ML20149J747
| ML20149J747 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 02/17/1988 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20149J750 | List: |
| References | |
| NUDOCS 8802230202 | |
| Download: ML20149J747 (10) | |
Text
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CAROLINA POWER & LIGHT COMFAhY, et al.
DOCKET NO. 50-325 ERUNSWICK STEAM ELECTRIL PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATILG LICENSE Amencnent No. 11C License Na, DPR-71 1.
The Nuclear Reguldtory Comission (the Comission) has found that:
A.
The application for amendnent filed Ly Carolina Fcwer & Light Company (the licensee), cated August 5, 1987, corplies with the standaros and requirements of the Atonic Energy Act of 1954, as amended (the Act),
and the Cce.tssion's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the prcvisions of the Act, and the rules and regulations of the h
Comission; C.
There is reasonable assurance: (1) that the activities authorized by this avendment can be conducted without endangering the health and safety of tne public, and (11) that such activities will be conductec in ccmpliance with the Comission's regulations; D.
The issuance of this arenement will not be inimical to the cernon defense and security or to the health and safety of the public; or.d E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations ana all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to thir. license amendment; and paragraph 2.C.(2) of Facility Operating License Nc. DPR-71 is hereby atended to read as follows:
BB022*/opc2 880217 DR A;%cg 0s0003g4 p
. I (2) Technical Specifications i
The Technical Specifications contained in Appendices A and B as revised through Amendment ho. 116, are hereby incorporated in the license. Carolina Power & Light Company shall operate the facility l
in accordance with the Technical Specifications, 3.
This license amendment is effectiva as of the date of its issuance and shall be irplemented within 60 days of issuance, j
FOR THE NOCLEAR REGULATORY COMMISSION Elinor G. Adensaa, Director Project Directorate !!-1 l
Division of Reactor Projects 1/11 l
Attachment:
Changes to the Technical Specifications
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Date of Issuance: February 17, 1938
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ATTACHPENT TO LICENSE AMENtMENT NO.116 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Replace the following 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 8 3/4 6-1 B 3/4 6-1
I e
l 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAIhHENT 3/4.6.1.1 PRIKARY CONTAIhHENT INTECRITY j
Primary CONTAIbHENT INTEGRITY ensures that the release of radioactive materials from the centsinment 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 i
site boundary radiation doses to within the limits of 10 CFR Part 100 during j
accident conditions.
3/4.6.1.2 PRIMARY CONTAIhHENT LEAXACE The limitations on primary containment leakage rates ensure that the tout I
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 I
g conservatism, the measured overall integrated leakage rate is further limited l
to less than or equal to 0.75 L, or 0.75 L, as applicable, during perfermance of the periodic tests to acteunt forpossibledegradationofthecontainment leakage barriers between leakege tests.
Operating experience with the main steam line isolation valves has indicated that degradation has occasionhily occurred in the leak tightness c6 i
the valvest therefore, thi special requirement for testing these valves.
Exemptions from the requirements of 10 CrR Part 50 bave been granted for main steam isolation valve leak testing, testing of airlocks after each
)
opening, and leakage calculation methods.
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Appendix J, paragraph III.A.3 requires that all Type A (Containment I
Integrated Leak Rate) tests be performed in accordance with ANSI W45.4-1972, "Leakage Rate Testing of Containment Structures for Nuclear Reactors."
ANSI N45.4-1972 requires that leakage calculations be performed using the Point-to-Peint or Total Time method. ANSI N45.4-1972 has beea revised to a new standard, ANSI /ANS 56.8-1981, "Containment System Leakage Testing," which incorporates the Mass-Point method for leakage calculations. Type A tests l
will be performed in conformance with ANSI N45.4-1972 but.will use the l
Mass-Point method for calculation of leakage rates as described in l
ANSI /ANS 56.8-1981.
t i
3/4.6.1.3 PRIMARY CONTA!hYENT AIR LOCVS The limitations on closure and leak rate for the containment air locks see l
required to meet the restrictions on PRIMARY CONTAINMEN7 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.
IRI.TNS'='ICK - UNIT 1 g 3/4 6-1 Amendment No. /l,116
t CONTAINMENT SYSTEMS LIMITINC CONDITION FOR OPERATION (Concinued)
ACTION (Continued)
The leakage rate to less than er 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 shan be determined in conformance with the criteria specified in Appen11x J of ifl CFR 50 using the l
methods and provisions of ANSI N45.4-1972, except that leakage rates for Type A tests shall be calculated using the Mass-Point method as specified in ANSI /ANS 56.!-1981*.
The primary centainment leakage rates shall be l
demonstrated at the following test schedule l
Three Type A Overall Integrated Containment Leakage Rate tests shall 1
a.
be conducted at 40 + 10 month intervals during shutdown at P 49 psig, or P. 25 psis, during each 10 year service period.,,The j
g third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection, If any periodic Type A test fails to meet either 0.75 L, or 0.75 L b.
scheduleforsubsequentTypeAtestsshallbereviewedank.
l the test approved by the Comission.
If two consecutive Type A tests fall re 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 l
first, until two consecutive Type A testa 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 whicht I
2 1.
Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Tyee A test data is within 0.25 L, or 0.25 L.
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2.
Has duration sufficient to establish. accurately the change in leakage rate between the Typa A test and the supplemental test.
1 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 pois, or P. 25 peig.
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- Exemption from Appendix J of 10CFR50.
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BRUNSii!CK - UNIT 1 3/4 6-3 Amendment No. H,116 a
8 CAROLINA POWER & LIGHT COMPANY, et al.
CCCXET h0. 50-324 BRUNSWICK STEAM ELECTRIC PLANT UhlT 2 ANENCLENT TO FACILITY OPERATIhG LICENSE Arenenent Nu 144 License No. OPR-62 1.
The huclear Regulatory Comission (the Connission) has found that:
A.
The application for amencrent filed by Carolina Pcwer & Light Company (the licensee), dated August 5, 1987, corplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Ccmission's rules and regulations set forth in 10 CFR Chapter 1; b.
The facility will operate in conformity with the application, the provisions of the Act, anc the rules and regulatiens of the Comis sicn; C.
There is reasonable assurance: (1) that the activities authori:eo by this ar.endment can be conducted without endangering the health dnd Sofety of the public, and (ii) that such activities will be cenducted in ccepliance with the Comission's regulations; D.
The issuance of this arenament will not be inimical to the corecn defense and security or to the beelth ano safety o! the public; cr,d 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 satisfieo.
2.
Acccrdingly, the license is emended by changes to the Technical Specifications, as indicated in the attachment, to this license amenoment; and paragraph 2.C.(2) of Fact 11ty Operating License No. OPR-62 is hereby ar4nded to read as follows:
k i
(2) Technical Specifications The Technical Specifications centained in Appendices A and B, as i
revised through Amendrent No.144, are hereby incorporated in the Ifcense.
Carolina Power & Light Company shall operate tie facility i
in accordance with the Technical Specificatiors.
3.
This license amendnent is effective as of the date of its issuance and shall be implemented within 60 days of issuance.
l FOR THE NUCLEAR REGULATORY Com ISSION i
Elinor G. Adensam. Director Project Directorate 11-1 Division of Feactor Projects I/II l
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Attachment:
i Changes to the Technical Specifications t
Date of Issuance: February 17, 1988 l
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I ATTACPMENT TO LICENSE AMENDMENT NO.144 FACILITY OPERATING LICENSE NO. DPR-62 i
DOCKET NO. 50-324 1
1 Replace the following pages of the Apperdix A Technical Specifications with the enclosed pages. The revised areas are indicated by carginal lines, Pemove Pages Insert Pages l
3/4 6-3 3/4 6-3 B 3/4 6-1 B 3/4 6-1 i
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3/4.6 CONTAINMENT SYSTEMS BASES j
)!
3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTECRITY I
Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials f' rom 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 l
site boundary radiation doses to within the limits of 10 CFR Part 100 during I
accident conditions.
i l
l 3/4.6.1.2 PRIMARY CONTAINMENT LEAXACE The limitations on primary containment leakage rates ensure that the total f
i 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 conservatiam, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, or 0.75 L, as applicable, during perforr.ance l
of the periodic tests to account for possible degradation of the containment l
leakage barriers between leakage tests.
1 Operating experience with the main steam line isolation valves has indicated that degradation has occasienally occurred in the leak tightness of the valvest therefore, the special requirsment for testing these valves.
l Exe ptions from the requirements of 10 CFR Part 50 have been granted for j
sain steam isolation valve leak testing, testing of airlocks after each j
sponing, and leakage calculatior methods.
I Appendix J, paragraph III.A.3 requires tbst all Type A (Containment Integrated Leak Rate) tests be performed in at.cordance with ANSI W45.4-1972, i
"Leskage Rate Testing of Containment Structuret for Nuclear Reactors."
ANSI W45.4-1972 requires that leakage calculations be performed using the Point-to-Point or Total Time method.
ANSI W45.A-1972 has been revised to a new standard, ANS!/ANS 56.8-1981, "Containment System Leakass Testing," which incorscrates the Mass-Point method for leakage cd culations. Type A tests will 1e performed in conformance with ANSI W45.4-1972 but will use the Mass-Piint method for calculation of leakage rates as described in ANSI /APS 56.8-1981.
3/4.6.1.3 PAIMARY COMTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on PRIMARY CONTAIWMENT INTECRITY and leak rate given in Specifications 3.6.1.1 and 3.6.1.2.
The specification makes j
allowances for the fact that there say be ung periods of time when the att j
i locks will be in a closed and secured position during reactor operation.
i j
j BRUNSWICK - UNIT 2 g 3/4 6-1 Amendment No.3, 144 2
1 J CONTA!WENT SYSTEMS LIMITINC CONDITION FOR OPERATION (Continued) ) i ACTION (Continued) j i i c. The leakage rate to less than or equal to 11.5 sef per hour for ary l one main steam line isolation valve, l I i prior to increasing reactor coolant system temperaturc above 212'F. i j i SURVEILLANCE REQUIREMENTS r 4.6.1.2 The primary containment leakage rates shall be determined in I ~ conformance with the criteria specified in Appendix J of 10 CFR 50 using the l j motheds and provisions of ANSI N45.4-1972, except that leakage rates for j l Type A tests shall be calculated using the Mass-Point eethod as specified in l l ANSI /ANS 56.5-1981*. The primary containment leakage ratis shall be demonstrated at the foMoving schedulet I i a. Three Type A Overall Integrated Containment Leakage Rate tests shalt be conducted at 40 2 10 month intervals during shutdown at P I l 49 psig, or P. 25 psis, during each 10 year service period.,,The g f } third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection. j b. If any periodic Type A test falls to meet either 0.75 L or 0.75 L l the test schedule for subsequent Type A tests shall be, reviewed ank, j i approved by the Comission. If two consecutive Type A tests fail tn { eeet 0.75 L, or 0.75 L, a Type A test shall be performed at each l plant shutdownforrefbelingorevery18 months,whicheveroccurs first, until two consecutive Type A tests meet 0.75 L, or 0.75 L, at l j g which time the above test schedule any be resumed. 3 1 i ) c. The accuracy of each Type A test shall be verified by a supplemental i test whicht j 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. g i l l 2. Was duration sufficient to establish accurately the change in i i leakage rate between the Type A test and the supplemental test. 3. Reguires 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 seasured leakage i st P,, 49 pois or P. 25 psig. g b
- Euemption from Appendix J of 10CFR50.
f I } 3RUNSWICK - UNIT 2 3/4 6-3 Amendment No. 7)r, R? l l } .}}