ML20207L031
| ML20207L031 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/24/1986 |
| From: | Hosey C, Revsin B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207L025 | List: |
| References | |
| 50-413-86-20, 50-414-86-23, NUDOCS 8607300173 | |
| Download: ML20207L031 (9) | |
See also: IR 05000413/1986020
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET.N.W.
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Report Nos.: 50-413/86-20 and 50-414/86-23
Licensee: Duk'e Power Company
422 South Church Street
Charlotte, NC 28242
Docket Nos.:
50-413 and 50-414
License Nos.: NPF-35 and NPF-48
Facility Name: Catawba 1 and 2
Inspection Conducted: Ma 13-16 and June 2, 1986
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Inspector:
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Q B. K. Revsifi V
Dats Signed
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Accompanying Personnel:
C. H. Bassett
Approved by:
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C. M. Hosey7 Section Chief
Date' Signed
Division of Radiatiqn Safety and Safeguards
SUMMARY
Scope: This routine, unannounced inspection involved onsite inspection in the
area of radiation protection including:
startup shielding surveys; internal
exposure control and assessment; external exposure control and personal
dosimetry; radioactive materials control, posting and labeling; solid radioactive
waste classification and characterization;
transportation of radioactive
materials; and program for maintaining exposure as low as reasonably achievable
(ALARA).
Results:
One violation - Failure to maintain radiation exposure records in
accordance with instructions contained on Form NRC-5.
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8607300173 860715
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REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- J. W. Cox, Technical Services Superintendent
- W.
P. Deal, Station Health Physicist
- C. L. Hartzell, Compliance Engineer
- P. G. LeRoy, Licensing Engineer
- R. E. Sorber, Health Physicist, General Office
L. D. Schlise, Health Physics Supervisor
H. F. McInvale, ALARA Supervisor
G. A. VanderVelde, Health Physics Supervisor
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M. C. Couch, Administrative Health Physics Supervisor
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G. T. Mode, Health Physics Coordinator
- T. Keane, Health Physics, General Office
- W. M. Carter, Radiological Health Supervisor, Applied Science Center
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- P. S. Wingo, System Environmentalist, Applied Science Center
NRC Resident Inspectors
- P. K. VanDoorn, Senior Resident Inspector
- M. Lesser, Resident Inspector
P. Skinner, Resident Inspector
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- Attended first exit interview.
- Attended second exit interview.
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2.
Exit Interview
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The inspection scope and findings of the May 13-16, 1986 inspection were
summarized on May 16, 1986, in the first exit interview with those persons
indicated in Paragraph 1 above.
An unresolved item * (Paragraph 11)
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concerning the density thickness at which the licensee was assessing worker
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whole body exposure when the lenses of the eyes were not shielded by at
least 700 milligrams per square centimeter (mg/cm2) density thickness was
discussed in detail. The licensee acknowledged the inspection findings and
took no exceptions. The licensee did not identify as proprietary any of the
materials provided to or reviewed by the inspector during this inspection.
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At a second exit interview conducted on June 2, 1986, and attended by those
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persons indicated in Paragraph 1 above, licensee management was informed
that the unresolved item would Se considered closed. The licensee was also
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informed that failure to maintain radiation exposure records in accordance
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- An Unresolved Item is a matter about which more information is required to
determine whether it is acceptable or may involve a violation or deviation.
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with instructions contained on Form NRC-5 would be considered an apparent
violation of 10 CFR 20.401 (Paragraph 11).
Licensee management took
exception to this finding based on their opinion that measurement of whole
body radiation exposure through a tissue equivalent density thickness of
1000 mg/cm , rather than the required 300 mg/cm2, would not result in a
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significantly different whole body dose.
3.
Followup on Inspector Identified Problems (92701)
(Closed) Inspector Followup Item (50-413/84-41-01) Snubber Inspection in
High Radiation areas. The licensee has snubbers located in High Radiation
Areas which require periodic inspection as required by Technical Specification 3.4.7.
The licensee applied for an exemption from this
inspection requirement for certain snubbers but were denied on the basis of
ALARA practices being insufficient justification for deleting inspections.
Consequently, the licensee reanalyzed 11 snubbers as rigid supports and
issued Nuclear Station Modification Request No.10736 to implement the
modification.
Completion is expected by February 1,1987.
4.
Control of Radioactive Materials and Contamination, Surveys and Monitoring
(83526)
The licensee was required by 10 CFR 20.201(b) and 20.401 to perform surveys
and to maintain records of such surveys necessary to show compliance with
regulatory limits.
Survey methods and instrumentation were outlined in the
Final
Safety Analysis Report (FSAR), Chapter 12, while Technical Specification (TS) 6.11 provided the requirement for adherence to written
procedures.
Radiological control procedures further delineated survey
methods and frequencies.
During plant tours, the inspector examined radiation level and contamination
survey results outside selected rooms and cubicles. The inspector performed
independent radiation level surveys of selected areas and compared them with
licensee survey results.
The inspector noted that locked high radiation
areas outside containment were maintained as required by TS 6.12.
The
inspector noted that there were numerous contamination recepticles in use
about the Auxiliary Building and was informed by the licensee that they had
implemented a very vigorous program to minimize contaminated areas in the
plant, and consequently, whenever possible leakage was identified, it was
captured or run into a drain.
During tours of the plant, the inspector observed the use of survey
instruments by plant staff and compared plant survey instrument results with
readings taken by the inspector using NRC equipment. The inspector examined
calibration stickers on radiation protection instruments in use throughout
the facility.
No violations or deviations were identified.
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5.
Transportation (86721)
10 CFR 71.5 requires that licensees who transport licensed material outside
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the confines of their plants or other places of use, or who deliver licensed
material to a carrier for transport, comply with the applicable requirements
of the regulations appropriate to the mode of transport of the Department of
Transportation (DOT) in 49 CFR Parts 170 through 189.
The licensee stated that at the time of the inspection, only two shipments
of radioactive waste had been made. They were Shipment No. CNS 85-08 made
on October 17, 1985 and Shipment No. CNS 86-10, made on February 27, 1986.
Both shipments contained 14 B-25 boxes of compacted dry active waste (DAW).
The inspector verified that the requirements for 49 CFR Parts 170 through
189 had been met for these shipments.
No violations or deviations were ' identified.
6.
Solid Radioactive Waste (84722)
10 CFR 20.311 requires that the licensee maintain a tracking system for
radioactive waste shipments to verify that shipments had been received
without undue delay by the intended recipient. The inspector reviewed the
tracking metnodology used by the licensee and examined the documented
receipt acknowledgements for Shipment Nos. CNS 85-08 and CNS 86-10.
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10 CFR 20,311 requires a licensee who transfers radioactive waste to a land
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disposal facility to prepare all waste so that the waste is classified
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according to 10 CFR 61.55.
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10 CFR 61.55(a)(8) states that the concentration of a radionuclide may be
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determined by indirect methods such as the use of scaling factors which
relate the inferred concentration of one radionuclide to another that is
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measured if there is reasonable assurance that the indirect methods can be
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correlated with actual measurements.
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The licensee had been using scaling factors developed for waste streams at
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its sister station, McGuire, since Catawba had not been operational long
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enough to develop plant specific scaling factors for its waste streams.
licensee representatives stated that plant specific waste streams had been
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sampled in February 1986, and that vendor analyses of the samples had been
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completed.
It was anticipated that the plant specific scaling factors would
be available for use in January 1987. The inspector questioned the licensee
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about the lag time between sample analyses and implementation of new scaling
factors.
The licensee stated that scaling factors were computed by the
General Office (GO) and since they also generated the scaling factors for
the other two plant sites, the GO preferred to put them all on a similar
schedule. The inspector stated that at the time of the inspection this area
was not an issue for the licensee since all shipments thus far and those
anticipated for the near future were DAW.
However, with the upcoming
refueling outage scheduled for late summer, 1986, this situation could
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readily change. The inspector reviewed Procedure HP/0/8/1006/05, Periodic
Sampling of Radioactive Waste for 10 CFR 61 Scaling Factor Determination.
The inspector discussed with licensee representatives the program for waste
characterization. The licensee did not anticipate that waste solidification
would be performed frequently, and at the time of the inspection, some resins
were anticipated to be possible candidates for solidification.
All
waste processing equipment would be vendor operated.
In general, when
waste stability was required, the licensee planned to use approved
containers.
No violations or deviations were identified.
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7.
As low As is Reasonably Achievable Program (ALARA) (83728)
10 CFR 20.1(c) specifies that licensees should implement programs to keep
worker's doses ALARA. FSAR, Chapter 12, also contains licensee commitments
regarding ALARA actions.
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The licensee had dedicated six . full time individuals for their program for
maintaining doses ALARA. All work with estimated exposure greater than one
man-rem required ALARA preplanning.
Through the preplanning review,
shielding recommendations were made, job surveys were delineated and overall
evaluation of the work to be performed was made from a radiological point of
view.
Post-job ALARA reviews were required for jobs whose dose estimates
exceeded 50 man-rem and for those jobs where dose estimates were exceeded by
25 percent.
The inspector reviewed the minutes of -the ALARA committee meetings for
October 16, 1985, and January 2,
1986.
The licensee stated that the
committee met on a quarterly. basis and was composed of representatives from
each plant work group.
These members, while not from the ranks of upper
management, had the authority to make commitments for their group.
Plant management commitment to ALARA appeared to be strong. ALARA reports
are sent to plant management monthly and include such data as man-rem by
work group, station dose distribution by work group and a graphic
representation of estimated versus actual exposure received for the year.
Additionally, station section managers have had ALARA goals incorporated
into their performance appraisal goals.
The ALARA group had begun preparation for the refueling outage planned for
August 1986.
The inspector reviewed work packages, WR No. 6335 SWR, Reactor
Coolant Pump Seal Work, Unit 1, and WR No. 5311 SWR, Unit 1, Conoseal Work,
that were under preparation.
Scheduling and planning for the outage were ongoing and as more information
of work to be performed became available, the ALARA group would continue to
develop work packages.
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The ALARA Section had also developed a library of slides and video tapes of
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various rooms showing valve locations and various job evolutions to orient
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and familiarize workers with the jobs to be performed. While not required
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meetings.
No violations or deviations were identified.
8.
Radiation Protection-Startup (83521)
The inspector reviewed Procedure TP/2/B/2200/01, Biological Shield Survey.
This procedure required facility surveys for gamma and neutr'on radiation at
the following power levels: 0%, 30%, 50%, 75% and 100%. Background surveys
prior to fuel loading were also required. The licensee had completed the
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required surveys for background and 0% power levels. All values were found
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to be below the Lower Limit of Detection for the instrument used for
measurement except for the area of the Auxiliary Building on the 543 foot
level in Mechanical Penetration Room No. 227. Gamma radiation levels were
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found to be 0.1 mR/ hour due to piping from the floor drain tank from Unit 1.
The licensee stated that these surveys would be completed as power ascension
continued.
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No violations or deviations were identified.
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9.
Internal Exposure Control and Assessment (83725)
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The licensee is required by 10 CFR 20.103, 20.201(b), 20.401, 20.403 and
20.405 to control uptakes of radioactive material, as ess such uptakes, and
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keep records of and make reports of such uptakes.
FSAR, Chapter 12, also
includes commitments regarding internal exposure control and assessment.
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Due to the limited operating history of the facility and the active
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containment program in operation, the licensee has had few occasions for
use of respiratory protective equipment. The inspector reviewed the daily
exposure computer printout (REC) and found the highest Maximum Permissible
Concentration (MPC) hours assigned to an individual was 1.64. The inspector
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reviewed the methodology the licensee had in place for MPC hour tracking and
determined that until an individual accumulated 0.1 MPC hours, the computer
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program is set up so that the hours do not appear on the REC. However, a
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printout of MPC hours by individual Health Physics (HP) number showed that
all these data were present in the individual's dose file. The inspector
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reviewed procedure HP/0/B/1000/04, Preparation of Health Physics Radiation
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Work Permits and Standing Radiation Work Permits, May 5, 1986 which provided
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the guidelines for issuance of respiratory protective equipment.
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No violations or deviations were identified.
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10. Staffing (83722)
HP staffing levels were discussed with the Station Health Physicist (SHP).
At the time of the inspection,104 staff positions were authorized,103 of
which were filled. Ir4 addition, 19 HP contractor technicians were onsite to
augment the experience level of the staff. At present, 27 HP technicians
were ANSI qualified. ' Shift coverage is provided by three to five technicians
along with one ANSI qualified technician.
The primary responsibility of
the HP shift crew was liquid and gaseous releases.
Duke Power was in the process of reducing the size of their construction
group and as a consequence, 18 of the 104 staff persons at Catawba had come
from the construction crews. It was anticipated that 10 more would be added
to the HP staff in the future. At the time of the inspection, the training
for these individuals consisted of Independent Radiation Worker (IRW)
training.
IRW training consisted of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of classroom instruction
followed by three weeks of on-the-job training during which time trainees
were task qualified.
It was anticipated that as slots opened at the
Training Center, these individuals would rotate through the regular HP
technician training program. The SHP stated that for the upcoming outage
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for Unit 1, these individuals would be assigned to decontamination work.
Also, 60 contract HP technicians will be hired for the outage.
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No violations or deviations were identified.
11. External Occupation'al Exposure Control and Personal Dosimetry (83724)
a.
10 CFR 20.202 requires that appropriate personnel monitoring devices be
worn by personnel likely to receive exposures in excess of 25 percent
of the limits specified by 10 CFR 20.201 or who enter a high radiation
area, and to require the use of such devices.
During tours of the
facility the inspector observed personnel wearing monitoring devices as
required.
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b.
10 CFR 20.401 requires that each licensee maintain records showing the
radiation exposures of all individuals for whom personnel monitoring is
required and that such records be kept on Form NRC-5 in accordance with
the instructions contained in that form.
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Form NRC-5, Item 5 requires that unless the lenses of the eyes are
protected with eye shields having a tissue equivalent density thickness
of at least 700 mg/cm , dose recorded as whole body dose should include
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the dose delivered through a tissue equivalent absorber having a
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density thickness of 300 mg/cm ,
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Licensee representatives stated that thermoluminescent dosimeters
(TLDs) were provided to the plant by the dosimetry laboratory of the
Applied Science Center of the Training and Technology Center located
near Huntsville, NC.
All dosimetry processing was performed at the
center which maintained its own staff with additional technical
assistance provided by the G0 health physics group.
The licensee
informed the inspector that Teledyne Isotopes TLDs and processing
equipment were in use by the Center and that these TLDs utilized
calcium sulfate as their phosphor.
The inspector visited the dosimetry laboratory and was informed that
the TLD card utilized in the badges had eight possible readout areas,
four main read out areas and four backup readout areas. Filtration of
the TLD card was provided by the badge case such that the effective
density thicknesses at which radiation levels were measured were:
Area 1 - 7 mg/cm2
Area 2 - approximately 1000 mg/cm2
Area 3 - approximately 1600 mg/cm2
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Area 4 - approximately 1600 mg/cm
The licensee stated that the ratio of Area 2 to Area 4 was used to
provide information concerning the hardness of the gamma radiation
field and that this ratio was utilized through various algorithms to
treat the data from the four areas. Skin dose was assigned from Area 1
after manipulation by the algorithm and whole body or deep dose was
assigned - from Area 4 which, after treatment with the algorithm,
provided a whole body dose equivalent to that delivered through a
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density thickness of 1000 mg/cm ,
The inspector questioned the licensee about density thickness of eye
protection worn by workers at the plant.
The licensee stated that
safety glasses were required to be worn at all times while in the plant
which included work in containment. The licensee had determined that
both safety glasses and goggles had a density thickness of
approximately 200 mg/cm2 and that respirator face plates had a density
thickness of 260-270 mg/cm2
The inspector asked the licensee how
whole body dose through a tissue equivalent absorber of 300 mg/cm2 was
assessed since the eye protection provided the workers did not meet the
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2 as specified by Form NRC-5. The licensee responded that
700 mg/cm
whole body dose was assessed at 1000 mg/cm2 and that no assessments
were performed at 300 mg/cm . The licensee also stated that each of
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the Duke Power facilities had an ongoing beta radiation program
designed to characterize the beta spectrum at each unit of each
facility and that this data permitted them to conclude that whole body
dose measured at 1000 mg/cm2 was not significantly different from whole
body dose measured at 300 mg/cm2,
The inspector reviewed the results of the half value layer (HVL)
studies performed at the various facilities and found that at 300
mg/cm2 attenuation, all but approximately one to twelve percent of the
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beta radiation dose was attenuated. Consequently, for the worst case,
the maximum that whole body dose could have been underestimated was 12
percent.
The licensee reviewed their dosimetry records since the
installation of the Teledyne system in 1976 and found that using this
conservative percentage of 12, no individual was likely to have
received an overexposure to the whole body. The inspector informed the
licensee that failure to assess whole body dose through a tissue
equivalent absorber of 300 mg/cm2 when the lenses of the eyes were not
shielded by at least 700 mg/cm2 would be considered an apparent
violation of 10 CFR 20.401 (50-413/86-20-01).
12.
IE Information Notices (92717)
The following IE Information Notices were reviewed to ensure receipt and
review by appropriate licensee management.
85-46
Clarification of Several Aspects of Removable Radioactive
Surface Contamination Limits for Transport Packages
86-22
Underresponse of Radiation Survey Instrument to High
Radiation Fields
86-23
Excessive Skin Exposures Due to Contamination with Hot
Particles
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Plant Statistics
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a.
Collective Dose
For 1985, the collective dose for the facility as measured by TLD was
61 man-rem. A collective dose of 97 man-rem had been projected for the
year.
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For 1986, a collective dose of 540 man-rem was projected, and as of
February 28, 1986, 19 man-rem had been received.
The increase in
expected man-rem for 1986 is due to the refueling outage scheduled
later in the year for Unit 1.
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