ML19312B876

Proposed Tech Specs 4.2.10 & 4.2.11 Re Reinstitution of Surveillance Program

ML19312B876
Person / Time
Site:	Oconee
Issue date:	03/10/1977
From:	DUKE POWER CO.
To:
Shared Package
ML19312B874	List: ML19312B873 ML19312B876
References
NUDOCS 7911250072
Download: ML19312B876 (2)

Text

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4.2.3 Ths structural integrity of ths Reactor Coolant System bo"ndary ,

shall be maintained at the level required by the original accep- l tance standards throughout the life of the station. Any evidence, as a result of the tests outlined in Table IS-261 of Section XI of the code, that defects have developed or grown, shall be investiga-ted, including evaluation of comparable areas of the Reactor Coolant System.

4.2.4 The results of the Inservice Inspections performed pursuant to Specifications 4.2.1, 4.2.2, and 4.2.3 shall be reported to the Commission within 90 days of completion.

4.2.5 To assure the structural integrity of the reactor internals throug-out the life of the unit, the two 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 inspec-ted 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.

4.2.6 Sufficient records of each inspection shall be kept to allow compari-son and evaluation of future inspections.

4.2.7 The inservice inspection program shall be reviewed at the end of five years to consider incorporation of new inspection techniques and equipment which have been proved practical and the conclusions of this review and evaluation shall be discussed with the NRC/DRL.

4.2.8 At approximately three-year intervals, the bore and keyway of each reactor coolant pump flywheel shall be subjected to an in-place, volumetric examination. Whenever maintenance or repair activities necessitate flywheel removal, a surface examination of exposed surfaces and a complete volumetric examination shall be performed, if the interval measured from the previous such inspection is greater than 6 2/3 years.

l 4.2.9 Reactor vessel material surveillance specimens representative of the materials present in the reactor vessel beltline region shall be irradiated in the Oconee or similar power reactor vessels or in test reactors. Insofar as possible, the irradiation withdrawal and examination of these specimens shall be scheduled to provide surveil-lance results necessary for updating Technical Specificatica 3.1.2.

The program shall be reviewed annually to assure that necessary data will be available when needed.

4.2.10 During the first two refueling periods, two reactor coolant system piping elbows shall be ultrasonically inspected along their longitu-dinal welds (4 inches beyond each side) for clad bonding and for cracks in both the clad and base metal. The elbows to be inspected are identified in B&W Report 1364 dated December, 1970.

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4.2.11 To eccura that reactor internals vant valves are not opaning during operation, all vent valves will be inspected during each refueling outage to confirm that i:o vent valve is stuck open and that each valve operates freely.

Bases The surveillance program has been developed to comply with Section XI of the ASME Boiler and Pressure Vessel Code, Inservice Inspection of Nuclear Reactor Coolant Systems, 1970, including 1970 winter addenda, edition. The program places major emphasis on the area of highest stress concentrations and on areas where fast neutron irradiation might be sufficient to change material properties.

Irradiation of reactor vessel material surveillance specimens representative of the materials present in the Oconee reactors provides the capability of determining radiation induced changes in the mechanical and impact properties in the region of the reactor vessel surrounding the core. Test specimens will be installed in holder tubes placed inside the reactor vessel at Crystal River Unit 3 and other similar power reactors and test reactors.

The program will provide sufficient data on the radiation effects on the toughness properties of the irradiated materials to allow an evaluation of the toughness properties of this reactor vessel throughout its service life and determine safe operating pressure-temperature limits.

To assure the availability of adequate surveillance data for the Oconee reactor vessels, the program will be reviewed annually.

Early inspection of Reactor Coolant System piping elbows is considered desirable in order to reconfirm the integrity of the carbon steel base metal when explosively clad with sensitized stainless steel. If no degrada-tion is observed during the two annual inspections, surveillance requirements will revert to Section XI of the ASME Boiler and Pressure Vessel Code.

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