HomeMy WebLinkAbout851184.tiff SSINS No. : 6835
IN 85-91
UNITED STATES
NUCLEAR REGULATORY COMMISSION Q C ` `tl- `
OFFICE OF INSPECTION AND ENFORCEME s DEC 6 1985
WASHINGTON, D.C. 20555
JLJ
November 27, 1985 k�er� Coco.
IE INFORMATION NOTICE NO. 85-91: LOAD SEQUENCERS FOR EMERGENCY DIESEL GENERATORS
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice is provided to advise licensees and applicants of
potential design deficiencies that could bypass load sequencers, thereby.
causing loss of redundant emergency diesel generators (EDGs). Recipients are
expected to review the information for applicability to their facilities and
consider actions, if appropriate, to preclude similar problems occurring at
their facilities. However, suggestions contained in this information notice
do not constitute NRC requirements; therefore, no specific action or written
response is required.
Description of Circumstances:
On August 22, 1985, the licensee for the Duane Arnold nuclear plant discovered
that an accident signal and the loss of the standby transformer (a source of
offsite electric power) would cause engineered safety feature (ESF) loads to
be applied as a single block load onto the EDGs (the sources of onsite electric
power), which would likely cause loss of both EDGs.
Pending replacement of the unit auxiliary transformer (lost in a transformer
fire in October 1984), the licensee was operating the plant with the non-
safety-related loads on the station startup transformer and the safety-related
loads on the station standby transformer. The plant design objective was to
sequence the ESF loads onto the EDGs if offsite power to the ESF buses should
be lost and an accident signal was present. The licensee' s training staff
realized that the logic and sensors used to determine the availability of off-
site power were such that the offsite power feeder breakers to the ESF buses
could be tripped, but offsite power would be indicated as being still available.
Under these conditions the design would cause the ESF diesel generator load
sequencers to be bypassed.
To justify continued safe operation, the licensee has temporarily placed
certain sequencer test switches in the test position, which forces the sequen-
cers to function even though offsite power is sensed as being available.
851184
8511250335
M
IN 85-91
November 27, 1985
Page 2 of 2
For the longer term, the licensee is developing a permanent design change
which is to be reviewed by the NRC.
Discussion:
The design of the electric power system at the Duane Arnold nuclear plant
includes features to sequence ESF loads onto the EDGs, but not to sequence
loads onto offsite power. In a sense, these design objectives are in conflict;
that is, one is for sequencing and the other is for not sequencing. When
design objectives are potentially conflicting, careful analysis is necessary
to ensure that failures of various types do not result in implementation of
the improper objective. In this case, the logic was designed so that if
source of offsite power is "available" (such as at either the standby trans-
former or the startup transformer) the ESF load sequencers would be bypassed.
Thus, if the standby transformer were lost, causing a loss of power to the
safety-related loads, the logic would still indicate offsite power as available.
This design was provided by Bechtel Corporation.
The result was the potential for an interaction between the offsite electric
power system and the onsite electric power system that could have caused the
loss of redundant sources of onsite power. Such an interaction is incompatible
with the requirements of 10 CFR 50, Appendix A, General Design Criterion No. 17,
"Electric Power Systems. " The Duane Arnold original design was such that the
availability of offsite electric power was determined indirectly; that is, by
an upstream measurement rather than directly at the ESF buses. This deficiency
existed in the original plant design and was not discovered when the design was
reviewed again by the licensee after the loss of the unit auxiliary transformer
in October 1984.
No specific action or written response is required by this information notice.
If you have questions about this matter, please contact the Regional Adminis-
trator of the appropriate NRC regional office or this office.
Edward L ordan, Director
Divisio f Emergency Preparedness
and E gineering Response
Office of Inspection and Enforcement
Technical Contacts: J. T. Beard, NRR
(301) 492-7465
Eric Weiss, IE
(301) 492-9005
Attachment: List of Recently Issued IE Information Notices
Attachment 1
IN 85-91
November 27, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-58 Failue Of A General Electric 11/19/85 All power reactor
Sup. 1 Type AK-2-25 Reactor Trip facilities designed
Breaker by B&W and CE holding
an OL or CP
85-90 Use Of Sealing Compounds In 11/19/85 All power reactor
An Operating System facilities holding
an OL or CP
85-89 Potential Loss Of Solid-State 11/19/85 All power reactor
Instrumentation Following facilities holding
Failure Of Control Room an OL or CP
Cooling
85-88 Licensee Control Of 11/18/85 All power reactor
Contracted Services Providing facilities holding
Training an OL or CP
85-87 Hazards Of Inerting 11/18/85 All power reactor
Atmospheres facilities holding
an OL or CP; and
fuel facilities
85-86 Lightning Strikes At Nuclear 11/5/85 All power reactor
Power Generating Stations facilities holding
an OL or CP
85-85 Systems Interaction Event 10/31/85 All power reactor
Resulting In Reactor System facilities holding
Safety Relief Valve Opening an OL or CP
Following A Fire-Protection
Deluge System Malfunction
85-84 Inadequate Inservice Testing 10/30/85 All power reactor
Of Main Steam Isolation Valves facilities holding
an OL or CP
85-83 Potential Failures Of General 10/30/85 All power reactor
Electric PK-2 Test Blocks facilities holding
an OL or CP
OL = Operating License
CP = Construction Permit
•
SSINS No. : 6835
IN 85-58, Supplement 1
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
November 19, 1985 SEC 41985
IE INFORMATION NOTICE NO. 85-58, SUPPLEMENT 1: FAILURE OF W%N€(2#L ELECTRIC
TYPE AK-2-25 REACTOR TRIP
BREAKER
Addressees:
All nuclear power reactor facilities designed by Babcock and Wilcox (B&W) and
Combustion Engineering (CE) and holding an operating license (OL) or a construc-
tion permit (CP).
Purpose:
IE Information Notice 85-58, "Failure of a General Electric Type AK-2-25
Reactor Trip Breaker," described the failure of a refurbished reactor trip
breaker (RTB) at the Rancho Seco Nuclear Power Generating Station. This
supplement to the notice provides information on other recent failures of
General Electric (GE) AK-2-25 type RTBs. It is expected that recipients will
review the information for applicability to their facilities and consider
actions, if appropriate, to preclude similar problems at their facilities.
However, suggestions contained in this supplement do not constitute NRC
requirements; therefore, no specific action or written response is required.
Description of Circumstances and Discussion:
Calvert Cliffs Nuclear Power Plant
Recently, there have been two failures of undervoltage (UV) trip devices in the
GE AK-2-25 type RTBs at Calvert Cliffs. The first failure was discovered in
February 1985 when the response time of an UV trip device measured 628 milli-
seconds, well above the licensee' s acceptance criteria. This UV trip device
was installed in October 1984. The analysis of the failure revealed that
several laminated sections that are part of the armature had slipped down and
effectively eliminated the air gap between the movable armature and the pole
face. By design, there must be an air gap between the laminations and the pole
face. The physical contact between the laminations and pole face allowed the
armature to be held down by residual magnetism after dc power was removed,
resulting in the slow response time.
The second failure was discovered during preventive maintenance of an RTB in
July 1985. In this case, laminations had moved down only slightly to make the
air gap below tolerance. This did not affect the response time. However, it
did affect the pickup and dropout voltages of the UV trip device causing them
to be low.
8511150089
IN 85-58, Supplement 1
Page 2 of 3
November 19, 1985
The licensee' s corrective actions included replacing the UV trip devices and
instituting a program to measure the air gap between the laminations and pole
face on a yearly basis. It was pointed out by the licensee that the air gap
measurement is not part of the checks recommended in the GE Service Advice
Letter. Therefore, it is likely that this is not being performed in the
industry.
Oconee Nuclear Station
On July 22, 1985, at the Oconee Nuclear Station Unit 1, one of the GE AK-2-25
dc RTBs failed its trip response time during on-line testing of a reactor
protection system channel while the unit was operating at 100% power. The trip
response time of the breaker was 1738 milliseconds, well over the licensee' s
acceptance criteria.
On July 23, 1985, the licensee exercised the failed breaker numerous times, but
no failure occurred. However, a detailed inspection of the breaker showed a
metal burr on the head of one of the mounting studs for the UV trip device.
The licensee concluded that the probable cause of the failure was the armature
of the UV device touching the stud as it moved toward the trip position.
The failed UV trip device was a new device installed on the breaker. It was
discovered that the mounting stud heads of the new devices had square edges
rather than round ones like the old devices. According to the licensee, the
possible reduced clearance between the armature and the heads of the mounting
bracket studs could have caused the contact and, thus, the slow RTB response
time. The licensee' s corrective actions included the replacement of the failed
RTB with an operable spare RTB, and the on-line testing and operability verifi-
cation of all RTBs.
The licensee' s review of previous RTB failures has indicated that a new UV
device installed in the same breaker had failed once before on April 29, 1985
because of mechanical binding. The cause of the previous failure, which also
resulted in a slow trip response time, was thought to be some particles,
possibly paint chips or metal shavings, stuck in the pivot point of the UV
device. The licensee' s corrective actions included the replacement of the
failed RTB with an operable spare RTB, and the revision of the RTB inspection
and maintenance procedure to include detailed inspections of the UV and shunt
devices and other binding points.
In addition to the RTB failures discussed above, GE notified the NRC and
affected facilities on September 13, 1985 of certain defects in the UV trip
devices supplied for use on AK and AKR type low-voltage power circuit breakers.
Subsequently, GE issued Service Advice Letter No. 300 on September 26, 1985
that outlines the actions to be taken with respect to those defects. One of
the defects addressed by GE involved insufficient clearance between the armature
and mounting stud, similar to the Oconee problem. The other defect involved
improper painting of the mating surfaces of the armature and pole pieces in ac
powered UV devices.
IN 85-58, Supplement 1
Page 3 of 3
November 19, 1985
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional
Administrator of the appropriate regional office or this office.
1/Efrit
dan, Director
Divisiomergency Preparedness
andring Response
Office of Inspection and Enforcement
Technical Contact: R. N. Singh, IE
(301) 492-4149
Attachments: List of Recently Issued Information Notices
Ok
Attachment 1
IN 85-58, Supplement 1
November 19, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-90 Use Of Sealing Compounds In 11/19/85 All power reactor
An Operating System facilities holding
an OL or CP
85-89 Potential Loss Of Solid-State 11/19/85 All power reactor
Instrumentation Following facilities holding
Failure Of Control Room an OL or CP
Cooling
85-88 Licensee Control Of 11/18/85 All power reactor
Contracted Services Providing facilities holding
Training an OL or CP
85-87 Hazards Of Inerting 11/18/85 All power reactor
Atmospheres facilities holding
an OL or CP; and
fuel facilities
85-86 Lightning Strikes At Nuclear 11/5/85 All power reactor
Power Generating Stations facilities holding
an OL or CP
85-85 Systems Interaction Event 10/31/85 All power reactor
Resulting In Reactor System facilities holding
Safety Relief Valve Opening an OL or CP
Following A Fire-Protection
Deluge System Malfunction
85-84 Inadequate Inservice Testing 10/30/85 All power reactor
Of Main Steam Isolation Valves facilities holding
an OL or CP
85-83 Potential Failures Of General 10/30/85 All power reactor
Electric PK-2 Test Blocks facilities holding
an OL or CP
85-82 Diesel Generator Differen- 10/18/85 All power reactor
tial Protection Relay Not facilities holding
Seismically Qualified an OL or CP
OL = Operating License
CP = Construction Permit
SSINS No. : 6835
IN 85-90
UNITED STATES r7 `" ` `
NUCLEAR REGULATORY COMMISSION Q
ir
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555 DEC 4 1985 ' H
November 19, 1985 1JJ
CRELL`-K, CoLO. _ 7
IE INFORMATION NOTICE NO. 85-90: USE OF SEALING COMPOUNDS' IN AN OPERATING
SYSTEM
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice is provided to alert recipients of a potentially
significant problem pertaining to the injection of a sealing compound into an
operating system. It is expected that recipients will review the information
for applicability to their facilities and consider actions, if appropriate, to
preclude a similar problem occurring at their facilities. However, suggestions
contained in this information notice do not constitute NRC requirements;
therefore, no specific action or written response is required.
Description of Circumstances:
Catawba Nuclear Station Unit 1 has a closed-loop component cooling water system
(CCWS) with equipment receiving cooling flow arranged in two parallel circuits
(trains). Each train provides cooling water to one train of redundant engi-
neered safety equipment (essential header). In addition, cooling water for the
nonessential header can be provided by branches from either train of the CCWS.
The nonessential header is isolated from both essential headers by a 20-inch,
butterfly-type, motor-operated valve in the branch for each train (See
Figure 1).
In May 1985, operations personnel identified that there was excessive leakage
past the seats of both of the 20-inch butterfly valves. An attempt was made to
shut down train B and the nonessential header to allow for work on the butter-
fly valve in train B. However, because of the excessive valve seat leakage
from train A, train B could not be depressurized. A decision was made to
inject a sealing compound into the branch piping of the operating train
(train A) immediately upstream of the butterfly valve. Because the valve was
located in a section of branch piping just downstream from a dead leg (approxi-
mately a 5-foot drop in piping), the conjecture was that the sealing compound
would migrate to the leaking valve seat area, thus stopping the leak without
being carried to any component being cooled by the main run of train A.
8511150120 IL
11 C� :'�Tc.) 12-9-
'�
IN 85-90
November 19, 1985
Page 2 of 3
To inject the sealing compound, holes were drilled in the valve body at an
angle so that the injection would be on the upstream side of the valve and
below the dead leg. Sealing compound of various consistencies was then inject-
ed into this area. A total of 146 boxes (2 pounds per box) was injected into i
this area without stopping the leakage.
Following this unsuccessful effort, a mechanical plugging technique was used to
i
successfully plug the A train branch line downstream of the valve. The leaking
butterfly valves were then removed and replaced sequentially. Also replaced
during this time was a 1-inch drain line and valve located between the two
20-inch butterfly valves. The 1-inch line and valve had become plugged with
sealing compound that had passed the seat of the butterfly valve. In the
course of replacing the valves, it was found that most of the sealing compound
settled as a large plug just upstream of the 20-inch A train valve, but this
plug had not sealed the leak in the valve.
Subsequent to replacement of the two 20-inch butterfly valves and the smaller
piping, the CCWS was returned to service. With the system operating, higher
than normal differential pressures were identified on some train A components.
Further, the required cooling water flow could not be achieved. As a result of
these findings, various heat exchangers were inspected and cleaned. It was
found that a small amount of sealing compound had gone back up the dead leg of
the branch and into the main run piping. Sealing compound was found in varying
amounts in these heat exchangers. The train A coolant charging pump motor
cooler had approximately 2 pounds of sealing compound (maximum amount identi-
fied) and some coolers had none.
Other problems associated with the use of sealing compounds have occurred in
the past. IE Information Notice No. 82-06, "Failure of Steam Generator Primary
Side Manway Closure Studs," was issued March 12, 1982 and addressed another
situation where the use of sealants to stop a leak may have contributed to
other unforeseen problems. In that situation, the sealing compound was inject-
ed using a procedure which essentially created an enclosure around the bolt
circle between the flanges. This resulted in an autoclave type of atmosphere
where corrodents from the coolant could concentrate and contribute to degrada-
tion of the flange bolting.
The results of tests sponsored by the Electric Power Research Institute (EPRI)
on sealants and on line sealing procedures also show additional problems that
may develop with the use of sealants. These results are published in EPRI
NP-3111, "Testing and Evaluation of On-Line Leak Sealing Methods. "
Licensees are reminded that ANSI/ANS-3. 2-1982, "Administrative Controls and
Quality Assurance for the Operational Phase of Nuclear Power Plants," part
5. 2.71, Maintenance Programs, requires that "planning for maintenance shall
include evaluation of the use of special processes, equipment and materials in
performance of the task, including assessment of potential hazards to personnel
and equipment. "
IN 85-90
November 19, 1985
Page 3 of 3
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional
Administrator of the appropriate regional office or this office.
and Ly ordan, Director
Division,; f Emergency Preparedness
and En ineering Response
Office of Inspection and Enforcement
Technical Contact: William F. Anderson, IE
(301) 492-4819
Attachments:
1. Figure 1, "Component Cooling Water Header Interconnections"
2. List of Recently Issued IE Information Notices
Attachment 2
IN 85-90
November 19, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-89 Potential Loss Of Solid-State 11/19/85 All power reactor
Instrumentation Following facilities holding
Failure Of Control Room an OL or CP
Cooling
85-88 Licensee Control Of 11/18/85 All power reactor
Contracted Services Providing facilities holding
Training an OL or CP
85-87 Hazards Of Inerting 11/18/85 All power reactor
Atmospheres facilities holding
an OL or CP; and
fuel facilities
85-86 Lightning Strikes At Nuclear 11/5/85 All power reactor
Power Generating Stations facilities holding
an OL or CP
85-85 Systems Interaction Event 10/31/85 All power reactor
Resulting In Reactor System facilities holding
Safety Relief Valve Opening an OL or CP
Following A Fire-Protection
Deluge System Malfunction
85-84 Inadequate Inservice Testing 10/30/85 All power reactor
Of Main Steam Isolation Valves facilities holding
an OL or CP
85-83 Potential Failures Of General 10/30/85 All power reactor
Electric PK-2 Test Blocks facilities holding
an OL or CP
85-82 Diesel Generator Differen- 10/18/85 All power reactor
tial Protection Relay Not facilities holding
Seismically Qualified an OL or CP
OL = Operating License
CP = Construction Permit
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SSI.NS No. : 6835
D n
UNITED STATES 1 �1 ^'y
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT a 4 ig$5 !I
WASHINGTON, D.C. 20555
November 19, 1985
IE INFORMATION NOTICE NO. 85-89: POTENTIAL LOSS OF SOLID-STATE INSTRUMENTATION
FOLLOWING FAILURE OF CONTROL ROOM COOLING
Addressees:
All nuclear power reactor facilities holding an operating license (OL) or a
construction permit (CP).
Purpose:
This information notice is to alert recipients of a potentially significant
problem involving the loss of solid-state instrumentation following the failure
of control room cooling. Actions taken by the licensee to mitigate the event
also are discussed. It is expected that recipients will review this informa-
tion for applicability to their facilities and consider actions, if appropri-
ate, to preclude a similar problem occurring at their facilities. However,
suggestions contained in this information notice do not constitute NRC require-
ments; therefore, no specific action or written response is required.
Description of Circumstances:
On June 4, 1984, both units of the McGuire Nuclear Station were operating at
100% power with one of the two main control room ventilation units out of
service for maintenance. At 8:02 p.m. , the remaining chiller tripped on low
oil level resulting in a total loss of main control room cooling. At approxi-
mately 8:45 p.m. , as the control room temperature increased, numerous alarms on
Unit 1 high reactor coolant loop C Tave were received, as well as alarms on
Unit 1 pressurizer level . Attempts to restore the air conditioning were
unsuccessful and at 9:00 p.m. the air conditioning was declared inoperable. At
10:00 p.m. , the operators opened the doors between the control room and the
computer room, which still had cooling available. Operators also opened the
doors of the Westinghouse PCS 7300 cabinets, which contain the solid-state
circuit cards generating the alarms. The licensee then used portable fans with
ducting to provide cooling from the computer room to the PCS 7300 cabinets.
The required technical specification power reduction was started at 10:05 p.m.
and terminated at 10:55 p.m. when one of the air conditioning units was re-
turned to service. The solid-state instrumentation returned to normal follow-
ing restoration of the air conditioning.
8511150114 12/1h5
•
oaN
IN 85-89
November 19, 1985
Page 2 of 3
Discussion:
Before June 4, 1984, the McGuire Nuclear Station had experienced numerous
printed solid-state circuit card failures with the Westinghouse PCS 7300
cabinets and associated solid-state protection system (SSPS). The card fail-
ures, which involved reactor trips and spurious instrument indications, were
attributed, by the licensee, to overheating in the PCS 7300 cabinets. In some
cases, the spurious instrumentation indications disappeared when adequate
ventilation was provided to the cabinets; however, in other cases, continued
erratic instrumentation indicated that the overheating had significantly
shortened the life expectancy of the solid state components. The licensee also
had previously reported that the air chillers develop oil level problems when
loaded at less than full capacity. The heat load calculated during plant
design was too large compared to the actual heat load resulting in oversized
chillers.
Following the event, the licensee took temperature measurements inside the PCS
7300 cabinets and determined that with an ambient temperature of about 72°F the
cabinets had internal temperatures of up to 125°F on the top rack. The McGuire
operators estimate that the ambient temperature, during the event, reached 90°F
before alternate cooling was provided. The licensee has rebalanced the airflow
in the control area ventilation system to provide additional cooling to the PCS
cabinets. Though the licensee' s remedial actions to provide better normal
cooling appear to have increased the reliability of the_solid-state cabinets
under design operating conditions, the safety concern following loss of all
control room HVAC units remains.
The McGuire operators, alerted by prior experience, took prompt action to
provide alternate cooling to the solid-state equipment during the event.
Without such action, the possible loss of some instrumentation and erratic
instrument readings may have made it difficult to bring the plant to a safe
condition, such as hot shutdown. If no control room cooling is available to
the solid-state cabinets, it may not be prudent to delay in going to a hot
shutdown condition even though the plant technical specifications may allow
appreciable time to achieve the shutdown. The failure rate of the instrumenta-
tion can be expected to increase as the control room temperature increases and
the erratic instrumentation may cause a reactor trip at the same time that the
instrumentation is unreliable or unavailable to assist the operators.
Loss of all control room cooling may be more likely than previously thought.
In addition to the McGuire event, there has been recent identification of other
reported possible common-mode HVAC failures at Browns Ferry and Limerick.
Therefore, licensees should be alert for the possibility of the loss of control
room cooling and the impact this may have on their solid-state instrumentation.
IN 85-89
November 19, 1985
Page 3 of 3
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional
Administrator of the appropriate regional office or this office.
414:11r6/14
dan, Director
Division f Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: James Stewart, IE
(301)492-9061
Attachment: List of Recently Issued IE Information Notices
Attachment 1
IN 85-89
November 19, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-88 Licensee Control Of 11/18/85 All power reactor
Contracted Services Providing facilities holding
Training an OL or CP
85-87 Hazards 0f Inerting 11/18/85 All power reactor
Atmospheres facilities holding
an OL or CP; and
fuel facilities
85-86 Lightning Strikes At Nuclear 11/5/85 All power reactor
Power Generating Stations facilities holding
an OL or CP
85-85 Systems Interaction Event 10/31/85 All power reactor
Resulting In Reactor System facilities holding
Safety Relief Valve Opening an OL or CP
Following A Fire-Protection
Deluge System Malfunction
85-84 Inadequate Inservice Testing 10/30/85 All power reactor
Of Main Steam Isolation Valves facilities holding
an OL or CP
85-83 Potential Failures Of General 10/30/85 All power reactor
Electric PK-2 Test Blocks facilities holding
an OL or CP
85-82 Diesel Generator Differen- 10/18/85 All power reactor
tial Protection Relay Not facilities holding
Seismically Qualified an OL or CP
85-81 Problems Resulting In 10/17/85 All power reactor
Erroneously High Reading facilities holding
With Panasonic 800 Series an OL or CP and
Thermoluminescent Dosimeters certain material
and fuel cycle
licensees
OL = Operating License
CP = Construction Permit
I
j
SSINS No. : 6835
IN 85-88
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT r
WASHINGTON, D.C. 20555 4 //;'
November 18, 1985tesL V
. COLD.
IE INFORMATION NOTICE NO. 85-88: LICENSEE CONTROL OF CONTRACTED SERVICES
PROVIDING TRAINING
Addressees:
All nuclear power facilities holding an operating license (OL) or a construction
permit (CP).
Purpose:
This information notice is provided to emphasize to licensees their responsibility
for the content of safety-related training courses and materials provided by
consultants to utility employees. It is expected that recipients will review
the information for applicability to their facilities and consider actions, if
appropriate, to preclude the occurrence of similar problems at their facilities.
However, suggestions contained in this information notice do not constitute NRC
requirements; therefore, no specific action or written response is required.
Discussion:
In August 1985, the NRC learned that potentially misleading course material was
presented to licensee employees by a consulting firm as part of a training course
on containment leak rate testing. This material appeared to suggest and/or
condone practices that could be misleading in the conduct of an NRC inspection.
The following are some excerpts from the training course material entitled
"Interactions With the NRC. "
Should the utility inform the NRC of contemplated program
changes? This is debatable.
Alerting NRC opens up utility for comments and second
thoughts.
Springing changes on NRC has benefit of surprise.
Encourage [NRC] inspector to witness a Type C test, but. . .
don't be foolish:
Note: *Perform demo on an "easy" valve which has
traditionally not been a "problem leaker. "
8511150107 0 ii'
1
T �Z
IN 85-88
November 18, 1985
Page 2 of 3
*NRC will want to concentrate on past problem areas and pet
peeves.
Traditional industry approach to ILRT testing problems has been
predicated on not stating to NRC:
When the test began (thus allowing for repairs after
pressurization commenced)
The "Type A" test failed (since some smooth talkers have
managed to get out of failures)
We' ll do it over (since it has been possible to obtain
NRC agreement with statements such as, "you saw the test
before and didn'.t comment; why this time when we've done
it even better?")
*These statements appeared in the revised (March 1984) version
of the course as well as the November 1983 version.
During an inspection of the consulting firm, which took place after the NRC
became aware of the problem, it was found that the development and presentation
of this course material was an isolated instance. However, the inappropriateness
of this material would have been detected had the consulting firm's management
reviewed the quality of their product or had the licensee' s management reviewed
the training material before it was presented to their employees. Further,
although several of the licensee' s employees raised concerns (through course
evaluation forms) to the consulting firm about the appropriateness of the
presentation on interactions with the NRC, these concerns were apparently not
brought to the attention of licensee management in a timely fashion.
Licensees are responsible for the correctness of the material presented in
training courses at their facilities. All information and points of view
should accurately reflect a licensee' s position. As a result of the described
incident, in addition to specific corrective actions, the involved licensees
are reviewing their internal controls over contractor-provided training and
training material . Although during NRC followup the licensees involved and the
consultant' s training organization have stated that it was not their intent to
tell the licensee' s personnel how to mislead the NRC, the course attendees may
have received this impression. Because open and honest communications with the
NRC are a cornerstone of the regulatory process, it is imperative that licensees
assure themselves that all safety-related training materials accurately reflect
their position and philosophy.
IN 85-88
November 18, 1985
Page 3 of 3
No specific action or response is required by this information notice. If
you have any questions regarding this notice, please contact the Regional
Administrator of the appropriate NRC regional office or the technical contact
listed below.
Edward L. rdan, Director
Division o Emergency Preparedness
and Engineering Response
Office of Inspection and Enforcement
Technical Contact: E. W. Merschoff, IE
301-402-9045
Attachments: List of Recently Issued IE Information Notices
Or "`-
Attachment 1
IN 85-88
November 18, 1985
LIST OF RECENTLY ISSUED
IE INFORMATION NOTICES
Information Date of
Notice No. Subject Issue Issued to
85-87 Hazards Of Inerting 11/18/85 All power reactor
Atmospheres facilities holding
an OL or CP; and
fuel facilities
85-86 Lightning Strikes At Nuclear 11/5/85 All power reactor
Power Generating Stations facilities holding
an OL or CP
85-85 Systems Interaction Event 10/31/85 All power reactor
Resulting In Reactor System facilities holding
Safety Relief Valve Opening an OL or CP
Following A Fire-Protection
Deluge System Malfunction
85-84 Inadequate Inservice Testing 10/30/85 All power reactor
Of Main Steam Isolation Valves facilities holding
an OL or CP
85-83 Potential Failures Of General 10/30/85 All power reactor
Electric PK-2 Test Blocks facilities holding
an OL or CP
85-82 Diesel Generator Differen- 10/18/85 All power reactor
tial Protection Relay Not facilities holding
Seismically Qualified an OL or CP
85-81 Problems Resulting In 10/17/85 All power reactor
Erroneously High Reading facilities holding
With Panasonic 800 Series an OL or CP and
Thermoluminescent Dosimeters certain material
and fuel cycle
licensees
85-80 Timely Declaration Of An 10/15/85 All power reactor
Emergency Class Implementa- facilities holding
tion Of An Emergency Plan, an OL or CP
And Emergency Notifications
85-17 Possible Sticking Of ASCO 10/1/85 All power reactor
Sup. 1 Solenoid Valves facilities holding
an OL or CP
OL = Operating License
CP = Construction Permit
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