Information Notice No. 86-40: Degraded Ability to Isolate the Reactor Coolant System from Low-Pressure Coolant Systems in BWRS

                                                           SSins No:  6835 
                                                           IN 86-40        

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

                                June 5, 1986

Information Notice No. NO 86-40:   DEGRADED ABILITY TO ISOLATE THE REACTOR 
                                   COOLANT SYSTEM FROM LOW-PRESSURE COOLANT 
                                   SYSTEMS IN BWRS 

Addressees: 

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

Purpose: 

This notice is provided as a supplement to Information Notice (IN) 84-74 on 
interfacing systems loss-of-coolant accidents (LOCA) in boiling water 
reactors (BWRs) which would bypass primary containment Two recent events 
are described where the high-pressure reactor coolant system could not be 
fully isolated from low-pressure piping systems outside of 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 from occurring at their facilities However, suggestions 
contained in this information notice do not constitute NRC requirements; 
therefore, no specific action or written response is required 

Description of Circumstances: 

Pilgrim 

On February 13, 1986, Pilgrim experienced a "design pressure" alarm on the B 
loop of the residual heat removal (RHR) system because of leakage through a 
check valve (1001-68B) and a motor-operated isolation valve (1001-28B, the 
outboard, normally closed, isolation valve) in the line (see Figure 1) The 
RHR system has a design pressure of 450 psig compared with a 1250 psig 
design pressure of the reactor coolant system In addition to design 
pressure alarms, the piping up to the 28B valve had become warm Several 
design pressure alarms had occurred during the preceding several weeks The 
licensee's prior corrective action had been to vent the lines On February 
13, the licensee closed the normally open 1001-29B valve and opened the 28B 
valve to isolate the system The licensee planned to operate in this 
configuration until a scheduled refueling and maintenance outage in 
September 1986 



8606030014


                                                               IN 86-40    
                                                               June 5, 1986 
                                                               Page 2 of 3 

On April 11 and 12, 1986, the closed 29B valve began leaking Several high 
pressure alarms were reported The licensee bled off the line to reduce 
pressure and began an orderly shutdown, but within a short period of time, 
the high pressure alarm was again received The 28B valve was closed, but 
the leakage problem continued The shutdown rate was then increased until 
the unit scrammed because of other problems An NRC augmented inspection 
team was dispatched to the site to investigate these RHR valve problems and 
two problems in other systems 

Duane Arnold 

On March 15, 1986, while reducing power for an outage, Duane Arnold reported
closing the outboard LPCI (low-pressure coolant injection mode of RHR) 
isolation valve MO-2004 due to leakage through the inboard isolation valve 
MO-2003 (see Figure 2) 

Because the leakage flow was relatively small, the p across the check
valve inside of containment (CV-2002) was not high enough to seat the check 
valve The problem was discovered when it was noted that the pressure 
controllers on the RHR heat exchangers were indicating a pressure of 450 
psig rather than the normal 70 psig RHR system relief valves had lifted to 
keep pressure at or below 450 psig Closing the MO-2004 valve stopped the 
leakage, but because the plant has loop selection logic, both loops of LPCI 
were rendered inoperable The plant continued its planned shutdown and 
reached cold shutdown on March 16 The leaking valves were repaired during 
the outage 

Discussion: 

The underlying cause of this problem is leaking valves, one of which is 
inaccessible with the plant at power Possible solutions include increased 
surveillance, preventive maintenance, and reliability-based replacement 
Other BWR systems that can be subject to similar problems include but are 
not limited to core spray, high pressure coolant injection system (see IE 
Information Notice 84-74), and reactor core isolation cooling 

It is possible for leakage to exist from the reactor coolant system to a low
pressure system without causing a high pressure alarm or lifting of safety 
valves in the low pressure system For example, for the preceding events, if
the check valve at the discharge of one of the RHR pumps is leaking, the 
reactor coolant will flow to the suppression pool Hence, slowly increasing 
level in the suppression pool is one indication that there is leakage from 
the reactor coolant system to the low pressure RHR system, and degradation 
of the reactor coolant pressure boundary If leakage through the degraded 
valves were to increase suddenly, a severe accident could result, as 
described in the following paragraph Further, it should be noted that such 
leakage does not meet the intent of general design criteria 14, 30, and 54 
of Appendix A to 10 CFR Part 50 


                                                               IN 86-40    
                                                               June 5, 1986 
                                                               Page 3 of 3 

The leaking of primary coolant into RHR lines that were never meant to 
contain fluid at that temperature and pressure can cause a number of 
incidents: over-pressurization with possible faulting of the low pressure 
line and a LOCA, steam binding of one or more of the RHR pumps, and 
waterhammer The safety significance of these events is the increased 
probability of core melt and releases in excess of 10 CFR 100 limits The 
Office of Nuclear Reactor Regulations has designated this topic Generic 
Issue number 105, "Interfacing Systems LOCA at Boiling Water Reactors," and 
has given it a "high" priority A generic letter concerning staff actions 
relating to this topic is being considered 

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:  Mary S Wegner 
                    (301) 492-4511 

Attachments: 
1   Simplified RHR Diagrams 
2   List of Recently Issued IE Information Notices