ML20072P860

Proposed TS 15.4.2, ISI of Safety Class Components, Incorporating Use of Acceptance Criteria Described in WCAP-14157 to Allow Sleeved Tubes W/Certain Upper Sleeve Joint Parent Tube Indications to Remain in Svc

ML20072P860
Person / Time
Site:	Point Beach
Issue date:	08/26/1994
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML19353C445	List: ML20072P860
References
NUDOCS 9409080260
Download: ML20072P860 (2)

Text

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Defect is an imperfection of such severity that it exceeds the minimum acceptable tube wall thickness of 50%. A tube containing a defect is defective.

Pluaaina Limit is the imperfectica depth beyond which the tube must be removed from service or repaired, because the tube may become defective prior to the next scheduled inspection. The plugging limit is 40% of the nominal tube wall thickness.

6. Corrective Measures All tubes that leak or have degradation exceeding the plugging limit shall be plugged or repaired by a process such as sleeving

• prior to return to power from a refueling or inservice inspection condition. Sleeved tubes having sleeve degradation exceeding 40% of the nominal sleeve wall thickness -

shall be_pl.ugged.

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7. Reports (a) After each inservice examination, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission as soon as practicable.

(b) The complete results of the steam generator tube inservice inspection l shall be included in the Annual Results and Data Report for the l period in which the inspection was completed.

Reports shall include:

1. Number and extent of tubes inspected.

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2. Location and percent of all thickness penetration for each I indication.  !

3. Identification of tubes plugged or repaired.

(c) Reports required by Table 15.4.2 Steam Generator Tube Inspection shall provide the information required by Specification 15.4.2.A.7(b) and a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence. The report shall be submitted to the Commission prior to resumption of plant operation.

Brazed joints shall not be employed. Tubes previously subject to explosive plugging shall not be sleeved.

Unit 1 - Amendment No. 15.4.2-4 Unit 2 - Amendment P 9409080260 940826 r PDR ADOCK 05000266 P PDR

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B. In-Service Inspection of Safety Class Components Other than Steam Generator Tubes

1. Inservice inspection of ASME Code Cla:;s 1, Class 2 and Class 3 componente shall be performed 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) modified by Section 50.55a(b), except where specific written relief is granted by the NRC, pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

2. Containment isolation valves will be tested i.1 accordance with Technical Specification 15.4.4 instead of Section IWV-3420, Valve Leak Rate Test.

ILalh The steam generator tube inspection requirements are based on the guidance given in NRC Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes." ASME Section XI Appendix IV is being used for defining the basic requirements or the inspection method. However, at the present time, changes and improvements in steam generator eddy current inspection are occurring faster than the code can be revised. Thus, in order to ensure that the best possible exam of the tubing and/or sleeves is being done, the technique utilized will, in general, be the latest industry-accepted technique. This means that complete word-for-word compliance with Appendix IV may not be possible. However, the basic requirements and intent will be met to the extent practical. WCAPsfM57 E"IsehninsTT EVallisti 6iE6 f!HjbMdi.EspassI6~d V6'inth(HEJ)fSlesisdiTdbsi?containing1Indibstions?Withinuhe?phsr?J61'ntiZon#F U

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As stated in 15.4.2.B.1, safety class components, other than the steam gener-ator tubing, will be inspected in accordance with ASME Section XI. The code edition / addenda utilized for the inspection interval will be as defined in 10 CFR 50. The same code is utilized for both Unit 1 and Unit 2. Safety related components are classified as safety Class 1, 2, or 3. The code boundaries are defined based upon the following documents:

(a) Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants."

(b) American National Standard N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."

(c) Point Beach Nuclear Plant Units 1 & 2 Final Safety Analysis Report.

Code classified components are tabulated showing each specific examination area l and the examination requirements in an inspection interval long-term plan.

This plan is completely revised for each ten-ycar inspection interval.

l Unit 1 - Amendment No. 15.4.2-5 l Unit 2 - Amendment No.

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