ML20149F548
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| ML20149F548 | |
| Person / Time | |
|---|---|
| Site: | Neely Research Reactor |
| Issue date: | 02/10/1988 |
| From: | Burnett P, Dan Collins, Fredrickson P, Herdt A, Kuzo G, Verrelli D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20149F517 | List: |
| References | |
| 50-160-87-08, 50-160-87-8, NUDOCS 8802170242 | |
| Download: ML20149F548 (20) | |
See also: IR 05000160/1987008
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Report No.: 50 160/87-08
Licensee: Gecrgia Institute of Technology
225 North Avenue
Atlanta, GA 30332
Docket No.: 50-160 License No.: R-97
Facility Name: Georgia Institute of Technology Research Reactor (GTRR)
Inspection Co uc d: December 16, 1987, anuary 4-5 and 14-22, 1988
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Inspectors:
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. Fredrickson, Team Leader
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Accompanying Person : A. am . P ef:t Ma 4.r
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/"D M. Verrelli, Branch 'ief '
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D. H. Collins, Branch Chief Date Signed
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Di s n f Radiation Safety and Safeguards
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A. . Herdt, Branch Chief Dat6 Signed
Division of Reactor Safety '
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SUMMARY
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Scope: This special team inspection involved onsite review of GTRR operations
and health physics activities regarding a contamination event which occurred ,
during the week of August 17, 1987, and followup of allegations, Technical i
Specification changes, and licensee actions regarding previous enforcement
issues. ,
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Results:. . Four violations with multiple examples for each were identified -
failure to' have adequate procedures and failure to follow precedures for
handling and manipulating experiment material and for surveying and evaluating ;
potential radiological hazards; failure to conduct adequate radiation surveys ,
of the reactor. building and GTRR personnel and their personal property for
evaluation of exposure to radioactive material; failure to conduct adequate air
sampling and bioassay analyses for evaluation of personnel exposure to airborne
radioactive material during experiment and decontamination actlyities; and
failure to document and maintain records of radioactive material contamination
surveys conducted. .;
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1. .' Persons Contacted 's s
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License'e 'fmployee s s ,
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*R. Boyd, Manager,:0f fic'e of Radiation Safety
*W. Downs', Senior Reactor Operator
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B. Kahn, Radiation Safety Officer ' .'
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#*R5'Karam, Director, Nuclear Research Center r."
#*R. ?iacDonald, Associate Director, Nuclear Research Center
*L. McDovcil,' Manager of Reactor Operations .
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*S. Millspaugh, Health Physicist
A. Moore, Director, Research Communications Office
*P. Sharpe~, Heal.th Physicist
#T. Stolson, Vici President for Research ,
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Other licensee employees contacted included technicians auds offico
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Nuclear Regulatory Commission attending the second exit intervien
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#A. Adams, Project Manager, Nuclear Reactor Regulation, Washington,'C. O "c
#M. Ernst, Deputy Regional Administrator, Region II
* Attended exit interview on January 5, 1988
# Attended exit interview on January 22, 1988 - "
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2. Exit Interview (30703)
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The inspec' tion scope and findings were summarized on January 5.1988, and
January 22, 1988, with those persons indicated in Paragraph 1 abov6; The
inspector 4iescribed the areas inspected and discussed in detail the
inspection findings. On Jinuary 5,1988, the inspector discussed four
apparent violations with multiple examples for each in regard to a topaz
- i rradiatio(' experiment which resulted in the subsequent release of
contaminated material within the reactor containment building during the
week of August 17, 1987. The violations involved the failure to have ,,
adequate procedures and the failure to follow procedures for handling and
manipulating experiment mater W and for surveying and evaluating
potential radiological hazafds (Paragraphs 3.b, 3.c, and 3.d); failure to
. conduct adequate radiation surveys' of the reactor building and GTRR
# personnel and their personal property potentially exposed to radioactive
contamination (Paragraphs ,3.c and 3.d); , failure to conduct adequate air
sampling and bioassay analyses to evaluate personnel exposure to airborne r '
radioactive contamination durino. the experiment and decontamination
activities in the recctor containment building (Paragraph 3.d); and
. failure to document and maintain records of radiological contamination and ~~
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personnel surveys conducted (Paragraphs 3.c and 3.d). At'this meeting the
inspector also reviewed an appbrent violation concerning ariministrative
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changes.made to the licensee organization without an approved amendment to
the Technical Specifications (TS). A violation regarding the
reorganization will not be issued at this . time since the licensee had
informed the NRC during a May 4,1987, Enforcement Conference, of the
. reorganization and since the licensee had taken the position during the
Conference that this change was necessary to enhance safety. Further, NRC
representatives stated in the January 22, 1988 exit-interview, that the
present organizational structure was being reviewed by NRC, Office of
Nuclear Reactor Regulation (NRR) for its adequacy. NRC management stated
that the inspection findings indicate that there had been no improvement
in the licensee's performance 'since .the May 4, 1987 enforcement
conference, and may indicate problems in management control. Licensee
representatives stated that they believed management control had been
strengthened since July 1,1987, when the GTRR management reorganization
occurred. Tne licensee agreed to review the violations and respond
appropriately. The licensee did not identify as proprietary any of the
material provided to or reviewed by the inspector during this inspection.
3. Followup of the August 17-21, 1987, Contamiration Event (93700)
a. Event Identification
On December 16, 1987, while reviewing management reorganization
concerns for the GTRR program, the inspector identified a
contamination event which occurred at the reactor facility during the
week of August 17, 1987. The event involved the handling and
manipulation of topaz irradiation experiment material which resulted
in the release of radioactive contamination, initially identified as
Cadmium-115 (Cd-115), within the reactor containment building.
Licensee staff stated that subsequent decontamination efforts were
required to reduce contamination to normal levels in specific areas
of the reactor building. At the time of the December 1987 NRC
insp ction, a detailed description and evaluation of the event had
not been prepared by licensee staff or management. This inspection
was continued on January 4-5, 1988, and continued with a team
inspection conducted from January 14-22, 1988. The purpose of the
< inspection was to assess the adequacy of the licensee's operations
and health physics programs, and to evaluate licensee management's
ability to address and evaluate any potential health and safety
issues associated with the design, completion and evaluation of
e irradiation experiments.
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b. Topaz Irradiation Experiments
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The inspector reviewed selected procedures, records, and memoranda,
and interviewed GTRR operations and health physics staff and
management concerning topaz irradiation experiments and the
subsequent August 1987 radioactive contamination event.
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TS 6.4.b(2) and 6.4.b(6) require written procedures to be provided
and utilized for installation and removal of experiments and
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experimental facilities; and radiation and radioactive contamination
control.
Topaz gemstones had been irradiated using the GTRR facilities since -
' April.1987. Topaz gemstones to be irradiated ~were placed within an
aluminum (A1) canister which was lined with a layer of cadmium (Cd).
The topaz experiment criteria were specified in licensee procedure,
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Request for Minor Experiment Approval, GTRR Reference No. R7420,
dated April 4, 1987, which detailed experiment conditions and
expected radiation levels. Experiment criteria included limiting
irradiation times to 30 megawatt hours (MW-hrs.) and listed the .
expected radiation levels from the cadmium and aluminum material used
in the experiment to be "nil." After an initial topaz irradiation
conducted during April 1987, surveys showed that a maximum measured
radiation dose of 3 rads per hour (R/hr) at 1 inch from the
experiment material (outside of the Al canister) was recorded
following various irradiation times. Licensee procedure 3102,
Quality Assurance for Experiments, dated October 28, 1982, specified
that the primary Quality Assurance (QA) document was the
Experimenters Checklist and Schedule Form. This procedure required
that for Ca*.egory 4 Experiments - Internal, for example, the topaz
irradiation experiment, the approval form to address quantitative
controls of the experiment (e.g. irradiation time, flux levels,
reactivity effects, etc.). From discussion with cognizant licensee
personnel and review of topaz irradiation experiment data, the
inspector determined that calculations for the activation of the
cadmium and aluminum were not included in the evaluation of the
expected radiation levels from the experiment. Furthermore, the
unexpected high dose rates recorded for the 'nitial April 1987
experiment were ignored by the operators during the QA review of the
experiment results. The failure to follow procedures to properly
evaluate the expected activation of irradiated material' and
subsequent dose rates for the topaz irradiation experiment was
identified as an example of a violation of Technical
Specification 6.4.b(6) (50-160/87-08-01a). In addition, the
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inspector noted that the procedure did not specify a time frame for
such reviews to be conducted or a method of review of findings by
management. Based on the lack of these essential elements in the
procedure, the procedure was inadequate and was identified as another .
example of a violation of Technical Specification 6.4.b(6)
(50-160/87-08-01a).
l For the August 1987 event, the topaz experiment was irradiated ! intermittently from July 17, 1987, to August 14, 1987. The inspector
noted that the actual irradiati3n conditions, that is, dates,
ir.adiation duration and power, and reactivity effects were not
provided on the Experimenter's Checklist and Schedule form as required.
Review of the control room operations log indicated that the
August 1987 topaz experiment was irradiated for 41.8 MW-hrs. compared
to the 30 MW-hrs. allowed by the Experiment Approval Form. The failure
to follow procedural time limitation for irradiation of the topaz
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experiment was identified as an additional example of a violation of
TS 6.4.b(2) requirements (50-160/87-08-01b).
For ' irradiation, the Al canister had been .placed in the reactor
through vertical ports accessed by removal of-shielded plugs located
on the top of the reactor shield. _ Subsequent to irradiation, the
topaz experiment materials were maintained in the shut-down reactor
to allow short-live activation products to decay. For the August
1987 topaz experiment, irradiation was terminated on August 14, 1987,
and the experiment material was maintained in the reactor .until
August 17, 1987. On August 17, 1987, the material was removed from
the reactor vertical port using licensee procedure, Extended
Radiation Work Permit, Insertion and Removal of Routine Samples from
Vertical Irradiation Facilities by Licensed Reactor Operators,_ dated
July 23, 1981. Upon removal of the material from the reactor, the
total beta gamma dose rate at orie foot from the experiment material
(aluminum canister) measured approximately 3 R/hr. From discussions
with licensee personnel and review of training and experiment
records, the removal of material from the reactor was conducted as
authorized by the licensee RWP procedure. The experiment material -
was transferred for storage to a shielded lead cask (pig) located
nearby on the southeast edge of the top of the reactor shield.
On August 18, 1987, the operator removed the experiment material
from, and placed the material on top of, the storage pig. In an
attempt to determine the source of the unexpected high dose rates .;
recorded for the experiment material, the operator opened the
aluminum canister and transferred the topaz into a glass beaker. The '
dose rates measured for the topaz and the aluminum canister which .
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still contained the cadmium layer were not recorded. However, the
operator stated that both the cenister tnd topaz material showed high
radiation levels. The operator replaced the aluminum canister
containing the cadmium material into the storage pig and the topaz
behind a shielded area and exited from the area on. top of the reactor
shield. The operator stated that no procedures, other than the RWP
utilized for removal of the experiment from the reactor, were used to
detail the handling and manipulation of the experiment material. The
inspector noted that external dose rates were expected to be
significant, approximately 3 R/hr at 1 foot from the material on
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August 16, 1987, and questioned what criteria were followed to limit
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personnel whole body and extremity exposures during handling of the
material. The operator stated that health physics training and good
practices were followed. The inspector discussed training with
licensee staff and management and noted that the operator had
received self-monitoring approval which allowed his handling of the
material. The operator stated that tongs (12-18" in length) were
utilized for handling the canisters to decrease extremity exposure,
and that the manipulation and transfer of the topaz from the canister
was conducted in less than 30 minutes. Monitoring of extremity
exposures was not conducted and whole body exposures were monitored
using film badges (see Paragraph 3.d).
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- During a tour of the area on top of the . reactor on January 4,1988,
the inspector noted that tongs and other equipment were not available
for use. Also, the inspector noted that the area on top of the
reactor shield was maintained as a contaminated area and required
shoe covers and gloves as minimum protective clothing- (PC). -The
operator stated that on August 18, 1987, only the minimum PCs were -
worn. The inspector discussed personnel surveys conducted by the-
operator on August.18, 1987, subsequent to handling the experiment
material. The operator could not remember if he had conducted a
. personal radiation survey (frisk) prior to exiting the area on top of
the reactor and indicated that survey meters often were not available
at the access step-off pad area. Furthermore, -review of procedures.
indicated that frisks were not required to be performed. The absence
of survey equipment in this area was confirmed during the tour of the
reactor building conducted on January 4, 1988. Prior to exiting the
reactor containment building, the operator surveyed (frisked) his
hands and shoe bottoms as required at the reactor containment'
building access area and stated that radioactive contamination was
not detected ~ at that time. The operator then departed from the
campus and returned to his residence via public transportation.
According to the operator involved, other individuals were not in the
immediate area on the top of the reactor shield or on the ground
/loor area immediately beneath where the experiment material was
handled. In addition, the operator confirmed that he was the last
individual to leave the reactor containment building on August 18,
1987.
The inspector reviewed selected Radiation Work Permits (RWPs), and
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health physics and operations procedures and discussed the handling
and manipulation of experiment materials in open areas, such as on
top of the reactor shield, subsequent to irradiation experiment:,.
Licensee representatives stated that irradiated samples had been
handled (opened) and manipulated in the area on top of the reactor
shield. This area did not have fume hoods or containment structures
available to minimize the spread of contamination while handling
experiment material. The inspector noted that neither the health
physics nor operations procedures addressed the proper methods for
radiation contamination control and exposure control associated with
these activities in this open area. Furthermore, licensee
Procedure 9280, Personnel Surveying, indicated that surveys "should"
be conducted immediately af ter exiting any contaminated area or
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working with any contaminated material but did not require such
surveys. The inspector noted that to control the spread of
contamination, surveys must be conducted at exit points to all
contaminated or contamination control areas. The failure to have
adequate operation and health physics procedures to control or
prevent the spread of radioactive contamination and to control
personnel exposure while handling and manipulating irradiated
experiment material was identified as an additional example of a
violation of TS 6.4.b(6) (50-160/87-08-01c).
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c. Identification of Contamination
10.CFR 20.201(b) requires the licensee .to make or 'cause to be made
such surveys as (1) may be necessary for the licensee to comply with
regulations > in this .part, and (2) are reasonable under. the ,
circumstances to evaluate the extent of radiation hazards that may be
present.
10 CFR 20.401(b) requires each licensee to maintain records in the
same units used in this part, showing the results of surveys required
by 20.201(b). 10 CFR 20.401(c) requires that records of results of
surveys and monitoring maintained pursuant to Paragraph 20.201(b) of
this section shall be preserved for two years after completion of the
survey.
In the afternoon of August 19, 1987, radioactive contamination,
approximately 100-200 counts per minute (cpm) ab)ve-background, was
found'on the first (main) floor of the reactor c ntainment building
during routine surveillance activities. Discussion with cognizant
licensee health physics staff indicated that followup surveys of the
licensee facilities showed radioactive contamination in the south to
southwest areas of the reactor containment building including the
floor area on top of the reactor shield near the storage pigs; the
main floor and on equipment located there; and on a- catwalk located
approximately 30 feet across from, and at the same elevation as the
top of the reactor shield.
From discussion with cognizant licensee representatives the inspector
determined that from approximately one-fourth to one-third of the.
reactor containment building had measurable contamination above
background levels. Licensee representatives stated that they
believed the spread of contamination occurred on August 18, 1987,
when the operator opened the Al canister and manipulated the topaz
experiment material in the area on top of the reactor shield.
Additional supp rt for this conclusion was based on review of
activities conducted at the GTRR prior to discovery of the
contamination, the location of the highest radiation levels, and the
direction of the spread of contamination. That is, ventilation to
the reactor building exhausted its flow in a southerly direction
across the reactor top area where the experiment material was
handled. This air flow was believed to have caused the distribution
of contamination in the reactor containment building.
The inspector reviewed licensee records of weekly gross radiological
surveys conducted during August 1987. For a August 19, 1987 survey,
only the south main floor area, directly beneath the top of the
reactor shield area where the experiment material was handled,
indicated radiation levels approximately 100-200 cpm above
background. The inspector noted that these survey records did not
show survey results for all areas in the reactor containment, for
example, the area on top of the reactor. From discussion with
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cognizant licensee representatives, and review'of a memorandum from
R. Boyd, Manager, Office of Radiation Safety to, R. Karam, Director,
GTRR, dated August 20, 1987, the highest measurement of radioactive
contamination was 20 mR/hr noted on.a maslin wipe'of a large area on
top of the reactor shield where the irradiated topaz experiment was
~ handled on August 18, 1987 Initial licensee gamma spectroscopy
analyses of some contaminated material indicated Cadmium-115 (Cd-115)
as the main contaminant. Although surveys .of contamination were
made, licensee representatives stated that detailed surveys defining
the area and quantitative amounts'of radioactive contamination on all
surfaces (for example walls and vertical surfaces potentially exposed
to the airborne contamination in the reactor building) were not
conducted. The failure to conduct adequate surveys of the reactor
building following the discovery of contamination to evaluate the
extent of radiation hazards that may be present was identified as a
violation of 10 CFR 20.201(b) requirements (50-160/87-02a). In
addition, the failure to maintain records of the limited surveys
conducted on August 19-20 as a result of the contamination event was
identified as a Violation of 10 CFR 20.401 requirements
(50-160/87-08-03a).
On August 19-20, 1987, decontamination of the reactor containment
bu!1 ding was conducted by the GTRR operators with assistance provided
by the health physics staff. On August 19, 1987, one of the
operators, the individual involved in handling the topaz experiment
material the previous day, detected contamination on.his pants cuff
and shoe. The shoe was decontaminated but all contamination
(approximately 500 cpm) on the pants could not be readily removed.
The contaminated p?nt material was collected and discarded as
radioactive waste.' Licensee individuais concluded that the
individual became contaminated during decontamination efforts.
However, these pants had been worn by the operator on August 18,
1987, and the inspector questioned if the contamination could-
have occurred during handling of the experiment material. The
inspector noted that the operator Fad not worn a lab coat nor had he
conducted a whole body frisk on August 18, 1987. The inspector
questioned if the licensee had surveyed the operator's residence and
personal possessions to verify that radioactive material had not been
transferred from the reactor buildirg on the operator's clothing.
Licensee representatives stated that on August 19, 1987, the operator
was instructed to conduct a survey of his residence and personal
property using the appropriate instrumentation. Results of this
survey were not available to the inspector. During the pre-exit
meeting conducted on January 5, 1988, with licensee staf f, the
operator stated that the survey of his residence and personal
property had not been conducted as licensee personnel were led to
believe. The failure to conduct a survey to verify the absence of
radioactive contamination at the operator's residence was identified
as additional example of violation of 10 CFR 20.201(b) requirements
(50-160/87-08-02b).
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The inspector reviewed the August 21, 1987, survey results for the
reactor building following decontamination activities. Surveys from
the ' upper level clean area and first floor of the containment
building showed levels of less than 200 disintegrations per minute
per 100 square centimeters-(dpm/100 cm2 ). Surveys of the floor.on
top of the reactor shield and the top of the storage pigs maintained
there measured 300 dpm/100 cm2 and 400 dpm/100 cm2 respectively.
During the inspection, licensee representatives . stated that an
evaluation into the cause of the apparently loose radioactive
particulates that were released from the Al canister would be
conducted after radiation levels had subsided. On January 20, 1988,
radiation levels at 6 inches from the canister were approximately
600 mR/hr. The inspector noted that based on the initial August 17,
1987, radiation levels and the known half-life of Cd-115
(approximately 54 hours), the radiation _ levels were not reduced - as
expected for the January 20, 1988 measurement and were indicative of
the presence of other radionuclides. Although the inspector was
unable to sample the contaminants directly because of the high
radiation levels and the possibility of spreading loose contamination
by manipulation of the canister, a wipe sample was collected from the
storage pig which was contaminated. An NRC laboratory qualitative
gamma spectroscopy analysis showed the following: Zinc-65, 245 day
half-life; Cadmium-109, 453 day half-life; Tin-113, 115 day
half-life; and Antimony-124, 60.4 day half-life. Based on
discussions with licensee representatives, another radionuclide
identified, i.e., Cobalt-60, was attributable to previous
contamination of the storage pig. From the new radioactive materials
identified, both licensee and NRC representatives concluded that the
contaminants were activation products of the cadmium and soldering
material used in fabricating the cadmium layer used in the
experiment. The inspector noted that Technical Specification 3.4
required materials of construction, fabrication and assembly
techniques be specified. Licensee representatives stated that
materials and assembly techniques were not specified for the topar
irradiation experiment. The failure to have procedures for the
construction and fabrication of the experiment material was
identified as another example of a violation of TS 6.4.b(6)
(50-160/87-08-01d).
d. Personnel Exposure
10 CFR 20.103(a) requires that no licensee shall possess, use, or
transfer licensed material in such a manner as to permit any
individual in a restricted area to inhair a quantity of radioactive
material in any period of one calendar quarter greater than the
quantity which 'rould result from inhalation for 40 hours per week for
13 weeks at uniform concentrations of radioactive material in air
specified in Appendix B, Table 1, Column 1. 10 CFR 20.103(a)(3)
requires that for purposes of determining compliance with the
requirements of this section the licensee shall use suitable
measurements of concentrations of radioactive materials in air for
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detecting and evaluating airborne radioactivity in restricted areas'
and in addition, . as : . appropriate, shall 'use measurements of
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radioactivity 'in .the body, measurements of radioactivity excreted
from the body, ~ or any combination of such ' measurements as may be
necessary for timely detection and assessment of individual intakes
of radioactivity of exposed individuals.
From discussion with cognizant licensee representatives and review of
the. distribution and levels of -radioactive contamination, the.
inspector determined that licensee personnel were potentially exposed
to airborne radioactive contamination on August 18, 1987, during
manipulation of the experiment material and also- during
decontamination activities conducted on August 19-20, 1987. The
inspector reviewed and-discussed the licensee's' health physics survey
and personnel monitoring procedures used to evaluate .. potential
personnel exposure to airborne radioactive material.
Licensee representatives stated that a . continuous air sample was
collected in the reactor containment building during the week of
August 17, 1987. Licensee management stated that radiological
analyses of daily air samples indicated that during .the week of
August 17, 1987, levels of airborne radioactive contaminants were
below 10 CFR Part 20 Appendix B concentrations for gross activity,
air concentration averaging approximately 1 E-10 uC1/cc.
Procedures and rnethodology used to conduct the samnling and
subsequent analyses of airborne radioactive contamination were
reviewed and discussed. Licensee representatives stated that only
one air sampler was located in the reactor building at the northeast
area of the first floor beneath the control room. The inspector
noted that this sampler was not et the areas of the building where -
the major contamination was identified. Thus, the air samples
analyzed were not representative of concentrations of radioactive
material to which personnel were exposed. Furthermore, the licensee
could not verify with data that the sampler was representative of the
general air concentrations in the reactor building. The failure to
use suitable measurements of concentrations of radioactive materials
in the air for detecting and evaluating airborne radioactivity during
the week of August 17, 1987, was identified as an apparent violation
of 10 CFR 20.103(a)(3) requirements (50-160/87-08-04a).
In addition, the inspector reviewed licensee procedure No. 9038, Air
Sample Analysis. The procedure as written was inadequate in that if
following initial analyses, samples which exceeded 10 CFR Part 20
Appendix B limits for gross activity were not required to be
maintained for a sufficient time period to allow decay of naturally
occurring radon daughter products and then be recounted to accurately
evaluate the actual radiation hazards present as a result of GTRR
operations. The failure to have adequate procedures for air sampling
analyses was identified as an additional example of TS 6.4.b(6)
violation (50-160/87-08-01e).
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Licensee representatives conducted an "in-vivo" chest survey analysis
for the operator potentially exposed to airborne radioactive
materials. The in-vivo chest survey was conducted on both August 19
and August 20, 1987, using a 3 X 3-inch sodium iodide (NaI) detector
held against the operator's chest for approximately 5 minutes.
Licensee representatives indicated in a memorandum from R. Boyd,
Manager of the Office of Radiation Safety (MORS), to R. Karam,
Director GTRR, that the August 19, 1987, data of these surveys may
have shown a "slight positive indication of Cd-115" however, the
results were inconclusive as to whether it was external or internal
contamination. Furthermore, no indication of positive results was
noted for the August 20, 1987 analysis. These data were not
retained. The failure to maintain records of the "in-vivo" chest
surveys was identified as another example of a violatio1 of
10 CFR 20.401(b) requirements (50-160/87-08-03b).
Furthermore, the lower limits of detection for these chest tnalyses
had not been calculated. The inspector noted that there was no
procedure to provide guidance for calibration and operation of the
NaI detector to specify analytical detection capabilities for the use
of "in-vivo" chest surveys to meet 10 CFR 20.103 requirements. The
failure to have adequate procedures for the "in-vivo" chest survey
was identified as another example of a violation of TS 6.4.b(6)
(50-160/87-08-01f).
A urine sample was collected from the operator on August 20, 1987,.to
evaluate potential uptake of radioactive materials. The sample was
analyzed using liquid scintillation counting (LSC) methodology. The '
inspector reviewed the urine bioassay procedure 9036, E.ioassay
Analysis Methodology, and results of radiological analysi!. of the
operator's urine sample. The LSC methodology had been established
only for the analysis of tritium and other beta emitting isotopes,
for example, Carbon-14. Although the licensee identified Cd-115 as
the major airborne radioactive contamicant, the operator's urine
sample was only analyzed for tritium as noted on the licensee's
sample processing report. The failure to properly analyze the urine
sample for the expected contaminants sas identified as another
example of a violation of 10 CFR 20,201(b) survey requirements
(50-160/87-08-02c).
The inspector noted that quantitative results for a beta gamma
emitting nuclide could not be conducted with the licensee's present
methodology. The procedure failed to address standard practices in
calibration of the LSC or use of other instrumentation for beta gamma
quantitative analyses. The procedure also failed to addre ss standard
methodology, for example, the use of distillation to minimize
quenching in LSC analyses, to quantify esults. Furthermore the
appropriate biological retention models (for example, Report of
Committee II on Permissible Dose for Internal Radiation,1959 [ICRP
II] methodology) to relate calculated body burdens to estimate a
person's intake of airborne radioactive contaminants were not
specified. The f ailure to have adequate urine bioassay analyses and
internal exposure evaluation procedures were noted ts another example
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of a violation of TS 6.4.b(6) (50-160/87-08-01g). In addition, the
failure to have adequate bioassay analyses, that is, whole body
counting or urine bioassays, conducted to evaluate the operator's
exposure to airborne radioactive contamination was identified as an
additional example of a violation of 10 CFR 20.103 requirements
(50-160/87-08-04b).
External dose estimates for personnel involved in the August 1987
event were discussed. For the measurement period of August 1-31,
1987, the whole body dose as measured by film badge was 30 millirem
for the operator involved in the contamination event. For this
operator an estimate and/or direct measurement of extremity doses was
not conducted. For all other individuals involved in decontamination
activities, film badge whole body dose measurements did not exceed
30 millirem for the August 1-31, 1987 period.
e. Followup Activities
On January 11, 1988, a radiological contamination survey was made by
the licensee of the residence of the operator involved in the
contamination event. Direct radiation measurements of selected areas
and personal items, for example, floors, counter tops, personnel
clothes and shoes, in the residence using a GM detector were at
background levels. In addition, wipes (smears) using filter paper
discs were taken of surfaces within the residence and then andlyzed
for radioactive contamination using a low beta gas proportional
counter. All smear results were at background levels.
4. Technical Specification Changes
Independent of allegations discussed in Paragraph 6, and independent of
the NRR review, the inspector audited several of the changes incorporated
into the organization and plant operations for the proposed TS changes.
The existing organization was reviewed with respect to the previous
organization. The organizational change as made on July 1, 1987, and the
TS change request was submitted on August 6, 1987. On December 7, 1987,
NRR requested additional information or this change request. As of
January 22, 1988, the licensee had not responded to this request.
Although NRR will make the final deciCon on this change request, the
inspection focused on the existin ' struc'.ure without regard to the TS
change request.
The inspection identified one area in the new organization that did not
appear to be functioning satisfactorily. In the present structure, the
position of Chairman, Nuclear Safeguards Committee and Radiation Safety
Officer are held by the same individual and the organization chart shows
that (old) position reporting to the facility Director. This arrangement
eliminated the independent facility oversight responsibility usually
associated with this type of position. It was clear from discustions with
the RSO that the RSO was involved in oversight of activities at GTRR only
on request of the Director, GTRR. The Director stated that a change to
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the reporting level within Georgia Tech for 'these positions would be
considered.
Also, a portion. of the TS change involving a change from a helium to a
nitrogen reactor vessel cover gas is discussed in Paragraph 7.b.
No violations or deviations were identified.
5. Personal Interviews
During the inspection, interviews were conducted with the facility staff
to evaluate working relationships with respect to the August 1987 event
and also to evaluate the effect of changes made to the facility since the
enforcement conference of May 4, 1987. Although the interviews did
generate numerous comments and personal opinions of those ' interviewed,
.several general observations can be made:
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All personnel appeared to be conscientious in wanting to perform
their job in a safe manner.
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Working attitudes between health physics and operations have
continued to deteriorate. .
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The facility relies on informal training in lieu of procedures for
many routine tasks.
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The Deput/ Diractor position is a positive change to the facility. !
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Operations personnel appear satisfied with the Director's management
efforts.
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Health physics personnel appear to believe that the Director is
involved too much in day-to-day health physics activities to the
detriment of those activities.
The interviews, although not identifying any regulatory concerns, did
reveal that overall performance at the facility appears to have degraded
since the May 4, 1987 enforcement conference.
No violations or deviations were identified.
6. Allegation Followup (99014)
a. Allegation (RII-87-A-0090)
The alleger stated that the management reorganization made in regard
to the Neely Nuclear Research Center, Georgia Institute of Technology
Research Reactor (GTRR) program on or about July 1, 1987, was in
violation of NRC regulations. The specific concerns were as follows:
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(1) An amendment to incorporate the administrative changes int : ;he
licensee Technical Specifications (TS) was not-submitted at the
time the changes were implemented and thus, the subsequent
approval from the proper regulatory group (NRC) was not received
prior to implementing the reorganization.
(2) The GTRR health physics group does not have the ability to
terminate work at the facility which could . lead to potentially .
dangerous conditions.
(3) A new Radiological Safety Officer (RS0) was appointed and the
previous RSO demoted to the position of Manager of-the Office of
Radiation Safety (MORS). The actual duties of the new RSO did
not meet the professionally defined criteria for a RSO.
(4) The Office of Radiological Safety was renamed the Radiation
Safety Office without proper approval.
(5) The Nuclear Safeguards Committee (NSC) has been disbanded.
(6) The Director of the GTRR, a licensed isotope user, has authority
over the group meant to keep uses of radioactive materials in
check. Only this person can terminate work being done for his
own uses.
(7) The Radiation Protection Committee no longer controls the use of
isotopes on campus. The Director of the GTRR r.ow has that
responsibility,
b. Discussion and Findings
The inspector discussed the above concerns with the GTRR
management and staff, and interviewed the campus RSO. Current ,
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licensee procedures regarding authority to address potential
radiological health and safety issues at the GTRR also were reviewed.
In addition, licensee correspondence sent to the NRC, and management
issues discussed during an enforcement conference held with NRC
Region II personnel were reviewed and discussed. Information
collected as a result of personnel interviews, and the procedure and
correspondence reviews, was audited against applicable regulatory
codes, licensee TSs and the GTRR Safety Analysis Report (SAR).
(1) TS 6.2 and the GTRR SAR Section 6.1 detail the licensee
management organization. A proposal to change the GTRR
management organization was discussed during an enforcement
conference conducted on May 4, 1987 (IE Report
No. 50-160/87-06). At that time, NRC Region II management was
informed that the GTRR organization would be changed on or about
July 1, 1987. The licensee implemented the change on July 1,
1987, and provided the proposed TS change to the NRC Region II
staff in a letter dated July 6, 1987. A formal application to
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change the licensee Technical Specifications to reflect - the
present management structure was submitted to the NRC, Office of
Nuclear Reactor Regulation (NRR), by letter - dated- August 6,
1987.
The GTRR organizational change was . made on July 1, 1987, as
stated. Currently, the new GTRR management organization, its
functional responsibilities and adequacy to protect public
health and safety are being reviewed by NRR. This review
includes evaluation of the adequacy of the present GTRR
management organization and staff to meet the TS and - SAR
criteria.
(2) TS 6.4.b(6) requires written procedures to be provided and
utilized for radiation and radioactive contamination control.
The Georgia Institute of Technology (GT) Radiation Safety-
Manual, Section III, dated July 1987, indicates that the Director
of the GTRR has the authority to terminate any experiment which
uses radiation sources if the methods and/or procedures used in
such experiments are declared unsafe and contrary to
regulations. In addition, Section It/ authorizes the Manager of
the Office of Radiation Safety to supervise the campus
health physics program. The manual specifies that the MORS can
suspend on a temporary basis, any operation causing excessive
radiation hazards as rapidly as possible. From interviews of
reactor facility personnel and from review of research conducted
at the GTRR, no examples were identified where health physics
had recommended cessation of work and were overruled.
(3) TS 6.1.c details that the RSO, organizationally independent of
the GTRR operations staff, advises the director of the nuclear
research center in inatters pertaining to radiological safety at
the GTRR. The GTRR SAR, Section 6.1, details that all health
physics personnel are responsible to the campus RSO. Further,
the RSO is directly responsible to the President of the
Institute.
A new RSO was appointed on July 1, 1987. The recently appointed
RSO is not a member of the iRR staff but he reports directly to
the Director, GTRR, for RSO functions according to the new
organizational chart. Thus, TS requirements and SAR criteria
describing the RS0's independence from the GTRR organization and
his functional control over the health physics staff were not
met. This organizational change of TS and the apparent
deviation from SAR commitments concerning the lack of health
physics staff independence from the current GTRR management
structure is being reviewed by NRR.
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(4) The GTRR TS do not require an Office of Radiological Safety.
However, licensee SAR Sections 6.1 and 6.4 specify that the
Office of Radiological Safety authority is derived from the~'
President of. the Institute and is, responsible -for radiation
safety _ training, radioactive materials inventory, surveys,
effluent releases and personnel monitoring at the reactor
facility.
The Office of Radiological Safety was abolished and renamed the
Office of Radiation Safety on July 1,1987. This office was
integrated as a functional unit into the GTRR organization and
mandated to conduct required training, monitoring, and survey
requirements. However, the current GTRR administrative
authority over the Office of Radiation Safety does not appear
meet the intent of the GTRR SAR. This change in organization
currently is being reviewed by NRR.
(5) TS 6.2 details the licensee responsibilities and requirements
for a Nuclear Safeguards Committee (NSC). The GT Radiation
Safety Manual, dated July 1987, requires the Committee to meet
quarterly and as circumstances warrant.
Interviews with selected GTRR personnel and the campus RSO
indicated that the NSC is operational and has met as required by
the licensee's TSs.
(6) The GTRR TSs and SAR do not prohibit the Director of the reactor
research center from utilizing the nuclear research reactor
facilities or isotopes. However, the licensee SAR, Section 6.1,
details that the operating staff is completely dhorced from the
research programs and the Reactor Supervisor is in charge of the
reactor operations group.
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The Director of the GTRf, has conducted gem irradiation research
experiments through the reactor operations staff. The Director
has authority over the GTRR operations. Thus, use of the
operations staff to conduct research for the Director of the
reactor facility apparently deviates from the intent of the GTRR
SAR. The GT Radiation Safety Manual, Sections III and IV,
authorize that either the Director of the GTRR center or the
MORS can suspend any operation causing unsafe or excessive
radiation hazards. The present organization structure in
relation to its intended safety function is presently undergoing
review by NRR.
(7) The licensee's TS criteria and SAR specifications do not require
establishment of a Radiation Protection Committee nor prevent
the Director of the GTRR from having authority over use of
radioactive materials regulated by the state license.
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The Radiation Protection Committee was abolished and the.
Director, GTRR, was authorized responsibility for all nuclear
research and safety issues on July 1, 1987. The Health Physics
Manual, dated July 1987, Section III, lists the responsibilities
of the Director of the Neely Nuclear Research Center,- including
the direction and operation of the GTRR and all nuclear research
and radiation safety issues on the GT campus.
c. Conclusion
Several of the issues in the allegation were substant'.ated in that
the administrative, organizational and management changes occurred as
stated. These management and organizational changes to the GTRR TS
requirements and SAR commitments were made by the licensee. These
changes. currently are being reviewed by NRR. Other issues
identified, that is, disbandenment of the NSC and only the Director
of the GTRR having control of experiments were not substantiated.-
No violations were identified.
7. Licensee Action on Previous Enforcement Matters (92702)
Violaticns identified in Inspection Report 50-160/87-01 and responded to
in licenAee 'etters dated May 25, 1987, and July 15, 1987, were reviewed.
The identification of the violation in the discussion is consistent with
that used in the Notice of Violation enclosed with the report.
a. (0 pen) Violation 50-160/87-01-01, Failure to Provide or Utilize
Procedures.
(1) A.1.a: There was no approved procedure to measure excess
reactivity to assure it cid not exceed the limits ~ of
Technical Specification 3.1.e. Procedure 7246, Control Element
Reactivity Worth, was revised on October 30, 1987. Step V
established an excess reactivity calculation, using the measured
worths, on the Reactivity Worth Report Form (Page 3 of the
procedure), and stated an acceptance criterion that excess
reactivity be less than 11.9% delta-K/K. The procedure did
require a plot of the integral control element worth curves, but
did not provide data sheets to capture the reactor period and
associated reactivity worth data obtained in the measurements
and used to plot the data and to satisfy the acceptance
criteria. Review of some of the completed procedures in the
files revealed that general purpose data sheets, appropriately
headed and filled out, had been used to capture the data and
were attached to the completed procedures. This portion of the
violation is closed.
(2) A.1.b: There was no approved procedure to measure the 0,
concentration in the cover gas to assure it was less than 2% by
volume before making the reactor critical as specified in
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Technical Specification 3.6.e. The licensee purchased a- gas
chromatograph and prepared Procedure 4400, '03 Analysis in
Reactor Cover Gas (issued October 30, 1987) for its use. The
, procedure had been performed once, and the measured
concentration was much less than the limit. The procedure
specified an annual performance frequency, but no technical
justification of that frequency was provided by the licensee.
The reactor operated for less than 200 megawatt hours in the
recent year, but.is capable of thousands of megawatt hours of
operation in a year. Increased operation and other activities-
not considered by the licensee may affect the rate of Da
production. Although no longer surveillance interval is
acceptable, other considerations may dictate a shorter interval
is necessary to maintain confidence Technical Specification
3.6.e is always satisfied. The licensee agreed to perform a
quantitative evaluation of the frequency of cover gas
surveillance (Inspector Followup Item 160/87-08-05). This
portion of the violation is closed.
(3) A.1.c: The requirement of TS 3.5.b.6 that the containment
isolation valve closing time.be less than five seconds was not
addressed'in a surveillance procedure. Procedure 7220, Building
Isolation Test,. was revised to include Step D to measure the
valve closing time and an acceptance criterion of less.than five i
seconds was specified. The procedure did not include data
sheets for recording the measurements, nor did it specify the
method used to determine valve closing time. Discussion with
management and the operator who performed the test confirmed
that the measurement was based upon observing stem travel time.
Management agreed to revise the procedure accordingly. This
portion of the violation remains open.
(4) A.2.a: Procedures 2350 and 2400 referred to helium as the cover
gas although nitrogen is currently oeing used (see also
violation D). The procedures have been changed to indicate
nitrogen as the cover gas. This portion of the violation is
closed.
(5) A.3.a: The operators failed to log data required by procedure.
The operators were counselled by management on the requirements
to follow procedures. Audits of the operators' log by
management have confirmed acceptable log keeping. This portion
of the violation is closed.
(6) A.3.b: Procedure 2210 was not performed with the required
frequency. Review of the procedure and its requirement to run
water through the cooling tower weekly led licensee management
to the conclusion that neither were required in the current and
anticipated modes of operation. The procedure was deleted with
the approval of the NSC. This portion of the violation is
closed.
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b. (0 pen) Violation 50-160/87-01-04, (Violatien 0) The licensee changed
.the cover gas specified in TS .3.6.e from helium to nitrogen without
first obtaining the appropriate change to the specifications. The
lice'nsee subsequently submitted a proposed- change to the.
specifications, but has yet to submit an acceptable safety evaluation.
to justify the change. Nevertheless, the licensee' continues to use
nitrogen as the cover gas. The NRC has requested additional
infornation regarding this proposal in a letter dated December 7,
1987.
c. Part B.1 of Violation 50-160/87-01-02 stated that, contrary to the
requirements of Technical Specification 6.3.a(1), estimates of
isotopic activities were not being entered in the space provided on
Request for Minor Experiment Approval. form. In their letter of
July 15, 1987, the licensee denied the violation with the argument
that it was not the intent of the form to require calculation of the
activities expected from all irradiations. Region II management
concurred in the licensee's position and deleted the violation in a
Region II letter of August 31, 1987. In the event of the week of
August 17, 1987, that same form was filled out inaccurately in that all
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activities were listed as "nil" when in fact they were not
(Paragraph 3.b). In further review of minor experiment forms
emoleted since August 1987, the inspector determined that many of
tit were filled out inaccurately. Only the activity of interest to
the experimenter was listed and the other activities, from other
isotopes of the same element, that would be produced were not
calculated in many cases. Clearly the form is not being used in a
manner that would promote safe handling of the irradiated package,
and there appears to be no other control imposed by tne licensee to
accomplish that end. This activity will continue to be reviewed
during subsequent inspections to further assess the safety
significance of providing the calculations.
The procedures reviewed, including those revised to respond to
violations, were barely acceptable. Only an experienced person could
hope to complete the procedures satisfactorily. They would be of
limited use in training new operators for the facility, particularly
those enge;ed in self study.
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