Text

Proposed Tech Specs Re Containment Leakage Rate Testing
	ML20094Q259
Person / Time
Site:	Brunswick
Issue date:	11/27/1995
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20094Q255	List: BSEP-95-0605, Amends 950913 License Amend Application Re Containment Leakage Rate Testing,Based on Issuance of Final Rule for Primary Reactor Containment Leakage Testing for water-cooled Power Reactors on 950926 & Discussions W/Nrc; ML20094Q259;
References
	NUDOCS 9512010108
	Download: ML20094Q259 (60)
	v • d • e

i ENCLOSURE 5 5

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 MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 1 p

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INDEX

-ADMIN.ISTRATIVE' CONTROLS- +

LSECTION PA_.GE

'6. 5 i REVIEW AND AUDIT'(Continued)

.6.5.4 NUCLEAR ASSESSMENT SECTION INDEPENDENT REVIEW PROGRAM 'l Function........ ........... ...... ............ .... ... 6-10 Organization....... ...................-... .............. 6-10 Re v i ew . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . 6 Records.......... ..... .. .................. ..... ...... 6-12

6.5.5 NUCLEAR ASSESSMENT SECTION ASSESSMENT PROGRAM............. 6-13 1

6;5.6 0UTSIDE AGENCY INSPECTION AND AUDIT PROGRAM............ .. 6-15 6.6 R E PORTABL E EVENT ACTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-15 6,7 SAFETY LIMIT VIOLATION............... ..... ................. 6-15 6.8 PROCEDURES M e PROGRAMS. @ ,f.Y.u g .. .. , ............... 6-16 rp A 6: 9 REPORTING REQUIREMENTS Routine Reports. .. .............. ..................... 6-17a.

Startup Reports...... ........ . .. . . ....... .... 6-17q, Annual Reports.. ..... ..... ..... .. ......... ........ 6-18 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 4

.Special Reports........... ... ..... . . ....... ........ 6-22 Core Operating Limits Report. . . . . .. . ...... ...... 6-22 BRUNSWICK - UNIT 1 XV Amendment No. 4JJ.

i subject tot yp e B and C tests when pressurized to P. (n accordance wu.h the Primary i

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. CONTAINHENT SYSTEMS Con (ainment l ta e Rate TesMn3 P ropm l PRIMARY CONTAINMENT LEAKAG dtScribed in $pecificaticn b13.k 3

LIMITING CONDITION FOR OPE RATION l l

3.6.1.2 Primary containme leakage rates shall be limited to:

a. An overall integrat leakage rate of:

1. Less than or equa to L , 0.5 percent by weight of the containment air pe 24 $ours at *P , 49 psig g O Delabad.

2. L::: :h:2 q 1 :: i , 0 ??"' ;;=::: 17 =ist: cf ti:

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b. A combined leakage rate of .ess than or equal to 0.60 L,for eu.

1';: d i; T Li. 2. ',. 1, except for ein penetrations and +44r valvesgy Ase-steam line isolation valves *g erbj :* t "fre

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

APPLICABILITY: , W hen PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1.

ACTION:

Subject hIype nod C MGks in OCCordance Wikh P

the Primaryconunment Leaka3e % Tee $ rgram.,

y1g,,

a. The measured l lverall integrated primary containment leakage rate exceeding O. 5Lg: 0. 7 5 gL , .: ;;;1i =il:, or 7

b. The me ured combined leakage race f or -eM. penetrations and eM.

valves li:n ! in T il; 2.',. 1, except for main steam line isolation t valves *, cri. ::: :: ?;;: " crf C n :: exceeding 0.60 L,, or

c. The measured eakage race exceeding 11.5 scf per hour for any one

' main steam lin isolation valve, restore. :o

a. The overall integrated leakage rate (s) to less than or equal 0.75 L, :r 0 'S Lg , = ;;;1i = il:, and

b. The combined leakage rate for eH- penetracions and *M- valves He+ed-i; Teil; 2.5.? i, except for main steam line isolation valves ,

and

ij :: n ?fp; " _;d C n::: to less than or equal to 0.60 L,, l

Exemption to Appendix "J" of 10 CFR 50.

I BRUNSWICK - UNIT 1 3/4 6-2 A=endment No. H-l

. \

1 CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

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

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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. 1J,1 ,

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 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*:

1. Within 72 7das h:;y:: following each closing, except when the air lock is being used for multiple entries, then at least once per 44k 4weev, and 30 kys Prior to establishing PRIMARY CONTAINMENT INTEGRITY when'the air 2;

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 with,in its limit:

1. At least once per :! ---t' g, and 30 mordhs

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.

Exemption to Appendix J of 10 CFR 50.

Tr., r.. 1;i;r.; ef "recificatier. i.0.2 ;;; r.:: ;;;11;:ble.

BRUNSWICK - UNIT 1 3/4 6-5 Amendment No. 44WL

. CONTAINMENT SYSTEMS PRIMARY CONTAINMENT STRUCTURAL INTECRITY LIMITING CONDITION FOR OPERATION ;

i 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. ;

SURVEILLANCE REQUIREMENTS 2 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 a visual inspection of those surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no

apparent changes in appearance or other abnormal degradati .

4.6.1.4.2 Reports Any abnormal degradat rimary containment structure detected during ' . required inspections shall be reported to i the Commission purs to Specification 6.9.2. This Special Report shall include a des ; tion of the condition of the concrete, the inspection e tolerances on cracking, and the corrective actions taken.

procedure e Otnd durin two other tv,Eue,lin outo, es before the next Type A kest if the interval rkhe ge A kest has been exte.nded to Ib year 3 M

d 4

}

BRUNSWICK - UNIT 1 3/4 6-6 Amendment No. s+-

3/4.6 CONTAINMENT SYSTEMS -

)

i' BASES l

3/4.6.1 PRIMARY CONTAINMENT !

3/4.6.1.1 PRIMARY CONTAINMENT INTECRITY )

l Primary CONTAINMENT INTECRITY 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 1erkage 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 CONTAINM'ENT LEAKAGE

~

The ions.on primary containment Leakage rates ens t the total p containment leakag me will not exceed the val med in the accident j analyses at the peak acci e saure of ig, P . As an added w conservatism, the measured over ted leakag,e rate is further limited v to less than or equal . L, or 0.75 L , licable, during performance of the perio ststoaccountforpossibledegraa f the containm

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Operati'ng experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of c the valves; therefore, the special requirement for testing these valves.

> Exemptions from the requirements of 10 CFR Part 50 have been granted for 1 mainsteamisolationvalveleaktesting/tertir-;cfci'^9: c h u c;.:P -

-:;:: % >and leakage calculation methods.

[ App 1x J, paragraph III.A.3 requires that all Type A (Contai nt fIntegrated k Rate) tests be conducted in accordance with Ame 'can National Standard (ANSI 45.4-1972, "Leakag'e Rate Testing of Contai nt Structures for Nuclear Reacto " March 16, 1972. In addition to t Total Time and N Point-to-Point methods scribed in that standard, th ass Point method, when i

% used with a test duration at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , i n acceptable method to use F to calculate leakage rates. pical descrip ' n of the Mass Point method 6 lcan be found in ANSI /ANS 56.8-19 "Contai ent System Leakage Testing CD Requirements," January 20, 1987. Re duration Type A tests may be E performed using the criteria and Tot T1 method specified in Bechtel

g. Topical Report BN-TOP-1, Revisio , Novembe , 1972 (References 1 and 2).

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'I

References:

w g 1. CP&L Letter to . D. B. Vassallo, " Integrated Leak Rat est,"

.4 October 20, ' 3.

n.

.w u- 2. NRC L er from Mr. D. B. Vassallo to Mr. E. E. Utley, December 9, 83.

BRUNSWICK - UNIT 1 B 3/4 6-1 Amendment No. -t%

INSERT #1:

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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. Revision 0, 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 flowneters with a 5 percent of full scale accuracy. Readings are compensated for back pressure, temperature, and test medium variables. To overcome the less accurate flowneter 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 cumulative containment leakage total

}

-INSERT #1: (Continued) accounts for instrument inaccuracy. No such instrument error analysis or acccJnting is required per ANSI /ANS 56.8-1994.

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.

< a INSERT #2:

NRC Regulatory Guide 1.163. Revision 0 (Reference 2) endorses NEI 94-01 (Reference 3) which in turn identifies ANSI /ANS 56.8-1994, " Containment System i 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.

2. NRC Regulatory Guide 1.163. Revision 0, 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."

4. ANSI /ANS 56,8-1994, " Containment System Leakage Testing Requirements".

5. CP&L Letter to Mr. D. B. Vassallo, " Integrated Leak Rate Test," l October 20, 1983. l l

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

7. Updated FSAR, Section 6.2.

8. Updated FSAR, Section 15.6.4.

l

1 CONTAINMENT SYSTEMS ~

t 3ASES ,

D9 3/4.6.1.3 PRIMARY CONTAINPCiT AIR LCCX3 F* In ttattons on closure and lese rate for the containment si '. As ars re;utred to meu. *he restrictions on PR* MARY CONTAINMENT INT .6t and leak Oh jj rate given in 3pecitt _ 'ons 3.6.i.1 and 3.6.1.2. Eh ecification makes

.ailowances for the fact that - re may be 'on . .tods of time when the air

.ccxs will be in a closed and secur o;o1 during reactor operation. Only f, maintain tne integrity of the tre closed door in eacn air loc requte 3 In the av - st an inoperable door -lock, locking shut cne

.: ntainmene.. 7 i

- a .e containment integrity unite permt - access to the gg ky.nnerdoorwil*

s gck for - . anance and surveillance testing.

CC J/4.6.1.4 PRIMARY CONTAINMENT STRUCTURAL INTECRITY This limitation ensures that the, structural integrity of the primary containment steel vessel will be maintained comparable to the original design I standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand tne maximum pressure of 49 psig in the event of a LOCA. A visual inspection in conjunction withYL!;72 * ' " ; > -= -*

is sufficient to demonstrate this capability.

--9> INSERT #+ Se Prhn tunWnment Leok Ate.TesBn Pro ram) 3/4.6.1.5 PRIMARY CONTAINMENT INT AL PRESSURE I

The limitations of primary containment internal pressure ensure that the ,

containment peak ,,ressure of 49 psig does not exceed the design pressure of 62 l 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.

l 3/4.5.1.6 PRIMARY CONTAIRMENT AVERACE AIR TEMPERATURE

\

I 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.

3R'3S'4 ICV - UNI T ! B 3/4 6-2 Ate ndien t No. L'*ke - ,' - :-

INSERT #3:

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 that 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 controllad 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 per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the maximum peak containment pressure (P.) of 49 psig. This allowable leakage rate forms the basis for the acceptance criteria imposed on the surveillance requirements 4 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 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 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, Revision 0, 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 approved exceptions (References 2 and 3).

o .

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.

References:

1. 10 CFR Part 50. Appendix J.

2. NRC Regulatory Guide 1.163. Revision 0, 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."

r 1

I I

INSERT #4:

References:

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

2. NRC Regulatory Guide 1.163, Revision 0, dated September 1995,

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

(

ADMINISTRATIVE CONTROLS .

6.8 PROCEDURES _ m PROCRAMS d6

't 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,

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

c-f. Fire Protection Program implementation.

g. OFFSITE DOSE CALCULATION MANUAL implementations' ,

h. PROCESS CONTROL PROGRAM implementation.

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

6.8.2 Temporary changes to procedures of Specification 6.8.1 above, any other procedures that affect nuclear safety, and proposed tests or experiments may be 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 previousi designated alternate within 14 days of implementation.

6.B.2 lbrams and danuals_

dbHbrF The"following programs shall be established, i.,plemented, and maintained: .

l 6.t.3.1 f

-6. Primary Coolant Sources Outside Containment I A program to reduce leakage from those portions of systems outside i 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.Whh>

.o ADMINISTRATIVE CONTROLS .

, AND M AeduAl.S PROCEDURES,AN9 PROGRAMS (Continued) 9 A

1. Preventive maintenance and periodic visual inspection requirements, and

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

b.S.3. 2. In-Plant Radiation Monitoring Jh.

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,

2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equipment, 6.6.3.3

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

1. Training of personnel,

2. Procedures for sampling and analysis, and

3. Provisions for maintenance of sampling and analysis equipment.

~

W lP4GERT #.T 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.

\

BRUNSWICK - UNIT 1 6-17 Amendment No . -4&&-

l 0

l INSERT #5:

6.8.3.4 Primary Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate I testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, as modified by approved exemptions. This program shall be in accordance with the 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. ,

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,.

.g ' 4

~

ENCLOSURE 6 -

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 MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 2 i i

t t

f T

i 9

i l

l I

t

INDEX ADMINISTRATIVE CONTROLS SECTION !

_P._Afd 6.5 REVIEW AND AUDIT (Continued) 6.5.4 NUCLEAR ASSESSMENT SECTION INDEPENDENT REVIEW PROGRAM l Function............................... ................. 6-10 >

Organization............... ................ ........... 6-10 Review................................................... 6-11 Records......... ........ .... ..................... .... 6-12 i

6.5.5 NUCLEAR ASSESSMENT SECTION ASSESSMENT PROGRAM......... . 6-13 l 6.5.6 0UTSIDE AGENCY INSPECTION AND AUDIT PROGRAM.......... ..

6-15 '

66 REPORTABLE EVENT ACTION............... . . ...... ..... .. 6-15 i 6.7 SAFETY LIMIT VIOLATION. ...................... ......... . 6-15 ,

6.8 PROCEDURESAG PROGRAMS. At4D MA>4UALS

............. ........ ..... . ..... 6-16 9 b ,

6.9 REPORTING REOUIREMENTS i Routine Reports.......... . ................ ............ 6-17a.

Startup Reports. .......... . ..... .............. ...... 6-17a, Annual Reports.. ..... .................................. 6-18 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 i Core Operating Limits Report..... ... .............. .. 6-22 Y

1 BRUNSWICK - UNIT 2 XV Amendment No. 200-

-__ . ,- .e. -

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

..- _. . - . . _ _ 1 1

4 i

i me*1!, and C tests uben f CONTAllGENT SYST_ EMS Su%ect k.o Ty '

~

Eressurized to 'a in acct,rdcxnc.e with hhe.

PRIMARY CONTAINMENT LEAEAGE l

Primary Conte.6 ment Ltaka3e Ro.ka ying '

l, LIMITING CONDITION FOR OPER TION Procgram dese.ribed in Specifitdon 6.%.1.4 3

~

1

,i j 3.6.1.2 Primary containmen' leakage rates shall be limited to: :

j a. An overall integrat leakage rate oft 1

3 1. 14se than or eq to L , 0.5 percent by weight o the containment air 24 $ours at P,, 49 poi I / Deleted.

j 2. 4 in ; e - :: ;; 21 . Q . 0.357 ; ::: 57 ::1# : :f ;'

: 2 f-- :: ci ;r

. :: :: : ::d:::f en::::: Of ?g, j .s y..g.

i ,

i

b. A combined leakage rate of oess than or equal to 0.60 L for e n j penetrations and e&& valvesg li:=d i T.L1.1.0.0 ^ , ex, cept f or main ,

' steam line isolation valves *[dj::: : *;;: 5 :-f C :::t: d:

- er tri: :d :: r,, e ,.:..

I

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

1 i

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

l Q

* "' subject to yp T e 3 ancl C lesh in accordance wie P c

the Primory ontainment Leako3* Rate TesHn3 rogram, l

With:

! a. The maast ed overall 1stegrated primary containment leakage rate j exceeding :: 0.75 ig , ; ;ppii.;;;;, or 0.75L,k i

j b. The me d combined leakage rate for 4&& penetrations and -e&4 - '

l T:il: 2.5.2 1, except for main steam line isolation i!

valves valves *, __.bdi:

'::t :: *;;: ! crf C ::::: exceeding 0.60 L,, or !j

! c. The measured le ate exceeding 11.5 scf per hour for any one main steam line isolac alve, restore:

. a. The overall integrated leakage race (s) to less than or eq 1 to 0.75 L :: 0_'5 E g, ;; ppli;;il:, and

. *0

b. The combined leakage rate for.e4&, penetrations and en valvesj 46esed.

ie T il: 2.5.?-1, except for main steam line isolation valvesd, j ::ij::: :: *;;; 5 - ' C t::t to less than or equal to 0.60 L,, and i

t

Exemption to Appendix "J" of 10 CFR 50.

BRUNSWICK - UNIT 2 3/4 6-2 Amendment No. 91 1

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) .

1-_

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

.I vu-4.6.1.2g -_4_, . ,,m ,4n___. ,-,c,,,_ _,.--

asaw yi a uru s J wvisuussuaNaiw avunuyw awwww sa su s i uw uvsuvissus wwww u,,, u ;_____._,.ma 4 sai s--m-Asnom i ,4 + b +ba - ,, L m A . 1 m ,mA ,. 4&- : - ---2t2J .- in erm en a---J2.. 1 uwwur vuI sw w su i vii wrsw wwiavuuIwn ussu wi 5 6wIe au Jyww4 5---Jwu 6ea AV wi I\ sv. nyyUIlu i A U.

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J 2- 1^ een en AnnanA4v 1 rr--' '-

ng etg, Perform reptred primary containment leakay L m 1 AmJ rate testin in accordance via the Primar Conkoinment Les e Rahe Testin Program

c. Dc'eted, described in Speci ication 6.t.2.

a ,

+me.+..,, w,,, sm _- m n. a . . - + na . ,4. +. s ,+ o

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

n 7

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,,+m.,. . ...- s. .e. .,, x,,,,.,,.,.i,,,+ma..-4,,,,7

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1 l

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

BRUNSWICK - UNIT 2 3/4 6-3 Amendment No. GO+

1

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CONTAINMENT SYSTEMS

SURVEILLANCE REQUIREMENTS 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*:

1. Within ?Y dej;sh_ : following each closing, except when the air lock

- is being used for multiple entries, then at lease once per 72 h::::,.and 30 days

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 that at P,, 49 psig, and by verifying that the overall air lock leakage is within its limit:

. 30rnonths

h: g, nd

1. At least once per ci: a 4 ll

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

c. By verification of air lock interlock OPERABILITY; l

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.

I i

C

Exemption of Appendix J of 10 CFR 50.

6 ___,..__c,_

.t-- r - -----.

.r. .___sz,__

-r----------

.__ >n . - . __.

-- -- -- rr-------

BRUNSWICK - UNIT 2 3/4 6-5 Amendment No. k&4 a -. -- __________________l

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

d i

?

i CONTAINMENT SYSTEMS

~

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

i ACTION:

! With the structural integrity of the primary containment not conforming to the

~

above requirements, restore the structural integrity to withih the limits j

prior to increasing the Reactor Coolant System temperature above 212*F.

4

SURVEILLANCE REQUIREMENTS ,

i j 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 a visual inspection of those surfaces. This inspection shall be

' performed prior to the Type A containment leakage rate testYto verify no . ._ c s '

j apparent changes in appearance or other abnormal degradation .

{ 4.6.1.4.2 Reports Any abnormal degradation of the pr M containment structure detected durin th- stru m uiruu inspectiuns shall be reported '

to the Commission pur to Specification 6.9.2. This Special Report shall l include a des ption of the condition of the concrete, the inspection procedure, e tolerances on cracking, and the corrective actions taken.

t j and during Wo other refue\ing outages before the next Ty pe A ted if the interval b the '

T yp A lesh has been edended to 10 years.,

i 4

i BRUNSWICK - UNIT 2 3/4 6-6 Amendment No. 444-4 a

i i i ?

}*, i t

3/4.6' CONTAINMENT SYSTEMS- '

BASES '

~

i 4 i 3/4.6.1 PRIMARY CONTAINMENT i 3/4.6.1.1 PRIMARY CONTAINMENT INTECRITY f ;

} Primary CONTAINMENT INTEGRITY ensures that the release of radioactive e

materials from the' containment atmosphere will be restricted to those leakage i paths and associated leak rates assumed in the accident analyses. This i

% restriction, in conjunction with the leakage rate limitation, will limit the a b- site boundary radiation doses to within the limits of 10 CFR Part 100 during i accident conditions. '

u 4

7 3/4.6.1.2 PRIMARY CONTAINMENT LEAKACE.

t .

f V l

2

'mitations on primary containment leakage rates ensure that th ntainment volume will not exceed the value assumed

accident

)

kJ U

analyses at the peak acci essure of 49 psi s an added j conservatism, the measured overall t Leakag,.e rate is further limited i

to less than or equal e , or 0.75 L , as

ble, during performance !

W of the per*

stoaccountforpossibledegradationo ntainment Of- l e arriers between leakage tests. '

/ Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of j the valves; therefore, the special requiremenc for testing these valves.

1 Exemptions from the requirements of 0 CFR Part 50 have been granted for main steam isolation valve leak testing i:: tin; cf :!:12:E: cfter :: h j ;--:-gandleakagecalculationmethods.

I i Q Appendix J, paragraph III.A.J requi.res that all Type A (Containment 1 l

4

% Intes ed Leak Rate) tests be conducted in accordance with American Natt p Standard SI) N45.4-1972, " Leakage Rate Testing of Containment Str res i i

et for Nuclear R tors," March 16, 1972. In addition to the Total

and

d Point-to-Point me s described in that standard, the Mass nt method, when 2 used with a test dura '

~

of at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , is an a ptable method to use to calculate leakage rate . A typical description the Mass Point method f can be found in ANSI /ANS 56.8- 7, "Contai System Leakage Testing 4

Requirements," January 20, 1987. uced ration Type A tests may be 2 performed using the criteria and Tot me method specified in Bechtel w Topical Report BN-TOP-1, Revisio , Novem 1, 1972 (References 1 and 2).

. u i-4 l g3 References 4

w OC L 1. CP&L Lett o Mr. D. B. Vassallo, " Integrated Leak Rat est,"

Octo 0, 1983.

. ( NRC Letter from Mr. D. B. Vassallo to Mr. E. E. Utley, December 9, 3 e

INSERT #1:

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 containmerit 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 Revision 0, 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 cumulative containment leakage total

INSERT #1: (Continued) accounts for instrument inaccuracy. No such instrument error analysis or accounting is required per ANSI /ANS 56.8-1994.

The leakage rate acceptance criteria of s 0.60 L, for the conbined 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.

INSERT #2:

NRC Regulatory Guide 1.163. Revision 0 (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.

2. NRC Regulatory Guide 1.163, Revision 0, 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."

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.

CONTAINMENT SYSTEMS 3ASES M

%g :/5.s.l.3 ?RIMARY ~0NTADeEMT AIR LCCTS

,E, .

tatta::cns on :.:sure inc iean rate for :he :entain. ment si ..sts ar_

m re:atred to .... :ne res: rte: ions an ?RIMAAY ::0NTAIUMENT :PT"" ...Y ana less 2 race given in 3peett'... ' -ns :.5.1.1 ana :.6.1.2. *

. 5:ec;ft:sti:n makes y ail:wances for :ne fac: :nat .. may :e '- periocs )f time snen :ne air F-

.oces will be in a closes anc secu- .s -

";on ,suring rea::cr ::erstion. Oniv

te closec :oce in esca a . rain :he intezeity af the-

][

.ven is recutres to ,

,, :entainment. :n "' _ vent of an inoperaoie door interio 'acKing snut :ne o inner do , ensure containment integri:y unile permitting ac: . -

the 6 (J _ :or maintenance and surveillanca -><- , -

n.

w oc 3/t.6.1.4 PRIMARY CONTAINMENT STRUCTURAL "NTECRITY This limitation ensures enat :ne structurai integrity of :ne primary

cntainment steel vessel will be T.aintaaned :omparaole to the :: ginal design stancards for the life af the facility. 3truc: ural integrity :s required to ensure that the vessel will witnstano the maximum pressure of '9 psig in the event of a LOCA. A visuat inspection in conjunction with Tjp; '

i;;;;g; :c;;;

is sufficient to demonstrate this capability.

P 9 gg g .o..

a g 3 he frimary Dontainment Leakage Rale.TesHn3 PRIMARY CONTAINMENT INTERNAL ?RESSURE rogrard The limitations of primary containment internal pressure ensure that the containment peak pressure of 49 psig does not exceed the design pressure of 62 I psig during LOCA conditions. The limit of 1.75 psig, for initial positive containment pressure will limit the total pressure to 49 psig, vnich is less chan the design pressure and is consistent with the accident analyses.

3/4.6.1.6 PRIMARY CONTAINMENT AVERACE AIR TEMPERATURE -

The limi:ation 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.

BRUNS *a'ICK - UNIT 2 5 3 'c 5-2 2mendment No. 4++"

I

INSERT #3:

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 that 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 containment is designed with a maximum allowable leakage rate (L.) of 0.5 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the maximum peak containment pressure (P,,) of 49 psig. This allowable 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 must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, anu 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 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(0) 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, Revision 0, 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 approved exceptions (References 2 and 3).

. l l

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 containnent. 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.

References:

1. 10 CFR Part 50, Appendix J.

t

2. NRC Regulatory Guide 1.163 Revision 0, dated September 1995,

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

3. Nuclear Energy Institute Guideline 94-01, Revision 0, dated July 26, 4 1995, " Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J."

i l

I

INSERT #4:

References:

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

2. NRC Regulatory Guide 1.163. Revision 0, dated Septender 1995.

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

1 I

I

1 .'

s ADMINISTRATIVE CONTROLS 6.8 PROCEDURESAND.PROGRAMSgAND MANUAL.S 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 ,

l Cuide 1.33, November 1972.

b. Refueling operations.

i

c. Surveillance and test activities of safety related equipment. i i

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 ef fluent 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 procedures that affect nuclear safety, and proposed tests or experiments may be 3 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 4 and 6.5.2.2 and approved by the Ceneral Manager - Brunswies. Plant or his previously designated alternate within 14 days of implementation.

G.B.3 Proerams and MonuoAs 4,4ri- The Tollowing programs shall be established, implemented, and maintained: t 6.8.5.l

.a Primary Coolant Sources Outside Containment 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. W i

J '

f. * .

r o

I '

2 ,

ADMINISTRATIVE CONTROLS

_ AND MANUALS .

PROCEDURES,AND PROCRAMS (Continued) ;

'i h i

- 1. Preventive maintenance and periodic visual inspection !

requirements, and

+

~ 2. Integrated leak test requirements for each system at refueling j s cycle intervals or less. _j

48.3.2.

-h. .In-Plant Radiation Monitoring 1 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:

I

1. Training of personnel, !

i d.

2. Procedures for monitoring, and !

3. Provisions for maintenance of sampling and analysis equipment.

}' G.S.S.$

c Post-Accident Sampling i ,

A program which will ensure the capability to obtain and analyze

' reactor coolant, radioactive iodi, pes and particulates in plant gaseous l effluents, and containment atmosphere samples under accident ;

j conditions. The program shall include the followings j i

. 1. Training of personnel, l 2. Procedures for sampling and analysis, and

[ 3. Provisions for maintenance of sampling and analysis equipment.

! 4 INSE.RT l 6.9 REPORTING REQUIREMENTS )

i l ROUTINE REPORTS l 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.

i STARTUP REPORTS

6.9.1.1 A susmary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to i che 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 j i nuclear, thermal, or hydraulic performance of the plant. i

+

l

BRUNSWICK - UNIT 2 6-17 Amendment' No. +P9-

4 C INSERT #5:

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, as modified by approved exemptions. This program shall Le in accordance with the 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.

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

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

i ENCLOSURE 7 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 TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1 L

c ,

1NDfl ,

j ADMINISTRATIVE CONTROLS SECTION P.eff !

1 f6.5 - REVIEW AND AUDIT (Continued). ;

6.5.4 NUCLEAR ASSESSMENT SECTION INDEPENDENT REVIEW PROGRAM l Function................................................. 6-10 !

Organization............ ................................ 6-10 -!

6 '

Review................................................... !

t Records..... .................................... ....... 6-12 6.5.5 NUCLEAR ASSESSMENT.SECTION ASSESSMENT PROGRAM............ 6-13 :

E6.5.6 0UTSIDE AGENCY INSPECTION AND AUDIT PROGRAM........... 6 !

,6.6 REPORTABLE EVENT ACTION........ .......................... .6-15 !

s ,

6.7 - SAFETY LIMIT VIOLATION.......... ........................ 6 :

6.8 PROCEDURES. PROGRAMS. AND MANUALS........................ 6 l 6.9 REPORTING REQUIREMENTS Routine Reports.................... . ................... 6-17a Startup Reports......................... .......... ..... 6-17a Annual Reports.......... .......... ......... ........... 6-18 Personnel Exposure and Monitoring Report. . . . . . . . . . . . . . 6-18 Annual Radiological Environmental Operating Report. ... . 6-19 i.

Semiannual Radioactive Effluent Release Report. . . . . . . . . 6-20 t

i Monthly Operating Reports..................... . . . 6-21 r~

Special Reports.......................... ............. . 6-22 Core Operating Limits Report....... ..................... 6-22 ;

r 3

i i l

' BRUNSWICK - UNIT 1 - XV Amendment No. l

l l

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 , 0.5 percent by weight of the containment air per 24 flours at P,. 49 psig. I

2. Deleted. I ,

b. A combined leakage rate of less than or equal to 0.60 L for penetrationsandvalvessubjecttoTypeBandCtestswfien 3ressurized to P in accordance with the Primary Containment Leakage late Testing Pro,g ram 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.

ACTION:

With:

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

BRUNSWICK - IINIT 1 3/4'6-2 Amendment No. I b- 1

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

.l CONTAINMENT SYSTEMS i l

LJMITING CONDITION FOR OPERATION (Continued) ;

i ACTION (Continued) J

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

I prior to increasing reactor coolant system temperature above 212 F. i

) l c SURVEILLANCE REQUIREMENTS !

l 1 . . . .. .. . ..

1 i 4.6.1.2.1 Perform required primary containment leakage rate testing 'in .. l accordance with the Primary Containment Leakage Rate Testing Program described j

in Soecification 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. j i !

i >

i i

i I-i i

i !

1 2

i

! I i- '

L t-I i

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

. BRUNSWICK - UNIT 1 3/4 6-3 Amendment No. l m, - . - - - . , , , . - - , , . ~ .

- .,-w - _ .-

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS r-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*:

'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 air lock and

3. When thL .r 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

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.

Exemption to Appendix J of 10 CFR 50.

I BRUNSWICK - IINIT 1 2/4 6-5 Amendment No. I

CONTAINMENT SYSTEMS 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.

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 i test by visual inspection of those surfaces. This inspection shall be performed prior to the Type A contcinment 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 13 years, to verify no apparent changes in appearance or other abnormal degradation. l 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.

I BRUNSWICK - UNIT 1 3/4 6-6 Amendmenc No. I a _.

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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. Revision 0.

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

3/4.6 CONTAINMENT SYSTEMS BASES i M F'-

3/4.6.1 PRIMARY CONulNMENT 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. +

The leakage rate acceptance criteria of s 0.60 L, for the combined Type B and C tests and 5 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. I NRC Regulatory Guide 1.163. Revision 0 (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.

2. NRC Regulatory Guide 1.163. Revision 0. 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."

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 B 3/4 6-la 1endment No. I

CONTAINMENT SYSTEMS BASES 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The primary containment air lock forms part of the prim.sry 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 that 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 containment is designed with a maximum allowable leakage rate (L.) of 0.5 percent by weight of the containment air ser 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 must be OPERABLE.

the air lock must be in compliance with the Type B air lock leakage test and both Sir 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 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 Prirery 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. Revision 0, 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 approved exceptions (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 mairtain 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 5/4 6-2 Amendment No. I

a J

~'

CONTAINMENT SYSTEMS ,

BASES ;

3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS (Continued) :

References:

j

1. 10 CFR Part 50. Appendix J.

2. NRC Regulatory Guide 1.163. Revision 0. dated September 1995. l

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

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

" Industry Guideline for Implementing' Performance-Based Option of 10 CFR 50 i 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 i event of a LOCA. A visual inspection in conjunction with the Primary i Containment Leakage Rate Testing Program is sufficient to demonstrate this :

capability. !

i

References:

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

2. NRC Regulatory Guide 1.163. Revision 0. dated 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 aeak 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 :

containment pressure will limit the total pressure to 49 psig, which is less than the design pressure and is consistent with the accident analyses.

1 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 i

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

-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.

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.

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.

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 ae 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 Proarams and Manuals l The following programs shall be established, implemented, and maintained:

6.8.3.1 Primary Coolant Sources Outside Containment 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

I ADMINISTRATIVE CONTROLS PROCEDURES.: PROGRAMS. AND MANUALS (Continued) 'l l

1. Preventive maintenance and periodic visual inspection l requirements, and !

t

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

6.8.3.2 In flant Radiation Monitoring I !

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. .

2. Procedures for monitoring, and j i

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 atmosriere samples under accident l conditions. The program shall include t1e following:

1. Training of personnel.. t 1

2. Procedures for sampling and analysis, and :

3. Provisions for maintenance of sampling and analysis equipment, i ,

l 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.

f i

Appendix J. as modified by approved exemptions. This program shall !

. be in accordance with the guidelines contained in Regulatory :

Guide 1.163. " Performance-Based Containment Leak-Test Program." dated

[

September 1995 as modified by the following exceptions:

1

1. Compensation of instrument inaccuracies applied to the ;

containment leakage total per ANSI /ANS 56.8-1987 instead of ANSI /ANS 56.8-1994. l The peak calculated containment internal pressure for the design !

1 basis loss of coolant accident. P, is 49 psig. l t

The maximum allowable primary containment leakage rate. L, shall be !

O.5% of primary containment air weight per day at P,. ,

l

i I

!- BRilNSWICK . UNIT 1

. 6-17 Amendment No. l ,

i

ADMINISTRATIVE CONTROLS 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 tb.at may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

BRUNSWICK - IINTT 1 6-17a Amendment No. I

I pi

i ENCLOSURE 8 ,

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

- CONTAINMENT LEAKAGE RATE TESTING

' 1YPED TECHNICAL SPECIFIChTION PAGES - UNIT 2 I

C 3

i

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

.m 5' -'

. l ADMINISTRATIVE CONTROLS' q

i

SECTION' .Pgf l m ;

6.5 REVIEW AND AUDIT (Continued)- s. i 6.5.4 NUCLEAR ASSESSMENT SECTION INDEPENDENT REVIEW PROGRAM Fu nc t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-10 j 1

Organization............... .............................

. 6-10 :- ;

Review.......... . ......................................... 6-11" Records...;.. ....................... .................... '6-12 ;

i 6.5.5 NUCLEAR ASSESSMENT SECTION ASSESSMENT PROGRAM.. ..... ... 6-13 i 16.5.6 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 i i

6.9 ~ REPORTING REQUIREMENTS !

Routine Reports................ ......................... 6-17a Startup Reports.......................................... 6-17a !

l Annual Reports............................. . . . . . . ....... 6-18 ;

Personnel Exposure and Monitoring . Report. . . . . . . . . . . . . . . . . 6-18 Annual Radiological Environmental Operating Report. . . . . . . 6-19 ,

Semiannual Radioactive Effluent Release Report. . . . . . . . . . . 6-20 l

. Monthly Operating Reports................................ 6-21 :

F Special Reports......................... ................ 6-22 f Core Operating Limits Report............................. 6-22 l i

1 i

i i

BRUNSWICK - UNIT 2 XV l Amendment No. l .

J l

1

-___-.-----L--. - - ,, ,-

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 penetrationsandvalvessubjecttoTypeBandCtestswben 3ressurized 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.

ACTION:

With:

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

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.

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

i 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. !

1 l

. SURVEILLANCE REQUIREMENTS l 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 1 per 18 months, i P

1 l

1 I

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

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

i CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 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 per hour when the gap between the door seals is pressurized to_10

.psig*: '

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

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.

r

b. By conducting an overall air lock leakage test at P . 49 psig, and by .

. verifying that the overall air lock leakage is with,in its limit: !

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

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: ;

4

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.

1 1

i

Exemption to Appendix J of 10 CFR 50.

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

CONTAINMENT SYSTEMS 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.

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 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 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.

BRUNSWICK - UNIT 1 3/4 6-6 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 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.

Analytkal methods and assumptions involving the primary containment are presente0 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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, Revision 0, 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 compensa+1on 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 leatage 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.

I

3/4.6 LCONTAINMENT 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. ;

a The leakage rate acceptance criteria of s 0.60 L, for the combined Type B i 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. l Primary containment operability is maintained by limiting leakage to s 1.0 La. j

. 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 l 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. I NRC Regulatory Guide 1 163. Revision 0 (Reference 2) endorses NEI 94-01 I (Reference 3) which in turn identifies ANSI /ANS 56.8-1994. " Containment System l Leakage Testing Requirements" (Reference 4) as an acceptable standard i regarding leakage-rate test methods, procedures, and analyses. Reduced l 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. l 1972 (References 5 and 6). !

References:

1. 10 CFR Part 50. Appendix J. ;

2. NRC Regulatory Guide 1.163. Revision 0, 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 i Appendix J."

l

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

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

October 20, 1983.

l

_6._

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

i

7. Updated FSAR. Section 6.2. :

t

'8. Updated FSAR. Section 15.6.4. i i

i BRUNSWICK - UNIT 2 B 3/4 6-la Amendment No. 1 [

t 0

CONTAINMENT SYSTEMS:

BASES.

t 13/4.6.1:3 ' PRIMARY CONTAINMENT-AIR LOCKS The primary containment air lock forms part of the 3rimary containment

pressure boundary. As such, air lock integrity and leac 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 i result in a leakage rate in excess of that assumed in unit safety analysis. !

4 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 1 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 i containment is designed with a maximum allowable leakage rate (L.) of i 0.5 percent by weight of the containment air aer 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the maximum peak 7 4

containment pressure (P,) of 49 psig. This a'10wable leakage rate forms the

! basis for the acceptance criteria imposed on the surveillance requirements associated with the air lock. I i

j 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 !

i required to be.0PERABLE. 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 :

i Part 50. Appendix J. Option B (Reference 1). The Primary Containment Leakage 4

Rate Testing Program conforms with NRC Regulatory Guide 1.163. Revision 0.

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 approved exceptions (References 2 and 3).

4 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 8 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.

2. NRC Regulator Guide 1.163. Revision 0, dated September 1995.

" Performance- ased 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."

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. Revision 0, dated 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 )eak 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 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.

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

' N ADMINISTRATIVE CONTROLS w.

6.8' PROCEDURES. PROGRAMS. AND MANUALS I

- 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.

c. Surveillance and test activities of safety related equipment,

d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g .- 0FFSITE DOSE CALCULATION MANUAL implementation.

h. PROCESS CONTROL PROGRAM implementation.

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.

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

]e 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 i unit affected, j

c. The change is documented. reviewed pursuant to Specifications 6.5.2.1 l' 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 Proarams and Manuals I The following programs shall be established, implemented, and l maintained: j 6.8.3.1 Primary Coolant Sources Outside Containment I !

I A program to reduce leakage from those portions of systems outside i containment that could contain highly radioactive fluids during a !

serious transient or accident to as low as practical levels.. The i program shall include the following- ,

l l

8RUNSWICK - UNIT 2. 6-16 Amendment No. I !

4

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

~ ' '

ADMINISTRATIVE CONTROLS' PROCEDURES.' PROGRAMS. AND MANUALS (Continued) l. }

1. Preventive maintenance and periodic visual inspection l requirements.'and i i

l2. Integrated leak test requirements for each system at' refueling j cycle intervals or'less. !

i 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 i conditions. This program shall include the following:

}

1. ' Training of personnel..

2. Procedures for' monitoring, and i i

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 j reactor coolant, radioactive iodines and ) articulates in plant ;

gaseous effluents, and containment atmos)1ere samples under accident :

conditions. The program shall include tie following: ;

1. Training of personnel, f l

2. Procedures for sampling and analysis, and l l

3. Provisions for maintenance of sampling and analysis equipment. !

6.8.3.4 Primary Containment Leakage Rate Testing Program 3 i

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

E of the containment as required by 10 CFR 50.54(o) and 10 CFR 50. i Appendix J. as modified by approved exemptions. This program shall l be in accordance with the guidelines contained in Regulatory l Guide 1.163. " Performance-Based Containment Le.sk-Test Program." dated '

September 1995 as modified by the following e ceptions:

l

1. Compensation of instrument inaccuracies applied to the !

-containment leakage total per ANSI /ANS 56.8-1987 instead of ;

, ANSI /ANS 56.8-1994, i

[ The peak calculated containment internal pressure for the design

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

J, The maximum allowable primary containment leakage rate. L , shall be ;

r 0.5% of primary containment air weight per day at P,. l I

a

DRUNSWICK - UNIT 2- 6-17 Amendment No. l !

ADMINISTRATIVE CONTROLS-6.9 REPORTING REOUIREMENTS I

ROUTINE REPORTS i

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.- l

! i STARTUP REPORTS i !

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

nuclear, thermal, or hydraulic performance of the plant. l l !

4

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

!

ML20094Q259

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