IR 05000320/1987003
| ML20214V008 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/27/1987 |
| From: | Bell J, Dan Collins, Cowgill C, Moslak T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214U977 | List: |
| References | |
| 50-320-87-03, 50-320-87-3, IEIN-86-023, IEIN-86-23, NUDOCS 8706110407 | |
| Download: ML20214V008 (11) | |
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U. S. NUCLEAR REGULATORY COMMISSION Report No.
50-320/87-03 Docket No.
50-320 Category C
License No. DPR-73 Priority
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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 Inspection C n ed: Aari) 1, M87s-May 8, 1987 Inspectors:
K-/a'7/27 T. Moslak', Resident Inspector (TMI-2)
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1, Se f r Radiation Spbcialist
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D. Coll'ns, a
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L S/27/77 Approved By:
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C. CowgT11, Chief, T11-2 Project Section dite Yigned Inspection Summary:
Areas Inspected:
Routine safety inspection by site inspectors of plant operations, defueling operations, the conduct of opp ations within the Reactor Building, surveillance testing, shipments of radioactive materials, and the implementation of radiological controls. Site inspectors examined the specific circumstances that surrounded two incidents resulting in personnel contamination. The first involved the " hot particle" contamination of a worker's moustache. The second involved the rupture of a hose in the
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three workers.
Results: No violations were identified.
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DETAILS
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1.0 Routine Plant Operations Inspections of the facility were conducted to assess compliance with the requirements of the 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; 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.
Inspections of the control room were performed during regular and backshift hours.
The Shift Foreman's Log and selected portions of the Control Room Operator's Log were reviewed for the period April 1,1987 through May 8, 1987. Other logs reviewed during the inspection period included the Submerged Demineralizer System Operations Log, Radiological Controls Foreman's Log, and Auxiliary Operator's Daily Log Sheets.
Operability of components in systems required to be available for response to emergencies was reviewed to verify that they could perform their intended functions. The inspector attended selected licensee planning meetings. Shift staffing for licensed operators, non-licensed personnel, and fire brigade members was determined to be adequate.
No violations were identified.
2.0 Defueling Operations During this reporting period, a variety of light duty and heavy duty tools were used to remove fuel assembly sections and place them in canisters.
The airlift tool was routinely used to fill the interstitial spaces in the partially loaded cans with smaller size debris. A damaged fuel assembly. B-6, was removed from its position using a hook tool and subsequently loaded into a canister.
Several " rocks" were removed from the top of the debris bed and placed in the canister loading funnel. An air driven chisel was then used to break these, allowing the pieces to fall through the funnel into the defueling canister. This initial use of the air chisel proved successful as all the available " rocks" were broken into a loadable size. During use of the air chisel the sealing surface on one of the canisters was damaged.
The extent and effect of the damage is currently being evaluated. A new tool, the fuel assembly Jack, was recently installed in the reactor vessel.
It is to be used to insert hydraulic rams between the lower grid and lower end fittings of partial length assemblies and hydraulically raise these assemblies.
To date, approximately 93,000 lbs. of material has been removed from the reactor vessel out of a total of approximately 300,000 lbs. The total weight includes the weight of the core, 207,000 lbs.; structural and absorber material, 78,000 lbs.; and the additional mass added by
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oxidation of core and structural material, and mass added by partial melting of the core support assembly and core former walls.
3.0 Monthly Surveillance On April 22, 1987, the inspector witnessed the performance of On/0ff Site Power Supply Checks (4210-SUR-3700.01). The tests are performed, pursuant to Technical Specifications 4.8.1.1.1, 4.8.2.1, and 4.8.2.2.1, once per seven days, to verify that each of the required independent circuits between the offsite transmission network and thc onsite Class IE distribution system are operable. Operability is demonstrated for the AC/DC busses by verifying correct breaker alignment and indicated power availability with tie breakers open between redundant busses. The inspector accompanied a Control Room operator during the performance of the surveillance. The inspector verified that the most current revision of the controlled procedure was being used, that the exceptions to the procedure were processed as required, and that test acceptance criteria were met.
The inspector noted that the only exception to the performance of this surveillance was that bus 2-36 was cross-tied to bus 2-46.
This was done following the failure of the 2-36 transformer. Cross-tieing was permitted per Special Operating Procedure (SOP) 4210-3730-87-61, effective February 19, 1987.
The S0P was approved by the NRC pursuant to Technical Specification 6.8.2.
No violations were identified.
4.0 Reactor Building Entry On April 10, 1987, the inspector entered the Reactor Building (RB) to evaluate the overall radiological and housekeeping conditions. The inspector accompanied a Radiological Controls Technician (RCT) while the RCT performed routine checks and filter changeouts of AMS-3 monitors and conducted radiological control coverage for the removal from the RB basement, and subsequent decontamination, of a Remote Reconnaissance Vehicle (RRV-1).
Specific observations were made of the following activities:
measurement of the steam generator, secondary side, water level
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changeout of particulate filters for AMS-3 monitors, located on the
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305' and 347' elevations removal of the shielded hatch (plug) on the 305' elevation, in
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preparation for removal of the RRV-1 from the RB basement removal of the RRV-1 from the RB basement and its subsequent
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decontamination washdown.
From these observations, the inspector determined that the activities were conducted in accordance with the procedures, radiation work permits (RWP) and unit work instructions (UWI), applicable to the task being
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performed. The inspector also determined that the sequence of operations agreed with that described in the pre-job briefings.
The inspector conducted several radiation measurements in work areas and near radwaste storage areas and determined that the radiation levels agree with licensee measurements.
The inspector identified what appeared to be a minor leak in a hose carrying water used for decontamination operations. The inspector informed operations personnel of this fact and they acted expeditiously to correct the situation.
In support of this entry into the reactor building, the inspector reviewed the RWP and determined that the protective clothing, respirators, and dosimetry specified were appropriate for the tasks being performed. The inspector attended pre-job briefings and determined that the appropriate personnel were present, and that the briefings were thorough and organized.
The inspector identified a weakness in the overall conduct of the operations in that the RCT was assigned two tasks. The RCT was to perform routine radiological surveys and /,MS-3 filter changeouts, and to provide health physics coverage for removal / decontamination of the RRV-1.
The inspector observed that after the RCT provided initial coverage for removal of the shielded hatch on the 305' level in preparation for RRV-1 removal, the RCT went to the 347' level to changeout a filter on an AMS-3.
Following the filter changeout, the RCT was to return to the 305'
level and provide coverage for removal /washdown of the robot. When the RCT returned to the 305' level, the washdown of the RRV-1 was underway and the RCT observed a maintenance technician in that area conducting an inventory of a tool locker.
Since the maintenance technician was not wearing the protective clothing / respiratory protection prescribed for the washdown area, the RCT ordered the technician to immediately leave the area. Had the RCT been assigned the single task of providing coverage for removing /deconning the RRV-1, he would not have permitted the technician to initially enter the washdown area.
No violations were identified.
5.0 Health Physics and Environmental Review a.
Plant Tours The NRC site Radiation Specialists performed inspection tours of the plant, including all radiological control points and selected radiologically controlled areas. Among the areas inspected were:
the Auxiliary and Fuel Handling Buildings, EPICOR-II, Radiochemistry Laboratories, radioactive waste storage facilities, the Respirator Cleaning and Laundry Facility, the Radiological Controls Instrument Facility and the Waste Handling and Packaging Facility.
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Among the items inspected were:
Access control to radiologically controlled areas
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Adherence to Radiation Work Permit (RWP) requirements
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Proper use and storage of routinely used respirators and
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associated equipment Maintenance and storage of emergency-respiratory equipment-
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Adherence to radiation protection procedures
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Use of survey meters and other radiological instruments.
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The inspectors reviewed the application of raciological controls during normal hours, on backshifts, and on weekends. Log books maintained by Radiological Controls Field Operations and Radiological Engineering to record activities in the reactor building and the balance of the plant were reviewed. All of the log books contained appropriate entries.
No violations were identified.
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Radioactive Material Shipments The NRC site Radiation Specialists inspected radioactive materials shipments on April 6, 16, and 27, 1987.
The inspector's review covered:
Verification that the recipient is appropriately licensed
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Verification of compliance with the 10 CFR 20.311 radioactive
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shipment manifest and tracking requirements Compliance with approved packaging and shipping procedures
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Proper preparation of shipping papers, including certification
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that the radioactive materials had been properly classified,
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described, packaged and marked for transport
Warning labels on packages and placarding of vehicles
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Compliance with regulatory limits for radioactive contamination i
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and radiation dose rates.
f The inspector's review consisted of (1) examinations of shipping papers, procedures, packages and vehicles, and (2) performance of radiation and contamination surveys.
i No violations were identified.
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Reactor Building Entries i
The inspector monitored the licensee's conduct of reactor building (RB) work during the inspection period.
The following were reviewed on a sampling basis during the inspection period:
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RB entries were planned &nd coordinated so as to ensure that
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ALARA review, personnel training, and equipment testing had been conducted.
Radiological precautions were planned and implemented,
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including: use of an RWP, locked high radiation area access authorization, specific work instructions, alarming self-reading dosimeters, and breathing zone air samplers.
Individuals making entries into the RB had been properly
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trained in emergency procedures, and possessed appropriate communications equipment.
Unique tasks were performed using specifically developed
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procedures, and mock-up training had been previously conducted where warranted.
The inspector observed radiological and industrial safety conditions in the reactor building on April 30, 1987. A crew was at work on the shielded work platform. The inspector noted that special notice was being taken by workers in the building of the tripping hazards associated with the temporary placement of the new 4-inch hoses for the Defueling Water Cleanup System. The hoses will be eventually mounted to their wall hangers.
The inspector noted that a temporary stairway arrangement had been placed on the lower open stairwell sections from the 305' to the 347' elevation. This arrangement provides for a second pathway from the 347' elevation during emergencies.
The licensee has also placed the emergency stretcher basket inside a fluorescent orange cover for contamination control and identification.
The licensee has also completed work to provide a safety railing around the elevator shaft roof. The inspector noted that the housecleaning and radiological conditions in the tool repair area were satisfactory.
No violation was identified.
d.
Removal of a Section of the Open Stairwell in the Reactor Building The inspector reviewed the preparations made for the removal of a section of the open stairwell in the Reactor Building. This task was to be performed in accordance with ALARA Review 70047 and Radiation Work Permit (RWP) 15884 on April 22 and 23, 1987.
ALARA Review 70047 was written to evaluate and control the radiation doses to personnel performing the task. Among the considerations used in forcasting doses were the physical layout and actions required by personnel. A full-scale mockup of the stairwell area was constructed on the site, and the assigned people practiced performing the task. Worker and engineer comments regarding the task were resolved, and a specific timetable was established under work control number 4730-3100-87-C1459 Revision 0.
RWP 15884 was issued to provide radiological controls and authorization to perform the work, i
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7-f On April.22, 1987, the first stair section, from the 305' elevation to the 29/'. elevation, was removed. Over the two days used to
< remove.and cuttup the stairway section, a total of 2.573-person-rem were used.-2.700 person-rem had been initially estimated.. The highest exposed individual wa's' the one who cut the stair section
' free on the 297'. elevation, he received 355 mrem in about four minutes.
. A ' ramp will be placed on the lowest stair section so that a robot vehicle. can access the' 282' area and permit examination _ of the conditions there. Tr4 robot exploration is tentatively set in mid-June 1987.
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No violations were identified.
e.
Defueling.Ilater Cleanup System Hose Break The inspector reviewed actions taken as a result of a May 3,1987 (0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />) break in a 3-inch diameter reinforced rubber hose line used in the,Defueling Water Cleanup System (DWCS). The hose was the pump dischapge line going to the filtering canisters and.the break was located at the upper edge of the canal cavity. southwest end at the 347' elevation. Total radioactivity concentration in the unfiltered primary water was 3.36 pC1/ml, about half Sr-90 and half
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Three individuals working on the defueling shielded work platform (SWP) at the 327! elevation were sprayed with the water (temperature about 75* F). The protective' clothing worn by these workers was i
only partially successful in protecting the individuals. Skin
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l contamination on one worker was easily removed. One worker had some residual contamination in his scalp and hair. The individual's
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scalp and hair were gradually decontaminated over the next few days.
l-Selective trimming of hair also reduced the contamination levels-of L
this individual'. The licensee's medical / radiological' advisors i
agreed with the non-aggressive course of actions for removing the L
. contamination.
The third worker was released after being initially E
decontaminated, but was found to have slight contamination the next l
day. He was subsequently decontaminated.
Radiological Engineering j
will perform'an assessment of dose received by the contaminated
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individuals.
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A critique was held at 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> on May 3, 1987 to evaluate the l'
causes of the incident, establish corrective actions, and set
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guidance for returning ~to work on tre SWP. Corrective actions included a reinspection of all hoses in the system and a survey and l
subsequent decontamination of the SWP. The inspection also included assessing possible damage to electrical components caused by the j
water spray. Surface contamination on the SWP was found to exceed normal contamination levels.
Contamination levels on the 347' and 305' elevations had reached about 100,000 dpm/100 cm2 after the hose
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I break. Alpha contamination was found to be about 1000 dpm/100 cm2 and 400 dpm/100 cm2 on the SWP and walkways, respecti.vely.
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A cleanup, inspection, and survey of the platform were performed prior to 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br />, May 3,1987. A radiological survey did ir,dicate loose contamination levels to be in the range of 10-15 mrad /hr; however, defueling was resumed at 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br /> and continued
until end-of-day. On May 4 Radiological Controls management restricted resumption of defueling until the SWP was decontaminated to normal working levels.
Decontamination of the SWP and the walkways on the 347' and 305'
elevations was completed on May 5, 1987 and defueling was then resumed.
The inspector will continue review of the event.
(320/87-03-01)
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Records Review The inspector reviewed selected radiological records during the period to assure the accuracy and completeness of the licensee's documentation of occupational exposure. The records reviewed included Radiation Work Permits (RWPs), Dosimetry Investigative Reports, TMI-2 Incident / Event Reports, Radiological Awareness Reports, and Dosimetry Exception Reports.
The inspector also reviewed various licensee records and periodic reports concerning the radiological controls program, including
- current data and trends in such areas as person-rem per RWP hour, decontamination status, skin contaminations, environmental monitoring, radiological events, whole body counting, training, dosimetry, shipments, progress toward achievement of goals and objectives, storage tank radioactivity content, airborne radioactivity, and person-rem by work category, effluent releases, (including sump releases and sources of sump contamination), and the cumulative dose (person-rem) to plant personnel.
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No violations were identified.
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Licenste Audits Th3 inspector reviewed a report of an audi.t of the licensee's t,diological controls program by an outside consultant, I ternational Technology Corporation. The report characterized the ogram as adequate, but makes some recommendations for improvement.
Y e licensee is acting on these recommendations and the inspector wilh evaluate any resulting changes to the program.
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Limited Transfer of Radiological Controls Activities
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The inspector noted that as of April 27, 1987 some radiological controls responsibilities had changed in the Unit 2 owner-controlled area. Unit 1 Radiological Controls has assumed responsibility for all radiological operations at the-Respirator and Laundry Maintenance Facility, including surveys on respiratory protective equipment, and incoming / outgoing laundry drum and vehicle surveys.
Unit 1 also has taken responsibility for radiological surveys for
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radioactive receipts at, and shipments, except rail casks, from TMI, the radiological instrument shop, and daily survey of non-contaminated compacted trash and non-compactible dump units within the Unit 1 owner-controlled area.
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Personnel Contaminations There were a total of 106 skin contaminations (i.e. more than 100 counts per minute above background using an Eberline HP-210 probe with an Eberline RM-14 rate meter) recorded during this period. Of this total, 12 resulted in skin doses (calculated) of 10 millfrem (mrem) or more with all but two being 25 mrem or less. The two highest doses were calculated by the licensee to be 412 mrem and 1.2 rem.
The skin contamination rate (number per hour worked under a radiation work permit) has increased significantly in 1987 compared to 1986.
The increase may be attributable to increased work with highly contaminated equipment associated with defueling activities.
The inspector reviewed the circumstances associated with ten personnel contaminations (nine in the reactor building) occurring in the period January through April of 1987. The licensee's actions were adequate in nine of the ten cases reviewed and were consistent with written procedures and generally accepted good radiological safety practices. One case is under evaluation as noted below.
No violations were identified.
Evaluation of the 1.2 rem Dose Event The 1.2 rem dose event is currently under evaluation. Contamination of a small area of the face of a worker was discovered as he passed through a portal monitor after exiting from a contaminated area at the reactor building personnel hatch anteroom.
The licensee determined by means of a detailed whole body survey (frisk) that the worker had contamination on his face and a much smaller amount on his chest.
It was further detennined that the contamination on his face was limited to a very small area of his moustache.
Using generally accepted means of personnel decontamination, the licensee effectively and efficiently decontaminated the worker.
The contamination (characterized as a single particle) on the moustache was flushed away during the decontamination process. The licensee " mocked up" the contamination by obtaining a sample of radioactive surface contamination from the area in the reactor building where the contamination found on the worker apparently originated. Survey meter readings of the sample anJ the individual's contamination were compared and a radiochemical analysis of the sample was performed.
Based on these data, the licensee calculated an average dose to one square centimeter of the skin of the worker to be 1.2 rem.
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The licensee's evaluation of the circumstances leading up to the contamination determined that the worker's contamination may have occurred as he was assisting other workers in the removal of contaminated boots at the reactor building personnel hatch.
It was on this basis that the types and amounts of radioactive contamination and the radiation dose to a small area of the worker's skin was determined. The licensee has since conducted additional surveys of the area in which the worker carried out his activities and found " hot particles" with a radionuclide composition significantly different from that of the sample on which its dose-calculations were based. Because of this and other considerations affecting the determination of the dose and its relation to applicable dose limits, this item is unresolved.
(320/87-03-02)
Inspector Findings The inspector noted that the worker's contamination occurred March 12, 1987; and that licensee recognition of the event as having the potential for an overexposure occurred several days later.
The circumstances of the contamination were not usual in that the contamination was apparently not in contact with the skin and the radiations from the contamination were apparently attenuated by the worker's moustache.
Because of this, well established and previously used routine methods for determining the dose from contamination on the skin were not directly applicable. As a result, the dose determination task was relatively difficult.
However, the attention recently directed to the hot particle problem (NRC IE Information Notice 86-23) and the recognized potential for this type of occurrence should have resulted in the licensee being prepared to deal with the problem more expeditiously. Specifically, Procedure 9200-ADM-4330.02, " Personnel Contamination Monitoring and Decontamination", although recently revised (effective March 12, 1987), does not provide guidance for assessing dose under the condition's experienced in this case. The licensee is considering revisions to the procedure to provide more effective guidance in performing dose assessments. The inspector will follow licensee efforts in this matter.
(320/87-03-03)
The worker's contamination was detected and evaluated as he passed through a portal monitor at a Radiological Controls control point after leaving his work area. Although a whole body frisk was accomplished by an attending Radiological Control Technician at this time, the worker did not perform a whole body frisk "immediately-upon exiting a posted contamination area..." (his work area) as is required by Procedure 9200-ADM-4330.02. Subsequent observations by the inspector confirmed that workers do routinely perform whole body frisks utilizing equipment provided specifically for this purpose at the exit from the work area. The failure to perform a whole body frisk upon exiting from the contaminated area was determined to be an isolated occurrence and to have had an insignificant effect on the worker's dose.
The inspector will continue to observe licensee practices in this are [
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Licensee Corrective Actions The inspector noted that the licensee is taking several steps to define the extent and magnitude of the hot particle problem, to prevent personnel contaminations, and to mitigate the effects of contaminations should they occur. These efforts include additional training of both workers and radiological controls personnel with respect to the detection, handling and analysis of particles, and improved work practices; surveys with adhesive paper for particles on plant surfaces; the definition of " hot particle areas" and the use of respiratory protective equipment in these areas; increased emphasis on surveying of
" clean" protective clothing to locate " hot particles" before they may be transferred to those wearing the clothing; frequent frisking of exposed skin areas of workers in areas in which hot particles are likely to be present; and the provision of quantitative guidance (criteria) to radiological controls personnel to aid in judging the seriousness of a contamination and to help ensure prompt initiation of dose assessments by Radiological Engineering. The inspector will continue to follow licensee actions in these areas, i
6.0 Inspector Follow Items
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Inspector follow items are inspector concerns or perceived weaknesses in the licensee's conduct of operations (hardware or progransnatic) that could lead to violations if left uncorrected.
Inspector follow items are addressed in paragraphs 5.e. and 5.1.
7.0 Unresolved Items Unresolved items are findings about which more information is needed to ascertain whether they are violations, deviations, or acceptable. An unresolved item is addressed in paragraph 5.1.
8.0 Exit Interview The inspectors met periodically with licensee representatives to discuss inspection findings. On May 15, 1987, the site inspectors summarized the inspection findings in a meeting with the following personnel:
C. Dell, Licensing Technical Analyst L. Edwards, 00A Monitor S. Levin, Defueling Operations / Support Director W. Potts, Acting Site Operations Director R. Rogan, Director, Licensing and Nuclear Safety J. Tarpinian, Radiological Engineering Manager D. Turner, Director, Radiological Controls, TMI-2 At no time during the inspection was written material provided to the licensee by the TMICPD staff except for procedure reviews pursuant to Technical Specification 6.8.2.