IR 05000219/1990020

Insp Rept 50-219/90-20 on 901022-26.No Violations Noted. Major Areas Inspected:Internal Exposure Control & Review of Procedures for Drywell Occupancy During Fuel Movement

ML20058J050
Person / Time
Site:	Oyster Creek
Issue date:	11/02/1990
From:	Michael Kunowski, Pasciak W, Sherbini S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20058J048	List: IR 05000219/1990020 ML20058J045
References
50-219-90-20, NUDOCS 9011270161
Download: ML20058J050 (12)

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i U. S. NUCLEAR REGUIATORY COMMISSION

REGION I

Report N /90-20 D>cket No.- 50-219 License N DPR-16 Licensee: .GPU Nuclear Corooration P. O. Box 388 Forked River. New Jersev 08731 Facility name Qyster Creek Nuclear Generatina Station Inspection Att Forked River. New Jersey Inspection Conducted: October 22 - 26, 1990 L Inspectors < MW b ll/l /9 0 M. Kunowski, Radiation Specialist date !

O(DP Facilities Radiation Protection Section, Region III

-. 'h C Wes ". IIlll'? O S. Sherbini, Senior Radiation Specialist date Facilities Radiation Protection Section Approved by: .

& b 11 2 C) O Pasciak, Chief, Facilities Radiation date Protection Section, DRSS Insoection Summary ' Inspection on October 22-26, 1990 (Report N /90-20)-

Areas Insoected: A routine, unannounced inspection of the radiological controls program on site. Areas inspected included ,

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review of previously identified items, internal exposure control, preparations for the upcoming refueling outage, and review of

. procedures for drywell occupancy during fuel movemen Results: Within the scope of this inspection, no violations were identifie PDR ADOCK 05000219 o PDC

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DETAILS l Personnel ContacteG Licensee Personr.el

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p * R. Barrett, Director, Plant Operations

• J. Barton, Deputy Director, Oyster Creek R. Brown, Respirator Maintenance Technician W. Cooper, Radioloci J. Derby, Radiologic. al cal Engineer Engineer R. Giordano, Manager, Outage Radiological Engineering L

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• Heller, Licensing Engineer Hopson, Radiological Engineer Hurley, Radiological Engineer

• MacFarlane, Manager, Site Audit Miller, Radiological Engineer Muehleisen, Maintenance Support Superintendent Parry, Radiological Engineer Pollard, Manager, Rad. Con. Field Operations

• Slobodien, Director, Radiological controls

• Wolf, Manager, Radiological Engineering NRC Personnel L E. Collins, Senior Resident Inspector

• M. Banerjee, Resident Inspector

• Denotes attendance at the exit meetin i Status of Previously Identified Items (Closed) Follow-Up Item 87-02-03 This item-addressed the licersee's actions in response to NRC Bulletin 80-10, "Contarination of nonradioactive systems and resulting potential frr unmonitored, uncontrolled release of radioactivity ..n the environment". This item was reviewed but left'open in Inspection Report 50-219/90-14. At l that time, the licensee had developed an action plan to resolve any remaining issues. This work was not complete at

the time of the current inspection. However, this item is

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being closed because completion of the work is being tracked

- under the Resident Inspector's open item 90-06-05, which addresses the same issue, namely, work related to Bulletin j 80-10.

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2.2 (closed) Noncomoliance Item 89-16-01

This item was opened in connection with violation of the -

requirements of 10 CFR 50, Appendix B, " Quality Assurance criteria For Nuclear Power Plants And Fuel Reprocessing Plants", Criterion XVI, " Corrective Action". The violation addressed the fact that there had been 10 violations of the locked high radiation area door requirements over the previous ten-month period and that the licensee's corrective actions had not been sufficient to prevent recurrence. The licensee's corrective actions following the violation included the followingt o The procedure dor control of locked high radiation areas, 9300-ADM-4110.06, " Control of Locked High Radiation Areas" was revised to improve control of these area o A second persort, called a verifier, is now required to verify that the door to the locked high radiation area that had just been occupied is locked at the completion of work in that area, and he is required to document this verificatio These actions appear to have been effective; there have been no locked high radiation area door violations since the .

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third quarter of 1989, that is, for a period of over a yea The licensee is also inspecting all locking mechanisms on locked high radiation area doors with the intent of replacing those that are not easy to lock or are defective in any way. Based on these_ actions, this item is considered close ,

2.3 (Closed) NoncomollADee Item 90-09-02' '

This item addressed a two-part violation for failure to perform adequate radiation surveys. One part referred to work 3rformed under the vessel by instrumentation technicians working on a carousel which is raised about 8 feet off the floor. Work on the carousel places the workers closer to the reactor vessel, which is a radiation sourc '

The carousel area had not been surveyed prior to start of the. work. The licensee attributed the incident to miscommunication between the workers and the radiological co'strols supervisor who conducted the pre-job briefing. The l'.censee's corrective actions includedt

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o The parts of the inspection report dealing with the incident were made required reading for all maintenance and l radiological controls personnel.

l o The issues identified as a result of the investigation of the incident were made a part of the cyclic training for the radiological controls technicians (RCT).

l o Management discussed the incident with the radiological controls staf o Disciplinary action was take i The second part of'the violation addressed failure to identify a hot spot reading 2.8 R/hr on contact and 600 i mR/hr at one foot during surveys of the Shutdown Cooling i

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Room. Work performed in the vicinity of the hot spot led to

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a dose of 40 mrem each to a worker and an RCT. The licensee )

attributed the cause to error by the RCT in performing the I survey. Corrective actions were similar to those for the i first part of the violation. Based on these actions, this ,

item is considered close ] (Closed) Noncomoliance Item 90-09-03

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This item refers to a violation of a radiation work permit (RWP) requirement for entry-into a high radiation area. A maintenance worker entered the area without a dose rate meter, as required by the RWP. The violation was brought to NRC attention through an allegation which claimed that a supervisor had directed a plant worker to enter a high radiation area without a dose rate meter. The NRC and licensee investigations confirmed the allegation concerning entry without a dose rate meter, and also that this deficiency had been identified but not documented by a RCT at the time it happened. However, the investigations i concluded that the supervisor had not directed the worker to ,

enter the area without a meter. It-was concluded that the

. worker was unaware et entry requirements for high radiation areas and had not understood the requirements specified on  ;

the RWP. The investigations also faulted the RCT for not documenting the incident. The licensee's corrective actions  ;

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o Counseling for the individuals involved.

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o Incorporation of the incident into General Employee L

Training (GET) and emphasizing high radiation area entry requirement o A memo was issued to all RCTs reminding them of the policy following identification of incidents or deviation Based on these actions, this item is considered close .0 Internal Excesure Control ,

The inspector reviewed the licensee's implementation of its internal exposure control program, including the adequacy of the assessments of intakes of radioactive material, L calibration and maintenance of equipment, and efforts to l maintain exposure ALARA. The status of previously identified

,. items in this area was also reviewed. No major weaknesses

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were identified within the areas inspecte .

3.1 Whole Body countert Th9 operation and quality control of the whole body counter facility were found to be good. The operation is supervised by a radiological engineer,-and the ,

staff was found.to be knowledgeable, well trained, and dedicated, and most have many years of experience operating >

the equipment. A review of the procedures applicable to this area showed the following items for improvemen o Procedure 9310-ADM-4246.01, " Operation of the Canberra Whole Body Counting System", Attachment 10.1, Section 3.2,

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requires that the Co-60/Cs-137 check source be placed 48" above the counter base plate. The correct practice, currently used by the licensen, is to place the source 44" i above.the plate, o The same Attachment above, Section 4.1, incorrectly refers to Sections 4.1 through 4.3 for instructions to verify L proper preparation of the whole body counter. The correct sections are 7.1 through o Procedure 9300-ADM-4025.01, " Bioassay Procedure", refers to Procedure 9300-ADM-4246.01, " Operation, Calibration and

Quality Assurance of the Canberra Whole Body Counting System". However, this procedure is no longer in use; the proper procedures are 9310-ADM-4246.01 and 9311-PMI-4222.0 i

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6 The licensee initiated corrections for the above items during the inspection, and most of the items were corrected before the and of the inspectio . Resniratory Protection Proarant A review of the training and qualifications of the operating staff of the respirator l maintenance facility showed that the staff was well trained i and qualified even though a significant number of them have I only recently been appointed to work in that facility. A tour of the facility and observation of some ongoing activities showed no problems in that operation. The applicable procedures were also reviewed and were found to be good. The licensee stated that they are in the process of discontinuing the use of the corn oil respirator fit booth j in favor of the recently developed fit testing system that operates on ambient dust to test for fit. It is expected i that implementation will be complete before the start of the upcoming refueling outage. A significant recent organizational change was the reassignment of the major responsibility for the respiratory protection program on sit The acjor responsibility was shifted from the corporate Respiratory Protection Supervisor to the site i radiological engineer who previously was responsible for the l day-to-day operation of the site progra A review of internal exposure records showed that there have been few internal exposures in the recent past, and those that did occur were relatively low, generally less than 10 i MPC-hours. The licensee's action and calculations following a recent hot particle ingestion were also reviewed and were found to be good. All the necessary whole body and excreta analyses were performed and the calculations adequately addressed the various factors involved in the inciden However, some concerns were identified in connection with-this incidents o-The licensee's analysis focused mainly on doses received l

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as a result of the contamination incident, and little effort was devoted to accurately assess the worker's intake to determine compliance with applicable regulations. The l- regulations limit intakes rather than delivered doses in l l' internal contamination situation o The workers involved in the incident knew they were

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,. entering a highly contaminated area of the drywell and they L had no respirators because of the extreme heat in the areas l

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entered. The license's policy in such situations is t- >

<e the workers wear dust masks as protection against larg*

particles of radioactive material. The workers did not wear *

these masks because none were available at the tim o The workers spent a significant amount of time looking for a valve in a cramped and contaminated area with relatively elevated radiation fields. They were apparently inadequately briefed regarding the exact position of the valve. In addition, the valves in that location, and many others, are ,

not labelled for ease of identificatio The licensee stated that they have already taken action to i correct these deficiencies. In particular, the licensee stated.that they were in the midst of an extensive labelling program that involves placing clearly visible labels on components and valves throughout the plant to facilitate identification. The licensee stated that the drywell labelling portion of the program will be implemented during the upcoming outage. Progress in this area will be reviewed during future inspection .0 Qutaae Prenarations Preparations for the upcoming refueling outage (13R)

included a reorganization of the Radiological Controls Department, expansion of access facilities and installation of.new remote monitoring and communications hardware, and radiological review of upcoming wor .1' outaae oraanizatiED: Westinghouse (vendor) has been assigned to perform a major portion of the outage work, including all work in the drywell, the refueling floor, and the turbine deck. This work is expected to account for'about 85% of the dose accumulated during the outage. Radiological controls for the outage, however, will be under the direction of the site radiological controls organization. A new section has been added to the Radiological controls Department on site r for the duration of the outage. The section is the Outage Radiological Engineering section, and the section manager, a Westinghouse engineer with extensive radiological experience, reports to the site Radiological Controls Director. The staff in the section cs.trently includes four contractor radiological engineers and three licensee radiological engineers on temporary assignment to the section. This staff of seven radiological engineers is

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expected to be augmented to ten during the outage, and will be known as ALARA Liaison engineers. The existing site Radiological Engineering Section remains unchanged, and its functions will be mainly work not directly related to the outage. It will, however, perform some outage support work, such as dose calculations, shielding calculations, and ,

respiratory protection suppor l The functions of the Outage Radiological Engineering section during the planning phase includet o Review job scope and recommend changes if necessar ,

o Review job packages and associated procedures to ensure that the proposed work is to be done in a radiologically sound manne o Generate ALARA reviews and estimate exposure for each job, and provide this information for use in generating RWP During the outage, the section functions will expand to

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include the following:

o Follow the progress of jobs and track exposure o Identify and investigate reasons for deviations from estimated exposure for each jo o Advise Field Operations staff on actions to correct deviation o Take part in pre-job briefings and mockup trainin An ALARA Liaison Engineer will be assigned to each of the main work areas, including the drywell, refueling floor, and turbine deck, and will provide continuous coverage on a two shift rotatio The Radiological Controls Field Operations (RCFO)

organization will remain essentially unchanged except for l augmentation of the technician staff. A Group Radiological *

Controls Supervisor (GRCS) will be provided each shift to i supervise radiological controls in each major work area, such as the drywell, turbine building, refueling floor, l reactor building, and the yard. The same GRCS will always be I assigned to the same area to ensure continuity. A RCFO I l

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manager or acting manager will be on site at all times, and will be assisted by an assistant manager and a Shift GRCS Coordinator, who will supervise the activities of all the other GRCSs on the shif .2 Facilitiest There are currently two access facilities ic +.he radiological controls areas (RCA): a general access facility and a drywell access facility. Both of these facilities will be expanded to improve access control and expedite acces The expansions will include more space for dressing out,  ;

more access computer terminals, more space for briefings, and more frisking equipment for exit contamination contro In addition, the licensee is installing a drywell command  !

center..This center will contain closed circuit TV monitors t and a base radio station. Approximately 15 TV cameras will be installed in the drywell and monitored from the command center. The base radio station will allow radio communication with personnel in drywell. The licensee stated '

that the drywell RCTs and some job supervisors will be equipped with communicaticos headsets to communicate with the command center. The ceamand center will be manned by a drywell coordinator and a radiological controls person. The command center equipment was obtained-from TMI-2 where it was used during the defueling operation .3 outaae Work scoce: In addition to the normal refueling, maintenance, and inspection activities, the outage will l include chemical decontamination of the recirculation piping and the Reactor Water Cleanup System (RWCU) . Exposure reduction modifications are also planned for the drywell, including installation of shielding, better lighting, a chiller, and communications and observation (TV) system The total outage exposure is estimated to be a little over 1400 man-rem, but this figure is still subject to chatsg I The major job groupings and their associated exposure estimates are shown below, o Refueling 105 man-rem ,

o Support (rad con, maintenance) 285 " !

o Major inspections / testing 240 "

l o Major valve work '65 "

o control rod drive work 90 " '

L o Scaffolding / insulation / shielding 140 "

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o Chemical decontamination 120 "

l o Major piping work (Isocond./RWCU) 95 "

l o Asbestos removal 40 "

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o Stress improvement / pipe repair 50 "

Total 1230 man-rem The balance of the estimated total 1400 man-rem is taken up by small repair and maintenance jobs, or less dose intensive i wor i

' Drywell Occunancy Durina Fuel Movement The licensee plans to continue work in the drywell during fuel movement from the reactor vessel to the fuel storage '

pool or the reverse. To guard against excessive radiological exposures in case of a fuel drop accident, the licensee has developed two radiological controls procedures for control of work in the drywell during fuel movement: I o Procedure 9300-ADM-4110.13, "Drywell Occupancy During Fuel Handling Operations". l o Procedure 9300-OPS-4224.05, "Drywell Fuel handling Radiation Monitoring Testing".

A review of Procedure 9300-ADM-4110.13 identified several areas of concern. These are discussed belo I J

o Section 5.2.1 states tt t " Persons who require access to I the drywell during refueling will be trained on this procedure". However, the procedura does not include any required training for RCTs in the drywell and on the refueling floor, or for workers involved in fuel movemen I The licensee stated that they will include these personnel in appropriate training to explain to them the requirements j of the procedures as well as the hazards involved in moving ,

fuel while work is in progress in the drywel l, o Section 6.2 states that the dose rates at the 75'

elevation in the drywell would be about 20 R/ min in a worst case fuel drop accident, dropping to about 1.2 R/ min at the drywell access elevation. The inspector asked about the time to evacuate from that elevation. The licensee stated that they estimate about 30 seconds. Therefore, in such a worst case accident, the exposure received during evacuation from the 75 elevation would be less than 10 rem. The licenkse also stated that they do not intend to allow access above '

the 46' elevation during fuel movement except in an

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emergency situation. The exposure rate at the 46' elevation in the worst case drop accident is estimated to be R/ mi o section 6.3 states that during fuel movement "...no person l is allowed ...on or above drywell elevation 46', unless l concurred in by the drywell coordinator and authorized by l

the GRCS...". The inspector stated that this statement in ;

effect authorized access above the 46' elevation during fuel )

movement, contrary to the licensee's statement above. The J licensee stated that they do not intend to allow access above the 46' elevation except in an emergency, as stated l above, and that this section was not properly worde J o Section 6.4 states that, in case of delays in fuel J handling, personnel may go above the 46' elevation in the l drywell and that " Prior to resuming fuel movement, the 119' '

(refueling floor) GRCS will notify the Drywell GRCS and Drywell coordinator". The inspector stated that this provision does not clearly identify the lines of authority .:

for control of work in the drywell and resumption.of fuel movement after a delay. Resumption of fuel movement should :

not proceed until the drywell coordinator or GRCS authorized such resumption. Such authorization must be based on knowledge of the location of workers in the drywell and completion of necessary preparations for resumption of fuel movement, and can only come.from supervisors in charge of I drywell operation. The licensee stated that they will clarify this line of authorit o Section-7.1.4 states that "The drywell radiation monitors ,

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will trigger a " klaxon" (or equivalent) continuous alarm that will sound as long as the radiation level exceeds the i set point". Section 7.3 states that "The radiological controls personnel at the drywell entrance will utilize a megaphone or air horn to evacuate personnel in the drywell".

The inspector asked if these alarms can be heard in all locations in the drywell, under normal working conditions, where access will permitted. The licensee stated that the ,

alarms will be tested for audibility in the drywel o The procedure did not mention the pattern of fuel movement .

to^be used for removal of the fuel from the core or insertion of the fuel into the core. The inspector stated that persoanal working at or above the top of the biological shield or close to penetrations in the biological shield

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would be exposed to high radiation fields if the fuel is allowed to approach the inside wall of the pressure vessel I during fuel movement. The licensee stated that this matter has not yet been resolved, and that in any case workers will not be allowed anywhere close to the top of the bioshield or to penetrations in that shield that would represent a high radiation sourco, o The procedure did not state, as a minimum requirement for !

access anywhere in the drywell during fuel movement, that the fuel shute shield must be in place. The liceasee stated that installation of the fuel shute shield is a requirement for fuel movement that is specified in a refueling procedure. That procedure was not reviewed during this inspectio o Procedure 9300-OPS-4224.05 specifies the requirements for N

testing the monitors and alarm set po..Its. The inspector stated that the tests appeared to be function checks of the electronics that by-passed the detector and therefore did not assure detector operability. The licensee stated that this was the case and that they introduced a separate background check to ensure detector operability. In that check, the background at the detector location is measured

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using a survey instrument and the detector background '

reading must be within 20% of that measuremen '

o The detectors inside the area monitors to be used for evacuation alarms are GM tubes. These tub'es are susceptible to saturation if e:: posed to high radiation fields, and such saturation would cause their readings to be much lower than the actual field, or even-close to zero, thus defeating their alarming function. The licensee stated that they will I

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take appropriate precautions to guard against this phenomeno .0 Exit Meeting The' inspector met with licensee representatives at the end of the inspection on October 26, 1990. The inspector reviewed the purpose and scope of the inspection and discussed the inspection findings, i

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