Information Notice No. 92-36: Intersystem LOCA Outside Containment
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
May 7, 1992
NRC INFORMATION NOTICE 92-36: INTERSYSTEM LOCA OUTSIDE CONTAINMENT
All holders of operating licenses or construction permits for nuclear power
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees of potential plant vulnerabilities to intersystem
loss-of-coolant accidents (ISLOCAs). It is expected that recipients will
review the information for applicability to their facilities and consider
actions, as appropriate. However, suggestions contained in this information
notice are not NRC requirements; therefore, no specific action or written
response is required.
This information notice provides information gathered during a concerted NRC
staff effort to study plant vulnerabilities to ISLOCAs. The staff gathered
this information by performing (a) detailed evaluations of operating events,
(b) inspections of a limited sample of pressurized water reactors (PWRs),
and (c) extensive analyses of the sample PWRs. The information may be of
use in recipients' individual plant examination (IPE) programs.
The ISLOCA is a class of accidents in which a break occurs in a system con-
nected to the reactor coolant system (RCS), causing a loss of the primary
system inventory. This type of accident can occur when a low pressure
system, such as the residual heat removal (RHR) system, is inadvertently
exposed to high RCS pressures beyond its capacity. ISLOCAs of most concern
are those that can discharge the break flow outside the reactor containment
building, primarily because they can result in high offsite radiological
consequences but also because the RCS inventory lost cannot be retrieved for
long-term core cooling during the recirculation phase.
In the "Reactor Safety Study," (WASH-1400), published in 1975, and in
NUREG-1150, "Severe Accident Risks: An Assessment for Five U.S. Nuclear
Power Plants," the NRC described the ISLOCA outside containment as an event
of low core damage frequency, but as one of the main contributors to plant
risk. In those studies the NRC referred to the ISLOCA as "Event-V." Most
probabilistic risk assessments (PRAs) have also shown that the ISLOCA is
very unlikely. However, these PRAs typically have modelled only those
Event-V sequences that include only the catastrophic failure of check valves
that isolate the RCS from
May 7, 1992
Page 2 of 4
low pressure systems. These PRAs included little consideration of human
errors leading to an ISLOCA. Also, most existing PRAs have given little or
no credit for operator actions to terminate an ISLOCA or to mitigate its
radiological consequences if core melt were to occur.
On January 22, 1992, the Virginia Electric Power Company, licensee for the
North Anna Power Station, reported that the RHR relief valves would not pass
the design-basis flow to relieve an overpressurization of the RHR system
when the latter is aligned to the RCS. The function of these relief valves
is important when the RCS is water solid and therefore susceptible to
overpressurization events, such as from a charging-letdown flow mismatch or
a temperature change.
The licensee made this report after conducting an engineering evaluation to
respond to a notification by the nuclear steam supply vendor, the
Westinghouse Electric Corporation. In February 1990, Westinghouse reviewed
the RHR relief valve design basis for the Westinghouse Owners Group and
recommended that its customers review the following three items:
The adequacy of the RHR relief valves for protecting against cold
Discharge capability of relief valves for probable back pressures
Design basis commitments for valve specifications, commitments in the
final safety analysis report, and technical specifications
The NRC has issued several information notices to discuss certain
operational events regarding ISLOCAs. In IN 90-05, "Inter-system Discharge
of Reactor Coolant," the staff discussed an event during which about 68,000
gallons of reactor water was discharged outside the containment. The staff
has also analyzed operational experience and documented its findings in
augmented inspection team (AIT) reports. On October 23, 1990, the staff
issued AIT Report 50-456/90-020 on an event at Braidwood that resulted in
primary water leakage outside the containment and in the contamination of
three personnel, one of whom received a second degree burn. Table 3 is a
selected list of information notices and AIT reports that the staff has
issued on ISLOCAs and related events.
Although no ISLOCA has caused core damage, accumulated operational
experience, both in the United States and abroad, indicates that ISLOCA-like
events have occurred at a rate higher than expected. In conducting this
study, the staff defined an ISLOCA-like event, or an ISLOCA precursor, as an
event that results from the failure, degradation, or inadvertent opening of
the pressure isolation valves (PIVs) between the RCS and lower pressure
systems. An ISLOCA precursor may become an ISLOCA if it occurs during
different plant conditions, or if some of the failures occur together.
The NRC staff conducted root cause analyses of ISLOCA precursors, extensive
plant inspections, and detailed analyses of a sample of PWRs. These
May 7, 1992
Page 3 of 4
included thermal-hydraulic analyses, fragility analyses to determine the
likely sizes and locations of a break, and human reliability analyses. The
staff used the results of these analyses in PRAs to gain insights about the
significant contributors to ISLOCA risk.
The staff directed the studies described in this information notice towards
finding vulnerabilities of PWR plants to ISLOCAs, since the primary
pressures present in PWRs are greater than those found in boiling water
reactors (BWRs), while the design pressures of low pressure systems are
about the same in both PWRs and BWRs. However, BWR licensees also may find
this information to be relevant to their plants.
Upon conducting these studies, the staff made the following observations on
the ISLOCA risk at nuclear power plants:
1. The estimated core damage frequency caused by ISLOCAs could be greater
than was estimated in PRAs for some plants.
The ISLOCA risk depends on both the accident initiators and the
capabilities for recovery. These factors vary from plant to plant.
The main contributors to ISLOCA initiation and/or recovery include (a)
human errors and (b) the effects of the accident-caused harsh
environment on plant equipment and recovery activities. Both factors
have significant uncertainties. Existing PRAs have provided little or
no treatment of these factors. Plants that are particularly vulnerable
to either of these two factors could have a higher ISLOCA risk than
indicated by existing PRAs.
2. Most plants lack contingency plans to provide backup water supplies
that can be transferred readily to provide long-term core cooling after
By examining a plant's emergency procedures, a licensee can find
insights for improving the plant's features to address the concerns for
both ISLOCAs and other accidents.
3. The root cause analyses of operational events indicate that ISLOCA
precursors most likely would be initiated by human errors, notably
during testing and maintenance or because of procedural deficiencies.
This may be attributed to the general lack of awareness of the
possibility or consequences of an ISLOCA.
Licensees may significantly reduce the probability of ISLOCA precursors
by improving the ability of operators and maintenance personnel to
recognize ISLOCAs, mechanisms that can cause them, actions to prevent
them, and methods to manage them if they occur.
4. Most observed ISLOCA precursors have low public risk consequences.
However, an ISLOCA precursor can require a shutdown or extension of a
shutdown, require radioactivity cleanup operations, and cause personnel
May 7, 1992
Page 4 of 4
Table 1 presents the staff's observations from root cause analyses and plant
inspections. Table 2 presents insights gained from the ISLOCA PRAs.
The staff is completing its ISLOCA research program under Generic Issue 105,
"Intersystem Loss of Coolant Accidents in Light Water Reactors." Upon
completing this research, the staff may issue further generic correspondence
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.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical contacts: Kazimieras Campe, NRR
Sammy Diab, RES
Gary Burdick, RES
1. Table 1. "Observed Plant Vulnerabilities to ISLOCA Precursors"
2. Table 2. "ISLOCA Risk Insights"
3. Table 3. "A Selected List of ISLOCA Reports and References"
4. List of Recently Issued NRC Information Notices
May 7, 1992
Page 1 of 1
Table 1. Observed Plant Vulnerabilities to ISLOCA Precursors
(Obtained from root cause analyses of ISLOCA precursors and plant
1. Lack of awareness of the nature or consequences of ISLOCAs
2. Inadequate emergency procedures for ISLOCA outside containment,
especially for non-power operational modes
3. Poor or incorrect valve labels
4. Different nomenclature used for the same equipment in the same plant
5. Poor coordination between concurrently run tests
6. Miscommunications between the control room operators and auxiliary
operators ("get the valve" is meant as "crack open then close," but
understood to mean "open")
7. Poor shift turn-over communications
8. Poor post-maintenance testing or operability checks
9. Inadequate application of independent verification
10. Tendency not to check diverse instrument indications
11. Tendency to commit personnel to extensive overtime work, especially
during shutdown and startup operations, thus increasing the fatigue
level and the likelihood of errors
May 7, 1992
Page 1 of 1
Table 2. ISLOCA Risk Insights
(Obtained from ISLOCA PRAs)
1. The staff's studies suggest that the core damage frequency caused by an
ISLOCA could be substantially greater than previous PRA estimates for
some plants. This is primarily caused by the effects of operator
errors and harsh environments caused by the accident. Valve alignment
errors during transition between operating modes can be particularly
2. Equipment qualified for a harsh environment is likely to survive the
adverse ISLOCA temperature and humidity, but not the possible
submersion caused by flooding.
3. Multiple system failures may result from the ISLOCA harsh environment
or flooding, depending on the size and location of the break in
relation to affected equipment, the separation of redundant trains, and
the effect of fire sprays on flooding.
4. ISLOCA recovery is limited by harsh environments, which may damage
essential equipment thus complicating long-term cooling, and the rate
of loss of reactor water outside the containment. If the water is not
quickly replenished, an ISLOCA may lead to core damage, even after the
leak has been isolated.
5. Symptom-based procedures may lead the operator to realize that an
ISLOCA has occurred. However, unless the emergency procedures refer to
plant provisions for conserving and replenishing water, the operator
may have difficulty managing the accident.
6. Most observed ISLOCA precursors have low risk consequences, primarily
because of the presence of one or more of the following conditions:
small leak size, redundant means of detecting and isolating a leak, and
low power or shutdown conditions.
May 7, 1992
Page 1 of 1
Table 3. A Selected List of ISLOCA Reports and References
Identification Title or Subject Date
IN 90-64 Potential for Common-Mode Failure of 10/04/90
HPSI Pumps or Release of Reactor Coolant
Outside Containment During a LOCA
IN 90-05 Inter-system Discharge of Reactor Coolant 01/29/90
IN 89-73 Potential Overpressurization of Low 11/01/89
AIT Report An assessment of the 10/4/90 Braidwood 10/23/90
50-456/90-20 loss of reactor coolant inventory and
personnel contamination and injury
AIT Report An assessment of the 4/12/89 Pilgrim 05/08/89
50-293/89-80 overpressurization event, which occurred
during the conduct of the RCIC logic test
Inspection ISLOCA Program Inspection of the Waterford 09/14/90
Inspection ISLOCA Program Inspection of the Catawba 06/11/90
Inspection ISLOCA Program Inspection of the Davis 12/21/89
Report Besse plant
Audit Haddam Neck ISLOCA Audit Report: July 24 - 09/20/89
Report August 4, 1989, Enclosure to Memorandum
Docket No. 50-213 from Frank J. Congel, NRC, to
Steven A. Varga, NRC*
NUREG/CR-5745 Assessment for ISLOCA Risks - June 91
Draft Methodology and Application:
Combustion Engineering Plant
NUREG/CR-5744 Assessment for ISLOCA Risks - Feb 91
Draft Methodology and Application:
Westinghouse Four-Loop Ice Condenser Plant
NUREG/CR-5604 Assessment for ISLOCA Risks - Feb 91
Draft Methodology and Application: Babcock
and Wilcox Nuclear Power Station
NUREG/CR-5124 Interfacing Systems LOCA, Boiling Feb 89
NUREG/CR-5102 Interfacing Systems LOCA, Pressurized Feb 89
*A copy of this report is available in the NRC Public Document Room, 2120 L
Street, N.W., Washington, DC.
Page Last Reviewed/Updated Friday, May 22, 2015