Information Notice 2017-04, High Energy Arc Faults in Electrical Equipment Containing Aluminum Components
| ML17058A343 | |
| Person / Time | |
|---|---|
| Issue date: | 08/21/2017 |
| From: | Louise Lund, Mcginty T Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking |
| To: | |
| Harris B, NRR/DPR/PGCB, 301-415-2277 | |
| References | |
| TAC MF9285 IN-17-004 | |
| Download: ML17058A343 (14) | |
ML17058A343 UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
OFFICE OF NEW REACTORS
WASHINGTON, DC 20555-0001
August 21, 2017
NRC INFORMATION NOTICE 2017-04: HIGH ENERGY ARCING FAULTS IN ELECTRICAL
EQUIPMENT CONTAINING ALUMINUM
COMPONENTS
ADDRESSEES
All holders of an operating license or construction permit for a nuclear power reactor under
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of
Production and Utilization Facilities, except those who have permanently ceased operations
and have certified that fuel has been permanently removed from the reactor vessel.
All holders of an operating license for a non-power reactor (research reactor, test reactor, or
critical assembly) under 10 CFR Part 50, Domestic Licensing of Production and Utilization
Facilities, except those who have permanently ceased operations.
All holders of and applicants for a combined license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
addressees of operating experience and recent NRC testing results pertaining to the magnitude
of arc fault hazards in electrical equipment containing aluminum components. The NRC
expects that recipients will review the information for applicability to their facilities and consider
actions, as appropriate. The suggestions in this information notice are not NRC requirements;
therefore, the NRC requires no specific action or written response.
DESCRIPTION OF CIRCUMSTANCES
NRC-Led International Testing Program Insights
From 2014 to 2016, the Office of Nuclear Regulatory Research (RES) collaborated in a testing
program with the Nuclear Energy Agency (NEA), the National Institute of Standards and
Technology, and various international partners through the Organization for Economic and
Cooperative Development (OECD). The purpose of the testing program was to confirm the
zone of influence (ZOI) for high energy arcing fault (HEAF) events that was specified in
NUREG/CR-6850 EPRI 1011989 EPRI/NRC-RES Fire PRA Methodology for Nuclear Power
Facilities, Volume 2: Detailed Methodology (Agencywide Documents Access and Management
System (ADAMS) Accession No. ML15167A411). RES initiated the testing program after
reviewing international HEAF events documented in the OECD Fire Project - Topical Report
No.1, Analysis of High Energy Arcing Fault (HEAF) Fire Events, published in June 2013 (https://www.oecd-nea.org/nsd/docs/2013/csni-r2013-6.pdf). The testing protocol initiated the arc by installing a bare copper wire across the three phases of
power in accordance with IEEE C37.20.7-2007/Cor 1-2010, Guide for Testing Metal-Enclosed
Switchgear Rated up to 38 kV for Internal Arcing Faults Corrigendum 1. No accelerants were
used to initiate the fault or ensuing fire conditions. The testing was performed using actual
electrical equipment from nuclear power plants. There were four primary controlled parameters
that were varied as part of the testing programarc location, fault current, voltage and fault
duration:
1. Arc location - operating experience from HEAF events was used to select representative
arc locations within the enclosure. Typically the arc locations were across breaker
stabs, at enclosure power supply entry locations, or along bolted connections within the
enclosure that could be subject to fatigue failures.
2. Fault current - the arc fault currents were selected to replicate fault capacities of typical
electrical distribution systems within nuclear power plants.
3. Voltage - The voltages were selected from typical voltages used in nuclear power
plants power distribution systems. The program tested both low and medium voltage
equipment.
4. Arc duration - The duration selected for the HEAF tests was a controlled parameter and
based on a review of operating experience where arc duration information was readily
available from fault recording devices, or could be determined from post arc fault
investigation.
This program has shown that HEAF tests involving aluminum resulted in a significantly larger
release of energy than HEAF tests involving copper. This aluminum involvement includes
components, subcomponents, or parts that form part of the normal current-carrying pathway
(such as bus bars, breaker armatures, contacts, cable, etc.), or components, subcomponents, or parts that could become involved in the fault current pathway as a result of a ground fault
(housings, structural framework, adapters, cradles, wireways, conduits, draw-out or racking
mechanisms, etc.). In addition to larger energy release during the HEAF event when aluminum
is involved, RES staff also observed dispersal of electrically conductive aluminum byproducts
throughout the area. This byproduct was conductive and caused short circuiting and grounding
of electrical equipment in the area. Through the testing program, RES staff observed that HEAF
tests involving aluminum damaged test measurement and recording equipment and the
electrical supply of the test facility well beyond the damage limits approximated in
NUREG/CR-6850 (EPRI 1011989).
A total of 26 tests were performed, consisting mostly of electrical equipment that contained
copper conductors. The equipment with copper components exhibited similar damage states to
those postulated in the current methodology presented in NUREG/CR-6850, Appendix M.
However, results obtained for equipment containing aluminum components exhibited damage
states well beyond current HEAF damage models. These results are summarized as follows:
Test 23 of the HEAF experimental series consisted of a low-voltage, 480 VAC
switchgear cabinet that contained an aluminum bus bar. The test resulted in failure of
the cabinet walls such that the plasma created by the arcing fault was expelled a
significant distance out of the cabinet. Inconel temperature sensors located
approximately 0.9 m (3 ft.) from the enclosure (which is the NUREG/CR-6850 horizontal
ZOI limit) were melted and destroyed during the test. The vapor/plasma cloud migrated
well beyond the current limit of 0.9 m (3 ft.) horizontal and 1.5 m (5 ft.) vertical distance
for the ZOI. In addition to the physical damage caused by the HEAF event, there was a
layer of aluminum byproducts that coated structures and components within the test cell. This aluminum byproduct layer caused shorting problems in the test facility electrical
power system and required major repair. This test is an outlier from the other
low-voltage tests performed as part of the experimental program. It was the only low
voltage experimental test to exhibit a level of damage beyond the NUREG/CR-6850,
Appendix M, ZOI methodology. In all other low-voltage tests, the temperature sensors
located 0.9 m (3 ft.) horizontally from the cabinet were not damaged, recorded data, and
were able to be used in subsequent experimental tests.
Test 26 of the HEAF experimental series involved a 4.16 kV bus duct section removed
from a decommissioned United States nuclear plant. The bus duct section consisted of
non-segregated copper bus bars enclosed within an aluminum duct. The bus duct
section was secured to the floor of the test enclosure using wooden structural members
and the open ends were secured with a sheet of electrically insulating fiber board. This
piece of fiber board was installed to create a pressure boundary within the bus duct unit
and limit the free release of energy once the fault was initiated. Immediately upon
initiating the HEAF event, the hot gas mixture forced the fiber board away, allowing the
arc and associated hot gas/plasma mixture to jet out of the bus duct, blowing hot
gas/molten metal linearly from the bus duct opening. The hot gases/plasma extended
approximately 9 m (30 ft.) horizontally. The instruments associated with temperature
and heat flux, located approximately 0.9 m (3 ft.) away from the duct, were damaged.
The test facility was coated with an electrically conductive aluminum byproduct and
several electrical components in the test cell required replacement.
The increased physical damage to the test specimens, measurement devices, and the testing
facility observed during tests 23 and 26 can be attributed to the involvement of aluminum during
the HEAF. The presence of aluminum in the components, subcomponents, or parts that form
part of the normal current carrying pathway causes a more energetic plasma development when
consumed during the arcing process. The increased energetic plasma causes a larger amount
of cabinet damage and/or the transport of gaseous high energy particles/plasma farther than
previously assumed. These two tests included aluminum components which resulted in
exceeding the ZOI documented in NUREG/CR-6850 and its Supplement 1. In both tests, a
major impact from the HEAF was a layer of aluminum byproducts that plated out on most
surfaces within the test enclosure, including electrical equipment. This aluminum byproduct
layer caused shorting problems in the test facility electrical power supply and required a major
repair and cleanup activity. The extent of damage observed from tests 23 and 26, which
contained aluminum components, far exceeded the damage observed from other tests which
did not contain aluminum components.
The OECD/NEA test report is publically available and can be downloaded at
https://www.oecd-nea.org/nsd/docs/indexcsni.html NEA/CSNI/R (2017) 7-Report on the Testing
Phase (2014-2016) of the High Energy Arcing Fault Events (HEAF) Project: Experimental
Results from the International Energy Arcing Fault Research Programme.
NRC Generic Issue Program
On May 6, 2016, RES submitted a possible generic issue involving an identified vulnerability of
electrical equipment containing aluminum components during a HEAF event (ADAMS
Accession No. ML16126A091). The issue was entered into the generic issue process on May
12, 2016, (ADAMS Accession No. ML16132A116) and an initial evaluation was performed
(ADAMS Accession No. ML16064A250). On May 20, 2016, a generic issue review panel was
formed (ADAMS Accession No. ML16209A007). The committee is currently in the screening phase, in
accordance with the NRCs Management Directive 6.4, Generic Issues Program (ADAMS
Accession No. ML14245A048).
United States Nuclear Power Plant Operating Experience
Fort Calhoun Station, Unit 1
On June 7, 2011, a switchgear fire occurred at the Fort Calhoun Station while the plant was
shut down for a planned refueling outage. The fire resulted in a loss of power to six of nine
safety-related 480 VAC electrical distribution buses, one of two safety-related 4160 VAC buses
and one of two non-safety related 4160 VAC buses. The event resulted in the loss of the spent
fuel pool cooling function and could have resulted in the loss of a safety function or multiple
failures in systems used to mitigate a situation had the event occurred at power. Significant
unexpected system interactions also occurred. Specifically, combustion products from the fire
caused a fault across an open bus-tie breaker on an island bus; as a result a feeder breaker
tripped unexpectedly resulting in loss of power to the opposite train bus. Also, the event
resulted in grounds on both trains of safety-related direct current (DC) power used for breaker
operation and electrical protection.
Fort Calhoun Station originally used General Electric AKD-5 Powermaster Low Voltage Drawout
Switchgear, with a welded aluminum bus bar structure that transitioned to copper bus stabs in
each breaker cell. The original AK-50 circuit breakers connected directly to the silver-plated
areas on the line and load stabs. The new Nuclear Logistics Incorporated/Square-D circuit
breaker design was an integrated unit consisting of a circuit breaker and cradle assembly. The
cradle assembly converted the internal vertical breaker connectors in the new breaker to top
and bottom spring-loaded horizontal finger assemblies which connected to the original
switchgear bus stabs.
The fire was caused by the catastrophic failure of the feeder breaker for 480 VAC load center
1B4A in the west switchgear room. A large quantity of soot and smoke was produced by the fire
which migrated into the conducting connections associated with the non-segregated bus duct (a
metal enclosure containing the bus bars for all three electrical phases) connecting to island bus
1B3A-4A, even though the bus-tie breaker was open. The smoke and soot were sufficiently
conductive that arcing occurred between the bus bars such that island bus 1B3A-4A was
affected and the other connected train load center, 1B3A, was affected by incorrect breaker
sequencing.
The licensee's root cause investigation concluded that the initial fire in the load center was the
result of high resistance connections on the line side of the feeder breaker cubicle, which
caused overheating and failure of the cradle finger clusters, resulting in bus grounding and
phase-to-phase shorting. The investigation also determined that combustion products from the
fire caused a fault on the island bus side of the bus-tie breaker (BT-1B4A), which resulted in an
overcurrent condition through two breakers (feeder breaker 1B3A and bus tie breaker BT-
1B3A). The licensees initial investigation determined that breaker 1B3A tripped prematurely on
a short-time overcurrent fault and concluded that since the breaker protection scheme did not
operate as designed, bus 1B3A de-energizing resulted in the loss of multiple electrical power
distribution system trains from a single fire. Appropriate breaker coordination was required by
the licensees fire protection program to ensure that the plant could achieve and maintain safe
shutdown conditions following a fire. The maximum conditional core damage probability for the event was estimated to be
3.4 x 10-4. At the time of the fire, Fort Calhoun Station was experiencing impacts from flooding
of the Missouri River and had declared a Notice of Unusual Event on June 6, 2011. Additional
information is available in Licensee Event Report (LER) 2011-008, Revision 1, for the Fort
Calhoun Station, Fire in Safety Related 480 Volt Electrical Bus (ADAMS Accession
No. ML113010208), and Fort Calhoun Station - NRC Special Inspection Report 05000285/2011014; Finding of Preliminary High Safety Significance (ADAMS Accession
No. ML12072A128).
Columbia Generating Station
On August 5, 2009, at 7:49 a.m., while operating at 100 percent rated thermal power, the main
generator experienced a differential current lockout. This resulted in a main generator overall
differential lockout trip, causing a turbine trip and a fast transfer from the output of the main
generator to the startup transformer (normal breakers opened and start up breakers closed). As
a result of the turbine trip with reactor power above 30 percent, an automatic reactor scram
occurred.
At 7:51 a.m., the control room received a report from the turbine building of a fire in the electrical
bus duct above switchgear SM-3. This resulted from a fault on the non-segregated 6900 VAC
bus that supplies power from the normal auxiliary transformer to switchgears SH-5 and SH-6.
The bus catastrophically failed and vaporized all three phases of the conductor in the location of
a flexible bus connection link. The faulted bus section was located above switchgear SM-3.
Operators notified the fire brigade to respond. At 7:53 a.m., operators closed the inboard main
steam isolation valves to halt the reactor vessel depressurization and cooldown, however, the
technical specification cooldown limit of 100°F was exceeded. Operators called the offsite fire
department for assistance due to heavy smoke in the turbine building.
The licensees root cause evaluation determined that the fault was located on the
non-safety-related, non-segregated, 6900 VAC N2/X bus. The bus catastrophically failed, melting all three phases of conductor in the location of a flexible link. The bus bars are
completely enclosed in aluminum ducts, with all three phases in the same duct
(non-segregated). The bars were a combination of aluminum square hollow tubing and copper
connection plates and were mounted on porcelain insulators with spring washers. The bars
were assembled together using a series of rigid and flexible connections.
The faulted bus section was located above the medium voltage switchgear SM-3 and damage
from the molten metal produced by the fault included another high voltage bus and other cables
in the area of switchgears SH-5 and SH-6. The licensee inspected the damage and noted the
explosion melted and removed approximately 1.2 m (4 ft.) of each of the three buses and 2.4 m
(8 ft.) of the bus duct enclosure. Both the bus and the bus duct enclosure were made of
aluminum. This bus duct was located at approximately 4.6 m (15 ft.) above finished floor. The
molten aluminum and debris was thrown in the general vicinity. Photographs revealed that a
circular pattern of material remained. The licensee evaluation concluded that the fault likely
originated on the center flexible link of the bus connection. However, the licensee concluded
that the root cause of the failure was indeterminate, because the catastrophic failure of the bus
destroyed any evidence that would have provided an indication as to the cause. The remaining
bus duct showed signs of high heat exposure. Observed from the ground were smoke streaks at all metal joints and covers for a distance of 6 m (20 ft.) south and about 3 m (10 ft.) north of
the missing section.
The licensee determined that the most probable cause of the event was the
relaxation/loosening of a bolted connection on the flexible link from repeated thermal cycles
over time. The relaxation/loosening of the bolted connection resulted in degradation and
overheating of the connection. The insulation continued to degrade to the point where a short
occurred between two phases of the bus. The short destroyed the bus and the surrounding bus
enclosure. The molten aluminum and copper splattered nearby switchgear cabinets but did not
cause any internal damage. A possible contributor to the non-segregated bus failure was
attributed to running the bus near its original rating of 2500 amps. The nominal bus loading at
90 percent capacity was closer to the rating than the other four buses, and, therefore, could
have caused higher temperatures and connection degradation. The licensee performed a bus
uprate analysis in 1994 and increased the rating to 2806 amps to account for the worst case
loading of 2610 amps.
Additional information is available in LER 2009-004-00, for Columbia Generating Station, 6.9 kV Non-Segregated Electrical Bus Failure (ADAMS Accession No. ML092870468), and
Columbia Generating Station - NRC Special Inspection Report 05000397/2009010 (ADAMS
Accession No. ML093280158).
Diablo Canyon Nuclear Power Plant, Unit 1
On May 15, 2000, at 12:25 a.m., at 100 percent power, Diablo Canyon, Unit 1 experienced a
turbine/reactor trip. The cause of the unit trip was an electrical phase-to-phase fault on the
12 kV bus in an overhead bus duct supplied by auxiliary transformer 1-1. The switchyard and
main generator field breaker opened immediately following the unit trip; however, coast down of
the main generator continued to feed the arc fault. The fault and subsequent arcing damaged a
4 kV startup bus duct located immediately above the faulted 12 kV bus. Damage to the 4 kV
bus induced a second arcing fault in the 4 kV bus duct resulting in a differential trip of startup
transformer 1-2 that occurred 11 seconds after the initial fault. The loss of both offsite sources
of power to all 4 kV loads resulted in an undervoltage condition that caused the emergency
diesel generators to start and load successfully.
The cause of the bus failure could not be conclusively determined due to the absence of
physical evidence. The immediate cause was postulated to be a thermal failure of the bolted
connection of the center conductor of the 12 kV bus. The faulted nonsegregated bus duct
consisted of three conductors of aluminum in a sealed bus duct enclosure. The bus
connections consist of silver-plated splice plates bolted to the silver-plated bus bars and
insulated with polyvinyl chloride (PVC) boots or Raychem tape. It is postulated that the PVC
boot failed due to the excessive heat and created smoke. The radiant heat from the center
conductor caused the insulation to fail on adjacent conductors. The smoke provided an ionized
environment for a phase-to-phase arc from the center conductor to the south conductor.
Subsequently, there was arcing between all conductors.
Prior to returning Unit 1 to service, the licensee performed extensive inspections, tests, maintenance and repairs, including an effort to increase the capacity by replacing the 12 kV
aluminum bus conductors inside the turbine building with copper, from where the bus entered to
pass the damaged section. The licensee upgraded buses with little design margin from
aluminum to copper and installed large copper splice plates in select locations. Corrective
actions included a new preventative maintenance program and upgrades to the 4 kV and 12 kV nonsegregated buses on both units. Additional information is available in LER 1-2000-004-00,
Unit 1 Unusual Event Due to a 12 kV Bus Fault (ADAMS Accession No. ML003725220).
Zion Nuclear Power Station, Units 1 and 2
On April 3, 1994, at 4:30 a.m., Zion Nuclear Power Station, Unit 1 was synchronized to the
electrical distribution system after the completion of refueling outage Z1R13. The unit was
stabilized at 25 percent reactor power at 5:11 a.m. At 6:13 a.m., a 3 percent load increase was
initiated as required by fuel conditioning guidelines. At 6:18 a.m., a loud noise emanating from
the vicinity of the main generator was heard in the control room. Almost simultaneously, a main
generator lockout trip and subsequent turbine and reactor trips occurred with the main
generator tripping on differential current.
At 6:19 a.m., the operating personnel received a report of a fire emanating from the generator
lead box. The station fire alarm was sounded and the fire brigade was dispatched to extinguish
the fire. The shift engineer was notified at 6:30 a.m. that the fire brigade could not contain the
fire. The plant then requested offsite fire department assistance.
After the termination of the fire, the licensee performed an assessment of the damage to the
main generator and associated bus ducts. The A and B phase isophase bus ducts showed
signs of excessive arcing. The corners of the 90-degree turns on both phase housings were
blown outward, and aluminum spatter covered the general area of the fault. Large amounts of
white powder were also found in the A and B phase duct work (later determined to be aluminum
oxide). The conductors on both phases were also damaged. The bus duct cooling boundaries
for the A and B phases at the insulator penetration on the generator end were blown inward.
The retaining clamps for the A and B phase housings, located at the generator end, appeared to
have separated. One broken and several damaged stand-off insulators were also found. The C
phase bus duct sustained only minor damage in this event. An internal inspection of the
C phase bus duct, conducted by the licensee, found an oil film on the first and second standoff
insulators from the generator end.
The licensee also inspected the generator lead box and surrounding areas. The licensee
determined that the C phase bushing was severely damaged at the isophase shorting plate and
internal to the lead box. The A and B phase bushings sustained damage external to the lead
box. Molten metal was found in the area where the C phase bushing exits the lead box. The
micarta cleats which separate the generator from the lead box were blown upward into the
generator. Evidence of arcing between the top flange of the C phase bushing and the structural
ribbing of the lead box was discovered. Soot covered the inside of the lead box and appeared
to cover the lower third of the generator stator. The licensee also found evidence of arcing on
the neutral bus enclosure in the vicinity of the generator lead box.
The licensee determined that the root cause of the electrical fault was a main generator phase
C bushing failure. The licensee investigation found that the inner ceramic surface inside the
bushing insulator had been contaminated with an oily substance, and this contamination lead to
the failure. The high current arc from the C phase to the lead box resulted in the rupture of the
bushing internal and external to the lead box. This allowed hydrogen to escape into the
atmosphere. The hydrogen was ignited by the flashover and started to burn. The fault currents
generated were significant enough in magnitude to cause the fault damage observed on the
A phase winding of the west main power transformer. Although this scenario provides an
explanation for the high side neutral residual current recorded during the event, the physical damage observed during the internal inspection of the transformer is not of the magnitude seen
on other documented failures of this nature.
Samples of molten metal and fire residue obtained from the A and B phase bus ducts and the
C phase lead box were analyzed. The intent of this examination was to identify the presence of
any conductive foreign material which may have contributed to the flashover of the A and B
phases. Nothing out of the ordinary was found in this investigation. The majority of the material
examined was identified as aluminum. The white powder found in the ducts was identified as
aluminum oxide. The aluminum deposits were a result of the arcing that occurred on the A and
B phases, which are fabricated from aluminum. Additional information is available in LER
94-005-01 Unit 1 Main Reactor Trip and Subsequent Start of 1A Auxiliary Feedwater Pump
Following a Generator Fire and Isophase Bus Duct Fault (Legacy ADAMS Accession
No. 9801210070).
Shearon Harris Nuclear Power Plant, Unit 1
On October 9, 1989, at 11:05 p.m., with the plant at full power, a turbine/reactor trip occurred at
Shearon Harris Nuclear Power Plant, Unit 1, due to an electrical fault. The reactor was
stabilized in hot standby. The licensee determined that the initiating cause of the turbine/reactor
trip was multiple ground faults. The ground faults destroyed the neutral grounding bus and
caused three fires: (1) an oil fire at the B main power transformer; (2) a hydrogen fire
underneath the main generator; and (3) a small oil fire in the generator housing. The licensee
declared an alert and the site fire brigade responded to the fires. The site fire brigade was later
assisted by offsite fire departments.
The licensee determined that the ground faults were caused by aluminum debris carried down
the bus duct by the forced air cooling system. The aluminum debris entered the bus duct as a
result of previous damper failures in the bus duct cooling system in 1988 and 1989. Arcing
between the conductor and the enclosure occurred over a 15 m (50 ft.) length of the A phase
bus immediately upstream of the B main power transformer. Ionization from this arcing reduced
the dielectric strength of the cooling air which was carried into the bushing box of the B main
power transformer. This caused an A phase to ground flashover in the bushing box, which
immediately propagated to the B phase bushing. The fault cracked both low voltage bushings, causing oil to leak from the bushings and ignited the leaking oil.
The fault at the B main power transformer created magnetic forces in the main section of the
isolated phase bus that damaged insulators in A and B phase. The 58 cm (23 in.) diameter
tubular conductor is suspended in the middle of the 104 cm (41 in.) diameter aluminum
enclosure by the ceramic insulators. In the A bus duct, five insulators pulled apart, allowing the
conductor to come in contact with the grounded enclosure, creating a phase-to-ground fault.
As part of the corrective actions, the licensee evaluated the design of the isolated phase bus
duct bus supports and found it to be adequate. However, the sections in which the conductor
supports failed were rotated 180 degrees to place the supports underneath the conductor as a
prudent design measure. The licensee also revised the design of the isolated phase bus duct
cooling system to preclude debris intrusion into the ducts. The licensee has replaced the
damaged duct. Additional information is available in the Shearon Harris Nuclear Power Plant
LER 89-017-01, Electrical Fault on Main Generator Output Bus Causing Plant Trip and Fire
Damage in Turbine Building (Legacy ADAMS Accession No. 8912130259). Kewaunee Power Station
On July 10, 1987, at 11:43 a.m., with the plant at full power, a reactor/turbine trip occurred due
to an undervoltage transient on the 4160-volt electrical buses, which supply the reactor coolant
pumps and main feedwater pumps.
The control room operators received an unexpected reactor/turbine trip. Following normal
practice, the plant dispatched an equipment operator to verify that the auxiliary feedwater
pumps and diesel generators had started and were operating normally. Upon investigation, the
equipment operator noticed smoke and fire coming from the vicinity of the electrical bus bar
located on the eastern end of the turbine building, and immediately notified the shift supervisor.
The bus fire terminated once the transformer was de-energized. A smaller fire occurred on a
maintenance cart involving rags and rubber goods ignited by falling aluminum slag. The
equipment operator quickly extinguished this fire. At 11:46 a.m., the equipment operator
reported to the control room that the fire had been extinguished.
The licensee determined that the event was caused by a phase-to-ground fault which occurred
on the bus bar (1.3 cm by 10.2 cm [0.5-inch by 4-inch] flat aluminum bus bar rated at
3000 amps) routed from the main auxiliary transformer to buses 1-3, 1-4, 1-5 and 1-6. The fault
caused an undervoltage condition resulting in a reactor trip and subsequent turbine trip. The
bus undervoltage reactor trip is set at less than 77 percent of the rated bus voltage for greater
than 0.1 second. Plant systems performed as designed. The fast transfer prevented the
reactor coolant pumps from tripping on the undervoltage condition because of the time delay of
approximately 5 seconds in the pump breaker trip logic. A time delay relay of approximately
5 seconds is designed in the reactor coolant pump logic to prevent tripping of the pumps before
fast transfer to an offsite power supply.
The licensee determined that the root cause of the event was an insulation failure on the bus
bar compounded by accumulation of particulate debris. The bus bar runs perpendicular to
turbine building ventilation fans mounted in the turbine building exterior wall. Suction from the
fan pulled dust-filled air through the bus duct. Dust and metallic powder debris collected on the
outside of the bus bar insulation. The insulation failure combined with the accumulated dirt
provided a tracking path from phase to ground. The phase-to-ground fault progressed to a
phase-to-phase fault which accounted for the extensive bus damage. The event damaged a
9 meter (30 ft.) section of the bus bar from the main auxiliary transformer to buses 1-3, 1-4, 1-5 and 1-6. The equipment operator quickly extinguished the related fire. There was no other
equipment damage as a result of this event.
The damaged section of bus bar was isolated. The bus bar and main auxiliary transformer were
inspected to determine the extent of damage, and a temporary change request was
implemented to supply electrical buses 1-3 and 1-4 from the reserve auxiliary transformer. The
main auxiliary transformer supplied buses 1-1 and 1-2. Safeguards buses 1-5 and 1-6 remained in their normal lineup. In addition to the immediate corrective actions taken to replace
damaged components, the licensee inspected and cleaned various sections of the bus bar
ducting. The licensee implemented a triennial inspection procedure covering all bus bar
ducting. Additional information is available in LER 87-009-00, Electrical Bus Bar Failure
Causes Undervoltage on RXCP Buses and Reactor Trip (ADAMS Accession
No. ML111661536).
BACKGROUND
HEAF events have occurred in both United States and foreign nuclear power plants. HEAFs
are energetic and explosive faults in electrical equipment characterized by a rapid release of
energy in the form of heat, light, vaporized metal, and pressure increase due to high current
arcs between energized electrical conductors or energized conductors and ground or neutral.
HEAFs have the potential to cause extensive damage to the failed electrical components and
distribution systems along with adjacent equipment and cables. Furthermore, the significant
energy released during a HEAF event can act as an ignition source of nearby combustibles
resulting in an ensuing fire, which could potentially affect the performance of nearby structures, systems, and components (SSCs) important to safety.
The current method for estimating the damage from a HEAF for electrical enclosures is
contained in an industry/NRC fire probabilistic risk assessment methodology document, EPRI
1011989/NUREG/CR-6850 Appendix M. This method is based on a single well-documented
event that occurred at San Onofre Unit 3 in 2001. NUREG/CR-6850 Supplement 1 FAQ
07-0035 provides the current methodology for estimating damage specific to bus duct failures.
Recent international HEAF testing has been performed to better characterize the damage
potential and modes of failure. RES initiated this testing project after investigating international
HEAF events documented in the OECD Fire Project - Topical Report No.1, Analysis of High
Energy Arcing Fault (HEAF) Fire Events, NEA/CSNI/R (2013) published in June 2013 (https://www.oecd-nea.org/nsd/docs/2013/csni-r2013-6.pdf). The experimental test program
identified that the presence of aluminum can cause a more energetic plasma arc and
subsequent metal fire. Under some circumstances this may cause a larger amount of cabinet
damage and/or cause the transport of conductive aluminum byproducts further than previously
assumed.
Each operating nuclear power plant is required to have a fire protection plan that satisfies
General Design Criteria (GDC) 3 in Appendix A of 10 CFR Part 50. This is to ensure that SSCs
important to safety are designed and located to minimize the probability and effects of fires and
explosions. Licensees who have chosen to voluntarily adopt a risk-informed, performance-based fire protection program in accordance with 10 CFR 50.48(c) have typically
evaluated the hazard posed by HEAF events through the use of the EPRI/NRC-RES fire
Probabilistic Risk Assessment document, EPRI 1011989/NUREG/CR-6850. Licensees
adhering to a deterministic fire protection program either meet the prescriptive requirements of
Section III.G in Appendix R of 10.CFR 50.48(b) (or corresponding NRC staff guidance in
NUREG-0800) and approved exemptions, or have a NRC staff-approved safety evaluation
report on the fire protection program.
Related NRC Generic Communications and Documents:
The following NRC generic communications contains additional operating experience
addressing general electrical distribution equipment complications (non-aluminum related):
IN 1999-13, Insights from NRC Inspections of Low- and Medium-Voltage Circuit
Breaker Maintenance Programs,dated April 29, 1999 (ADAMS Accession
No. ML031040447).
IN 2000-14, Non-Vital Bus Fault Leads to Fire and Loss of Offsite Power, dated
September 27, 2000 (ADAMS Accession No. ML003748744). *
IN 2002-01, Metalclad Switchgear Failures and Consequent Losses of Offsite Power, dated January 8, 2002 (ADAMS Accession No. ML013540193).
IN 2002-27, Recent Fires at Commercial Nuclear Power Plants in the United States, dated September 20, 2002 (ADAMS Accession No. ML022630147).
IN 2005-21, Plant Trip and Loss of Preferred AC Power from Inadequate Switchyard
Maintenance, dated July 21, 2005 (ADAMS Accession No. ML051740051).
IN 2005-15, Three-Unit Trip and Loss of Offsite Power at Palo Verde Nuclear
Generating Station, dated June 1, 2005 (ADAMS Accession No. ML050490364).
IN 2006-18, Supplement 1, Significant Loss of Safety-Related Electrical Power at
Forsmark Unit 1 in Sweden, August 10, 2007 (ADAMS Accession No. ML071900368).
IN 2006-31, Inadequate Fault Interrupting Rating of Breakers, dated
December 26, 2006 (ADAMS Accession No. ML063000104).
IN 2007-14, Loss of Offsite Power and Dual-Unit Trip at Catawba Nuclear Generating
Station, dated March 30, 2007(ADAMS Accession No. ML070610424).
NUREG/IA-0470, Volume 1 Nuclear Regulatory Authority Experimental Program to
Characterize and Understand High Energy Arcing Fault (HEAF) Phenomena, dated
August 2016 (ADAMS Accession No. ML16235A163).
DISCUSSION
Title 10 CFR Part 50, Appendix A, GDC 3, Fire Protection states, in part, Structures, systems
and components important to safety shall be designed and located to minimize, consistent with
other safety requirements, the probability and effect of fires and explosions
Improper operation and maintenance practices represent the most significant contributors to
HEAF events. Additional contributing causes include the use of improper tools and equipment
during maintenance, introduction of conductive foreign material, insulation breakdown due to
aging, failure of tripping mechanisms, loose bolted connections due to vibrations or inadequate
maintenance practices, corrosion, improper work practices due to inadequate training, and a
lack of preventative maintenance. Any of these contributors could lead to the development of a
low resistance path between conductors or between conductors and ground, creating a HEAF
event.
For plants that have transitioned to a performance-based fire protection program under
10 CFR 50.48(c) and National Fire Protection Association Standard 805, HEAF events
constitute some of the high-risk scenarios. Recent testing suggests HEAFs involving aluminum
components may exceed the ZOI methodology presented in NUREG/CR-6850.
For plants that have not transitioned to a performance-based fire protection program, fire hazard
analyses performed to satisfy 10 CFR Part 50, Appendix A, GDC 3, may not have considered
HEAFs as potential failure modes, including explosions or the dispersion of aluminum
byproducts. These regulations include 10 CFR 50.48(a) and (b) and Appendix R to
10 CFR Part 50, Section III.G (or corresponding NRC staff guidance in NUREG-0800) and/or exemptions/deviations from these requirements granted in accordance with 10 CFR 50.12 or
established in the licensing basis.
The operating experience and recent testing presented in this information notice demonstrates
that the hazards from a HEAF may be substantially greater for electrical equipment that
contains aluminum components than for those with copper components. Additionally, the
operating experience and testing highlights the spread of an electrically conductive coating of
aluminum byproducts which could lead to additional failures. A summary of the aluminum
impact from operational experience is stated below:
The Fort Calhoun event illustrates the adverse effects caused by large quantities of
conductive aluminum byproducts in the smoke produced by HEAF events involving
aluminum which can adversely affect adjacent equipment. The event further resulted in
significant unexpected system interactions. Specifically, loss of power to both train A
and train B buses occurred. Also, the event resulted in grounds on both trains of
safety-related DC power used for breaker operation and electrical protection.
The Columbia event involved aluminum bus bars completely enclosed in aluminum
ducts. The event vaporized approximately 1.2 m (4 ft.) of each of the three buses and
2.4 m (8 ft.) of the bus duct enclosure and smoke and heat effects were observed at all
metal joints and covers for a distance of 6 m (20 ft.) south, and about 3 m (10 ft.) north of
the missing section.
The Diablo Canyon HEAF event damaged both the 12 kV bus duct and the 4 kV bus
duct. Both bus duct conductors were made of aluminum. The event led to the loss of
both offsite electrical sources and the reliance on emergency diesel generators.
The Zion HEAF event initiated a fire that the onsite fire brigade could not control. The
bus duct was made of aluminum and the A and B phase isophase bus ducts showed
signs of excessive arcing. The licensee found extensive aluminum spatter in the general
area of the fault as well as large amounts of white powder that were later determined to
be aluminum oxide. Additionally, the licensee stated that the physical damage observed
during inspections was greater than other documented failures of this nature.
The Shearon Harris event damaged over a 15.2 m (50 ft.) section of the phase A bus.
The bus duct enclosure was made of aluminum. The event also destroyed the neutral
grounding bus and caused three fires: (1) an oil fire at the B main power transformer, (2)
a hydrogen fire underneath the main generator, and (3) a third small oil fire in the
generator housing.
The Kewaunee Power Station HEAF event damaged a 9.1 m (30 ft.) section of the bus
bar and the licensee observed the spread of a metallic dust. The bus bar conductor was
made of aluminum.
As previously discussed, the NRC has entered the issue of high-energy arc faults involving
aluminum into the generic issue process. As part of this process, the NRC staff will obtain
additional data and will perform additional evaluations to inform future agency actions, as
appropriate. As described in a memorandum dated March 4, 2016, the NRC staff had also
performed an initial evaluation and determined that there was no immediate safety concern
(ADAMS Accession No. ML16064A250). Although not explicitly required, the NRC encourages addressees to review the information in
this information notice for applicability and consider actions, as appropriate. Suggestions
contained in this information notice are not NRC requirements; therefore, the agency requires
no specific action or written response.
CONTACT
S
Please direct any questions about this matter to the technical contact(s) listed below or the
appropriate Office of Nuclear Reactor Regulation (NRR) or Office of New Reactors (NRO)
project manager.
/RA/ (Paul G. Krohn for)
/RA/
Timothy J. McGinty, Director
Louise Lund, Director
Division of Construction Inspection
Division of Policy and Rulemaking
and Operational Programs
Office of Nuclear Reactor Regulation
Office of New Reactors
Technical Contact:
Nicholas Melly, RES/DRA
Mark Henry Salley RES/DRA
301-415-2392
301-415-2474
E-mail: Nicholas.Melly@nrc.gov
E-mail- MarkHenry.Salley@nrc.gov
TAC No. MF9285 OFFICE
NRR/DPR/PGCB/LA
RES/DRA/FXHAB
QTE
NRO/DCIP/QVIB-1*
NRR/DRA/APLB
NRR/DRA/D
NAME
ELee
NMelly
TJackson
JRobinson
JGiitter
DATE
03/09/2017
04/03/2017
03/16/2017
06/06/2017
06/13/2017
06/13/2017 OFFICE
RES/DRA/FXHAB
RES/DRA/D*
NRR/DPR/PGCB/PM*
NRR/DPR/PGCB/LA*
NRR/DPR/PGCB/BC
NRO/DCIP/D
NAME
MSalley
MCheok
BHarris
ELee
AGarmoe
TMcGinty
(PKrohn for)
DATE
06/14/2017
07/21/2017
07/25/2017
07/26/2017
08/01/2017
08/02/2017 OFFICE
NRR/DPR/D
NAME
LLund
DATE
08/21/2017