Information Notice No. 94-72: Increased Control Rod Drop Time from Crud Buildup

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                            WASHINGTON, D.C.  20555

                                October 5, 1994

                               CRUD BUILDUP


All holders of operating licenses or construction permits for pressurized-
water reactors.


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to potential problems resulting in increase of
control rod drop times caused by the buildup of crud in control rod drive
mechanisms designed by the Babcock and Wilcox Company (B&W).  It is expected
that recipients will review the information for applicability to their
facilities and consider actions, as appropriate, to avoid similar problems. 
However, suggestions contained in this information notice are not NRC
requirements; therefore, no specific action or written response is required.  

Description of Circumstances

Oconee Nuclear Station Units 1 and 2

On April 29, 1993, the licensee for Oconee Nuclear Station Unit 2 conducted
control rod drop time tests and found that drop time for control rod Group 3
rod 8 (rod 3-8) exceeded the technical specification limit of 1.66 seconds. 
This rod had to be dropped several times during startup in the previous
refueling outage to achieve the drop time acceptance criteria.   Later, the
control rod drive mechanism (CRDM) was sent to the B&W facility for
disassembly and inspection.

A review of the rod drop data at Oconee Unit 1 revealed a similar problem for
rod 1-8 and rod 2-5.  On May 4, 1993, based on a licensee request, NRC issued
enforcement discretion to allow a drop time of 2 seconds for both rods for one
fuel cycle.  On August 25, 1993, following a plant trip, rod 2-5 had a drop
time of 1.938 seconds, and the time increased to 2.063 seconds when the rod
was drop tested on November 3, 1993.  The licensee safety evaluation,
following B&W inspection, concluded that the degraded drop time was the result
of crud deposits that caused sticking of four ball check valves in the thermal
barrier region of the CRDM and that the drop time would not exceed 3.0 seconds
if all four ball check valves were stuck.  Based on this analysis, a technical
specification amendment was issued on January 11, 1994, to allow control rod
insertion time of 3.0 seconds for rods 2-5 and 1-8 for the remainder of the
core operating cycle. 

9409280139.                                                            IN 94-72
                                                            October 5, 1994
                                                            Page 2 of 3

Three Mile Island Nuclear Station Unit 1

On October 14, 1993, the licensee for Three Mile Island Nuclear Station   
Unit 1 (TMI-1) performed control rod drop time testing.  There are seven
control rod groups; five groups have 8 rods, one group has 9 rods and one
group has 12 rods.  One rod in each of the rod groups 1, 3 and 4 initially
failed to comply with the technical specification requirement of 1.66 seconds. 
Drop times varied from 1.72 to 1.83 seconds.

As a followup to the October tests, the licensee conducted rod drop tests on
March 17, 1994.  Twelve rods failed to meet the technical specification
requirements.  The actual drop times ranged from 2.06 to 2.88 seconds.  With
the assistance of B&W, the licensee conducted an extensive study of the
probable causes.  Repetitive rod drops (ranging from 1 to 47) improved the
drop time for all rods to values well below the technical specification
requirements.  After further evaluation, the licensee and B&W concluded that
the excessive drop times had been caused by crud buildup in the thermal
barrier region.

After bringing all control rods within the drop time required in the technical
specification, plant operation resumed.  On March 29, 1994, the NRC issued a
confirmatory action letter documenting the following actions:  (1) withdrawal
of licensee request to raise the technical specification limit, (2) increase
of lithium concentration to minimize crud buildup, (3) performance of rod drop
time tests within three months of reactor startup, and (4) removal and
inspection of a CRDM if the drop time exceeded technical specification limits
in those tests.

The rod drop tests conducted in June 1994 after three months of operation
showed three rods with drop times ranging from 2.01 seconds to 2.2 seconds. 
The subsequent CRDM inspection revealed crud buildup in the thermal barrier
region and stuck ball check valves.


In B&W plants, the operation of at least one of four ball check valves,
located at the thermal barrier, is needed to attain rod drop times within
technical specification limits.  These check valves move up to allow the
reactor coolant to quickly flow into the upper CRDM area, replacing the void
created by the downward movement of the lead screw.  Reactor coolant also
rises through the narrow clearance between the control rod and the guide tube. 
During the surveillance and power control movements of the rod, the check
valves are not fully challenged since the coolant flows to the upper structure
at a much lower rate and in a much smaller quantity.

The inspections of CRDMs by both licensees identified crud deposits on the
thermal barrier ball check valves, and at the clearance between the outer
diameter of the lead screw and inside diameter of the thermal barrier bushing. 
The latter condition could result in significant increase in drop time if the
four check valves also are stuck.  Both licensees attribute the crud buildup
to be the combined result of 24-month fuel cycles, inadequate reactor coolant
pH control, and low clearance for the ball check valve movement..                                                            IN 94-72
                                                            October 5, 1994
                                                            Page 3 of 3

At TMI-1, the drop time degradation began after the plant started operating
with 24-month fuel cycles.  The 24-month fuel cycles require a higher boron
concentration in the coolant for reactivity control during the early part of
the cycle.  Lithium is used to buffer boric acid and to maintain pH in the
allowable range.  Compliance with a lithium concentration limit of 2.2 ppm
(recommended by the fuel vendor to avoid excessive fuel cladding corrosion
rates) resulted in pH values below 6.9 (7.0 is the neutral point between
acidity and alkalinity) during the long fuel cycle.  This condition may
explain increased crud precipitation and deposition.  The licensee has since
obtained approval from the vendor to operate at higher lithium concentrations
(allowing a pH of 6.9 and above).  

The corrosion products generally consisted of nickel-substituted spinels of
magnetite.  The crud solubility depends on pH level and the coolant
temperature.  A pH of 6.9 or greater should result in relatively less
formation of crud.

The TMI-1 licensee also observed that certain rods that were frequently
exercised and that were in the central part of the core did not have
significant crud buildup.  The licensee committed to exercise all rods through
10 percent of their length every 2 weeks in order to cycle the fluid trapped
in the upper CRDM area.  Such long strokes expose the lead screw area to the
reactor vessel environment and increase the coolant exchange with the upper
CRDM area.

The B&W thermal barrier design for TMI-1 had narrow tolerances for the ball
check valve movement.  New thermal barriers with larger tolerances were
installed at the four locations where slow rod drop time was observed in the
past.  The licensee is considering replacement of all thermal barriers and
working with the B&W owners group in exploring other options such as chemical
cleaning of the thermal barriers.

Primary chemistry control and narrow tolerances at the ball check valve appear
to be significant factors in the observed degradation of control rod drop
times.  Trending of "as found" rod drop times may lead to early detection of
potential problems.

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.

                                   Brian K. Grimes, Director
                                   Division of Project Support
                                   Office of Nuclear Reactor Regulation

Technical contacts:  Thomas Koshy, NRR            Howard Richings, NRR
                     (301) 504-1176               (301) 504-2888

Attachment:  List of Recently Issued NRC Information Notices

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