ML19318A586

Ro:On 800515,during Primary Sys Leakage Insp,Leakage Detected from Reactor Coolant Pump RC-3C.Caused by Corrosion Damage to Closure Studs.Studs W/Significant Corrosion Will Be Replaced & Metallurgical Exams Performed

ML19318A586
Person / Time
Site:	Fort Calhoun
Issue date:	05/22/1980
From:	Stevens S OMAHA PUBLIC POWER DISTRICT
To:
References
FC-494-80, NUDOCS 8006230414
Download: ML19318A586 (3)

Text

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1623 MA4NEY e OMAHA, NESMASMA 68102 s TELEPHONE 536 4000 AREA CCDE 402 May 22, 1980 FC-494-80 Mr. K. V. Seyfrit U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011

Dear Mr. Seyfrit:

This special maintenance report concerning reactor coolant pump leakage at the Fort Calhoun Station, is being submitted in accordance with Section 5.9.3.d of the station Technical Specifications. This document is intended only as an interim report. A final report will be submitted to the Commission following completion of maintenance activities and a discussion of the leakage problem with the nuclear steam supply system vendor, F.ombustion Engineering; the pump manufacturer, Byron Jackson and the District's inservice inspection agency, Southwest Research Institute.

On the evening of May 15, 1980, plant operating personnel were per-forming a cold pressure test of the reactor coolant system prior to plant startup. The primary system was pressurized to approximately 180 psig using one low pressure safety injection pump. A routine primary system leakage inspection by operations personnel revealed leakage coming from reactor coolant pump RC-3C. At this time the low pressure safety injection pump was secured. On the morning of May 16, 1980, the system was again pressurized using the low pressure safety injection pump. An inspection by maintenance personnel revealed leakage from the shaft seal on pump RC-3C. This seal was subsequently tightened which reduced but did not terminate the leakage. Further investigation re-vealed leakage emanating from the gasketed surface between the pump casing and the pump cover. At this time, technical assistance was requested from the pump manufacturer, Byron Jackson. On the afternoon of May 16, tho other three reactor coolant pumps RC-3A, RC-3B and RC-3D were inspected. Leakage was found to be coming from RC-3A and RC-3B.

In addition, it was noted that corrosion damage had occurred to a number of the closure studs on pumps RC-3A and RC-3B. No stud damage was evident on RC-3D.

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lb:. K. V. Seyfrit Page Two FC-494-80 The pump cover and the pump casing for the reactor coolant pumps are constructed of ASIM A-351, Grade CF8M stainless steel. Sealing between the pump cover and the pump casing is accomplished by means of two concentric 304 stainless steel Flexicallic gaskets. A leak-off line is included between the two gaskets and is presently blanked off with a pipe plug. To aid in determining the source of the leakage, the leak-off line for each pump was unplugged and the system was repressurized to approximately 180 psig. Pump RC-3A showed slight leakage through the leak-off line. Pumps RC-3B and RC-3C showed considerably more leakage.

No leakage was noted for RC-3D.

As of May 21, 1980, the closure studs have been removed from pumps RC-3B and RC-3C. Each pump makes use of 16 closure studs constructed of ASIH A-193, Grade B7 carbon steel, chrome plated in the thread area, and phosphate coated in the shank area. The studs are approximately 3-1/2" in diameter and approxi=ately 29" in length. During assembly the studs are hydraulically tensioned to approximately 23,500 psi. The residual stress in the studs, when they are held by the nuts, is approximately

'* 15,500 psi.

Inspection of the closure studs for RC-3B revealed three studs with substanical corrosion in the shank area. Nominal diameter in this area was 1 to 1-1/2" minimum. In addition, one stud was substantially corroded, although not to the extent of the previous three. The closure studs for RC-3C, although exhibiting some corrosion in several cases, were in re-latively good condition. At present, three studs have been identified on RC-3A which exhibit corrosion similar to the worst three on RC-3B.

However, a final determination of stud condition on RC-3A cannot be made until the studs are removed. The studs on RC-3D have been found to be acceptible for continued service. It should be noted that although corrosion had significantly reduced the stress in a number of the studs on RC-3B, full hydraulic tension pressure was required for removal of all studs. All studs on RC-3B and RC-3C were removed intact.

The pump cover has been removed from RC-3C. The gasketed surfaces on the pump cover and the pump casing are in good condition with no evidence of steam cutting. No damage is obvious to the inner gasket on RC-3C; the outer gasket exhibits some signs of deterioration of the asbestos filler, although no complete voids were noted in this gasket.

The pump cover has not yet been removed from RC-3B. Disassembly of RC-3A is still in the early stages.

Prior to reassembly of these pumps, all studs exhibiting significant corrosion will be replaced. All new or acceptable used studs will be subjected to examination by ultrasonic, visual and magnetic particle techniques prior to installation. Any studs which contain flaws greater than allowed by ASME Section XI will not be used. Consideration is being given to installing some type of monitoring device on the leak-off lines between the primary and secondary gaskets. In order to preclude a recurrence of the corrosion noted in this inspection, future inservice inspection, presently limited to ultrasonic examination, will be

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Mr. K. V. Seyfrit Page Three FC-494-80 supplemented with visual examination of the studs installed in the reactor coolant pumps. It is notable that a straight beam ultrasonic examination, was performed on the RC-3A and RC-3B studs during the 1980 refueling outage. Sufficient portions of the studs can be viewed with the pump assembled to allow discovery of any corrosion similar to that which has occurred thus far. It is emphasized that even with the damaged studs discovered during this inspection, the one GPM reactor coolant system leak rate limit discussed in the station Technical Specifications was never exceeded due to leakage from the reactor coolant pump casing to pump cover joint.

In order to assist with the investigation into the cause of the stud damage, at least one stud will be submitted for metallurgical examination.

Sincerely i f j

hik t. \ U I)br ender C.\Stevens nager Fort Calhoun Station SCS/JJF:jbk cc: Director, Office of Management Information and Program Control U. S. Nuclear Regulatory Commission Washington, DC 20555 (3)

Director, Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, DC 20555 (30)

Mr. W. C. Jones Mr. R. L. Andrews SARC Chairman PRC Chairman

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