ML19259A735

Forwards Info as Requested in NUREG-0312 Outlining Proposed Insp Plan Re Planned Feedwater Nozzle Insp Program for April 1979.W/insp Results & Summary of Repairs Made in Fall of 1976

ML19259A735
Person / Time
Site:	Cooper
Issue date:	12/27/1978
From:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
To:	Ippolito T Office of Nuclear Reactor Regulation
References
NUDOCS 7901100233
Download: ML19259A735 (4)

Text

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. - . COOPER NUCLEAR STATION

• P.o. BOX 98, B ROWNVILLE, NEDR ASKA 68 321 c

4 Nebraska Public Power o Distn. t 1EtEeno~E i o2> .2s-3 11 December 27, 1978 Mr. Thomas Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Planned Feedwater Nozzle Inspection Program, April 1979

References:

(1) NUREG-0312 " Interim Technical Report on BWR Feedwater and Control Rod Drive Return Line Nozzle Cracking" (2) NEDD-21821 " Boiling Water Reactor Feedwater Nozzle /Sparger Final Report", General Electric (3) BWR Service Information Letter, SIL No. 207

Dear Sir:

This letter is being submitted as requested by Reference (1) to outline the proposed inspection plans at Cooper Nuclear Station during the upcoming April 1979 refueling outage.

Prior to the installation of the interference fit feedwater spargers in October 1976, Cooper Nuclear Station had gone through 71 startup/ shutdown cycles. The original spargers were the " junction box" type with a flared thermal sleeve. The presently installed spargers are the forged-T design with thermal sleeves machined to provide a nominal .010 inch interference fit in the nozzle. In October 1976, the feedwater nozzles' inner blend radii were liquid penetrant (PT) examined per G.E. Field Disposition Instruction recommendations, and all crack indications were removed. Attachment 1 outlines the results of the inspections and re pairs . The largest crack indication, less than 3/16", was well below the average maximum crack depth for other plants as documented by Ref-e rence (2). The above can possibly be attributed to control of water chemistry and the minimal time of unheated feedwater flow or unstable feedwater flow. The water chemistry control at Cooper Nuclear Station is amplified by the fact that routine analysis of the reactor vessel water during normal operation indicate values of < 30 ppb Cl and < 0.15 pmho/cm conductivity with normal operating limits of < 200 ppb Cl and

< l.0 pmho/cm respectively. A review of the records verified that the water chemistry specifications delineated as limiting conditions for operation in Section 3.6.B of the Technical Specifications have not been exceeded during the time period from October 1976 to date.

7901100n3 qt C9

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Mr. Thomas Ippolito December 27, 1978 Page 2.

A contributing factor to maintenance of these low limits is the lack of a high chloride ion intrusion potential in the circulating water system.

Cooper Nuclear Station uses the Missouri River as a circulating water supply and the residual chloride ion concentration is low with respect to sea water which is used at many other facilities.

Since the refueling outage of fall 1976, Cooper Nuclear Station has logged thirteen (13) startup/ shutdown cycles, which is less than the previously designated 20. During this time period, the maximum number of hours of feedwater flow during startup at less than design temper-ature is 161, based on the time period from reactor critical to turbine generator on the line. Feedwater flow is routed through the feedwater heaters for all plant operating configurations; therefore, there has been no period of time when the feedwater flow was totally unheated.

Our system design does not have an automatic heater bypass feature, therefore, rapid changes in feedwater temperature do no* occur at Cooper Nuclear Station. There have been six (6) unscheduled scrams from power since the installation of the new forged-T design spargers. During a scram, reactor vessel water level is controlled by either the HPCI or RCIC. There have been six (6) HPCI and/or RCIC initiations and the time frame of each initiation has been in the magnitude of minutes. There has been no period of unstable feedwater flow.

The required liquid penetrant examination (PT) of the feedwater nozzles at the earlier of (a) every other scheduled refueling outage, or (b) at the scheduled refueling outage after 20 startup/ shutdown cycles after the last PT examination, stated in References (1) and (3), is based on a fracture mechanics analysis performed by General Electric in early 1976.

This analysis showed that if the crack tip ic left in the root of a grind cavity or if a new crack reinitiates quickly, it will only take about 20 startup/ shutdown cycles for the crack growth rate to behave as though the original crack had never been repaired. The probability of leaving a tip of a crack in a repair area on the CNS reactor vessel is considered remote.

Based on the above discussions, ALARA compliance, and the possibility of removing the feedwater nozzle cladding in the Spring Outage 1980, plans are being made to externally volumetrically inspect the feedwater nozzle safe-ends as required for the normal in-service inspection program in April 1979. In addition to the above, General Electric is submitting for our review, when in final form, their procedure MIUR 12-S752, Rev.

O, " Manual In-service Ultrasonic Examination of RPV Nozzle Bore and Blend Radius Section" for possible use at Cooper Nuclear Station. A

• Mr. Thomas Ippolito '

December 27, 1978 Page 3.

liquid penetrant examination of the feedaater nozzles' inner blend radii will not be performed. The above stated inspection plan for April 1979 is considered adequate. This submittal fulfills the requirement set forth in Reference 1.

If there are any questions concerning this submittal please contact me.

Sincerely, M b J. . Pilant Director of Licensing and Quality Assurance JMP:cg

Attachment:

Inspection Results and Summary of Repairs, Fall 1976

e ATTACHMENT 1

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4 . .

" INSPECTION RESULTS AND

SUMMARY

OF REPAIRS, FALL 1976" The following table (T,iile 1) presents the results of the initial dye penetrant inspection om the feedwater nozzle inner blend radius. Note that all of the cracks noted were removed with less than 1/8 inch of clad grinding except one. One of the grind-outs on the 225 nozzle was deeper than 1/8 inches but less than 3/16 inches.

TABLE 1: N0ZZLE CRACKS Number of Cracks Nozzle Azimuth Found on Initial P?

45 3 135 3

• 225o 3

• 315 12

• HPCI Injection Lines A dye penetrant test was performed on the cover plate-to-junction-box weld and both sparger-leg-to-junction-box welds on the old 315o sparger.

No cracks were found. No other dye penetrant tests were done on the spargers because of personnel exposure considerations.