NS OPERATING EXPERIENCE WEEKLY SUMMARY 93-16
April 16 - April 22, 1993
The purpose of the NS Operating Experience (OE) Weekly Summary is to enhance
safety throughout the DOE complex by promoting feedback of operating experience
and encouraging the exchange of information among DOE nuclear facilities. The OE
Weekly Summary is distributed for information only. No specific actions or responses
are required solely as a result of this document.
Readers are cautioned that review of the OE Weekly Summary should not be relied
upon as a substitute for a thorough review of the interim and final Occurrence Reports.
The following events were reviewed during the week of April 16 - 22, 1993.
ITEM PAGE
1. MISPOSITIONED RELAY CAUSES LOSS OF CRITICALITY ALARM
HORN CAPABILITY 2
2. FAILURE TO PROPERLY WEAR RESPIRATOR 3
3. PERSONNEL OVEREXPOSURE FROM MISHANDLING
RADIOGRAPHY SOURCE 4
4. REACTOR SCRAM WHILE PERFORMING ON-THE-JOB TRAINING 5
5. GREASE SOLIDIFICATION CAUSES MOLDED CASE CIRCUIT
BREAKER FAILURE 6
6. FRENCH EXECUTIVES ON TRIAL FOR NUCLEAR ACCIDENT 7
7. COOPER-BESSEMER DIESEL GENERATOR FUEL PUMP 10CFR21
NOTIFICATION 7
8. UNABLE TO VERIFY CRITICALITY SAFETY-LIMIT COMPLIANCE 8
9. LEVEL INSTRUMENTATION INACCURACIES OBSERVED DURING
NORMAL PLANT DEPRESSURIZATIONS 9
10. OPERATOR INCORRECTLY STARTS WRONG EMERGENCY
DIESEL GENERATOR 10
ADDITIONAL INFORMATION RELATED TO FOLLOWUP ACTIVITIES
1. WORKER CONTAMINATIONS DURING CONSTRUCTION ACTIVITIES 11
1. MISPOSITIONED RELAY CAUSES LOSS OF CRITICALITY ALARM HORN CAPABILITY
On April 14, 1993, an instrument technician at Plant 9 on the Fernald Environmental
Management Project (FEMP) site found a mispositioned alarm-timer relay for the
radiation-detection alarm system, resulting in loss of criticality alarm-horn capability
for all of Plant 9 and Plant 81, contrary to facility operational safety requirements.
Automatic, remote, criticality-alarm capability was maintained to the site
communications center. Local criticality-alarm capability was maintained by a warning
bell and lights at the radiation alarm station inside Plant 9. Also, the capability of the
site-wide emergency message system to notify personnel to evacuate the area was
maintained. The technician found the relay removed from its designated position in
an alarm-relay box and lying on the bottom of the box.
The technician located the mispositoned relay while doing a follow-up inspection,
related to a March 31 event, after restoration of plant air to a Plant 9 alarm horn. The
technician installed the relay in the proper location and completed a successful
functional test of the horns. Results of a prior inspection and functional test of the
alarm system by radiation technicians on April 1 showed that all Plant 9 and Plant 81
radiation-detection horns were operable. The system was not designed to signal a
trouble alarm for a missing relay or open relay-box door, nor was it designed for
locking the door. The radiation detection alarm system is part of the site-wide,
emergency-alarm system, safety class A equipment. Plant 9, a former uranium
processing facility, was unoccupied and under modification to change its function to
vitrification. Plant 81, a Resource Conservation and Recovery Act storage facility,
was also unoccupied. Personnel report that prior material characterizations in the
areas of Plant 9 and Plant 81 indicate that a criticality event there is not possible.
(ORPS Report ORO-WMCO-FMPC-1993-0026)
The Office of Nuclear Safety reported the March 31 event, concerning an
unauthorized isolation of plant air to a Plant 9 air horn in Operating Experience Weekly
Summary 93-14 (ORPS Report ORO-WMCO-FMPC-1993-0021). In that event, a Plant 9 horn
was found isolated from the plant air system and the back-up supply nitrogen bottle
was missing.
While loss of alarm horn capability was not a concern for this event, it caused
concerns because a loss of horn capability could have occurred in an area of the site
where a criticality event was possible. Facility personnel have not identified the
cause of the mispositioned relay. They report that radiation technicians routinely
remove radiation-detection-alarm relays when troubleshooting the alarms or doing
routine inspections. Also, they report that relay removal was not part of the
procedures, but that technicians remove them because of prior occurrences of false
activations of alarms while inspecting and troubleshooting. The technician
responsible for the April 1 inspection and testing reported that he remembers
removing and reinstalling the relay. Also, site personnel reported construction activity
in and around Plant 9.
Prior to the event, facility personnel found that the procedure used to accomplish the
April 1 inspection partially duplicated the scope of another procedure, in that both
included quarterly inspections. Apparently, radiation technicians used both
procedures for quarterly inspections. Facility personnel had initiated action to
eliminate duplication in the procedures by revising one procedure and eliminating the
other. Personnel reported on planned training of radiation technicians on the
procedure revision and on the importance of following procedures. Personnel also
reported that training of radiation technicians on procedure revisions was standard
practice. Also, as a precaution, facility personnel inspected all other radiation-
detection-alarm relay boxes and found no other relays out of place.
DOE Order 4330.4A, "Maintenance Management Program," provides policy,
objectives, and criteria for establishment of maintenance programs. The stated
objective for surveillance and preventive maintenance addresses the need for
activities to ensure that equipment needed for safe and reliable facility operation is
inspected, tested, and maintained to keep that equipment within design operating
condition. One criteria calls for documentation to record activities performed, data
collected, and, when appropriate, the "as-found" and "as-left" condition of equipment
(page I-40). The Office of Nuclear Safety is developing a safety notice to address
uses of independent verification for equipment positioning.
2. FAILURE TO PROPERLY WEAR RESPIRATOR
On April 14, 1993, personnel at the Oak Ridge Alpha Handling facility identified a
worker wearing a respirator without the required filter canisters, increasing the risk
of airborne contamination exposure. Upon discovery, facility personnel discontinued
work activities and escorted the worker from the contaminated area. Nasal smears
and lung counts performed on the individual were negative. The worker had been
trained in the proper use of respirators and had previously used this type of
equipment.
The worker was preparing to bag and transfer waste in a glovebox for further disposal
when the incident occurred. The radiation work permit and a facility procedure for
glovebox operations specified respirator use as a precautionary measure. While there
is a potential for airborne activity, the area is routinely monitored and facility
personnel report that no airborne activity was discovered. The worker had obtained
the respirator and filter cartridges from the industrial hygienist department. Facility
procedures require workers to assemble and test respirators prior to use. Although
investigation of the root cause is in progress, the worker apparently forgot to properly
assemble and test the respirator before initiating work in the glovebox.
Serious consequences could have resulted if there had been airborne activity present.
As part of immediate corrective actions, the worker was given counseling, and facility
personnel suspended all activities requiring respirator use until respiratory protection
refresher training could be provided to operators. In addition, quality assurance
personnel plan to perform surveillances of work activities requiring respiratory
protection. (ORPS Report ORO--MMES-X10CHEMTEC-1993-0005)
Another event associated with the misuse of respiratory equipment occurred on July
18, 1991, at the Idaho National Engineering Laboratory Uranium Conversion/
Processing and Handling Facility. In this incident, a worker complained that his
respirator was not filtering paint fumes properly. The follow-up investigation
determined that the wrong filter canisters were used. Facility personnel determined
that a lack of training resulted in the worker requesting the wrong type of respiratory
filter canister. (ORPS Report ID--MKF-FPRCONST-1991-1011).
These events emphasize the importance of respirator qualification training and the
need for properly specifying, assembling, and inspecting respiratory equipment prior
to use. DOE Order 5480.4, "Environmental Protection, Safety, and Health Protection
Standards," mandates the requirements contained in American National Standards
Institute Standard Z88.2, "Practices for Respiratory Protection," and 29 CFR
1910.134 for implementation of the Respiratory Protection Program and associated
training of personnel. The DOE Radiological Control Manual provides additional
guidance for implementation of a respiratory protection program.
3. PERSONNEL OVEREXPOSURE FROM MISHANDLING RADIOGRAPHY SOURCE
On April 16, 1993, Scientific Inspection of Hixson, Tennessee, notified the Nuclear
Regulatory Commission (NRC) that one of their employees received significant
radiation exposure when he tried to manually adjust a stuck 99-curie iridium-92
radiography source. The calculated exposure to the individual's hand was 1,869 rem
and the calculated whole body dose was 1.217 rem. The employee was taken to a
hospital where his blood was tested and he was checked for contamination.
The employee was taking industrial radiography pictures at a cogeneration plant in
New Jersey when the source became stuck as he was cranking it out of the guide
tube. The area had been roped off before beginning the radiography. Suspecting it
was jammed, the employee approached the source with a radiation survey meter
which went off-scale high. The meter is normally set on the 10X scale, which reads
from 1 to 100 mrem/hour. He realized the source was out, returned to his crank
station and tried to crank the source. However, the source would not move in either
direction. The employee subsequently made three trips to the source attempting to
free it, spending an estimated half-minute in the immediate area of the source with
his hands about a half-inch from the source. After the third trip, he was able to crank
the source back into its stored position. He then notified personnel at the job site
what had happened and was transported to the hospital. (NRC Event Report No. 25409)
There have been incidents involving problems with radiography equipment at DOE
facilities. On February 4, 1992, a door limit switch failed on an X-ray generator being
used for radiography at the Oak Ridge Y-12 Site causing the generator to be activated
before the shielding door was fully closed (ORPS Report ORO--MMES-Y12QUALSER-1992-
0002). On February 14, 1991, radiography technicians at the Idaho Chemical
Processing Plant experienced difficulty in retracting a radiography source after leaving
the source guide tube bent at an angle and radius that did not allow the source to be
freely retrieved into the shielded portion of the camera. Facility personnel identified
the contributing cause and root cause as a less-than-adequate procedure and less-
than-adequate training, respectively (ORPS Report ID--WINC-ICPP-1991-0030). There were
no abnormal radiation exposures from either of these events. Descriptions of similar
events in the commercial nuclear industry can be found in NRC Information Notice 91-
23, "Accidental Radiation Overexposures to Personnel Due To Industrial Radiography
Accessory Equipment Malfunctions." Information Notice 91-49, "Enforcement of
Safety Requirements for Radiographers," describes NRC mandated radiography safety
requirements for NRC-licensed radiographers.
Articles 365 and 665 of the DOE Radiation Control Manual (DOE/HT-0265T, June
1992) contain requirements applicable to conducting radiography at DOE facilities.
Article 365 specifies the requirements related to operation and control of radiation
generating devices and references the following documents: (1) DOE 5480.4,
"Environmental Protection, Safety, and Health Protection Standards;" (2) "General
Safety Standard for Installations Using Non-Medical X-Ray and Sealed Gamma-Ray
Sources, Energies up to 10MeV;" (3) ANSI N43.2, "Radiation Safety for X-Ray
Diffraction and Fluorescence Analysis Equipment;" and (4) 10CFR34, "Licenses for
Radiography and Radiation Safety Requirements for Radiographic Operations." Article
655 specifies that "radiographers shall have training in accordance with 10 CFR
34.31. Radiation Generating Device Operators should have training comparable to
that required by 10 CFR 34.31."
DOE facilities involved in radiography or other operations of radiation generating
devices should consider reviewing their facility compliance with the applicable
sections of the DOE Radiological Control Manual.
4. REACTOR SCRAM WHILE PERFORMING ON-THE-JOB TRAINING
On April 16, 1993, the Advanced Test Reactor (ATR) in Idaho automatically
shutdown when an electrician conducting on-the-job training inadvertently tripped a
480-volt circuit breaker that supplied commercial power to components in three
experimental test loops. Facility personnel initiated applicable emergency procedures,
verified the cause of the scram, and ensured the reactor was in a safe shutdown
condition. After verifying no problems with the circuit breaker existed, ATR operators
restarted the reactor on April 18 following xenon decay.
The electrician, instructing trainees on circuit breaker rack-in and rack-out procedures,
received permission to operate on a spare 4160-volt breaker in the plant; he did so
without incident. However, the instructor then decided to show the trainees, without
actually operating circuit breakers, the differences in the rack-in/out procedures for
different types of breakers. He did not notify operations personnel of his intent to do
this. The instructor opened front panels for different types of breakers before
reaching the subject circuit breaker. As he was pointing out a rackout latch
mechanism, the instructor touched it with enough force to actuate it and cause the
breaker to open but not rack out. (ORPS Report ID--EGG-ATR-1993-0015)
During the event investigation, facility personnel determined that the electrician
conducting the training knowingly violated Conduct of Operations guidelines and
appropriate disciplinary actions are being considered. Investigators also identified
weaknesses in the guidelines for conduct of on-the-job training by maintenance
personnel in the plant and have suspended on-the-job training evolutions by those
personnel until the guidelines are strengthened. Finally, a Test Reactor Area Safety
Bulletin will be issued to communicate lessons learned from this event to facility
personnel.
In OE Weekly Summary 93-8, the Office of Nuclear Safety reported a similar event
that occurred on February 18, 1993, at the Sequoyah commercial nuclear power
plant where an electrical instructor caused a main turbine trip and a reactor scram
while conducting on-the-job training. The instructor was using the main generator
exciter field circuit breaker as an example to explain the operation of a certain type
of circuit breaker. The plant was operating at 100 percent power at the time. The
instructor received permission from the main control room to open the breaker cabinet
door. After opening the door, he reached into the cabinet to point out a latching
mechanism. As he was doing this, he inadvertently bumped the latching mechanism,
tripped the breaker, and caused the turbine trip and reactor scram. (NRC Event Number
25101)
These events demonstrate the importance of guidelines to control training activities
on operating equipment. Permission should always be obtained for operations prior
to entering energized panels or instrument panels. Control room and other
supervisory personnel should carefully consider the potential consequences when
granting permission to use operating equipment for on-the-job training. Guidelines
related to control of training activities at DOE facilities may be found in DOE Order
5480.19, Conduct of Operations Requirements for DOE Facilities, Chapter V,
Control of On-Shift Training."
5. GREASE SOLIDIFICATION CAUSES MOLDED CASE CIRCUIT BREAKER FAILURE
On April 7, 1993, the Nuclear Regulatory Commission (NRC) issued Information
Notice (IN) 93-26, "Grease Solidification Causes Molded Case Circuit Breaker Failure
to Close," to inform commercial nuclear power plant licensees of a problem with a
400-amp frame, 600-Vac molded-case circuit breaker manufactured by General
Electric Corporation (part number TJK436Y400). On March 26, 1992, a circuit
breaker of this type failed to close while transferring electrical loads from an
uninterruptible power supply (UPS) to the maintenance power supply at the Nine Mile
Point Unit 2 commercial nuclear power plant. Failure of the breaker to close caused
a loss of power to the standby gas treatment system radiation monitoring cabinet, a
false high-radiation signal, a primary containment isolation, and the loss of a control
room fire panel annunciator. Operators manually closed the circuit breaker and
restored power to the affected equipment within approximately 12 minutes.
Plant personnel determined that the breaker did not close because the grease used
at the pivot points inside the breaker had dried out and solidified. When the grease
dried out, it caused increasing friction and gouging at the metal-to-metal contact
areas. This friction caused the breaker to become increasingly more difficult to close
until finally the breaker would not close at all. Plant personnel located all breakers of
the same make, model, and year as the failed one and scheduled their replacement.
General Electric stated this type of molded case circuit breaker manufactured prior to
1985 used a soap-based or clay-based grease that could solidify with age, and they
recommended field testing to identify such solidification. Since 1985, General
Electric has replaced the soap-based and clay-based greases with synthetic grease
that does not dry out and solidify with age. However, it is possible the older molded-
case circuit breakers using the clay-based and soap-based grease may still be used
at other facilities and could result in similar failures. Facilities utilizing General Electric
molded case circuit breakers manufactured prior to 1985 should consider measures
such as inspection, grease replacement, or breaker replacement. Preventative
maintenance programs should include periodic inspection of electrical and mechanical
component lubricant condition.
6. FRENCH EXECUTIVES ON TRIAL FOR NUCLEAR ACCIDENT
On April 19, 1993, the former head of a French chemical company and two
colleagues went on trial charged with causing actual bodily harm to three repairmen
who suffered radiation poisoning in August 1991. The repairmen said they were not
warned about a particle accelerator left on in a room where they worked. Untrained
for working in a nuclear environment and unable to read instructions on the
accelerator written in English, they were exposed in a few minutes to as much as 20
times the maximum radiation dosage considered safe for one year. All three
repairmen suffered burns, hair loss, and skin irritation. All three individuals on trial
face up to one year in prison if convicted.
The incident is a reminder that personnel responsible for and involved in the operation
of DOE facilities can, under certain circumstances, be held personally liable for actions
that involve willful and deliberate violations of established regulations or law,
particularly when such actions result in deaths, injuries, or other serious
consequences. More importantly, it is an example of the need to enforce worker
safety training requirements as mandated by Section 31.31(b) of the National Defense
Authorizations Act for 1992 and 1993.
7. COOPER-BESSEMER DIESEL GENERATOR FUEL PUMP 10CFR21 NOTIFICATION
On April 14, 1993, the Cooper-Bessemer Reciprocating Products Division of Cooper
Industries sent the Nuclear Regulatory Commission (NRC) a 10CFR21 notification
concerning defective fuel pumps in the "KSV" emergency standby diesel generator
systems. The condition is significant since the defective fuel pumps may fail after
only ten hours of operation. Cooper-Bessemer engineering personnel contend that the
KSV emergency diesels can produce sufficient torque without one fuel injection pump
(and therefore one cylinder out-of-service) to drive the attached electric generator and
satisfy all plant design-basis electrical conditions. The defective fuel pump (part
number 2-50F-049-001) was supplied to Cooper-Bessemer by the Haynes
Corporation, Naples, Florida. The notification listed the Byron, Braidwood, Zion, Nine
Mile Point, South Texas, Cooper, and Waterford commercial nuclear facilities as
affected by the 10CFR21.
The NRC event notification also stated that since November 1989, twelve pump
failures have been reported at various facilities. The Cooper-Bessemer Owners Group
that investigated these incidents attributed the cause of failure to the galling of the
plunger and barrel in the fuel pump by silica and alumina particles. Most of these
failures occurred within ten hours of the initial operation of the pump. The Owners
Group determined the root cause to be a temporary breakdown of quality assurance
requirements at the Haynes Corporation Naples facility during the manufacturing and
rebuilding of the fuel pumps. (NRC Event Notification Number 25391)
Cooper-Bessemer personnel reported that the breakdown in quality assurance
requirements occurred during 1989 shortly after the fuel pump manufacturer moved
operations from another location. They suspect that the alumina and silica
contamination came from shot blast used to clean the pump hoses. Despite cleaning
requirements, trace amounts of the abrasive material remained in the threaded areas
of the fuel pump housing. As part of corrective actions, the fuel pump manufacturer
has eliminated shot blast operations on the fuel pump housings and initiated an acid
cleaning process. They also have initiated inspections prior to pump assembly to
check for dirt or grit.
Department of Energy facilities using diesels manufactured by Cooper-Bessemer
should determine the applicability of the 10CFR21 notification and contact the vendor
for further information. Cooper-Bessemer recommends that defective pumps be
removed and returned to the manufacturer for rebuild.
8. UNABLE TO VERIFY CRITICALITY SAFETY-LIMIT COMPLIANCE
On April 16, 1993, at the compact reactor fuel area of the uranium fuel fabrication
facility in Lynchburg, VA, Babcock & Wilcox personnel notified the Nuclear Regulatory
Commission when they determined that compliance with a nuclear criticality safety
limit could not be shown because of inability to demonstrate compliance with a
posted administrative limit on concentration of uranium-235. This condition was
discovered during a practice run for a new process. A placard in bay 13A in the fuel
washing station of the compact reactor fuel area required operators to limit uranium-
235 concentrations to 1,500 grams per liter of highly enriched uranium. While a
majority of the uranium handled in bay 13A for the last two to three years was
depleted, some was highly enriched. The posting was active during this time. A
criticality safety evaluation for the posted area relied on two barriers: the posting, and
minimum spacing of fuel washing cans. The safety evaluation, prepared by the
nuclear criticality safety group, specified the posting limit and it was reviewed and
approved by affected operations and engineering groups. Facility personnel reported
that can spacing was maintained by passive, geometric-type, engineering controls.
Personnel also reported that fuel-washing operations make certain physical changes
to the fuel. This meant it was impossible for the bay 13A operators to verify
compliance with the posting. The placard stated, in part: "If this limit cannot be met,
notify Nuclear Criticality Safety." No such notifications were recorded. Also,
according to facility personnel, applicable operating procedures stated that if any step
in the procedure could not be met, then the operator must stop work and notify the
supervisor. The facility has been shut down until an evaluation is performed to
determine correct, verifiable wording on the sign. (NRC Event Number 25405)
For Department of Energy facility contractors, DOE Order 5480.24, "Nuclear
Criticality Safety," provides direction on establishing nuclear criticality safety
program requirements. This order establishes a double contingency principle to
prevent a criticality accident and provides necessary elements for criticality safety
programs. Also, the order invokes several American Nuclear Society Standards for
basic elements and control parameters of programs for nuclear criticality safety. DOE
Order 5480.19, "Conduct for Operations Requirements for DOE Facilities," provides
necessary elements for operator aid postings. As stated on pages I-95 and I-96, a
program should be established to ensure that postings "...are current, correct, and
useful," and that postings "...should be reviewed periodically to ensure they are still
correct and necessary." This event also demonstrates the need for a questioning
attitude on the workers' part. Workers should routinely ask themselves if an
instruction, such as a posting, is understandable and achievable, and, when in doubt,
ask supervision for clarification or direction before proceeding. Also, a training needs
analysis of procedures and operator aids that identify necessary worker skills,
knowledge, and abilities can be helpful in finding unclear and erroneous instructions.
9. LEVEL INSTRUMENTATION INACCURACIES OBSERVED DURING NORMAL PLANT
DEPRESSURIZATIONS
On April 8, 1993, the Nuclear Regulatory Commission issued Information Notice 93-
27, "Level Instrumentation Inaccuracies Observed During Normal Plant
Depressurization." The Information Notice describes an event at a commercial nuclear
power plant and provides background information on inaccuracies in reactor-vessel
level indication and how these errors could result in complicated long-term operator
actions and possible loss of automatic actuation of safety systems while the reactor
is in a reduced-pressure condition.
On January 21, 1993, during a plant cooldown following a reactor scram at
Washington Nuclear Power Plant Unit 2, "notching" of the level indication was
observed on at least two of four channels of the reactor-vessel narrow-range level
instrumentation. ("Notching" is a momentary increase in indicated water level.) This
increase occurred when a gas bubble moved through a vertical portion of the
reference leg and caused a temporary decrease in the static head in the reference leg.
The notching was first observed on one channel at a pressure of approximately 120
psig and on another starting at approximately 50 psig. At these pressures, the level
error was in the approximate range of 4 to 7 inches. Beginning at a pressure of about
35 psig, level indication from one channel became erratic and, as the plant continued
to depressurize, a 32-inch error occurred. The 32-inch level error was sustained and
gradually recovered over a period of two hours.
The licensee postulated this large error in level indication was caused by gas released
into the reference leg displacing about 40 percent of the water volume. The licensee
also postulated that the slow recovery of correct level indication resulted from the
time needed for steam to condense in the condensate chamber and refill the reference
leg. The licensee inspected the reference leg for the channel with the 32-inch level
error and discovered leakage through reference leg fittings. This could have been a
contributing factor to accumulation of dissolved noncondensible gas in that reference
leg.
The licensee determined that the false high level indication observed could result in
a failure to automatically isolate a leak in the residual heat removal system during
shutdown cooling. The design basis for the unit included a postulated leak in the
residual heat removal system piping outside containment during shutdown cooling.
The licensee implemented compensatory measures for future plant cooldowns to
ensure that a leak could be promptly isolated.
The Nuclear Regulatory Commission previously issued Information Notice 92-54,
"Level Instrumentation Inaccuracies Caused by Rapid Depressurizations," and Generic
Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level
Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)," to discuss how
noncondensible gas may become dissolved in the reference leg of water level
instrumentation and lead to false indications of high level after a rapid
depressurization event.
These events underscore the importance of considering all factors that could affect
proper instrument operation. Unaccounted-for fluid transients could prevent proper
instrument operation and thereby prevent or degrade safety-related equipment
operation.
10. OPERATOR INCORRECTLY STARTS WRONG EMERGENCY DIESEL GENERATOR
On April 20, 1993, an operator at the Quad Cities commercial nuclear power facility
incorrectly started the wrong diesel generator, causing an inadvertent operation of an
emergency safety system. The operator started the unit 1 diesel instead of the unit
1/2 swing diesel as part of an output breaker test. The unit 1 diesel started, but did
not load. Upon discovery, facility personnel secured the swing diesel. Reactor
operations were not affected by this event, and there was no damage to the swing
diesel or other plant equipment.
The control switches for the unit 1 emergency diesel and the swing diesel are located
on the same panel approximately 20-inches apart, with the unit 1 diesel operating
switch painted black and the swing diesel painted orange. The operator was directed
to start the swing diesel and load it to unit 2 to test the operation of the output
breaker. The operator had reviewed and walked through the procedure, looking at
the panel indications. Another operator was sent to confirm the readiness of the
swing diesel prior to beginning the test. After receiving confirmation that the swing
diesel was ready, the operator made an announcement over the plant page system
that the swing diesel was being started. He then walked over to the panel and turned
the switch for the unit 1 diesel instead of the swing diesel. (NRC Event Notification Number
25417)
Personnel errors are a significant contributor to many events at Department of Energy
facilities and at commercial nuclear power plants. The Office of Nuclear Safety
reported other occurrences involving personnel error and self checking in Operating
Experience Weekly Summaries 92-17, 92-19, 92-24, 92-33, and 93-08. As
previously reported, the commercial industry, with guidance from the Institute for
Nuclear Power Operations, is addressing the problem with a program called
Self-Checking. This technique is described in Good Practice OE-908, Self
Checking. Self Checking is a risk management tool designed to reduce the potential
for human error by helping facility personnel focus their attention on the details of the
task at hand. This event emphasizes the importance of self-checking to reduce
human error. This event may not have occurred if the operator had focused on the
proper switch to activate. Self Checking techniques are used widely at commercial
nuclear power plants to prevent human performance problems and can identify a
potential human performance error before it occurs.
The Institute of Nuclear Power Operations also recently issued Significant Operating
Experience Report 92-1, Reducing the Occurrence of Plant Events Through Improved
Human Performance. This document describes lessons learned by member
commercial utilities in addressing the principal causes of human performance
problems, and also describes essential elements identified from a review of several
successful commercial utility programs to address human performance problems.
ADDITIONAL INFORMATION RELATED TO FOLLOWUP ACTIVITIES
1. WORKER CONTAMINATIONS DURING CONSTRUCTION ACTIVITIES
In OE Weekly Summary 93-1, the Office of Nuclear Safety reported a contamination
event that occurred on December 29, 1992, at the Bethel Valley Liquid Low-Level
Waste (BVLLW) Upgrade Project at Oak Ridge National Laboratory. Nineteen workers
were contaminated and low-level contamination from the event was spread to
surrounding grade-level areas. At the time, the cause of the contaminations was
attributed to a leaking pipe. (ORPS Report ORO--MKFO-X10CENTENG-1992-0026)
Site personnel have since completed a Type B investigation of the event and issued
report DOE/OR-1082, "Type B Investigation of the Contamination Accident at the
BVLLW Upgrade Project."
According to the report, the workers were contaminated while performing
construction activities at a new but not yet operating radioactive waste-collection
facility. Waste drains from an operating radioanalytical laboratory had been
connected to the new waste receiving facility in early December. Radioactive waste
was inappropriately introduced into the new but not yet authorized laboratory drain
system, and water was later released into the construction site during installation of
valve instrumentation in the drain line. Additionally, a mixture of rainwater and
contaminated liquid was pumped to a nearby grade-level area which was
subsequently surveyed and posted as a contamination area. Surveys confirmed that
no contamination reached the storm drain.
The Monitoring and Control Station (MCS), which is a part of the BVLLW Upgrade
Project, is intended to receive low-level radioactive waste from the Radioactive
Materials Analytical Laboratory (RMAL) in Building 2026. Workers came in contact
with radioactive liquid in the MCS tank and valve vault areas while performing various
construction activities. The liquid exited through taped-over piping when a valve was
cycled several times. The workers noticed the liquid but assumed it was rainwater
or condensate. One worker removed the tape thinking the liquid would empty out;
however, a continuous trickle was observed, and the tape was placed back in
position.
The workers were not aware of their contamination until the end of the work shift
when employee comments regarding the liquid coming from the cut pipe were
overheard by a fellow employee. The employee was aware that Building 2026
handled radioactive materials and that the MCS was tied into Building 2026. After
confirming that the workers had contacted the liquid, he contacted supervision to
initiate worker contamination surveys. Several workers were found to be
contaminated and efforts were initiated to recall potentially contaminated workers
who had already left the site.
The Investigation Board used the Management Oversight and Risk Tree (MORT)
method and Barrier Analysis to investigate the event. According to the report, the
direct cause was introduction of liquid low-level radioactive waste into the new drain
system in Building 2026, Laboratory 100, which leads to the MCS where construction
personnel were working. The new drain system was not ready for authorized use.
The Board identified 13 contributing causes in the areas of Risk Assessment, Job
Safety Analysis, Conduct of Operations, and Management and Oversight. Finally, the
Board concluded there were three root causes: (1) a failure to identify the hazards and
properly analyze job safety, (2) a failure to apply Conduct of Operations principles
related to personnel protection, and (3) a failure to provide proper safety oversight.
The report provided 16 specific recommended measures to prevent recurrence of a
similar event. In general, the Board recommended that management review their
health and safety programs to ensure a safe working environment for employees
involved in concurrent operations and construction activities, and that the review be
an integral part of every phase of a construction project, from initial planning through
operation. The report further recommended that risk assessment of tie-ins to
operating areas be performed to understand and respond to potential vulnerabilities.
Other recommendations, including one for a special emphasis on Conduct of
Operations during concurrent construction and operations activities, were made in the
report.
SAFETY NOTICES UNDER DEVELOPMENT
Note: The Office of Nuclear Safety encourages input related to the development of Safety
Notices. If you have any questions, comments, or information concerning events or
issues similar to the following, please contact Ivon Fergus, Office of Nuclear Safety
at (301) 903-6364.
1. NS has identified a number of events related to the loss of annunciators and
other safety-related equipment because of problems involving 120-VAC/125-
VDC systems at DOE and commercial facilities. NS is reviewing potential
generic problems associated with the adequacy of 120-VAC/125-VDC systems
at DOE facilities.
2. NS evaluated three events associated with the temporary diesel generator at the
Rocky Flats Plant, Building 707. The lessons learned from these events,
particularly as they relate to the control of temporary modifications, are being
considered for dissemination in an NS Safety Notice.
3. NS is developing a Safety Notice concerning problems with Uninterruptible
Power Supplies (UPS).
4. NS is considering development of a Safety Notice related to control of work at
electrical substations and switchyards.
5. NS is working with Lawrence Livermore National Laboratory and DOE-SF
personnel to develop a Safety Notice on cracking in ventilation ducting.
6. NS is considering developing a Safety Notice related to fuel oil supplies for
Emergency Diesel Generators (EDGs).
7. NS is developing a Safety Notice to address uses of independent verification for
equipment positioning.
8. NS is developing a Safety Notice in regards to maintaining important alarm and
monitoring systems at facilities undergoing Transition and Decontamination &
Decommissioning.
SAFETY NOTICES PREVIOUSLY ISSUED
Safety Notice No. 91-1, "Criticality Safety Moderator Hazards," September 1991
Safety Notice No. 92-1, "Criticality Safety Hazards Associated With Large Vessels,"
February 1992
Safety Notice No. 92-2, "Radiation Streaming at Hot Cells," August 1992
Safety Notice No. 92-3, "Explosion Hazards of Uranium-Zirconium Alloys," August
1992
Safety Notice No. 92-4, "Facility Logs and Records," September 1992
Safety Notice No. 92-5, "Discharge of Fire Water Into a Critical Mass Lab," October
1992
Safety Notice No. 92-6, "Estimated Critical Positions (ECPs)," November
1992
Safety Notice No. 93-1, "Fire, Explosion, and High-Pressure Hazards Associated with
Drums and Containers," February 1993
Copies of NS Safety Notices may be requested from: Nuclear Safety Information Center,
Office of Nuclear Safety, U.S. Department of Energy, Room S161, GTN, Washington, DC
20585