Intersystem LOCA Outside ContainmentML031200356 |
Person / Time |
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Site: |
Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
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Issue date: |
05/07/1992 |
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From: |
Rossi C E Office of Nuclear Reactor Regulation |
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To: |
|
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References |
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IN-92-036, NUDOCS 9205010045 |
Download: ML031200356 (10) |
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Similar Documents at Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
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Category:NRC Information Notice
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Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination 2020-09-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>. |
UNITED STATES NUCLEAR REGULATORY
COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555 May 7, 1992 NRC INFORMATION
NOTICE 92-36: INTERSYSTEM
LOCA OUTSIDE CONTAINMENT
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
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.Background
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 92050_ 45 AMA
IN 92-36 May 7, 1992 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 overpressur- ization 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 overpressure
events 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.Discussion
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 (PIYs) 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 analyses
IN 92-36 May 7, 1992 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 capabili-ties 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
uncer-tainties.
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 an ISLOCA.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 precur-sors 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 recog-nize 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
injury.
IN 92-36 May 7, 1992 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 complet-ing this research, the staff may issue further generic correspondence
to licensees.
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.es ERossi, Drectr Division of Operational
Events Assessment
Office of Nuclear Reactor Regulation
Technical
contacts:
Kazimieras
Campe, NRR (301) 504-1092 Sammy Diab, RES (301) 492-3914 Gary Burdick, RES (301) 492-3812 Attachments:
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
Attachment
1 IN 92-36 May 7, 1992 Table 1. Observed Plant Vulnerabilities
to ISLOCA Precursors (Obtained
from root cause analyses of ISLOCA precursors
and plant inspections)
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
opera-tors ("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
Attachment
2 IN 92-36 May 7, 1992 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
important.
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 essen-tial 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.
Attachment
3 IN 92-36 May 7, 1992 Table 3. A Selected List of ISLOCA Reports and References
Identification
IN 90-64 IN 90-05 IN 89-73 Title or Subject Potential
for Common-Mode
Failure of HPSI Pumps or Release of Reactor Coolant Outside Containment
During a LOCA Inter-system
Discharge
of Reactor Coolant Potential
Overpressurization
of Low Pressure Systems Date 10/04/90 01/29/90 11/01/89 10/23/90 05/08/89 AIT Report 50-456/90-20
AIT Report 50-293/89-80
An assessment
of the 10/4/90 Braidwood loss of reactor coolant inventory
and personnel
contamination
and injury An assessment
of the 4/12/89 Pilgrim overpressurization
event, which occurred during the conduct of the RCIC logic test Inspection
Report 50-382/90-200
Inspection
Report 50-413,414/90-200
Inspection
Report 50-346/89-201
ISLOCA Program Inspection
plant ISLOCA Program Inspection
plants ISLOCA Program Inspection
Besse plant of the Waterford of the Catawba of the Davis 09/14/90 06/11/90 12/21/89 Audit Report Docket No. 50-213 NUREG/CR-5745 NUREG/CR-5744 NUREG/CR-5604 NUREG/CR-5124 NUREG/CR-5102 Haddam Neck ISLOCA Audit Report: July 24 -August 4, 1989, Enclosure
to Memorandum
from Frank J. Congel, NRC, to Steven A. Varga, NRC*Assessment
for ISLOCA Risks -Draft Methodology
and Application:
Combustion
Engineering
Plant Assessment
for ISLOCA Risks -Draft Methodology
and Application:
Westinghouse
Four-Loop
Ice Condenser
Plant Assessment
for ISLOCA Risks -Draft Methodology
and Application:
Babcock and Wilcox Nuclear Power Station Interfacing
Systems LOCA, Boiling Water Reactors Interfacing
Systems LOCA, Pressurized
Water Reactors 09/20/89 June 91 Feb 91 Feb 91 Feb 89 Feb 89-A COpy OT this report is available
in the NRC Public Document Room, 2120 L Street, N.W., Washington, DC.
Attachment
4 IN 92-36 May 7, 1992 LIST OF RECENTLY ISSUED NRC INFORMATION
NOTICES Information
Date of Notice No. Subject Issuance Issued to 92-35 92-34 92-33 92-32 92-31 92-30 Higher Than Predicted
Ero-sion/Corrosion
in Unisol-able Reactor Coolant Pres-sure Boundary Piping Inside Containment
at A Boiling Water Reactor New Exposure Limits for Airborne Uranium and Thorium Increased
Instrument
Response Time When Pressure Dampening Devices are Installed Problems Identified
with Emergency
Ventilation
Systems for Near-Site (Within 10 Miles) Emer-gency Operations
Facili-ties and Technical
Support Centers Electrical
Connection
Problem in Johnson Yokogawa Corporation
YS-80 Programmable
Indi-cating Controllers
Falsification
of Plant Records Spent Fuel Pool Re-activity Calculations
05/06/92 05/06/92 04/30/92 04/29/92 04/27/92 04/23/92 04/22/92 All holders of OLs or CPs for nuclear power reactors.All licensees
whose opera-tions can cause airborne concentrations
of uranium and thorium.All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for nuclear power reactors.All holders of OLs or CPs for nuclear power reactors and all licensed operators and senior operators.
All holders of OLs or CPs for nuclear power reactors.92-21, Supp. 1 OL = Operating
License CP = Construction
Permit
IN 92-36 May 7, 1992 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 complet-ing this research, the staff may issue further generic correspondence
to licensees.
This information
notice requires you have any questions
about the of the technical
contacts listed Reactor Regulation (NRR) project no specific action or written response.
If information
in this notice, please contact one below or the appropriate
Office of Nuclear manager.Original Signed by Charles E Rei Charles E. Rossi, Director Division of Operational
Events Assessment
Office of Nuclear Reactor Regulation
Technical
contacts: Kazimieras
Campe, NRR (301) 504-1092 Sammy Diab, RES (301) 492-3914 Gary Burdick, RES (301) 492-3812 Attachments:
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 Document Name: IN 92-36*See previous concurrence.
C/OGCB:DOEA:NRR
D/DOEA:NRR
04/24/92 RPB:ADM*TechEd 04/09/92 D/DSIR:RES
- WMinners 04/15/92 C/RPSIB:DSIR:RES
- KKniel 04/14/92 RPSIB:DSIR:RES
- GBurdick 04/13/92 C/EIB:DSIR:RES
- RLBaer 04/13/92 OGCB:DOEA:NRR
- CVHodge 04/08/92 SC/RAB:DREP:NRR
- KCampe 04/09/92 C/RAB:DREP:NRR
- WBeckner 04/09/92 D/DREP:NRR
- FCongel 04/09/92 EIB:DSIR:RES
- SDiab 04/13/92 IN 92-XX April xx, 1992 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 complet-ing this research, the staff may issue further generic correspondence
to licensees.
This information
notice requires no speci you have any questions
about the informat of the technical
contacts listed below or Reactor Regulation (NRR) project manager.fic action or written response.
If ion in this notice, please contact one the appropriate
Office of Nuclear Charles E. Rossi, Director Division of Operational
Events Assessment
Office of Nuclear Reactor Regulation
Technical
contacts: Kazimieras
Campe, NRR (301) 504-1092 Sammy Diab, RES (301) 492-3914 Gary Burdick, RES (301) 492-3812 Attachments:
1. Table 1. "Observed
Plant Vulnerabilities
to ISLOCA Precursors" 2. Table 2. "ISLOCA Risk Insights" 3. Table 3. "A Partial List of ISLOCA Reports and References" 4. List of Recently Issued NRC Information
Notices Document Name: C/OGCB:DOEA:NRR
D/D1 CHBerling~*
fj, CER 04/21/92gq"'
04/RPB:ADM D/D TechEd J7Hh9q W" 04/ q/92 04/OGCB:DOEA:
RR SC/I CVHodge US9 KCai 04/od/92 04/ISLOCA REV 2 DOEA: NRR tossi l g kS 15~/ 92 RAB: REP:NRR imp I9/9 C L SI R:RES 04/A//92 C/RAB:DREP:IER
WBeckner Xyt 04/A /92 R Q DSIR:RES GB k 04//3/92 D/DREP:N FCongel 04/9 /92 C/EIB:D IRRES RLBaerXiF'
04//3/92 EIB:DSIR L>SDiab 04/,3/92/
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list | - Information Notice 1992-01, Cable Damage Caused by Inadequate Cable Installation Procedures and Controls (3 January 1992)
- Information Notice 1992-02, Relap5/MOD3 Computer Code Error Associated with the Conservation of Energy Equation (3 January 1992)
- Information Notice 1992-02, Relap5/Mod3 Computer Code Error Associated with the Conservation of Energy Equation (3 January 1992)
- Information Notice 1992-03, Remote Trip Function Failures in General Electric F-Frame Molded-Case Circuit Breakers (6 January 1992)
- Information Notice 1992-04, Potter and Brumfield Model Mdr Rotary Relay Failures (6 January 1992, Topic: Probabilistic Risk Assessment)
- Information Notice 1992-04, Potter and Brumfield Model MDR Rotary Relay Failures (6 January 1992, Topic: Probabilistic Risk Assessment)
- Information Notice 1992-05, Potential Coil Insulations Breakdown in Abs RXMH2 Relays (8 January 1992)
- Information Notice 1992-05, Potential Coil Insulations Breakdown in Abs Rxmh2 Relays (8 January 1992)
- Information Notice 1992-05, Potential Coil Insulations Breakdown in ABS RXMH2 Relays (8 January 1992)
- Information Notice 1992-06, Reliability of ATWS Mitigation System and Other NRC Required Equipment Not Controlled by Plant Technical Specifications (15 January 1992)
- Information Notice 1992-06, Reliability of ATWS Mitigation System and Other NRC Required Equipment not Controlled by Plant Technical Specifications (15 January 1992)
- Information Notice 1992-07, Rapid Flow-induced Erosion/Corrosion of Feedwater Piping (9 January 1992)
- Information Notice 1992-08, Revised Protective Action Guidance for Nuclear Incidents (23 January 1992)
- Information Notice 1992-09, Overloading and Subsequent Lock Out of Electrical Buses During Accident Conditions (30 January 1992)
- Information Notice 1992-10, Brachytherapy Incidents Involving Iridium-192 Wire Used in Endobronchial Treatments (31 January 1992, Topic: Brachytherapy)
- Information Notice 1992-10, Brachytherapy Incidents Involving Iridium-192 Wire used in Endobronchial Treatments (31 January 1992, Topic: Brachytherapy)
- Information Notice 1992-11, Soil and Water Contamination at Fuel Cycle Facilities (5 February 1992, Topic: Brachytherapy)
- Information Notice 1992-12, Effects of Cable Leakage Currents on Instrument Settings and Indications (10 February 1992, Topic: Brachytherapy)
- Information Notice 1992-13, Inadequate Control Over Vehicular Traffic at Nuclear Power Plant Sites (18 February 1992, Topic: Brachytherapy)
- Information Notice 1992-14, Uranium Oxide Fires at Fuel Cycle Facilities (21 February 1992, Topic: Brachytherapy)
- Information Notice 1992-15, Failure of Primary Systems Compression Fitting (24 February 1992)
- Information Notice 1992-16, Loss of Flow from the Residual Heat Removal Pump During Refueling Cavity Draindown (25 February 1992, Topic: Reactor Vessel Water Level, Temporary Modification, Brachytherapy)
- Information Notice 1992-17, NRC Inspections of Programs Being Developed at Nuclear Power Plants in Response to Generic Letter 89-10 (26 February 1992, Topic: Stroke time)
- Information Notice 1992-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire (28 February 1992, Topic: Hot Short, Safe Shutdown)
- Information Notice 1992-19, Misapplication of Potter and Brumfield Mdr Rotary Relays (2 March 1992)
- Information Notice 1992-19, Misapplication of Potter and Brumfield MDR Rotary Relays (2 March 1992)
- Information Notice 1992-20, Inadequate Local Leak Rate Testing (3 March 1992)
- Information Notice 1992-21, Spent Fuel Pool Reactivity Calculations (24 March 1992)
- Information Notice 1992-23, Results of Validation Testing of Motor-Operated Valve Diagnostic Equipment (27 March 1992)
- Information Notice 1992-24, Distributor Modification to Certain Commercial-Grade Agastat Electrical Relays (30 March 1992)
- Information Notice 1992-25, Pressure Locking of Motor-Operated Flexible Wedge Gate Valves (2 April 1992, Topic: Stroke time, Hydrostatic)
- Information Notice 1992-27, Thermally Induced Accelerated Aging and Failure of ITE/Gould A.C. Relays used in Safety-Related Applications (3 April 1992)
- Information Notice 1992-27, Thermally Induced Accelerated Aging and Failure of Ite/Gould A.C. Relays Used in Safety-Related Applications (3 April 1992)
- Information Notice 1992-28, Inadequate Fire Suppression System Testing (8 April 1992, Topic: Safe Shutdown)
- Information Notice 1992-29, Potential Breaker Miscoordination Caused by Instantaneous Trip Circuitry (17 April 1992)
- Information Notice 1992-30, Falsification of Plant Records (23 April 1992)
- Information Notice 1992-31, Electrical Connection Problem in Johnson Yokogawa Corporation YS-80 Programmable Indicating Controllers (27 April 1992)
- Information Notice 1992-32, Problems Identified with Emergency Ventilation Systems for Near-Site (Within 10 Miles) Emergency Operations Facilities and Technical Support Centers (29 April 1992)
- Information Notice 1992-32, Problems Identified with Emergency Ventilation Systems for Near-Site (within 10 Miles) Emergency Operations Facilities and Technical Support Centers (29 April 1992)
- Information Notice 1992-33, Increased Instrument Response Time When Pressure Dampening Devices Are Installed (30 April 1992)
- Information Notice 1992-33, Increased Instrument Response Time When Pressure Dampening Devices are Installed (30 April 1992)
- Information Notice 1992-34, New Exposure Limits for Airborne Uranium and Thorium (6 May 1992)
- Information Notice 1992-35, Higher than Predicted Erosion/Corrosion in Unisolable Reactor Coolant Pressure Boundary Piping Inside Containment at a Boiling Water Reactor (6 May 1992)
- Information Notice 1992-35, Higher than Predicted Erosion/Corrosion in Unisolable Reactor Coolant Pressure Boundary Piping inside Containment at a Boiling Water Reactor (6 May 1992)
- Information Notice 1992-36, Intersystem LOCA Outside Containment (7 May 1992)
- Information Notice 1992-37, Implementation of the Deliberate Misconduct Rule (8 May 1992)
- Information Notice 1992-38, Implementation Date for the Revision to the EPA Manual of Protective Action Guides and Protective Actions for Nuclear Incidents (26 May 1992, Topic: Brachytherapy)
- Information Notice 1992-39, Unplanned Return to Criticality During Reactor Shutdown (13 May 1992, Topic: Fuel cladding)
- Information Notice 1992-40, Inadequate Testing of Emergency Bus Undervoltage Logic Circuitry (27 May 1992)
- Information Notice 1992-41, Consideration of Stem Rejection Load In Calculation of Required Valve Thrust (29 May 1992, Topic: Anchor Darling)
... further results |
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