United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 92-86: Unexpected Restriction to Thermal Growth of Reactor Coolant Piping

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

                               December 24, 1992

                               OF REACTOR COOLANT PIPING 


All holders of operating licenses and construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to problems that may be caused by restricted
thermal growth of reactor coolant system (RCS) piping and components.  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

Description of Circumstances

On February 28, 1992, the licensee for the Wolf Creek Nuclear Generating
Station (Wolf Creek) was heating the RCS to prepare for a plant startup.  The
RCS was at 15.51 MPa [2250 psia] and 280.6�C [537�F] when plant personnel
heard a loud metallic noise and felt vibration in the reactor containment. 
Both the seismic and the loose parts and vibration monitoring (LPVM) system
alarms annunciated in the control room.  The licensee halted the heat-up and
initiated an incident investigation team to investigate and evaluate the cause
of the noise. 


The initial investigation by the licensee included (1) a system walkdown, 
(2) interviews of plant personnel and (3) a review of relevant plant data. 
Additionally, the licensee team began to search past events that may have
caused similar effects.  The walkdown inspection did not find any piping or
pipe support discrepancies attributable to the noise event.  However, the
search of past records found two events with similar effects.  These events
occurred on January 9, 1992, and on May 10, 1990.  Later investigation showed
that the event on May 10, 1990 was not related.  Based on the interviews and
the review of plant data, the licensee team concluded that the noise event
that occurred on February 28 was the result of a thermal expansion event
caused by restrictions to RCS thermal growth.  The licensee initiated a 


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                                                            December 24, 1992
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program to closely monitor RCS component behavior during plant heat-up.  The
licensee placed instrumentation to measure the pressure, temperature, and
displacement of portions of RCS piping and components on each of the four RCS
loops.  The licensee monitoring program addressed many possible causes for the
noise event, such as the upper and lower steam generator (SG) supports and the
reactor coolant pump (RCP) tie rods; however, as the investigation progressed,
the licensee team began to focus on the movement of the RCS crossover leg
saddle blocks in relation to their individual thrust block supports.  The
saddle blocks are attached to the crossover leg elbows beneath each SG and
each RCP (see Figures 1 and 2). 

On March 8, the licensee began a controlled plant heat-up to normal operating
temperature.  Personnel were stationed in the containment building to perform
system walkdowns and to record data.  The heat-up was halted at selected
plateaus to allow time for recording data.  As the heat-up progressed, the
licensee team found that, except for the saddle block on the SG side of
loop D, all of the RCS crossover leg saddles (both SG and RCP sides) came into
contact with their supports at approximately 260 to 282�C [500 to 540�F].  The
saddle block on the SG side in loop D made contact with its support at 
226.7�C [440�F].  On March 16, with the plant at 15.51 MPa [2250 psia] and
288.3�C [551�F], the licensee experienced another noise event.  Seismic alarms
and LPVM alarms annunciated in the control room and personnel in the
containment building heard a metallic noise and felt vibration.  

After reviewing the data from this heat-up, the licensee concluded that the
mechanism for the noise events was (1) the  restrained thermal growth of the
RCS piping caused by binding between the RCS crossover leg saddle blocks and
the support blocks and (2) the sudden release of energy that resulted when the
restrained thermal growth overcame the frictional resistance.  This conclusion
was supported by visual inspection of the surfaces of the shim plates
installed on the support blocks.   

After reaching this conclusion, the licensee initiated actions to increase the
clearance between the crossover leg saddle blocks and the support blocks. 
With the performance of a RCS integrity analysis, the support block shim
plates were removed and machined to allow for a minimum clearance between the
saddle blocks and the support blocks at normal operating temperature and

On March 23, the licensee conducted another heat-up of the RCS.  As the saddle
blocks approached the support blocks because of thermal growth in the piping,
the licensee temporarily removed shims from the support block surfaces to
ensure there would not be any contact between the saddles and the support
blocks.  The licensee observed that the crossover leg piping was free to
expand in a uniform manner with no unusual movement.  The licensee concluded .

                                                            IN 92-86
                                                            December 24, 1992
                                                            Page 3 of 4

that this further demonstrated that the cause of the previous noise events was
binding between the saddle blocks and the shims on the support blocks.  The 
licensee has noted no unusual noise or vibration during subsequent plant

As stated in the licensee team Report 92-01, the licensee has performed other
corrective actions and is considering implementing additional actions to
preclude recurrence. 

Although the root cause for the insufficient clearance was not determined, the
following potential contributors were identified:

     1.  Quality assurance did not verify the clearances for the saddle block
         to thrust block support during the initial hot functional testing.

     2.  No inspections were performed to monitor the clearances.

     3.  Housekeeping deficiencies from previous modifications in this area   
         may have resulted in debris build up in the gap.

     4.  The shim plates may have been deformed during previous heat ups.

Unanticipated restriction to thermal growth of the RCS may result in support
damage and levels of piping stress and fatigue that were not considered in the
original design of the system.  Excessive stress and fatigue can lead to
functional impairment of piping and components.  The staff addressed similar
concerns in NRC Information Notice (IN) 88-80, "Unexpected Piping Movement
Attributed to Thermal Stratification," October 7, 1988, and in IN 91-38,
"Thermal Stratification in Feedwater System Piping," June 13, 1991.

In its Report 92-01, the licensee team recommended, "perform required analysis
and remove shims and saddles from crossover legs if possible."  However, these
components can only be removed after the NRC has approved the piping for leak
before break criteria in accordance with the requirements of General Design
Criterion 4 of Appendix A to 10 CFR Part 50.  The licensee subsequent action
was to machine down and reinstall the shim plates in a manner that returned
the installation to its original design.

                                                            IN 92-86
                                                            December 24, 1992
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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 (NRR) project manager.

                                      ORIGINAL SIGNED BY

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

Technical contacts:  Cheng-Ih Wu, NRR
                     (301) 504-2764

                     Terence Chan, NRR
                     (301) 504-2169

1.  Figure 1.  Crossover-Leg Saddle Supports Overview
2.  Figure 2.  Crossover-Leg Saddle Supports
3.  List of Recently Issued NRC Information Notices.
Page Last Reviewed/Updated Tuesday, November 12, 2013