IR 05000293/1993010
| ML20045D424 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/15/1993 |
| From: | Noggle J, Pasciak W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20045D412 | List: |
| References | |
| 50-293-93-10, NUDOCS 9306290016 | |
| Preceding documents: |
|
| Download: ML20045D424 (12) | |
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i U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
10-293/93-10 j
l Docket No.
50-293 License No.
DPR-63 Licensee:
Boston Edison Company RFD #1 Rocky Hill Road
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Plymouth. Massachusetts 02360
Facility Name:
Pilgrim Nuclear Power Station Inspection At:
Plymouth. Massachusetts
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Inspection Conducted:
May 10 - 14.1993-1 Q
[p $/f3 Inspector:
J. Nokp4e, Sen(dr] Radiation Specialist Date Facilities Radiation Protection Section e
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d Yd f8 Approved by:
W. Pasciak, Chief, Facilities ]
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Date Radiation Protection Section Areas Insnected:
Areas covered in this inspection-included a review of:
audits and surveillances, collective personnel exposure status, outage radiological controls, and the implementation of the internal and external exposure radiological programs during a refueling and maintenance outage.
Ecluks: The licensee has demonstrated generally strong performance in the areas inspected.
Significant resources have been devoted to shielding various high radiation sources to reduce
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personnel exposures, although some program weaknesses in this area were identified. Two
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violations were identified relative to an entry into a high radiation area on April 24, 1993 (Section 6.0). It appears that them continues to be problems with procedure violations and non-compliances with radiation work permit requirements.
9306290016 930617 PDR ADOCK 05000293 PDR O
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DETAUS i
R 1.0 Persons Contacted
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1.1 Boston Edison Company
- J. Bellefeuille, Deputy Plant Manager
W. Caithness, Supervisor Radiological Operations
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M. Christopher, Supervisor Radiological Operations S. Chugh, Corporate Engineer
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W. Coady, ALARA Planning Specialist R. Cook, Problem Report Coordinator L. Dooley, Technical Training Section Manager i
R. Fairbank, Corporate Engineering Manager
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- ' F. Famulori, Quality Assurance Department Manager l
G. Gordon, ALARA Planning Specialist
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S. Landahl, Radiological Operations Support Division Manager
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J. McClellan, Senior Quality Assurance Engineer
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- W. Munro, Senior Compliance Engineer
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R. O'Neill, Technical Programs Division Manager G. Reidinger, Maintenance Planning Division Manager
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- W. Rothert, Technical General Manager
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L. Sasser, Senior Radiological Protection Technician
- L. Schmeling, Plant Manager -
'l G. Vasquez, Lead Radiological Engineer
J. Walker, Problem Report Coordinator
- L. Wetherell, Radiological Section Manager D. Zembel, Maintenance Section Manager 1.2 USNRC Personnel
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- R. Eaton, Nuclear Reactor Regulation Pilgrim Project Manager
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- Denotes attendance at the exit meeting on May 14, 1993.
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2.0 BLync This inspection was an announced safety inspection of the Pilgrim Nuclear Power Station radiation control programs during a scheduled refueling and maintenance outage.
3.0
. Organization The Radiological Section has maintained a stable organization since the last inspection
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at a permanent staffing level of 74, consisting of 35 HP technicians, 25 technical suppon personnel,10 supervisors, 3 division managers, and the Radiation Protection Manager.
There were 60 additional senior HP techuicians and 29 decontamination technicians brought in for the outage, which began on April 3,1993. This was considered to be a relatively small number of additional contractor HP technicians to cover outage work activities when compared to similar nuclear power plant outages; however, the.
technicians appeared to have been used efficiently and no adverse effects on program performance were observed.
4.0 Training and Oualifications The licensee's training and qualification program for contractor HP technicians was myiewed through discussions with personnel and through the review of qualification
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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.
Through a sampling of contractor HP technician msumes, the licensee was found to be in full compliance with the experience requirements of ANSI N18.1-1971. The inspector i
also examined the training program for the temporary contractor HP technicians. The licensee has adopted the use of the Middle Atlantic Nuclear Training Group (MANTG)
HP technician fundamentals screening exam to qualify HP technicians for generic HP
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technician msponsibilities based on a minimum 80% score on the exam. A site specific i
training course was provided, consisting of approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of classroom training on station procedures, to fully qualify HP technicians to perform radiological safety work at Pilgrim Station. The inspector reviewed the msults of both the MANTG generic HP.
exam and the Pilgrim site specific qualification sign-offs for selected HP contractor technicians and found all conditions for qualification had been met.
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5.0 Audits and Surveillances The latest licensee audit (No. 92-11) of the radiological protection program was conducted between September 19,1992 and October 16,1992. The audit team included j
a non-liccasee technical specialist as well as licensee technical specialists, all of whom
were independent of the audited program. Three deficiencies were reponed:
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Failure of personnel to follow instmetions and to myiew surveys as required by the Radiological Work Permit (RWP);
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Failure of personnel to follow posted radiological instmetions and perform proper frisking techniques;
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Failure to perform or record radiation / contamination surveys in accordance with procedures.
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In addition, the audit reported that the work force had less than adequate concern for radiological postings and housekeeping. The audit reported that worker attention toward radiological protection requirements was on a decline.
In response to the audit findings, the licensee developed and implemented a three-hour Advanced Rad Worker Training Course, which was given to 120 constmetion personnel just prior to the outage to help offset some of the radiation worker deficiencies that had been observed. This course addressed such topics as cobalt reduction, causes of facial contamination, instmetion on multi-layered protecdve clothing undress pmcedures, use
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of high efficiency particulate air (HEPA) filter systems, use of catch containments, hot panicle zone procedures, and suggested some radwaste reduction techniques.
In addition to the annual radiation protection program audit, the licensee had generated
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29 quality assurance (QA) surveillances in the radiation protection and radwaste areas
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from January 1,1993 until May 10, 1993. The inspector reviewed all 24 surveillance repons that related to HP areas. No safety significant findings were identified by the surveillances. Various program areas were reviewed and previous deficiencies were l
being followed and closed.
The lack of significant findings reported from these j
surveillances was in contrast to the audit observation that poor radiation worker practices
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q were mereasmg.
6.0 Radiological Problem Reports Approximately one year ago, the licensee combined several station-wide deficiency j
tracking programs into a single problem report program. Radiological problem repons
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(RPRs), as described in procedure No. 6.1-209, Rev. I1, provides an input into the l
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station's problem repon program. It is this program that determines safety significance, assigns the commensurate level of resources for investigation, root cause determination, assignment of responsibility for corrective action determination, acceptance of proposed i
corrective actions, and tracking of open items through closure. The HP section, tracks and trends the number of RPRs that go into the problem report program and reviews the i
corrective actions prior to closure of each issue.
The inspector reviewed the latest Trend Report covering the first calendar quarter of 1993. The licensee generated 46 RPRs during the first calendar quarter of 1993 compared to 79 RPRs written during the entire year of 1992. The report noted a large variety of RPR categories represented with the most common category due to procedure violations, which include contaminated material found outside of the radiological
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controlled area, RWP violations and inadequacies, posting violations, and use and control
of dosimetry. The Trend Report indicates an increasing number of radiological work
practice incidents as compared with 1992. The licensee's Advanced Rad Worker training initiative was not given until March 1993. Therefore, its effectiveness in improving radiation worker work practices may not yet be apparent. The inspector will continue to monitor the licensee's resolution of this concern.
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The inspector reviewed several RPRs in detail. Two RPRs of significance are described
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The RPR No. 93.0322 involved a licensee identified Technical Specification 6.13.1 violation of an unauthorized high radiation area entry. On April 24,1993, three electricians signed onto a general RWP, which did not require an HP briefing of
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radiological conditions, and proceeded across a high radiation area barricade into the "B" i
Quad. The geneml RWP (No. 5011) did not allow work in the area the workers had entered nor did it allow entry into a high radiation area. The general RWP required a
self-briefing of radiological conditions using current survey maps of the areas to be
worked, which was not performed. The workers crossed over what they thought was only a contamination boundary rope (posted as a high radiation area) and proceeded to i
the work location and pulled cable through a wall penetration. The alarming pocket dosimeter of one worker alarmed continuously in the work area and the worker checked-i the dose reading occasionally, but continued to work in the area.
The workers i
encountered a blockage in one of the penetrations and as a result, left the area. After
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leaving the ama they reported the alarming dosimeter condition to an HP technician. The dose received by the worker with the alarming dosimeter was 22 millirem.
The area being worked was subsequently surveyed by an HP technician and the worker closest to the penetration was found to be working in a high radiation area as he was-
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laying on a pipe, which was reading 200 mR/hr contact with the pipe and 120 mR/hr at
30 cm. This was a violation of Technical Specification 6.13.1 as the workers' did not enter the area under the proper RWP.
The inspector reviewed the licensee's immediate corrective actions, which consisted of suspending the job and terminating employment of the three workers involved. The licensee conducted meetings with the construction crafts to explain the Alnor alarming dosimeter fm,ctions and related worker responsibilities, and discussed the seriousness of high radiation area entry requirements. At the time of this inspection, the licensee's i
problem report group expressed dissatisfaction with the corrective actions described l
above and the issue was still open at the end of the inspection. Two violations of j
regulatory requirements associated with this entry are described telow.
t The workers entered on the wrong RWP. It only required a self-briefing of mdiological i
conditions. The three workers entered a posted high radiation area without authorization
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and did not meet the requirements of Technical Specification 6.13.1, which requires that worker entry into high radiation areas is controlled by procedures (RWPs). This is in noncompliance with Technical Specification 6.13.1 (50-293/93-10-01). Had the workers used the correct RWP, an HP technician briefing would have informed the workers of
the dose rates in the work area. As a result, the workers were not informed of the i
radiological hazards associated with their planned work. This is in noncompliance with 10 CFR 19.12, "Instmetion to workers" which requires workers to be instructed in the radiological hazards associated with their work (50-293/93-10-01). While the non-i
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compliances are against separate regulatory requirements, they are aggregated as a single violation because they are the result of the same failure, i.e., to use the proper RWP for the job.
The second violation involved the disregard of the alarming pocket dosimeter which,
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according to HP instructions, required the worker to leave the area and contact HP.
i,cconfing to procedure No.1.3.106, " Conduct of Radiological Operations," when an alarming pocket dosimeter alarms due to exceeding the high dose rate alarm setpoint, the worker shall move to a lower dose rate area or report to radiation protection personnel.
In General Employee Training, workers are instmeted to leave an area if an alarming
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dosimeter continuously alarms in the area.
Technical Specification 6.11 requires procedures for personnel radiation protection to be approved, maintained and adhered to for all operations involving personnel radiation exposure. Failure to adhere to procedure No.1.3.106 is a violation (50-293/93-10-02).
Another significant RPR that was reviewed by the inspector had not been resolved by the licensee at the time of this inspection. The RPR No. 92.0224, is a conriktion of several recurring instances of finding contaminated tools outside of the RCA. bince 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 were recurring instances of contaminated tools being found in " clean" tool boxes. Due to the chronic lack of contaminated tool control, the licensee
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raised the level of attention of this issue and has combined the individual contaminated.
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tool incidents into this one RPR. The licensee has now established a contaminated tool r
issue depot inside the RCA in an attempt to meet the demand for tools in the RCA and reduce the number of tools into and out of the area. This tool control change has been.
determined by the licensee as appropriate to correct the problem, however, due to the
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magnitude of the repeat occurrences, the licensee has decided to perform a complete t
investigation, a root cause analysis and a re-review of the corrective actions before
closing this issue. The inspector will continue to follow -
resolution of this issue in
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a future inspection.
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The inspector reviewed the station practices and procedures for release of material from the station and was satisfied that sufficient controls existed to prevent the inadvertent release of contaminated material from the station.
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In summary, the licensee has a strong radiological pmblem report program that includes a process for thoroughly reviewing and trending radiological incidents. The problem
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report group is able to impose corrective actions to address station-wide issues. It
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appears that there continue to be problems with procedure violations and non-compliances with radiation work permit requirements. Additional attention in establishing effective
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corrective actions for identified problems is warranted.
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7 7.0 ALARA The licensee has established a 1993 personnel exposure goal of 405 person-rem, consisting of 278 person-rem of outage exposure,116 person-rem during non-outage periods, and 11 person-mm for contingency or emergent work. By May 13,1993, the licensee had accumulated 56 person-rem of non-outage exposure, and 232 person-rem from the outage for a total of 288 person-rem by day 41 of a 59 day outage. Pilgrim Station's previous refueling outage year in 1991 resulted in total annual exposures of 605 person-rem. At the current rate of dose accumulation, this outage year appears to be appmaching the same value.
7.1 Shielding
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As a result of a management commitment to reduce doses to station workers, the licensee I
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personnel exposures. In spite of the strong effort to provide temporary shielding for the j
workers, there were a few missed opportunities for dose reduction. The inspector toured areas inside the drywell and the reactor water cleanup (RWCU) heat exchanger room and noted some locations where shielding could be beneficial 7.1.1 Drvwell Shielding General area dose rates inside the drywell appeared to be high considering the j
significant shielding installations in the area. The inspector noted the installation of temporary shielding included portions of the recirculation piping system and the RHR piping systems. The inspector also noted the lack of continuity between shielded sections and not enough shielding to substantially mduce general field dose rates in other locations. In particular, the recirculation riser pipes were shielded at the elbow near the N2 penetration nozzles, but the vertical riser pipes
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that connect to the elbows were not shielded. Water bladder shields were i
installed at the base of these risers at their junction with the ring header piping l
and the ring header itself was dmped with lead blankets. The unshielded vertical riser pipes were measured by the inspector to be 400 - 600 mR/hr at contact and there are 10 of these pipes located around the 51 foot elevation of the drywell.
Due to their unshielded condition they contributed to the general background dose rates of approximately 100 mR/hr for this work area.
j On the drywell entry level (23 foot elevation), the large recirculation suction and discharge piping significantly contributed to the general area dose rates. These vertical recirculation pipes had been partially shielded by the licensee by hanging lead blankets from a U-shaped frame attached to structural members above. The shadow shielding covered the piping 6 feet above the grating level while the high radiation recirculation piping continued vertically overhead. The inspector noted
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a mduction in general ama dose rates in the vicinity of the shielding at waist
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level, however, at the upper chest and head level, the general dose rates were as if the piping system was unshielded, at approximately 100 mR/hr.
The principal means of installing lead blanket shielding was to wrap it around
existing piping and to secure it with clastic bungy cords. In some cases, there was a significant amount of lead shielding hung in the overhead secured with
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bungy cords to prevent the shielding from falling on personnel and on safety related components below. This was identified as a potential industrial safety
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hazard. The licensee stated that in the futum, significant shielding installations
would be inspected by engineering or quality assurance personnel for adequacy
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of restraints. Existing shielding installations will be removed prior to plant stanup.
i 7.1.2 RWCU Heat Exchanger Shielding The inspector also toured the RWCU heat exchanger room where some pipe replacement work had occurred during the outage.
The mgenerative heat cxchangers were draped with one layer oflead blankets, the non-regenerative heat exchangers were draped with 2 layers of lead blankets and an overhead section of RWCU piping was also wrapped in 2 layers of lead blankets. The inspector
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observed that a significant amount of work had been performed directly alongside the regenerative heat exchangers v ' tre the RWCU outlet piping had been
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removed and replaced. The regener
/e heat exchangers were surveyed by the
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inspector and found to be appro:r
!y 400 mR/hr on contact. Through the i
attenuation of one layer of lead F m
(%" equivalent lead thickness), the area where the pipe replacement e, cred was approximately 80 - 160 mR/hr. In the
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area of the non-regenerative hat exchangers, two layers of lead blankets nduced
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a 60 mR/hr heat exchanger radiation source down to 30 mR/hr contact resulting
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in 35 mR/hr in this general area. The majority of the work had been performed near the regenerative heat exchangers, yet more shielding had been applied to the non-regenerative heat exchangers. Funher, the dose rates fmm the regenerative heat exchangers were higher than from the non-regenerative heat exchangers.
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This reflected a weakness in the utilization of shieldmg.
j When the inspector informed the ALARA specialists of these observations, they inspected the RWCU heat exchanger room and found that the two layers of blankets that had originally been installed had been reduced to one layer, probably due to job interferences. No one in the HP section or the ALARA planning group had authorized the removal of the shielding and a higher than planned dose i
rate field existed for some unknown time period. Based on the May 14, 1993 outage dose repon, RWP No. 20?7 for replacirg RWCU outlet piping was 20%
overbudget on dose expenditures (currently at Ic.2 person-rem). HP technicians were required to survey the RWCU heat exchanger room daily when worked and to brief workers on dose rates, however, no one noticed the higher than normal
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working dose rate field or raised the issue. The ALARA planning group had made several reviews to investigate the higher accumulated doses and identified the large scope of scaffold building and some weld rework problems. During the
previous H.P inspection in this area, the inspector had also noted similar inadequacies in the ALARA shielding program, which continue to exist. It appears that there were some communication weaknesses that led to the shield installation weaknesses.
7.2 ALARA Proeram The ALARA planning group consisted of two ALARA specialists who report to the
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maintenance planning supervisor. Close association with maintenance planning has l
provided the ALARA planning group with advance notice of station maintenance requests l
to allow planning for the necessary shielding or other ALARA engineering controls that are prescribed for specific jobs. While the ALARA planning group does get man-hour data for specific jobs, they are not always able to obtain this data relative to plant locations, which would allow shielding resource prioritization decisions to be made.
The inspector reviewed the four ALARA procedures that constitute the current ALARA pmgram:
Procedure No. 3.W.1-1, Rev. 2, "ALARA Work Review,"
Procedure No. 3.W.1-2, Rev. 0," Station ALARA Performance,"
Procedure No. 3.W.1-3, Rev. 0,"ALARA Engineering Controls," and Procedure No. 3.W.1-4, Rev.1,"ALARA Planning Assessments" The inspector determined that these procedures contained most of the fundamental elements for an effective ALARA program, however, there was a lack of shielding design acceptance criteria.
The inspector also reviewed procedure, No. NEDW 390, Rev. 0," Radiation and Contamination Reduction at Pilgrim Nuclear Power Station." Also, the ALARA training program for engineers that was based on this procedure was also reviewed. In general, the training program addresses certain ALARA design considerations for construction of
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new facilities. It was detennined to have less of an application to an operating facility, i
except perhaps with respect to major plant modifications.
The inspector reviewed the shielding design documentation with associated engineering service requests provided by the ALARA planning gmup and engineering responses for the drywell and RWCU heat exchanger shielding designs. The inspector also discussed 3 Inspection report No. 50-293/92-24
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these issues with the cognizant engineer and engineering manager. The inspector made the following findings:
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The engineering shielding requests were written by the ALARA planning group in terms of appmval for weight loading of plant systems or structures. There was no shield value relayed to the engineer, either in dose savings or in dollar values, and the shielding desired was not adequately characterized to allow development of an engineered design.
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The Nuclear Engineering Department Work Instruction No. 239, Rev. 6," Criteria for Evaluation and Appmval of Radiation Shielding," contains only limited guidance for providing the engineering calculations necessary to approve a significant shielding design.
No guidance is provided for minor shielding applications.
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Significant shielding designs are normally routed through civil engineering, mechanical engineering, and systems and safety assessment engineering prior to obtaining final approval. Due to the time constraint involved in such a matrix approval process, long lead times are requimd for these approvals. This process does not accommodate the iterative shielding design approach when outage time is shon.
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In-field engineering review is currently not part of the process. The licensee does have plans for relocating the engineering group from Braintme to Pilgrim Station upon completion of a new building that is currently under constmetion.
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As mentioned above, there were many scaffolds erected which had to be taken down prior to RWCU system stanup. Some scaffold platforms were evidently crected to reach several motor operated valves to allow for routine maintenance.
There are no permanent platforms installed for these valves. Instead, repetitive scaffold work is perfonned in high radiation areas. No evaluation has been perfonned to address the permanent platform requirements in these areas in order to reduce scaffolding erection needs and to reduce exposures. The licensee -
indicated plans to review this issue.
The licensee agreed to review the extent of engineering involvement necessary for shielding design and installation in order to more clearly establish the need for engineering resources.
The licensee has provided a significant number of ALARA reviews forjobs in-progress.
The need for ALARA job reviews were partly determined by the daily outage dose repon. When the accumulated dose for a particular job exceeded the estimated dose by greater than 25 % or, if the ratio of percent of estimated dose used versus percent ofjob completion was greater than 1.25, then an ALARA in-process myiew was perfonne _
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The ALARA planning gmup actually performed a great deal more in-process reviews than was required. This was a good tool that allowed the licensee an opportunity to cormet a dose overrun situation before the work was completed. These reviews also pennitted the licensee an opportunity to gain experience and apply it to ongoing and future jobs. This was a very good ALARA program element.
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In summary, the ALARA shielding effort was strong. Enhancements could be made to increase the communication between health physics, ALARA planning, work scheduling,
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engineering, and the workers in order to maximize shielding effectiveness. The ALARA l
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program procedures used by the ALARA planning gmup were well developed except for the lack of any shield design acceptance criteria. The engineering shielding procedure
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consists only of an elemental procedure for evaluating shielding proposals and, as a result, does not always provide adequate guidance. In addition, the engineering review process does not always match the short tenn outage needs for in-field shielding. Other
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ALARA shielding pmblems are also described above.
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8.0 Ooerational HP The outage opentional HP group of approximately 35 permanent HP technicians was supplemented with 60 temporary contractor HP technicians. The permanent HP staff filled the lead HP organization positions. There wem several satellite HP control points
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established with HP technician assignments listed below for a work shift.
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" Red line" RCA access point, 2 HP technicians assigned
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Refueling floor,4 HP technicians
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First floor of the mactor building,4 HP technicians
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Drywell, 4 HP technicians
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Reactor water cleanup level of the reactor building, 3 HP technicians
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Turbine operating deck, 3 HP technicians
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Turbine and Auxiliary bay doors,2 HP technicians
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Decontamination building, I HP technician
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" Balance-of-plant", 3 HP technicians Air sample records were reviewed for selected days during the outage. On average, 60
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air samples were taken during each day of the outage. This number is relatively low, and may result from the small number of outage HP technicians, small radiological workscope during the outage, or other factors. Based on inspector observations, the adequacy of the air sampling program observed during the inspection week was acceptable considering the radiological work requirements. From January 1,1993 through May 14,1993, the licensee had recorded 42 air samples that were greater than
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t 0.25 maximum permissible conec.itration (MPC)2 In all of these cases respiratory protection devices were worn. There was only one case when MPC-hours were assigned because the protection factor of the respirator was exceeded. This occurred during solid CO grit blasting of reactor cavity grating components, when an air sample of 190 MPCs
was obtained. This decontamination work was performed inside a containment structure, however, the HEPA ventilation unit was not properly sized to meet the airborne contamination generation capability of the decontamination method. The decontamination worker was in the area for approximately 40 minutes wearing a full face filter respirator and was assigned 2.53 MPC-hours.
This was the only case of internal exposure
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assignment based on air sample results for the year to date. The maximum MPC-hour assignment for the outage has been 16.4 MPC-hours (as assigned from bioassay results).
In general, the air sampling program was considered adequate for the outage.
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The radiological work areas were posted as required. However, no dose rate information
or hot spot signs posted in the areas to warn workers of radiological hazards. The licensee relies on the effectiveness of HP briefings to workers to communicate the
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location and level of radiological hazards. General RWP requirements call for a daily work area survey to be performed when work is to be performed.
While full i
documented surveys are not always performed on a daily basis, daily verifications of conditions am performed, thereby meeting the intent of the RWP requirements. The HP log book entries were often used to document initial job start " verification" of radiological conditions.
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In general, the operational HP performance for the outage was strong. Sufficient HP technician oversight was provided for the significant work areas and sufficiently detailed radiological controls were generally prescribed and enforced.
Area postings and
housekeeping were good. No violations were noted.
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9.0 Exit Meeting
The inspector met with licensee representatives at the end of the inspection, on May 14,
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i 1993. The inspector reviewed the purpose and scope of the inspection and discussed the
i findings. The licensee acknowledged the inspection findings.
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2.;>_0.25 MPC defines an airborne radioactivity area
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