ML20134F184
| ML20134F184 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 08/08/1985 |
| From: | Greger L, Lovendale P, Miller D, Nicholson N, Paul R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20134F179 | List: |
| References | |
| 50-454-85-22, 50-455-85-20, NUDOCS 8508210010 | |
| Download: ML20134F184 (25) | |
See also: IR 05000454/1985022
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-454/85022(DRSS); 50-455/85020(DRSS)
Docket Nos. 50-454; 50-455
Licenses No. NPF-37; CPPR-131
Licensee:
Commonwealth Edison Company
Post Office Box 767
Chicago, Illinois 60690
Facility Name:
Byron Nuclear Power Station, Units 1 and 2
Inspection At:
Byron Site; Byron, Illinois
Inspection Conducted:
May 6, 7, and 10; June 5-7, 10-14 and 27, July 8-9,
and 22, 1985
8N!$I
Inspectors:
D. E. Mil er
Date
8/7!6
R. A.
ul
Date
nb
8/7/8I
N. A.
i ho
Date
8/8//f~
P. C. Lovendale
Oate
.R.Gege,$ief
6/7//75"
Approved By:
Facilities Radiation Protection
Date
Section
Inspection Summary
Inspection on May 6, 7, and 10; June 5-7, 10-14, and 27; July 8-9, and 22, 1985
50-454/85022(ORSS); 50-455/85020(ORSS))
(Reports No.
Nonroutine, announced inspection of the radiation protection
Areas Inspected:
circumstances surrounding personal exposures greater than
program including:
station limits; a personal contamination incident; the radiation protection
program associated with startup activities; ALARA program; exposure controis;
training and qualifications; organization and management controls; two unplanned
volume control tank releases; an unplanned boric acid evaporator release;
Unit 2 condensate sump overflow; and licensee action on previous findings.
The inspection involved 182 trispector-hours onsite by five NRC inspectors.
8508210010 850000
ADOCM 05000454
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Results:
Several violations were identified (failure to follow radiation
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protection procedures - Sections 5, 6, 7, and 13; inadequate instructions to
workers - Section 5; inadequate high radiation area controls - Section 5;
inadequate evaluation of radiological conditions - Section 6; lack of adequate
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procedures - Sections 5 and 14).
Enforcement conferences were held June 27,
1985 and July 22, 1985 to address the inspection findings.
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DETAILS
1.
Persons Contacted
W. D. Britton, Quality Insurance Inspector
W. Burkamper, Quality Assurance Supervisor
R. A. Chrzanowski, Security Administrator
T. P. Joyce, Operating Engineer
J. Langan, Licensing
W. McNeill, General Instructor
R. E. Querio, Station Superintendent
F. Rescek, Technical Health Physics Supervisor, CECO
D. St. Clair, Technical Staff Supervisor
M. Snow, Compliance Department
J. R. Van Laere, Radiation-Chemistry Supervisor
G. Wagner, Power Operations Manager, Ceco
R. C. Ward, Assistant Superintendent Administration and Support Services
K. T. Weaver, Station Health Physicist
J. A. Hinds, Senior Resident Inspector, NRC
The above personnel attended the June 14, 1985 exit meeting.
The inspectors also contacted members of the operations, rad / chem,
technical, mechanical, training, security, and engineering staffs during
this inspection.
2.
General
This inspection, which began at 10:30 a.m. May 6,1985, included reviews
of the circumstances surrounding a personnel entry into the incore motor
drive area resulting in exposures exceeding station limits, personnel
entry into containment and subsequent exposures greater than administrative
limits, and a contamination incident resulting from maintenance work on a
CVCS valve.
Also reviewed were neutron surveys conducted at specified
power levels; two unplanned VCT releases; circumstances regarding an
unplanned boric acid evaporator release and overflow into the Unit 2
condensate pump area; routine operations of the ALARA, exposure control,
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contamination control, and training programs;
and rad / chem staff
qualifications and stability.
Direct surveys of the plant taken during
tours were in general agreement with licensee data.
3.
Licensee Action on Previous Findings
(0 pen) Open Item (454/85014-020):
Review neutron survey data.
The
licensee continues to conduct and review neutron surveys.
This is
discussed more completely in Section 8.a.
4.
Organization and Management Controls
The inspector reviewed the licensee's organization and management
controls for the radiation protection and radwaste programs including
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changes in the organizational structure and staffing, and effectiveness
of procedures and other management techniques used to implement these
programs.
Twenty-seven qualified radiation-chemistry technicians (RCTs), are
available for shift work; RCTs perform both chemistry and health physics
functions.
Seven RCTs in training are expected to be qualified by July,
1985.
RCTs report to one chemistry foreman and six health physics
foremen who in turn report to a lead foreman.
The professional staff
includes health physics and chemistry support, in addition to ALARA,
-GSEP, and TLD coordinators.
Management and professional staff report to
the Station Health Physicist who in turn reports to the Radiation-Chemistry
Supervisor.
The Radiation-Chemistry Supervisor is the station's radiation
protection manager as defined by Regulatory Guide 1.8.
The Radiation-
Chemistry Supervisor reports to the Assistant Superintendent for
Administration and Support Services in accordance with TS 6.2.1.
Staff stability was reviewed.
This department has experienced a fairly
low turnover rate during the past two years; two RCTs left the organiza-
tion and three were promoted.
A total of 42 RCT positions are allotted
for Unit 1 operations and Unit 2 startup testing.
Until those positions
can be filled, the licensee plans to rely on contracted technicians, who
are expected to be onsite within the next month.
A significant amount of
overtime has been necessary to complete assigned radiation protection
coverage and technical specification surveillances and samples; the RCT
staff has been working 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts, five days a week since January
1985.
Relief is anticipated with the newly qualified RCTs and contracted
technicians.
No apparent violations were identified.
5.
Personal Exposures Greater Than Administrative Limits
On May 1, 1985, the licensee notified the resident inspectors that two
station electrical maintenance (EM) workers exceeded their administrative
dose limits (100 mrem / day) while working in a high radiation area (HRA)
earlier that day.
The following scenario was developed based on regional
inspectors' interviews with the following involved personrel:
shift
engineer (SE), technical staff engineer, health physics (HP) foreman,
EM foreman, two RCTs, and two ems.
At approximately 11:00 p.m. , the "A" incore detector became stuck behind
the seal table shield wall during Unit 1 flux map startup testing conducted
on the backshift of April 30-May 1, 1985.
(Refer to Attachment I for a
diagram of the incore detector drive train.) The reactor was at 48% power.
A faulty electrical relay in the "A" incore drive motor on the 411' level
of containment was thought to be the problem.
The SE initiated a work
request for two ems to locally reset the relay.
Relatively low radiation
levels were expected in the vicinity of the incore drives since the
incore detectors, other than the stuck incore, were in their shielded
storage positions and the stuck incore was behind a shield wall.
However, before the work commenced, the technical staff developed a
temporary procedure to remotely free the detector, anti the work request
for containment entry was cancelled.
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The temporary procedure failed to release the "A" detector, and at
appro.Timately 2:30 a.m., the SE telephoned the HP foreman to reinstate
a worf. request for the local incore drive repair.
The SE expressed an
urgency to complete this repair.
The HP foreman outlined Radiation
Work Permit (RWP) options, including a Type II RWP for emergency entry
which did not require a prejob survey but which specified continual RCT
surveillance.
This type of RWP was selected and used for the first time
in the plant's brief operating history.
During the attempt to remotely
free the stuck detector using the improvised temporary procedure, the
other five moveable detectors were withdrawn from their previously
shielded positions to the 5 path positions at the incore drive motor
area, near the relay which was suspected to have malfunctioned.
The
technical staff and control room personnel involved in the flux mapping
procedure were aware of the new detector positions and the associated
potential radiological hazard, but did not convey this information to the
Although the radiation environment was discussed
earlier in preparation for the canceled work, no comparable discussion
occurred between the SE and the HP foreman for this work. The HP foreman
was not aware of the change in the detector positions nor the
corresponding high radiation fields.
His actions were based on previous
survey data of the drive area (approximately 5 mR/hr) when the detectors
were in a shielded position.
The SE was also unaware of the changed
incore detector positions and the resultant radiological hazard in the
vicinity of the incore drive motors.
The SE notified the EM department to prepare for the job.
The HP foreman
began to complete RWP 50147 (emergency) for this entry.
He authorized
the ems doses to 100 mrem / day on the RWP in accordance with BRP 1140,
Radiation Work Permit.
The RCTs assigned to this job signed in on standing
radiation / chemistry prejob survey RWP 50106 instead of RWP 50147 (emergency).
The HP foreman instructed the RCTs to conduct a jobsite survey at
the incore motor drive area, in addition to a routine containment survey
and sample collection.
He told the RCTs to telephone the jobsite dose
rates to him for RWP 50147 completion, even though dose rate information
is not needed to initiate an emergency RWP by licensee procedures.
When
questioned, the HP foreman stated he confused the RWP for emergency entry
requirements with those of a prejob survey.
Specific instructions
regarding the required continual RCT surveillance of the ems were not
discussed with the RCTs, who were unaware that the EM's entry was to be
made under a Type II RWP for emergency entries.
The prejob instructions
were inconsistent with RWP 50147 requirements.
The RCTs departed from the health physics office.
The HP foreman then
briefed the EM foreman and ems.
The HP foreman issued audible, integrating,
alarming dosimeters (digidoses) to the ems and stated the alarm setpoint
was 90 mrem.
He did not inform the ems that the digidoses' increasing
chirp rate corresponded to increasing radiation levels, nor did he
specifically instruct the ems to leave the area if the digidoses alarmed.
Because the digidoses had been used for only a short time at the Byron
Station, their use was not addressed during NGET training according to a
station instructor.
One of the ems, however, had previously used digi-
dose instruments at another Ceco facility.
The ems briefly reviewed
RWP 50147; the HP foreman indicated this was an emergency entry RWP with RCT
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coverage, but did not specifically state that continuous RCT surveillance
was required.
Jobsite radiation dose rates, based on previous surveys.
(5 mR/hr) were briefly discussed.
Neither dose rates nor stay times
were recorded on RWP 50147 at that time.
The EM foreman stated he was
told these dose rates were low, approximately 5 mP/hr; he stated he was
not aware of the detectors' withdrawn positions.
The two RCTs entered containment at approximately 3:09 a.m.
In accordance
with the HP foreman's instructions, routine airborne samples and direct
surveys enroute to the jobsite were taken.
Dose rates increased to 200
mR/hr as the RCTs descended the stairs to the 411' inccre drive area;
a significant increased digidose chirp rate was noted.
Dose rates at the
bottom of the stairs a few feet from the drive units were 5 R/hr; general
area readings in the vicinity of the incore drive units were approximately
5-7 R/hr.
The RCTs exited the area after an estimated 30-second stay
time.
The survey results were not called back to the HP foreman because
of poor phone capability, the high noise level, and wearing of full-face
respirators.
The RCTs collected another sample and exited containment
via an alternate route to avoid the high fields at the incore drive area.
At approximately 3:24 a.m., the EM's entered containment and proceeded
directly to the incore drive motor area.
Although they did not find the
RCTs at the jobsite as anticipated, they proceeded with the job because
the previous survey results discussed with them indicated low dose rates;
they assumed these levels did not warrant RCT coverage.
They apparently
were not alerted to the elevated dose rates because they wore earplugs in
addition to full-facepiece respirators and could not hear the increasing
digidose chirp rate.
The ems removed the incore drive motor cover and
reset the relay.
They remained in the area "briefly" to finish the job
after hearing their digidoses alarm.
It could not be determined at what
time the ems first heard the alarms.
They were in the area of the incore
drives for approximately 3 minutes; the alarms should have sounded after
approximately 1 minute in the area.
When the RCTs exited containment the
security guard and the EM foreman at the personnel hatch informed the
RCTs that the ems had entered containment earlier and were still inside.
The RCT with the lowest dose immediately returned to containment to
retrieve the ems, whom he met a fcw feet from the hatch; both ems'
digidoses were alarming.
They all exited containment.
The ems' pencil dosimeters (0-200 mrem) were offscale.
Initial digidose
readouts for the two ems were 299 and 279 mrem. Whole body film badge
results, received at 4:00 p.m. that day, indicated doses of 340 and
280 mrem, respectively.
The film badges should be reflective of the
actual whole body dose based on their location (chest) and the radiation
surveys of the work location.
Extremity doses should not have been
significantly higher than the whole body doses since the ems were not
working in close proximity to an individual incore detector.
The dose
rates could have been higher, however, had the incore detectors been
exposed at a higher flux rate or for a longer time period, or if the work
had been conducted sooner after retraction of the incore detectors from
the core. Work area dose rates could have been approximately an order of
magnitude higher under less favorable circumstances.
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Several problems apparently led to this unplanned exposure.
The SE, HP foreman, RCTs, and ems were unaware of the withdrawn
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position, of the incore detectors, and the associated high radiation
fields while planning this entry.
This resulted from (1) inadequate
communications between the technical and control room staff involved
in the flux mapping procedure, who were aware of the potential
radiological hazard, and the SE, who stated he was unaware of the
incore detector locations and the radiological hazards, and
(2) inadequate followup by the SE and HP foreman to verify the
incore detectors' positions.
Also the HP foreman and RCTs did not
recognize the inherent radiological hazards associated with the
incore detectors.
A review of the rad % tion-chemistry training
program indicated that no formal training addressing reactor systems
and associated radiological conditions had been provided to radiation
protection personnel.
The incore detectors, other than the stuck detector, should have been
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returned to their shielded storage positions prior to a containment
entry.
They were not, partially due to the communication and evalu-
ation problems noted aNove, but also because of the lack of adequate
precautions in the promedure used for flux mapping.
Flux map
testing was conducted in accordance with B0P-IC-03, Incore Moveable
Detectors - Partial Core Flux Mapping, just before the detector
became stuck.
This procedure did not include a restriction
prohibiting containment entry while the incore detectors were
withdrawn as did a comparable procedure, BOP-IC-01:
Incore Moveable
Detectors - Flux Mapping Procedure.
It is standard practice to
prohibit containment entry with the incore detectors in unshielded
locations.
Use of the RWP system and worker instructions by the HP foreman
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were not adequate.
As noted above, the HP foreman did not possess
adequate knowledge of the potential radiological hazards associated
with the proposed work.
However, appropriate implementation of
radiological controls still could have minimized worker exposures.
The Type II RWP for emergency entry was used to meet an " urgent"
situation, as perceived by the SE.
This RWP use was initially
justified by licensee personnel to prevent a reactor trip, one of
the conditions specified by BRP 1140-1, Radiation Work Permit, that
permits an emergency entry; however, a reactor trip was apparently
not an immediate threat as indicated by maintained steady power
levels throughout the shift.
The lack of familiarity with the
emergency RWP provisions and its first implementation at Byron led
to considerable confusion.
This was reflected by the HP foreman's
inconsistent directions to the ems and RCTs, the RCTs' use of a
standard RWP for prejob surveys and the ems' use of an emergency RWP
requiring continual RCT surveillance, the unfamiliarity of all
involved personnel with procedures surrounding use of the emergency
RWP, and the inadequato instructions given to the ems concerning
digidose use and RCT surveillance.
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The ems did not terminate their work and exit containment immediately
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upon recognition that their digidose monitors were alarming.
No area monitor was located at the incore drive area; therefore local
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radiological conditions could not be predicted before entry.
An area
monitor was located in the 401' level seal table room, which is
shielded from the incore drive area by approximately 36 inches of
concrete.
This monitor readout reflected the stuck incore detector
but not the other five incore detectors at their five path position
near the jobsite.
Although it was determined that the rad / chem
personnel involved in this event were unaware of the monitor's
location, this weakness did not affect the licensee's involvement
in this incident since the monitor readout was not evaluated by the
rad / chem personnel before this entry.
The following apparent violations were identified based on this incident.
10 CFR 50, Appendix B, Criterion V states activities affecting
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quality will be prescribed by documented instructions, procedures,
or drawings appropriate to the circumstances.
The operating
procedure in use did not restrict containment entry by personnel
when the incore detectors were unshielded.
(Violation 454/85022-01)
10 CFR 19.12 requires instructions in radiological conditions and
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precautions be given to individuals entering the restricted area
commensurate with potential radiological health conditions.
The two
EM workers were not adequately instructed regarding:
(1) continual
RCT attendance; (2) digidose use; and (3) current jobsite radiation
levels.
(Violatior 454/85022-02)
TS 6.12.2. states an approved RWP will specify dose rates and stay
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times for individuals entering radiation fields greater than
1,000 mR/hr; in lieu of stay times, but not of dose rates, continuous
RCT surveillance may be used.
RWP 50147 for the May 1, 1985 incore
detector repair job did not specify stay times or dose rates, nor
was continual RCT surveillance provided for two EM workers who
entered fields above 1,000 mR/hr.
(Violation 454/85022-03)
BRP 1140-1, Radiation Work Permit, Section E.1 states emergency
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entries are to be made with an RCT in continual attendance at the
jobsite.
Section C.3 of this procedure states workers signing an
RWP must comply with its requirements.
Continual RCT attendance at
the jobsite was not maintained for two EM workers who made emergency
entries into containment.
Further, these workers, who signed
RWP 50147 for this entry, exceeded the 100 mrem / day limit specified
by this RWP.
(Violation 454/85022-04)
BRP 1000-A1, Work In Controlled Areas - Personnel Conduct, states a
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worker should leave the controlled area as quickly as possible when
the dose equivalent equals the exposure authorized for the job.
The
EM vorkers remained in the controlled area after exceeding their
authorized exposure limits.
(Violation 454/85022-05)
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Several apparent violations were identified.
6.
Personal Contamination Incident
On July 1,1985, a mechanical maintenance crew consisting of a foreman, an
"A" mechanic, and a "B" mechanic, removed contaminated insulation from a
leaking CVCS valve without using protective clothing and without prior
notification of the Rad / Chem Department. As a result, the "B" mechanic's
hands and clothing were contaminated, and the "A" mechanic's shoes were
contaminated.
Based on discussions with the involved personnel, including
the mechanical maintenance foreman, the "B" mechanic, the "A" mechanic,
and radiation protection personnel, the following scenario was developed.
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The mechanical maintenance foreman was assigned to evaluate needed repairs
for a leaking CVCS relief valve located about 20 feet above the floor of
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the 364-foot elevation piping penetration area.
At about 2:00 p.m. the
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foreman entered the area under a general entry radiation work permit (RWP
5-0004) and climbed into the overhead to observe the leaking valve.
He
determined that the insulation around the valve had to be removed before
he could determine the exact source of the leak.
At about 2:30 p.m., the
foreman returned to the area with the two mechanics.
All three workers
climbed into the overhead and removed the insulation from the leaking
valve and the surrounding piping.
The "B" mechanic removed the majority
of the insulation with some assistance from the foreman.
The "A" mechanic
held open a plastic bag for the waste material.
After completing the
insulation removal, all three workers exited the area and proceeded to the
401-foot elevation where they used a personal contamination monitor
(frisker) to survey themselves. When the "B" mechanic moved his hand
within about four inches of the probe, the alarm sounded.
All three
workers then proceeded up the stairs to the 426-foot elevation where they
entered the decon room, and at the direction of the foreman, was' led their
hands.
An RCT passing by the decon room observed the workers washing
their hands and, after summoning assistance, took charge of the
decontamination.
All but very small amounts of contamination were removed
from the "B" mechanic's hands.
No further traces of the contamination
could be detected by July 3, 1985.
Shoe ccntamination on the "A" mechanic
was also removed.
All three workers received whole body counts.
No
internal radioactivity was detected.
The exact magnitude of the skin
contamination is not known since the Rad / Chem Department was not given the
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opportunity to perform an initial survey of the mechanic's hands.
on a contamination survey of the insulation material removed, the licensee
estimated the skin contamination to have been about 125,000 dpm per 100
square centimeters.
Before the initial entry into the 364-foot piping penetration area to
inspect the leaking valve, the foreman claims to have contacted the
rad / chem office by phone and then in person on his way into the plant.
According to the foreman, on both occasions he asked the rad / chem
representative, "any problem in the 364-foot piping penetration area?", to
However, none of the
which the rad / chem representative replied, "no".
eight rad / chem personnel on shift at that time could recall any contact
Even if he had contacted rad / chem, the information
with the foreman.
exchanged was not adequate to assess the radiological hazards associated
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with the foreman's task of climbing into the overhead (unsurveyed area) to
inspect the leaking valve.
The foreman could not identify any of the
rad / chem representatives that he stated he had contacted either by name or
by physical description.
During the interview with the inspectors, the
foreman stated that he was in possession of the work request, which
clearly indicated that the valve was leaking radioactive liquid and was
causing the surrounding area to become contaminated, that he was aware
that the chemical and volume control system (CVCS) contained radioactive
liquid, and that the door leading to the 364-foot piping penetration area
was posted with a sign which read, " Contact Rad / Chem Before Entry." Based
on this, it appears that the foreman should have been aware that
information regarding protective clothing requirements, area dose rates,
and special radiological precautions should Lave been requested from the
rad / chem office before he entered to inspect the leaking valve under the
general entry RWP.
Technical Specificatico 6.11 states that radiation protection procedures
shall be approved, maintained and adhered to.
Procedure BRP 1000-A1,
Radiological Control Standards, states that the job supervisor shall
contact the Rad / Chem Department for protective clothing requirements, dose
rates, and special radiological precautions before entering a controlled
10 CFR 20.201(b) requires an evaluation or survey be conducted of
area.
the rrdiological conditions to determine the extent of the radiation
hazards that may be present.
The foreman's failure to obtain adequate
information from the Rad / Chem Department regarding the radiological
conditions in the vicinity of the leaking CVCS relief valve before
entering the area is a violation of this procedural requirement and 10 CFR 20.201(b).
(Violation 454/85022-06)
After arranging for the two mechanics to assist him, all three proceeded
to the 364-foot piping penetration area.
They 'iid not contact the
rad / chem office regarding their intent to remove the insulation from the
leaking CVCS relief valve, but the "B" mechanic reportedly asked the
foreman while enroute to the work area if they should stop and confer with
a rad / chem representative regarding their task.
The foreman reportedly
answered "no", and did not indicate that he had already contacted
rad / chem.
Upon arrival at the 364-foot piping penetration area. the
mechanics observed that the door to the area was posted with a sign
stating " Contact Rad / Chem Before Entry," and that the leaking valve was
dripping liquid into a funnel normally used for containment of radioactive
A hose connected to the funnel was directing the liquid to a
liquids.
floor drain and was held in place by a sticker which read " Radioactive
The "B" mechanic reportedly again asked the foreman if they
Material."
should contact the rad / chem office or wear protective clothing to remove
The foreman reportedly answered "no" and stated that "it
the insulation.
wasn't that bad", an apparent reference to the amount of contamination
As the job of removing the insulation progressed, the foreman
present.
reportedly stated that "maybe we should have worn gloves, but it's too
late now."
At no time did the foreman stop the job, even though he
apparently knew they were handling radioactively contaminated material
without the proper protective clothing.
Technical Specification 6.11 states that radiation protection procedures
Procedure BRP 1000-A1,
shall be approved, maintained and adhered to.
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Radiological Control Standards, states rules to minimize the spread of
contamination as follows:
(1) observe radiological precautions on all
signs and labels; (2) assume surfaces are contaminated unless otherwise
indicated; (3) utilize work practices to minimize contamination spread;
and (4) consult the Rad / Chem Department before uncovering contaminated
equipment or disassembling potentially contaminated material.
In
addition, this procedure states the job supervisor shall contact the
Rad / Chem Department for protective clothing requirements, dose rates, and
special radiological precautions before entering a controlled area.
10 CFR 20.201(b) requires an evaluation or survey be conducted of the
radiological conditions to determine the extent of the radiation hazards
that may be present.
Tne foreman's failure to observe posted radiological
precautions, observe good work practices, and consult with the Rad / Chem
Department concerning evaluation of the radiological hazards present is a
violation of these procedural requirements and 10 CFR 20.201(b).
(Violation 454/85022-06)
After removing the insulation from around the leaking valve, all three
workers proceeded to the 401-foot elevation of the auxiliary building
where they surveyed themselves for contamination.
The "B" mechanic stated
he thought he was contaminated, even before he began the survey, because
he was sure the insulation was contaminated.
When he moved his hand
within about four inches of the probe, the alarm sounded.
All three, at
the direction of the foreman, proceeded to the decontamination room
located on the 426-foot elevation.
Both mechanics and the foreman stated
that they knew of the posting at the personal contamination monitoring
station which requires immediate notification of the rad / chem office if
they find they are contaminated.
However, the foreman directed the two
mechanics to proceed to the decontaminatica room even though the "B"
mechanic questioned whether rad / chem should be notified.
Upon entering
the decon room, all three workers began washing their hands without a
rad / chem representative in attendance.
As a result, the Rad / Chem
Department was unable to determine the extent of contamination initially
present on the mechanic's hand and had to estimate skin dose (which was
minimal based on surveys of the material handled).
Although the foreman
stated that he left the two mechanics in the decon room and went to the
rad / chem office to summon assistance, and that an RCT returned with him to
the decon room, the mechanics stated that the foreman remained in the
decon room with them from the time they first entered until after a
passing RCT noticed their activities and assumed the decontamination
responsibilities.
None of the rad / chem personnel on shift recalled any
contact with the foreman, nor could the foreman identify the RCT he stated
he contacted by name or by physical description.
Technical Specification 6.11 states that radiation protection procedures
shall be approved, maintained and adhered to.
Procedure BRP 1000-A1,
Radiological Control Standards, requires that workers observe radiological
precautions on all signs and labels.
Procedure BRP 1470-1, Personal
Decontamination, states deconning methods should only be conducted under
the direction of trained individuals, usually rad / chem personnel.
Failure
to adhere to the reporting requirements posted at the personal monitoring
station concerning rad / chem notification, and performing the decon
procedure without a rad / chem representative in attendance is a violation
of these procedural. requirements.
(Violation 454/85022-07)
11
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The licensee initiated followup and investigative actions after this
event.
A dose assessment for the "B" mechanic's skin showed less than 5
mrem, well within the 10 CFR 20 limit of 7.5 rem per quarter.
The three
mechanical maintenance personnel were whole body counted July 1, 1985; the
"B" mechanic had a followup count July 2, 1985.
No internal depositions
were identified.
The Station Superintendent met with the maintenance and rad / chem
departments on July 3, emphasizing radiation workers' responsibilities.
On July 4,1985, the decontamination room door was reportedly posted with
'
a notice to contact rad / chem before initiating personal decontamination.
-
A Radiation Occurrence Report (85026) briefly describing this event was
written immediately after the Rad / Chem Department was notified at
approximately 2:30 p.m. July 1, 1985.
The Station Health Physicist
conducted a group interview with involved personnel the following morning
to obtain an accourt of the event.
Individual interviews of the mechanics
were hampered because of their vacation schedules.
All three mechanical
maintenance personnel had been interviewed by July 9, 1985,
eight days
after the incident.
The station's account did not differ significantly
from the inspectors'.
Ceco corporate health physicists conducted a two
hour onsite review July 3, primarily interviewing the mechanical
maintenance foreman and Station Health Physicist.
Several apparent violations were identified.
7.
April 17, 1985 Containment Entry
A review of licensee Radiation Occurrence Reports (R0Rs) identified
another case of exceeding administrative dose limits during a containment
entry when the reactor was at 20% power.
This occurred April 17, 1985,
when a shift foreman (SF) and an equipment attendant (EA), accompanied by
an RCT, entered containment to locate an RCS leak.
Dose limits for the
entry were approved to 200 mrem under a Type II RWP with continual RCT
attendance for dose monitoring.
The three wore full sets of protective
clothing and respirators; digidoses with alarm setpoints of 150 mrem were
issued.
This crew entered containment and initially inspected low dose
areas outside the missile barrier, accumulating a total dose of < 5 mrem
for each individual.
At the shift foreman's direction, the three crossed
inside the missile barrier, into a significantly higher dose area, to
inspect the "D" reactor coolant pump (RCP).
The RCT and SF ascended a
ladder to inspect a piping area (general fields 2-3 R/hr) near the RCP;
the EA remained at the base of the ladder acting as a safety attendant.
Doses at the ladder base were approximately 60 mrem.
Following a brief
tour of the piping area with the RCT, the SF proceeded approximately ten
feet toward the pressurizer to inspect a valve without the RCT.
He
returned to the RCT and the three exited containment; the RCT's and SF's
,
digidoses were alarming. While the RCT remained at containment access
removing his protective clothing, the SF and EA reentered containment to
retrieve the SF's flashlight which had been mistakenly left behind.
The
SF was concerned that the last flashlight might violate TS 4.52.c, which
prohibits loose debris inside containment during plant operation.
At this
point, licensee reports indicate the SF's dose was approximately 180 mrem,
the EA's approximately 120 mrem, and the RCT's approximately 250 mrem.
12
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The SF and EA, aware they were approaching their limits, initially
,
remained outside the missile barrier; however, they did not find the
flashlight and decided further search inside the missile barrier was
necessary.
Since the EA had a lower accumulated dose, he initiated a
search of the piping area the SF and RCT had toured earlier.
The EA
i
reportedly exited the piping area when he noted his digidose readout was
approaching the 200 mrem limit.
Meanwhile, the SF, concerned that the EA
had not returned, entered the piping area to search for the EA.
The SF
exited the area and met the EA at the missile barrier.
Both exited
containment without having found the flashlight.
Final doses were 260
mrem (SF), 295 mrem (EA), and 254 mrem (RCT). These doses indicate the
l
three remained in the controlled area after exceeding their authorized
dose limits, an apparent violation of BRP 1000-Al which states a worker
I
should leave the controlled area as quickly as possible when reaching his
authorized exposure.
(Violation 454/85022-05)
The involved personnel failed to follow RWP directions.
Although they
were aware of their dose limits and took cursory measures to remain within
these limits, they did not heed the specific requirements of the RWP.
These actions, and resulting consequences, were in violation of BRP 1140,
Radiation Work Permit, which states the requirements of the RWP must be
t
complied with.
Contrary to this, doses for the three individuals exceeded
the 200 mrem limit authorized on the RWP.
In addition, the RCT did not
remain in continual attendance with the SF and EA as specified by the RWP.
Lack of continuous RCT surveillance for workers entering radiation fields
greater than 1000 mR/hr is a violation of TS 6.12.2. (Violation
454/85022-08)
Management followup on this event was not ini.tiated until May 14, 1985,
apparently because the ROR describing the event had been misplaced before
management review.
Further investigations were delayed because of time
conflicts with the May 1,1985 incore area entry (Section 5).
Comprehensive management review was initiated July 13, 1985.
t
l
Violations were identified as noted.
i
8.
Radiation Protection - Startup
!
a.
Startup Surveys
i
Neutron surveys at 75% power levels were reviewed.
Neutron levels
l
throughout containment and at the personnel hatch remain higher than
i
anticipated presumably because of a nozzle shield cover design.1
This situation and applicable dose reduction resolutions were
addressed in an April 30, 1985 memo to the station superintendent
from the ALARA coordinator.
A permanent design modification,
anticipcted to reduce neutron streaming, has been approved.
This
modification has been rescheduled for the November outage, after
equipment unavailability and time constraints precluded installation
during the July shutdown.
If a significant reduction in observed
1
IE Report No. 50-454/85014; 50-455/85009
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neutron levels result from this modification, it would be implemented
at Byron Unit 2 and Braidwood Units 1 and 2.
Interim dose reduction
,
measures at Byron Unit 1 include installation of water shields
outside_the hatch and proposed poly shielding to be hung inside the
'
hatch.
'
'
A contractor study was recently completed of general neutron dose
rates, fields and energies at various power levels.
The final report
has not yet been published.
b.
Facilities
!
. Facility and equipment differences between the two units were
i
evaluated to determine an impact on radiological practices and/or
procedures.
The Radiation-Chemistry Supervisor did not identify any
significant differences.
,
No apparent violations were noted.
'
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9.
Training and Qualifications
-The inspector reviewed the training and qualifications aspects of the
licensee's radiation protection, radwaste, and transportation programs,
including:
changes in responsibilities, policies, goals, programs, and
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methods; qualifications of newly hired or promoted radiation protection
personnel; and provision of appropriate radiation protection, radwaste,
i
and transportation training for station personnel.
Also reviewed were
i
management techniques used to implement these programs and experience
concerning self-identification and correction of program implementation
weaknesses.
,
!
To meet ANSI N18.1-1971 criteria referenced by TS 6.3.1, twenty-seven RCTs
have completed the training / certification program described below.
Those
>
with less than two years' experience work under appropriate rad / chem
i.
supervision.
These RCTs are certified for backshift coverage.
All
chemistry and health physics foremen meet ANSI N18.1-1971 qualifications
for non-licensed supervisors.
The Radiation Chemistry Supervisor (the
designated Radiation Protection Manager) and the Station Health Physicist
'
meet Regulatory Guide 1.8 RPM qualifications in accordance with TS 6.3.1.
!-
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These 27 certified RCTs have completed a training / qualifications program
!
of twelve weeks coursework, four weeks laboratory / work exercises, four
(
~
weeks of OJT, and completion of certification guides in accordance with
BRP 1920, RCT Certification Guides.
Tests scores, and certification
,
sheets were reviewed; no problems were noted.
l
The inspector noted that over half of the RCT certification sheets for
shipment surveys (BRP 1930, T8) had not been completed; waivers had been
issued in accordance with applicable procedures.
The Station Health
,
Physicist is establishing a schedule to complete these waivers.
According
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to the Rad / Chem Supervisor, currently these surveys are completed under a
I
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foreman's supervision.
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Coursework was reviewed.
No specific, formal training on radiological
hazards associated with plant systems was given to RCTs or HP foremen.
This was perceived as a programmatic weakness, contributing to the May 1,
1985 incore drii/e area entry (Section 5).
This was discussed at the exit.
The majority of RCTs do not have extensive operational nuclear plant
experience; their backgrounds are comprised primarily of preoperational
and startup experience.
Approximately 18 RCTs worked at Quad Cities
Nuclear Station during an outage to gain operational experience.2 Other
RCTs completed the above training program with brief tours of other CECO
facilities.
No apparent violations were identified.
10.
Maintaining Occupational Exposures ALARA
The inspector reviewed the licensee's program for maintaining occupational
exposures ALARA, including:
changes in ALARA policy and procedures;
worker awareness and involvement in the ALARA program; establishment of
goals and objectives, and effectiveness in meeting them.
Also reviewed
were management techniques used to implement the program and experience
concerning self-identification and correction of program implementation
weaknesses.
A formal ALARA program is being developed in accordance with BAP 700,
ALARA, under the direction of the ALARA coordinator, a former SR0 at
Braidwood with equipment attendant and health physics experience at the
Byron station.
Exposure goals for each department for 1985 have been
established.
The station ALARA committee, represented by the three
departmental superintendents, meets quarterly; minutes reviewed indicated
dose reduction considerations were addressed at the last two meetings.
A job history file is being assembled to facilitate future job planning
which will track individual job data by RWP, component / equipment number,
job, and workers' names.
Radiological information stored and retrieved
includes airborne and contamination levels, workers' doses, contact doses,
stay times, and man-rem estimates.
This job file will be expanded when
the station gains computer access to the corporate ALARA program around
September 1, 1985.
Currently, the ALARA coordinator relies on job histories maintained by the
Zion station for planning activities.
ALARA reviews of selected RWPs in
accordance with BAP 700-2, ALARA Reviews, were reviewed by the inspector.
A checklist of dose reduction measures is followed; no problems were
noted.
Dose estimates were comparable to actual job
doses.
A prejob briefing was held for the ALARA reviews.
Rad / cham
representatives attend the morning meeting as a job planning and awareness
measure.
Photographs of equipment have been posted outside higher dose rooms
throughout the auxiliary building to reduce workers' stay times in these
areas.
A complete set is kept by the rad / chem department.
2
IE Report No. 50-454/83008; 50-455/83006
15
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To promote worker awareness of ALARA, an ALARA suggestion box has been
established with awards given to notable dose saving measures.
Posters
promoting ALARA are planned.
The contamiration control program was reviewed.
Currently, the ALARA
coordinatcc has defined 4400 square feet of the auxiliary building as
2
contaminated (> 1000 dpm/100 cm ) or less than 1% of the total area.
The
plant goal is to maintain the contaminated portion of the auxiliary
building at less than SL
All of containment is posted as a contaminated
area.
Daily surveys are reviewed; contaminated areas are targeted for
cleanup by four designated deconners.
Decontaminated areas are tracked;
their location and source of contamination noted.
No apparent violations were identified.
11.
Exposure Controls
a.
External Controls
The inspector reviewed the licensee's external exposure control and
personal dosimetry programs, including:
changes in facilities,
equipment, personnel, and procedures; adequacy of the dosimetry
program to meet routine and emergency needs; planning and preparation
for maintenance and refueling tasks including ALARA considerations;
required records, reports, and notifications; effectiveness of
management techniques used to implement these programs; and
experience concerning self-identification and cerrection of program
implementation weaknesses.
The licensee conducts the external exposure control program in
accordance with BRP 1210-1, Personnel Monitoring for External
Exposures.
Film badges are clipped to security badges and issued to
each individual to provide positive controls over badge distribution.
The inspector reviewed exposure results for the second quarter 1985;
no exposures exceeding NRC limits were identified.
Five individuals
exceeded administrative limits as described in Sections 5 and 7.
review daily updates to identify individuals approaching their limit
and contact the appropriate work supervisor and access control
personnel when limits are approached.
These updates will be
maintained at access control.
Neutron dosimetry is issued to individuals entering containment and
working around the personnel hatch where elevated neutron readings
have been identified.
Timekeeping is also used for neutron dose
assessment; the higher of the two exposures is recorded as the
official dose.
According to a health physicist, exposures for lost and/or missing
badges are assessed in accordance with BRP 1200-T5, Radiation
Investigation Sheet.
New badges are assigned for the remainder of
the two week badge period.
This process was followed by rad / chem
personnel when an inspector's badge was inadvertently lost; no
problems were noted.
16
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b.
Internal Controls
The inspector reviewed the licensee's internal exposure control and
assessment programs including:
facilities; equipment; personnel;
procedures affecting internal exposure control; and assessment of
individual intakes.
A standup whole body counter is used to determine internal
deposition.
No depositions approaching the licensee's investigation
level (>2% MPBB) have been identified, nor has the 40-MPC hour
evaluation threshold been approached.
The licensee has developed a
conversion from internal activity deposited to MPC-hours, to verify
compliance with 10 CFR 20.103.
However, this conversion has not been
established in the program, either by procedure or in the computer
software used for whole body counting.
Licensee representatives
agreed to formally incorporate this conversion in the program (0 pen
Item 50-454/85022-09; 50-455/85020-01)
No apparent violations were identified.
12.
Licensee Event Report (LER) Followup
The inspector reviewed LER 85045 which identified a reactor coolant sample
collected and analyzed on April 10, 1985, in excess of the six-hour period
allotted after a reactor trip by Technical Specifications (TS) 3/4.4.8,
Table 4.4-4.
The surveillance form, BAP 1400-T5, initiating the sample
request was completed about 90 minutes after the trip; however, this
sample was not collected imnediately because of increased sampling
activity associated with the boric acid evaporator rupture disk event
discussed in Section 16.
The sample was collected and analyzed seven
hours and 45 minutes after the trip.
No problems were identified with the
analytical results.
Surveillance sheets for TS required samples are now placed in color coded
'
'
folders readily identified as a priority saniple.
The HP foremen have been
- -
instructed in TS required samples and appropriate followup.
A status
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board in the access control office lists TS required nonroutine samples.
!
Routine sample collection is assigned from a daily printout by the
l
chemistry supervisor.
Since this system was initiated, fifteen samples
~
required by this technical specification were collected and analyzed
within this specified time frame.
The licensee's corrective actions were
appropriate and timely.
,
No violations were identified by the inspectors.
13.
Allegation (RIII-85-A-0100)
!
The inspector reviewed an allegation submitted by a contractor that
l
individuals had exited the security building through an alarming portal
monitor on April 22, 1985 around 6:20 p.m., a common quitting time.
The
'
alleger stated he had contacted licensee representatives regarding this
matter, but that he subsequently noted various occasions during high
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traffic flow periods when no security officer was posted at the portals to
monitor personnel egress.
The station security plan does not require this designated post (post 8)
to be staffed; however, posted personnel can most effectively respond to
portal alarms, since the officer on the. opposite side of the turnstiles is
preoccupied collecting badges and security personnel within the adjacent
glass enclosed office may not be capable of responding timely.
Instructions concerning security guards' specific actions to portal
monitor alarms are outlined in distributed post orders and addressed
during basic guard training.
Discussions with security personnel
indicated they were aware of these actions.
The inspector observed portal monitor surveillance on June 6, 1985, at the
times specified by the alleger.
At 4:20 p.m., guards were posted at the
portals; no alarms were actuated.
At 5:26 p.m., a guard was not posted to
observe correct portal monitor exit procedures; the inspector observed a
portal monitor alarm actuation, but no attempt was made by security
personnel to either intercept the individual (to require another attempt
to pass through the portal monitor) or to restore the monitor before other
personnel exited.
A guard from the glass enclosed office reset the alarm
after several workers exited through the portal.
This is an apparent
violation of BRP 1460-3, Operation of the IRT Portal Monitors,
Section E.3, which states security (personnel) will stop any personnel who
alarm the portals from leaving the plant and notify the
Radiation / Chemistry Department for further action (Violation 454/85022-10;
455/85020-02).
This allegation was substantiated.
One apparent violation was identified.
14.
Unit 2 Condensate Pit Overflow
On May 24, 1985, diluted reactor coolant (RCS) water was inadvertently
transferred to the Unit 2 condensate pit sump and the construction runoff
pond via an unproceduralized valve lineup (Radiation Occurrence Report
85017).
This lineup was initiated to transfer recycle holdup tank (HUT)
water to the chemical drain tank (CDT).
This is an alternate flowpath;
the HUT inventory is normally treated by the boric acid evaporators which
were out of service at the time.
Operating engineers, in an effort to
reduce the HUT inventory, transferred this water to the CDT to near
,
l
capacity using P&ID lineups.
A permanent procedure did not address this
!
lineup nor was a temporary change initiated.
This is an apparent
l
violation of 10 CFR 50, Appendix B, Criterion V which states activities
!
affecting quality will be prescribed by documented instructions,
procedures, or drawings appropriate to the circumstances (Violation
454/85022-11; 455/85020-02).
Three valves (0AB 8630, 0AB 8591, and 0AB 0003) to the CDT remained open
following the transfer; they were not restored to their normal closed M-65
,
configuration.
Following the transfer, Unit 1 operators initiated a
dilution of RCS boron concentrations with letdown from the Volume Control
!
Tank (VCT) to the HUTS.
The above referenced valve lineup created an
alternate flowpath by gravity flow to the CDT.
The CDT reached capacity
and overflowed.
Overflow was directed through the CDT filtered vent
18
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system, equipped with loop seal drains to remove water accumulations; the
coolant was then collected by the Unit 2 auxiliary building equipment
drain sump, which overflowed into the floor drain system, and ultimately
to the condensate sump which flows to the construction runoff pond (CRO).
An estimated 6,000 gallons of water contaminated sections of the turbine
building and Unit 2 auxiliary building.
No personal contaminations were
reported during this occurrence.
Survey results reviewed indicated no
measurable activity was detected from the CR0 outlet.
Extensive
decontamination effort.; of approximacely 600 square feet of plant area,
including temporary office space, were initiated.
The floor area was
deconned by water washings and acid etchings to 200 dpm/100 cm2 fixed
contamination; sealers were applied to the floor and surrounding walls.
Drains were flushed, but fixed contamination remains.
The inspector took
direct reading surveys of the sump.
No levels significantly above
background were identified.
The sump remains posted and barricaded.
Condensate sump samples indicated a total isotopic concentration
significantly less than one MPC fraction (unrestricted).
No measurable
activity was identified in samples taken of the flume and CR0 outlet
following the overflow.
The CR0 inlet sample indicated a total activity
less than one MPC fraction (unrestricted).
Followup sampling was
conducted during the inspection.
No isotopic migration was identified.
No regulatory or technical specifications limits were exceeded.
On May 26, 1985, the licensee initiated a temporary procedure and a 10 CFR 50.59 review to address this lineup.
The licensee is completing further
actions in response to this occurrence.
This will be reviewed in a future
inspection.
One apparent violation was identified.
15.
VCT Releases
Two unrelated volume control tank (VCT) releases occurred, one on May 28,
1985 and the other June 4, 1985.
In both cases, no technical specifica-
tions limits were approached or exceeded.
Noble gases and associated
short lived daughter products were involved.
The auxiliary building
ventilation system was out of service at the time of the releases, thereby
impeding the dilution and clearance of the area.
A license condition
required the auxiliary building ventilation system to be fully operational
by July 1,1985.
Until that date, system testing was being conducted.
Neither release was reportable by regulatory requirements.
On May 28, 1985, Unit 1 operators were routinely venting the VCT; a
diaphragm valve in the release path developed a leak around 9:50 a.m.
and
by approximately 10:00 a.m. the VCT was secured.
Elevated background
levels were noted on an area monitor approximately 60-70 feet from the VCT
cubicle.
Rad / chem staff cleared the area immediately adjacent to the VCT
cubicle and valve area.
Samples of the VCT vicinity were collected over a
four-hour period.
Initial maximum values indicated the 40 MPC-hour value
was not exceeded; however, access to the auxiliary building was restricted
for approximately four hours as a precautionary measure.
The clothes of
seven people were contaminated with noble gas daughter isotopes; they were
detained by rad / chem personnel until this short lived activity decayed.
A
19
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1
maximum of 0.477 Ci of noble gas products were released, less than 0.1% of
the technical specifications limits.
A slight increase of activity was
noted on the vent stack monitor; no alarms were actuated.
,
i
At approximately 1:05 p.m. on June 4, 1935, a release from the VCT
occurred via a pressure reculating valve on a sample line.
This line had
been opened for a routine sample collection.
This is a separate line from
that involved during the May 28, 1985 release; no generic deficiency
between the two releases or associated valves was identified.
Elevated
readings were noted on laundry room friskers, approximately 100 feet from
the VCT valve gallery.
At 1:10 p.m., rad / chem personnel ordered a
precautionary evacuation of the 426' elevation and began efforts to
initiate the auxiliary building exhaust system.
The valve was isolated at
1:45 p.m., and by 2:30 p.m., the entire auxiliary building was evacuated.
Four maintenance workers were in the VCT valve gallery aisle at the time,
repairing the valve attributed to the May 28, 1985 release.
They finished
i
this job and exited this area at 1:10 p.m.
Whole body counts indicated no
uptakes occurred; 10 MPC hours were assigned to these individuals and
'
three RCTs who sampled the area.
Whole body exposures were approximately
10 mrem by pencil dosimeter for the maintenance workers and RCTs.
The
i
clothing of five workers near the valve gallery, in addition to the four
maintenance workers and RCTs, were contaminated with noble gas daughter
isotopes.
The twelve were detained and their clothing aerated until this
short lived activity decayed.
Evacuating personnel were surveyed by RCTs
before release.
At 2:50 p.m. the ventilation system was activated,
clearing the auxiliary building; reentry was authorized at 3:45 p.m.
A
total of 0.684 Ci were released, approximately 0.1% of technical
specifications limits.
Both valves were repaired and testing completed by June 5, 1985.
No
.
problems were noted.
!
.
No apparent violations were identified.
16.
Boric Acid Evaporator Release
,
[
On April 10, 1985, a release occurred during testing of the 0A boric acid
'
evaporator (BAE) vent lines. A temporary alteration (M85-0-093) was
1
installed to bypass a suspected leaking check valve (0AB037); an open
manual valve on the bypass line maintained the normal vent path from the
OA BAE to the gaseous waste vent header.
This temporary alteration was
completed in accordance with BAP 300-T5, Temporary Alteration, and was
!
reviewed by the licensee per 10 CFR 50.59.
The testing required the BAE
to be in a recycle mode, venting to the gaseous waste header.
Concurrently, the Unit 1 operator vented the VCT to the gaseous header for
'
approximately two minutes, unaware of the BAE venting.
A pressure spike
resulted in the header, overpressurizing the bypass line and BAE, causing
the BAE rupture disk to fail.
Approximately 80 gallons of BAE inventory
flashed to steam, partially condensed to liquid, and followed the inplace
flowpath to the OB evaporator room.
Since the auxiliary building
ventilation system was not operational at the time, steam was vented
through the discharge piping to the auxiliary building floor drain.
Condensing steam appeared at a 346' level drain near the elevator.
Process area monitors alarmed, indicating a potential airborr,e problem,
20
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primarily noble gases and decay products.
The auxiliary building was
evacuated and secured; contaminated areas at the OB BAE room were
barricaded.
Five individuals were contaminated.
They were decontaminated
by routine washings to normal background levels; no internal deposition
was identified by whole body counts. The supply fans were restored
approximately one hour after the event.
No detectable contamination was
measured at the affected area approximately 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the leak.
Access to the auxiliary building was restored around 4:00 p.m. when
samples and air monitors indicted normal levels had been established for
approximately one hour.
No radiological technical specifications limits
were approached or exceeded.
This matter will be reviewed further to
determine the adequacy of the licensee's alteration review and
communication with operations personnel. (0 pen Item 454/85022-12)
No apparent violations were identified; one open item was established for
future review.
17.
Enforcement Conferences
An enforcement conference was held June 27, 1985, to discuss the entry
into the incore detector motor drive area, Region III staff's concerns
about the problems contributing to the entry, and the associated
violations.
Enforcement options under consideration were addressed.
The
meeting, held at the Region III office, was attended by Mr. J. G. Keppler,
Regional Administrator, NRC Region III, Mr. C. Reed, Vice President,
Nuclear Operations, Commonwealth Edison Company, and members of their
respective staffs.
Region III personnel indicated their concern that performance by the
involved licensee control room operators, technical staff, shift engineer,
HP foreman, and electrical maintenance personnel ranged from only
marginally acceptable to largely unacceptable in this event.
The extent
of personnel performance weaknesses and the licensee's weak initial
management response to this event were emphasized.
The need for improved
attention to radiation protection practices by all workers, including
supervisors, was stressed.
Licensee representatives acknowledged that they were concerned with the
problems which led to this entry and stated their intent to strengthen the
Byron radiation protection program.
Specific corrective actions were
discussed including specialized systems training for radiation / chemistry
personnel, administrative measures to verify incore detectors' positions
before entry, procedural revisions for RWP use, and installation of a
locked gate at the incore drive area.
These actions will be reviewed
during future inspections.
In response to the NRC's concern that
investigation and followup actions were delayed, licensee representatives
acknowledged that corporate and station management did not initially
recognize the appropriate significance of this occurrence.
Licensee upper
management representatives emphasized that followup investigations and
corrective actions of future radiological occurrences would be more timely
and thorough at both station and corporate levels.
An enforcement conference was held July 22, 1985 in the Regional Office to
discuss the July 1, 1985 contamination occurrence and the April 17, 1985
21
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containment entry.
Enforcement options were discussed.
The meeting was
attended by Mr. J. G. Keppler, Mr. C. Reed, and members of their
respective staffs.
Licensee representatives described the circumstances
that led to these events based on their followup investigations and
interviews.
Corrective actions for these events were addressed, including
departmental staff meetings to emphasize procedural adherence and worker
responsibility, and disciplinary actions.
The licensee concluded the
mechanical maintenance foreman's actions were not willful but were
reflective of unsatisfactory supervisory performance in the July 1 event.
The foreman was removed from his supervisory position, according to
licensee personnel, in addition to other disciplinary action.
Region III's evaluation of this incident did not totally concur with the
licensee's conclusions, particularly concerning the licensee's conclusion
that training was a significant contributor to the event; the inspectors'
interviews with personnel indicated they were knowledgeable of the correct
radiation protection practices and procedures.
Licensee investigations
identified several root causes of these events, including staff
motivational problems and inadequate senior management commitment te the
radiation protection program.
The station superintendent reiterated his
increased support of this program and stated that he has recently held
meetings with various departments to promote radiation safety.
Region III personnel indicated that the three events (April 17, May 1, and
July 1) were of particular significance because of the failings of
supervisory personnel to follow established radiation protection
procedures and practices.
These supervisory personnel not only were
responsible to varying degrees for the radiation safety of other
personnel, but also in their supervisory roles they are influential in
shaping the behavior of other personnel.
Additionally, the frequency and
extent of these supervisory fallings, and management's' failure initially
to take timely actions to understand and correct them, imply significant
management weaknesses in the support of their radiation protection
program.
The inspector discussed the likely informational content of the inspection
report, with regard to documents or processes reviewed, with licensee
representatives on June 14, 1985.
The licensee did not identify any such
documents or processes as proprietary.
22
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