Information Notice 1986-19, Reactor Coolant Pump Shaft Failure at Crystal River

Reactor Coolant Pump Shaft Failure at Crystal River

ML031220620
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill
Issue date:	03/21/1986
From:	Jordan E NRC/IE
To:
References
IN-86-019, NUDOCS 8603210162
Download: ML031220620 (4)

LIS ORI NAL UNITED STATES

SSINS No.: 6835 IN 86-19 NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C. 20555 March 21, 1986 IE INFORMATION NOTICE NO. 86-19: REACTOR COOLANT PUMP SHAFT FAILURE AT

CRYSTAL RIVER

Addressees

All nuclear power reactor facilities holding an operating license (OL) or a

construction permit (CP).

Purpose

This information notice provides notification of failure of reactor coolant

pump shafts manufactured by Byron-Jackson (BJ) Company. It is expected that

recipients will review the information for applicability to their facilities

and consider actions, if appropriate, to detect a similar problem at their

facilities. However, suggestions contained in this notice do not constitute

NRC requirements; therefore, no specific action or written response is required.

NRC is continuing to obtain and evaluate pertinent information. If specific

actions are determined to be required by NRC, an additional communication will

be issued.

Description of Circumstances

On January 1, 1986, the Crystal River Unit 3 (CR-3) reactor tripped because of

low flow in reactor coolant system loop A. Just before the reactor trip

occurred, the reactor coolant pump (RCP) motor frame vibration monitor showed

high vibration. This was followed by a RCP thrust bearing upper shoes high

temperature alarm (which activates at temperatures greater than 1850 F). The

licensee manually tripped the RCP motor, which resulted in a reactor trip. On

January 4, 1986, the licensee entered the reactor building to inspect the RCP

and found no evidence to indicate that the pump had sustained damage.

On January 6, 1986, the licensee began preliminary troubleshooting on loop A

RCP motor shaft, coupling, and seal cover. Following these checks, the pump

shaft was uncoupled from the motor and an unsuccessful attempt was made to

rotate the pump shaft. Also, various attempts to raise the shaft with hydrau- lic pressure failed.

On January 14, 1986, an ultrasonic examination of the shaft in place identified

a major reflector at a distance of approximately 50 in. from the top. The

8603210162

IN 86-19 March 21, 1986 reflector was evident throughout the entire circumference of the shaft. Finally, on January 15, 1986, after having lifted the RCP motor and removed other

interferences, the licensee removed the upper shaft remnant from the pump.

Preliminary visual inspection of the removed shaft section showed that the

fracture occurred in the location of a machined, flat-bottom circumferential

groove measuring approximately 0.375 in. x 0.200 in. This groove is located

just below the multigroove section on the shaft that is identified as a thermal

barrier.

Other operating units with essentially identical pumps are Davis-Besse and

Arkansas Unit 1. Both these sites have been notified of the findings at Crystal

River. At Davis-Besse, currently in an extended outage, the licensee performed

similar UT examinations and reports confirmed cracking in one shaft, with

probable cracks in the other three. The licensee has ordered four replacement

shafts. Arkansas Unit 1 has also ordered four replacement shafts (about 12 week

delivery) and plans to continue in operation pending delivery. Midland Units 1 and 2 have the same pumps, but work is currently suspended on these partially

constructed facilities.

Discussions with cognizant Crystal River personnel disclosed that currently the

groove in question serves no functional purpose on the shaft assembly. It is

NRC's understanding that this groove was intended for a split ring that was

deleted by a design change after the groove had been machined in the shaft.

All four pumps at CR-3 have shafts with this machined-in groove. Following

verification of the shaft's failure, the licensee conducted an ultrasonic

examination of the three remaining RC pump shafts and determined that the shaft

in RCP B exhibited circumferential crack indications in the same location as

RCP A. The indications exceeded minimum calibration notch depth dimensions of

0.226 in. and were noted from 1800 to 2000 around the circumference. Subse- quently, PT confirmed the crack in the pump B shaft. Ultrasonic examination of

the C and D pump shafts showed indications of cracks. As a result, all four

shafts are being replaced.

The failed shaft(s) were made from precipitation hardening stainless steel

material produced to ASTM Specification A461-65 Grade 660 requirements and

inspected per ASME Section III (68,S69), paragraph N-322.1, N-627.

Currently, the licensee attributes the shaft failure on pump A at Crystal River

to residual fabrication stresses coupled with thermal stresses from cool seal

water injection. The pump B shatt crack is being attributed to local assembly

weld stresses compounded by thermal stresses. The shaft material is difficult

to weld successfully.

IN 86-19 March 21, 1986 A metallurgical investigation is being conducted by Babcock and Wilcox (B&W),

Lynchburg, Virginia, to determine the cause of failure. Region II metallurgi- cal staff is following up this investigation. To date, the only information

from this investigation is that in pump A all four socket head capscrews that

join the shaft and impeller were found to be broken. Two alignment pins were

not broken. Further information on shaft B is not yet available, other than

the cap screws on pump B assembly were either cracked or broken.

The cap screw failures are attributed to intergranular stress corrosion crack- ing (IGSCC).

A similar event involving the capscrews in a BJ pump at the Palisades nuclear

plant is discussed in Information Notice 85-03 and Supplement 1 to that infor- mation notice. The pumps at Palisades are a different size from those at

Crystal River, but the designs are apparently similar.

At Palisades, the shaft did not fail but separated from the impeller. The

shaft is normally secured by eight sockethead capscrews and four alignment

pins. All eight capscrews and two of the four alignment pins were broken. The

two other pins were distorted. The cause of failure was stated to be insuffi- cient preload on the capscrews caused by rough threads, which resulted in the

prescribed tightening torque not achieving the desired preload.

No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the Regional

Administrator of the appropriate regional office or this office.

X~w~ar t j

Divisio f Emergency Preparedness

and Engineering Response

Office of Inspection and Enforcement

Technical Contacts: Jim Henderson, IE

(301) 492-9654 Nick Economos, RII

(404) 331-5580

Attachment: List of Recently Issued IE Information Notices

Attachment 1 IN 86-19 March 21, 1986 LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Information Date of

Notice No. Subject Issue Issued to

86-18 NRC On-Scene Response During 3/26/86 All power reactor

A Major Emergency facilities holding

an OL or CP

86-17 Update Of Failure Of Auto- 3/24/86 All power reactor

matic Sprinkler System Valves facilities holding

To Operate an OL or CP

86-16 Failures To Identify Contain- 3/11/86 All power reactor

ment Leakage Due To Inadequate facilities holding

Local Testing Of BWR Vacuum an OL or CP

Relief System Valves

86-15 Loss Of Offsite Power Caused 3/10/86 All power reactor

By Problems In Fiber Optics facilities holding

Systems an OL or CP

86-14 PWR Auxiliary Feedwater Pump 3/10/86 All power reactor

Turbine Control Problems facilities holding

an OL or CP

86-13 Standby Liquid Control 2/21/86 All BWR facilities

System Squib Valves Failure holding an OL or CP

To Fire

86-12 Target Rock Two-Stage SRV 2/25/86 All power reactor

Setpoint Drift facilities hording

an OL or CP

86-11 Inadequate Service Water 2/25/86 All power reactor

Protection Against Core Melt facilities holding

Frequency an OL or CP

84-69 Operation Of Emergency Diesel 2/24/86 All power reactor

Sup. 1 Generators facilities holding

an OL or CP

86-10 Safety Parameter Display 2/13/86 All power reactor

System Malfunctions facilities holding

an OL or CP

OL = Operating License

CP = Construction Permit