ML19320A251

Proposed Tech Specs 6.12 & 4.2 Re Reporting Requirement for Annual Operating Rept

ML19320A251
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/01/1977
From:	ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML19320A246	List: ML19320A245 ML19320A251
References
NUDOCS 8004210542
Download: ML19320A251 (7)

Text

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b 6.12 REPORTING REQUIREMEN'IS 6.12.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted.

6.12.2 Routine Reports 6.12.2.1 Startup Report A summary report of plant startup and power escalation testing shall bc

.ubmitted 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. The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or comence-ment of commercial power operation, or (3) 9 months following initial criti- -

cality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

6.12.2.2 Occupational Exposure Data Report _1]

An Jccupational Exposure Data Report for the previous calendar year shall be submitted prior to March 1 of each year. The report shall contain a tabulation on an annual basis of the number of station, utility :uid other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions, 2f i.g., reactor operations and surveillance, inservice in-spection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling JJ A single submittal may be made for a multiple unit station. The sub-mittal should combine those sections that are common to all units at the station.

2f This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.

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The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted "or. In the aggregate, at least 80*6 of the total whole bode dose received from external sources shall be assigned to specific major work functions.

6.12.2.3 Monthly Operating Report Routine reports of operating statistics which include:

(1) Average Daily Unit Power Level (2) Operating Data Report (3) Unit Shutdowns and Power Reductions (4) Narrative Summary of Operating Experience Shall be submitted on a monthly basis to the Director, Office of Manage-ment Infomation and Program Control, U. S. Nuclear Regulatory Commission, Nashington, D. C. 20555, with a copy to the appropriate Regional Office, by the tenth of each month following the calendar month covered by the report.

6.12.3 Peportable Occurrences Reportable occurrences, including corrective actions and measures to prevent reoccurrence, 'shall be reported to the NRC as required below.

' Supplemental reports may be required to fully describe final resolution -

of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the orig-inal report date.

6.12.3.1 Prompt Notification With Written Followup  ;

The types of events listed below shall be reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by telephone and confimed by telegraph, ,

mailgram, or facsimile transmission to the Director of the appropriate  :

Regional Office, or his designate no later than the first working day

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following the event, with a written followup report within two weeks. A copy of the confimation and the written followup report shall also be sent to the' Director, Office of Management Infomation and Program Control, USNRC. The written report shall include, as a minimum, a completed copy of {

l a licensee event report form. Infomation provided on the licensee event report fom shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

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-l (a) Failure of the reactor- protection _ system or other systens subject to limiting safety systcra -settings to initiate the 're:ttui. red protective function by the time a monitored parameter reaches the setpoint speci-fled as - the limi ting safety system set ting in :t he. rechnical speci fi-cations or failure to complete the required protective function.

NOTE:

' Instrument drift discovered as a result of testing need not be reported under this-item but my be . reportable under items . (e), _ (f), or 6.12.3.2(a).

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IS-261 Item Component Exception 6.4 Bolting 26 Not Applicable 6.6 Integrally Welded Not Applicable Valve Supports A report covering results of each inspection shall be submitted as required by ASME Section XI.

4.2.3 .The structural integrity of the tsactor coolant system boundary shall be maintained at the level required by the original accep-tance standards throughout the life of the station. Any evidence, as a result of the tests outline 1 in Table 15-261 of Section XI of the code, that def ects have de , eloped or grown, shall be investigated.

4.2.4 To assu'e_the r structural integrite of the reactor internals through-out the life of the unit, the twe sets of main internals bolts (connecting the core barrel to the core support, shield.and to the lower grid cylinder) shall remain in place and under tension. This will be verified by visual inspection to determine that the welded bolt locking caps remain in place. All locking caps will be inspect-ed after hot functional testing and whenever the internals are removed f rom 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 .

cecparison and evaluation of future. inspections.

4.h.6 Complete surface and volumetric examination of the reactor coolant pump flywheels will be conducted coincident with refueling or maintenance shutdowns such that within a 10 year period af ter start-up all four reactor coolant pump flywheels will be examined.

4.2.7 The reactor vessel material irradiation surveillance specimens removed from the rcactor vessel in 1976 shall be installed, '

irradiated in and withdrawn from the Davis-Ecsse Unit No. 1 l

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 - submit ting a report of test results in accordance with'10 CFR 50, Appendix H.

Amendment No. Z2, 20, 22 77 l

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Successive Inspection Intervals ..

Every 10 years thereafter (or Volumetric inspection of two of nearest refueling outage) the welds at the expiration of i

i cach 1/3 of the inspection inter-val with a cumulative 100%

coverage of all welds.

Note - The welds s.lccted during each inspection period shall be distributed among the total number to be examined to provide a representative sampling of the conditions of the wells, 4.15.3 In the event repairs of any welds are required following any examin-

-at. ion during successive inspection intervals, the inspection schedule for the repaired welds will revert back to the first 10 year inspection program.

l 4.15.4 Examinations that reveal unacceptable structural defects in a weld I during an inspection tmder 4.15.2 should be extended to require an additional inspection of another 1/3 of the welds. If further ' l unacceptable defects are detected in the second sampling, the remainder of the welds shall be inspected.

4.15.5 Repairs, reexamination and piping pressure tests shall be conducted in accordance with Section Xi of the ASME Code.

4.15.6 lf A report covering results of each. inspection'shall Be submitted as required by ASME Section XI.

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8. . ' Tube Inspection means an inspection of the steam generator tube trom tne point of entry completely to the point of exit,

b. The steam generator shall be determined operable after completing the corresponding actions (plug all tubes exceeding the plugging

'imit and 'all tubes containing through-wall cracks) required by

~ able 4.18.2.

4.18.6 Reports Following each inservice inspection of steam generator tubes, a report covering results shall be submitted as required by ASME Section XI unless previously reported by LER per 6.12.3. Results of steam generator tube inspections which fall into Category C-2 or C-3 and require prompt notifi-cation of the NRC shall be reported per Specification 6.12.3.

Bases

-The surveillance requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS wil1Ebe maintained. The program for inservice inspection of steam gen- ~

erator tubes is based on a modification of Regulatory Guide 1.83, Revi-sion 1. Inservice inspection of steam generator tubing is essential in order to maintain-surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degrad-ation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. ~

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