ML17292B093
| ML17292B093 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 10/21/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17292B091 | List: |
| References | |
| 50-397-97-16, NUDOCS 9710240310 | |
| Download: ML17292B093 (24) | |
See also: IR 05000397/1997016
Text
ENCLOSURE 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.:
License No.:
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved By:
50-397
50-397/97-1 6
Washington Public Power Supply System
Washington Nuclear Project-2
Richland, Washington
August 17 through September
27, 1997
S. A. Boynton, Senior Resident Inspector
G. D. Replogle, Resident Inspector
H. J. Wong, Chief, Reactor Projects Branch
E
Attachment:
Supplemental
Information
97i02403f.0 97i021
PDR'DOCK 05000397
8
EXECUTIVE SUMMARY
Washington Nuclear Project-2
NRC Inspection Report 50-397/97-16
.~Oerations
Operations
personnel
did not adequately
assess
available in'formation indicating
degraded
performance of the installed instrumentation for measuring reactor coolant
system
(RCS) identified leakage.
As a result, actions required to perform an
alternate method for s'atisfying the surveillance were delayed and the Technical
Specifications
(TS) required surveillance interval for evaluating
RCS operational
leakage was exceeded
(Section 01.2).
Operations
and engineering
personnel
did not demonstrate
a questioning attitude
following the identification of a broken lockwire on a safety-related
pressure control
valve associated
with the automatic depressurization
system (ADS). Verification of
the valve's pressure control setpoint was delayed
2 days.
The as-found setpoint
was determined to be below that to support long term operability of the associated
ADS valves (Section 08.1).
Maintenance
The implementation of the Foreign Material Control (FMC) Program was poor during
Refueling Outage R12.
Previous corrective actions to prevent recurrence
were
considered
weak (Section M8.1).
Encnineering
Problem Evaluation Requests
(PERs) were appropriately written in the majority of
instances when they were required.
However, one violation of procedures,
with
two exam'ples, was identified for the failure to write PERs for problems with
safety-related
equipment.
Some engineers
did not have an appropriate
understanding
of PER requirements
and were not using the applicable procedure
(Section E8.3).
Plant Su
ort
The failure of licensee personnel to recognize and address
potential radiological
concerns
associated
with several work activities resulted in unplanned
personnel
contaminations
and exposures.
A violation with three examples was identified.
In
an event involving the surveillance tests in the equipment drains radioactive
(EDR)
sump area, health physics (HP) and operations
personnel
did not properly address
the source of the contamination
and, as a result, a second equipment operator
(EO)
was contaminated
(Section R1.1).
0
0
Re ort Details
Summar
of Plant Status
The inspection period began with the reactor at 100 percent power.
On August 26, 1997,
power was temporarily reduced to 95 percent to compensate
for high main condenser
steam jet air ejector temperature that was a result of high ambient temperatures.
On
August 29, power was reduced to approximately 80 percent at the request of the
Bonneville Power Administration.
The plant was returned to full power on September
1.
On September
6 and 12, power was temporarily reduced to 90 percent to support gain
adjustments
to the digital feedwater control system.
During the weekends of
September
20-21 and September
27-28, power was again reduced at the request of
Bonneville Power Administration. At the end of the inspection period the plant was
operating at 80 percent power in economic dispatch.
I. 0 erations
01
Conduct of Operations
01.'I
Genera( Comments
71707
Using Inspection Procedure 71707, the inspectors conducted
frequent reviews of
ongoing plant operations.
The conduct of operations was generally professional
and
safety conscious.
01.2
Missed Surveillance for RCS Total Leaka
e
a.
Ins ection Sco
e 61726
During discussions with operations
personnel
on September
5, 1997,'the inspector
noted that TS Surveillance Requirement
(SR) for RCS total leakage had been
performed using an instrument which was not operating reliably. The inspector
conducted
followup to this observation.
b.
Observations
and Findin s
Background:
TS SR 3.4.5.1, in part, requires the licensee to monitor RCS total
leakage every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The licensee would normally perform this surveillance
utilizing, in part, the identified leakage totalizer (FDR-FQ-38) in the control room.
On September
4, 1997, operators noted a step change
in the identified leakage rate
surveillance results.
While normal leakage was approximately 2.5 gpm, the
surveillance indicated that leakage was approximately 0.1 gpm.
Operations
questioned
the operability of the totalizer and initiated steps to measure identified
leakage using an alternate method ("bucket test," in accordance
with Plant
Procedure
Manual (PPM) 2.11.3, "Equipment Drain System" ).
-2-
An EO performed the first bucket test at approximately 6 a.m. on September
4.
However, operations
did not have confidence
in this initial test because
the EO had
observed
a flow surge during the surveillance.
Also, upon exiting the controlled
surface contamination
area, the EO was found to be contaminated.
At approximately'7 p.m. on September
4, following a flush of the EDR flow
instrument line, an EO completed
a second bucket test and found identified leakage
to be about 2.5 gpm.
Upon exiting the controlled surface contamination
area, the
EO was again found to be contaminated
(see Section R1.1 for a detailed discussion
of the contamination events).
Later in the shift, qualitative evaluations
by the
operating crew on the performance
of the leakage totalizer indicated that the
instrument was not measuring
an identified leakage flow consistent with the latest
bucket test or'historical instrument readings.
However, the operating crew did not
declare the flow totalizer inoperable,
and consequently
did not convey to HP
personnel the nee'd to be able to perfom another bucket test early in the day shift to
meet the TS SR.
At 7 a.m. on September
5, 1997, in consideration of the two contamination events,
the shift manager postponed
the next bucket test until additional HP controls could
be established
for the job.
Operators performed TS SR 3.4.5.1 utilizing FDR-FQ-38,
but operations recognized that the surveillance was not valid, based
on the
instrument's erroneous
indications.
At 8 a.m., FDR-FQ-38 was officially declared
HP established
improved contamination controls for the job, and at
approximately 3:30 p.m. the surveillance was successfully performed.
NRC Assessment:
Based upon the valid bucket test performed on the evening of
September 4 and taking into consideration
the 25 percent grace period allowed by
TS 3.0.2, the subsequent
surveillance for RCS total leakage was required to be
performed no later than 10 a.m., September
5. The failure to perform the
surveillance within the required interval was identified as a violation of
TS SR 3.4.5.1
(VIO 50-397/97016-01).
The inspectors
considered
the failure of Operations
personnel
to promptly inform HP
of the need to establish better HP controls for the bucket tests to be a key
contributor to the violation.
Operations
had clear indication that FDR-FQ-38 was
- inoperable on the evening of September 4, but did not inform HP of the need to
enhance
the HP controls for the test until the FDR-FQ-38 was officially declared
inoperable at 8 a.m. on September
5.
Conclusions
Operations personnel
did not adequately
assess
available information indicating
degraded
performance of the installed instrumentation for measuring
RCS identified
leakage.
As a result, radiological controls were 'not promptly established
to support
an alternate, manual method for measuring
leakage and, as a result, the 'I S required
surveillance interval for evaluating
RCS operational leakage was exceeded.
-3-
Operational Status of Facilities and Equipment
02.1
En ineered Safet
Feature
S stem Walkdowns
71707
The inspectors walked down accessible
portions of the following engineered
safety
feature systems:
Standby Service Water Loop A
Reactor Core Isolation Cooling
Emergency
Diesel Generators,
Divisions I, II, and III
Residual Heat Removal System, Trains A, B, and C
Low Pressure
Core Spray System
The systems were found to be properly aligned for the plant conditions with no
notable material condition deficiencies.
08
Miscellaneous Operations Issues (92901)
08.1
Closed
Licensee Event Re ort
LER 97-008-00:
inoperability of four ADS valves
due to improper setpoint of containment instrument air (CIA) pressure control valve.
On July 16, 1997, the licensee identified, through
a system engineer walkdown, a
broken lockwire on Valve CIA-PCV-2B. Subsequent
investigation on July 17 also
identified that the valve stem locknut was loose.
Valve CIA-PCV-2B provides a
supply of nitrogen to the four Subsystem
B ADS valves from the safety-related
backup nitrogen bottles.
The supply of nitrogen from these bottles is designed to
provide an actuating force to open the ADS valves in support of long-term alternate
core cooling.
Normal supply to the ADS valves is from the nonsafety-related
containment nitrogen system, which is not relied upon in the licensee's
loss-of-coolant accident analyses.
In addition, each ADS valve has
a safety-related
accumulator, which provides for at least one, and up to five, actuations of the ADS
valve for depressurizing
the reactor pressure
vessel.
Troubleshooting
performed on July 18 found that the setpoint of Valve CIA-PCV-2B
was 63 psig, well below the normal setpoint of 180 psig.
The licensee determined
that the set pressure
was insufficient in supporting long-term alternate core cooling
with the Subsystem
B ADS valves.
The valve was promptly readjusted to the
appropriate
180 psig setpoint, following the troubleshooting,
to restore operability.
Based upon the licensee's identification of the broken lockwire on July 16, and the
subsequent
setpoint restoration on July 18, the Subsystem
B ADS valves were
determined to have been inoperable for long-term cooling purposes for at least
52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br />.
TS 3.5.1.G requires the plant to be placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
when two or more ADS valves are inoperable.
Thus, the condition identified by the
licensee was a condition prohibited by TS and reportable under the requirements
of
The inspector noted that the normal nitrogen supply and each individual ADS valve
accumulator were available for the duration of the time the valve was improperly
set.
Additionally, the licensee's
analyses
have shown that the operability of the
three Subsystem
A ADS valves would have been sufficient to provide the ADS
function of long-term alternate core cooling if the normal nitrogen supply was lost.
Therefore, the ADS depressurization
and long-term cooling safety functions were
maintained throughout this time period and the actual safety significance of the
event was considered
to be low.
Identification of the broken lockwire by the system'ngineer
demonstrated
a good
practice of inplant monitoring.
However, based upon the information available to
the licensee on July 16 and 17, and the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> action time provided by
TS 3.5.1.G, the timeliness of troubleshooting
efforts to determine the valve's
setpoint was inconsistent with the potential impact on ADS operability.
The weak
follow-through on the identified discrepancy
by operations
and engineering resulted
in the four Subsystem
B ADS valves being inoperable for an extended
period of time
that could have been avoided.
A violation of 10 CFR Part 50, Appendix B,
Criterion XVI was identified (VIO 50-397/97-16-02).
As a result of the licensee's investigation of this event, the licensee was unable to
determine the specific cause of the misadjustment of Valve CIA-PCV-2B and there
was no clear evidence that would indicate tampering was involved.
II. Maintenance
M1
Conduct of IVlaintenance
M1.1
General Comments
a.
Ins ection Sco
e 62703
61726
The inspectors observed
the following work activities:
Work Order HFW3, Replacement
of SW-V-49, low pressure coolant spray
pump motor cooler throttle valve
~
PPM 2.11.3, Drywell Identified Leakage Bucket Surveillance
In general, work was appropriately performed.
However, problems associated
with
the drywell identified leakage bucket surveillances
are discussed
in Sections 01.2
and R1.1.
-5-
M8
Miscellaneous Maintenance Issues (92902)
M8.1
0 en
Ins ection Followu
Item 50-397/97009-02:
implementation of the FMC
Program.
The inspector identified that the craftsmen and supervisors
associated
with the Diesel Generator
2 cooling water heat exchanger work had an inadequate
level of knowledge to ensure proper implementation of the WNP-2 FMC
requirements.
Additionally, the work instructions associated
with the job did not
contain steps to ensure that required FMC inspections were performed.
This
inspection followup item was established
to track additional followup regarding the
overall implementation of the licensee's
FMC program.
The FMC controls for WNP-2 are specified in PPM 10.1.13, "FMC for Systems
and
Components,"
Revision 14.
The program requirements
apply to both safety- and
nonsafety-related
work. As a minimum, the procedure
requires
a documented
FMC
inspection by a craft supervisor prior to system/component
closure.
In response
to the inspector's
original finding, the licensee retrained all craft
supervisors
regarding the WNP-2 FMC requirements.
Craft supervisors
were
instructed to ensure that an appropriate
FMC inspection was documented
in the
"Work Performed" section of the work order.
This documentation
was required
even if the work document did not itself specify an FMC inspection.
WNP-2
management
stated that the individual work orders did not need to specify when
the inspections
were required,
as the craft supervisors
were already trained to
perform and document these inspections without such prompting.
As part of the followup to the inspector's
concern, the quality assurance
department
performed additional inspection of work packages to check compliance
with FMC requirements.
The quality assurance
inspectors reviewed 15 work
packages
in which the scope of work would have required an FMC inspection and
.found that an FMC inspection was not performed for eight of the jobs (all jobs were
from Refueling Outage R12).
Furthermore,
an FMC inspection was only performed
when the work package specifically required it. When craft supervisors
were
expected
to perform the inspections without being prompted by the work
document,
the supervisors
consistently failed to perform the inspections.
The
finding was documented
in PER 297-0683.
Since the sample of work orders reviewed was composed
primarily of
nonsafety-related
work, and all of the instances where FMC requirements
were not
met were associated
with nonsafety-related
work, no violations of NRC
requirements
were identified.
Nonetheless,
the inspectors considered
the overall
implementation of the licensee's
FMC program during Refueling Outage R12 to be
poor.
The inspectors
also noted that there were no indications of foreign materials
in systems following R12.
-6-
As corrective measures
for the issue, the licensee initiated plans to retrain all
maintenance
personnel
and planners regarding FMC requirements
prior to the next
outage.
The inspectors considered
the corrective measures
to be weak.
Specifically, training the maintenance
staff was performed on two previous
occasions
(once before the fast outage and once after the inspectors identified that
FMC program requirements
were not being met), but maintenance
personnel were
still not implementing the FMC Program requirements,
as demonstrated
by
PER 297-0683.
The inspector considered
the absence
of specific guidance
in the
work request (to specify and document the inspections)
as a significant contributor
to this problem.
The licensee's
corrective actions had not addressed
this
contributor.
In response
to the inspector's concern, the licensee planned to
strengthen the corrective actions.
This item will remain open pending further
review of the licensee's
FMC practices by the NRC.
M8.2
Reo
en
VIO 50-397 95020-02:
Inspection Report 50-397/97-12 erroneously
closed this item.
The item number closed should be VIO 50-397/95020-01, which
refers to the issue (qualitative/quantitative
acceptance
criteria) discussed
in the
report.
M8.3
Closed
VIO 50-397/95020-01:
See preceding paragraph.
III. En ineerin
E8
IVliscellaneous Engineering Issues (92903)
E8.1
Closed
Unresolved Item 50-397 97003-03:
inoperable reactor water cleanup
instruments.
On February 11, 1997, the licensee identified that reactor water
cleanup flow Switches LF-FS-15 and LD-FS-16 (Division I and II) were inoperable
since initial calibration in Spring 1995
~ The setpoints were found to be 276.5 gpm,
while TS permitted a maximum setting of 271.7 gpm.
The licensee reported the finding to the NRC in LER 97-001.
This issue will be
reviewed and tracked in conjunction with the LER. This unresolved
item is
administratively closed.
E8.2
Closed
Unresolved Item 50-397 96017-01:
This item pertained to the licensee's
deferral of one test associated
with the reactor recirculation control (RRC) and
reactor feedwater (RFW) systems.
The licensee planned to trip one RFW pump
from 100 percent reactor power to verify proper operation of RFW and RRC scram
avoidance capabilities.
The test was deferred until the end of the operating cycle.
On March 27, operators performed the subject test.
Due to the unexpected
operation of the plant, operators manually scrammed the reactor.
The NRC
subsequently
conducted
a special inspection of the event (NRC Inspection
Report 50-397/97-10).
This item is closed based
on that inspection effort.
0
-7-
E8.3
Closed
Unresolved Item 50-397/96024-03:
failure to write PERs.
The licensee
had identified repetitive instances
where plant personnel were not initiating PERs
when required.
PERs 296-0834, 295-1195 and 196-0357 each documented
multiple examples of the procedural noncompliance.
Background:
At WNP-2, the PER program is governed
by PPM 1.3.12, "Problem
Evaluation Reports."
The program applies to both nonsafety-related
and safety-
related system, structures,
and components.
PPM 1.3.12 requires
PERs to be
written, in part, for the following conditions:
Conditions adverse to quality, such as failures, malfunctions, deficiencies,
deviations, defective material and equipment,
and nonconformances
associated
with safety-related,
augmented
quality items, Maintenance
Rule
scoped systems,
and those used in emergency operation procedures.
Corrective work on an item because
it does not meet specified requirements,
unless the work is rework.
System, structures,
and components
malfunction, damage,
or degradation
considered
sudden or unexpected,
or outside the anticipated performance of
the item.
Although other processes
are redundant to the PER process with regard to
corrective actions (i.e. work requests),
PERs are still required in most cases to
ensure that plant problems are appropriately addressed
and trended, i.e.,
Maintenance
Rule.
Additionally, plant managers
review the new PERs each day.
Part of the information considered
at the PER meeting is previous, but similar, PERs.
Failure to write a PER could result in masking problems from plant management,
thus not allowing management
the opportunity to ensure that appropriate corrective
actions are taken.
Furthermore, management
may not get an accurate perception of
equipment performance
when past and current PERs are reviewed during the PER
meeting.
NRC Assessment:
The inspector audited
a 6-week sample of operator logs (July 5
to August 16, 1997) to determine if PERs were being written, when appropriate, for
equipment problems.
The inspector found that in most cases
PERs were written when required.
However, performance was not consistent.
For example, the inspector identified
the following conditions that met the PER criteria, but PERs were not written.
~
Hydraulic Control Unit 4619 experienced
low accumulator pressure
alarms
on six occasions
between July 10 and August 10.
In response
to each
alarm, the accumulator was secured
and recharged.
Although a PER was
required to be written to document tl e degraded
condition (leaking
-8-
no PER was written until the accumulator failed on August
11
when it could not be recharged.
A PER would have likely prompted
management
into taking more effective corrective actions to preclude failure
of the degraded
The failure to initiate a PER for the degraded
condition (prior to accumulator malfunction) is the first example of a violation
of 10 CFR Part 50, Appendix B, Criterion V which requires that procedures
be implemented
(VIO 50-397)97016-03).
~
On July 21, 1997 the containment hydrogen monitor (CMS-SR-14) failed and
was declared inoperable.
The hydrogen monitor is described
as
safety-related
in the Final Safety Analysis Report.
The failure to initiate a
PER for the inoperable hydrogen monitor is another example of a violation of
10 CFR Part 50, Appendix B, Criterion V (VIO 50-397/97016-03).
~
On July 23, during the RFW pump trip test, the reactor vessel level control
system unexpectedly tripped to single element control.
Adjustable Speed Drive 1B1, gate turn-off problems were noted on July 8,
10, and 15.
(The gate turn-offs are solid state devices that convert DC
current to an AC signal to drive the RRC pumps.)
On July 21, the low pressure
core spray keepfill pump bearing oil resevoir
was empty.
This was unexpected
because
the oil reservoirs
are verified to be
at least half full twice a day and the bearings do not normally use a
significant amount of oil. Additionally, keepfill pump bearings have suffered
repetitive problems at WNP-2. Most recently, on October 16, 1996, the
bearing associated
with the RHR-P-3 keepfill pump failed and rendered
residual heat removal Train C inoperable.
Off-gas explosive monitors were found to be inoperable
on July 21, 28, and
30.
On July 28, a coupling associated
with control rod drive Pump 1A was found
in a damaged
condition.
Compensatory
steps were established
to prevent
pump failure, but no PER was written.
For the above examples, the inspector discussed
the issues with the cognizant
engineers.
The inspector observed that in all cases the engineers
were generally
aware of the issues, but did not have a proper understanding
of PER requirements.
One engineer stated that he was only required to write a PER if the system failure
resulted in a plant power reduction.
Another stated that he did not believe that the
failure of the safety-related
containment hydrogen monitor warranted
a PER.
None
of the engineers
had used the PER procedure when deciding whether or not to write
a PER.
-9-
C.
Conclusions
PERs were appropriately written in the majority of instances when they were
required.
However, one violation of procedures,
with two examples, was identified
for the failure to write PERs for safety-related
component
problems.
Several other
examples were also identified where PERs were not generated
for nonsafety-related
equipment problems.
Some engineers
did not have an appropriate understanding
of
PER requirements
and were not using the applicable procedure.
IV. Plant Su
ort
R1
Radiological Protection and Chemistry Controls
R1.1
Inade
uate Job Plannin
for Establishin
Radiolo ical Controls
a.
Ins ection Sco
e 71750
The inspectors reviewed the circumstances
surrounding
several recent plant
activities that resulted in unplanned
contaminations
and exposures
to personnel.
The review included the planning aspects
of the activities and the actions taken in
response
to the events.
10 CFR 20.1501 requires that surveys be made to evaluate the extent of radiation
levels and the potential radiological hazards that could be present to ensure
compliance with the requirements of 10 CFR Part 20.
10 CFR 20.1902 requires the
posting as a high radiation area if areas accessible
to personnel for radiation areas
greater than 0.1 rem/hr at 30 cm.
specifies limits for
occupational
exposures.
b.
Observations
and Findin s
Inadequate Area Posting:
On July 29, 1997, two contract workers were working in
the pit area of condensate
Filter/Demineralizer 1B, posted only as a radiation area.
During their work they received dose rate alarms on their electronic dosimetry
(indicating a dose rate ) 50 mr/hr) and exited the area.
Subsequent
surveys of the
pit area found general area radiation levels up to 150 mr/hr. As a result of the
workers'rompt action the dose they received was relatively low (9 mrem total).
Prior to the workers entering the pit area, the filter/demineralizer had been returned
to service following planned maintenance
performed several days before.
However,
neither the workers nor the HP technicians who granted them access to the area
recognized that there had been
a change
in plant conditions.
This resulted in the
workers performing work under an inappropriate radiation work permit (RWP).
Had
-10-
it been recognized that the area was, in fact, a high radiation area, additional
reviews of the activity would have been required to determine the appropriate
level
of radiological controls.
The root cause of this event appeared
to be poor communications
between
operations
and HP personnel with regards to changes
in plant conditions.
Additionally, HP personnel
did not demonstrate
the appropriate sensitivity towards
the need to verify that radiological conditions had not changed
prior to granting
access to the condensate
filter/demineralizer pit. The failure to survey the pit area
to evaluate the extent of radiation levels is the first example of a violation of
10 CFR 20.1501(a)(2)(iii) (VIO 50-397/97016-04).
Significant Personnel Contamination During Nonroutine Surveillance:
On
September 4, 1997, an EO was contaminated
while performing a manual
determination of identified RCS leakage into the drywell ~ A subsequent
whole body
count also showed an uptake of a small amount of Cobalt-60 (approximately
30 nCi).
Identified leakage into the drywell is collected by the equipment drain radioactive
system and directed to the EDR sump located on 422'evel of the reactor building.
Radioactive contamination
in the EDR system has resulted in the EDR sump area to
be posted
as a contaminated
The EO was required to enter this
area to perform the manual leakage determination.
Although the activity was a nonroutine surveillance being performed in an area that
is not normally entered, the EO performed the work under a Group RWP for routine
equipment operation in high and high-high radiation areas.
The RWP did not provide
any specific information on the radiological hazards
required the HP prejob brief to include a review of the most recent surveys of the
area.
The most recent survey available for the EDR sump was performed in
May 1997 and indicated contamination levels between 20,000 and
150,000 dpm/100 cm~.
The RWP also required catch containers when breaching
contaminated
liquid systems unless the liquid is directed to an approved drain
system.
The RWP did not provide any requirements for the type of catch container
or drain system that would be acceptable.
The use of an open polyethylene bottle
to collect the EDR flow was an ineffective catch container and a key contributor in
the EO's contamination.
Contamination on the EO's hands can also be attributed to
the removal of his protective gloves to read his dosimetry while in the contaminated
area.
HP surveys performed following the contamination event showed
contamination
levels between 80,000 and 4,000,000 dpm/100 cm~ in the EDR
sump area. The EO's egress from the area also resulted in spread of contamination
outside of the posted area.
HP personnel decontaminated
the affected areas
reducing levels around the EDR sump to (80,000 dpm/100 cm'.
The inspector considered
the root cause of this event to be poor radiological work
planning and practices
in relation to the potential radiological risks.
Additionally, HP
0
-11-
personnel
were not sensitive to the need to verify the radiological conditions around
the EDR sump prior to entry into the area by the EO. The failure to perform surveys
to evaluate the potential radiological hazards prior to the EO performing the
surveillance
is the second example of a violation of 10 CFR 20.1501(a)(2)(iii)
(VIO 50-397/9701 6-04).
Second Personnel Contamination in EDR Sump Area:
During the evening of
September
4, 1997,
a second
EO entered the EDR'sump area to perform another
manual determination of drywell identified leakage.
HP personnel,
believing the
contamination of the first EO was due to poor work practices, accompanied
the
second
EO to the job site, but did not require any additional radiological controls.
In
fact, the EO utilized the same Group RWP used by the first operator earlier that day.
The HP technician accompanying
the EO did not identify any concerns with the
EO's radiological work practices.
However, upon exiting the area the second
was also found to be contaminated.
A subsequent
whole body count also showed
an uptake of Cobalt-60 at a similar level to that of the first EO.
The root cause of the contamination appeared
to be from the significant
contamination
levels in the EDR system.
Based upon whole body counts performed
on each of the EOs, the licensee believed that the contamination
levels generated
airborne activity when the system was breached
and drained into the open bottle.
The erroneous
assumption
on the part of HP personnel that the initial contamination
was due only to poor radiological work practices prevented
them from appropriately
considering the radiological risks involved with breaching the system and taking
action to minimize that risk.
Following the second contamination event, HP personnel
again decontaminated
the
Additional engineered
and personnel
radiological controls were
established
under a specific RWP approved
by the radiation protection manager.
Subsequent
entries into the EDR sump area by EOs showed that the additional
controls were effective in minimizing the potential for personnel contamination.
Inadequate Surveys Resulted in Unplanned Personnel Contaminations:
On
September
11, 1997, four individuals were found to have low levels of
contamination (1000 - 8000 dpm/100 cm') after performing work on control rod
drive hydraulic Pump 1A. Subsequent
surveys by HP personnel found
contamination around the pump motor couplings and gear box.
These components
were not identified as potentially contaminated
prior to the job and no prejob
surveys had been performed.
From discussions
with the radiation protection manager it was iden.ified that the HP
supervisor was aware of the impending job, but directions to an HP technician to
evaluate the job site was not adequately
communicated.
Thus, no direct HP
coverage
was provided during the work. It was also identified that two of the
maintenance
personnel
involved with the job were qualified to perform
contamination surveys.
-1 2-
The root cause of this event was a lack of sensitivity on the part of both
maintenance
and HP personnel for the need to verify radiological conditions before
starting the work. The failure to survey the work area is the third example of a
violation of 10 CFR 20.1501(a)(2)(iii) (VIO 50-397/97016-04).
Conclusions
The failure to recognize and address
potential radiological concerns associated
with
several work activities resulted in unplanned
personnel contaminations
and
exposures.
In the event involving the surveillance tests in the EDR sump area, HP and
operations
personnel
did not properly address
the source of the contamination
and,
as a result, a second
EO was contaminated.
V. Mana ement Meetin s
X1
Exit Meeting Summary
The inspectors presented, the inspection results to members of licensee management
after
the conclusion of the inspection on October 2, 1997.
The licensee acknowledged
the
findings presented.
The inspectors
asked the licensee whether any materials examined during the inspection
should be considered
proprietary.
No proprietary information was identified.
ATTACHMENT
Supplemental
Information
PARTIAL LIST OF PERSONS CONTACTED
Licensee
P. Bemis, Vice President for Nuclear Operations
D. Coleman, Acting Regulatory Affairs Manager
D. Hillyer, Radiation Protection Manager
J. Hunter, ALARASupervisor
P. Inserra, Licensing Manager
T. Messersmith,
Corporate Emergency Preparedness,
Safety and Health Officer
M. Monopoli, Operations Manager
G. Smith, Plant General Manager
J. Swailes, Engineering
Manager
R. Webring, Vice President Operations Support
INSPECTION PROCEDURES USED
IP 37551:
IP 61726:
IP 62707:
IP 71707:
IP 71750:
IP 92901:
IP 92902:
IP 92903:
Onsite Engineering
Surveillance Observations
Maintenance
Observations
Plant Operations
Plant Support
Followup - Operations
Followup - Maintenance
Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
~Oened
50-397/9701 6-01
failure to perform TS in a timely manner required surveillance
for identified leakage
50-397/9701 6-02
inadequate
corrective actions related to improperly adjusted
CIA valve for ADS valves
50-397/97016-03
failure to write PERs when required
50-397/97016-04
failure to perform radiological surveys
-2-
Closed
50-397/97008-00
LER
inoperability of four ADS valves due to CIA-PCV-28 pressure
setpoint discovered set less than required
50-397/97003-03
inoperable reactor water cleanup isolation instruments
50-397/96017-01
deferral of reactor feedwater pump trip test
50-397/96024-03
failure to write PERs
50-397/95020-01
inappropriate qualitative/quantitative
acceptance
criter for
control rod drive housing support installation (closed as
50-397/95020-02
in IR 50-397/97-12}
Discussed
50-397/97009-02
IFI
50-397/95020-02
implementation of the FMC Program
(reopened
due to erroneous
closure in Inspection
Report 50-397/97-12}
LIST OF ACRONYMS USED
CIA
FMC
IFI
LER
NRC
PER
SR
TS
WNP-2
automatic depressurization
system
containment instrument air
equipment drains radioactive
equipment operator
foreign material control
health physics
inspection followup item
licensee event report
U.S. Nuclear Regulatory Commission
problem evaluation request
Plant Procedures
Manual
reactor feedwater
reactor recirculation control
radiation work permit
surveillance requirements
Technical Specifications
unresolved item
violation
Washington Nuclear Project-2