Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
SSINS No: 6835
IN 86-11
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, DC 20555
February 25, 1986
Information Notice No. NO 86-11: INADEQUATE SERVICE WATER PROTECTION
AGAINST CORE MELT FREQUENCY
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP)
Purpose:
This notice is to alert recipients of a potentially significant problem
concerning the possible failure to provide sufficient redundancy in the
essential service water system (ESW) Failure of all ESW may be an
accident-initiating event that could lead to a core melt It is expected
that recipients will review the information for applicability to their
facilities and consider action, if appropriate, to preclude a similar
problem 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
Past Related Correspondence:
Circular 78-13, "Inoperability of Service Water Pumps," July 10, 1978
Description of Circumstances:
In May 1984, the Byron Unit 1 licensee, Commonwealth Edison, submitted to
the NRC a probabilistic risk assessment (PRA) to justify extending allowable
outage times for certain equipment from 3 days to 7 days The NRC reviewed
the study and determined that loss of both ESW pumps on Unit 1 was not
considered as an accident-initiating event At present, Byron Unit 1 is
operating and has two ESW pumps--one operating and one on standby If the
operating train failed and the standby train would not start, the component
cooling water system (CCW) would heat up The nuclear steam supply system
vendor, Westinghouse, has estimated that the heatup of the CCW would trip
the CCW pumps in 6 minutes CCW is essential for cooling the reactor coolant
pump (RCP) seals, either directly or via the charging pumps, which also are
cooled by CCW Without cooling, the RCP seals may possibly fail and cause a
loss-of-coolant accident (LOCA) 1 Assuming that case in the PRA the ECCS
pumps needed to mitigate the ensuing
1 The NRC Office of Nuclear Reactor Regulation currently has the subject of
RCP seal failure under study in its Generic Issue 23
8602200397
IN 86-11
February 25, 1986
Page 2 of 2
LOCA, also would fail without CCW Thus, loss of ESW could result in a core
melt
On reevaluation of the study under the assumption that loss of ESW is a
LOCA-initiating event, the core melt frequency was estimated at 0001 per
year This result also assumed that additional pumps, such as the ESW pumps
for Unit 2, would not be available to mitigate the LOCA at Unit 1 To lower
the estimated core melt frequency, the licensee committed to make at least
one of the Unit 2 ESW pumps available to Unit 1 by means of a crosstie
piping arrangement in the event of either of the Unit 1 ESW pumps becoming
inoperable The availability of the Unit 2 ESW pump reduces the core melt
frequency estimated from this sequence of events by a factor of 25, and the
overall core melt frequency by a factor of 5 These estimates reaffirm the
perceived weakness of the two-train system and the desirability of making
the Unit 2 ESW pump available
Discussion:
In the Byron design, each of the two Unit 1 ESW pumps will supply 100
percent of the system's requirements for Unit 1 The system was licensed as
meeting single failure safety criteria The PRA, however, identified a
circumstance in which even such a licensed system may represent significant
risk, so the licensee remedied the situation by committing to make available
an additional 100 percent capacity ESW pump from Unit 2 when one Unit 1 ESW
pump becomes inoperable
The NRC has surveyed pressurized water reactor, (PWR) designs that are in
operation and under construction as to availability of sufficient pumping
capacity for ESW This preliminary information indicates only one of these
PWR designs may be subject to the same question discussed here for Byron
This is being pursued by the NRC staff Boiling water reactor (BWR) designs
have not been similarly surveyed; it has not been determined by a PRA
whether loss of ESW could similarly result in a high core melt frequency for
BWRs Thus, due to uncertainty resulting from the incompleteness of this
survey, this notice is being published
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: Vern Hodge, IE
(301) 492-7275
Leonard Olshan, NRR
(301) 492-4937
Attachment: List of Recently Issued IE Information Notices
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