Information Notice 1980-27, Degradation of Reactor Coolant Pump Studs

Degradation of Reactor Coolant Pump Studs

ML031180384
Person / Time
Site:	05000000
Issue date:	06/11/1980
From:	NRC/OI
To:
References
IN-80-027, NUDOCS 8005050068
Download: ML031180384 (3)

SSINS No.: 6835 Accession No.:

UNITED STATES 8005050068 NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C. 20555 June 11, 1980

IE Information Notice No. 80-27

Description of Circumstances

On May 17, 1980, the NRC staff was informed by Omaha Public Power District (OPPO)

that severe corrosion damage was found on a number of closure studs in two of the

four Byron Jackson reactor coolant pumps at Fort Calhoun Unit 1 (PWR).

At the time, the reactor coolant system was undergoing a routine low pressure

leak test (180 psig) and visual inspection prior to plant restart after a four- month outage for refueling, pipe.support modifications and scheduled inservice

inspection. During the visual inspection, saturated and dripping insulation was

observed at one of the Byron Jackson reactor coolant pump flange regions. Upon

removal of the insulation, evidence of coolant leakage was found emanating from

the seating surfaces between the pump casing and the pump cover. Further

investigation of the three remaining pumps indicated similar coolant leakage past

both inner and outer flange gaskets on two of the three pumps. After complete

removal of the nonmetallic insulation, further visual observations revealed three

studs located side-by-side on one pump and three studs similarly located on the

other pump had significant corrosion wastage in the shank area next to the lower

thread section in the pump casing flange. Wastage of approximately 50% of the

original diameter of the stud giving them an "hour-glass" type appearance was

observed. The corroded studs were located In the vicinity of a component cooling

water line on both affected pumps but no direct correlation of this fact has been

established. Although not confirmed by metallurgical analysis, the cause of the

stud wastage is thought to be corrosive attack by hot boric acid from the primary

coolant.

The pump cover and casing for these pumps are constructed of ASTH A-351, Grade

CF8M stainless steel. Sealing between the cover and casing is achieved by two

concentric 304 stainless steel flexitallic gaskets. A leak-off line installed

between the gaskets on each pump was plugged and not instrumented. The leak-off

line was not in use and therefore, no indication of RCS leakage from the inner

seal was available. Each pump has 16 closure studs, consisting of ASTH A-193 Grade 87 low alloy steel, which are chrome plated in the thread area and

phosphate coated In the shank area. The studs are approximately 3-1/2 inches

in diameter and about 29 inches long.

No maintenance requiring removal of the pump casing studs had been performed on

the reactor coolant pumps since initial construction. The studs were covered

IE Information Notice No. 80-27 June 11, 1980 with block type Insulation since construction thereby limiting them from view.

The studs on two pumps were ultrasonically inspected in place in accordance with

the applicable ASME Section XI code rules. These ultrasonic examinations were

intended to locate cracks in bolting and were not effective in revealing wastage

of the studs.

The three affected pumps will be disassembled for further cleaning and inspection

of the studs and mating surfaces. Prior to reassembly, all studs exhibiting

significant corrosion will be replaced. All new, or acceptable used studs, will be subjected to u-ltrasonic, visual and magnetic particle examinations.

Installation of instrumentation for actively monitoring the leak-off lines

between the flexitallic gaskets is being performed. Future inservice inspec- tions, presently limited to ultrasonic examination, will be supplemented with

visual examination of the studs installed in the reactor coolant pumps. Replace- ment insulation will be in the form of a removable blanket to facilitate visual

examination.

The condition of the studs discovered at Ft. Calhoun raises concerns that such

severe corrosion, if undetected, could lead to stud failures which could result

in loss of integrity of the reactor coolant pressure boundary. The lack of

effectiveness of current ultrasonic examinations in revealing wastage emphasizes

the need for supplemental visual inspections and use of instrumented leak detec- tion systems to preclude unacceptable stud degradation going undetected. Licensees

should consider that the potential for undetected wastage of carbon steel bolting

by a similar mechanism could exist in other components such as valves.

This IE Information Notice is provided as an early notification of a signi- ficant matter that is still under review by the NRC staff. It is expected

that recipients will review the information for possible applicability to

their facilities.

No specific action or written response to this IE Information Notice is

required. If NRC evaluations so indicate, further licensee actions may be

required.

4

IE Information Notice No. 80-27 Enclosure

June 11, 1980

RECENTLY ISSUED

IE INFORMATION NOTICES

Information Subject Date Issued To

Notice No. Issued

80-26 Evaluation of Contractor 6/10/80 All Part 50 Licensees

QA Programs

80-25 Transportation of 5/30/80 Material Licensee in

Pyrophoric Uranium Priority/Categories II-A,

II-D, III-* and IV-DI;

Agreement State Licensees

in equivalent categories

80-24 Low Level Radioactive 5/30/80 All NRC and Agreement

Waste Burial Criteria State Licensees

80-23 Loss of Suction to 5/29/80 All power reactor

to Emergency Feedwater facilities with an

Pumps OL or CP

80-22 Breakdown In Contamination 5/28/80 All power reactor

Control Programs OLs and near term CPs

80-21 Anchorage and Support of 5/16/80 All power reactor

Safety-Related Electrical facilities with an

Equipment OL or CP

80-20 Loss of Decay Heat Removal 5/8/80 All light water reactor

Capability at Davis-Besse facilities holding

Unit 1 While in a Refueling power reactor OLs or CPs

Mode

80-19 NIOSH Recall of Recircu- 5/6/80 All holders of a power

lattng-Mode (Closed-Clrcuit) reactor OL, Research

Self-Contained Breathing Reactor License, Fuel

Apparatus (Rebreathers) Cycle Facility License

and Priority I Material

License

80-18 Possible Weapons Smuggling 5/5/80 All power reactor

Pouch facilities with an OL,

fuel fabrication and

processing facilities

and Materials Priority I

licensees (processors

and distributors)

I o -- _ . ._