeding 11.5 scf per hour for any one main steam line isolation valve, restore:

a. The overall integrated leakage rate (s) to less than or equal to 0.75 L,. and I

b. The combined leakage rate for penetrations and valves subject to Type B and C tests in accordance with the Primary Containment Leakage Rate Testing Program. except for main steam line isolation valves *, to less than or equal to 0.60 L, and Exemption to Appendix "J" of 10 CFR 50.

i BRUNSWICK - UNIT 1 3/4 6-2 Amendment No. l

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' CONTAINMENT SYSTEMS

LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued) l

c. The leakage rate to less than or equal to 11.5 scf per hour for any one main steam line isolation valve.

prior to increasing reactor coolant system temperature above 212 F.

SURVEILLANCE REQUIREMENTS l

l l 4.6.1.2.1 Perform required primary containment leakage rate testing in I I

accordance with the Primary Containment Leakage Rate Testing Program described in Specification 6.8.3.4.

4.6.1.2.2 Main steam line isolation valves shall be leak tested at least once i I per 18 months.

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(Pages 3/4 6-3A and 3/4 6-3B have been deleted.)

BRUNSWICK - UNIT 1 3/4 6-3 Amendment No. I

l* CONTAINMENT SYSTEMS

SURVEILLANCE REQUIREMENTS u 1 1

f 4.6.1.3 Each primary containment air lock shall be demonstrated OPERABLE:

a. By verifying the seal leakage rate to be less than or equal to 5 scf per hour when the gap between the door seals is pressurized to 10-psig: I

1. Within 7 days following each closing, except when the air lock l- 1s being used for multiple entries, then at least once per 30 days, and

2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air lock has been used and no maintenance has been performed on the air lock, and

3. When the air lock seal has been replaced.

b. By conducting an overall air lock leakage test at P,. 49 psig, and by verifying that the overall air lock leakage is within its limit:

1. At least once per 30 months, and l

2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when maintenance (except for seal replacement) has been performed on the air lock that would affect the air lock sealing capability. I

c. By verification of air lock interlock OPERABILITY:

1. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air l lock has been used, and

2. prior to and following a drywell entry when PRIMARY CONTAINMENT INTEGRITY is required, and 4

3. Foilowing the performance of maintenance affecting the air lock interlock.

I BRUNSWICK - UNIT 1 3/4 6-5 Amendment No. l

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' CONTAINMENT SYSTEMS l PRIMARY CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.4 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.4.

APPLICABILITY: OPERATIONAL CONDITIONS 1. 2, and 3.

ACTION:

With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 212 F.

l SURVEILLANCE REQUIREMENTS l 1

4.6.1.4.1 The structural integrity of the exposed accessible interior and exterior surfaces of the primary containment, including the liner plate, shall be determined during the shutdown for each Type A containment leakage rate test by visual inspection of those surfaces. This inspection shall be performed prior to the Type A containment leakage rate test and during two other refueling outages before the next Type A test if the interval for the Type A test has been extended to 10 years, to verify no apparent changes in appearance or other abnormal degradation.

4.6.1.4.2 Reoorts Any abnormal degradation of the primary containment structure detected during the above required inspections shall be reported to i the Commission pursuant to Specification 6.9.2. This Special Report shall '

include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.

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1 BRUNSWICK - UNIT 1 3/4 6-6 Amendment No. I

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'3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures 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.

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The safety design basis for the primary containment is that it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary containment are presented in References 6 and 7.

The maximum allowable leakage rate for the primary containment (L.) is 0.5 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the maximum peak containment pressure (P,) of 49 psig.

A Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50, Appendix J Option B (Reference 1). The Primary Containment Leakage Rate Testing Program conforms with NRC Regulatory Guide 1.163, dated September 1995, " Performance-Based Containment Leak-Rate Testing Program" (Reference 2) and Nuclear Energy Institute (NEI) 94-01, Revision 0, dated July 26,1995. " Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J" (Reference 3) with the exception of:

1. NEI 94-01 Section 8.0, " Testing Methodologies for Type A, B and C Tests" states that " Type A Type B and Type C tests should be performed using the technical methods and techniques specified in ANSI /ANS 56.8-1994, or other alternative testing methods that have been approved by the NRC." The Brunswick Plant takes exception to ANSI 56.8 flowmeter accuracy requirements based upon compensation of instrument inaccuracies applied to the containment leakage total per the previous revision of the standard.

Brunswick Plant administrative procedures and databases already effectively address instrument error. Brunswick Plant uses standard glass tube and ball type flowmeters with a 5 percent of full scale accuracy.

Readings are compensated for back pressure, temperature, and test medium variables. To overcome the less accurate flowmeter use, an equipment error is applied to the results of each test. The square root of the sum of the squares of the equi) ment errors for the tests is also added to the cumulative containment leacage total. This method is consistent with ANSI 56.8-1987 Appendix E and provides conservative assurance that the BRUNSWICK - UNIT 1 B 3/4 6-1 Amendment No. l l

' 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE (Continued) cumulative containment leakage total accounts for instrument inaccuracy.

No such instrument error analysis or accounting is reauired per ANSI /ANS 56.8-1994 I

2. NEI 94-01. Section 10.2.2.2. " Repairs or Adjustments of Airlocks" states that following maintenance on an air lock pressure retaining boundary, one of the following tests shall be completed:

a. The air lock shall be tested at a pressure of not less than P,. or

-b. Leakage rate testing at P, shall be performed on the affected area or component.

A previously approved exemption to 10 CFR 50. Appendix J that allows the performance of air lock door seal leakage rate testing at a pressure less than P following door seal replacement instead of air lock testing at P, has be,en retained and is listed as an exception in Technical Specification 6.8.3.4.

The leakage rate acceptance criteria of s 0.60 L, for the combined Type B and C tests and s 0.75 L for the Type A test ensures a primary containment configuration. including, equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

Primary containment operability is maintained by limiting leakage to s 1.0 L,.

Individual leakage rates specified for the primary containment air lock are addressed in Specification 3.6.1.3.

0]erating experience with the main steam line isolation valves has indicated tlat degradation has occasionally occurred in the leak tightness of the valves; therefore. the special requirement for testing these valves. j Exemptions from the requirements of 10 CFR Part 50 have been granted for the 1 main steam isolation valve leak testing and leakage calculations. 1

--NRC Regulatory Guide l.163. (Reference 2) endorses NEI 94-01 (Reference 3) ,

which in turn identifies ANSI /ANS 56.8-1994, " Containment System Leakage  !

Testing Requirements" (Reference 4) as an acceptable standard regarding leakage-rate test methods, procedures, and analyses. Reduced duration Type A tests may be performed using the criteria and Total Time Method specified in l Bechtel Topical Report BN-TOP-1. Revision 1. November 1, 1972 (References 5 l and 6).

References:

1. 10 CFR Part 50. Appendix J. Option B.

2. NRC Regulatory Guide 1.163. dated September 1995. " Performance-Based Containment Leak-Rate Testing Program."

BRUNSWICK - UNIT 1 B 3/4 6-la Amendment No. l

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3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE (Continued)

3. Nuclear Energy Institute Guideline 94-01. Revision 0, dated July 26, 1995.

" Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J." l 4

4. ANSI /ANS 56.8-1994, " Containment System Leakage Testing Requirements". 1

5. CP&L Letter to Mr. D. B. Vassallo, " Integrated Leak Rate Test,"

October 20, 1983.

]

6. NRC Letter from Mr. D. B. Vassallo to Mr. E. E. Utley. December 9. 1983.

7. Updated FSAR. Section 6.2.

8. Updated FSAR. Section 15.6.4.

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l BRUNSWICK - UNIT 1 B 3/4 6-lb Amendment No. I i

' CONTAINMENT SYSTEMS BASES 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The primary containment air lock forms part of the primary containment pressure boundary. As such. air lock integrity and leak tightness are I essential for maintaining primary containment leakage rate to within limits in the event of a DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in unit safety analysis. l l

The DBA that postulates the maximum release of radioactive material within i primary containment is a LOCA. In analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage. In the analysis of this accident, it is assumed that primary containment is OPERABLE. such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L.) of ,

0.5 percent by weight of the containment air 3er 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the maximum peak l containment pressure (P ) of 49 psig. This a'10wable leakage rate forms the I basis for the acceptance criteria imposed on the surveillance requirements '

associated with the air lock.

The primary containment air lock is required to be OPERABLE. For the air lock to be considered OPERABLE. the air lock interlock mechanism must be OPERABLE.

the air lock must be in compliance with the Type B air lock leakage test, and 1 both air lock doors must be OPERABLE. The interlock allows only one air lock l door to be opened at a time. This provision ensures that a gross breach of l primary containment does not exist when primary containment is required to be l OPERABLE. Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry and exit from primary containment.

Maintaining primary containment air locks OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50. Appendix J as established in the Primary Containment Leakage Rate Testing Program. The Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50. Appendix J.

Option B (Reference 1). The Primary Containment Leakage Rate Testing Program conforms with NRC Regulatory Guide 1.163. dated September 1995, " Performance-Based Containment Leak-Rate Testing Program" and Nuclear Energy Institute (NEI) 94-01. Revision 0, dated July 26. 1995. " Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J" as modified by the exceptions listed in Specification 6.8.3.4 (References 2 and 3).

An inoperable air lock door does not invalidate the 3revious successful performance of the overall air lock leakage test. T11s is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA.

Only one closed door in each air lock is required to maintain the integrity of the containment. In the event of an inoperable door interlock. locking shut the inner door will ensure containment integrity while permitting access to the lock for maintenance and surveillance testing.

BRUNSWICK - UNIT 1 B 3/4 6-2 Amendment No. I

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' CONTAINMENT SYSTEMS-BASES j 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS (Continued)

References- I l '. 10 CFR Part 50, Appendix J. Option B.

2. NRC Regulatory Guide 1.163, dated September 1995. " Performance-Based Containment Leak-Rate Testing Program."  !

3. Nuclear Energy Institute Guideline 94-01 Revision 0, dated Jul" 26, 1995.

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" Industry Guideline for Implementing Performance-Based Option of 10 CFR 50

-Appendix J."-

3/4.6.1.4 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the primary containment steel vessel will be maintained comparable to the original. design standards for the life of the facility. Structural integrity is required to -

ensure that the vessel will withstand the maximum pressure of 49 psig in the event of a LOCA. A visual inspection in conjunction with the Primary Containment Leakage Rate Testing Program is sufficient to demonstrate this capability.

References:

1. 10 CFR Part 50, Appendix J. Option B.Section III.A.

2. NRC Regulatory Guide 1.163. dated September 1595. " Performance-Based Containment Leak-Rate Testing Program."

3/4.6.1.5 PRIMARY CONTAINMENT INTERNAL PRESSURE The limitations of primary containment internal pressure ensure that the l containment Jeak pressure of 49 psig does not exceed the design pressure of 62 i psig during _0CA conditions. The limit of 1.75 psig, for ini+.ial positive  ;

containment pressure will limit the total pressure to 49 psig, which is less  !

than the design pressure and is consistent with the accident analyses. j 3/4.6.1.6 ' PRIMARY CONTAINMENT AVERAGE AIR TEMPERATURE The limitation in containment average air temperature ensures that the ,

containment peak air temperature does not exceed the design temperature of 300*F during LOCA conditions and is consistent with the accident analyses.

BRUNSWICK - UNIT 1 B 3/4 6-2a Amendment No. l

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• ADMINISTRATIVE CONTROLS 6.8 PROCEDURES. PROGRAMS. AND MANUALS l I 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

4 a. The applicable procedures recommended in Appendix "A" of Regulatory j Guide 1.33. November 1972.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g. OFFSITE DOSE CALCULATION MANUAL implementation.

h. PROCESS CONTROL PROGRAM implementation.

i. Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 1.21. Revision 1. June 1974, and Regulatory Guide 4.1. Revision 1. April 1975.

6.8.2 Temporary changes to procedures of Specification 6.8.1 above, any other 3rocedures that affect nuclear safety, and proposed tests or experiments may 3e made provided:

a. The intent of the original procedure, proposed test or experiment is not altered,

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator License on the unit affected. i

c. The change is documented, reviewed pursuant to Specifications 6.5.2.1 and 6.5.2.2 and approved by the General Manager - Brunsw1ck Plant or his previously designated alternate within 14 days of implementation.

6.8.3 Proarams and Manuals l The following programs-shall be established, implemented, and )

maintained:

6.8.3.1 Primary Coolant Sources Outside Containment l l

A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The program shall include the following:

BRUNSWICK - UNIT 1 6-16 Amendment No. I

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• ADMINISTRATIVE CONTROLS PROCEDURESL PROGRAMS. AND MANUALS (Continued) 1.  ;

1. Preventive maintenance and periodic visual inspection requirements, and 2 Integrated leak test requirements for each system at refu'eling l cycle intervals or less.

6.8.3.2 In-Plant Radiation Monitoring l A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following: )

1. Training'of personnel. l

2. Procedures for monitoring, and  !

3. Provisions for maintenance of sampling and analysis equipment.

6.8.3.3 Post-Accident Sampling l A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and ) articulates in plant gaseous effluents. and containment atmosnere samples under accident conditions. The program shall include t1e following:

1. Training of personnel.

2. Procedures for sampling and analysis, and

3. Provisions for maintenance of sampling and analysis equipment.

6.8.3.4 Primary Containment Leakage Rate Testing Program I

A program shall be established to implement the leakage rate testing  !

of the containment as required by 10 CFR 50.54(o) and 10 CFR 50. l Appendix J. Option B.as modified by a) proved exemptions. This 3rogram shall be in accordance with tie guidelines contained in Regulatory Guide 1.163. " Performance-Based Containment Leak-Test '

Program." dated September 1995 as modified by the following exceptions:  !

1. Compensation of instrument inaccuracies applied to the containment leakage total per ANSI /ANS 56.8-1987 instead of ANSI /ANS 56.8-1994.

2. Following air lock door seal replacement, performance of door '

seal leakage rate testing with the gap between the door seals pressurized to 10 psig instead of air lock testing at P, as specified in Nuclear Energy Institute Guideline 94-01.

Revision 0.

BRUNSWICK - UNIT 1 6-17 Amendment No. l

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• ADMINXSTRATIVE CONTROLS 4

i PROCEDURES.' PROGRAMS. AND MANUALS (Continued) l The peak calculated containment internal pressure for the design basis loss of coolant accident. P., is 49 psig.

The maximum allowable primary containment leakage rate, L., shall be 0.5% of primary containment air weight per day at P,.

6.9 REPORTING REOUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office unless otherwise noted.

STARTUP REPORTS 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license. (2) amendment to the license. involving a planned increase in power level. (3) installation of fuel that has a different design or has been manufactured by a different fuel

' supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

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i BRUNSWICK - UNIT 1 6-17a Amendment No. I l

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ENCLOSURE 6 l BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 SUPPLEMENT TO REQUESTS FOR LICENSE AMENDMENTS CONTAINMENT LEAKAGE RATE TESTING  !

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.5 REVIEW AND AUDIT (Continued) 6.5.4 NUCLEAR ASSESSMENT SECTION INDEPENDENT REVIEW PROGRAM Function.. ....... ... . ... . .. . . ....... .. 6-10 Organization.... . .. .. .... . . ....... . 6-10 Review........... ...... .. . .. .......... .. . ... .. 6-11 Records.. ......... ... .. . .. .. ... ........ ... 6-12 1

6.5.5 NUCLEAR ASSESSMENT SECTION ASSESSMENT PROGRAM.. .. . .... 6-13 i l

6.5.6 0UTSIDE AGENCY INSPECTION AND AUDIT PROGRAM.. . . . 6-15 6.6 REPORTABLE EVENT ACTION.. . . . .... .. . 6-15 6.7 SAFETY LIMIT VIOLATION. .. .... ..... .. .. . . .. . 6-15 6.8 PROCEDURES. PROGRAMS. AND MANUALS. . . . ... ... 6-16 l 6.9 REPORTING REQUIREMENTS Routine Reports.... . .. .... . ... ....... ... 6-17a Startup Reports. ...... . . ... . ....... .. .. 6-17a Annual Reports. ... ....... . .. .. . ..... .. .. .. . 6-18 l l

Personnel Exposure and Monitoring Report.. . .. ...... 6-18 l Annual Radiological Environmental Operating Report.. . . 6-19 Semiannual Radioactive Effluent Release Report. . . 6-20 Monthly Operating Reports.. . .. .. .. . ... 6-21 Special Reports. . . .. .. . .. . 6-22 Core Operating Limits Report.... .. . .... . . .. 6-22 BRUNSWICK - UNIT 2 XV Amendment No. I

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' CONTAINMENT SYSTEMS 1

PRIMARY CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:

a. An overall integrated leakage rate of:

1. Less than or equal to L , 0.5 percent by weight of the containment air per 24 flours at P, 49 psig. I

2. Deleted. I i

b. A combined leakage rate of less than or equal to 0.60 L for penetrationsandvalvessubjecttoTypeBandCtestswfien pressurized to P, in accordance with the Primary Containment Leakage Rate Testing Program described in Specification 6.8.3.4. except for main steam line isolation valves *.

c. *Less than or equal to 11.5'scf per hour for any one main steam line isolation valve when tested at 25 psig.

APPLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per  !

Specification 3.6.1.1. l ACTION:

With:

a. The measured overall integrated primary containment leakage rate exceeding 0.75 L, or l l

b. The measured combined leakage rate for penetrations and valves I subject to Type B and C tests in accordance with the Primary l Containment Leakage Rate Testing Program, except for main steam line isolation valves *, exceeding 0.60 L, or

c. The measured leakage rate exceeding 11.5 scf per hour for any one main steam line isolation valve, restore:

a. The overall integrated leakage rate (s) to less than or equal to 0.75 L,. and I

b. The combined leakage rate for penetrations and valves subject to Type B and C tests in accordance with the Primary Containment Leakage Rate Testing Program, except for main steam line isolation valves *. to less than or equal to 0.60 L., and Exemption to Appendix "J" of 10 CFR 50.

BRUNSWICK - UNIT 2 3/4 6-2 Amendment No. I

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' CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) l ACTION (Continued)

c. The leakage rate to less than or equal to 11.5 scf per hour for any one main steam line isolation valve, prior to increasing reactor coolant system temperature above 212 F.

SURVEILLANCE REQUIREMENTS 4.6.1.2.1 Perform required primary containment leakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program described in Specification 6.8.3.4.

4.6.1.2.2 Main steam line isolation valves shall be leak tested at least once I per 18 months.

(Pages 3/4 6-3A has been deleted.)

BRUNSWICK - UNIT 2 3/4 6-3 Amendment No. I

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' CONTAINMENT SYSTEMS-l SURVEILLANCE REQUIREMENTS 4

! 4.6.1.3 Each primary containment air lock shall be demonstrated OPERABLE:

i

a. By verifying the seal leakage rate to be less than or equal to 5 scf j .per hour when the gap between the door seals is pressurized to 10 i psig: I

1. Within 7 days following each closing, except when the air lock

is being used for multiple entries, then at least once per 30 days, and i 2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air

. lock has been used and no maintenance has been performed on the air lock, and l

3. When the air lock seal has been replaced.

$ b. By conducting an overall air lock leakage test at P,. 49 psig, and by verifying that the overall air lock leakage is within its limit:

i j 1. At least once per 30 months, and i

2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when maintenance (except for seal replacement) has been performed on 4

the air lock that would affect the air lock sealing capability. I

! c. By verification of air lock interlock OPERABILITY:

1. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when the air

! lock has been used, and i

1 2. prior to and following a drywell entry when PRIMARY CONTAINMENT INTEGRITY is required, and l

j 3. Following the performance of maintenance affecting the air lock interlock.

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BRUNSWICK - UNIT 2 3/4 6-5 Amendment No. 1

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CONTAINMENT SYSTEMS l

PRIMARY CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.4 The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.4.

APPLICABILITY: OPERATIONAL CONDITIONS 1. 2. and 3. l l

ACTION. j With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 212 F.

1 SURVEILLANCE REQUIREMENTS 4.6.1.4.1 The structural integrity of the exposed accessible interior and exterior surfaces of the primary containment including the liner plate shall be determined during the shutdown for each Type A containment leakage rate test by visual inspection of those surfaces. This inspection shall be performed prior to the Type A containment leakage rate test and during two other refueling outages before the next Ty Type A test has been extended to 10 years,pe A test to verify if the interval no apparent for the changes in appearance or other abnormal degradation.

4.6.1.4.2 Reoorts Any abnormal degradation of the primary containment structure detected during the above required inspections shall be reported to the Commission pursuant to Specification 6.9.2. This Special Report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.

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i BRUNSWICK - UNIT 2 3/4 6-6 Amendment No. I

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

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3/4.6 CONTAINMENT SYSTEMS BASES j 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY l Primary CONTAINMENT INTEGRITY ensures 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 4 accident conditions. )

, i 3/4.6.1.2 PRIMARY CONTAINMENT LEAXAGE )

The safety design basis for the primary containment is that it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate. l The DBA that postulates the maximum release of radioactive material within

-primary containment is a LOCA. In analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.

Analytical methods and assumptions involving the primary containment are presented in References 6 and 7.

The maximum allowable leakage rate for the primary containment (L.) is 0.5 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the maximum peak containment pressure (P,) of 49 psig.

A Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50. Appendix J. Option B (Reference 1). The Primary Containment Leakage Rate Testing Program conforms with NRC Regulatory Guide 1.163. dated September 1995. " Performance-Based Containment Leak-Rate Testing Program" (Reference 2) and Nuclear Energy Institute (NEI) 94-01. Revision 0, dated July 26.1995. " Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J~ (Reference 3) with the exception of:

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1. NEI 94-01. Section 8.0, " Testing Methodologies for Type A. 8 and C Tests" l states that " Type A. Type B and Type C tests should be performed using the technical methods and techniques specified in ANSI /ANS 56.8-1994, or other alternative testing methods that have been approved by the NRC." The Brunswick Plant takes exception to ANSI 56.8 flowmeter accuracy requirements based upon compensation of instrument inaccuracies applied to the containment leakage total per the previous revision of the standard.

Brunswick Plant administrative procedures and databases already  ;

. effectively address instrument error. Brunswick Plant uses standard glass  ;

tube and ball type flowmeters with a 5 percent of full scale accuracy. I Readings are compensated for back pressure, temperature and test medium variables. To overcome the less accurate flowmeter use, an equipment ,

error is applied to the results of each test. The square root of the sum l of the squares of the equi) ment errors for the tests is also added to the cumulative containment leacage total. This method is consistent with ANSI N56.8-1987 Appendix E and provides conservative assurance that the BRUNSWICK - UNIT 2 B 3/4 6-1 Amendment No.

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• 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE (Continued) cumulative containment leakage total accounts for instrument inaccuracy.

No such instrument error analysis or accounting is required per ANSI /ANS 56.8-1994.

2. NEI 94-01. Section 10.2.2.2. " Repairs or Adjustments of Airlocks" states that following maintenance on an air lock pressure retaining boundary, one of the following tests shall be completed:

a. The air lock shall be tested at a pressure of not less than P , or

b. Leakage rate testing at P, shall be performed 'on the affected area or component.

! A previously approved exemption to 10 CFR 50. Appendix J that allows the performance of air lock door seal leakage rate testing at a pressure less than P, following door seal replacement instead of air lock testing at P, has been retained and is listed as an exception in Technical Specification 6.8.3.4.

The leakage rate acceptance criteria of s 0.60 L, for the combined Type B and C tests and s 0.75 L, for the Type A test ensures a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

Primary containment operability is maintained by limiting leakage to s 1.0 L,.

Individual leakage rates specified for the primary containment air lock are addressed in Specification 3.6.1.3.

0]erating experience with the main steam line isolation valves has indicated tlat degradation has occasionally occurred in the leak tightness of the valves; therefore, the special requirement for testing these valves.

Exemptions from the requirements of 10 CFR Part 50 have been granted for the main steam isolation valve leak testing and leakage calculations.

NRC Regulatory Guide 1.163 (Reference 2) endorses NEI 94-01 (Reference 3) I which in turn identifies ANSI /ANS 56.8-1994. " Containment System Leakage Testing Requirements" (Reference 4) as an acceptable standard regarding leakage-rate test methods, procedures, and analyses. Reduced duration Type A tests may be performed using the criteria and Total Time Method specified in Bechtel Topical Report BN-TOP-1. Revision 1. November 1. 1972 (References 5 and 6).

References:

1. 10 CFR Part 50. Appendix J. Option B.

2. NRC Regulatory Guide 1.163. dated September 1995. " Performance-Based Containment Leak-Rate Testing Program."

BRUNSWICK - UNIT 2 B 3/4 6-la Amendment No. I

• 3/4.6 CONTAINMENT SYSTEMS I

BASES 3/4.6.1 PRIMARY CONTAINMENT ,

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE (Continued)

3. Nuclear Energy Institute Guideline 94-01. Revision 0, dated July 26, 1995.

" Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J."

4. ANSI /ANS 56.8-1994. " Containment System Leakage Testing Requirements".

5. CP&L Letter to Mr. D. B. Vassallo. " Integrated Leak Rate Test."  ;

October 20, 1983.

6. NRC Letter from Mr. D. B. Vassallo to Mr. E. E. Utley, December 9, 1983.

7. Updated FSAR. Section 6.2. l l

8. Updated FSAR. Section 15.6.4.

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BRUNSWICK - UNIT 2 8 3/4 6-lb Amendment No. I

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' CONTAINMENT SYSTEMS BASES 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The primary containment air lock forms part of the 3rimary containment i pressure boundary. As such, air lock integrity and leac tightness are  !

essential for maintaining primary containment leakage rate to within limits in l the event of a DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in unit safety analysis.

The DBA that postulates the maximum release of radioactive material within l primary containment is a LOCA. In analysis of this accident, it is assumed l that primary containment is OPERABLE such that release of fission products to '

the environment is controlled by the rate of primary containment leakage. In i the analysis of this accident, it is assumed that primary containment is '

OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L ) of 0.5 percent by weight of the containment air aer 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the maximum peak containment pressure (P,) of 49 psig. .This a~ 10wable leakage rate forms the l basis for the acceptance criteria imposed on the surveillance requirements associated with the air lock. l The 3rimary containment air lock is required to be OPERABLE. For the air lock to 3e considered OPERABLE. the air lock interlock mechanism must be OPERABLE. the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE. Closure of a single door in each air lock is sufficient to 3rovide a leak tight barrier following postulated events.

Nevertheless. Joth doors are kept closed when the air lock is not being used

.for normal entry and exit from primary containment.

Maintaining primary containment air locks OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50. Appendix J as established in the Primary Containment Leakage Rate Testing Program. The

. Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50. Appendix J. Option B (Reference 1). The Primary Containment Leakage Rate Testing Program conforms with NRC Regulatory Guide 1.163. dated September 1995. " Performance-Based Containment Leak-Rate Testing Program" and Nuclear Energy Institute (NEI) 94-01 Revision 0, dated July 26. 1995.

" Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 A)pendix J" as modified by the exceptions listed in Specification 6.8.3.4 (References 2 and 3).

An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock. door is capable of providing a fission product barrier in the event of a DBA.

Only one closed door in each air lock is required to maintain the integrity of the containment. In the event of an inoperable door interlock, locking shut the inner door will ensure containment integrity while permitting access to the lock for maintenance and surveillance testing.

-BRUNSWICK - UNIT 2 B 3/4 6-2 Amendment No. I

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• CONTAINMENT SYSTEMS BASES 1

3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS (Continued) l

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References:

1. 10 CFR Part 50. Appendix J. Option B.

2. NRC Regulatory Guide 1.163, dated September 1995. " Performance-Based Containment Leak-Rate Testing Program." l

3. Institute Guideline 94-01. Revision 0. dated July 26. 1995. l

' NuclearEnerbelineforImplementingPerformance-BasedOptionof10CFR50

" Industry Gu Appendix J." l l

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4 3/4.6.1.4 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY l 1

This limitation ensures that the structural integrity of the primary containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 49 psig in the event of a LOCA. A visual inspection in conjunction with the Primary Containment Leakage Rate Testing Program is sufficient to demonstrate this capability.

References:

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1. 10 CFR Part 50. Appendix J. Option B.Section III.A.

2. NRC Regulatory Guide 1.163, dated September 1995. " Performance-Based Containment Leak-Rate Testing Program.

3/4.6.1.5 PRIMARY CONTAINMENT INTERNAL PRESSURE i

The limitations of primary containment internal pressure ensure that the containment Jeak pressure of 49 s does not exceed the design pressure of 62

c psig during _0CA conditions. Th mit of 1.75 psig, for initial positive containment pressure will limit the total pressure to 49 psig. which is less than the design pressure and is consistent with the accident analyses.

i 3/4.6.1.6 PRIMARY CONTAINMENT AVERAGE AIR TEMPERATURE The limitation in containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 300*F during LOCA conditions and is consistent with the accident analyses. -

BRUNSWICK - UNIT 2 B 3/4 6-2a Amendment No. I

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' ADMINISTRATIVE CONTROLS 6.8 PROCEDURES. PROGRAMS. AND MANUALS l 4

6.8.1. Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33. November 1972.

b. Refueling operations.

Surveillance and test activities of safety related equipment.

c.

d. Security Plan implementation.

-e. Emergency Plan implementation.

f. Fire Protection Program implementation.

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g. 0FFSITE DOSE CALCULATION MANUAL implementation. ,

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h. PROCESS CONTROL PROGRAM implementation. l

1. Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 1.21. Revision 1. June 1974, and Regulatory Guide 4.1. Revision 1. April 1975.

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6.8.2 Temporary changes to procedures of Specification 6.8.1 above. any other 1 3rocedures that affect nuclear safety, and proposed tests or experiments may 3e made provided:

a. The intent of the original procedure, proposed test or experiment is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator License on the unit affected.

! c. The change is documented, reviewed pursuant to Specifications 6.5.2.1 and 6.5.2.2 and approved by the General Manager - Brunswick Plant or his previously designated alternate within 14 days of implementation.

6.8.3 Programs and Manuals l The following programs shall be established, implemented, and

. maintained:

6.8.3.1 Primary Coolant Sources Outside Containment I A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The program shall include the following:

BRUNSWICK - UNIT 2 6-16 Amendment No. I

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" ADMINISTRATIVE CONTROLS PROCEDURES. PROGRAMS. AND MANUALS (Continued) l 1-. Preventive maintenance and periodic visual inspection requirements. and

2. Integrated leak test requirements for each system at refueling cycle intervals or less.

6.8.3.2 In-Plant Radiation Monitoring l A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1. Training of personnel.

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2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equipment.

6.8.3.3 Post-Accident Sampling l A program which will' ensure the capability to obtain and analyze reactor coolant, radioactive iodines and 3 articulates in plant.

gaseous effluents, and containment atmos 31ere samples under accident conditions. The program shall include tie following:

1. Training of personnel.

2. Procedures for sampling and analysis, and

3. Provisions for maintenance of sampling and analysis equipment.

6.8.3.4 Primary Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50.

Appendix J. Option 8 as modified by a) proved exemptions. This

3rogram shall be in accordance with tie guidelines contained in Regulatory Guide 1.163. " Performance-Based Containment Leak-Test Program." dated September 1995 as modified by the following exceptions

1. Compensation of instrument inaccuracies applied to the containment leakage total per ANSI /ANS 56.8-1987 instead of

.l ANSI /ANS 56.8-1994.

2. Following air lock door seal replacement, performance of door seal leakage rate testing with the gap between the door seals pressurized to 10 psig instead of air lock testing at P, as specified in Nuclear Energy Institute Guideline 94-01.

Revision 0.

< BRUNSWICK - UNIT 2 6-17 Amendment No. I

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' ADMINISTRATIVE CONTROLS PROCEDURES. PROGRAMS. AND MANUALS (Continued) l The peak calculated containment internal pressure for the design basis loss of coolant accident. P,. is 49 psig.

The maximum allowable primary containment leakage rate. L., shall be 0.5% of primary containment air weight per day at P,.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10. Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office unless otherwise noted.

STARTUP REPORTS 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license. (2) amendment to the license involving a planned increase in power level. (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

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I BRUNSWICK - UNIT 2 6-17a Amendment No. I