ML20127E124
| ML20127E124 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/12/1985 |
| From: | Barr K, Bell J, Dan Collins, Cook R, Cowgill C, Moslak T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20127E075 | List: |
| References | |
| 50-320-85-10, NUDOCS 8506240442 | |
| Download: ML20127E124 (13) | |
See also: IR 05000320/1985010
Text
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U. S. NUCLEAR REGULATORY COMMISSION
DCS No.
50320-850327
Report No.
50-320/85-10
Docket No. 50-320
License No. DPR-73
Priority
Category
C
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Licensee:
GPU Nuclear Corporation
P.O. Box 480
Middletown, Pennsylvania 17057
Facility Name:
Three Mile Island Nuclear Station, Unit 2
Inspection At: Middletown, Pennsylvania
Inspe.: tion Conducted: April 11 - May 17, 1985
Inspector -
/, [mk
[//MW
lyok,5 enc
ident Inspector (TMI-2)
da~te signed
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L /it l T&~
T. Moslak', Residtht Inspector (TMI-2)
date figned
k'
& //tlW
x-
L D11, Senior Radiation Specialist
c ate ligned
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JtDO-
c //</C
. Barr, Raciation Specialist
date signed
10. PAN
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D. Collins, Radiation Specialist
date signed
p
k/2 Als"
Approved By:
e
C.'fowgill, Chigf, TMI-2 Project Section
cite signed
Inspection Sumary:
Areas Inspected:
Routine safety inspection by site inspectors of plant
operations (long term shutdown) including licensee action on previous
inspection findings' quality assurance annual assessment; witnessing of fire
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brigade drill; operator training; review of licensee event reports;
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preparations associated with plenum assembly removal including modified fuel
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transfer system turnover, flooding of the deep end of the fuel transfer canal,
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polar crane inspection, installation of fuel transfer canal dam, inspection of
the fuel transfer canal prior to flooding, refurbishment and testing of the
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polar crane auxiliary hoist; removal of the plenum assembly; reactor building
entry by NRC to examine "A" D-ring; routine health physics ano environmental
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reviews and security. The inspection involved 417 inspector-hours.
Results:
Failure to comply with procedures regarding security practices.
8506240442 850618
ADOCK 05000320
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DETAILS
1.0 Ongoing Recovery Operations
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Routine Plant Operations
Inspections of the facility were conducted to assess compliance with
the requirements of the Proposed Technical Specifications and
Recovery Operations Plan in the following areas:
licensee review of
selected plant parameters for abnormal trends; plant status from a
maintenance / modification viewpoint, including plant cleanliness,
control of switching and tagging, and fire protection; licensee
control of routine and special evolutions, including control room
personnel awareness of these evolutions; control of documents,
including log keeping practices; radiological controls; and security
plan implementation.
Random inspections of the control room during regular and back shift
hours were routinely conducted. The Shift Foreman's Log and
selected portions of the Control Room Operator's Log were reviewed
for the period April 11 - May 17,1985. Other logs reviewed during
the inspection period included the Submerged Demineralizer System
(SDS) Operations Log, Radiological Controls Foreman's Log, and
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Auxiliary Operator's Daily Log Sheets.
Operability of components in systems required to be available for
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response to emergencies was reviewed to verify that they could
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perfom their intended functions. The inspectors attended selected
licensee planning meetings. Shift staffing for licensed operators,
non-licensed personnel, and fire brigade members was observed.
During plant walk arounds, the inspector noted improvements need to
be made in overall plant housekeeping.
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No violations were identified.
2.0 Licensee Action on Previous Inspection Findings
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(Closed)InspectorFollowItem(320/84-24-01): Adequacy of Radiological
Review for SDS standpipe pump disassembly and cutting in preparation for
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disposal.
Radiological Awareness Report (RAR)85-005, was initiated by the licensee
following the initial attempt to prepare and package one of two SDS pumps
for disposal. Radiological conditions developed during the work that
resulted in personnel and area (compactor room) contamination and the
establishing of the compactor room as a locked high radiation area. The
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jobwasstoppedandtheRadiationWorkPermit(RWP) terminated. A
critique was conducted by the licensee, a second radiological review and
RWP were completed, and the preparation and packaging of both pumps was
completed without further incident. The inspector attended the critique,
reviewed the processing of the RAR, including closeout action taken by
Radiological Engineering, Radiological Field Operations, and Radiological
Training, and determined that appropriate actions were taken by the
licensee. Consequently, this item is closed.
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(Closed)UnresolvedItem(320/84-09-03):
Sump Sanpling/ Transfer
Requirements.
In order to improve the representativeness of sump samples and refine the
accounting of radioactive releases, the licensee has been developing and
implementing improved sump sampling and sample analysis methods.
Toward
this end, the licensee has rewritten its sump pumping and sampling
procedure to better control liquid transfers and improve the
representativeness of samples.
The revised procedure requires
recirculation of controlled sumps prior to sampling and the sending of
samples offsite for enhanced sensitivity analysis. Sump recirculation is
accomplished through the use of temporary systems, but hard piping is now
being installed under an Engineering Change Authorization.
In
consideration of these actions, this item is closed.
3.0 Quality Assurance Annual Assessment of 1984
On April 25,1985, in accordance with Quality Assurance (QA) Plan
requirements, the licensee's QA department presented its yearly
assessment of the effectiveness of the QA program to senior management.
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The inspector attended this presentation. Managers of sections within
the QA department, including QA Engineering, Site Welding, Quality
Control, Operations QA, Site Audits and QA System Engineering, identified
their respective program's strengths and weaknesses, then made
recommendations for improving performance. A total of 21 recommendations
were made. Examples of the recommended actions included:
Improve controls for modification tie-in to existing systems by
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revising procedures to more clearly define responsibility for
control of tie-ins and turnover
Reduce the inconsistency in supplying welding requirements in Unit
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Work Instructions (UWI) by providing more training to responsible
personnel and closer interfacing between Recovery Operations and the
Maintenance Departments
Improve procedural compliance in properly completing the lifted lead
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electrical jumper and temporary mechanical modification log by
providing renewed emphasis in plant operations training
Reduce the redundancy and duplication of procedures by more
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critically reviewing procedures during the biennial review.
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The inspector determined that presentations were candid and that the
assessment was accurate and consistent with findings identified in NRC
inspections.
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4.0 Witnessing of Fire Brigade Drill
On April 15, 1985, the Resident Inspector observed an unannounced Fire
Brigade drill. The drill simulated extinguishing an electrical fire in
the Cable Room on the 305' elevation of the Control Building.
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The inspector considered the drill to be satisfactory in terms of the
quickness of.the brigade response and actions taken at the scene to
extinguish the simulated fire.
The Brigade Leader took charge
imediately and initiated actions that were considered appropriate and
effective.
Brigade members reported promptly to the scene with their
assigned equipment.
Personnel suited up in turn out gear including,
Self-Contained Breathing apparatus. Two additional shift personnel were
at the scene assisting the Brigade Leader. The drill was monitored by
three members of the Fire Protection Department and two members of the
Training Department. Subsequent to securing the drill, a critique was
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conducted at the scene by the monitoring team. The inspector determined
that the critique was thorough.
The inspector identified no major deficiencies in the conduct of the
drill.
5.0 Operator Training
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On April 24, 27, and May 17, 1985, the inspector evaluated Senior Reactor
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Operator (SR0) and Fuel Handling Limited Senior Reactor Operator (FHSR0)
training pertaining to systems to be used in defueling operations. On
the respective dates, the inspector attended presentations on the Vacuum
Defueling System, Canister Positioning / Storage Systems, and In-Vessel
Lighting System / Core Boring System.
The inspector detemined that these
presentations met Training department objectives by introducing the
students to the equipment that would be used for defueling.
The
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handouts, audio visual aids, and lesson plans had sufficient detail to
provide a general overview of how components of the defueling systems
would be arranged and there basic principles of operations.
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The inspector detemined that the references used for development of this
training have been restricted to final design reports, design drawings
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and testing documents. Accordingly, the inclusion of several key areas
in systems training has been incomplete. These key areas are operating
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procedures, administrative procedures, technical specifications, and
limits and precautions.
These aspects of the systems training are to be
incorporated as the respective documents are drafted and finalized.
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Through discussions with the licensee's Operating Training Manager, the
inspector detemined that more detailed presentations, addressing the key
areas, are scheduled to be made by Design Engineers from the Bechtel
organization in mid-June.
Subsequent to completing this classroom
training, SR0's and FHSR0's will attend Penn State University (PSU) for
reactor training. This training using the PSU Braezale Reactor is to
accomplish four objectives:
Simulate the collapse of fuel to provide the operator with a first
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hand observation of neutron population change and criticality safety
margin reduction
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Remove fuel from various regions of the core to simulate a defueling
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operation
Simulate a deboration until criticality is reached to emphasize the
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criticality safety at high boron concentrations
Load fuel rods into a specific area of the core to simulate loading
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fuel fragments into a canister submerged under water.
Following the PSU reactor training, three additional phases will be
conducted to complete the system's training.
These phases are mockup
training using defueling equipment on the Defueling Test Assembly located
in the Turbine Building, hands-on training using the Fuel Canister
Handling Bridge in the Reactor Building, and Integrated
Systems / Procedures Classroom.
The NRC will continue to monitor the licensee's training program for
qualifying SR0's and FHSR0's to perform core alterations.
6.0 Licensee Event Report (LER) Review
The inspector reviewed the LER listed below which was submitted to the
NRC TMI Program Office to verify that the details of the event were
clearly reported, including the accuracy of the description of the cause
and the adequacy of corrective action.
The inspector determined whether
further infonnation was required from the licensee, whether the event
should be classified as an Abnonnal Occurrence, whether generic
implications were indicated, and whether the event warranted onsite
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followup.
LER 85-06 dated April 25, 1985, addressed a licensee identified
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condition in which a containment isolation valve is maintained in
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the open position without an NRC approved procedure so authorizing
this position. The root cause of this administrative discrepancy
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was an apparent typographical error which resulted in the position
of the valve, RR-V-20E (an instrument isolation valve located on the
inboard side of penetration R-588 for the Reactor Building (RB)
cooling coils), being listed as " closed" vice "open" in Revision 1
of 2104-5.1.
The inspector determined that this event resulted from an
administrative error and did not involve an actual mispositioning of
the valve. Accordingly, the inspector concluded that the LER was
submitted to meet the administrative reporting requirements of
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10 CFR 650.73 and had no effect on system operation nor the health
and safety of plant personnel and the public.
7.0 Plenum Assembly Removal
On May 15, 1985, the plenum assembly was removed from the reactor vessel
and placed in storage, under water, in the deep end of the refueling
transfer canal (RTC). Prior to completing this evolution the licensee
had to complete the following pre-requisites:
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Startup testing on the modified Fuel Transfer System
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Installation and testing of the dam that would separate the shallow
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end of the RTC from the deep end
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Annual preventative maintenance on the RB polar crane
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Refurbishment and testing of the auxiliary hoist of the polar crane
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Flooding of the deep end of the RTC.
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Site NRC inspectors performed the following as a verification that the
prerequisites had been properly completed.
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Modified Fuel Transfer System (FTS) Turnover: At various times
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during the three week testing phase, the inspector discussed test
progress with the lead Startup and Test Engineer.
The inspector
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determined that some of the remedial actions that were taken during
the testing prior to completing a Return-To-Service (RTS) to plant
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operations included changeout of a winch assembly due to excessive
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vibration and noise, repositioning of a lifting eye on the canister
basket due to a mechanical interference, and various mechanical
alignments to provide smoother operation.
Prior to completing the
RTS, plant operators were trained on the operation of the FTS and
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had cycled the West FTS approximately 40 times without any
operational difficulties. The inspector observed the West FTS being
operated three times and noted that the winch drives, horizontal
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transfer mechanism and hydraulic upenders worked satisfactorily.
Subsequent to the testing, the inspector observed blind flanges
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being installed on the transfer tubes in the "A" spent fuel pool and
leak tested to verify that unacceptable leakage would not occur from
the deep end of the RTC (when flooded) to the "A" spent fuel pool.
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The inspector discussed with the Site Operations Director, the
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statusoftheitemsontheIncompleteWorkList(IWL),priorto
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system turnover. Through this discussion, the inspector determined
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that none of the items on the IWL would preclude the FTS from
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perfoming its intended function nor would they be detrimental to
safe operation. The FTS was turned over to the Plant Operations
department on May 14, 1985.
No violations were identified.
Flooding of the Deep End of the RTC: On May 14, 1985, following
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turnover of the FT5 and dam to the Plant Operations department, the
licensee commenced transferring approximately 64,500 gallons of
borated water from the Borated Water Storage Tank (BWST) to the deep
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end of the RTC.
Prior to beginning this evolution, the inspector
discussed valve line-ups with licensed operators, reviewed the
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applicable opereting procedure 4210-0PS-3254.01 (Operation of Canal
Fill and Tool Flushing System), and examined system drawings to
verify flow path. The inspector determined that the controlled
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procedure had been reviewed and approved by the NRC per the
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requirements of TS 6.8.2, that the BWST volume and chemistry
parameters were as required by the TS and that the RTC level
indication met the procedural requirements of 4210-0PS-3254.01.
No violations were identified.
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Polar Crane Inspection:
During the reporting period, the resident
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inspector reviewed the polar crane annual preventative maintenance
procedure 4220-PMG-3882.01, " Periodic Inspection of Polar Crane,"
and the completed data sheets for performing these annual
inspections of the polar crane. As a result of these inspections,
contacts in the main hoist high speed main contactors on the bridge
were replaced because they were found in a degraded state.
The slow
speed main hoist contactor had cracks in the insulating plastic
housing for the auxiliary contacts. The auxiliary contact assembly
was replaced.
The slow speed motor clutch for No. 4 bridge drive
had cracks in the commutator brush holders. The brush holders and
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brushes were replaced. A portion of the bridge rails located at
position No. 12, which is located near the parked position,
exhibited signs of flaking and/or motion.
The gap at this location
was measured at 11/16 inch with a 5/8 inch offset.
Since the crane
is controlled using a pendent switch, the licensee is making a
concerted effort to avoid leaving the crane in the parked oosition,
thus minimizing wheel engagement of the flaking rail.
As a prerequisite to plenum lifting, the resident inspector made an
inspection of the physical condition of the polar crane on May 10,
1985.
It was noted that the plastic tie wrap used to ensure that
the brake shoes of the bridge hydraulic brake do not contact the
drum was broken on the brake at the end opposite from the cab. This
broken tie wrap was replaced prior to the inspector leaving the
polar crane.
It was also noted that there was more trash on the
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crane than anticipated. The trash was picked up and brought from
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the polar crane by the licensee at the time the inspector left the
The licensee has generated internal memoranda to bring staff
crane.
awareness to the need for good housekeeping practices. While on the
crane, the inspector did note that the original bumpers and some
electrical testing equipment which had been previously stored on the
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polar crane had been removed.
Fuel Transfer Canal Dam: Prior to lifting the plenum from the
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reactor vessel, the licensee installed a metal dam across the deep
end of the fuel pool. This dam, which is nominally 5 feet high, is
sealed'by two adjacent inflatable rubber seals located around the
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peripheral sealing surface.
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The resident inspector made an RB entry on April 30, 1985, to
witness inflating of the seals and the hydrostatic test accomplished
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by putting borated water in the annulus area between the seals.
When the bottled nitrogen was energized to the seals, several leaks
were detected using a soap solution. These leaking joints were
remade and/or tightened to effect zero detectable leakage. However,
when the nitrogen gas was energized to the seals, the seals did not
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inflate. The seals could not inflate because inline check valves
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located at the dam seals were installed in a reverse flow direction
which preclude nitrogen gas from entering the seals. The resident
inspector witnessed the inline check valves being reoriented and the
subsequent satisfactory leak test of the affected joints using a
soap solution. When the annulus area between the seals was filled
with borated water, water inadvertently overflowed into a trough
located on the outside of each seal. Therefore, the zero leakage
requirement of the seals could not be determined.
During a subsequent RB entry on May 10, 1985, the resident inspector
was able to observe dry conditions in the trough area while a water
level was maintained in the annulus area between the seals.
Thus, a
zero leakage condition was observed prior to flooding of the deep
end of the fuel transfer canal.
Fuel Transfer Canal Inspection: On May 14, 1985, the resident
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inspector made an RB entry to examine the material condition of the
fuel transfer canal.
The deep end had some minor pieces of debris
which were removed at the time of the inspection. The shallow end
had substantial amounts of debris and other material which could
interfere with pumping and purification systems in the event that
the shallow eno of the pool required flooding.
Prior to flooding
the deep end of the pool to a level of engaging the dam, the
licensee removed materials that could encrouch on safe flooding and
subsequent cleanup of the shallow portion of the fuel transfer
canal.
Polar Crane Auxiliary Hoist: During the inspection of the polar
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crane, the resident inspector examined portions of the polar crane
auxiliary hoist which had been recently refurbished and load tested.
It appeared that the auxiliary hoist had been refurbished as
described by the licensee in their memorandum, Sumary of the Polar
Crane Auxiliary Hoist Mechanical Component Refurbishment, dated
April 17, 1985. The resident inspector inquired whether the clips
which attach the wire rope to the drum are covered in any of the
preventative maintenance examinations as the auxiliary hoist has the
capability of running out most of the wire rope within some portions
of the lifting envelope. The licensee stated they were in the
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process of incorporating surveillance checks on these clips in their
annual preventative maintenance inspection program.
Subsequent to
examining the auxiliary hoist, the resident inspector examined
accessory replacement parts for Amerigear flexible couplings. The
locking nuts used on the couplings were a crimped styled nut. The
locking feature appears quite satisfactory.
During the load testing of the auxiliary hoist, it appeared that the
Dillon Load Cell used was out of calibration as the actual load
indication was above the anticipated test load. Whiting
Corporation, the crane supplier, had authorized testing of the
auxiliary hoist to 65,000 pounds. However, after the questionable
indication on the load cell, another load cell was used. The second
load cell had claimed accuracy of 1%. Considering the load cell
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accuracy and the questions associated with the actual load which may
have been placed on the crane during the attempted first test, there
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was some concern expressed as to whether the 65,000 pounds load
limit imposed by the Whiting Corporation had been (or could be)
exceeded. The Whiting Corporation gave verbal authorization,
followed by a written authorization to exceed the original load test
of 65,000 pounds to 67,500 pounds or 135% over the 50,000 pounds
rated capacity.
After the load test, the licensee sent both Dillon Load Cells to
Morehouse Instrument Company for calibration. The load cell used
for the test (accuracy of 1%), was within the allowable accuracy
limits. The load cell which had indicated higher than anticipated
loads was found out of calibration with readings nominally 3.5% high
at the load test weight. The acclaimed accuracy for this load cell
was 10.5%.
A review of the documentation associated with the auxiliary hoist
load test reveals that no load greater than 66,000 pounds could have
been lifted and the 67,500 pounds one time load limit was not
exceeded. The NRC authorized use of the reactor building polar
crane's auxiliary hoist to loads no greater than 25 tons by letter
dated, April 22, 1985.
8.0 NRC Reactor Building Entry
On April 19, 1985, an NRC inspector made an RB entry to assess radiation
safety, industrial safety, general housekeeping and work activities in
the "A" D-ring. The inspection included the 349', 330', 323' and 309'
elevations inside the "A" D-ring, as well as several small intermediate
level platforms. A photo survey was made of the 349' and 308' elevations
to record the condition of work areas and equipment staged in surrounding
areas to be used in the characterization of any fuel in pressurizer
relief and safety valves (349' elevation), in the base of the
pressurizer, and in the horizontal runs of the pressurizer surge line
(308' elevation).
In addition, on the 308' elevation, observations were
performed of the planned access routes for equipment drops to
characterize the "J" legs of the steam generator. Verification was
performed of the installed winches on each elevation for personnel
rescue.
The entry was made under Unit Work Instruction (UWI) 4370-305A-85-0072,
ALARA review 51030 and Radiation Work Permit (RWP) 011431.
Supporting
paperwork was reviewed for accuracy and appropriateness of specified
clothing, dosimetry and respiratory protection. All requirements were
found adequate.
The inspector attended the required task supervisor's meeting for job
planning and scheduling on the day prior to the entry and the job
briefing and Red Tag Key (high rad beta) area briefings just prior to the
entry. All briefings were found to be complete, done in appropriate
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detail and accurately reflected conditions encountered.
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The protective clothing, safety equipment, dosimetry and respiratory
protection requirements were found to be conservative. A problem did
arise during the suiting up process related to the acceptability of the
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use of a small size MSA respirator mask with a powered air purifier
respirator (PAPR). A Breezer respirator (same protection factor as PAPR)
was substituted for the PAPR.
Following the entry, it was detennined
that the use of the small size mask with the MSA PAPR is acceptable.
The
confusion was determined to be the result of the technician
misunderstanding the acceptable letter codes for mask use with particular
respirators. NRC will review the technician training program which
addresses the issuance of respiratory protection equipment to verify its
adequacy.
(320/85-10-01)
The inspection of the "A" D-ring resulted in the following observations:
Safety equipment, including lighting, was adequate
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Radiation dose rates were confinned to be as stated on the RWP and
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recent surveys
Wires impeding access on the 340' elevation to the ladder on the
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340' elevation (320/85-10-02)
The step-off pad at the top of the D-ring ladder presents a fall
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risk (320/85-10-03)
There were sheets of mirror insulation on the 349' elevation inside
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the D-ring. The insulation slows work in the area and presents a
potential increased contamination problem. The inspector will
follow licensee progress on removal of the material.
(320/85-10-04)
9.0 Routine Health Physics and Environmental Review
a.
Plant Tours
The NRC site radiation specialists perfonned routine plant
inspection tours. These inspections included all radiological
control points and selected radiologically controlled areas.
Items
inspected included:
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Access control to radiologically controlled areas
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Adherence to RWP requirements
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Proper use of respiratory protection equipment
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Adherence to radiation protection procedures
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Use of survey meters including personnel frisking techniques
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Cleanliness and housekeeping conditions
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Fire protection measures.
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No violations were identified. However, during a routine plant
walk-through inspection, the inspector noted a worker in the
personnel hatch anteroom in a relaxed position (sitting, leaning
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back against the wall, head resting on shoulder, eyes closed).
The
inspector noted a second worker speak to the worker.
The resting
worker seemed to respond appropriately.
The inspector noted that
both workers in the anteroom wore protective clothing and was
infonned by a licensee representative that these workers are
required to be present at the personnel hatch in order to make an
immediate entry in case of an emergency in the reactor building
(Procedure 4000-ADM-3240.01, Revision 1). The GPU Nuclear
Corporation Radiation Protection Plan (1000-PLN-4010.01, Revision 0)
states in Article 2 - Responsibilities of Workers, Item 18, that
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workers are to " Assure a mentally alert and physically sound
condition for performing assigned work." The suitability of
workers' condition for perfonning assigned work will be reviewed
during future inspections.
(320/85-10-05)
b.
Measurement Verification
Measurements were independently made by the inspector to verify the
quality of licensee performance in the areas of radioactive material
shipping, radiological control, radiation and contamination surveys,
and onsite environmental air and water sampling analyses. The
inspector reviewed the radiological controls applied within the
plant. Appropriate postings, surveys, and controls were observed
during inspector tours during day-shift and off-shift hours.
No violations were identified.
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c.
Reactor Building Entries
The site staff monitored RB entries conducted during the inspection
period. The inspection activities included review of selected
documents and direct observations of RB entries. The following
items were verified on a sampling basis.
The RB entry was properly planned and coordinated to assure
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that task implementation including adequate As Low As Is
Reasonably Achievable (ALARA) review, personnel training, and
equipment testing.
Radiological precautions were planned and implemented including
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the use of a RWP and specific work instructions.
Entries 583 through 614 were conducted during this inspection
period.
On May 15, 1985, the licensee removed the plenum assembly from the
reactor vessel. The plenum was placed under about 5 feet of
shielding water in the deep end of the fuel transfer canal in the
RB.
Radiological Controls included precautions for personnel access,
fall back provisions and area monitoring during the lift.
The
highest dose rate at the lift station atop the "A" D-ring was
30 mR/hr, double the normal readings.
No personnel were in
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line-of-sight of the plenum at any time that it was out of the
Total man-rem was 3.033; latching, 1.095 man-rem; the lift
water.
and transfer, 0.710 man-rem; and, unlatching,1.228 man-rem.
The
highest dose rate recorded was 80 R/hr with the bottom of the plenum
directly above and about 4 feet from the detector.
Reactor Buil
airborne particulate radioactivity concentrations were 1.8 x 10'gng
uCi/cc, within the range noted during routine activities conducted
in the RB. The inspector observed the RB vent monitoring readout in
the Control Room during plenum removal.
Readings were within the
normal range.
No violations were identified,
d.
Radioactive Material Shipments
The NRC site radiation specialists inspected TMI-2 radioactive
material shipments during the inspection period to verify the items
listed below.
The licensee had complied with approved packaging and shipping
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procedures.
The licensee had prepared shipping papers, which certifieu that
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the radioactive materials were properly classified, described,
packaged, and marked for transport.
The licensee had applied warning labels to all packages and had
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placarded vehicles.
The licensee had controlled the radioactive contamination and
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dose rates below the regulatory limits.
Inspector review of this area consisted of (1) examination of
shipping papers, procedures, packages, and vehicles, and (2) performance
of radiation and contamination surveys of the shipments which were
inspected.
No violations were identified
10.0 _ Security
THESE PARAGRAPHS, CONTAINING SAFEGUARDS INFORMATION, NOT FOR PUBLIC
DISCLOSURE, ARE INTENTIONALLY LEFT BLANK.
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11.0 Inspector Follow Items
Inspector follow items are inspector concerns or perceived weaknesses in
the licensee's conduct of operation (hardware or programmatic) that could
lead to violations if left uncorrected.
Inspector follow items are
addressed in paragraphs 2, 9 and 10.
12.0 Unresolved Items
Unresolved items are findings about which more information is needed to
ascertain whether they are violations, deviations, or acceptable.
Unresolved items are addressed in paragraph 2.
13.0 Exit Interview
The inspectors met periodically with licensee representatives to discuss
inspection findings. On May 22, 1985, the inspector summarized the
inspection findings to the following personnel at the exit meeting:
J. Auger, Licensing Engineer
J. Byrne, Manager, Licensing TMI-2
W. Craft, (Acting) Manager, Radiological Field Operation
S. Levin, Site Operations Director
F. Standerfer, Vice President / Director, TMI-2
At no time during the inspection was written material provided to the
licensee by the TMIPO staff except for procedure reviews pursuant to
Technical Specification 6.8.2.
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