Information Notice No. 89-30: High Temperature Environments at Nuclear Power Plants
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
All holders of operating licenses or construction permits for nuclear power
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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March 15, 1989
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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
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
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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March 15, 1989
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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
T. Greene, NRR
Attachment: List of Recently Issued NRC Information Notices
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.
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
89-26 Instrument Air Supply to 3/7/89 All holders of OLs
Safety-Related Equipment or CPs for nuclear
89-25 Unauthorized Transfer of 3/7/89 All U.S. NRC
Ownership or Control of source, byproduct,
Licensed Activities and special
89-24 Nuclear Criticality Safety 3/6/89 All fuel cycle
licensees and other
mass quantities of
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
Page Last Reviewed/Updated Friday, May 22, 2015