Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
August 9, 1993
NRC INFORMATION NOTICE 93-61: EXCESSIVE REACTOR COOLANT LEAKAGE
FOLLOWING A SEAL FAILURE IN A REACTOR
COOLANT PUMP OR REACTOR RECIRCULATION PUMP
Addressees
All holders of operating licenses or construction permits for nuclear power
reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to the potential for excessive reactor coolant
leakage following a seal failure in a reactor coolant pump or reactor
recirculation pump. 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
Oconee Nuclear Station, Unit 1
On May 24, 1992, the licensee commenced a reactor shutdown from 100 percent
power because of excessive leakage from the 1A2 Reactor Coolant Pump seal.
The maximum leakage was approximately 23 liters per minute [6 gpm]. The seal
failed because of the premature degradation of obsolete seal parts that had
mistakenly been installed.
Westinghouse supplied the Unit 1 reactor coolant pumps, incorporating a three-
stage seal series arrangement to limit coolant flow up the pump shaft.
Although Westinghouse had provided the necessary information on the design
change of the seal, the information was not properly communicated to plant
personnel. As a result, the obsolete seal parts were not removed from the
station stock and appropriate maintenance procedures were not revised to
reflect the change. These deficiencies contributed to the fact that
maintenance personnel inadvertently installed the obsolete seal parts.
Further details can be found in Licensee Event Report (LER) 50-269/92-09 and
NRC Inspection Report No. 50-269/92-13.
9308030200.
IN 93-61
August 9, 1993
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Millstone Unit 1
On May 25, 1989, the licensee started up Unit 1 for Cycle 13 operation.
During plant heatup on May 27, 1989, operators noted indications of
intermittent seal failure for the "A" Reactor Recirculation Pump inner seal.
The licensee continued power escalation and cycle startup testing while making
plans to replace the seal. On May 29, 1989, while at full-power operation,
the drywell leakage exceeded the Technical Specifications limit and the
licensee commenced plant shutdown. The leakage was approximately 34 liters
per minute [9 gpm] at the start of the event and increased to about 174 liters
per minute [46 gpm] during the shutdown process. After reviewing the failed
seal and consulting with the pump manufacturer, Byron Jackson, the licensee
still did not identify the exact cause of the seal failure. However, the
licensee did determine that the pump seal had probably failed as a result of
one or more of the following causes: (1) improper seal handling prior to
installation, (2) introduction of debris and corrosion products into the seal
cavity, and (3) improper depressurization following hydrostatic testing of the
seal assembly.
Further details can be found in LER 50-245/89-14, Revision 1, and in NRC
Inspection Report No. 50-245/89-12.
Clinton Unit 1
On May 21, 1989, the licensee took the reactor to critical for Cycle 2
operation. On May 25, 1989, the pressure in the seal outer cavity decreased
to approximately 414 kPa gauge [60 psig], indicating failure of the upper seal
stage. Approximately 10 hours later, the seal appeared to reseal and operated
properly. On June 1, 1989, with the reactor at about 42-percent power during
power ascension, upon shifting the "B" Reactor Recirculation Pump speed from
low to high, the operators immediately noted indications that both the upper
and lower seals in the pump had failed. The seal failures resulted in
increased flow from the drywell floor drain sump inlet; the leakage reached a
maximum of 242 liters per minute [64 gpm]. The licensee then initiated plant
shutdown. Although the exact cause of the seal failure was not determined,
the licensee indicated that the probable cause was improper assembly or
improper installation.
Further details can be found in the licensee special report submitted to NRC
on June 30, 1989, and in NRC Inspection Report No. 50-461/89-21.
Discussion
Both reactor coolant pumps and reactor recirculation pumps use a series of
primary and secondary seals to limit the reactor coolant leakage to
containment. A loss-of-coolant accident (LOCA) can occur if leakage through
the seals of reactor coolant pump or reactor recirculation pump exceeds the
capacity of the normal makeup systems. Thus, the failure of these seals can
represent a significant degradation of the reactor coolant pressure boundary.
.
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August 9, 1993
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The NRC has, over a period of years, evaluated the issue of reactor coolant
pump and reactor recirculation pump seal-related problems (Generic Issue 23)
and the need for additional licensing requirements to reduce the potential
core-melt risk resulting from the failure of these pump seals. An evaluation
program was initiated to resolve the generic issue and address several reactor
coolant pump seal leaks that occurred in the late 1970s and the early 1980s.
Analysis performed in conjunction with the evaluation indicated that the
overall probability of core-melt due to small-break LOCAs could be dominated
by reactor coolant pump seal failures. The two conditions under which seals
have failed or could fail, normal operating conditions and off-normal
operating conditions, are addressed below:
Seal performance under normal operating conditions
Based on the review of LERs and feedback from industry, some licensees appear
to have recently made major improvements in reactor coolant pump and reactor
recirculation pump seal performance. This improvement is attributed to a
combination of factors, including the following: enhanced seal quality
assurance programs, modified seal design, improved maintenance procedures and
training, closer attention to detail, improved seal operating procedures, more
knowledgeable personnel involved in seal maintenance, reduction in frequency
of transients that stress the seals, and seal handling and installation
equipment designed with the appropriate precision. However, not all plants
have implemented such measures, and some seal failures have caused substantial
reactor coolant leakage (as described above).
Development and implementation of appropriate procedures and training can help
assure correct operator action for operational conditions related to seal
degradation and can assist to identify seal degradation in a timely manner.
This can thereby prevent or mitigate cascade failure of multistage seal
assemblies.
Section III of the ASME Boiler and Pressure Vessel Code has included specific
exclusions for seal components under NB-3411.2 and NB-2121(b) relative to
design requirements. However, code exclusions by themselves do not relieve
licensees from other pertinent regulatory requirements such as Appendix B to
10 CFR Part 50, as applicable. For those items covered by Appendix B, as
reflected in plant-specific licensing bases, a quality assurance program is
required. The staff is considering additional generic action to address
whether all licensees should treat certain seal components as safety-related.
Seal performance under off-normal operating conditions
With respect to off-normal operating conditions, particularly loss of all seal
cooling water which can be caused by station blackout, loss of component
cooling water or loss of service water, the major concerns involve seal
failures due to adverse temperature effects on elastomer seal materials and
performance instabilities at the primary seal face related to coolant flashing
and two-phase flow. The staff is proceeding with rulemaking to address issues
of ensuring reactor coolant pump seal function or compensating for seal
failure during loss of seal cooling events, including station blackout.
.
IN 93-61
August 9, 1993
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This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.
ORIGINAL SIGNED BY
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts: Jai Raj N. Rajan, NRR
(301) 504-2788
Peter C. Wen, NRR
(301) 504-2832
Attachment:
List of Recently Issued NRC Information Notices .
