United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 84-74: Isolation of Reactor Coolant System from Low-pressure Systems Outside Containment

                                                          SSINS No.: 6835  
                                                          IN 84-74         

                               UNITED STATES 
                       NUCLEAR REGULATORY COMMISSION 
                    OFFICE OF INSPECTION AND ENFORCEMENT
                           WASHINGTON, DC 20555  

                             September 28, 1984

Information Notice No. 84-74:   ISOLATION OF REACTOR COOLANT SYSTEM FROM 
                                   LOW-PRESSURE SYSTEMS OUTSIDE CONTAINMENT 

Addressees: 

All nuclear power reactor facilities holding an operating license (OL) or a 
construction permit (CP). 

Purpose: 

This information notice is provided as a notification of potentially 
significant problems in maintaining isolation boundaries between the 
high-pressure reactor coolant system (RCS) and low-pressure piping systems 
outside containment. These problems contribute to an increased likelihood of
an intersystem loss-of-coolant accident (LOCA) which would bypass primary 
containment. 

It is expected that recipients will review the information for applicability
to their facilities and consider actions, if appropriate, to preclude 
similar problems occurring at their facilities. However, suggestions 
contained in this information notice do not constitute NRC requirements and, 
therefore, no specific action or written response is required. 

Description of Circumstances: 

Pilgrim 

At Pilgrim, on September 29, 1983, the low-pressure section of the high 
pressure coolant injection (HPCI) suction piping was overpressurized during 
functional testing of the HPCI system logic. The cause was personnel error 
while conducting more than one surveillance test at the same time. The 
prerequisites and initial conditions were not met for all steps. This led to
the simultaneous opening of two motor-operated HPCI discharge valves. 

A testable injection check valve should have isolated the HPCI from the RCS.
However, the movable internals of this valve were bound by rust. Apparently,
this condition held the valve partially open during normal operation, but 
did not prevent closing when pressure was applied. As a result, a sudden but
brief overpressurization of the HPCI piping occurred. 

The Pilgrim check valve was repaired and tested. The licensee held a 
critique with operators and instrumentation and control technicians and 
initiated administrative actions to ensure strict compliance with 
surveillance procedure action steps. 

8409270525  
.

                                                       IN 84-74            
                                                       September 28, 1984  
                                                       Page 2 of 4         

Hatch Unit 2 

At Hatch Unit 2 in October 1983, it was found that a testable check valve 
had been held open for about 4 months by an incorrectly assembled actuator. 
This is a swing-type testable check valve with an air actuator controlled by 
a four-way pilot solenoid valve. It is installed on a 24-inch low pressure 
coolant injection (LPCI) line. The second isolation device on this line is a 
normally closed motor-operated gate valve. The gate valve automatically 
receives a signal to open upon a LPCI actuation signal but has independent 
diverse interlocks to prevent opening at high differential pressure. 

The Hatch event resulted from a series of errors. On June 7, 1983, during 
maintenance on the valve actuator, the two air supply lines were installed 
backwards. The air supply line to the right-hand cylinder of the actuator 
was incorrectly connected to the left-hand cylinder, and vice versa. Failure 
to use a vendor maintenance manual appears to have contributed to this 
error. Inadequate post-maintenance functional testing of the valve allowed 
the initial error to go undetected. The check valve position is indicated in 
the control room. It is not known with certainty why this did not lead to 
early detection. However, it appears likely that, after maintenance, the 
indication was readjusted to show a closed position in the belief that the 
check valve must actually be closed. 

A Hatch plant maintenance worker was counseled by utility management on the 
importance of performing correct maintenance and the importance of using 
maintenance manuals and performing thorough post maintenance testing before 
returning components to service, particularly for components that are 
safety-related. For the longer term, the licensee is considering alternative 
methods of testing the check valve using shutdown cooling flow. This could 
allow permanently deactivating the actuator without interfering with check 
valve operability or position indication. 

Browns Ferry Unit 1 

At Browns Ferry Unit 1, on August 14, 1984, the core spray system was 
overpressurized. In December 1983 or earlier, during maintenance on the 
pilot solenoid valve for the testable check valve, a plunger with reversed 
air ports was apparently installed in the solenoid valve. This resulted in 
the check valve being held open. Then in August, while performing a 
semi-annual logic functional test, the operators failed to electrically 
disarm the motor-operated injection valve. The motor-operated valve opened 
with the testable check valve open. The core spray system, designed for 500 
psi, was pressurized above 500 psi, lifting the small relief valve installed 
on the line. The maximum core spray system pressure is not known. Operators 
observed that the control room pressure gauge read off-scale (above 500 psi) 
and the pressure might have approached primary coolant system pressure 
(about 1050 psi). 

The piping was not damaged, probably because of substantial design margins. 
The pump discharge check valves in combination with open pump suction valves
apparently prevented overpressurizing the pump suction piping which is 
designed for 150 psi. 
.

                                                       IN 84-74            
                                                       September 28, 1984  
                                                       Page 3 of 4         

The Browns Ferry occurrence was caused by errors similar to those in the 
Hatch event. The solenoid valve had been reassembled incorrectly and without
using a maintenance manual. The proximity switch and the air actuator had 
been readjusted providing an erroneous closed position indication. 
Post-maintenance testing was inadequate. Finally, during the logic 
functional test, there was a failure to properly follow the test procedure 
and to disarm the outboard motor-operated valve. The licensee's final 
corrective actions are yet to be determined. 

Discussion: 

The events described above are considered to be significant because they 
substantially reduced safety margins for preventing an intersystem LOCA that
bypasses containment. When the testable check valve is open, a postulated 
failure or inadvertent opening of, the motor-operated valve could allow 
discharge of high-pressure reactor coolant into low-pressure systems. 

The consequences of such an event are not certain. The flowrate through the 
motor-operated valve could vary from a small amount of leakage to a massive 
discharge. If the flow force were moderate, it could close the check valve 
despite the actuator. This would effectively terminate the event. If, 
however, the forces were large, the movable internal portions of the check 
valve could be severely damaged. A substantial failure of the low-pressure 
system, if it were to occur, would lead to a LOCA that bypasses the 
containment and could flood the low-pressure ECCS pumps. This would be an 
accident exceeding current design basis with radioactive material discharged
outside the primary containment. 

Other plants were thought to have a valve configuration similar to that of 
Hatch. Following the Hatch event, the NRC's Office of Analysis and 
Evaluation of Operational Data (AEOD) prepared Engineering Evaluation Report 
E414, "Stuck Open Isolation Check Valve on the Residual Heat Removal System 
at Hatch Unit 2," on May 31, 1984. This report confirmed that a number of 
BWRs have a similar residual heat removal (RHR) system configuration (i.e., 
a testable check valve inside primary containment and a motor-operated 
injection valve outside primary containment). Additional plants found with 
this configuration include Duane Arnold, Brunswick 1 and 2, Cooper, Dresden 
2 and 3, Hatch 1, FitzPatrick, Monticello, Peach Bottom 2 and 3, Pilgrim, 
and Quad Cities 1 and 2. 

In the Pilgrim and Browns Ferry events, the low-pressure section of the HPCI
suction piping and one loop of the core spray system, respectively, were 
actually overpressurized. The potential for an accident was increased by 
degradation of the barriers. 

It is suggested that licensees of nuclear reactor facilities consider 
reviewing their practices and controls in the area of maintenance activities 
involving air-operated testable check valves, especially where such valves 
provide isolation barriers for the reactor coolant system.  It is also 
suggested that licensees consider reviewing their practices and controls to 
ensure that instrument and logic tests do not permit the inadvertent opening 
of motor-operated valves, especially where these valves also provide 
isolation barriers for the reactor coolant system.
.

                                                       IN 84-74            
                                                       September 28, 1984  
                                                       Page 4 of 4         


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 NRC Regional Office or this 
office.


                                   Edward L. Jordan, Director
                                   Division of Emergency Preparedness
                                     and Engineering Response
                                   Office of Inspection and Enforcement

Technical Contacts: M. S. Wegner, IE 
                    (301) 492-4511 

                    S. Newberry, NRR 
                    (301) 492-8932 

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