Information Notice No. 92-28: Inadequate Fire Suppression System Testing

                                UNITED STATES
                           WASHINGTON, D.C.  20555

                                April 8, 1992



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 to potential inadequate performance of carbon 
dioxide (CO2) and Halon fire suppression systems caused by excessive leakage 
from the protected enclosure or by deficient operation of the system's 
components.  Limited acceptance testing may not be adequate to identify 
these problems.  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.


In Section 50.48 of Title 10 of the Code of Federal Regulations, the NRC 
established fire protection requirements for operating nuclear power plants.  
This rule requires automatic and manual fire suppression systems to function 
so that the capability to safely shut down the plant is ensured.  Many 
licensees use total flooding CO2 and Halon fire suppression systems to 
protect systems necessary for safe shutdown.  In Branch Technical Position 
APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," the 
staff referenced National Fire Protection Association (NFPA) standards, NFPA 
12-1973, "Carbon Dioxide Extinguishing Systems," and NFPA 12A-1973, "Halon 
1301 Fire Extinguishing Systems."  These standards emphasized the need to 
minimize leakage from the enclosure in order to retain the fire suppressing 
agent for the required soak time and the importance of thoroughly inspecting 
the fire suppression system to ensure that it will operate properly.  
Licensees frequently use full discharge tests to demonstrate that fire 
suppression systems perform properly and that leakage from protected 
enclosures is acceptable.

Description of Circumstances

On February 23, 1988, the Connecticut Yankee Atomic Power Company, the 
licensee for the Haddam Neck Power Plant, performed a full discharge test of 
the CO2 fire suppression system for the containment cable vault.  The test 
results indicated that the CO2 concentration within the cable vault failed 
to meet NFPA 12 requirements.  Consequently, on February 27, the licensee 
declared the 


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fire suppression system for the cable vault inoperable.  The licensee deter-
mined that the root cause of the failure was excessive leakage of CO2 from 
the enclosure area through numerous unsealed electrical conduits in the 
lower level of the cable vault.  These conduits were in the original plant 
design, but were not considered in the design of the CO2 system.

While performing an inspection the week of April 3, 1989, at the Susquehanna 
Steam Electric Station (Susquehanna), the NRC found a concern regarding the 
adequacy of initial testing of the plant's CO2 fire suppression systems.  In 
1982, the Pennsylvania Power and Light Company (PP&L), the licensee for 
Susquehanna, had performed a full discharge test for one of seven areas pro-
tected by automatic CO2 fire suppression systems.  The test found that the 
required concentration of CO2 was not maintained in the enclosure for the 
required soak time.  The test results may have been caused solely by the 
failure of a temporary seal around an access door.  However, the licensee 
did not perform additional testing to confirm the cause of the test failure.  
The licensee then performed limited acceptance tests of the CO2 fire 
suppression systems. 

To address the NRC's concern, PP&L performed testing in the first quarter of 
1990 using room pressurization to measure enclosure leakage and to determine 
a projected agent retention time.  The licensee based the testing on the 
enclosure integrity procedure in Appendix B to NFPA 12A-1989.  The test 
results indicated that three of the seven areas included enclosures with 
leakage greater than that which would ensure retention of the required CO2 
concentration for the required soak time.  The failure of these enclosures 
was attributed to their small enclosed volume and the corresponding small 
allowable leakage area.  In general, a smaller allowable leakage area should 
be expected for small enclosures because of the higher ratio of boundary 
area to enclosed volume.  

On April 21, 1990, at the Catawba Nuclear Station (Catawba), an inadvertent 
steam release actuated a CO2 fire suppression system.  Although the fire 
suppression system is designed to discharge to only one area at a time, the 
three selector pilot valves installed in the system directed the CO2 
discharge to all three areas protected by the system.  Duke Power Company, 
the licensee for Catawba, investigated the incident and discovered that the 
solenoids operating the three selector pilot valves were installed 
backwards.  The licensee determined that the required CO2 concentration 
could not be obtained within the protected areas when the system discharged 
into more than one area at a time.  Therefore, the licensee declared the 
system inoperable.  The licensee attributed the improper solenoid 
installation, in part, to a preoperational test procedure which did not 
adequately test the system for the incorrectly installed components. 


Retaining an adequate concentration of fire suppressing agent for the 
required soak time is important for enclosures containing equipment that 
could develop "deep seated" fires.  In a study of deep seated cable fires, 
Sandia National Laboratory determined that, for certain configurations of 
cables qualified to Standard 383 of the Institute of Electrical and 
Electronic Engineers (IEEE), it 

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was necessary to retain a 50% concentration of CO2 for a minimum soak time 
of 15 minutes to extinguish fully developed fires.  Sandia National 
Laboratory documented the results of the study in NUREG/CR-3656, "Evaluation 
of Suppression Methods for Electrical Cable Fires," dated October 1986.  

Full discharge testing of CO2 fire suppression systems may present certain 
hazards at operating nuclear power plants.  These hazards include thermal 
shock to safety-related components, uncontrolled electrostatic discharge, 
and hazards to personnel from high concentrations of CO2.  Some licensees 
have used alternative testing methods which avoid these hazards.  For 
example, the licensee for the Vermont Yankee Atomic Power Station responded 
to the NRC's concern regarding the adequacy of initial tests of the plant's 
fire suppression systems by performing an alternative test that incorporated 
methodology from the enclosure integrity procedure in Appendix B to NFPA 
12A-1989.  That methodology is conservative because the effects of the 
thermal expansion of the mixture of CO2 and air are not included and a 
"worst case" distribution of measured leakage area is assumed.  The licensee 
also performed a rigorous engineering evaluation of the installed CO2 system 
to verify that the system would operate as designed to deliver a sufficient 
amount of CO2.

The testing described in Section 1-7.4 of NFPA 12A-1989 was developed to 
alleviate concerns for both the cost and the environmental damage associated 
with repeatedly performing full discharge tests of Halon fire suppression 
systems.  The testing described in NFPA 12A provides an alternative method 
to full discharge testing of Halon systems to demonstrate that the fire 
suppression system and the enclosure function as designed. 

This information notice requires no specific action or written response.  If 
you have any questions about the information in this notice, please contact 
the technical contact 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 contact:  S. R. Jones, NRR
                    (301) 504-2833 

Attachment:  List of Recently Issued NRC Information Notices

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