ML19259C488

Submits Info Re Feedwater Lines to Steam Generator in Response to 790525 Request.Feedwater Pipe to Feedwater Nozzle Welds to Be Examined for Cracking During First Week of Jul 1979

ML19259C488
Person / Time
Site:	Point Beach
Issue date:	06/20/1979
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Stello V Office of Nuclear Reactor Regulation
References
NUDOCS 7906220181
Download: ML19259C488 (8)

Text

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O Wisconsin Electnc rw couer 231 W. Ml0HIGAN. F.C. BOX 204E MILWAUKEE. W: 53201 June 20, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. N'JCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention: Mr. Victor Stello, Jr. , Director Division of Operating Reactors Gentlemen:

DOCKET NOS. 50-266 AND 50-301 INFORMATION CONCERNING FEEDWATER LINES POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Your letter dated May 25, 1979, forwarded a request for information concerning the feedwater lines at the Point Beach Nuclear Plant. This request was prompted by the discovery of circumferential throughwall cracks in the

' heat affected zones of the feedwater nozzle to pipe welds at another nuclear plant. Enclosed herewith are our responses to this infonnation request.

In addition to the specific problems identified in your letter, we have become aware of similar indications at another facility via a telephone call from Mr. Boyd of the I&E Region III Staff on June 5,1979. As a result of this information, we are making preparations to examine the feedwater pipe to feedwater nozzle welds at the Point Beach Nuclear Plant. We hava tentatively scheduled these examinations for the week of July 1,1979. If the results of these examinations provide any indications of similar cracking at Point Beach Nuclear Plant, we shall, of course, take corrective action. In any case, we will notify the Nuclear Regulatory Commission pmmptly.

Very truly yours ,

h' a C. W. Fay, Dir ctor Nuclear Power Department Enclosures 2281 069 M wey hehenetkfM ffooi gs w-m 7906220/8/ .

ENCLOSURE 1 RESPONSES TO INFORMATION RE0' JESTS FEEDWATER LINE5 DESIGN

1. Provide as-built piping or isometric drawings of the feedwater line to steam generator sparger within containment. Show details of the design such as dimensions, pipe schedule, support type and locations, pipe restraints, and valve (s).

RESPONSE

Attached hereto are drawings M-60, M-61, and M-409 (Sheets 1 and 2) for PBNP Unit 1 and drawings M-2060, M-2061, and M-2409 for PBNP Unit 2.

2. Provide the results of any stress or fatigue analyses which were performed for this systen.

RESPONSE

Attached is a letter from Bechtel Power Corporation which summarizes the results of stress analyses on the 16-inch and 3-inch feedwater lines inside containment and concludes that the thermal and seismic stresses are well within code allowables.

FABRICATION HISTORY

1. Supply a list of the materials for the steam generator sparger, steam generator feedwater nozzles and feedwater piping within containment.

RESPONSE

Attached is the piping class sheet for the feedwater piping at PBNP.

The feedwater piping is designated EB-9 and is ASTM A-106B schedule 80 pipe. The feedwater nozzle is type SA-503*,1.2 steel. The feedwater ring and thermal sleeve are SA-106 Grade B.

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2. Provide the details of the welding process (es) used to make the nor:le-to-pipe, pipe to soarger and piping welds. Include details of welding such as preheat, joint configuration (include with or without backing ring), and post weld treatment, if any.

RESPONSE

Attached are drawings M-250 (Sheet 2 of 2), M-81, and the prep details of the feecwater no::le weld. These drawings provide details of the piping welds. Detail 32 of drawing M-81 is the detail of the concentric reducer which joins the feedwater pipe to th: steam generator feedwater nozzle. The feedwater no::le to feedwater piping welds were field fabricated without backing rings. No post weld heat treatments were a; plied. Since the thermal sleeve is a slip fit with the nozzle ID.

no welding details are required to describe the nozzle to feedring joint.

3. Provide the NDE performed during and after fabrication of the weld joints requested in question 2.

RESPONSE

The fabrication specification for the shop fabricated portions of the feedwater lines required a 100% radiographic examination of all butt welds. This requirement was extended to field welding. The radiographs of the feedwater pipe to steam generator feedwater nozzle welds have been recovered from storage and have been reviewed.

4. Provide the Code edition to which the feedwater piping system was fabricated.

RESPONSE

The feedwater piping system was fabricated to USASI B31.1 - July 1967 Edition.

2281 071

5. State the fracture toughness requirements, if any, for the feedwater piping systen.

RESPONSE

No special fracture toughness requirements are known to have been LDecified.

PRESERVICE/ INSERVICE INSPECTION AND OPERATING HISTORY

1. State whether the feedwater system welds received a prerarvice inspection in accordance with ASME B&PV Code,Section XI.

RESPONSE

The feedwater system welds did not receive a preservice inspection in accordance with ASME B&PV Code,Section XI as this code was not written at the time of the PBNP preservice inspection. The welds were, however, 100. adiographed and hydrostatically tested as normal installation procedures under the USAS B31.1.

2. Provide the extent of inservice inspection performed on the feedwater pipe to steam generator nozzle welds. Include the results of the examinations, any corrective actions taken and causes of any failures.

RESPONSE

To date, no inservice inspections of the feedwater to steam generator nozzle welds have been conducted; however, at least annual 800 psi leak tests of the steam generator have been perfonned. As discussed in the accompanying letter, an inspection of these welds is scheduled for the week of July 1,1979.

3. Provide the schedule and extent of inservice inspection for the feedwater system for the next inspection interval.

RESPONSE

The complete requirements for inservice inspections for the next inspec-tion interval have not yet been finalized. These rcquirements will 2281 072 include, however, subjecting the feedwater system to the periodic inservice hydrostatic pressure test specified by ASME B&PV,Section XI.

4 Provide any history of water hammer or vibration in the feedwater system and design changes and/or actions taken to prevent these occurrences.

RESDONSE:

The Point Beach Nuclear Plant has not experienced any feedwater line water hammer inside containment. This infomation has been provided in the past. Please see our letter dated November 1,1977. More recently, we provided to Mr. Dwight Christiansen of EG&G Associates our verbal responses to a list of 24 questions concerning steam generator water hamer. As you know, the NRC has contracted EG&G Associates to provide an evaluation of the feedline water hammer.

Our infomation was provided in April 1979.

5. Provide a description of feedwater chemistry controls and a sumary of chemistry data.

RESPONSE

The PBNP steam generators use an all volatile chemistry trea+ ment.

Information regarding steam generator feedwater chemistry control, including chemistry data, has been provided in the past with our letter dated October 2,1975.

2281 073 h # a.cl " 0 S Ch s.L /

. Bechtel Power Corporation Engineers-Constructors Fit'y Beale Stree' San Fraesce. Cahtornia Ma# Acoress- P C Sox 3965. San Fran 4sec. OA 94119 August 16, 1974 PJ . T.J. Rodgers Project Administrator .~

Wisconsin Electric Power Company 231 West Michigan Milwaukee, Wisconsin 53201 Attention: Mr. D.J. Bell

Subject:

Wisconsin Electric Power Company Job No. 10447 Point Beach Nuclear Plant Feedwater Check Valve Replacement Gentlemen:

In reply to Don Bell's July 19, 1974 and August 1, 1974 telephoned request, and supplementing our May 13, 1974 letter on the above subject, we have reviewed the seismic analyses of the 16 inch and 3 inch feedwater lines inside containment and find that the maximum thermal stress and seismic stress are 13,636 psi and 7,403 psi respectively. The allowable stresses according to ANSI B31.1 are 22,500 psi for thermal and 11,392 psi for seismic loading. Therefore, the thermal and seismic stresses are well k within code allowables.

In response to Doug Dill's August 16, 1974 question we confirm that no piping in these systems was cold sprung.

Very truly urs, H.E. Morris Project Engineer HEM /bjr 2281 074

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