IR 05000293/1993016
| ML20058D486 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/19/1993 |
| From: | Joyner J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Boulette E BOSTON EDISON CO. |
| References | |
| NUDOCS 9312030165 | |
| Download: ML20058D486 (4) | |
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Docket No.
50-293 E. Thomas Boulette, PhD Senior Vice President-Nuclear Pilgrim Nuclear Power Station Boston Edison Company Rocky Hill Road Plymouth, Massachusetts 02360
Dear Dr. Boulette:
SUBJECT: NRC Inspection Report No. 50-293/93-16 An announced safety inspection of the radiological controls program at your facility was conducted by Mr. J. Noggle on August 30 - September 3,1993. Additional information was requested, with the inspection continuing in the NRC Region I office through October 15 for review of that information and for discussion and evaluation with the Occupational Health and Safety Administration (OSHA) of an industrial safety practice observed during the inspection. Preliminary inspection findings were discussed with members of your staff at your facility on September 3,1993, with additional inspection results communicated by telephone to Ms. Laurie Wetherell on September 9,1993. The inspection results were funher discussed during a November 17,1993 telephone conversation between you and J. H.
Joyner of this office.
Areas reviewed during the inspection are important to health and safety and are discussed in the enclosed inspection report. These areas included review of: audits and surveillances, training and qualifications, radiological problem repons and associated corrective actions, external exposure control, and internal exposure control programs.
Overall performance of the radiological controls program at Pilgrim Station has been good.
External radiation exposures have been relatively low, although slightly above your As low As Reasonably Achievable (ALARA) annual goal as of August 27,1993. Your performance in limiting internal exposures during 1993 has been excellent. Relatively high numbers of personnel contaminations were reported and poor radiological work practices were identified through your audit program before and during the outage period. We acknowledge the
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strong performance of your problem self-identification program. In some instances there was great thoroughness with which your staff addressed the issues, for example, that of contaminated tools found in clean areas. However, better worker accountability and i
additional training frequently have been the only actions prescribed by your staff to correct j
many of the radiological incidents when weakness of health physics controls were also
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implicated in the incidents. As described in the enclosed inspection report, however, we believe that occasional lapses in the control of radiological work by your health physics
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i 931203o165 93 nFFICIAL RECORD COPY l
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organization have been a factor in some of the problems identified by your staff.
Accordingly, we believe that additional attention to the establishment of the root causes of radiological problems and to the establishment of corrective actions appears warranted.
During this inspection, two violations of NRC requirements (one with two examples) were identified that were associated with the control of radiological activities. Due to the self-i identification of these violations by your staff and the timing of these violations relative to a
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previous violation in this area, we have elected to exercise discretion and not issue citations for these violations. We do feel, however, that your corrective actions have been too narrowly focused. One violation of 10 CFR 20.201, failure to survey, was identified and is described in the enclosed Notice of Violation and in the enclosed inspection report. You are required to respond to the Notice of Violation in accordance with the instructions contained therein.
One unresolved item was identified. This item was unresolved as a result of difficulties experienced by your staff with records retrieval. Specifically, documentation demonstrating HP technician training qualification in accordance with station procedures was not available during the inspection period. This matter will be further reviewed in a future inspection.
Further, a drywell entry into an oxygen-deficient environment did not appear to meet OSHA l
guidance for maintenance of communications. In accordance with a memorandum of j
understanding between OSHA and the NRC, this matter was reported to the regional OSHA l
office on October 15, 1993.
In addition, on September 1,1993, during a telephone discussion between you and Mr. C.
W. Hehl of this office, several additional corrective actions to a violation identified during NRC Inspection No. 50-293/93-10 were reviewed. Our understanding of those actions is described in Section 3.2 of the enclosed inspection report. If your understanding differs from that documented herein, please contact this office immediately.
Your cooperation with us is appreciated.
Sincerely, oiginalsigned By:
James H.Joyner James H. Joyner, Chief Facilities Radiological Safety and Safeguards Branch Division of Radiation Safety and Safeguards
Enclosures:
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NRC Inspection Report No. 50-293/93-16 OFFICIAL RECORD COPY
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.h0V 19 1993
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Boston Edison Company
REGION 1
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Report No.
50-293/93-16 l
Docket No.
50-293 i
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License No.
DPR-63 Licensee:
Boston Edison Company RFD #1 Rocky Hill Road j
Plymouth. Massachusetts 02360 i
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Facility Name:
Pilgrim Nuclear Power Station Inspection At:
Plymouth. Massachusetts j
Inspection Conducted:
August 30 through October 15.1993 N/
Inspector:
J. Nogle, Senikr@adiation Specialist IIate Facilities Radiation Protection Section
////f/93 Approved by:
R. Bores,# Chief, Facilities Date
Radiation Protection Section Areas Insoected: Areas covered in this inspection included a review of: audits and surveillances, training and qualifications, radiological problem reports and associated corrective actions, external exposure control, and internal exposure control programs.
Results: Overall performance of the radiological controls program at Pilgrim Station has been good. External radiation exposures have been relatively low, although slightly above the licensee's As Low As is Reasonably Achievable (ALARA) annual goal as of August 27, 1993. The licensee's performance in limiting internal exposures during 1993 has been excellent, but relatively high numbers of personnel contaminations were reported and poor radiological work practices were identified through the licensee's audit program before and during the outage period. The strong performance of the licensee's problem self-identification program is acknowledged. In some instances the addressed the issues with great thoroughness, for example, that of contaminated tools found in clean areas. However, better worker accountability and additional training frequently have been the only actions prescribed to correct many of the radiological incidents when weakness of health physics D@Cb
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controls were also implicated in the incidents. Occasional lapses in the control of radiological work by your health physics organization have been a factor in some of the problems identified by the licensee. However, this component of the root causes of such problems has not been identified or addressed in the establishment of corrective actions.
Additional attention in this area appears warranted. During this inspection, two violations of NRC requirements (one with two examples) were identified that were associated with the control of radiological activides. Due to the self-identification of these violations by your staff and the timing of these violations relative to a previous violation in this area, we have elected to exercise discretion and not issue citations for these violations. We do feel, however, that the licensee's corrective actions have been too narrowly focused. One violation of 10 CFR 20.201, failure to survey, was identified and is described in this inspection report.
One unresolved item was identified. This item was unresolved as a result of difficulties experienced by the licensee's staff with records retrieval. Specifically, documentation demonstrating HP technician training qualification in accordance with station procedures was not available during the inspection period. This matter will be further reviewed in a future inspection. Further, a drywell entry into an oxygen-deficient environment did not appear to meet OSHA guidance for maintenance of communications. In accordance with a memorandum of understanding between OSHA and the NRC, this matter was reported to the regional OSHA office on October 15, 1993.
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DETAILS 1.0 Individuals Contacted 1.1 Boston Edison Comoany
- G. Basilesco, Senior Compliance Engineer
- N. Desmond, Compliance Division Manager L. Dooley, Technical Training Section Manager
- F. Famulori, Quality Assurance Department Manager
- R. Gay, Senior Compliance Engineer
- E. Kraft, Jr., Vice President - Nuclear Operations S. Landahl, Radiological Operations Support Division Manager
- R. Lewis, Radiological Training Supervisor
- P. Markson, Communications Specialist
- D. Montt, Chemistry Division Manager
- M. Most, Chemistry Supervisor
- L. Schmeling, Plant Manager J. Walker, Problem Report Coordinator
- L. Wetherell, Radiological Section Manager
- A. Williams, Station Services Section Manager 1.2 USNRC Personnel
- R. Bores, Chief, Effluents Radiation Protection Section
- J. MacDonald, Senior Resident Inspector
- A. Cerne, Resident Inspector
- L. Peluso, Radiation Specialist
- Denotes attendance at the exit meeting on September 3,1993.
2.0 Purnose This inspection was an announced safety inspection of the Pilgrim Nuclear Power Station radiation control programs.
3.0 Previously Identified items 3.1 (Closed) Inspector Followup Item (50-293/92-11-01)
In a previous inspection, the inspector questioned the technical adequacy of using a large anicle survey monitor (bag monitor) for the free release of material from the site based on a 50,000 disintegrations per minute (dpm) aggregate activity alarm setpoint. The inspector reviewed current revisions of the applicable station procedures and noted that the bag monitor was not used as a stand alone survey tool.
Current station practice requires a direct frisk of material followed by a survey with
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l the bag monitor prior to the unrestricted release of material. The inspector was satisfied that appropriate controls were established for controlling radioactive material
and this item is now closed.
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3.2 (Closed) Severity Level IV Violation (50-293/93-10-01)
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On April 24,1993, three radiation workers made unauthorized entries into a posted high radiation area and one of the workers entered into a high radiation field without knowledge of the radiological conditions and without being on the correct Radiation Work Permit. Some confusion surrounded the proper response to the alarming pocket
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dosimeter alarms. A letter was received from the licensee dated July 26,1993 in response to the notice of violation. NRC Region I issued a reply letter, dated August
i 12, 1993, which identified additional root causes for the April 24,1993, event and requested further evaluation and corrective actions to resolve the violation. On September 1,1993, by telephone conversation between Dr. and staff of
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Boston Edison Company and Mr. C. Hehl and staff of this office, the licensee communicated additional corrective actions that they agreed to implement. These corrective actions are recorded below as Inspector Followup Items (IFIs) and will be reviewed during future inspection activities.
3.2.1 (Onen) IFI (50-293/93-16-01)
The licensee discussed its plans to implement a new automated exposure control system at Pilgrim Station by January 1994. The new computer software will restrict radiation work permit (RWP) use to only authorized users. According to the licensee, this will help prevent the incorrect use of RWPs by unauthorized individuals.
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3.2.2 (Open) IFI (50-293/93-16-02)
Procedure 6.1-022 provides the work supervisor RWP responsibilities. The licensee agreed to clarify the work supervisor's responsibilities by excluding from those responsibilities the authority to perform the formal radiological briefings.
3.2.3 (Onen) IFI (50-293/93-16-03)
Procedure 1.3-106 describes the alarming pocket dosimeter alarms and the required responses to each. The licensee agreed to clarify the language in Section 5.3 of this procedure to provide for unambiguous interpretation of alarm response required by the procedure.
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4.0 Audits and Surveillances
The inspector reviewed the latest Radiological Protection Program Audit No. 93-05, which was performed between April 20,1993, and May 26,1993. The auditors
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included an HP technical specialist from Pilgrim Station and one from Maine Yankee.
This audit reported that the radiological protection program at Pilgrim Station was
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implemented in compliance with established requirements. One deficiency report was issued for failure to properly document RWP required surveys within specified frequencies. In addition, poor radiation work practices were described as of concern
and to warrant licensee management attention. The audit was detailed, of good
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quality, and performance-based. The audit findings were documented, tracked, and assigned for appropriate close out.
The inspector reviewed the surveillances that had been performed since the last NRC HP inspection of May 14, 1993. A variety of technical HP areas were reviewed in nine surveillance reports. No recommendations or findings were reported. The apparent discrepancy between audit findings / radiological trend repons relative to identification of poor radiation worker practices and the lack of surveillance report findings was discussed with the licensee. As explained to the inspector, the licensee's surveillance program provided a select sampling of very specific elements of the HP program, and was not intended to provide an assessment of overall program performance. However, overall, the licensee's audits and surveillances demonstrated effective oversight and assessment of the radiation protection program.
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't 5.0 Trainine and Oualifications The licensee's training and qualification program for HP technicians was reviewed through discussions with personnel and through the review of qualification records.
Criteria used for this review included the American National Standard ANSI N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel", NUREG-1220,
" Training Review Criteria and Procedures", and the licensee's training program procedures.
From the qualification records, the inspector reviewed the experience level of each of the radiological operations permanent HP staff. The results indicated an average of 9.4 years of nuclear experience with each member fulfilling the ANSI criteria for health physics technicians. Currently, Pilgrim Station provides an HP technician training and qualification program upon initialjob assignment. This course requires approximately one year to complete and consists of approximately one-third classroom j
training and two-thirds on-the-job training. The inspector requested training qualification records on six randomly selected HP technicians. The licensee was not able to provide the requested records before the end of this inspection due to record retrieval difficulties. The licensee does not allow HP technicians to undertake safety protection responsibilities until HP qualifications have been completed and there was
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no indication (based on inspector observations and discussions with licensee HP technicians) that any on-duty HP technicians were unqualified. Technical Specification 6.2.4 requires that while fuel is in the reactor, an individual qualified in
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radiation protection procedures shall be on site at all times. The confirmation of station HP technician training qualifications remains unresolved pending the retrieval and subsequent NRC review of the qualification training records (Unreso!ved Item 50-293/93-16-04).
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Currently, once an HP technician completes the HP technician training course and becomes qualified on station procedures and practices, the individual is not trained or examined again in those areas by the training department. A review of core HP
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procedures by the radiation operations HP technicians is provided by a quarterly self-study requirement that the radiation operations HP technicians read selected procedures as determined by the Radiological Operations Division Manager. The inspector questioned whether HP technicians at the station were qualified in various HP procedures, instruments, and methodologies since some HP personnel qualified as long as 15 years ago to a substantially different HP program, whereas, others just recently completed the current qualification program. The training department had self-identified the need for addressing HP technician requalification and has committed to measuring these training needs through administration of a l
requalification exam before the next refueling outage. This will be reviewed during a
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subsequent inspection of this area.
Each year the training department has a goal of providing three weeks of continuing training to the qualified HP tmhnicians. The target goal of three weeks includes completion of the annual routine requirements of all radiation workers, such as a physical examination, whole body count, respirator training, general employee training, and radiation worker training. Continuing training subjects taught during 1993 included: hazardous material awareness, conduct of radiological operations, radwaste material packaging and handling, plant systems review, introduction to personal computers, gamma spectroscopy, solid state detectors, and review of industry events. Review of core health physics course material is not generally provided through the continuing training program.
In conclusion, the inspector was satisfied that an acceptable training program is in place with good direction towards the development of an HP technician requalification l
program. However, a weakness associated with rotation of personnel to new positions is discussed in Section 8.2.
l 6.0 Radiological Problem Reports (RPRs)
The inspector reviewed the latest RPR Trend Report, covering the second calendar
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quarter of 1993. The licensee generated 88 RPRs during the second quarter of 1993 compared to 79 RPRs written during the entire year of 1992 and 45 RPRs written l
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during the first quarter of 1993. The second quarter encompassed a refueling and maintenance outage and a higher number of Radiological Problem Reports could be expected. The latest trend report noted a large variety of RPR categories represented,
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with the most common categories due to personnel contaminations, clean area contaminations, procedure violations, and RWP violations. The total number of I
personnel contaminations was 314 for the outage, which resulted in 37 of the 88 RPRs. Of the other RPRs, seven were instances of contaminated tools found in clean areas of the station since the outage. Three RPRs were written for workers exceeding l
their alarming pocket dosimeter dose alarm setpoints (maximum of 179 mR over).
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One RPR was written for a worker inside the drywell (a high radiation area) with an alarming pocket dosimeter turned off. The licensee determined that the root causes for these incidents were human factors, inattention to detail, and personnel error.
Individual counselling and/or disciplinary action were prescribed to effect short-term corrective actions. In order to correct these worker deficiencies in the general workforce, the licensee plans to present two-day advanced radiation worker training for all Boston Edison Company craft personnel (mechanics, electrical maintenance, instrumentation and controls, custodians, and chemistry personnel) by the spring of 1994. Although the BECo radiation workers were considered a seasoned workforce,
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the licensee believes that additional training in radiation work practices, combined
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l with disciplinary action for non-compliance activities, will correct the trend toward poor radiation work practices at Pilgrim Station.
A previously identified significant RPR that was reviewed by the inspector during the
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l last inspection' had not been resolved by the licensee at the time of that inspection.
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RPR No. 92.0224 was a compilation of several recurring instances of finding
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contaminated tools outside of the RCA. Since March 1,1993, the licensee established a single access point to the RCA, which required all equipment and personnel leaving the RCA to pass through this point for contamination monitoring.
Prior to this change, the licensee surveyed the site for contaminated material to ensure contamination controls were in place. Prior to and after the single RCA access point change was made, there have been many recurring instances of contaminated tools being found in " clean" tool boxes. Due to the chronic lack of contaminated tool control, the licensee raised the level of attention of this issue and has combined the individual contaminated tool incidents into RPR No. 92.0224. The licensee established a contaminated tool issue depot inside the RCA in an attempt to meet the demand for tools in the RCA and reduce the traffic of tools into and out of the area.
Since the previous inspection, additional corrective actions were assigned by the licensee, which effected closure of the RPR. These include:
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bi-weekly surveys of tool storage locations;
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adoption of a general policy to not release tools from the RCA; l
' NRC Inspection No. 50-293/93-10, May 10 - 14,1993 I
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relocation of the HP office for optimum visibility of the RCA egress area;
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development of a procedure for a tool contamination monitor;
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periodic assignment of an HP technician to monitor tools as they are returned to the tool storage areas; and
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research of available industry methods for the automatic survey of materials.
The inspector determined that sufficient corrective actions were assigned to reasonably expect resolution of the chronic recurrence ofincidents of contaminated tools found in
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clean areas.
The inspector reviewed several recent RPRs in detail. These RPRs were not reviewed during the previous inspection, however they did occur during the same outage period. The licensee demonstrated a very effective incident self-identification program. Four RPRs of significance are described below. These examples demonstrate weaknesses in radiological controls of work activities, which were not directly addressed in the RPR corrective action recommendations.
RPR No. 93.0317 involved an April 23,1993, licensee-identified event when an ALARA Specialist, while walking down some shielding installations in the drywell, noticed that piping insulation had already been removed from the N9B nozzle without using the appropriate specific RWP and without a required radiological pre-job briefing. None of the drywell HP technicians were aware of this job and the specific RWP for the planned work had not been used. This exposed N9B piping was later surveyed to be 11 R/hr at contact and 700 - 1500 mR/hr at 30 centimeters. The licensee's event investigation failed to determine who removed the insulation after questioning all personnel who were signed in on RWPs and who logged into primary containment through the plant security system. No workers came forth to acknowledge having removed the insulation despite the licensee efforts to locate the individual (s) responsible. No specific corrective actions were adopted to resolve this incident. This RPR was closed with reference made to RPR 93.0322 corrective actions, which recommended two-day advonced radiation worker training.
The drywell was a high radiation area that was continuously controlled by the HP staff during this time period in the following manner. All workers were required to sign on the appropriate high radiation area RWP and receive a pre-job radiological briefing before entering the drywell. Additionally, any non-health physics personnel entering the drywell received an initial work area radiation survey to verify dose rates
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and establish intermittent HP coverage of the work by a roving HP technician who was in constant radio contact with the drywell control point. All of the applicable drywell HP technicians were interviewed by the licensee and none were aware of the N9B insulation removal activity. If the work crew misidentified a work location or insulation removal task, the roving HP technician could have identified this error and provided control of the work. The failure to provide control of drywell work by HP
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personnel was believed by the inspector to be a significant factor in this incident.
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Te hnical Specification (T.S.) 6.13.2 states that the requirements of T.S. 6.13.1 shall alm apply for high radiation areas in which the intensity of radiation is greater than
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1000 mrem /hr. T.S. 6.13.1 requires that entrance to each high radiation area shall be controlled by issuance of a Radiation Work Permit. The RWP that would have authorized this work was not used. The entry was made without this authorization and resulted in work on and the uncovering of a high radiation source. Due to the issuance of a similar violation issued for failure to use the correct RWP that occurred during the same time frame, this licensee-identified violation of Technical Specifications will not be cited.
Radiological Problem Report No. 93.0366 describes an April 30,1993 refueling floor incident in which three workers received measurable intakes of radioactive material while packaging highly contaminated main steam line plugs. The workers were properly signed in on the correct RWP and were dressed in protective clothing as specified by the RWP (no respirators or face shields were required). The workers did brief the HP technician at the refueling floor clean area checkpoint on the work they i
were going to perform, but did not interface with the roving HP technician. Upon entering the refueling floor cc.ataminated area, they proceeded to package the main steam line plugs for storage and left the area. The personnel contamination monitors alarmed and subsequent bioassay measurement and evaluations determined that the three workers, respectively, received 3.7,0.9, and 0.8 MPC-hours of internal radiation exposure.
At the time of the incident, the main steam line plugs had not been surveyed to determine the radiological hazard to the workers. Subsequent to the event, the
licensee determined that the plugs had contamination levels of 5 mrad /hr/100 cm smearable and a total contamination level of 30 mrem /hr measured at contact with the plugs. Had the main steam line plugs been surveyed prior to the job, appropriate safety precautions could have been prescribed on a specific RWP and an unplanned internal exposure event could have been prevented. This is a violation of 10 CFR 20.201(b), failure to survey, pursuant to 10 CFR 20.103 (NOV 50-293/93-16-05).
RPR No. 93.0377 involved a May 10,1993 incident in which four workers exceeded their alarming pocket dosimeter (APD) dose alarm setpoints. The four workers were dressed in protective clothing with their alarming pocket dosimeters placed just above their ankles and inside their protective clothing in accordance with HP instructions.
The workers entered the reactor cavity to perform some housekeeping activities and to install cavity grating. The APDs were set to alarm at 100 mrem. Upon exiting the cavity area the APDs of all four workers were found to be alarming and to read 113, 279,139, and 107 mrem, respectively. None of the workers heard their APD alarm and, as a result, each of the workers sustained a small unplanned exposure. The inspector noted the use of the APDs was not consistent with Regulatory Guide 8.28, Audible Alarming Dosimeters, Section C.2, in that the dosimeters were used in an area (and in such a manner) where they could not be heard. The licensee stated that
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10 the Technical Specification requf.rement for high radiation area entry was met by supplying continuous HP technician coverage from the refueling floor, however the HP technician also was unaware of the alarming APDs. Exposure control was not based on set stay times, but was dependent on the alarming dosimeters. The control of the four workers by the HP technician was not sufficient to ensure exposures were controlled within RWP limits of 100 mrem per entry. Procedure No. 6.1-022, Section 8.1.6.1, requires all individuals to comply with RWP instructions. This was
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a violation of Technical Specification 6.11, which specifies that procedures shall be adhered to for all operations involving personnel radiation exposure. Due to the issuance of a similar violation issued for failure to use the correct RWP that occurred during the same time frame, this licensee-identified violation will not be cited. This event demonstrated a lack of HP controls to ensure RWP conditions (100 miem) were met.
Another alarming pocket dosimeter-related incident occurred on May 29,1993.
Radiological Problem Report No. 93.0448 described an event in which a worker unintentionally used an APD in a high radiation area without turning it on. The worker had a computer printout 't*cket' indicating that the dosimeter had been turned
"on" by the computer from an earlier log-in entry. The worker used this ' ticket' to interface with the drywell HP technician to sign-in on the appropriate RWP. The worker entered the drywell for approximately 20 minutes. The mistake was noted upon exit. The licensee determined that the root cause was " inattention on the part of the worker" and the worker was counseled and the RPR was closed. Similar mistakes could be made by other workers due to the log-in process in use, and this was another exainple of lack of control of radiological work. The licensee did not address the radiological controls weakness associated with the RWP log-in process in the RPR documents. The drywell entry RWPs required functioning APDs for entry. The individual did not comply with Procedure No. 6.1-022, Section 8.1.6.1, which requires all individuals to comply with RWP instructions. This is a second example of a Technical Specification 6.11 violation that specifies that procedures shall be adhered to for all operations involving personnel radiation exposure. Due to the issuance of a similar violation issued for failure to use the correct RWP that occurred during the same time frame, this violation of Technical Specifications will not be cited. This is an example of a lack of HP control to ensure equipment was used properly.
In summary, the licensee's program for radiological problem reporting is strong and includes a process for thoroughly res. ewing and trending radiological incidents. The problem report group was authorized to impose corrective actions to address station-wide issues. The station continues to have incidents of radiation protection procedure violations and non-compliances with radiation work permit requirements. However, corrective actions have focused on worker accountability and additional training, i
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I without identification of potential, less-than-adequate health physics controls being a factor in the incident. Additional attention to establishing more comprehensive effective corrective actions, including improving HP controls, is warranted.
7.0 External Exposure Controls
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The licensee reported 1993 collective personnel exposures of 394 person-rem through August 27,1993 versus an ALARA goal of 405 person-rem through the end of the year. The ninth refueling outage occurred between April 3,1993 and May 30,1993 resulting in 326.5 person-rem versus an outage goal of 278 person-rem. 'Ihe approxi;nately 48 person-rem overrun was due to refueling support (24 person-rem over budget), operations support (10 person-rem over budget), health physics support
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I (8 person-rem over budget), and maintenance services (7 person-rem over budget).
The licensee estimated that 51 person-rem was saved due to temporary shielding installation. This was the lowest exposure attributed to a Pilgrim refueling outage since the first refueling outage in 1974. Enhanced maintenance planning and scheduling controls has improved outage efficiency. Historically, Pilgrim station refueling outages have run significantly longer than those of other boiling water reactors. Reducing the outage duration to 57 days was a major factor in significantly reducing the cumulative outage exposures.
Through numerous tours of the station, the inspector made the following observations.
The radiological work areas were posted as required. Some improvement has been made in posting general area dose rates in areas of elevated dose rates and in the use of "Do Not Imiter" signs in several of these areas. Use of these signs in combination with HP briefings helps to provide the workers with the knowledge needed to keep their exposures ALARA. No posting discrepancies were noted.
During the inspection, a shon unplanned outage occurred. The inspector observed the preparation, HP briefing, and departure of two individuals who made an initial entry into the primary containment (drywell). RWP No.93-272 was written appropriately for this entry and required the use of self-contained breathing apparatus and full protective clothing. An operator and an HP technician worked as a team surveying the major drywell elevations for steam leaks, radiation levels, and oxygen content. In accordance with precautions for an area classified as Immediately Dangerous to Life or Health (IDLH), they were cognizant of the need to maintain visual contact with one another, and they made regular page system calls to the outside receiving team.
The licensee maintained an additional operator and HP technician fully dressed out in standby should they be needed. Oxygen content was measured as well as gamma and neutron radiation fields. An industrial safety specialist had evaluated the drywell temperature (104*F) with respect to worker safety as well. In general, the inspector i
determined that the initial containment entry was well performed, however the lack of continuous communication with the entry team was apparently not consistent with Occupational Health and Safety Administration (OSHA) guidance (29 CFR I
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I 1910.134(e)(3)(i)). This matter was reported by the inspector to OSHA on October j
15,1993 in accordance with the Memorandum of Understanding between the NRC l
and OSHA. It should be noted that the practice of using the plant paging system intermittently for drywell entries under IDLH conditions was discussed in IE '
Information Notice No. 81-26, Part 4, and was described as an unacceptable practice.
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The inspector reviewed select aspects of the licensee's dosimetry program. The
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licensee did not process their own thermoluminescent dosimeters (TLDs), but obtained i
these services through Yankee Atomic Environmental Laboratory (YAEL). Through
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the inspection of licensee-provided records, the inspector verified that YAEL was l
accredited by the National Voluntary laboratory Accreditation Program (NVLAP)
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through October 1,1993 for processing the Panasonic Model UD814 TLD for categories 1-7 (beta and gamma radiation field types) and accredited for processing
the Panasome Model UD814 TLD for category 8 for neutron radiation monitoring.
l The inspector reviewed the latest quality assurance (QA) audit performed at YAEL by j
a five-utility consortium between September 21 - 23,1992.~ The audit was a very
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technical and detailed evaluation with no major findings reported. The inspector also i
reviewed the latest independent interlaboratory comparison report of YAEL's TLD
processing performance, which was performed by Battelle North West Laboratories,
-l dated August 12, 1993. Very good results were reported, indicating continuing.
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quality personnel exposure results were produced.
p The inspector reviewed non-TLD-determined external dose assignments to hcensee -
l personnel exposure records during'1993. The licensee demonstrated reasonable and j
conservative assumptions with clear documentation for each' dose assignment. 'Ihe j
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inspector checked further to ensure the calculated dose assignments were translated j
l into the official dose records for each affected employee. In each case, record ~
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l transcriptions had been made and accurate personnel exposure reco.ds were found.
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The inspector reviewed licensee records to ascertain whether personnel exposure.
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l accordance with 10 CFR 19.13 requirements. ' The inspector selected at random l
l approximately 50 recently terminated employees and verified that personnel exposure i
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In general, the inspector noted a well run dosimetry group, that appeared to provide
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all of the services necessary to track and report station personnel exposures m a
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timely fashion.
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8.0 Internal Exoosure Controls
The inspector reviewed the results of all air sample results collected between January 1,1993 and July 29,1993, which included the last refueling outage. There were 49 i
air samples with results greater than 0.25 maximum permissible concentration (MPC),
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three air samples were greater than 6 MPC, and one air sample measured 190 MPC.
This last air sample resulted in the assignment of 2.53 MPC-hours to one individual i
l who was wearing a filter cartridge respirator. Based on air sample results, the applicable RWP sign in sheets of worker entries into the airborne radioactivity areas, and due to the protection afforded by the use of respiratory protection equipment, no other MPC-hours were assigned. This indicates that, in all cases for the time period of January 1,1993 through July 29,1993, for air samples that represented airborne radioactivity areas, the affected individuals had been provided respirators and i
effectively minimized the internal exposure hazards.
To evaluate the adequacy of the air sampling program the inspector reviewed bioassay results. During the subject review period, there were a total of 17 bioassay results that were greater than 10% maximum permissible body burden (MPBB). These bioassays were conducted as the result of personnel contamination alarms while the individuals were leaving the RCA. These followup bioassays may indicate a few missed air sampling opportunities; however, external personnel contaminations also were represented, and therefore no conclusive assessment could be made. Final internal exposure assignments for this time period (21/3 quarters) totaled 61.9 MPC-hours collectively, and a maximum of 17.1 MPC-hours assigned to an individual.
Federal regulations limit internal exposures to 520 MPC-hours per quarter per individual. The station's performance in limiting internal exposures during 1993 has been excellent.
8.1 Respiratory Protection The inspector reviewed the respiratory protection program with respect to 10 CFR. 20, NUREG 0041, and ANSI Z88.2. The inspector reviewed the respirator storage and processing facility and applicable records.
The inspector identified that the following respiratory protection equipment was
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available for issue and use at the station. Applicable National Institute for Occupational Safety and Health / Mine Safety and Health Administration (NIOSH/MSHA) certifications were substantiated for all equipment.
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1 RESPIRATORY PROTECTION EQUIPMENT NIOSH/MSHA CERTIFICATION MSA Half Masks Comfo II TC-23C-47
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r MSA Ultravue 7-203-1 TC-21C-150
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MSA Filter Cartridges, Type H TC-23C-155 j
MSA Chemical Cartridges, Type GMA TC-23C-40C
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Scott-o-ramic 801450-40 TC-21C-149 Scott Filter 604100-50 TC-21C-149
Scott Presur-Pak 4.5 TC-13F-76, 212, %
l Driger Panorama Nova RA TC-21C-497 l
Driger Filter 4052828 TC-21C-497
Bubble Hoods, Defense Apparel HSQ-10 TC-19C-120
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MSA Filter 463284 TC-21-186 l
3M Acid Gas Respirator 8714 TC-23C-129 3M Organic Vapor Respirator 8712 TC-23C-123 Respirator fit testing of personnel was accomplished with a TSI PortaCount Model 8010 air particle detector. The licensee owns three instruments, with two in service and one shipped off for calibration at any one time. The inspector verified current TSI factory calibrations on the two in-service air particle detecting instruments. The
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inspector reviewed documentation dated May 19, 1993, certifying that the service air supply and bottled air compressor produced Grade D quality air, as defined by the Compressed Gas Association. The respirator processing and storage area was
reviewed and found to be a well maintained facility that provided quality safety equipment. No discrepancies were noted in this area.
8.2 Whole Body Counting The inspector reviewed the whole body counting operation and calibration status with respect to ANSI N343-1978, ANSI N42.12-1980, and ANSI N42.14-1978. The inspector reviewed calibration records, whole body count records and logs, and discussed whole body counter operations with the on-duty HP technician.
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The licensee utilizes a Canberra Fastscan sodium iodide whole body counter for routine in-vim bioassay measurements. The inspector reviewed appropriate documentation of the energy and efficiency calibration that was perforr.ed on December 16, 1992. The calibration was performed with a National Institute for Science and Technology (NIST)-traceable mixed gamma source. The calibradon data indicated an operating voltage of 700 Volts with a gain setting of 7.5. The inspector noted an as-found gain setting of 8.28. The change in gain was a result of daily adjustments to compensate for energy spectrum shifts. Recalibration would be performed upon failure of the routine source measurement to fall within three standard deviations of a known value.
The licensee utilizes a Canberra High Purity Intrinsic Germanium (HPIG) single detector body counter for investigational whole body counts. This instrument was last calibrated on July 1,1993 for three geometry configurations: lung, gastrointestinal tract, and thyroid. A NIST-traceable europium-152 source was used for these
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t calibrations. The calibration results indicated a detector operating voltage of 3500 Volts and a gain of 5.84.
The inspector discussed daily whole body counter routine background and source checks and the inspector determined the competency of the operator to perform the necessary routine instrument operations on the Fastscan unit. With regard to investigational body counting with the HPIG counter, the operating technician was not familiar with correct operating procedures of this instrument. The HPIG detector operates at near cryogenic temperatures through the use of a liquid nitrogen bath. On the external surface of the detector housing there was a Canberra factory sticker that read 4500 Volts. Since the body counter is not in use routinely, the detector is kept at room temperature (to conserve the liquid nitrogen supply) and the voltage to the detector is turned off. When asked how an investigational body count was performed, the operating technician explained that the detector would be bathed in liquid nitrogen and the voltage would be turned up to 4500 Volts (as indicated on the detector). The count would be started promptly. The calibration data indicated an operating voltage of 3500 Volts and a gain setting of 5.84. This information was not available in the whole body counting room, but was located in an office in another building inside the plant. Standard practice in initial startup of a germanium counting system includes a four to eight-hour cool down time for the detector crystal to stabilize pnor to counting. The inspector determined that the operating technician was not properly trained to correctly operate the HPIG body counter and important reference information (current calibration voltage and gain settings) was not available in the whole body count room. These are considered weaknesses in the internal exposure control program.
The licensee agreed to ensure that the necessary operating parameters of the whole body counters would be made available in the whole body count room. The licensee indicated that the training program was good, but not timely, as HP technicians are
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rotated through a temporary duty in the whole body count room without a whole body l
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counter operation training refresher prior to accepting the responsibility for
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performing bioassay measurements. The licensee agreed to review these pmctices and
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provide for the current whole body counter operator training inadequacies
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immediately. The inspector will review the licensee's' corrective actions in a future inspection (IFI 50-293/93-16-06).
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9.0 Exit Meetine
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The inspector met with licensee representatives at the end of the inspection, on :
September 3,1993. Additional inspection continued in the NRC Region I office
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through October 15 for review of that information and for discussion and evaluation j
i with OSHA of the industrial safety practice discussed _in Section 7.0 of the report.
Additional inspection results were communicated to a licensee representative by.
telephone on September 9,1993.
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