Bulletin 85-03: Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings

                                                  SSINS No.: 6820  
                                                  OMB No.: 3150-0011  
                                                  Expiration Date: 08/01/88 
                                                  IEB 85-03 

                                UNITED STATES
                            WASHINGTON, DC 20555

                              November 15, 1985



All holders of nuclear power reactor operating licenses (OLs) or 
construction permits (CPs) for action. 


The purpose of this bulletin is to request licensees to develop and 
implement a program to ensure that switch settings on certain safety-related
motor-operated valves are selected, set and maintained correctly to 
accommodate the maximum differential pressures expected on these valves 
during both normal and abnormal events within the design basis. 

Description of Circumstances: 

There have been two recent events, and a number of earlier events, during 
which motor-operated valves failed on demand, in a common mode, due to 
improper switch settings. 

Event 1, Davis-Besse Plant - On June 9, 1985, the Davis-Besse Plant 
experienced a complete loss of main and auxiliary feedwater. This event was 
described previously in IE Information Notice No. 85-50, "Complete Loss of 
Main and Auxiliary Feedwater at a PWR Designed by Babcock & Wilcox," and in 
NUREG-1154,  "Loss of Main and Auxiliary Feedwater Event at the Davis-Besse 
Plant on June 9, 1985.'" Normally open, Limitorque motor-operated auxiliary 
feedwater (AFW) gate valves failed to reopen on either an automatic or 
manual signal from the main control room after they were inadvertently 
closed during the event. While other failures also occurred in the AFW 
system, the failure of these two valves was itself enough to prevent AFW 
from reaching either steam generator. During the recovery from this event, 
the valves were opened with the handwheels. 

The results of licensee troubleshooting activities after the event led to 
the conclusion that the setting for the torque switch bypass limit (torque 
bypass) switch in each valve's control circuit had not been set to remain 
closed long enough to provide the necessary bypass function on valve opening
with differential pressure conditions across the valve. During the event, 
the valves experienced a high differential pressure after closing. The 
torque bypass 


                                                       IEB 85-03  
                                                       November 15, 1985 
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switch on both valves was improperly set, causing the torque switches to 
become operable prematurely. This condition stopped valve travel before the 
valve discs were fully off their seats. 

The torque bypass switches were set to drop out after the valves opened to 
5% full-stroke. The 5% full-stroke setting was based on a number of 
handwheel turns. In a 10 CFR 21 report, submitted subsequent to the event, 
Toledo Edison Company identified two reasons why the torque bypass switch 
settings were not adequate: (1) the 5% full-stroke settings were not 
adequate for unseating the AFW system discharge valves with large 
differential pressures across the valves and (2) the procedure for setting 
this switch was defective in that the 5% full-stroke was not specified to be 
in addition to the handwheel turns required to take up the motor-operator 
coast and backlash. The torque bypass switch setting errors were revealed 
only when high differential pressure conditions across the valves caused 
higher loadings. The valve failures were reproduced during tests performed 
by the licensee with differential pressures applied across the valves. 
During the tests, the valves operated properly when low differential 
pressures were applied across them, but failed to open when high 
differential pressures were applied. The valves were instrumented during 
these tests to obtain signature traces of critical parameters. 

Event 2, Sequoyah Plant Unit 2 - An event involving partial loss of main 
feedwater occurred on May 2, 1985, at Sequoyah Nuclear Plant Unit 2 while in
Mode 2 and returning to power after a reactor trip. Feedwater was being 
supplied through the main feedwater (MFW) system isolation valve bypass 
lines. Operators attempted to open the MFW system isolation valves to supply
water to the steam generators; however, two of the four MFW isolation valves
would not open. The startup was discontinued and the unit was returned to 
hot shutdown. 

During examination to determine the reason for the valve failures, the 
licensee discovered that both valve stems had sheared from their discs. The 
discs were found in the closed positioning within the valve seats. The stems
had suffered fracture failures through approximately three-quarters of the 
diameter of the shafts, in addition to stress failures of the remaining 
quarter. The Limitorque motor-operators on the valves use limit switches to 
control valve motion in the open direction. These MFW system isolation 
valves are large (18 inch diameter), fast acting (154 inches per minute 
travel speed) valves. Because of the high speed of these valves and the 
large mass of the discs, the selection of the limit switch setpoint needs to 
account for the large momentum of the disc and its continued motion after 
the limit switch deenergizes the valve motor-operator. The set point was not 
correctly established and the disc impacted the backseat during opening. The 
failure mechanism of these valves was identified by the licensee to be 
impact loading of the stem on the opening stroke as a result of the disc 
impacting the backseat, combined with a stress failure of the remaining 
portion of the stem on the opening stroke. Main feedwater valves are not 
included in the actions requested by this bulletin. The NRC is, however, 
continuing to evaluate the Sequoyah event. 

NRC Field Evaluation - As a part of the resolution of Generic Issue II.E.6 
"In-Situ Testing of Valves," the NRC contracted with the Oak Ridge National 
Laboratory in 1984 to perform a limited study to determine the effectiveness
of signature tracing techniques in determining the operational readiness of 

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                                                       November 15, 1985 
                                                       Page 3 of 6 

safety-related motor-operated valves. It was hoped also that this study 
could provide some insight as to current conditions of valve switch settings 
at nuclear power plants. Signature traces of motor current, torque and limit
switch actuations and axial motion of the worm gear (an indication of 
operator torque) were obtained from 36 motor-operated valves at 4 nuclear 
plant sites. Although the formal technical report [NUREG/CR-4380 "Evaluation
of the Motor-Operated Valve Analysis and Test System (MOVATS) to Detect 
Degradation, Incorrect Adjustments, and Other Abnormalities in 
Motor-Operated Valves"] has not been issued, the current draft of the report 
indicates that (1) this inspection method can be used to improve current 
ASME methods and (2) there were abnormalities with nearly every valve 

Table 1 contains a summary of the study's findings with respect to switch 
setting abnormalities. Of particular interest with respect to the events 
described above is the finding that 75% of the valves had improperly set 
torque bypass switches (56% of the valves had the close-to-open torque 
bypass switch set so that it was opening before the valve fully unseated) 
and 8% of the valves were unintentionally backseating. The abnormalities in 
Table 1 have not been fully evaluated at this time, and they should not be 
interpreted to mean that any abnormality resulted in an inoperable valve. 


The NRC has previously identified common mode failures, on demand, of 

IE Circular No. 77-01, "Malfunction of Limitorque Valve Operators," reported
that on October 28, 1976, two motor-operated (Limitorque) valves located 
between the refueling water storage tank and the charging pump suction at 
the Trojan Nuclear Plant failed to open in response to a spurious safety 
injection (SI) signal. The malfunction in both valves resulted from the 
torque switch in the opening circuit becoming activated before the valves 
were fully off their seats. The valves also were equipped with a torque 
bypass switch. Each of the valves that malfunctioned was found to have its 
torque bypass switch adjusted such that it allowed the torque switch to be 
operable in the circuit before the valve was moved from its seat. The 
licensee's investigation revealed that in each case the valve had been 
manually closed hard on its seat following a maintenance operation. 
Examination by the licensee revealed similar improper adjustments of the 
torque bypass switches on several other motor-operated valves in 
safety-related systems. 

IE Information Notice No. 81-31, "Failure of Safety Injection Valves To 
Operate Against Differential Pressure," reported on September 3, 1981, that 
both trains of the San Onofre Unit 1 safety injection system were found to 
be inoperable when challenged to operate against differential pressure. 
Improperly set switches were the principal cause of these failures. There 
were no adverse consequences in this particular event because there was no 
accident that required safety injection. The reactor pressure remained above
the safety injection pump's shutoff head; therefore, no actual injection of 
later would have occurred if the valves had opened. However, had reactor 
pressure decreased and actual injection been required, injection flow would 
not have been automatically available as designed. These valves had been 
regularly tested at each  

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                                                       November 15, 1985 
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refueling outage, but the tests were not required to be performed with 
differential pressure across these valves. 

Florida Power Corporation reported an event at Crystal River Unit 3 in LER 
77-9. During plant cooldown with the unit in hot shutdown, decay heat 
removal valves in the decay heat removal pump suction would not open with 
remote actuation. These failures were caused by pressure acting on the gate 
valve discs. The valves were opened manually with the handwheels. The torque
switches were reset. 

In addition to common mode valve failures on demand, there have been 
numerous common mode failures discovered during testing or as a result of 
investigating a single failure. NUREG/CR-2270, "Common Cause Fault Rates for
Valves," February 1983, contains reports of 99 common cause valve fault 
events from 1976 through 1980. 

The NRC has previously identified other problems with motor-operated valve 
switches in Bulletin No. 72-3, "Limitorque Valve Operator Failures"; IE 
Information Notice No. 79-03, "Limitorque Valve Geared Limit Switch 
Lubricant; Circular No. 81-13, "Torque Switch Electrical Bypass Circuit for 
Safeguard Service Valve Motors"; Information Notice No. 82-10, "Following Up
Symptomatic Repairs To Assure Resolution of the Problem"; and Information 
Notice No.84-10, "Motor-operated Valve Torque Switches Set Below the 
Manufacturer's Recommended Value."  

The failure and potential failure of Westinghouse Electro-Mechanical 
Division motor-operated gate valves to close are discussed in IE Bulletin 
No. 81-02 and IE Bulletin No. 81-02, Supplement 1, "Failure of Gate Type 
Valves To Close Against Differential Pressure." 

Copies of the above referenced NRC bulletins, circulars and information 
notices can be obtained from your local public document room. 

Actions for All Holders of Operating Licenses or Construction Permits: 

For motor-operated valves in the high pressure coolant injection/core spray 
and emergency feedwater systems (RCIC for BWRs) that are required to be 
tested for operational readiness in accordance with 10 CFR 50.55a(g), 
develop and implement a program to ensure that valve operator switches are 
selected, set and maintained properly. This should include the following 

a.   Review and document the design basis for the operation of each valve. 
     This documentation should include the maximum differential pressure 
     expected during both opening and closing the valve for both normal and 
     abnormal events to the extent that these valve operations and events 
     are included in the existing, approved design basis, (i.e., the design 
     basis documented in pertinent licensee submittals such as FSAR analyses 
     and fully-approved operating and emergency procedures, etc). When 
     determining the maximum differential pressure, those single equipment 
     failures and inadvertent equipment operations (such as inadvertent 
     valve closures or openings) that are within the plant design basis 
     should be assumed. 

                                                       IEB 85-03  
                                                       November 15, 1985 
                                                       Page 5 of 6 

b.   Using the results from item a above, establish the correct switch 
     settings. This shall include a program to review and revise, as 
     necessary, the methods for selecting and setting all switches (i.e., 
     torque, torque bypass, position limit, overload) for each valve 
     operation (opening and closing). 

     If the licensee determines that a valve is inoperable, the licensee 
     shall also make an appropriate justification for continued operation in
     accordance with the applicable technical specification. 

c.   Individual valve settings shall be changed, as appropriate, to those 
     established in item b, above. Whether the valve setting is changed or 
     not, the valve will be demonstrated to be operable by testing the valve
     at the maximum differential pressure determined in item a above with 
     the exception that testing motor-operated valves under conditions 
     simulating a break in the line containing the valve is not required. 
     Otherwise, justification should be provided for any cases where testing 
     with the maximum differential pressure cannot practicably be performed. 
     This justification should include the alternative to maximum 
     differential pressure testing which will be used to verify the correct 

     Note: This bulletin is not intended to establish a requirement for 
     valve testing for the condition simulating a break in the line 
     containing the valve. However, to the extent that such valve operation 
     is relied upon in the design basis, a break in the line containing the 
     valve should be considered in the analyses prescribed in items a and b 
     above. The resulting switch settings for pipe break conditions should 
     be verified, to the extent practical, by the same methods that would be 
     used to verify other settings (if any) that are not tested at the 
     maximum differential pressure. 

     Each valve shall be stroke tested, to the extent practical, to 
     verify,that the settings defined in item b above have been properly 
     implemented even if testing with differential pressure can not be 

d.   Prepare or revise procedures to ensure that correct switch settings are
     determined and maintained throughout the life of the plant. Ensure
     that applicable industry recommendations are considered in the 
     preparation of these procedures. 

e.   Within 180 days of the date of this bulletin, submit a written report 
     to the NRC that: (1) reports the results of item a and (2) contains the
     program to accomplish items b through d above including a schedule for 
     completion of these items. 

This item s intended to be completely consistent with action item 3.2, 
     "Post-Maintenance Testing (All Other Safety-Related Components)," of 
     Generic Letter 83-28, "Required Actions Based on Generic Implications 
     of Salem ATWS Events." These procedures should include provisions to 
     monitor valve performance to ensure the switch settings are correct. 
     This is particularly important if the torque or torque bypass switch 
     setting has been significantly raised above that required.  

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                                                       November 15, 1985 
                                                       Page 6 of 6 

     1.   For plants with an OL, the schedule shall ensure that these items 
          are completed as soon as practical and within two years from the 
          date of this bulletin. 

     2.   For plants with a CP, this schedule shall ensure that these items 
          are completed before the scheduled date for OL issuance or within 
          two years from the date of this bulletin, whichever is later. 

f.   Provide a written report on completion of the above program. This 
     report should provide (1) a verification of completion of the requested 
     program, (2) a summary of the findings as to valve operability prior to 
     any adjustments as a result of this bulletin, and (3) a summary of data 
     in accordance with Table 2, Suggested Data Summary Format. The NRC 
     staff intends to use this data to assist in the resolution of Generic 
     Issue II.E.6.1. This report shall be submitted to the NRC within 60 
     days of completion of the program. Table 2 should be expanded, if 
     appropriate, to include a summary of all data required to evaluate the 
     response to this bulletin. 

The written reports shall be submitted to the appropriate Regional 
Administrator under oath or affirmation under provisions of Section 182a, 
Atomic Energy Act of 1954, as amended. Also, the original copy of the cover 
letters and a copy of the reports shall be transmitted to the U.S. Nuclear 
Regulatory Commission, Document Control Desk, Washington, DC 20555 for 
reproduction and distribution. 

This request for information was approved by the Office of Management and 
Budget under a blanket clearance number 3150-0011. Comments on burden and 
duplication may be directed to the Office of Management and Budget, Reports 
Management, Room 3208, New Executive Office Building, Washington, DC 20503. 

Although no specific request or requirement is intended, the time required 
to complete each action item above would be helpful to the NRC in evaluating
the cost of this bulletin. 

If you have any questions regarding this matter, please contact the Regional
Administrator of the appropriate NRC regional office or the technical 
contact listed below. 

                              James M. Taylor, Director 
                              Office of Inspection and Enforcement 

Technical Contacts: H. A. Bailey, IE    R. J. Kiessel, IE 
                    (301) 492-9006      (301) 492-8119 

1. Table 1
2. Table 2 
3. List of Recently Issued IE Bulletins 

                                                       Attachment 1 
                                                       IEB 85-03 
                                                       November 15, 1985 

                                 TABLE 1 

                 Summary of Significant MOV Abnormalities 

                    Bypass switch improperly set       75 
                    Incorrect thrust                   50 
                    Unbalanced torque switch           33 
                    Valve backseating                   8
                    High motor current                  3
                    Torque switch abnormalities         2
                    Miscellaneous abnormalities        33 

Percent of valves experiencing abnormality. The total does not equal 
     100 percent as most valves had more than one abnormality. 

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