United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 93-62: Thermal Stratification of Water in BWR Reactor Vessels

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C. 20555

                                August 10, 1993


NRC INFORMATION NOTICE 93-62:  THERMAL STRATIFICATION OF WATER IN BWR REACTOR  
                               VESSELS
                                                                               

Addressees

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

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees that loss of forced circulation through the reactor
vessel coupled with isolation from the main condenser may allow cold water to
stratify in the bottom of the reactor vessel and cause temperatures to be
lower than allowable.  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

Hatch Unit 1

On August 27, 1992, at Unit 1 of the Hatch Nuclear Power Plant, a high
radiation signal from a main steamline radiation monitor initiated a Group 1
isolation.  The main steam isolation valves closed and the reactor
automatically scrammed from 100-percent power.  The resulting low water level
in the reactor vessel caused the recirculation pumps to trip thereby
terminating forced circulation through the reactor vessel.  There was no
circulation by the reactor water cleanup system because the licensee had
previously isolated that system for testing.  The reactor vessel water level
was restored by the steam-driven feedwater pumps and the reactor core
isolation cooling (RCIC) system.  

After steam was no longer available to the feedwater pump turbines, water
level was maintained primarily by the RCIC system.  Relatively cool water was
added to the reactor vessel by injection from the RCIC system into the
feedwater sparger and from the control rod drive system into the lower region
of the reactor vessel.  Initially, the operators used the temperature of the
coolant in the drain line from the bottom head of the reactor vessel to
monitor the temperature in the reactor vessel.  Later, the operators realized
that the temperature in the drain line would not be meaningful because the
reactor water cleanup system had been secured and there was no flow in the 

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drain line.  Because of this, the operators changed to another point on the 
reactor vessel.  That location was subsequently determined to be above the 
reactor vessel bottom head and not representative of the minimum temperature
in the vessel.  A vendor review of the event determined that at one point the
bottom head temperatures were 8�C [15�F] lower than were allowed by the
pressure-temperature limits in the technical specifications.

The operators could not restart the reactor recirculation pumps because the
difference in temperature between the reactor dome and the reactor bottom was
greater than the technical specification limit of 62.8�C [145�F].  After the
reactor was depressurized and while the coolant was still stratified, the
operators started one residual heat removal pump in the shutdown cooling mode. 
The temperature at the drain line increased 82�C [220�F] in 10 minutes from
its initial 50�C [90�F] temperature; this change exceeded the technical
specification limit of 37.8�C [100�F] change in one hour.

Peach Bottom Unit 3

On October 15, 1992, at Peach Bottom Unit 3, a half-isolation of the primary
containment occurred after operators had performed a surveillance test of
low-pressure switches on the main steamline.  While plant personnel were
checking the relays to determine the cause of the half-isolation signal, a
second half-isolation signal was received.  The main steam isolation valves
closed and the reactor scrammed from 100-percent power.  High-pressure coolant
injection and RCIC automatically initiated and, in conjunction with the main
safety relief valves, were used to control water level and system pressure. 
During recovery from the transient, a second reactor scram from high pressure
occurred.  Because of a delay in resetting the first scram and limited flow
through the reactor drain line, thermal stratification of the reactor coolant
occurred in the vessel.  The operators did not consider the temperature of the
drain line to be representative of the bottom head temperature because the
reduced flow rate through the drain line caused too low an indicated
temperature due to heat losses from the drain line.

The operators could not restart the recirculation pumps because the
temperature difference between the reactor dome and the drain line was greater
than the 62.8�C [145�F] allowed by the technical specifications for restarting
a recirculation pump.  Therefore, the operators proceeded to depressurize and
cool the reactor before restoring forced circulation.  Although other bottom
head metal temperature indications were available, they were not actively
monitored by the operators because of procedural deficiencies and lack of
training.  A subsequent review of recorded temperature data determined that
the temperature difference between the reactor dome and the bottom head drain
line was about 135�C [240�F] and that the pressure-temperature limits for the
reactor vessel bottom head had been violated.  With the reactor pressure about
4.14 MPa [600 psig], the reactor vessel metal temperature had decreased to
about 64�C [115�F], nearly 22�C [40�F] lower than the pressure-temperature
limit.

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                                                               Page 3 of 3


Discussion

These events demonstrate that isolation of the reactor vessel from the main   
condenser with loss of recirculation flow can be initiated by a variety of
causes.  Isolation of the reactor vessel causes the operators to take manual 
control to restore proper water level and system pressure.  A restart of the
recirculation pumps may be delayed because of procedural restrictions. 
General Electric has issued a number of communications to licensees regarding
the loss of forced circulation in the reactor vessel.  In those
communications, General Electric addressed issues such as potential operating
difficulties, concerns about reactor vessel temperature monitoring, and the
potential for thermal stratification within the reactor vessel.  Also, plant
technical specifications contain pressure-temperature limitations to ensure
the integrity and safe operation of the reactor coolant system.  

Once thermal stratification occurs, any rapid circulation of water could
result in a large step change in the temperature of the water adjacent to the
reactor bottom head penetrations.  This step change may violate the technical
specification limits for rate of temperature change.  A temperature
differential within the reactor vessel may be reduced by increasing coolant
flow out of the bottom head drain and reducing cold water flow through the 
control rod drive system, which enters the bottom region of the reactor
vessel.  

Another concern is maintaining operation within brittle fracture temperature
limits.  Once temperature differences develop in the reactor vessel that
restrict restoring forced circulation, operator actions that affect pressure-
temperature limits are critical.  

Correct and timely operator response to the above conditions depends upon
proper actions being specified in plant procedures and appropriate training
being provided to operators for those actions. 

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


                                       ORIGINAL SIGNED BY 
                                   
                                       Brian K. Grimes, Director
                                       Division of Operating Reactors Support
                                       Office of Nuclear Reactor Regulation

Technical contact:  J. Carter, NRR
                    (301) 504-1153

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