IR 05000320/1986015
| ML20211G042 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/11/1987 |
| From: | Bell J, Dan Collins, Cowgill C, Moslak T, Myers L NRC, NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211F997 | List: |
| References | |
| 50-320-86-15, NUDOCS 8702250260 | |
| Download: ML20211G042 (17) | |
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U. S. NUCLEAR REGULATORY COMMISSION Report No.
50-320/86-15 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 cted: eqecember 9,4986 - January 9, 1987
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^L / t B/7'7 Inspectors:
T. Mosfa'k,' Resident Inspector (TMI-2)
date sfgned J W GrGs -
2h/r7 Be
, Senior Radiation'5pecialist ddt'e/ sighed
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Th Yl D. Co lins, Radiation Specialist da'teAiigned
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(%su f4bitn 2/9/87
'L'./Myers./Radiati4nSpecialist date s1gndd
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Approved By:
Q1/ E 2/n/P7 C. C6wgi Chief, TMI-2 Project Section date signed Inspection Summary:
Areas Inspected:
Routine safety inspection by site inspectors of plant operations, defueling operations, defueling canister closure bolt undertorquing, radiological instrument calibrations, external exposure control and dosimetry, the unplanned internal exposure of a technician, and reactor building entries.
Results:
One violation was identified in that an evaluation of the hazards associated with entry to a contaminated area were not suitably evaluated (paragraph 8.0).
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0702230260 070217 PDR AuocK 05000320 G
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DETAILS 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 December 9, 1986 through January 9, 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 The licensee is continuing efforts to identify tooling and techniques which will facilitate defueling. Since completing the core drilling operation (on November 14,1986) no significant increase in the efficiency of removing fuel has occurred.
Several relatively large pieces of core material are hindering " cess to the drilled material.
These large rocks are believed to have rallen into the central region of the core from the undrilled 2-foot periphery of the core.
Defueling operators have identified highly localized regions of the fragmented bed which can be penetrated and have directed their efforts to using the spade bucket tool to remove material from these regions. The defueling rate has ranged from 3 to 4 canisters per week.
To improve water clarity, the Defueling Water Cleanup System (DWCS) has been modified to incorporate the use of organic polymer coagulants and body feed (diatomaceous earth) injection in parallel with DWCS filtration. Ongoing operations indicate that this modification substantially improves water clarity with a significant increase in DWCS filter life. The improvement in underwater visibility is expected to enhance defueling operations and assessments of core conditio '.
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The licensee is to install and operate a new defueling system, the. Air Lift System, in mid-January. The system works on the principle of an air eductor into which core debris will be drawn through a nozzle and piping system to a defueling canister.
The inspector will continue to closely follow licensee defueling activities.
3.0 Defueling Canisters On December 12, 1986, the licensee submitted an application to the NRC requesting that Certificate of Compliance No. 9200 for the Model No.
125-B shipping cask be amended to authorize the shipping of four (4) fuel canisters whose closure head bolt torque values were within the range 40 to 60 ft-lbs. rather than the previously approved range of 50 to 60 fuel canister that had one (of eight) quested authorization to ship one ft-lbs.
In addition, the licensee re closure bolts improperly seated.
Based upon a safety analysis, submitted with the application, that concluded that reducing the minimum specified torque from 50 to 40 ft-lbs. should have no adverse effect on the ability of the canister to remained sealed during normal and accident conditions of transport, the NRC approved that portion of the application.
The NRC did not approve shipping the canister that had an improperly seated lid bolt. The licensee subsequently removed and properly reseated the bolt prior to shipment.
Site inspectors evaluated the circumstances that led to the improper condition of the lid closure bolts to determine if a violation of regulatory requirements occurred.
Through interviews with licensee representatives and examination of selected records, the inspector determined that the nonconforming conditions were identified during the normal process of reviewing the Canister Vessel Traveller Data Package for the fuel canisters by Operations and Quality Assurance personnel prior to their shipment offsite.
As a result of this review, two Material Non-Conformance Reports (MNCR Nos. 246-86 and 247-86) were generated identifying the bolts' conditions and two Quality Deficiency Reports (QDR No. 124-86 and 125-86) were initiated that identified the administrative deficiencies that contributed to the condition. The inspector examined the MNCRs and QDRs. The inspector determined that the MNCRs/QDRs specified the appropriate level of site management to resolve the issues that lead to this condition.
The inspector examined the Safety Analysis (SA No. 4240-3252-86-255) that addressed the significance of the undertorqued condition of the bolts, and determined that the analysis was performed, reviewed, and approved by the appropriate cognizant engineers and management.
In resolving the MNCRs/QDRs, licensee management conducted a critique to establish the root cause resulting in the nonconforming conditions and to identify corrective actions that would be taken to prevent a recurrence.
Through interviews with licensee representatives, the inspector determined that the bolts were torqued to the lesser values as a result of miscommunication on the part of radwaste operations personnel resulting in a failure to follow the procedure (4215-0PS-3255.01) that details requirements for preparation of canisters for shipment. As corrective action, licensee management counselled the responsible individuals and established clearer lines of communication to preclude a
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recurrence. The effectiveness of the corrective actions will be evaluated in a future NRC inspection.
(320/86-15-01)
The inspector determined that the circumstances leading to the condition of the closure head bolts were contrary _to Technical Specification 6.8.1 that requires operating procedures be implemented. However, the violation meets the criteria of 10 CFR 2 Appendix C for not. issuing a Notice of Violation in that it was identified by the licensee, it was of minor safety significance, it was reported to the NRC, long term measures are being implemented to prevent a recurrence, and it has not been a recurring problem. Accordingly, a Notice of Violation will not be issued.
The inspector had no further questions regarding this matter.
4.0 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, and the Radiological Controls Instrument Facility.
Among the items inspected were:
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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 radiological controls during normal hours, on backshifts, and on weekends. Log books
maintained by Radiological Controls Field Operations and
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Radiological Engineering to record activities in the reactor
building and the balance of the plant were reviewed. All of the log l
books contained appropriate entries.
l No violations were identified, b.
_ Radioactive Material Shipments The NRC site Radiation Specialists inspected radioactive materials
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shipments on December 10, 14, 19, and 22, 1986, and January 7 and
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-11, 1987, including observation of the preparation and dispatch of a railcar-mounted cask shipment of Unit 2 core debris.
The inspector's review covered:
Compliance with appreved 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, 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
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and radiation dose rates.
The inspector's review consisted of (1) examinations of shipping papers, procedures, packages and vehicles, and (2) performance of
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radiation and contamination surveys.
j No violations were identified, c.
Reactor Building Work Reactor Building Entries The inspector monitored the licensee's conduct of reactor building (RB)workduringtheinspectionperiod. The following were reviewed on a sampling basis during the inspection period:
RB entries were planned and 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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informed, trained, understood 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 conducted where warranted.
A radiation specialist entered the RB during entry no. 1133 on December 22, 1986, to assess radiological and industrial safety conditions. The inspector accompanied a Radiological Controls Technician performing routine daily surveillance activities in support of RB entrie *
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4 General housekeeping in all areas observed on the 305' and 347'
levels, and the defueling platform was satisfactory. The inspector j
noted some minor trip hazards, e.g. a loose drain shield and an
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unsecured and unmarked plastic hose on the floor of the 305' level F
near the "A" D-ring. These were brought to the attention of the licensee and corrective actions were initiated. Various
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-surveillance and filter changeout activities associated with AMS-3
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air monitors were observed and were performed in a satisfactory
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manner.
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The inspector examined the temporary, moveable shield on the 347'
level that fell on a worker December 16, 1986. The inspector noted that a section of the temporary reactor coolant filter shield was unstable. The temporary shield stabilizers did not appear to be effective. As noted in another section of this report, the inspector will follow licensee efforts to improve shield wall stability.
Observed operations and facilities were found to be consistent with applicable Safety Evaluation Reports and implementing procedures.
As of the end of the inspection period, 1,151 entries had been made into the reactor building.
No violations were identified.
5.0 Personnel External Exposure Control and Dosimetry Program a.
Introduction and Inspection Scope The in:,pector reviewed selected portions of the licensee's program for measuring, controlling, recording, and reporting of personnel dose from external sources of radiation. Major areas covered included:
staffing; training; organization; equipment and facilities; procedures; records; and the specification, issue, processing, and use of personnel monitoring devices.
The licensee uses three types of personnel monitoring instruments to measure exposure to gamma, beta and neutron radiations:
thermoluminescent dosimeter " badges" (TLD), self-reading " pocket chamber" dosimeters (SRD) and " digital" or " alarming" dosimeters.
The SRDs and digitals are used as "real time" monitors of exposure while on the job. Also, the SRD readings are used as the dose of record until the TLD is read (normally once per month); at which time the individual's dose of record becomes that determined from the TLD.
b.
Organization Personnel external exposure control 'and dosimetry are accomplished within the Radiological Controls Department as shown in Figure 1.
" Radiological Controls Organization".
The Manager, Radiological Controls TMI-2 reports to the Vice President for Radiological and Environmental Controls.
The licensee has requested a technical specification change that includes the assignment of the
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RAD 10 LOGICAL CONTROLS ORGANIZATION ac.ager, aanoi tc46 wtrois r m o strater i
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- Techalcal Specification
- Effluent Asselsent Cualification Progres Progem Surveillance
- Trend Monitoring and Analysis
- Raclological Controls Prcfesslor.41
- Aut mated Radiological Controls Records * Control of Radiation herk
- Exposure Re&ction Progras (AUAA)
- Radioactive Waste $tre e Analysis Er.hancerent fralaine manaseent Systes
- Surveys:
- Ce M ral Ersleyee Ra3! alogical
- Resolratory Protection Progran
. Cirect Radiation
- Tomorary Shleidlag Progres controls tralaleg overstent Coordination / Supervision
. Atroorne Radlosctivity
- Shleiding Analysis
- IhPQ Accrecitation Radiological e Radiological controls Instrunent
. Surface contaalnation
- 4411ty Assurance /@ality control Prctection Areas Oversignt Calibretten Repair. Malatenance
- Erergency Response
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Accountability
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Radiological Health group's functions to Unit 1.
The group will continue to serve both units.
c.
External Exposure Control Radiation exposures are controlled, in part, through the use of the radiation exposure monitoring (REM) system, administrative exposure limits, and the process for reviewing and approving exposure limits in excess of routine administrative limits (" dose extensions"). The inspector observed technicians utilizing the REM system video displays to check worker's accumulated exposures against applicable limits and against estimated exposures for tasks covered by RWPs as a basis for signing workers on to RWPs. The inspector also examined records of dose extension requests and interviewed radiological engineers involved in the approval process. The interviews established that the radiological engineers were aware of the ALARA concept implications of dose extensions.
However, the inspector noted that statements in the justification section of the dose extension request form sometimes simply stated the need for additional dose rather than explaining why a particular individual should be allowed to receive additional exposure.
The inspector observed the process of estimating task-specific exposures in work areas, the specification of the use of digital alarming dosimeters with settings consistent with exposure estimates, and the use of these dosimeters to maintain awareness of accumulated personnel exposure. Those equipped with the dosimeters were observed to read them frequently and thus monitor their accumulated dose.
It was also noted that the readings correlated well with SRD and TLD readings.
Thelicenseehasinitiatedtheuseofthicker(7 millimeter)
compound face shields for those workers not wearing respirators, and supplementary lenses for those wearing respirators to improve worker protection against high energy beta radiation. The dosimetry program has been adjusted to accommodate this procedure.
For those using the thicker face shields and lenses, the whole body / eye lens exposure will be assessed at an equivalent tissue depth of 10 millimeters.
SRDs used by such workers have been modified to increase their wall thickness so as to give readings consistent with the TLDs. Workers assigned special dosimetry sets will be assessed whole body / lens of the eye exposures based on a deep dose read (1000 mg/cm ) of their TLDs. Additionally, workers issued a special a
dosimetry set will be required to use the augmented eye protection.
The inspector will review the coordination of special dosimetry sets with the augmented eye protection.
Area posting and access control, and labeling of containers of radioactive material were observed to be consistent with licensee procedures and regulatory requirements.
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d.
Personnel Monitoring Requirements Requirements for personnel monitoring are specified in radiation work permits (RWP) by Radiological Controls Field Operations personnel.
The inspector examined the process and bases for determining personnel monitoring requirements for incorporation into RWPs, including the availability and use of radiation survey results and other information on work site conditions; and the evaluation of proposed tasks and the specification of personnel monitoring by Radiological Engineering in "ALARA reviews". The inspector's examination included the review of records and the interviewing of managers, supervisors, radiological engineers and technicians involved in the process.
It was noted that Radiological Engineering is continuing to equip some individuals with nine TLDs distributed over his/her head, trunk and extremities as a means of determining the highest exposure locations and, therefore, the best location for a single "whole body" TLD.
No unacceptable conditions were identified.
e.
Dosimetry The inspector examined the TLD calibration, issue, collection and processing system, including personnel radiation exposure records.
The inspector noted that the licensee has acquired improved processing equi) ment; incorporated TLD glow curves into the dosimetry data )ank; and made algorithm modifications to change the minimum assessed deep dose from 10 mrem to 1 mrem. The dosimetry system was re-accredited under the National Voluntary Laboratory Accreditation Program in October 1986.
The inspector examined dosimetry records for selected individuals, including current exposure (NRC Form 5) and exposure histories (NRC Form 4). The licensee esteblishes a history for each individual who is issued a TLD for monitoring of occupational exposure and updates the history as a basis for approval of increases in an individual's administrative or regulatory limit. Also examined were records of
" Dosimetry Investigation Reports" involving such occurrences as unusually high TLD readings, differences of more than 25 percent between SRDs and TLD readings, and lost TLDs.
The inspector also examined training records for the two utility technicians and the contractor technician who were qualified since the last programmatic inspection.
Proceduro 9000-ADM-2622.05, "RST Training Procedure", includes a detailed outline of the training to be received by each Radiological Support Technician (RST).
During the review the inspector noted that training session attendance sheets were not specifically correlated with the training outline subjects; however, interviews of technicians established that the required training had been received.
Contractor technicians are typically restricted to specific, limited areas of work and are trained only in these areas.
The TLD system operation, maintenance and calibration were examine '.
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Licensee personnel were interviewed, procedures were reviewed and the operation of equipment was observed. A TLD calibration irradiation using the licensee's 1.2 curie cesium-137 source was observed.
TLD exposures are determined from the calibrated source strength and irradiation time; the time being controlled by an electronic timer and checked with a secondary timer. The magnitude of each exposure is confirmed by means of a radiation measurement device, typically a Victoreen Radocon.
Records of source and Radocon calibrations were examined. The calibration facility was operated and equipped in accordance with the applicable procedure and with regulatory requirements and all records were found to be in order.
No unacceptable conditions were identified.
6.0 Radiological Instrument Calibration
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The inspector reviewed the radiological instrument calibration program as it applies to portable radiological instruments and self-reading dosimeters. The program was reviewed against criteria in the regulations, the Corporate Radiation Protection Plan, appropriate regulatory guides, and industry standards.
In addition, licensee reports, including audits and reviews were examined.
The Radiological Instruments Group is the organizational unit responsible for calibration of all portable radiological instruments, self-reading dosimeters and certain Technical Specification surveillance instruments used at Units 1 and 2.
The Radiological Instruments Group reports to the Manager, Radiological Health who reports to the Unit 2 Radiological Controls Director. See Figure 1 " Radiological Controls Organization".
Radiological Instruments utilizes a computerized tracking system to track those instruments for which they are responsible. A weekly report is issued to instrument users advising them that certain instruments will require calibration within the next two weeks.
If an instrument requiring calibration is not at the calibration facility on or before the calibration date, an overdue report is sent to the user. The licensee audit of April 30, 1986, determined that Radiological Instruments did not pursue the overdue instruments to determine their disposition. Many of the overdue instruments were broken, lost, or otherwise unaccounted-for.
As a corrective action, Radiological Instruments has investigated the overdue instruments and purged lost and broken instruments from the list of overdue instruments.
The groups responsible for these instruments must now verify and document the disposition of each instrument to Radiological Instruments. This practice has been incorporated into the recall procedure.
During a tour of the spent fuel handling building on December 11, 1986, the inspector noted that an HPI area monitor was in use and overdue for calibration (due December 6, 1986). The inspector also noted that the instrument is source checked weekly.
The inspector informed the licensee of the overdue instrument and they removed it from service.
Subsequent calibration checks identified that the instrument was within calibration
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limits.
Inspector review identified that in April 1986, NRC inspectors identified a similar circumstance that was reported in Inspection Report 50-320/86-05.
Radiological Controls Field Operations (RCF0) is responsible for ensuring that instruments are returned for calibration.
The inspector confirmed that appropriate recall lists had been provided to RCF0. The inspector expressed concern that some instruments had not been returned for calibration on or before their due date.
This item is considered unresolved and will be reviewed in future inspections.
(320/86-15-02)
Radiological Instruments performs Technical Specification calibration of some instruments such as AMS-3 vent monitors. This effort is coordinated by the Instruments and Calibration group (I&C) under Site Operations which has overall Technical Specification surveillance responsibility.
Measurement and test instruments used by Radiological Instruments in their calibration of radiological instruments is also the responsibility of the Radiological Instruments Group. One individual is specifically assigned responsibility for the calibration of the maintenance and test equipment and the maintenance of the records for these instruments.
Instrument condition, records and procedures were found to be adequate.
The inspector had no other' questions in this area.
7.0 Unusual Event - Worker Injured and Contaminated in the RB An Unusual Event was declared at about 9:00 AM, December 16, 1986, when a worker was injured while working on the 347' elevation of the reactor building (RB).
The worker moved a temporary lead shield wall consisting of lead-filled curtains on steel frames mounted on wheels. The movement destabilized the shield wall which fell and pinned him underneath.
The worker was rescued and removed from the RB via the RB equipment hatch. A survey of the worker, done near the equipment hatch in an area with a high background radiation level, indicated some contamination of the worker. However, a re-survey of the worker in the ambulance found him to be free of contamination. A re-survey of the worker at the Hershey Medical Center confirmed the absence of contamination.
The worker was released from the Medical Center and returned to work.
Licensee procedures prohibit the movement of temporary shields by workers without Radiological Controls' approval. The worker was disciplined for moving the shield without prior approval.
Studies are being performed to determine how the stability of the moveable shields can be improved. The inspector will review the results of these studies.
(320/86-15-03)
The inspector noted that the licensee's response to the event was satisfactory.
However, the presence of unnecessary personnel at the RB
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exitcreatedco.kfusionduringresponsetotheevent. As a re 61t, the worker was characterized as contaminated when placed.in the-ambulance
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when, in fact, he was not contaminsted.
The inspector was informed that, in the future, personnel responding to emergencies will be limited to those required to deal with the emergency. Any unnecessary personnel will be required to leave the area.
The inspector will review future responses to emergencies.
(320/86-15-04)
8.0 Unplanned Intake of Radioactive Material by a Worker
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Description of the Event Background On Friday, January 2,1987, a work party was to entei Makeup Pump)
Room 1-C (MU-1-C) on the 281' level of the Auxiliary B611 ding (AB to decontaminate a highly contaminated robot previously used in the adjacent Seal Injection Valve Room'(SIVR).
See Figure 2,." Makeup Pump Room IC and Adjacent Areas"'.-
The Group Radiological Controls Supervisor (GRCS) assigned a well qualified, experienced Radiological Controls Technician (RCT) to cover the work activity. The RCT decided that a survey of thi two contamination control airlocks and MU-1-C was required prior to the planned entry. The RCT decided to use the rcutine HP/0PS surveillance and job coverage Radiation Work Permit (RWP), as opposed to the RWP specifically covering the decon.tamination work, for the survey entry.
The decontamination work RWP included requirements for speci'fic respiratory protective equipment (RPE) and breathing uone air samplers (BZA). The HP/0PS RWP did not. The HP/0PS RWP indicated respiratbry protective equipment "per GRCS";
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however, the RCT did not consult the GRCS concerning RPE.
The GRCS did not know that the RCT was going to make an entry to survey the area and did not give the RCT any specific direction as to which RWP or RPE to use.
The Entry The RCT briefed the workers, specifying PCs, RPE requirements and
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other details on the work-specific RWP under which the workers would make the entry for decontamination of the robot. He instructed the workers to don PCs and other protective equipment and meet him in the inner airlock outside of MU-1-C at'which time he would briet them on the radiation levels determined during his survey.
The RCT entered the highly contaminated (100-200 mrad,smearable)
MU-1-C without RPE or a BZA. Af ter completing the entry, and while briefing the workers, an AMS-3 air monitor in the AB corridor north of the outer airlock alarmed. A high volume air sampler was set up in the corridor and other air monitors were examined to determine the extent of the airborne contamination' event.
No other increase in airborne levels was found.
Based on the high volume air sample analysis results, the RCTs and workers evacuated the area.
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- Post-Entry Actions and Results When the RCT frisked out he found contamination on his chest and by nasal blow.
No one else was found to be contaminated.
The RCT was immediately given a whole body count and found to have a burden of 22.5 nanocuries of Cs-137.
Isotopic analyses of samples of airborne radioactive material from the MU-1-C established a ratio of Cs-137 to Sr-90 of about 1:5. The RCT was given a whole body re-count on Monday, January 5.
A preliminary assessment, based on the second WBC and the Cs-137-to-Sr-90 ratio, resulted in an initial assignment of an exposure of 78.7 MPC-hours to the RCT. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> urine sample was completed January 8.
The licensee will make a final determination of the MPC-hour exposure after receiving the results of the urinalysis. The inspector will review the results of the licensee's determination.
(320/86-15-05)
The first high volume air sample analysis, indicating that there was a very high airborne radioactivity concentration in the corridor outside of the outer airlock, was discovered by the licensee to be in error. An incorrect air pumping rate had been used. A re-analysis using the correct high volume air sampler pumping rate was done and showed a much lower airborne radioactivity concentration.
Contamination surveys showed that the only area contaminated as a
result of the event was the corridor north of the alarming AMS-3.
The corridor was subsequently decontaminated and work activities were resumed, except in MU-1-C.
The licensee initiated an Incident / Event Report and two critiques of the event were held on January 5 and 7 with the RCT in attendance at the second critique.
b.
Immediate Corrective Actions As a result of this incident, all GRCSs were instructed that all RCT job coverage survey entries must be made under the work-specific RWP. Also, the authority of GRCSs to specify whether the routine HP/0PS surveillance and jeb coverage RWP or a job-specific RWP will be used was removed with respect to highly contaminated areas such as the SIVR and MU-1-C.
Requirements are now placed on work-specific RWPs to control the exposure of workers to airborne radioactivity that may result from uncoordinated operation of local ventilation and the opening and closing of doors and airlocks.
In addition, because of concern with respect to the potential for upsets in normal ventilation flows, a study of the building ventilation was initiated and restrictions on opening of large doors were re-established. Such restrictions had been cancelled earlier when it appeared that the opening of such doors had no effect on the local ventilation of the airlocks and the SIVR (adjacent to MU-1-C).
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c.
Inspector Findings Management Controls and Comunication The GRCS who had the responsibility to supervise and instruct the RCT did not do so in an effective manner. The GRCS did not discuss pertinent aspects of the job coverage assignment with the RCT or specify which RWP to use for the RCT's entry. The GRCS relied on the RCT to take appropriate action based on the RCT's experience and to comunicate to him any questions or problems that the RCT had concerning the job.
Comunications were less than adequate in that the meaning of "per GRCS" on the HP/0PS RWP was not clear to the RCT as evidenced by the fact that the RCT did not discuss with the GRCS the RCT's decision to enter the MU-1-C without RPE. Also, the significance of the fact that no previous entries had been made to the MU-1-C without respiratory protection and a BZA, was apparently not made clear to the RCT.
Radiation Work Permit The routine HP/0PS RWP was inadequate for entry into areas such as the MU-1-C.
The routine RWP could be used for surveillance purposes in very highly contaminated areas with the potential for high airborne radioactivity levels.
However, no specific, explicit requirement is included on this RWP for consultation with the GRCS concerning requirements for continuous monitoring, PCs, RPE, and BZAs that may be necessary in the more hazardous areas such as the MU-1-C.
The job specific RWP, RWP No. 15288, required RPE and BZAs.
This RWP was adequate for entry to the MU-1-C.
The inspector will review the licensee corrective actions concerning the use of routine HP/0PS RWPs in highly contaminated areas.
(320/86-15-06)
Surveys 10 CFR 20.201 requires surveys as necessary to compliance with 10
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CFR 20 and to evaluate the extent of any radiation hazards that may
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be present.
10 CFR 20.103 requires, in part, that the licensee use suitable measurements of concentrations of radioactive materials in air for detecting and evaluating airborne radioactivity in restricted areas.
The RCT examined recent area surveys that indicated that the area to
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be entered was highly contaminated. Airborne levels had been l
measured at about 5 MPC. However, the significance of the airborne l
radioactivity concentrations and the potential for higher airborne radioactivity concentrations were not adequately evaluated.
No RPE or BZA was used and the RCT was exposed to unknown airborne radioactivity concentrations and received an intake of radioactive materials in excess of established limits. The failure to perform appropriate evaluations prior to and during the entry is a violation
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of 10 CFR 20.201 and 20.103.
(320/86-15-07)
The licensee's taking of a high volume air sample imediately following the event was appropriate. However, the mis-calculation l
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of the magnitude of the concentration of airborne radioactivity based on the high volume air sampler sample is of concern. The GRCS review of the calculation of the result discovered the error after the 281' elevation of the AB was evacuated due to the erroneous determination of a high concentration of airborne radioactivity.
The circumstances surrounding the use by a technician of an inappropriate air pumping rate will-be reviewed by the inspector.
(320/86-15-08)
9.0 Security Guard Inattentive to Duty During a routine walk through of the plant at about 12:30 PM, January 3, 1987, an. inspector observed a security guard inattentive to his duties in the Truck Bay. The Truck Bay was accessible only via the Auxiliary Building and the Model Room since the Truck Bay outside door was closed.
There were several workers on the fuel Shipping Cask Loading Platfom decontaminating a cask, but none on the floor of the Truck Bay where the guard was located.- The guard did not observe the inspector enter the Truck Bay and mount the platform. He did observe the inspector when the inspector dismounted the platform and left the Truck Bay.
The inspector informed licensee management of his observation.
Licensee response was to reinform guards of the severe disciplinary action that will result when a guard is observed inattentive to duty.
In addition, the licensee will reevaluate the Truck Bay guard location and provide additional guidance for the guard to follow to prevent inattentiveness.
The inspector will follow the licensee actions.
(320/86-15-09)
10.0 Inspector Follow Items Inspector follow items are inspector concerns or perceived weaknesses in the licensee's conduct of operations (hardware or programmatic) that could lead to violations if left uncorrected.
Inspector follow items are addressed in paragraphs 3.0, 7.0, 8.0 and 9.0.
11.0 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 6.0.
l 12.0 Exit Interview The inspectors met periodically with licensee representatives to discuss inspection findings.
On January 16, 1987, the site inspectors summarized the inspection findings in a meeting with the following personnel:
J. Byrne, Manager, TMI-2 Licensing
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l W. County, QA Lead Auditor W. Craft, Manager, RCF0, TMI-2 l
C. Dell, Licensing Technical Analyst l
A. Miller, Manager, Plant Operations l
F. Standerfer, Director, TMI-2 (
J. Tarpinian, Rad Engineering Manager
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At no time during the inspection was written material provided to the
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licensee by the TMICPD staff except for procedure reviews pursuant to Technical Specification 6.8.2.