ML20125C840
| ML20125C840 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/03/1992 |
| From: | Hague R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20125C810 | List: |
| References | |
| 50-373-92-27, 50-374-92-27, NUDOCS 9212140102 | |
| Download: ML20125C840 (23) | |
See also: IR 05000373/1992027
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U.S. NUCLEAR REGULATORY COMMISSION
REGION 111
Report No.
50-373/92027(DRP); 50-374/92027(DRP)
Docket Nos.
50-373; 50-374
License Nos. NPF-ll; NPF-18
Licensee:
Commonwealth Edison Company
Opus West III
1400 Opus P1 ace
Downers Grove, IL 60515
facility Name:
LaSalle County Station, Units I and 2
Inspection At:
LaSalle Site, Marseilles, Illinois
Inspection Conducted: October 14 through November 25, 1992
Inspectors.
D. Hills
C. Phillips
M. Miller
J. Roman, Illinois Department of Nuclear Safety
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Approved By:
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M [,/I.y
/2/3/# L
F. I~. Hague, ' Chief
Date
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Reactor Projects Section 1C
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inJpection Summary
inslection from October 14 throunh November 25. 1992 (Reports No. 50-373/92027
,(D_R_P_11_50-174 /92027 (DRP) ) t
Areas Inspected: A routine, unannounced safety inspection was conducted by
the resident inspectors and an Illinois Department of Nuclear Safety
inspector.
The inspection included followup on previously identified items
and licensee event reports; review of operationhl safety, monthly maintenance,
surveillance activities; safety assessment and quality verification; and
report review.
Results: Three violations were identified concerning the following:
Two examples of failure to follow fuel handling procedures
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caused by inattention to detail (paragraphs 4.d(3)(a) and
4.d(3)(b)). Two additional examples of failing to follow
procedure caused by inattention to detail involving the
reactor water cleanup (RWCU) and liquid radioactive waste
systems (paragraphs 4.a(2)(a) and 4.a(2)(b)).
9212140102 921204
ADOCK 00000373
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Three examples of inadequate procedures, all involving
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surveillances (paragraph 6.a).
Failure of the out-of-service (00S) program to ensure an
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adequate review of potential impacts prior to tuplementing
an 00S (paragraph 4.a(3)(a)).
Two non-cited violations were identified involving the following:
An inadequate work package for reactor vessel disassembly
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(paragraph 2).
Modifying the plant without ensuring updating of applicable
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design drawings (paragraph 2).
ELant Operations
Performance declined in this area.
Several examples of operational personnel
failure to follow procedure due to inattention to detail were noted. Two of
these examples involved fuel handling activities.
Similar problems were no;ed
in the two most recent, previous refuel outages.
Additional management
attention was warranted in this area.
Failures of operating personnel to
follow procedure were noted as a concern earlier this year. However, licensee
actions to address this previous concern appeared to be effective as the prior
examples were due to a different root cause.
In addition, operating shift
management knowledge of 005 effects prior to implementation was lacking in
some cases. This was contributed to by a programmatic deficiency.
Implementation of the licensee's shutdown risk guidelines was progressing
well.
Maintenanc.e/Surveill ancg
Performance remained steady in this area. The quality of procedures remained
a concern as three examples were identified thich resulted in adverse events
or inadequate testing during surveillances. An inadequate work package for
reactor vessel disassembly was also_ identified. However, a repeat
verification by a mechanical maintenance shift supervisor was conscientious
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and a good example of self-checking. The absence of actions to address a
power line with minimal vehicle clearance did not reflect an aggressive
attitude toward problem identification and resolution. Although licensee
actions to address control room ver.tilation radiation monitor spiking were
lacking previously, actions taken in 1992 were aggressive and showed a marked
improvement in sensitivity toward problem resolution.
1.ikewise, licensee
actions to address reactor core isolation cooling (RCIC) system failures were
much more aggressive than in previous years. Actions to prevent spurious RCIC
initiations during surveillarces were ineffective and not aggressive.
However, the licensee appeared to be addressing this problem in a more
reasonable manner following a recent September 1992 RCIC initiation. The
licensee continued to successfully reduce an already low backlog of corrective
maintenance.
Plant management remained committed to keeping control room
equipment in good working condition. Plant management was developing enhanced
indicators to better manage maintenance backlog.
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Enoineerino/ Technical Succoti
Performance remained steady in this area. A weakness in the licensee's
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document control system was f dentified which allowed plant modifications
without ensuring updating of applicable design documents.
Licensee actions in
response to a series of fuel pool ccoling pump trips were reasonable and
further corrective actions will be monitored for effectiveness.
Some
nonoutage corrective work requests assigned to the technical staff were not
addressed in a timely manner.
Actions taken to address spurious RCIC
initiations through a modification were reasonable but not particularly
aggressive until this year.
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Radioloaigal Controls
Performance remained steady in this area.
The licensee took more aggressive
dose reduction initiatives than in the previous Unit 2 refuel outage.
However, radworker performance and cyclical housekeeping remained a problem.
Safety Assessment /0uality Verification
The inspectors noted some progress being made but additional improvement was
still necessary in the timeliness of procedure revisions.
This adversely
impacted on the timeliness of corrective actions for quality verification
findings. The implementation of event screening meetings was a positive step
toward problem identification and resolution.
Recent quality verification
staff reductions did not appear to have an initial adverse effect but will
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continue to be evaluated for the longer term.
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PETAILS
1.
D n ons Contacted
G. J. Olederich, Manager, LaSalle Station
- W. R. Huntington, Technical Superintendent
- J. V. Schmeltz, Production Superintendent
D. S. Berkman, Assistant Superintendent, Work Planning
H. Hentschel, Assistant Superintendent, Operations
- J. Walkington, Services Director
J. Lockwood, Regulatory Assurance Supervisor
M. Santic, Assistant Superintendent, Maintenance
- K. Kociuba, Quality Verification Superintendent
- Denotes those attending the exit interview conducted on
November 25, 1992.
The inspectors also talked with and interviewed several other licensee
employees during the course of the inspection.
2.
Licenseo Aq. tion on Previously Identified items (92701 and 92702)
(Closed) Unresolved Item (50-373/92021-02 (DRP)): During reactor vessel
disassembly, the extension legs for the ste&m separator lifting rig were
installed and in the work package the latching mechanism was verified to
function properly.
Following shift turnover, the day-shif t mechanical
maintenance +,upervisor decided to recheck the latching mechanism
operation despite the completed verification step. He identified that
the latches did not operate properly and that the air actuation lines
were improperly connected.
The inspectors regarded the repeat
verification to be a conscientious effort on the part of the day shift
supervisor.
The supervisor who signed the verification step in error had completed a
prerequisite step on the preceding page of the work package which was
similar to the verification step in question.
The supervisor believed
the verification step was a repeat of the prerequisite step and therefore
signed the step.
While the work package steps were correct as written,
the steps were insufficiently clear as to ensure proper comprehension by
the workers.
Failure to provide instructions appropriate to the
circumstances is a violation of 10 CFR Part 50, Appendi.x B, Criterion V.
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lhe licensee planned to incorporate clearer instructions into the work
package. The licensee identified this violation and it is not being
cited because the criteria specified in Section VII.B.2. of tne " General
Statement of Policy and Procedures for NRC Enforcement Actions,"
(Enforcement Policy, 10 CFR Part 2, Appendix C (1992)) were satisfied.
(Closed) Unresolved Item (373/92021-01):
Review concerns with reactor
core isolation cooling (RCIC) system steam line leak detection
instramentation.
The inspectors verified the existence of diverse RCIC
leak detection and isolation methods including ambient and differential
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temperature and low steam line pressure.
In addition, the concern
involved only one division of the primary containment and reactor vessel
isolation control system. Adequate assurance existed that a RCIC steam
line leak would be detected and the system would automatically isolate.
Therefore, the inspector has no further concerns regarding the lack of
multiple differential pressure instruments to provide isolation of
specific valves upon high steam line flow rate.
However, the inspectors noted that a failure of a single switch would
necessitate closure of the isolation valve in accordance with technical
specifications, rending RCIC inoperable and having a negative impact on
system availability.
A review of licensee event reports (LER) indicated
that this occurred three times in 1990-1991.
Through a review of licensing basis documents, the inspectors could not
find any references or credit taken for the differential pressure
detection of a RCIC instrument line break and resulting automatic
isolation of RCIC.
Therefore, the inspectors have no further concerns
regarding an out-of-service isolation of these instruments.
The inspector determined that differential pressure switches lE31-N013BB,
IE31-N007AB, and IE31-N007B8 had been entirely removed in the field to be
used as spares elsewhere in the plant. This was accomplished through
work requests in August 1990. This process was in accordance with
LaSalle Administrative Procedure (LAP)-300-7, " Preparation and Control of
Nuclear Work Requests," Revision 7, step F.1.W and LAP-240-6, " Temporary
System Changes," Revision 25, step E,1.
However, the process did not
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ensure corresponding plant drawings including M-2101, sheets 1 and 3, IE-
1-4226AD, and IE-1-4226AF were revised to match the plar.t configuration.
The licensee planned to remove all RCIC instrument line break switches
including these through a modification which ultimately would have caused
the drawings to be revised. However, this modification still had not
been implemented at the time of the inspection.
ANSI N18.7-1976, " Administrative Controls and Quality Assurance for the
Operational Phase of Nuclear Power Plants," section 5.2.15, required
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procedures for control of documents and changes thereto to preclude the
possibility or use of outdated or inappropriate documents.
The
licensee's practice of modifying the plant (deleting plant equipment)
through work requests and the out-of-service program without ensuring
revision of the design documents is contrary to ANSI N18.7-1976 as it
allows the use of inappropriate documents which no longer reflect plant
design.
This is considered a violation of 10 CFR 50, Appendix B,
Criterion VI as implemented by Regulatory Guide 1.33, Revision 2, which
endorses ANSI N18.7-1976. The licensee initiated a revision to LAP-300-7
to require a temporary system change in similar cases.
Safety
significance was minimal in this particular case as these particular
switches were not required.
The violation was categorized as a Severity
Level V and it is not being cited because the criteria specified in
Section VII.B.1 of the " General Statement of Policy and Procedures for
NRC Enforcement Actions," (Enforcement Policy,10 CFR 2, Appendix C,
(1992)) were satisfied.
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(Closed) Open Item (373/83-29-01):
IE Notice.82-49, " Correction For.
Sample Condition For Air and Gas Monitoring," was closed in inspection
report 50-373/83-29; 50-374/83-28.
However, certain modifications were-
determined to be necessary and the completion of these modifications were
assigned the tracking numbers 373/83-29-01 and 374/83--28-04.
Inspection
report 50-373/89021;.50-374/89021 closed 374/83 28-04 administiatively.
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Due to an oversight, 373/83-29-01 was not added to the list of items that
were administratively closed. This item is also considered closed.
(Closed) Violation (373/92008-01):
Several examples of non-licensed
operators failing to follow procedure.
Review of licensee corrective
actions is described in paragraph 7.b.
(Closed) Open Item (373/92008-06):
Review licensee actions in regard to
encouraging worker initiated procedure changes. This. item is discussed
in paragraph 6.a of this report.
Further actions in this regard will be
tracked through violation 373/92027-03 (DRP). 'This item is considered
closed.
No cited violations, two non-cited violations, and no-deviations were
identified in this area.
3.
. Licensee Eylat. Beoorts fgJlqwn (9270A1.-
The following licensee event reports were reviewed to ensure that
reportability requirements were met, and that-corrective actions, both
immediate and to prevent recurrence, were accomplished in accordance with
the technical specifications:
(Closed) LER 373/92009 Spurious Auto Start of Control Room Ventilation
Emergency Make-Up Train Due to High Radiation Spike
(Closed) LER.374/92010 RCIC Steam Line Outboard Isolation Valve Motor
Damage Due To Torque Switch failure To Trip
(Closed) LER 374/92013 Reactor Core Isolation Cooling: System Spurious
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Initiation During-LIS-LC-403-Due To A Pressure Spiket
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(Closed) LER 373/92011 Wrong High Radiation Door Downgraded Due to
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Persnnnel Error
(Closed) LER 374/92011 Spurious Auto Start of Control Room Ventilation
Emergency Makeup Train Due to High Radiation Spike
(Closed) LER 373/92010 Unit 1 Automatic Reactor Scram Due to Low
Charging Header Pressure
In addition, recent deviation reports (DVRs) were reviewed in order to
monitor conditions related to plant or personnel performance and to -
detect J,ential development of trends. Appropriate-generation and-
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disposition of DVRs, in accordance with the Quality Assurance Manual,
were also reviewed.
No violations or deviations were identified in this area except as
identified in this or other inspection reports.
4.
Onerati_caglJaf11y Verif_itAtJ.gp_(1QJ_LQ_anJ 717Ql).
The inspectors reviewed the facility for conformance with the license and
regulatory requirements,
a.
On a sampling basis the inspectors observed control room activities
for proper control room staf fing, coordination of plant activities;
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adherence to procedures or technical specifications; operator
cognizance of plant parameters and alarms; electrical power
configuration; and ;be frequency of plant and control room visits by
station managers. Various logs and surveillance records were
reviewed for accuracy and completeness.
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Significant observations were:
(1) The inspectors observed several licensee shutdown risk review
board meetings conducted during the refuel outage.
This meeting,
held on a weekly basis, verified implementation and emphasis of
the licensee's shutdown risk guidelines. Licensee management's
ability to instill a shutdown risk sensitivity among plant
workers, was apparent. An example was a nuclear station operator
(NS0) ideritifying a single closed valve preventing draining of
the vessel, such that additional administrative controls were
placed upon that valve. Shutdown risk level assessment sheets
were being completed on a regular basis. The inspectors noted
good dissemination of r.hutdown risk status (such as current risk
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level and available systems) among plant workers.
(2) The inspectors noted two examples of failing to follow procedures
caused by inattentton to detall during operating organizat1on
activities described below.
Two additional examples are
discussed in paragraph 4.d(3) of this report. Although each
incident had minimal safety significance, in aggregate they
indicated inattention to datail during several different facets
of operating activities and a need for greater managewent
attention in this area.
(a) While unisolating RW:V filter demineralizer 2A por LaSalle
Operating Procedure (LOP)-RT-06, " Reactor Water Cleanup
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System Filter /Demineralizer Precoat," Revision 21, RWCU
automatically isolated on high differential flow. The safety
significance cf the event was minimal as the containment
isolation system performed its design function. Steps
F.15.a.2 and F 15.a.4 of the procedure required the upstrean
resin drain stop valves to be closed and step F.15.b required
verification that these valves were closed. Operators
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indicated these steps were performed.
Partial flew through
these valves caused the isolation, indicating inadequate
conduct of these steps.
(The licensee subsequently performed
a leak check of these valves which indicated no leakage.)
The failure to properly implement these steps was due to
inattention to detall and is an example of a violation (50-
374/92007-Ola (DRP)) of technical specification 6.2.A,1 which
required adherence to procedures.
(b) While transferring water from the Unit 2 chemical waste
collectino tank to the chemical waste process tank per
LaSalle Operating procedure (LOP)-WZ-03, " Chemical Waste
Collector Tank Transfer To Chemical Waste Process Tank And
Vice Versa", the Unit 2 waste sludge tank was overfilled and
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approximately 1200 gallons spilled onto the floor. The
safety significance of this event was minimal as the water
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was returned to the radwaste system and did not significantly
contribute to the radiation levels already existing in the
waste sludge tank room.
Step F.2 of the procedure required
the operator to blowdown the chemical waste collector tank to
the waste sludge tank per LOP-WZ-16, " Chemical Waste
Collector Tank Sludge Transfer To Waste Sludge Tank".
The
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operator performed that action, but subsequently returned the
line-up back to the waste sludge tank instead of the chemical
waste process tank.
Step F.5 of LOP-WZ-03 stated, " VERIFY
proper ficw by observing:
a.
Decrease on Chemical Waste
Collector Tank Level Recorder OLR W2051.
b.
Increase on
Chemical Waste Process Tank Level Recorder OLR-WZO52." The
operator failed to properly verify the increase in chemical
waste process tank level due to inattention to detail.
This
is an example of a violation (S0-374/92027-Olb (DRP)) of
technical specification 6.2. A.1 which required adherence to
procedures.
(3) The inspectus idadifica a concern regarding required operator
knowledge of 00S c'fects prior to implementation.
This concern
resulted from two events described below:
(a) A Unit 2 scram occurred on Nvember 16, 1992, during conduct
of 00S 2-2202-92 on the Unte 2 station air compressors (SAC)
to allow s aduled maintenaice. The Unit 0 and 1 SACS were
in an abn
~ al lineup, being supplled cooling water from the
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Unit 2 tuiLine building closed cooling water (TBCCW) system
through crosstie valves rather than the normal Unit 11BCCW
supply.
The DOS isolated Unit 2 TBCCW from the Unit 2 SAC
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which also isolated flow to the crosstic valves. The Unit 0
and 1 SACS tripped on high lube oil temperature.
The loss of
air pressure caused control problems with the turbine driven
reactor feed pumps such that the main turbine eventually
tripped on high reactor water level. The automatic scram
occurred due to the main turbine trip.
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The shift supervisor (SS) who assigned the tagout was aware
of the abnormal system configuration and that the 00S was for
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the Unit 2 SAC. However, he possessed insufficient knowledge
of the affects of the-005, as he failed to-realize it
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involved TBCCW.
The SS relied upon the facts that the 00S
was a scheduled activity, the shift engineer was aware it was
to be performed, and knowledge that the Unit 2 SAC was not
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operating. The operators assigned to hang the tagout were
unaware of the abnormal lineup.
LaSalle Administrative Procedure (LAP)-900-4, " Equipment Out-
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Of-Service Procedure," Revision 48, did not require any
review of an 00$ against the existing plant configuration for
possible adverse affects at the time of-implementation.
(A
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safety related or technical specification related 00S did
require a review only to ensure technical specification
requirements were still met.) As it was impossible when an
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00S was written to always anticipate the exact plant
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configuration and an informed, conscientious operator was the
last barrier to such an event, the inspectors regarded a
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procedural requirement for such a review to be imperative.
This included the need for specific implementation guidance.
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This appeared to be a programmatic problem at LaSalle.
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ANSI N18.7-1976, " Administrative Controls and Quality
Assurance for.the Operational Phase of Nuclear Power Plants,"
Section 5.2.6, required prior to granting permission for
release of equipment, operating personnel shall verify that
the equipment can be released, determine how long it may be
out-of-service, and to document such granting of permission.
As operating personnel did not properly verify that the
equipment could be released and document granting of that
permission, nor does the licensee's administrative program
require such action, this is considered a violation -(50-
374/92021-02 (DRP)) of 10 CFR 50, Appendix B, Criterion XIV
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as implemented by Regulatory Guide 1.33, Revision 2, which
endorses ANSI N18.7-1976.
(b) On October 29, 1992, while implementing 00S 1-1065-92 on IB
reactor protection system electrical bus, the standby gas
treatment (SBGT) dampers IVG001 and-2VG001 (inlet isolation
dampers) unexpectedly opened due to the loss of power.
Although personnel who had prepared the 00S expected this to.
occur, this knowledge was not adequately communicated to the
operating crew. Therefore, the operating crew had incomplete
knowledge of the .affects of the 00S.= An Attachment F,
'" Safety or Technical Specification Related-Equipment Outage
Form," was completed, indicating it.was reviewed for
technical specification implications and authorized under
that basis.
Safety significance was minimal as SBGT fan
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operation was previously defeated and there was no-adverse
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consequences of the dampers repositioning other than
initially confusing the operators,
b.
On a routine basis the inspectors toured accessible areas of the
facility to assess worker adherence to radiation controls and the
site security plan, housekeeping or cleanliness, and control of fic1d
activities in progress.
Significant observations were:
(1) The inspectors noted a deterioration of radworker practices,
similar to that noted during the last Unit 2 refuel outage.
Examples (including a violation) are discussed in inspection
report 50-373/92026; 50-374/92026.
Housekeeping was also noted
to be cyclical as it was during the previous refuel outage.
(2) The inspectors noted increased licensee dose reduction actions
during the refuel outage. A standpipe was placed in the drain of
the equipment drain sump underneath the vessel to maintain a
constant shielding level as the sump was a significant source of
dose. Also, a temporary filtering system for tha water in the
sump was installed. As the control rod drives (CRD) were
removed, a certain amount of crud was flushed out the bottom and
down into the sump. The filtration reduced dose due to crud
build up in the sump. Finally, a modification was performed to
lower the height of the CRD housir.g support structure under the
vessel.
It was estimated that the time to remove the steel for
CRD replacement was reduced by approximately an hour and one-
hal f.
These actions appeared to be very effective, reducing the
dose from CRD replacement by approximately 30 percent.
In
addition, the licensee substantially increased hydrolazing
activities to reduce dose in different areas of the plant.
Hydrolazing efforts were very minimal during the last Unit 2-
refuel outage.
c.
Walkdowns of select engineered safety features (ESF) were performed.
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The ESFs were reviewed for proper valve and electrical alignments.
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Components were inspected for leakage, lubrication, abnomi
corrosion, ventilation, and cooling water supply availab.:ity.
Tagouts and jumper records were reviewed for accuracy where
appropriate,
d.
Refueling Activities
The inspectors verified that. refueling activities were being
conducted and controlled as required by procedures except as noted
below. This was done on a sampling basis through direct
observations, discussions with licensee personnel, and review of work
requests and procedures.
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Significant observations were:
(1) CECO was informed on September 18.-1992', of a potential problem
with the fuel-received for the Unit I refuel- outage. As a-
precaution, General Electric (GE) requested that'the fuel be
shipped back to GE for additional _ testing. As a result,
replacement fuel arrived after the start of the refuel outage.
The inspectors observed new fuel receipt and inspection on a
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sampling basis to ensure adequate coordination of activities on
the refuel floor. At one point, fuel unload was temporarily
suspended while decontamination activities progressed for the new
fuel receipt 4nspection area. The inspectors regarded this as a
conservative action to limit distractions to_ fuel handlers during
-fuel unload. As GE supplied personnel-to handle new fuel.
receipt, the inspectors reviewed-the' training and qualifications
of these GE personnel. No problems were identified during this
review.
(2) While observing fuel unload, the inspectors noted that the
digital position indication for bridge movement was inoperable.
The licensee indicated that this system was no longer utilized
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because inaccuracies could be introduced over a short period of
use. Although not required, the inspectors regarded this
indication, when functioning properly, as a barrier to a fuel
bundle mispositioning event.. This was in consideration of a
previous Dresden fuel bundle mispositioning event -in which the
licensee indicated a similar condition was a contributing cause.
This importance was expressed to licensee management.
(3) Two adverse fuel handling incidents described below occurred
during fuel receipt and fuel unload.
A review of licensee's
deviation and informal reports for the last_two years indicated a
fuel bundle misorientation during the last-refuel outage and a
fuel bundle misposition in the refuel outage prior to that. Most
involved the fuel- pool and_ all were due to inattention to detail.
Although each incident had minimal safety significance, in
aggregate they indicated a need for greater management attention
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in this area.
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(a) On October 15, 1992, in the process of defueling Unit I a
fuel bundle struck the-cattleshoot approximately four-inches
down- from the bale handle. - The . safety significance- of the
event-was minimal as inspections indicated the fuel damage
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was very slight.
The cause of the event was inattention to detail on the part
of the fuel handling crew. A phone talker shift turnover
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taking place on the refuel bridge distracted both the fuel
handler and the senior reactor operator such that the fuel-
mast not being fully retracted was not noticed. The General
Electric (GE) cylindrical mast stopped movement in the upward
direction unless the fuel handler slowed the upward velocity
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when the bottom telescoping section was about to be fully
withdrawn into the next section.
When the fuel handler
heard the mast upward motion stop, he moved the refueling
bridge toward the cattleshoot without checking the digital
height indicator, " Normal Up Light," or looking at the fuel.
LaSalle Fuel Procedure (LFP)-400-1, " Fuel Movements Within
The Reactor And Spent Fuel Storage Pools", step F.1.n,
stated, " Raise grapple and fuel assembly by using the right
hand controller grapple raise / lower lever until hoist readout
indicates zero (grapple full up) or until the ' NORMAL UP
LIMIT' light illuminates on the left hand controller.". The
failure to perform this action is an example of a violation
(50-373/92027-Olc (DRP)) of technical specification 6.2.A.1
which required adherence to procedures.
(b) On October 24, 1992, the licensee identified that two new
unirradiated fuel bundles, YJ3709 and YJ3710, for the Unit I
refuel outage, had been placed in the wrong storage locations
in the Unit 2 fuei pool contrary to the nuclear component
transfer checklist. The fuel handler (a GE contractor)
improperly verified the serial numbers of the fuel bundles as
they were being passed to him from the new fuel inspection
stand. The fuel handler had been trained on how to properly
verify information for quality assurance records and the fuel
handlers were previously told what was expected of them when
handling fuel.
Periodic supervision was provided by GE
supervisors as opposed to Ceco personnel.
Safety
significance was minimal as the fuel bundles were virtually
identical as far as internal construction and the reactivity
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of the Unit 2 fuel pool was unaffected.
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LaSalle Fuel Procedure (LFP)-100-2, " Administrative Control
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Of Transfer Of Fuel Or Special Nuclear Material Between Or
Within the Spent Fuel Pool (s) Or Vaults", step F.3., stated,
'
"As the steps on the Nuclear Component Transfer List are
completed, they will be VERIFIED, initialed and dated by the
Fuel Handling Supervisor." Failure to follow the procedure
is considered an example of a violation (50-373/92027-Old
I
(DRP)) of technical specification 6.2.A.1.
In the process of
l
implementing one of the corrective actions for this event,
.
consisting of a more thorough audit of the new fuel bundles
l
in the fuel pool, the licensee also identified one
!
misoriented fuel bundle.
l
l
(4) On October 22, 1992, while moving new fuel from the new fuel
inspection stand to the fuel prep machine, the jib crane stopped
'
working. The fuel bundle was hanging from the jib crane over the
edge of the spent fuel pool. After ensuring the bundle was in no
danger of falling, the jib crane boom was hand cranked back over
the floor. The bundle was left suspended about two feet over the
new fuel vault cover while electrical maintenance repaired the
12
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jib crane. A jib crane power cable had caught on the edge of the
crane pedestal structure, causing a short and subsequent tripping
,
of the lifting motor breaker. The jib crane was repaired and the
fuel bundle placed back in the new fuel inspection stand.
(5) The inspectors reviewed fuel pool _ cooling / cleaning operations and
noted several fuel pool cooling pump trips in June, October, and
November 1992. The onsite nuclear safety group and plant
management reviewed these trips and proposed several corrective
actions. The inspectors will rialuate the effectiveness of these
corrective actions as the3 we incorporated.
The licensee
appeared to be addressing this problem in a reasonable manner.
Two violations, including one with four examples, and no deviations were
identified in this area.
5.
lignthly Maintenance Observation (62703)
,
,
l
Station maintenance activities affecting the safety-related and important
to safety systems and components listed below were observed or reviewed
to ascertain that they were conducted in.accordance with approved
procedures, regulatory guides and industry codes or standards,-and did
not conflict with technical specifications.
"
!
The following maintenance activities were observed and reviewed:
,
WR L83942
Steam condensing Pressure Switch Modification No, M-1-1-88-
026
WR L59953
High Pressure Core Spray (HPCS) Underground Piping
,
Modification M-1-1-86-072 Electrical Construction Test
Procedure #19 Control-Circuits
WR LO2281
Unit 1 Division III Battery Upgrade
WR L18135
Reactor Recirculation Pump IB Breaker 38 Did Not Trip During
Testing
'
WR L12225
Perform LaSalle Electrical- Procedure (LEP)-EQ-146 Motor -
Operated Valve (MOV) Inspection, Refurbishment, and Votes
Test of IE12-F0643
'
WR L13409
Disassemble, Inspect, and Reassemble IE51-F0ll
WR Ll4091
- 0- Diesel Generator Small Bore- Tubing Replacement
WR L15561
Replace the Degraded Voltage Relays With 1TE27N Undervoltage
Relays for Division I Switchgear 141Y
WR LO2279
Replace the Division III Battery Charger
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d
WR LO2281
Replace the Division III Batteries
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WR L12210
Perform Inspection Per LEP-AP-101 for Non-Segnented Bus Duct
from Unit Auxiliary Transformer 141 to Bus 151
i
i
WR L13359
Disconnect, Dress, and Reconnect Bolted Connections to Bus
143
WR LO5924
Apply Tape to Insulate All Exposed Bus Cable and Duct
!
Connections in the Rear of Switchgear 143
WR L59953
Electrically Modify High Pressure Core Spray So It Takes
Suction From the Suppression Pool
WR L97339
1B Diesel Generator Small Bore Tubing Replacement
,
1'
Significant observations included:
'
a.
On October 19, 1992, a maintenance mechanic driving a mobile crane,
l
struck a 480 volt power line.
The_ power line crossed an access road
on wooden utility poles and su) plied a nearby storage building.
The
l
crane boom was down as it was seing moved and a spotter was not being
utilized. The licensee's policy guide on operation of vehicles
l
within the protected area required the boom to be down unless
transporting a load. A spotter was required only when the boom was-
not down or if the vehicle was being driven within certain boundaries
,
l
around high voltage equipment.
The licensee had previously thought
sufficient clearance existed with the boom in the down position and
cranes had used that-path before. The crane may have hit a pothole
with the resulting bounce negating the minimal clearance for that
electrical line.
No personnel were injured and the vehicle was not
damaged as a result of this_ event. Other than a loss of power to the
storage building, the plant was not affected by this event.
The licensee identified other lines within the protected area with
the potential to cause overhead clearance problems. These were
,
marked with warning signs giving the clearance height.
In addition,.
l
the license permanently restricted further vehicle access to the
location of the event. The licensee also planned to implement a
periodic surveillance of line heights to ensure that over time they
do not sag below the height depicted on the warning signs.
,
A mechanical maintenance foreman stated that prior to the event, he
had casually mentioned in passing to a general foreman, that when
using this road, care should be taken under this electrical line.
The general foreman could not remember this conversation. The method
of conveying this information did not reflect an aggressive and
- proactive attitude toward problem identification and resolution.
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b.
The inspectors performed an historical review of licensee actions in
response to two recurring equipment problems noted below.
Both were
prevalent in 1992 as well as previous years.
The inspectors
concluded that the licensee was addressing these concerns in a much
more aggressive and proactive manner in 1992 than previously.
(1) The inspectors reviewed licensee actions as to the long history
of spurious control roem ventilation radiation monitor spikes
causing control room emergency ventilation automatic actuations.
Prior to 1992, this problem was not addressed in a proactive
manner. As a result of a July 17, 1988, and previous actuations,
a modification to change the actuation logic was planned as
indicated in LER 373/88016. Although this would not address
radiation monitor spiking for unknown reasons, it would prevent
resulting system actuations. This planned modification did not
receive adequate priority to be pursued aggressively.
LERs
373/91008 and 91010, depicting three additional actuations in
1991, again referred to the modification and estimated a late
1992 completion date. Other t'1an verifying correct operation
with surveillance procedures and replacing a power supply board,
no other corrective actions were mentioned in these LERs.
The licensee was much more aggressive in addressing this
recurring equipment problem in 1992 showing a marked improvement
in sensitivity.
Four additional actuations occurred between May
and September 1992 due to spurious radiation monitor spikes (LERs
373/92007, 373/92009, 374/92007, and 374/92011).
Various actions
'
were taken during this time to address the problem including
troubleshooting with a vendor technician and later a vendor
senior design engineer, taking apart, cleaning, and inspecting
the detectors, additional sealing of the units, replacing the GM
tubes and amphenol connectors, and modification of the
'
calibration procedure to change the way the background was set.
However, these actions still were not effective. The licensee
planned an additional modification of the detector electronics to
filter out fluctuations in the signals. The licensee also
,
started pursuing the languishing 1988 planned modification to the
I
actuation logic and estimated an early 1993 completion date.
(2) The inspectors noted several reactor core isolation cooling
(RCIC) system failures in 1992, resulting in increased
unavailability of RCIC. Certain aspects of those failures were
discussed in inspection reports 50-373/92010; 50-374/92010; 50-
373/92013; 50-374/92013; 50-373/92016; and 50-374/92016. As a
result, the inspectors reviewed LERs from 1990 through 1992 to
,
ascertain any common causes for these failures. A total of 11
.
safety system failures were reported for RCIC which could be
i
grouped into the following categories:
(a) Unit 1 RCIC overspeed trips due to the governor valve
sticking open.
Four occurrences were noted in June 18, 1990,
July 20 1991, October 23, 1991, and April 6, 1992.
Following
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the last failure, the licensee verified binding between the
valve stem and carbon ring.- This was believed due to steam
lenkage through the steam admission valve causing an
t
unfavorable environment within the governor valve.
(b) Unit 2 RCIC exhaust vacuum breaker upstream containment
1 solation valve tripped on thermal overloads necessitating
.
closure of isolation valve per technical specifications.
-
Following an initial trip on June 15, 1992, current trace
data, motor meggering data, motor winding resistance, motor
current data, and bus voltage data did not indicate any
~
4
problems.
Plans were instituted to provide more frequent
'
l
lubrication of the valve stem.
Following lubrication on July.
14, 1992, a similar failure occurred during valve stroking.
The licensee performed a VOTES test and current traces with
no problems noted.
Current traces taken on an increased
frequency indicated degradation in relation to the time
period between valve cycling. 'As a result, the valve was
-
repacked, upon which a junk-ring at the. bottom of the
stuffing box was found cocked and the valve stem slightly
worn in this area.
'
(c) Unit 2 steamline outboard isolation valve failure. This
occurred on August 10, 1992, due to a torque switch failure
as it was set-above the limitorque recommended value for the
operator due to Generic Letter 89-10, Supplement 3 concerns.
<
This event is being evaluated in more detail in conjunction
,
with the currently ongoing NRC motor operated valve
inspection at LaSalle.
,
(d) Unit 1 RCIC inoperable due to 250 volt battery low
,
electrolyte temperature.
This_ occurred only once on December
25, 1990, due to ventilation system design problems'.
l
l
(e) RCIC steamline high flow switch failure necessitating closure
of isolation valve per technical specifications. This
occurred on May 11, 1990, August 1, 1990, and June 21, 1991,
i
and is discussed-in paragraph 2 of this report.
Due to the excessive number of failures of RCIC, the licensee formed
a task group to concentrate on RCIC turbine problems. Three major
review initiatives resulted including a' maintenance history
performance review by General Electric, a system vulnerability
assessment by Bechtel, and evaluation of maintenance work practices
.
!
by Technicon.
Beside repair of the steam admission valve and RCIC
turbine overhaul, numerous extensive recommendations were developed
!
including component. replacements, modifications, and changes in
.
procedures and operational practices. The licensee was either
implementing or still evaluating these recommendations.
l
16
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y .,. , ,,..
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9
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__
_
_-_
_
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_-
i
The causes of failures were very diversified between the five
,
categories and did not :;ignify any obvious maintenance program
,
deficiencies. While the licensee's approach to the first two.
i
.
governor valve events was reasonable, it could not be described as
~
aggressive toward problem resolution. The licensee's response to the
t
third governor valve failure was more aggressive but stopped short of
addressing all potential root causes. The response to the latest
-
governor valve failure was very aggressive', reflected a more detailed
and defined approach to failure investigation, and was more
,
encompassing as to potential root causes.
The licensee's response to
i
the vacuum breaker isclation valve was reasonable although not
particularly aggressive as repacking was not done until after the
second failure and degradation was shown to exist.
I
c.
The licensee continued to successfully reduce the backlog of
corrective maintenance. The NRC operational safety team inspection
(OST!) conducted in November 1991, described the number of
1
outstanding nonoutage corrective work requests (690_at that time) as
six weeks of backlog and nioderately low.
Since the Spring 1992
refuel outage, this number averaged between 550 and 600 and was
,
reduced to less than 500 just prior to the start of the e,urrent
-
refueling outage.
However, the OSTI also noted about 50 percent of
the corrective work requests w He older than three months and
recommended evaluation of support to craft activities to help reduce
this time.
The licensee addressed this concern by periodically requesting status
of older work requests from the various departments and more recently
adding more emphasis by producing a ten oldest work request list for
departments. Although these efforts appear to have been partly
effective, further improvement was still warranted. While the
percentage of nonoutage corrective work requests greater than three
months old assigned to electrical, mechanical, and instrument
maintenance appeared to be reasonable,- those assigned to other work
groups was still high. These groups included engineering and
construction, operational analysis division, operations, substation
construction, and technical staff (TS).
The inspectors performed a inore detailed review of the backlog
concentrating on the TS for further insight. Of the 142 nonoutage,
corrective work requests greater than six months old, 42 were
assigned to the TS. The inspectors noted that total open TS
corrective nonoutage work requests had declined significantly over
the past year indicating good progress.
Hcwever, 68 per cent of the
TS backlog was greater than six months old, showing continued efforts
needed on older items. The inspectors selected a sample of four of
these work requests for specific review and concluded insufficient
justification existed for the long delay-in _two of these items.
Although they were of generally low safety significance, justifying a
lower priority, it was apparent that some items were still not being
addressed in a timely manner.
17
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The number of pending control room corrective work requests remained
4
low, showing a strong commitment to keeping control room equipment in
4
good working condition.
The inspector noted that plant management was developing and
evaluating enhancement to maintenance performance indicators to
sssist in management evaluation and control of work request backlog.
No violations or deviations were identified in this area.
6.
Monthly Surveillance Observation (61726)
Surveillance testing required by technical specifications, the safety
'
L
analysis report, maintenance activities or modification activities were
observed 2nd/or reviewed.
Areas of consideration while performing
observations were procedure adherence, calibration of test equipment,
j
identification of test deficiencies, and personnel qualification. Areas
of consideration while reviewing surveillance records.were completeness,
i
proper authorization / review signatures, test results properly
i
dispositioned, and independent verification documented. The following
,
activities-were observed / reviewed:
LaSalle Mechanical Surveillance (LMS)-DG-01 Main Emergency Diesel
Unit Surveillances
,
t
LaSalle Electrical Surveillance-(LES)-GM-308 Unit 1 Southern Division
.
Operational Analysis Division Inspections and Calibrations for the
'
Diesel Generator 0 System
LES-GM-301 Calibration and Functional Testing of Unit 14kV Emergency
Bus Loss of Voltage Relays by Operational Analysis ' Division
LES-GM-130 Inspection of Westinghouse Motor Control Center Equipment
!
LES-GM-109 Inspection of 480 Volt Klockner-Moeller Motor Control
l
Centers
I
i -
Significant- observations included:
l
a.
The inspectors noted the following three procedural inadequacies,'_two
resulting in unnecessary challenges to safety-related equipment, and
i
i
the other involving inadequate testing.
Numerous examples of
-
-
i
procedural weaknesses were noted in inspection report 50-373/92008;
50-374/92008.
The licensee had previously implemented a weekly joint
procedure review meeting to provide better quality and more timely
'
procedure reviews. This was especially. designed to show quick
i
i
turnaround of worker initiated procedure changes to encourage more
worker identification and ultimate correction of existing procedure:
deficiencies. These additional examples -indicated continued problems
'
with new and existing procedures.
However, due to the nature of the
licensee corrective-actions one would expect, not an immediate, but_a
'
-
more long term resolution of existing procedural- deficiencies.
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,
(1) On October 9, 1992, while performing LES-RD-102, "L' nit 1
Alternate Rod Insertion Division 1 Logic Functional Test,"
Revision 1, a Unit I scram occurred.
The unit was in a refueling
^
outage at the time with the reactcr vessel head removed.
,sa
prerequisite, the procedure rcforred to Attachment B with
information on the test and its effects upon plant operation.
The attachment listed conditions the plant must be in prior to
performing specific steps in the procedure.
Among these were the
mode switch being in the shutdown position and the control rod
drive (CRD) charging water header supply stop valve being c!csed.
The attachmont did not delineate any specific order for these
conditions.
The stop salvo closure was accomplished first,
resulting in low CRD charging water header pressure and a reactor
If the mode switch had been taken to shutdown prior to closing
the valve, the scram would not bsve occurred since it was active
only in operational conditions 2 and 5.
The operators' attention
to detail was lacking as they had been trained on this particular
scram signal and should have recognized the consequences of their
actions.
Safety significance of the event was minimal since the
reactor was already shut down, limiting affects upon the plant to
primarily unneeded wear on the control rod drives. This
'
procedure was inappropriate to the circumstances in that it did
not delineate a specific applicable steo sequence that would have
prevented the scram.
Therefore, this is considered an example of
a violation (373/92027-03a (DRP)) of 10 CFR 50, Appendix B,
Criterion V,
'
(2) Technical specification 3.6.1.2 required the RWCU discharge
valves 1G33-F040 and 2G33-F040 to be type C leak rate tested.
This was accomplished in accordance with LaSalle Technical
Surveillance (LTS)-100-ll, "Feedwater Outboard Stop and RWCU
Return Valves Local Leak Rate Test," Revision 8.
The inspector
noted a check valve between the RWCU discharge valve and its
required vent path for the local leak rate test.
This procedure
was inappropriate to the circumstances as the location of this
check valve invalidated the test results as performed by LTS-100-
11.
This is considered an example of a violation (50-373/92027-
03b (DRP)) of 10 CFR S0, Appendix B, Criterion V.
The licensee was unable to positively show that the 0.6 la for
combined type B and C leakage required by technical
specifications was met. The licensee subsequently tested 1G33-
F040 in the reverse direction with satisfactory results.
The
test on 2G33-F040 required partial draining of the Unit 2
feedwater system and Unit 2 was operating. Therefore, the
licensee requested and was granted a waiver of compliance from
the Office of Nuclear Reactor Regulation (NRR) until an emergency
technical specification change could be processed to delay the
testing of 2G33-F040 in the reverse direction until an outage of
sufficient duration.
The configuration of feedwater and RWCU, in
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addition to previous testing results, provided sufficient
confidence of containment operability in the interim.
(3) An unplanned emergency control room ventilation actuation-
occurred on October 16, 1992, due to causes other than radiation
monitor spiking. Because of a change in how the detectors were
calibrated, background radiation was not always sufficient to
perform the technical specification required channel check
surveillance with a unit shut down.
LaSalle Limited Procedure
(LLP)-92-159, " Alternate Method For Performing Channel Check of
VC Intake Radiation Monitors," Revision 0, was written to perform
the surveillance in those cases. The procedure war deficient as
it failed.to recognize that a greater than two second delay
between returning the monitor switch to the."0PERATE" position
and depressing the red tr_ip light during the surveillance would
cause the. actuation logic to seal in.
The technical staff group
leader who authorized the procedure and the onsite reviewers had
insufficient knowledge of detector operation to identify this
potential . As the procedure was inappropriate to the
circumstances, it is considered an example of s, violation (50-
373/92027-03c-(DRP)) of 10 CFR 50,-Appendix B, Criterion V.
This
represented an unnecessary challenge to safety-related equipment.
Another unplanned emergency control room ventilation actuation
occurred the following day, October 17, 1992, while performing
the same evolution. The operators were aware of the previous
event through the turnover process.
To prevent a spurious
initiation while performing the surveillance,.the operator placed
the train into pull-to-lock. Upon completing- the surveillance
but prior to taking the train out of pull-to-lock, the operator
depressed the reset push button.
By design, this should have
prevented an actuation. The button was either not depressed
sufficiently hard or-long enough to reset the signal. As a
i
result, another initiation occurred when the train was taken out
I
of pull-to-lock. The-inspectors regarded the second incident as
isolated and having a different root cause than the _first
incident.
(b) The inspectors performed an historical review of _ licensee actions in
response to recurring:RCIC system spurious-initiations.
LERs from
1990 through 1992 showed three RCIC initiations due to pressure
transients during surveillances and four RCIC initiations from
.
pressure transients during reactor scrams caused by a turbine trip.
l
l
Following a Unit 2 turbine trip on September 12, 1990, the licensee
l
initiated a study to. determine if a modification was feasible to
'
filter or delay short duration pressure spikes.
The corrective
actlons following a March 1,.1992, Unit I turbine trip referred to
l
!
the study being done for the previous: September 1990, event.
The
l
corrective actions following an August 27, 1992, Unit 2 turbine trip
l
discussed a modification, to be reviewed, to insert a time delay to
prevent a spurious pressure signal from being transmitted.
20
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o
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Corrective actions to address spurious RCIC initiation during the
latest November 16, 1992, turbine trip were not yet formalized.
The spurious RCIC initiations during surveillances all occurred while
conducting LaSalle Instrument Surveillance (LIS)-LC-303, " Unit 1 Main
Steam Isolation Valve Leakege Control Inboard Reactor Vessel Pressure
FJnctiOnal Tost," or corresponding procedure for Unit 2, LIS-LC-403.
The licennee determined the cause of the first initiation on July 24,
1991, to be a stuck isolation valve on a pressure transmitter.
The
corrective actions were geared toward replacement of the isolation
valve and identification of other such valves.
The cause of a second
event on June 22, 1992, was thought to be a packing leak on an
instrument stop valvs.
Corrective actions included training of
instrument .naintenance (IM) personni , restricting performance of
this surveillance to more experienced personnel, revision of the
nrocedure to include cautions as to this problem, and to evaluate
replacement of the stop valve with an anti-surge valve.
Cw rective
actions for a September 17, 1992, event were to replace the
instrunent stop valves with anti-surge valves, bypass the Rfl0
initiation logic during the performance of- this surveillance, and -
reference was made to the time delay modificatiun.
The inspectort discussed the history and progress of the review of
the level sereor time delay modification with the cognizant technical
staff engineer.
Initially, no hardware solution to the problem
appeared to Le available. When a hardware solution was found, the
initial scope of the modification was to install a time delay relay
in all reactor level signal circuitry.
This option was dropped
because it was considered too expensive and the problem did not
appear ".o be urgent.
Due to the marked increase in the rate of
actuaticas in 1992, the licensee approved a modification to install
time delay relays in the RCIC and high pressure core spray (HPCS)
-
logic circuitry.
The RCIC modification for Unit I was scheduled for
the current refuel outage.
The corrective actions taken by the licensee in regard to the time
delay modification were reasonable but not aarticularly aggressive
until this year.
The corrective action; ta an by the licensee in
regard to oreventing spurious RCIC Initiations during surveiliances
prior to t1e September 1992 event were ineffectivo and not
aggressive. The inspectors will evaluate the current set of
corrective actions as they are implemented for long term
'
effectiveness,
t:ne violation with three examples, anst no devlauons were identified in
this area.
7.
hf_giv Assenguttt_and Ouality Verification 14_Qi@l
a.
The inspectors reviewed tha open quality verift:ation (QV) corrective
action records to verify actions were being taken in a reasonable
mannen
Except for ore described below, r.o problems were noted for
21
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items belonging to the previous Nuclear Quality Programs (NQP)
Howrver, there were many item, past their original due date on the
onsite nuclear safety (ONS) list.
From review of the items and from
'
discussions with OkS personnel, the inspector ascertained many of the
,
items were procedure revisions.
The excessive time required to
complete procadore revisions at LaSL11e was previously identifie<l in
inspection report 50-373/92010; 50-374/p2010
falew of precedure
revision turnaround statistics indienteo a w pror/ess was being made
in this area but further progress was nw e 1ary.
The following two
,
examples were indicative of the problem:
'
>
(1) On February 26, 1992, a licensee quality verification (QV)
,
inspector identified loose 'natorial nn the catwalk of the reactor
,
building overhead crane which possed over the reactor cavity and
the spnt fuel pools. 7he items discovered were not accounted
i
for as loose items in the area of the reactor cavity and spent
-
fc 1 pools.
Licensee management agreed to monthly housekeeping
inspections of the crane as part of 1.nSalle Mechanical
Surveillance (LMS)-HC-0), " Station Cranes and Holst Preventative
1
Maintenance and Examinations." This change did not occur until
seven months later on September 29, 19924
(2) 1he inspectors noted that the licensee still had not developed
pclicies or procedures to govern the Lie down of large loads when
,
moving them within site boundaries.
A; indicated in inspection
report 50-373/92008; 50-374/92003, this concern originated from a
previous Dresden event in which a large radwaste shipping cask
.
fell off a truck whose path was near incoming power lines.
As a
l
result of the inspectors' inquiries, plant management requested
ONS to review this area.
This review, documented in the January
1992 ONS monthly report, resulted in action item record 373-352-
92-00404 to develop approp"iate guidance with an original due
date of August 1, 1992.
This would have been prior to the next
'
scheduled refueling outage in consideration of protention of
i
power sources and shutdown risk.
,
The licensee was developing changes to the procedure review process
and accountability to shorten the turnaround time.
Effectiveness of
those actions will be evaluated in future inspection reports.
b.
The inspectors reviewed licensee actions to address previous operator
procedural adherence o ncerns.
These concerns were discussed in
inspection report 50-373/92008; 50-374/92008.
The inspectors
verified completion of Ilcensee actions in response to violation 50-
l
373/92008-01.
In addition, the licensee formed a task group which
consolidated procedural adherence expectations from various
'
procedures into a single page, understandable documant and
disseminated it to plant personnel. The root cause of the procedura)
.
adherence violation examples described in paragraphs 4.a and 4.d of
<
this report differ from those of the previous concerns.
previously,
'
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they resulted from a combination of not understanding licensee
management procedural adherence expectations and a lack of respect or
22
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attitude toward procedural adherence.
Recent exemples involve
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inattention to detail errors.
The licensbe's prw ieus corrective
,
actions appear to have beon effective as no similar probitms to the
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previous concerns have been recently noted (except for radworker
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practices described in inspection report 50-373/92026; 50-374/92025.)
c.
To ascertain event screening capability, the inspectors attended
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several licensee daily events meetings.
The licensee had recently
.
instituted utilization of this arocess to ensure the correct tracking
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mechanisms in accordance with tie event significance and appropriate
1
level of needed followup. Attendees were from several pertinent work
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group:, ensuring a multidisciplined approach,
Review of inputs (such
as logs) since the last meeting were detailed, tending to provide a
lower threshold of event identification.
A review of deviation and
informal report generation numbers since instituting the daily event
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n,eeting supports this conclusion. Although the rate of deviation
report generation remained roughly steady, the rate for informal
reports increased substantially.
The inspectors regarded this
process as a positive step toward problem identification and
.
resolution.
d.
Through intt.rviews with plant and quality verification (QV) personnel
>
and review of various reports, the inspectors ascertained the effect
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of recent staffing reductions involving QV.
ONS and NQP groups were
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combined into a QV grou3 and total staffing cut in half.
Despite an
expected reduction in 11e number of audited activities, this groun'is
maintained positive attributes previously identified with ONS.
Th
included maintaining a big picture overview, trend identification,
and detailed evaluations with recommendations beyond those of plant
personnel.
The inspectors will continue to evaluate th's area for
more long term effects,
j
No violations or deviations were identified in this area.
,
8.
Report Review (9011H
During the inspection, the inspector reviewed selected licensee reports
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and determined that the information was technically adequate, and that it
satisfied the reporting requirements of the license, technical
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specifications and/or 10 CFR as appropriate.
!
No violations or deviations were identified in this area.
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9.
fjLit Intervigg
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The inspectors met with lieansee representatives (denoted in Paragraph 1)
during the inspection per 4d and at thn conclusion of the inspection
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period on November 25, If,J2. The inspectors summarized the scoao and
results of the inspectin and discussed the likely content of t11s
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inspection report. The licensee acknowledged the information and did not
indicate that any of the information disclosed during the inspection
could be considered proprietary in nature.
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