United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 89-30: High Temperature Environments at Nuclear Power Plants

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

                                 March 15, 1989


Information Notice No. 89-30:  HIGH TEMPERATURE ENVIRONMENTS AT 
                                   NUCLEAR POWER PLANTS


Addressees:

All holders of operating licenses or construction permits for nuclear power
reactors. 

Purpose:

This information notice is being provided to alert addressees to potential 
problems resulting from high temperature environments in areas that contain 
safety-related equipment or electrical cables.  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 con-
tained in this information notice do not constitute NRC requirements; 
therefore, no specific action or written response is required. 

Description of Circumstances:

In November 1988, while Duane Arnold Energy Center (DAEC) was shut down for 
refueling, the licensee for DAEC discovered 1 pinhole leak, 2 through-wall 
cracks, and 30 flaw indications on the control rod drive (CRD) insert lines 
inside the drywell.  The defects were caused by externally induced chloride 
stress corrosion cracking.  The area near the defects contained Rockbestos 
Firewall III radiation, cross-  linked, polyethylene-insulated, electrical 
cable with a Hypalon (Neoprene Chloroprene) jacket.  The cable had previously 
been degraded by exposure to local drywell temperatures in excess of 270�F.  
When the damaged electrical cable was replaced, loose degraded insulation 
lodged in the conduit and the field junction box.   Moisture from steam leaks 
condensed in and dripped through the conduit onto the CRD piping.  The conden-
sate contained chlorides that were leached from the insulation lodged in the 
conduit and the junction box.  There are several areas at a reactor facility 
where degradation of cables and leaching of chloride may occur because of high
temperature and humidity.  In addition to the drywell, the licensee for DAEC 
also found indications of chlorides leaching on the steam tunnel. 

During a refueling outage in November 1988, the licensee for Dresden Unit 2 
discovered evidence that paint inside the upper region of the drywell had 
been exposed to elevated temperatures.  Further investigation revealed that 
the Limitorque operators on the steam supply valves to the high-pressure 




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                                                            Page 2 of 3


coolant injection system and the isolation condenser (located in the same 
area) had indications of exceeding their environmental qualification (EQ) 
design temperature.  Grease samples taken from these valves showed significant 
degradation, and the lower main bearing of one valve operator was damaged.  
Other equipment affected by the high temperature included two vessel head vent 
valves and a standby liquid control valve.  Also, the electrical insulation on 
about 50 cables was cracked.   The root cause for the elevated temperature at 
Dresden was attributed to a deficiency in procedures that resulted in the 
ventilation ducts in the upper region of the drywell being left closed for 
about 18 months while the plant was in operation. 

In August 1987, the NRC became aware that Arkansas Nuclear One, Unit 1 
(ANO-1), had probably operated since it was licensed in 1974 with containment 
temper-atures ranging from 90�F to 180�F.  The bulk average temperature was 
roughly 140�F.  Safety-related electrical equipment is environmentally 
qualified to operate at temperatures up to 120�F.  Also, design basis accident 
scenarios had been analyzed assuming an initial containment temperature of 
110�F.  Over the years, the licensee for ANO-1 attempted to reduce the high 
containment temperature by installing improved insulation on the reactor 
coolant system and by acid cleaning of the chillers used for the containment 
cooling units.  These efforts resulted in a very limited temperature 
reduction. 

Discussion:

In the boiling-water reactor events described above, elevated drywell temper-
ature was responsible for degradation of safety-related equipment.  Electrical
cables are vulnerable to degradation when exposed to high temperatures that 
exceed their design EQ temperature even for a short period.  Regarding the 
DAEC event, the elevated temperature along with high humidity led to the 
degradation of safety-related components. 

In the ANO-1 event, the higher local temperatures exceeded some of the EQ 
temperatures for some of the safety and non-safety equipment and components.  
Also, the higher bulk temperature exceeded the ambient temperature assumed 
in some of the accident analyses.  Three of the analyses that were affected 
were:

     1. The reactor building peak pressure analysis.
     2. The inadvertent initiation of the containment spray system analysis.
     3. The internal containment subcompartment differential pressure analysis.

There has been a history of reports since 1982 of boiling-water reactors 
(BWRs) and pressurized-water reactors (PWRs) experiencing excessive heat load 
problems within the drywell and localized high temperature areas within 
containment.  On June 30, 1988, the NRC issued Temporary Instruction (TI) 
2515/98, "Information of High Temperature Inside Containment/Drywell in PWR 
and BWR Plants."  The objective of this TI was to determine whether or not 
high containment or drywell 
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                                                            Page 3 of 3


temperatures were a plant-specific problem or generic to all PWRs and BWRs.  
Preliminary findings from the TI showed that:

     1. BWRs, especially Mark I and II containments, routinely operate very 
        close to their EQ temperature limit.

     2. In the drywells of BWRs there may be substantial temperature gradients
        (i.e., 100�F or more) that may or may not be detected depending on the 
        location of instrumentation and circulation of the drywell air.

     3. The BWR drywell head region seems most susceptible to high temperature.

     4. Some PWRs experienced high containment temperatures but the licensees
        failed to recognize the safety significance and take corrective actions.

It is important for licensees to be aware that there are areas within the plant 
where the local temperature may exceed equipment qualification specifications 
even when the bulk temperature, as measured by a limited number of sensors, is 
indicating that it is lower than the qualification temperature.

No specific action or written response is required by this information notice.  
If you have any questions about this matter, please contact one of the 
technical contacts listed below or the Regional Administrator of the 
appropriate regional office. 




                                   Charles E. Rossi, Director
                                   Division of Operational Events Assessment
                                   Office of Nuclear Reactor Regulation

Technical Contacts:  R. Anand, NRR
                     (301) 492-0805

                     T. Greene, NRR
                     (301) 492-1176

Attachment:  List of Recently Issued NRC Information Notices
.                                                            Attachment 
                                                            IN 89-30
                                                            March 15, 1989
                                                            Page 1 of 1

                             LIST OF RECENTLY ISSUED
                             NRC INFORMATION NOTICES
_____________________________________________________________________________
Information                                  Date of 
Notice No._____Subject_______________________Issuance_______Issued to________

89-29          Potential Failure of          3/15/89        All holders of OLs
               ASEA Brown Boveri                            or CPs for nuclear
               Circuit Breakers                             power reactors.
               During Seismic Event

89-28          Weight and Center of          3/14/89        All holders of OLs
               Gravity Discrepancies                        or CPs for nuclear
               for Copes-Vulcan                             power reactors.
               Air-Operated Valves

89-27          Limitations on the Use        3/8/89         All holders of OLs
               of Waste Forms and High                      or CPs for nuclear
               Integrity Containers for                     power reactors, 
               the Disposal of Low-Level                    fuel cycle 
               Radioactive Waste                            licenses and 
                                                            certain by-product 
                                                            materials licenses.

89-26          Instrument Air Supply to      3/7/89         All holders of OLs
               Safety-Related Equipment                     or CPs for nuclear
                                                            power reactors.

89-25          Unauthorized Transfer of      3/7/89         All U.S. NRC 
               Ownership or Control of                      source, byproduct, 
               Licensed Activities                          and special 
                                                            nuclear material 
                                                            licensees.

89-24          Nuclear Criticality Safety    3/6/89         All fuel cycle
                                                            licensees and other
                                                            licensees 
                                                            possessing more 
                                                            than critical 
                                                            mass quantities of
                                                            special nuclear 
                                                            material.

89-23          Environmental Qualification   3/3/89         All holders of OLs
               of Litton-Veam CIR Series                    or CPs for nuclear
               Electrical Connectors                        power reactors.
_____________________________________________________________________________
OL = Operating License
CP = Construction Permit 
..
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