ML20236T066
| ML20236T066 | |
| Person / Time | |
|---|---|
| Site: | University of Missouri-Rolla |
| Issue date: | 11/23/1987 |
| From: | Greger L, Slawinski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20236T060 | List: |
| References | |
| 50-123-87-01, 50-123-87-1, IEIN-85-048, IEIN-85-081, IEIN-85-092, IEIN-85-48, IEIN-85-81, IEIN-85-92, IEIN-86-022, IEIN-86-024, IEIN-86-22, IEIN-86-24, NUDOCS 8711300276 | |
| Download: ML20236T066 (13) | |
See also: IR 05000123/1987001
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.U.
S. NUCLEAR-REGULATORY COM'iISSION
REGION III
Report No. 50-123/87001(DRSS)
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. Docket No. 50-123
License No. R-79
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' Licensee:
The Curators of the University
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of-Missouri - Rolla
Rolla, MO .65401
Facility Name:
University of Missouri - Rolla
' Nuclear Reactor Facility
Inspection At:
Rolla, Missouri
Inspection Conducted:
0ctober 28-30, 1987
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Inspector:
W. // Slawinski
' //- 13 - 77 -
Date
Approved By:
L.
e er, Chief
//-23967
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Facilities Radiation Protection
Date
Section
Inspection Summary
Inspection on October 28-30, 1987 (Report No.'50-123/87001(DRSS))
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Areas Inspected:
Routine, announced. inspection of operations, radiation
protection, and radwaste management programs, including:
records, logs and
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organization; review and audit; training and_requalification programs;
' procedures; surveillance; instruments and equipment; exposure controls;
material transfers; surveys; notifications and reports; and radwaste
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management.
Also reviewed were open inspection items and licensee response
to selected IE Information Notices.
Results:
One violation was identified (failure to follow laboratory rules
outlined in a Standard Operating Procedure - Section 7).
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8711300276 371124
ADOCK 05000123
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DETAILS
1.
Persons Contacted
- C. Barton, Senior Electronic Technician and Senior Operator
- A. Bolon, Ph.D., Nuclear Reactor Facility Director
- R. Bono, Health. Physicist and Director, Environmental Health / Safety
and Risk Management
- M. Straka,'Ph.D., Reactor Manager
- N. Tsoulfanidis, Ph.D., Campus Radiation Safety.0fficer
'*J. Williams, Laboratory' Mechanic
- Indicates those present at the exit meeting on October 30, 1987.
- Indicates those contacted by telephone on November 6, 1987.
'2.
General
This inspection, which began with visual observation of facilities and
equipment, posting, labeling, and access controls on October 28, 1987,
was conducted to examine the routine reactor operation, radiation
. protection, and radwaste management programs.
The inspector observed a
student participating in a reactor startup and power ascension, and
performed radiological surveys-(direct and smear) of various restricted
areas; no discrepancies from posted direct radiation survey readings were
noted.
No removable contamination was detected on any of the ten area
smears collected by the inspector.
3.
Licensee Action on Previous Inspection Findings
(Closed) Order (50-123/85001-01):
September 27, 1985 Order that nonpower
reactor licensees show cause why they should not be required to reduce
the quantity of highly enriched uranium (HEU) maintained onsite to the
amount necessary to maintain a normal schedule of operations.
The
' licensee does not possess any unirradiated HEU, they have several
standard TRIGA-MTR conversion fuel elements with an intent to convert
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to low-enrichment (<20% 2350) fuel.
These elements are in storage and
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consist of an aluminum guide piece, four stainless-steel clad TRIGA fuel
rods, and an aluminum handle.
Before loading the reactor with TRIGA fuel
or mixed MTR-TRIGA fuel, the licensee must prepare an appropriate safety
analysis and obtain authorization from the NRC.
This Order therefore does
not apply to the licensee; no further action is warranted.
(0 pen) Open Item (50-123/85002-01):
Reactor operator performance
evaluations performed by individuals other than specified in the
licensee's approved Requalification Training Program.
This matter is
considered unresolved pending NRR disposition.
See Section 6.
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4.
Organization, Logs; and Records
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. The facility organization was reviewed and verified to be consistent with
'the technical specifications and hazard analysis report. The minimum-
staffing requirements were verified to be present during. reactor operation
and fuel handling or refueling operations.
The n'uclear reactor facility
-staff consists of a Director, Reactor Manager, Maintenance Engineer,
Electronic Technician and Laboratory Mechanic.
These individuals, except
for the Laboratory Mechanic, are also set'or reactor. operators. .The
Director, Environmental Health / Safety and Risk. Management is the campus
health physicist and is. responsible for radiological safety at the
facility.
The health physicist is organizationally independent of the
reactor. operations group and reports to the campus radiation safety
officer.. A part-time health physics technician trainee (nuclear
engineering student) assists the campus' health physicist. .A Supervisor,
Hazardous Materials and Chemicals has been appointed but does not share
radiation protection responsibility as reported in Inspection Report
No. 50-123/85001.
The reactor is operated for'approximately 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> annually; the
majority-(70-75%) used for class instruction and training purposes.
About 20% of reactor run time is used for research related activities.
' Re' actor operations logs and records were reviewed to verify that:
a.
Records were available for inspection.
b.
Required entries were made.
c.
Significant problems or incidents were documented,
d.
The facility is being operated and maintained properly.
No violations or deviations were identified.
5.
Reviews and Audits
The licensee's Radiation Safety Committee is responsible for oversight of
reactor operations and assures the facility is operated in a manner
consistent with the requirements of the facility license and applicable
regislations.
The' committee is currently composed of seven voting and two
ex-officio members.
The members appear to possess adequate experience and
balanced knowledge of reactor operations, safety, and radiation protection
to ensure proper oversight of licensed activities.
A review of Radiation
Safety Committee meeting minutes for the period 1986 to present indicated
the committee is meeting all regulatory requirements,
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The licensee continues to implement a cooperative audit interchange with
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the Columbia Research Reactor.
Independent audits of reactor operations
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are typically performed by reactor operators and/or shift supervisors
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from the Columbia facility on an annual basis.
Audit reports for 1986
and 1987 were reviewed by the inspector.
Neither audit identified
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- significant concerns!"several' minor deficiencies including repetitive
items from the previous audit were reported.
The 1986 and.1987. audits
did not include.a review of radiation protection' activities nor are
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writtenLresponses to. audit' findings required.
The' desirability of
performing independent audits of reactor health physics activities,
and that future audit' reports be reviewed by all responsible 1 facility
personnel and written responses be made ' addressing audit findings / .
recommendations was discussed at the exit meeting and will be reviewed
further during a future inspection.
(0 pen Item:123/87001-01F
The licensee's review.'and audit program was examined by the. inspector to-
verify'that;
a.
Review of facility changes, operating and maintenance procedures,
design changes, and unreviewed experiments had been conducted by a
rafety review committee as required by the technical specifications.
b.
.The review committee is composed of qualified members and that-
quorum requirements'and frequency of. meetings had been met.
c.
Required safety audits had been conducted in accordance with
technical specification requirements.
No violations or deviations were identified.
6.
Training and Requalification Program-
a.
Indoctrination Training
All reactor-related personnel are given an indoctrination in
radiation safety before they assume their~ work responsibilities.
Additional radiation safety instructions are provided to those who
will be working directly with radiation or radioactive materials.
Students and new employees receive training pursuant to 10 CFR 19.12
by viewing a video tape.
Individuals sign a form attesting to their
understanding of the material presented in the tape.
The video-tape
was viewed by the inspector; no problems were noted.
In addition,
employees are provided and instructed to review Regulatory
Guides 8.13, " Instruction Concerning Prenatal Radiation
Exposure," and 8.29, " Instruction Concerning Risks From Occupational
Radiation Exposure," and the licensee's 50P-600, " Laboratory
Rules."
b.
Operator Requalification Program
The inspector reviewed procedures, logs and training records to
verify if the operator requalification program is conducted in
conformance with the licensee's NRC approved Requalification
Training Program dated March 11, 1981.
The requalification program is divided into three major areas as
described below:
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(1) A written examination developed and administered by the
licensee to verify an operator's knowledge level.
(2) On-the-job training to:
develop and enhance operator competence
.in manipulating plant controls and mechanisms' required by the
license; ensure cognizance of design, procedure and licensee
~ hanges; and to foster an understanding'of emergency procedures.
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(3) Observation and evaluation of operator performance to. actual
and simulated plant conditions.
An annual written examination is administered to all operators who
have had a reactor operator or senior reactor operator . license for
more than a year.
The examination contains questions covering ten
subject areas including principles of reactor operation' and theory,
radiation protection,.and operating characteristics.
Prior to 1986,
the requalification program considered an operator deficient in a
subject area if an exam grade of less than 80% was achieved.
The
miniraum acceptable cumulative grade for all exam sections was also
80%.
In 1986, the licensee modified the requalification program and
reduced the minimum acceptable exam grade to 70% for each section
and overall.
This change was made without prior NRC approval.
The
licensee reportedly performed.a 10 CFR 50.59 evaluation of this
change and determined it did not constitute and unreviewed safety
question 2
The requalification program is not described in the
licensee's Safety Analysis Report and therefore changes to it do
not appear to be within the purview of 10 CFR 50.59.
Since this change, two annual exams were administered to each of the
licensee's four reactor operators / senior operators.
In 1986, three
individuals scored less than 80% but more than 70% on at least two
exam sections.
In 1987, two individuals scored less than 80% but
more than 70% in one exam section.
No one scored an overall exam
grade.less than 80% in 1986 or 1987.
The acceptance criteria used
in the NRC certification examination is 70% in a section and 80%
overall.
Therefore, the licensee's 1986 and 1987 examination scores
meet NRC acceptance criteria and satisfy section 6.2 of
ANSI /ANS 15.4-1977.
The requalification program requires the Training Coordinator or
Reactor Supervisor (same individual) to conduct a semiannual
performance evaluation of all operators during one of their
reactivity manipulations.
The licensee recently modified this
portion of the program to allow the Training Coordinator, Reactor
Manager or their designee (but a licensed operator) to conduct the
annual performance evaluation.
This change, also made without
prior Commission approval, has not been fully implemented.
From
1986 to date, the licensee continued to perform semiannual
performance evaluations but plan to reduce these to annual should
their 10 CFR 50.59 evaluation show the frequency change not to
constitute an unreviewed safety question; however, four evaluations
were performed by senior operators and not the Training Coordinator
or Reactor Supervisor / Manager.
This latter issue was previously
identified in Inspection Report No. 50-123/85002 and tracked as Open
Item 123/85002-01.
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Pursuant to 10 CFR 50.54 (1-1), changes to NRC
approved operator requalification programs are not allowed without
prior Commission approval if the change decreases the scope of the
program.
It appears 10 CFR 50.59 evaluations are not applicable to
such changes.
This matter is considered to be unresolved pending
NRC determination whether the aforementioned changes decrease the
scope of the program.
This matter will be referred to NRR for final
resolution.
The licensee should abide by the requirements'of the
March 11, 1981 approved program until this matter is resolved.
(Unresolved Item 123/87001-01)
No violations or deviations were identified.
One unresolved item
remains open.
7.
Procedures
The inspector reviewed the licensee s procedures to determine if
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procedures were developed, approved and implemented in accordance with
technical specification requirements.
This review also verified:
a.
That operation and radiation protection procedure content is
adequate to safely operate and maintain the facility.
b.
That responsibilities were defined.
c.
That required checklists were used.
In response to a commitment to Region III in a letter dated March 5, 1984
(A. E. Bolon to C. E. Norelius), radiation protection procedures
.(600-series Standard Operating Procedures) have been upgraded where
necessary.
Most revised procedures were approved and issued in final
form in June 1985.
The licensee currently has sixteen radiation
protection procedures included in 600-series 50Ps.
50P-600, " Laboratory
Rules," revised June 19, 1985, outlines various precautions and rules to
ensure that all personnel abide by safe operating practices.
Item B(4)
of this procedure states, "Do not eat, drink, or smoke in the bay area
or counting room, or while handling radioactive materials." Contrary to
this rule, on October 28, 1987, the inspector observed a licensee staff
member smoking cigarettes in the reactor bay area.
The inspector was
later informed that drinking and smoking in the bay area was a routine
practice by this individual and an occasional practice by others.
Coffee
drinking in the bay area was also observed and reported during a previous
inspection (Inspection Report No. 50-123/80-03).
Item B(15) of 50P-600
states, "Always use the frisker station when leaving the bay area."
Contrary to this rule, individuals routinely bypass the frisker station
and egress the bay area through an alternate (emergency) exit.
The
inspector was informed that frisking is not required unless radioactive
materials were handled or contamination was suspect.
Failure to follow
SOP-600 rules is a violation of Technical Specification 6.3 which
requires that written radiation control procedures be prepared and
utilized (Violation 123/87001-01).
This matter was discussed at the exit
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meeting.
One violation was identified.
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8.
Surveillance Program
The inspector reviewed procedures,. surveillance test schedules and test
records, and discussed the surveillance program with responsible personnel
to verify:
a.
That, when necessary, procedures were available and adequate to
perform tests.
b.
That tests were completed within the technical specification required
time schedule.
c.
Test records were available and test results were within technical
specification limits.
Records of surveillance / tests results were selectively reviewed for the
period 1986 to date.
They included:
Reactor safety system channel checks.
Pool water resistivity determinations.
Ventilation system operability checks including visual checks of
fans and closure devices (inlet and exhaust duct louvers).
Area radiation monitor operability and setpoint verifications.
Control rod drop time measurements.
Calibrations of power range safety and period channels.
Technical Specification 3.5 requires a ventilation fan with a capacity
of at least 4500 cfm be operable when the reactor is at full power.
However, the capacity of the ventilation system fans have not been
verified since 1984.
The desirability of verifying fan capacities on
a periodic basis was discussed at the exit meeting and will be reviewed
during a future inspection.
(0 pen Item 123/87001-02)
No violations or deviations were identified.
9.
Instruments emi Equipment
a.
Portable Survey Instruments
The licensee appears to have an adequate supply of appropriate
portable survey instruments c pable of measuring beta, gamma and
neutron radiation.
Most portable survey instruments are calibrated
by the campus health physicist as authorized by NRC Material License
No. 24-00513-32.
Beta / gamma measuring instruments are calibrated
semiannually using a nominal 100 millicurie Cs-137 source.
High
range instruments (ionization chambers) are returned to the
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manufacturer for calibration.
Two neutron measuring instruments are
-typically calibrated by the manufacturer or response checked by the
licensee using a 5 curie Pu-Be source.
Calibration records for 1986
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to present were' reviewed; no problems were noted.
The inspector
examined several instruments maintained in the reactor facility;
each instrument was operable and had a current calibration sticker.
b.
Area Radiation Monitors
In accordance with technical specifications. the facility has
operable area radiation monitors located at the reactor bridge,
the demineralized, and in the experiment room.
Operability and
alarm setpoints are checked daily using internal check sources;
performance of these checks were confirmed by a selective review
of daily operational checklists.for 1987 to date.
In-situ monitor
calibrations are normally performed semiannually using the
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aforementioned Cs-137 standard.
c.
Frisking Station and Portal Monitor
A hand-held frisker is located at the reactor bay egress leading
into the counting room.
An avea posting and laboratory rules
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instruct workers to frisk prior to leaving the bay area.
Failure to
perform personal frisks is discussed in Section 7.
Frisker
operability is periodically verified using a check source;
calibration is performed using a Cs-137 source.
The licensee has one walk-through portal monitor located at the
counting room egress leading into the control room.
The monitor
contains nine conventional G. M. tubes (three on each side, one for
each foot, and one above the head) linked to an analog display with
audible alarm capability.
The alarm is not connected.
Personnel
normally walk through the monitor when leaving the counting room to
enter the control room or leave the facility.
The monitor is not
routinely
'"rce checked for operability nor is the alarm function
tested.
- ne nspector noted that most personnel pass through the
monitor without pausing and observing the analog display.
This does
not appear to be an acceptable practice for personnel contamination
detection, considering the alarm function is disconnected.
To
assure its effectiveness as a contamination detection instrument,
the monitor should be routinely source checked for operability, it's
efficiency determined, and alarm capabilities utilized and checked.
These matters were discussed at the exit meeting and will be reviewed
during a future inspection (0 pen Item 123/87001-03).
d.
Air Particulate Monitor
The reactor facility has one continuous air particulate monitor
(CAM) located in the reactor bay area.
The CAM is electronically
pulse checked periodically; however, alarm and source checks are
not normally performed.
The desirability of performing routine
operability checks on the CAM was discussed at the exit meeting.
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No violations or deviations were identified.
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10.
Exposure Controls
a.
External Exposure
'The personnel dosimetry services of R. S. Landauer, Jr. and Company
are utilized by the licensee on a bimonthly exchange basis.
All
nuclear reactor facility staff members are provided with whole body
film badges capable of detecting beta, gamma, fast and thermal
neutrons.
Self-reading dosimeters are prcvided to visitors,
temporary workers, and other personnel as warranted.
Extremity dosimeters have not been routinely provided for several
years.
One researcher is reportedly provided with TLD finger rings
during control rod inspections.
The licensee indicated that
extremity dosimeters (TLD finger rings) were routinely used several
years ago but discontinued because vendor analysis of the TLDs
showed no appreciable exposure.
The licensee has not established
specific procedures or guidelines which address usage of extremity
exposure monitoring devices.
The licensee should evaluate irradiated
sample and sealed source handling practices, and reactor maintenance
activities, to determine the need for extremity dosimetry use and
develop / implement guidelines for such use.
This matter vas discussed
at the exit meeting and will be reviewed during a future inspection
(0 pen Item 123/87001-04).
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The vendor's whole body dosimetry reports were reviewed for the
period January 1986 through September 1987 for reactor facility
workers.
All annual whole body exposures were less than 50 mrem.
Self-reading dosimeters are checked on an annual basis for drift and
response to a known cesium-137 radiation field.
No formal criteria
for considering a dosimeter unacceptable for use is.followed.
The
licensee was alerted to Regulatory Guide 8.4/ ANSI N13.5-1972,
" Performance Specifications for Direct Reading and Indirect Reading
Pocket Dosimeters for X-and Gamma Radiation," and indicated they
would consider adhering to this standard for future dosimeter checks.
b.
Internal Exposure
The licensee has no routine bioassay program.
They rely on
continuous airborne particulate monitor (CAM), gaseous effluent, and
reactor pool water samples to define any problems.
However, the
facilities CAM filter media is not analyzed or changed on a routine
basis; filters are changed after sufficient particulate buildup
significantly reduces air flow.
The licensee indicated that filters
have occasionally been analyzed for isotopic content ar4 no
significant activity, other then naturally occurring isotopes, was
detected.
During the exit meeting, the inspector discussed with the
licensee the desirability of collecting and analyzing filter media on
a routine basis.
This matter will be reviewed during a future
inspection (0 pen Item 123/87001-05).
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Pool water samples are collected every six months and analyzed for
tritium content.
The most recent tritium analysis of pool water
showed a concentration of about 3 E-6 uCi/ml; similar concentrations
were determined for the previous two samples.
Based on this
analysis and typical pool water usage (evaporation), less than
0.05 millicuries of tritium is released from evaporation of
reactor tank water per year.
The licensee does not have an approved (10 CFR 20.103) respiratory
protection program but maintains half and full face respirators and
SCBAs for emergency use.
No violations or deviations were identified.
11.
Material Transfers
Material'is occasionally irradiated in the reactor and transferred to
on-campus users (normally less than ten transfers per year).
Irradiated
samples are held at the reactor facility to allow short-lived isotopes
to decay, tnen surveyed (direct and smear) prior to being released to
users.
Samples for use in laboratories on the University campus are
transferred to NRC Byproduct Material License No. 24-00513-32.
Procedure
SOP-604, " Release of Byproduct Material on Campus," has been established
and provides guidelines for such transfers.
Records of transfers were
selectively reviewed for 1987; no problems were noted.
No violations or deviations were identified.
12.
Surveys
a.
Area Surveys
Direct radiation surveys and smear sampling are performed in various
reactor building areas on a monthly basis by the health physics
staff.
Special surveys are performed as needed to evaluate new or
changing experiments; thermal column and beam port experiments are
surveyed after any modifications.
The licensee relies on their area
monitors for indications of unexpected radiation levels.
The inspector reviewed records of smear and direct surveys from
June 1986 to date.
Smears are counted on the licensee's gas flow
proportional counter; significant removable contamination is rarely
detected.
No problems were noted.
b.
Air Samples
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The facility's continuous air particulate monitor (CAM) also
functions as an air sampler.
The desirability of performing routine
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operability checks of the CAM are described in Section 9.
Concerns
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regarding CAM filter media exchange and analysis are described in
Section 10.
No violations or deviations were identified.
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13.
Notifications and Reports
Review of records and discussion with licensee representatives indicated
no problems regarding comp 1.iance with 10 CFR 19 or 20 notification and
reporting requirements.
The inspector reviewed the following technical
specification required reports for timeliness of submittal and adequacy.
of information submitted:
a.
Progress Report 1985-86 Nuclear Reactor Facility.
b.
Progress Report 1986-87 Nuclear Reactor Facility.
No violations.or deviations were identified.
14.
IE Information Notices
The inspector reviewed the licensee's internal review of selected
IE Information Notices.
The licensee's evaluations and conclusions
are presented below:
Notice No. 85-48:
" Respirator Users Notice:
Defective Self-Contained
Breathing Apparatus Air Cylinders." The licensee does not possess the
subject a'ir cylinders for self-contained breathing apparatus.
Notice No. 85-81:
" Problems Resulting in Erroneously High Reading With
Panasonic 800 Series Thermoluminescent Dosimeters." The licensee does
not use Panasonic thermoluminescent dosimeters.
R. S. Landauer, Jr. and
Company film badges are used to monitor personnel exposures.
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Notice No. 85-92:
" Surveys of Wastes Before Disposal From Nuclear
Reactor Facilities." The licensee generates very little solid radwaste.
No solid radwastes are disposed of as normal " cold" trash.
See
Section 15.
Notice No. 86-22:
"Underresponse of Radiation Survey Instrument to High
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Radiation Fields." The licensee does not possess the subject radiation
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survey instrument.
Notice No. 86-24:
" Respirator Users Notice:
Increased Inspection
Frequency for Ceicain Self-Contained Breathing Apparatus Air Cylinders."
The licensee does not use the subject air cylinders for self-contained
breathing apparatus.
15.
Radwaste Management
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a.
Gaseous Radwaste
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The principle gaseous effluents produced are Ar-41 and neutron-
activated dust particles.
These are produced by the irradiation
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of air in the reactor pool water and, to a lesser extent, air and
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airborne particulate in the thermal column and other experimental
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facilities.
The air is swept from the reactor bay and experimental
areas by three exhaust fans located in the roof of the reactor
building.
Technical specifications require a ventilation fan with a
capacity of at least 4,500 cfm to be operable when-the reactor is at
full power.
The licensee collects a reactor bay air sample with the reactor at
full power on an annual basis.
The Ar-41 concentration in the
reactor bay at. full power was determined by the licensee to be less
than 2 E-8 uCi/ml.
The licensee evaluates airborne releases monthly
by reflecting operating times of building exhaust fans and reactor
power to.the annual measured air activity at full reactor power.
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The inspector reviewed monthly release data for 1986 to date.
The
maximum activity released was 139 mil 11 curies in May 1987 at a
concentration of less than 2 E-8 uCi/ml (less than 50% of technical
specification /10 CFR 20 concentration limits).
b.
Liquid Radwaste
All drains in the reactor bay and equipment areas lead to the
basement sump.
The largest volume of liquid radwaste is produced
by the regeneration of the demineralized.
Resin regeneration
are initia11y'discherged into two 300 gallon retention tanks and
allowed to decay.
The contents of the tanks are eventually released
to the basement sump, pumped to the mid-level sump, and released to
the sanitary sewer system if analysis shows that concentrations are
within 10 CFR 20.303 limits.
Licensee records show there were eight
liquid releases in 1986 totaling about 70 uCi (gross activity) in
2890 gallons.
In 1987 to date there were nine releases totaling
about 89 uCi (gross activity) in 4185 gallons; one release contained
cobalt-60 at a concentration of 2 E-5 uCi/ml.
c.
Solid Radwaste
The licensee generates very little solid radwaste.
The waste consists
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primarily of spent resins which are stored for eventual shipment to
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a licensed disposal agency.
Records showed that only one shipment
'
was made from 1986 to date and totaled about 2 uCi of Co-60/Cs-137
in three 55 gallon drums.
Solid short-lived radwaste is held for
decay and transferred to the Universities Columbia Campus for
incineration pursuant to the conditions of NRC Byproduct Material
License No. 24-00513-32.
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No violations or deviations were identified.
16.
Exit Meeting
The inspector met with the licensee representatives (denoted in
Section 1) at the conclusion of the inspection on October 30, 1987 and
summarized the scope and findings of the inspection.
The inspector also
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. - .
_. -
- - - _ _ _ - _ _ _ _ _ _ - - _ -
- e
'
' discussed:the'likely informational.. content of th'e inspection report with'
regard.to documents or processes reviewed during the inspection.
The
licensee did not identify-any documents or processes as proprietary.
In
- response to; certain matters discussed by. the inspector, the licensee:
a .'
Acknowledged the inspector's comments concerning independent audits
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(Section 5).
b.
' Acknowledged the inspector's comments concerning the operator
=
requalification program changes and that the issue was considered
l
unresolved (Section 6).
The licensee advised the inspector that
an adequate 10 CFR 50.59 evaluation had been performed prior to
. implementing the changes.
'
,
.e
c.
Acknowledged the. inspector's comments regarding failure to-adhere to.
50P-600, " Laboratory Rules" and that this represents a technical
specification violation (Section 7).
d.
Acknowledged the inspector's comment regarding verification of
ventilation system flow rates (Section 8).
g'
e.
. Acknowledged thel inspector's comments concerning operability
checks and alarm functioning of the portal monitor (Section 9.c).
- f.
Acknowledged the inspector's' comment regarding extremity monitoring
devices and the need to establish guidelines for their use
.(Section 10.a).
g.
Agreed that CAM filter media should be routinely changed and
analyzed (Section 10.b).
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