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HomeNRC LibraryDocument CollectionsGeneric CommunicationsInformation Notices1992

# Information Notice No. 92-36: Intersystem LOCA Outside Containment

```
                                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 


9205010045 
.

                                                            IN 92-36
                                                            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 
     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 (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 
analyses 
.

                                                            IN 92-36
                                                            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 
     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 
     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 
     injury. 

.

                                                            IN 92-36
                                                            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 
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. 




                                   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
.

                                                            Attachment 1
                                                            IN 92-36
                                                            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 
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 
     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 
     
.

                                                            Attachment 2
                                                            IN 92-36
                                                            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 
     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 
     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. 
     
.

                                                            Attachment 3
                                                            IN 92-36
                                                            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
                    Pressure Systems

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
Report              plant
50-382/90-200                      

Inspection          ISLOCA Program Inspection of the Catawba     06/11/90 
Report              plants
50-413,414/90-200                  

Inspection          ISLOCA Program Inspection of the Davis       12/21/89 
Report              Besse plant
50-346/89-201                      

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 
                    Water Reactors

NUREG/CR-5102       Interfacing Systems LOCA, Pressurized        Feb 89
                    Water Reactors

___________________
*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*

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