ML20023B913
| ML20023B913 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/18/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Arkansas Power & Light Co |
| Shared Package | |
| ML20023A491 | List: |
| References | |
| DPR-51-A-077 NUDOCS 8305090154 | |
| Download: ML20023B913 (7) | |
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NUCLEAR REGULATORY COMMISSION 3,
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ARVANSAS P0k'ER & LIGHT COMPANY DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT NO.1 AMENDMENT TO FACILJTY OPERATING LICENSE c
Amendment No. 77 License No. DPR-51 1.
The Nuclear Regulatchy Comission (the Comission) has found that:
A.
The application for amendment by Arkansas Power and Light Company (the licensee) dated October 19, 1977, as supplemented December 15, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act} and th_e Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 conducted in compliance with the Comission's regulations; D.
The issuance of this amend:nent 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 CPR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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A:cordingly, the license is amended by changes to the Technical Specifications as indicated ir. the attachment to this license amendment, and paragraph 2.c.(2) of Facility Operating License No. DpR-51 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.77, are hereby incorporated in the license. The licensee shall operate the fapility in accordance with the' Technical Specifications.
3.
This license amendment is e'ffective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION oh ThStolf, Chief -
Ope ating Reactors Br. h !4
- ision of Licensino *
Attachment:
Changes to the Technical Specifications Date of Issuance: April 18,1983 l-(
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s ATTACHMENT TO LICENSE AMENDMENT NO. 77 FACILITY OPERATING LICENSE NO. DPR-$1 DOCKET NO. 50-313 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment -
number and contain vertical lin,es indicating the area of change.,
Remove Insert 67 67' 76 76 77 77 77b 77b
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4 SURVEILLAf!CE REOUIREMENTS Specified surveillance intervals may be adjusted plus or minus 25 percent to accommodate normal test and surveillance schedules.
The maximum combined interval for any 3 consecutive tests shall not e.rceed 3.25 ' times the specified surveillance intervah Surveillance ennuirements are not applicable when the plant operati6g conditions are below those requiring operability of the designstad component.
Hewever, the required surveillance must be performed prior to reaching the operating conditions requiring operability.
For example, instrumentation requiring twice per week surveillance when the reactor is Mal need not have the required surveillance when the reactor i.:..dtdewn.
Inservice inspection of ASME' Code Class 1, 2, 3 components shafl be perfomed in accordance with Section XI of the ASME Boiler and pressure Vessel. Code and applicable addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted ty the !!RC pursuant to 10 CFP, 50, Section50.55a(g)(6)(1).
Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation.
Exceptions to these requirements are stated in the individual specifications.
Surveillance Requirements do not have to be performed on inoperabTe equiptr.ent.
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4.1 OPERATIONAL SAFETY ITEMS Apolicability Applies to items directly related to safety limits and limiting conditions for operation.
Obiective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.
Specification The minimum frequency and type of surveillance required for a.
reactor protective system and engineered safeguards system i
instrumentation when the reactor is critica_1 shall be as stated in l
Table 4.1-1.
b.
. Equipment and sampling test shall be performed as detailed in Tables 4.1-2 and 4.1-3.
Discrepancies noted during surveillance testing will be corrected c.
and recorded.
i d.
A power distribution map shall be made to verify the expected.
power distribution at periodic intervals at least every 10 effective full power days using the incore instrumentation j
detector system'.
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8 Amendment ilo. X,R, 77 67
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REACTOR C001R;T SYSTEM SifRVEILI.ANCE 4.2 i
Aerlicability ssure boundary.
Applies to the surveillance of the reactor coolant system pre Objective.
system pressure To assure the continued integrity of the reactor coolant boundary.
Specification operation, an ultrasonic test survey shall bc d
equired to Prior to initial unit made of reactor coolant system pressure boundary wel s as r 4.2.1 for future establish preoperational integrity and baseline data inspections.
in accordance Post-operational inspections of components shall be made i
XI of the AS!E with the methods and intervals indicated in Sect on 4.2.2 uired by Boiler and Pressure Vessel Code and applicable add lief has been granted by the NRC.
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1 76 Amendment I;o.7)f 77 e
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4.2.3 The structural integrity of the reactor coolant system boundary shall be maintained at the level required by the original acceptance standards throughout the life of the station. Any evidence, as a result of the tests outlined in Table 15-261 cf Section XI of the code, that defects have. developed or grown, shall be Tnvestigated.
4.2.4 To. assure the struclu'ral integrity of the reactor inte'rnals throughout the life of the unit, the two sets of main internals bolts (connecting the core barrel to the core support, shield and to the lowe'r grid cylinder) shall remain in place and under tension.
This will be verified by visual inspection to determine
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that the welded bolt locking caps remain in place.
All locking caps will be inspected after hot functional testing and whenever the internals are removed from the' vessel during a refueling or maintenance shutdown. The core barrel to core support shield caps will be inspected each refueling shutdown.
4.2.5 Sufficient records of each inspection shall be kept to allow comparison and evaluation of future inspections.
4.2.6 Surface and volumetric examination of-the reactor coolant pump flywheels will be conducted coincident with' refueling or maintenance shutdowns such that during 10 year intervals all four reactor coolant pump flywheels will be examined.
Suen examinations will be performed to the extent possible through the access ports, i.e., those areas of the flywheel accessible without motor disassembly. The surface and volumetric examination may be accomplished by Acoustic Emission Examination as an initial examination method.
Should the results of the Acoustic Emission Examination indicate that additional examination is necessary to ensure the structural integrity of the flywheel, then other appropriate NDE methods will be performed on the area of concern.
4.2.7 The reactor vessel material irradiation surveillance specimens 4
removed from the reactor vessel in 1976 shall be installed, irradiated in and withdrawn from the Davis-Besse Unit No. I reactor vessel in accordance with the schedule shown in Table 4.2-1.
Following withdrawal of each capsule listed in'.
Table 4.2-1, Arkansas Power & Light Company shall be responsible for testing the specimens and submitting a report of test results in accordance with 10CFR50, Appendix H.
77 Amendment No.12,fE,7E,J4", 77
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i Table 4.2-1 ANO-1 CAPSULE ASSDf3LY WITdDRAWAL SCHEDULE AT DAVIS-BESSE 1 t
I IkS5RTION/ WITHDRAWAL CAPSULE ANI-E '
Has been withdrawn for testing ANI-B Withdraw following 1st cycle at Davis-Besse 1 ANI-A-Withdraw following 3rd cycle at
. Davis-Desse 1 ANI-C Withdraw fo116 wing 7th Cycle at Davis-Besse 1 Insert in location 12 (upper) prior ANI-D to 4th cycle at Davis-Besse 1; withdraw following 12th cycle
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Insert in location YZ (upper) prior.
ANI-F to 4th. cycle at Davis-Besse 1; withdraw following lith cycle Bases The surveillance program has been developed to comply with the a'pplicable edition of Section XI and addenda of the AS}E Boiler and Pressure Vessel Code, Inservice Inspection of Nuclear Reactor Coolant Systems, as required by 10 CFR 50.55a, to the extent practicable within licitations of design, gec=etry and materials of construction.
The nu=ber of reactor vessel specimens and the frequencies for removing and testing these specimens are provided to assure compliance with~ the requirements of Appendix E to 10, CFR Part 50.
For the purpose of Technical Specification 4.2.8, the definition of Regulatory Guide 1.16, Revision 4 (August 1975) applies for the term l
"co=mercial operation".
Cumulative reactor utilization factor is defined as:
[(Cumulative thermal megawatt hours since attainment of commerical operation at 100% power) x 100] 4 [(licensed thermal power) x (cumulative hours since attainment of commercial operation at 100% power)].
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