ML20245H754
| ML20245H754 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 08/04/1989 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245H759 | List: |
| References | |
| NUDOCS 8908170268 | |
| Download: ML20245H754 (12) | |
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. UNITED STATES
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2" NUCLEAR REGULATORY COMMISSION -
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WASHINGTON, D. C. 20555 8
%,,.....f CAROLINA POWER & LIGHT COMPANY, et al.
DOCKET NC. 50-325 B_RUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.136 License No. DPR-71 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment filed by Carolina Power & Light Company (the licensee), dated May 1,1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance: (1) that the activities authorized by this amendment can be conducted wi.thout endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications, as-indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. DPR-71 is hereby amended to read as follows:
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(2) Technical Specifications
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The Technical Spec (fications-contained in Appendices A and B, as revised through Amendment No.136, are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
3.
This' license amendment is effective as'of the date of its issuance'and i
shall be implemented within 60 days of issuance.
]
FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By:
Elinor G. Adensam, Director Project Directorate II_1 i
Division of Reactor Projects I/II l
Office of Nuclear Reactor Regulation d
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Attachment:
Changes to the Technical Specifications Date'of Issuance:
August 4,1989 i
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. ATTACHMENT TO' LICENSE AMENDMENT NO. - 136 t
' FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325
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' Replace 'the following pages of-the Appendix A Technical Specifications with.
the. enclosed pages. The revised areas are indicated by marginal lines.
Remove Pages Ir. sert Pages 3/4 6-3' 3/4 6-3
- B 3/4 6-1 B 3/4 6-1 B 3/4 6-2 B 3/4 6-2 p
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CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued)
The leakage rate to less than or equal to 11.5 sef per hour for any c.
one main steam line isolation valve, prior to increasing reactor coolant system temperature above 212*F.
SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated at the following schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50:
Three Type A Overall Integrated Containment Leakage Rate tests shall a.
be conducted at 40 + 10 month intervals during shutdown at P 49 psig, or P, 25 psig, during each 10 year service period.,
g The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection.
b.
If any periodic Type A test fails to meet either 0.75 L, or 0.75 L,
the test scheduleforsubsequentTypeAtestsshallbereviewedank approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L, or 0.75 L, a Type A test shall be performed at each g
plant shutdown for refueling or every 18 months, whichever occurs first, until two consecutive Type A tests meet 0.75 L, or 0.75 L, at g
which time the above test schedule may be resumed.
c.
The accuracy of each Type A test shall be verified by a supplemental test which:
1.
Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L, or 0.25 L.
t 2.
Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.
3.
Requires the quantity of gas injected into the containment or bled from,the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage rate at P,, 49 psig, or P, 25 psig.
g BRUNSWICK - UNIT 1 3/4 6-3 Amendment No. ES, 775,136
)
3/4.6 CONTAINMENT SYSTEMS BASES j
l 3/4.6.1 PRIMARY CONTAINM g 3/4.6.1.1 PRIMARY CONTAINMENT INTECRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage j
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.
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1 3/4.6.1.2 PRIMARY CONTAINMENT LEAKACE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 49 psig, P. As an added conservatism, the measured overall integrated leakag,e rate is further limited l
to less than or equal to 0.75 L, or 0.75 L, as applicable, during performance i
t of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
Operating experience with the main steam line isolation valves has indicated that 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 main steam isolation valve leak testing, testing of airlocks after each opening, and leakage calculation methods.
Appendix J, paragraph III.A.3 requires that all Type A (Containment Integrated Leak Rate) tests be conducted in accordance with American National Standard (ANSI) N45.4-1972, " Leakage Rate Testing of Containment Structures for Nuclear Reactors," March 16, 1972. In addition to the Total Time and Point-to-Point methods described in that standard, the Mass Point method, when used with a test duration of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, is an acceptable method to use to calculate leakage rates. A typical description of the Mass Point method can be found in ANSI /ANS 56.8-1987, " Containment System Leakage Testing Requirements," January 20, 1987. 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 1 and 2).
References:
1.
CP&L Letter to Mr. D. B. Vassallo, " Integrated Leak Rate Test,"
October 20, 1983.
2.
NRC Letter from Mr. D. B. Vassallo to Mr. E. E. Utley, December 9, 1983.
BRUNSWICK - UNIT 1 B 3/4 6-1 Amendment No. $$, 7/$, 778 1 36
CONTAINMENT SYSTEMS-BASES 3/a.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The limitations.on closure and leak rate for the containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and leak rate given in Specifications 3.6.1.1 and 3.6.1.2.
The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation.
3/4.6.1.4 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 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 Type A leakage tests
~
is sufficient to demonstrate this capability.
3/4.6.1.5 PRIMARY CONTAINMENT INTERNAL PRESSURE The limitations of primary containment internal pressure ensure that the containment peak pressure of 49 psig does not exceed the design pressure of 62 psig during LOCA conditions. The limit of 1.75 psig for initial positive containment pressure will limit the total pressure to 49 psig, which is less than the design pressure and is consistent with the accident analyses.
3/4.6.1.6 PRIMARY CONTAINMENT AVERACE 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 6-2 Amendment No. H,136 l
[p aa, UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
y WASHINGTON, D. C. 20555
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CAROLINA POWER' & LIGHT COMPANY, et al.
DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE l
Amendnent No.166 l
License No. DPR-62 l
l 1.
The Nuclear Regulatory Consnission (the Commission) has found that:
l A.
The application for amendment filed by Carolina Power & Light Company (the licensee), dated May 1,1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conoucted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment; and paragraph 2.C.12) of Facility Operating License No. DPR-62 is hereby anended to read as follows:
I 1
- 2'.
(2) Technical Specifications The Technical Specifications contained'in~' Appendices A and B, as revised through Amendment No.166, are hereby incorporated in the license. Carolina Power & Light Company shall operate the f acility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of. its issuance and shall be. implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION' Original Signed By:
Elinor G. Adensam, Director Project Directorate 11-1 Divilion of Reactor Projects I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 4,1989 l
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OFFICIAL RECORD COPY L-___---___--______-____________--
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' ATTACHMENT TO LICENSE AMENDMENT NO. 166
~
FACILITY OPERATING LICENSE NO. DPR-62
'l
' DOCKET NO. 50-324
.l Replace thel f611owing pages of.the Appendix A Technical Specifications with
!the. enclosed.pages. The revised areas are indicated by marginal lines.
Remove Pages Insert Pages 3/4 6-3
. 3/4 6-3 B 3/4 6-1 B 3/4 6-1 B 3/4 6-2.
B 3/4 6.
_..._.n
_ _ ~ _ - -. _ _ _ _ - _ - _ _ _ - _ _ - - _ - - - _ - - - - - -
s t
CONTAINMENT SYSTEMS LIMITINC CONDITION FOR OPERATION (Continued)
ACTION (Continued)
The leakage rate to less than or equal to 11.5 scf per bour for any c.
one main steam line isolation valve, prior t'o increasing reactor coolant system temperature above 212'F.
SURVEILLANCE REQUIREMENTS-4.6,.1.2 The primary containment leakage rates shall be demonstrated at the following schedule and shall be determined in conformance with the criteria specified in Appendix J of 10CFR50:
Three Type A Overall Integrated Containment Leakage Rate tests shall a.
be conducted at 40 1 10 month intervals during shutdown at P 49 psig, or P, 25 psig, during each 10 year service period.,,The t
third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection.
b.
If any periodic Type A test fails to meet either 0.75 L or 0.75 L '
t the test schedule for subsequent Type A tests shall be, reviewed and approved by the Commission. If two consecutive Type A testa fail to meet 0.75 L, or 0.75 L, a Type A test shall be performed at each t
plant shutdown for refueling or every 18 months, whichever occurs first, until two consecutive Type A tests meet 0.75 L, or 0.75 L, at g
which time the above test schedule may be resumed.
c.
The accuracy of each Type A test shall be verified by a supplemental test which:
1.
Confirms the acceracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L, or 0.25 L.
t 2.
Has duration sufficient to establish accurately the change in 1eakage rate between the Type A test and the supplemental test.
3.
Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at P,, 49 peig or P, 25 prig.
t BRUNSWICK - UNIT 2 3/4 6-3 Amendment No. 97, J$$, 166 s
p-t 3/4.6 CONTAINMENT SYSTEMS BASES
' 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1' PRIMARY CONTAINMENT INTECRITY 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 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 limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 49 psig, P,.
As an added conservatism, the measured overall integrated leakage rate is further liraited to less than or equal to 0.75 L, or 0.75 L, as applicable, during performance of the periodic tests to account forpossibledegradationofthecontainment leakage barriers between leakage tests.
Operating experience with the main steam line isolation valven has indicated that 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 main steam isolation valve leak testing, testing of airlocks after each opening, and leakage calculation methods.
Appendix J, paragraph III.A.3 requires that all Type A (Containment Integrated Leak Rate) tests be conducted in accordance with American National Standard (ANSI) N45.4-1972, " Leakage Rate Testing of Containment Structures for Nuclear Reactors," March 16, 1972. In addition to the Total Time and Point-to-Point methods described in that standard, the Mass Point method, when s
used with a test duration of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, is an acceptable method to use I
to calculate leakage rates. A typical description of the Mass Point method can be found in ANSI /ANS 56.8-1987, " Containment System Leakage Testing Requirements," January 20, 1987. 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 1 and 2).
]
1
References:
1 1.
CP&L Letter to Mr. D. B. Vassallo, " Integrated Leak Rate Test,"
October 20, 1983.
]
2.
NRC Letter from Mr. D. B. Vassallo to Mr. E. E. Utley, December 9, 1983.
BRUNSWICK - UNIT 2 B 3/4 6-1 Amendment No. Pl. 147, ISS 166
.. =
I CONTAINMENT SYSTEMS BASES l
3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS i
The limitations on closure and leak rate for the containment air locks are i
required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and leak rate given in Specifications 3.6.1.1 and 3.6.1.2.
The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation.
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 Type A leakage tests is sufficient to demonstrate this capability.
3/4.6.1.5 PRIMARY CONTAINMENT INTERNAL PRESSURE The limitations of primary containment internal pressure ensure that the containment peak pressure of 49 psig does not exceed the design pressure of 62 psig during LOCA conditions. The limit of 1.75 psig, for initial positive containment pressure will limit the total pressure to 49 psig, which is less than the design pressure and is consistent with the accident analyses.
3/4.6.1.6 PRIMARY CONTAINMENT AVERACE 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 6-2 Amendment No. 779, l
166
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