Information Notice No. 86-19: Reactor Coolant Pump Shaft Failure at Crystal River
SSINS No: 6835
IN 86-19
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, DC 20555
March 21, 1986
Information Notice No. 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
185F) 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
hydraulic 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
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IN 86-19
March 21, 1986
Page 2 of 3
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 0375 in x 0200 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 0226 in and were noted from 180 to 200
around the circumference Subsequently, 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-3221, 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 shaft 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
Page 3 of 3
A metallurgical investigation is being conducted by Babcock and Wilcox
(B&W), Lynchburg, Virginia, to determine the cause of failure Region II
metallurgical 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
cracking (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 information 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
insufficient 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
Edward L Jordan Director
Division of Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: Jim Henderson, IE
(301) 492-9654
Nick Economos, RII
(404) 331-5580
Attachment: List of Recently Issued IE Information Notices
Page Last Reviewed/Updated Tuesday, March 09, 2021