IR 05000293/1981021
| ML20042B519 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/18/1982 |
| From: | Knapp P, Nimitz R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20042B511 | List: |
| References | |
| 50-293-81-21, NUDOCS 8203250388 | |
| Download: ML20042B519 (31) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-293/81-21 Docket No. 50-293 License No. DPR-35 Priority Category C
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Licensee:
Boston Edison Company M/C Nuclear
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800 Boylston Street Boston, Massachusetts 02199 Facility Name: Pilgrim Nuclear Power Station Inspection at: Plymouth, Massachusetts Inspection conducted: August 31-September 4 and September 28-October 2, 1981 Inspector:
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_._R L. Nimitz, RadJation Specialist (date)
'M PN Approved by: -d i'
Nrew p P!J.Knapp, Chief,FagityRadiological (date)
Protection Section Inspection Summary:
Inspection on August 31 - September 4, 1981 and September 28 - October 2, 1981 (Report No. 50-293/81-21)
Areas Inspected:
Routine, unannounced inspection by one regional based inspector of the Radiation Protection Program during refueling including:
licensee action on previous inspection findings, confirmation letter follow-up, train-ing and retraining, ALARA program, Health Physics Appraisal follow-up, radioactive and contaminated material control, high radiation area posting and control, review of high TLD badge reading, and procedures. The inspection involved 60 inspector-hours onsite by one regional based inspector.
Results: Of the 9 areas inspected, no items of noncompliance were identified in 5 areas; 3 items of noncompliarce were identified in 4 areas (Failure to post notices to workers as required t'y 10 CFR 19.11, Paragraph 4.b; Failure to adhere to radiation protection procedures (2 instances) as required by Technical Specification 6.11, Paragraphs 9d and 10; Failure to post a High Radiation Area in accordance with Technical Specification 6.13, Paragraph 8).
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DETAILS 1.
Persons Contacted i
Boston Edison Company
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(1) W. Armstrong, Deputy Manager, Nuclear Operations E. Grahm, Senior Plant Engineer
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(1) F. Gidello, Senior Plant Engineer R. Kuhn, Senior Radiological Engineer (1) (2) R. D. Machon, Nuclear Operations Manager (1) (2) C. J. Mathis, Deputy, Nuclear Operations Manager J. J. McCann, Watch Engineer l
D. L. Myers, Senior Planning Engineer
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W. F. Olson, Senior Nuclear Training Specialist E. O'Rorke, Watch Engineer
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R. A. Purdy, Radioactive Waste Co-ordinator (1) K. Roberts, Chief Maintenance Engineer (1) (2) P. D. Smith, Chief Technical Engineer (2) T. Sowden, Environmental and Radiological Health and Safety Group Leader (1) (2) A. R. Trudeau, Chief Radiological Engineer (2) E. Zemanski, Management Services Group Leader F
i Nuclear Regulatory Commission-(1) J. R. Johnson, Senior Resident Inspector Pilgrim Station
(2) H. Eichenholtz, Resident Inspector Pilgrim Station (1) denotes those persons attending the preliminary exit. interview on September 4, 1981 (2) denotes those persons attending the final exit interview on October
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2, 1981.
The inspector also contacted other personnel including members of the licensee's operations, maintenance, radiological controls and training organizations.
2.
Licensee Action on Previous Inspection Findings a.
(Closed) Noncompliance (50-293/80-05-04) Failure to maintain administrative control of high radiation area access keys. At
various times during the inspection, the inspector reviewed
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individual high radiation area access key accountability with
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respect to Pilgrim Nuclear Power Station Procedure No. 6.1-012, a
Revision 6, Access to High Radiation Areas, dated May 6, 1981. The i
review indicated the on-watch Health Physics technician was controlling issuance and accountability of keys, audits of keys were
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being performed at shift turnover, and the Health Physics Supervisor
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was reviewing the key accountability form for discrepancies.
b.
(0 pen) Noncompliance (b0-293/80-05-05) Failure to post high radia-tion areas.
Inspector tours of the controlled areas identified one high radiation area that was not posted (Details, paragraph 8).
c.
(Closed) Noncompliance (50-293/80-05-11) Failure of personnel to
monitor for contamination.
Inspector tours of the controlled areas and observation of control point monitoring stations indicated guards had been stationed at the major exit points to ensure proper frisking. The observations indicated personnel were properly frisking for contamination.
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d.
(0 pen) Noncompliance (50-293/80-05-09):
Failure to review and
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evaluate MPC-hours.
Inspector review of work involv:.,g removal of reactor internals and licensee calculation and evaluation of MPC-hours sustained therefrom indicated incorrect air sample data were being used to generate MPC-hour data. The licensee was crediting individuals with MPC-hour exposure not sustained.
The licensee's Chief Radiological Engineer plans to implement an MPC-hour sign-in sheet to ensure appropriate air samples and stay times are correlated properly, e.
(0 pen) Noncompliance (50-293/80-05-10):
Failure to have a procedure detailing precautionary actions to comply 111th the 10 CFR 103(b)(2)
40 MPC-hour control measure.
Inspector review of Pilgrim Nuclear Power Station Procedure 6.1-021, MPC Hours Determination, indicated the procedure has been revised to include MPC-hour alert and action points.
Inspector discussions with cognizant licensee representatives indicated training was given to appropriate personnel. Due to the licensee's microfilming of records, the inspector was unable to review training documentation.
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(0 pen) Noncompliance (50-293/80-05-08):
Failure to prevent recurrence of personnel exposures to airborne radioactivity in excess of 40 MPC-hours.
Inspector review of the licensee's actions taken indicated the licensee had revised a procedure (No. 6.1-021) to include MPC-hour alert and action points.
The review indicated that no procedural guidance for actions had been established (e.g., evaluation and implementation of corrective measures to preclude recurrence following identification of an exposure in excess of 40 MPC-hours.)
3.
Confirmation Letter Follow-up The inspector reviewed licensee implementation of actions to be under-taken as described in letters dated February 10, 1981 and July 27, 1981 from B. H. Grier, then Director of NRC Region I, to J. E. Howard,
Vice-President Nuclear, Boston Edison Company.
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Letter of February 10, 1981 This letter dealt with licensee actions following a January 17, 1981 resin spill at the Pilgrim Nuclear Power Station. The inspector review of the licensee's actions with respect to the above letter indicated the following:
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the licensee revised the condensate demineralizer system operating
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procedure (No. 2.2.97) on May 20, 1981 to identify resin hopper valves and to include a valve check-off list.
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the licensee, trained appropriate operating personnel on August
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2, 1981, in the above procedure change prior to new resin
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addition on August 3, 1981.
the licensee initiated changes to the condensate demineralizer
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resin regeneration system piping and instrumentation drawing on February 26, 1981 to reflect the resin addition hopper piping.
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A subsequent review of the resin iiopper piping by licensee
personnel identified additional drawing changes to be made, i.e., adding a valve to the drawing. A second drawing change was initiated on March 27, 1981.
the licensee did not perform a walk-down of all accessible
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portions of the condensate demineralizer resin regeneration system and did not revise system piping and instrumentation drawings to reflect this walk-down by September 1,1981 as stated in the above letter.
Regarding the system walk-down, based on discussions at the preliminary exit interview on September 4, 1981, the system walk-down and initiation of drawing revisions was to be completed by September 11, 1981.
The inspector review of this item on October 1, 1981 indicated the system walk-down had been performed on September 10, 1981. However, two system variations were identified and no revision to system drawings, as of October 1,1981, was initiated. This was brought to the licensee's attention at the final exit interview on October 2, 1981.
Based on review of a memorandum dated October 6, 1981, a change to the system drawings reflecting the walk-down was initiated October 2, 1981.
Additional inspector review of the licensee's actions taken with respect to the February 10, 1981 letter indicated the following:
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A policy statement was issued February 11, 1981 to all Pilgrim Nuclear Power Station personnel to provide guidance on personnel protection when repair / corrective action is performed during a radiation emergency.
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A procedure change natice was issued on February 2,1981 to require evaluation and appropriate action for personnel safety prior to taking actions which may be hazardous.
The changes to the system drawings referenced above will be reviewed during a subsequent inspection.
(50-293/81-21-09)
b.
Letter of July 27, 1981 This letter dealt with licensee actions so implement post accident effluent sampling requirements addressed in a January 2, 1980 NRC Confirmatory Order.
The inspector review of the licensee's actions with respect to the above letter indicated the following:
i the licensee established and implemented procedures for drywell i
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I atmosphere sampling, transport, and analyses under emergency conditions by September 1, 1981. The procedures were approved August 28, 1981.
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the licensee established and implemented procedures for sampling, transport and analyses of radioactive iodine and particulate samples of the Reactor Building, Main Stack and Turbine Build-ing effluents under accident conditions by September 1, 1981.
The procedures for Reactor Building and Main Stack procedures were approved August 28, 1981. The Turbine Building procedure-was approved September 1, 1981.
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the licensee revised and implemented orocedure 7.8.4 for collection, handling and transport of primary coolant samples under accident conditions by September 1,1981. The procedure was approved August 28, 1981.
the licensee established and implemented procedures for chemical
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and radiological analyses of primary coolant samples under accident conditions by September 1, 1981. The analysis procedures j
were included in the revision of procedure 7.8.4.
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the licensee reviewed the placement of the Turbine Building high range noble gas monitor by September 1, 1981.
Inspector review of the detector placement on August 31, 1981 indicated the licensee had repositioned the detector to ensure that the monitor provided a representative indication of noble gas effluent from th. Turbine Building.
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the licensee was training appropriate personnel in the imple-mentation of the newly established procedures and the revised procedure 7.8.4 within one week after September 1, 1981.
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Regarding the newly. established procedures and revised procedure 7.8.4, the inspector reviewed the following procedures on September 1 and 2, 1981:
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Pilgrim Nuclear Power Station Procedure No. 5.7.3.1, Revision 0, Primary Coolant Sampling, Transport, and Analyses under Emergency Conditions (Includes guidance formerly provided in procedure No. 7.8.4).
Pilgrim Nuclear Power Station Procedure No. 5.7.3.2, Revision
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0, Drywell Atmospheric Sampling, Transport, and Analyses Under Emergency Conditions.
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Pilgrim Nuclear Power Station Procedure No. 5.7.3.3, Revision 0, Sampling, Transport, and Analyses of Effluent Iodines and Particulates from the Main Stack under Accident Conditions.
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Pilgrim Nuclear Power Station Procedure No. 5.7.3.4, Revision 0, Sampling, Transport, and Analyses of Effluent Iodines and Particulates from the Reactor Building Vent under Emergency Conditions.
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Pilgrim Nuclear Power Station Procedure No. 5.7.3.5, Revision 0, Sampling, Transport, and Analyses of Effluent Iodines and
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Particulates from the Turbine Building under Accident
Conditions.
Based on this review, several procedural deficiencies were
identified which were primarily related to the adequacy of sample
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analysis capability and provisions for personnel to minimize their radiation exposure.
The inspector met with the licensee's Chief Technical Engineer subsequent to the procedure issuance and brought these matters to his attention. Among other items, the deficiencies identified included:
lack of sample analyses guidance if gamma spectroscopy system dead time exceeds 20 percent, lack of calculational methods
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relating to dilution of liquid samples to assure a specified dose rate (required by procedure), a procedural step refer'nced which did not exist, and lack of identif! cation of iodine sampling media to be used.
Based on discussions at the preliminary exit interview on September 4, 1981, the licensee was to review and revise the procedures to
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resolve the above matters by October 3, 1981.
(50-293/81-21-10)
4.
Training and Retraining The inspector reviewed selected areas of the training and retraining provided by the licensee to radiation workers and members of the Radiation Protection Organization. The review was made with respect to the following:
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Technic
.gecification 6.4, Training
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ANSI-N18.1, 1971, Selection and Training of Nuclear Power Plant
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Personnel
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10 CFR 19.12, Instructions to Workers
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Pilgrim Nuclear Power Station Training Manual, Revision 3
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Regulatory Guide 8.27, Radiation Protection Training for Personnel
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at Light-Water-Cooled Nuclear Power Plants a.
Radiation Worker Training The inspector audited the General Employee Training (GET) Course and discussed the course content and length with members of the licensee's training organization.
The inspector's attendance at tne course on September 28, 1981, revealed the course consisted of three separate video-tape modules.
Module 1 (duration approximately 40 minutes) provided general radiation protection information, Module 2 (duration approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
provided plant specific information, and Module 3 (duration approximately 30 minutes) provided respiratory protective equipment training.
The combination of information provided in Modules 1 and 2 was intended to provide the information required by 10 CFR 19.12.
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The audit of the General Employee Training course identified the following:
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The training course did not provide information relative to proper wearing of personnel monitoring devices for monitoring of beta radiation.
The training course did not elaborate on the workers'
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responsibility to read, understand and adhere to radiation work permits as evidenced by their signature _ thereon.
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The training course did provide guidance as to how to properly
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perform whole body frisking for contamination, i.e., time duration of frisk, probe distance, etc.
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Although the training course indicated hoods were to be worn and clothing openings were to be taped in an airborne radioactivity area, one individual was depicted on video-tape entering such an area without a hood and without taped openings.
The individual, with no head covering, was wearing a respirator.
The above matters were discussed with licensee training representatives.
The training representatives indicated the above matters would be reviewed and addressed in future training courses.
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In addition, the audit of the course indicated that the topics discussed in the combination video-tape and lecture modules appeared to lack sufficient training time, as compared to that recommended in Regulatory Guide 8.27.
The approximately 3-hour GET course was noted to provide 75% of the maximum training time (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) recommended for office workers not normally exposed to radiation, 21% of the maximum training time (14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />) recommended for closely supervised radiation workers, and 7% of the maximum training time (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />)
recommended for radiation workers not closely supervised.
The above matter was discussed with licensee training represen-tatives who indicated they had not as yet been directed ' a implement the recommendations of Regulatory Guide 8.27 (R.G. 8.27 2 to be implemented by March 1982).
The review of the radiation worker training also indicated that retraining was provided every two years in accordance with the Pilgrim Station Training Manual.
Regulatory Guide 8.27 recommends that this retraining occur annually.
Inspector discussions relative to Regulatory Guide 8.27 training with licensee representatives at the preliminary exit interview on September 4, 1981 indicated the licensee has yet to extensively review the training program described in the regulatory guide and compare the recommended training to what is now in place.
Licensee representatives indicated additional training personnel are to be ebtained who would assist in this area. As a result, licensee representatives could not comment on their actions regarding implementation of the training recommendations outlined in Regulatory Guide 8.27.
(50-293/91-21-01)
No violations were identified.
b.
Posting of Notices to Workers The inspector reviewed posting of notices to workers with respect to the requirements of 10 CFR 19.11.
10 CFR 19.11, Posting of Notices
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to Workers, requires in Section (a)(4) that, among other items, the licensee is to post copies of any notice of violation involving radiological working conditions and any response thereto. Section (e) of 10 CFR 19.11 requires that the licensee post the notice of violation within 2 working days of receipt and also post any response thereto within 2 days after dispatch of the response. Both the notice of violation and the licensee's response are to remain posted
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for a minimum of 5,vorking days.
On June 8, 1981, the licensee was issued a letter containing a notice of violation involving radiological working conditions. The licensee responded to the notice of violation in a letter dated July 29, 1981.
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During a review of controlled bulletin boards on August 31, 1981,
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the inspector noted that a notice of violation involving rz.diological conditions dated July 22, 1980, was posted as was the licensee's response dated August 29, 1980.
The review indicated however that neither the June 8, 1981, notice of violation nor the licensee;s response thereto were posted.
Inspector discussions with a management services representative indicated the two documents in question had not been posted.
The inspector indicated that failure to post notices to workers as required by 10 CFR 19.11 was a violation.
(50-293/81-21-02)
c.
Radiological Group Training.
The inspector reviewed the licensee's Radiological Group Training.
The results of this review are discussed in Section 5 of this report.
5.
ALARA Program
The inspector reviewed the licensee's program for maintaining occupational exposure as low as reasonably achievable (ALARA). The review was made with respect to the following:
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Technical Specification 6.11, Radiation Protection Program
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Regulatory Guide 8.8, Revision 3, Information Relevant to Ensuring that Occupational Radiation Exposure at Nuclear Power Stations Will Be as Low as Reasonably Achievable, dated June 1978.
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Regulatory Guide 8.10, Revision IR, Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as is Reasonably Achievable, dated September 1975.
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Pilgrim Nuclear Power Station Procedure No. 6.10-001, Revision 0, ALARA - Post Job Evaluation, dated June 4,1981.
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Pilgrim Nuclear Power Station Procedure No. 6.10-002, Revision 1, ALARA - Pre-Job Planning, dated January 14, 1981.
Pilgrim Nuclear Puwer Station Procedure No. 6.10-003, Revision 0,
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ALARA - Radiological Discrepancy Reports, dated June 4, 1980.
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Pilgrim Nuclear Poier Station Procedure No. 6.10-004, ALARA -
Decontamination, d-ted June 4, 1980.
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Pilgrim Nuclear Power Station Procedure No. 6.10-005, Revision 0, ALARA - Use of Portable and Permanent Shielding, dated June 4, 1980.
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Boston Edison Company, Nuclear Organization Policy Directive No.
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PDN3, Implementation of 10 CFR Part 20.l(c) Requirement to Maintain Radiation Exposure as Low as Reasonably Achievable, dated July 1978.
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Pilgrim Nuclear Power Station ALARA Guidelines (established approximately
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August 1979):
Radiation Work Permit ALARA Guideline
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Radiological Discrepancy Report ALARA Guideline Use of Portable and Permanent Shielding ALARA Guideline
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Regarding this review, the inspector noted that Regulatory Guide 8.8, Section C.1, Program for Maintaining Station Personnel Radiation Doses ALARA, states in part,
"To attain the integrated effort needed to keep exposures of station personnel ALARA, each applicant and licensee should develop an ALARA program that reflects the efforts to be taken by the utility, nuclear steam supply system vendor, and architect-engineer to maintain radiation exposure ALARA in all phases of a station's life. This program should be in written form and should enntain sections that cover the generally applicable guidance presented in this guide, as a minimum..." Section C.1,b(3) of the guide recommends developing plans, procedures and methods for keeping exposures ALARA.
In addition, Section C.l.b(1) of the guide recommends that, in view of the need for upper-level management support, responsibility and authority for implementing the ALARA program should be assigned to an individual or committee whose responsibilities and authorities include ensuring that an effective measurement system is established and used to determine the degree of success achieved by station operations with regard to ALARA program goals and specific objectives, a.
ALARA Policy Implementation During 1980 Outage The inspector reviewed the licensee's ALARA Program which was in place during the 1980 outage. The inspector review of the ALARA procedures in place during the outage and discussions with cognizant licensee personnel indicated that although a Nuclear Organization Policy Directive (PDN 3 referenced above) was issued July 1978 direct.ing the establishment of ALARA implementing procedures, no procedures were established until approximately August 1979.
These procedures consisted of three, one-to-two page non-station-approved ALARA guidelines (referenced above).
The guidelines were indicated as being approved by the Chief Radiological Engineer. However, documentation of approval of only one of the three guidelines utilized during the 1980 outage was evident.
Review of the guidelines indicated they provided limited guidance for an effective ALARA review in the areas of pre-job planning, on going job review and post job evaluation in that they lacked such items as:
estimation of total man-rem and recommendation of various means to reduce exposure (e.g., decontamination, mock-ups, etc.),
on going job review criteria and actions to be taken, post-job review and evaluation criteria / requirements (e.g., actual vs. sustained man rem), cost / benefit and recommendations for future work.
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Regarding monitoring and reporting of ALARA performance, the Nuclear Organization Policy Directive (referenced above) required the establishment and documentation of a process to monitor performance and periodically review the effectiveness of the ALARA program for operational nuclear power stations including a comparison of radiation exposure incurred at BECo's operational nuclear station to that incurred at other comparable operational units and the establishment of corrective measures as required.
The inspector review of the above indicated that reports which detailed ALARA efforts were prepared and forwarded to higher management for review. However, no indication of a means to monitor performance, periodically review the effectiveness of the ALARA program and implement corrective actions was evident.
The inspector did note the reports forwarded indicated significant efforts were directed to shielding and decontamination of high radiation and contaminated areas.
Based on discussions with cognizant licensee personnel the following were indicated as possible contributers to the substantial man-rem exposure during the 1980 outage:
A well defined outage plan, which provided information such as
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number of personnel, estimated job duration, etc., for each job to be performed, had not been developed. Con equently, adequate pre planning could not be performed.
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Jobs were rescheduled during the outage without sufficient time for pre-job ALARA planning.
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Man-rem tracking was performed only on the control rod drive work. Consequently, large numbers of personnel, working on relatively low exposure rate tasks, such as torus modification and nozzle work contributed significant exposure to the station's total man-rem without the licensee being immediately aware of the significance of this contribution.
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Use of inexperienced radiation workers in high exposure rate areas, such as those associated with the control rod drive work.
The inspector noted that a large number of new tasks were added during the outage which were not initially reviewed by the ALARA group prior to the outage.
b.
Current ALARA Program The inspector reviewed the current ALARA program with respect to the recommendations of Regulatory Guide 8.8 and 8.10 (referenced above).
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(1) Pre-Job ALARA The review of pre-job ALARA planning indicated no radiation work permits (RWP) are issued without an ALARA evaluation being performed. The evaluation requires input from the individual requesting the RWP for, among other items, estimated job man-hours, availability of tools and equipment, number of workers and need for mock-up training.
Subsequent to performance of radiation, contamination and airborne radioactivity surveys, an ALARA engineer evaluates the job with respect to estimated man-rem, shielding, decontamination, ventilation, high level waste disposal, and calculates a projected exposure savings.
The onsite ALARA group received an initial outage task plan and receives updates weekly. Man-hours are provided on the schedule for performance of ALARA reviews.
Using the detailed schedule, the ALARA group selects the RWP jobs for the following week and performs a pre-work ALARA evaluation. Using the recently initiated detailed " task sheets," the group can determine the
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estimated man-rem based on the number ci workers and man-hours presented on the " task sheet." The ALARA group maintains logs which provide information on the status of the jobs. The use of the logs and various ALARA forms are not formally incorporated into procedures.
The inspector review of the licensee's pre-job ALARA planning indicated, however, that no formal means was in place for consideration of alternate methods to perform a task with potentially lower man-rem expenditures prior to specifying the method to be used.
In addition, no formal method was in place to identify and implement major cost effective actions to substantially reduce man-rem over the life of the plant.
Further, no assignment of " dollar-value" per man rem was incorporated to facilitate cost-benefit determinations as part of the ALARA review.
Based on the above review, the inspector noted that additional steps can be taken to supplement the licensee's substantial effort to fully and formally implement all aspects of the pre-job ALARA program.
(2) Review of On-Going Work The inspector examination of the licensee's program for job tracking indicated that ALARA personnel were reviewing on going work.
However, no formal program or criteria were established for this review. An informal method was established for tracking jobs if the potential for one man-rem of exposure existed, otherwise, a decision was made on a per-job basis. The review of on going work would normally address matters, for example, such as changes in conditions, increase in numbers of personnel, changes in radiation levels, and approaching or exceeding man-rem estimate.
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The above was discussed with licensee ALARA group representa-tives who indicated these areas were being reviewed informally.
The licensee's representatives indicated that the tracking program was to be formalized and would include us; of a computer to track man-rem exposure and maintain a data bank for the jobs reviewed.
The licensee's Chief Radiological Engineer subsequently issued criteria to ALARA technicians for use in monitoring and evaluating new and ongoing tasks.
The criteria, issued October 1, 1981, requires ALARA tracking if a job is expected to exceed one man-rem, requires ALARA supervision review if the job results in greater than 50 millirem per man per hour but less than a total of one man-rem, and requires " frequent review" of ALARA tracking forms to ensure originally expected exposure rates remain the same.
If conditions change, the task is to be a
reevaluated.
The criteria did not include review and comparison of estimated man-rem with actual,
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Based on the above review the inspector noted that additional steps can be taken to fully and formally implement all aspects of an on going job ALARA review.
(3) Post-Job ALARA Evaluation The inspector review of the licensee's program for post-job evaluation indicated procedure No. 6.10-001 (referenced above)
was established for this purpose. The procedure requires calculation of man-rem savings realized from pre-job planning and preparation and completion of a Corrective Action Summary.
The summary requires determination of the effectiveness of the ALARA action and a determination of recommended changes to be made if the job were to be repeated. The Corrective Action Summary is to be reviewed by the Senior ALARA Engineer or his designated alternate.
The inspector review of the above indicated that no criteria were provided in the procedure to specify the point at which to bring to the attention of higher level management corrective actions which would result in substantial man-rem savings.
Nevertheless, these matters were apparently brought to the attention of higher level management at daily meetings.
(4) Management Overview of ALARA The inspector discussed the current ALARA program with the licensee's Nuclear Operations Manager (NOM).
The discussions indicated the NOM is cognizant of ALARA activities at the station. ALARA topics are discussed during daily planning
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meetings. The topics discussed include any changes in the number of personnel to be working on jobs having ALARA importance, and the radiation work permits which will be needed.
The licensee's Deputy NOM receives a weekly ALARA report which provides information on the number of radiological deficiency reports and the status of the report, the number of radiation work permit areas left in an unsatisfactory condition and who is responsible, the total number of radiation work permits issued and the number requiring ALARA action and the results o'f the ALARA action.
The NOM indicated that a monthly report is sent to corporate
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management which provides, among other items, the number of man-rem sustained during the month, the year to date man-rem.,
and a monthly man-rem goal. The NOM indicated that action is being taken to formalize the ALARA program on an organizational level, i.e., the establishment of organizational level procedures, to incorporate the ALARA review process into the maintenance request (MR) system, and to establish specific criteria for the need for an ALARA review.
In addition, the NOM provided a Nuclear Operations Policy Directive No. P-2, Personnel Conduct, dated August 27, 1981 which, among other matters, outlines radiation protection guidelines to which adherence is required.
Included in these guidelines are requirements to remain in as low a radiation
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area as practicable to accomplish the work and not to loiter in radiation or airborne radioactivity areas.
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The inspector discussions with the licensee's NOM indicated the NOM was encouraging support from all station personnel regarding ALARA, and was supporting the Chief Radiological Engineer in formulating and implementing a station-program in maintaining occupational radiation exposures ALARA.
The discussions with the NOM indicated a Policy Statement would
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be issued by November 1, 1981 which addressed, among other items, the need for personnel to maintain their occupational
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exposure ALARA.
Further, the NOM indicated the entire ALARA program would be completely formalized by April 1, 1982.
(50-293/81-21-03)
No violations were identified.
6.
Health Physics Appraisal The inspector reviewed the licensee's implementation of corrective action
for the findings identified during the NRC Health Pnysics Appraisal (HPA)
of Pilgrim Nuclear Power Station.
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References:
NRC Health Physics Appraisal Report No. 50-293/80-05, dated July 22, 1980 Licensee response to Health Physics Appraisal, BECo Letter No.80-202, dated August 29, 1980 Licensee Supplemental Response to Health Physics Appraisal, BECo
Letter No. 81-93, dated May 11, 1981 The following paragraphs provide the HPA findings, licensee responses, and the inspector's findings regarding the implementation of the licensee's corrective actions described in the response.
a.
Internal Exposure Control Program General Finding (50-293/80-05-07)
The overall program for internal exposure control was found to be inadequate and not effective.
Significant Finding (A.1)
1.
Lack of confidence in the direct measurement activities (whole body counter). There was a lack of technical oversight for this operation and weakness in personnel training and qualifications of those individuals assigned to operate and calibrate the whole body counter.
Licensee Response
"The whole body counting procedure for calibration, Procedure
- 6.5-130, " Calibration of Canberra Model 2230 Body Burden Analysis System," was revised, and it was approved by the ORC on February 1, 1980. The factory representative for the whole body counter reviewed and agreed with the above calibration procedure.
Formal training on the whole body counter will be given to all PNPS H.P. personnel and to selective contractor H.P. personnel as part of the formal training / retraining program to be implemented by January 1, 1981, as stated in Section B, Item 1 below."
Inspection Finding
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Procedure #6.5-130 was revised and approved by the Operations Review Committee on February 1, 1980. The procedure required use of known amounts of radioactivity in phantoms during the calibration.
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Procedure #6.4-130, Revision 1, Operation of the Whole Body Counter, dated December 27, 1978 requires a daily background check and a calibration when the system is being used.
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Regarding training, review of training records indicated 14 of 16 radiation protection technicians were trained in the operation
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of the whole body counter during the period June - August 1981.
The training duration was approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The licensee also provided extensive training in Health Physics, including internal dosimetry, to the major portion of the radiation protection staff during January and March of 1981. This training duration was approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
During the review of this item, the inspector noted the licensee obtained a contractor whole-body counter on September 7, 1981, to augment the whole-body counter capabilities of the Pilgrim Station..The counter was placed in operation on September 14, 1981. The inspector discussions with licensee representatives regarding this counter indicated the Itcensee had evaluated the capabilities of the counter prior to placing it in service.
The inspector noted, however, that as of October 2, 1981, the licensee had not established approved procedures for the counter.
This was brought to the attention of licensee radiation protection representatives who initiated a temporary procedure for the unit. The temporary procedure was approved October 2, 1981.
The licensee's revised whole body counter calibration procedure will be reviewed during a subsequent inspection (50-293/81-21-11)
Significant Finding (A.2)
2.
Lack of procedures to provide for proper collection, handling and analysis of indirect bioassay samples; together with a lack of procedures establishing biological models and calculational techniques necessary to evaluate monitoring data in terms of dose assessment and compliance with intake limitations set forth in 10 CFR 20.103.
Licensee Response
"In the month of June, four Health Physics Engineers attended a one week course on Internal Dosimetry at the University of Lowell, Lowell, MA.
At the present time, indirect bioassay sampling and inter-pretation of results is under evaluation.
Consultants are in the process of being contacted to assist Boston Edison in developing a comprehensive Internal Dosimetry Program.
The initial program is targeted for implementation by January 1, 1981."
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Licensee's Supplemental Response
"Following the completion of the Internal Dosimetry training course by members of its staff, Pilgrim Station's Health Physics
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Group initiated a reevaluation of its planned internal dosimetry program developments.
Considered during this reevaluation were the pending International Committee on Radiation Protection (ICRP) recommendations which had been issued in draft format during 1980.
It became obvious during the reevaluation process that the scope of the internal dosimetry program development was much larger than originally conceived and its nature more complex than initially anticipated.
In the late fall of 1980, a decision was made to expand the scope of the program to encompass the proposed ICRP recommendations as much as possible.
To achieve this goal, an eminent figure in the Health Physics Consulting area was contracted to develop the internal dosimetry program for Pilgrim Station which would incorporate those aspects of the ICRP draft recommendations deemed applicable.
Due to the complexities involved in this newly scoped program development the projected completion date for this effort is January 1,1982."
Inspection Finding The licensee did send several Health Physics Engineers to the Internal Dosimetry Course during June 1981.
The licensee utilizes the whole body counter (direct bio-assay)
for identifying and quantifying any intakes of radioactive materials. Based on the whole body coun+.er results, a contractor is contacted to provide guidance and analyses of urine and fecal samples (indirect bioassay). However, no guidance is included in the procedure as to when this contact should be made.
Inspector discussions with cognizant licensee representatives-indicated the contractor's development of the program is anticipated to be completed by January 1,1982.
The licensee's Internal Dosimetry Program will be reviewed
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during a subsequent inspection.
(50-293/81-21-04)
Significant Finding (A.3)
3.
Failure to ensure consideration of engineering controls for airbe ne radioactivity areas or to evaluate and document the
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practicability of applying process or engineering controls in airborne radioactivity areas.
Excessively high loose radioactive contamination levels existed in many areas of the plant and a program to reduce and maintain significantly lower levels was not implemented.
Licensee Response
"The consideration of engineering or process controls and the practicality of their use in airborne radioactivity areas is currently being implemented through the ALARA Program Procedures.
In addition, these procedures address the initial decontamination of areas containing high levels of radioactive contamination as well as making reasonable efforts to maintain low contamination levels.
The Respiratory Protection Program and accompanying procedures, which are in the process of being written, will-incorporate the consideration and use of engineering and process controls in airborne radioactivity areas on an expanded scale to meet the guidelines of NUREG-0041.
Full implementation of the ALARA procedures was achieved on August 4, 1980, and full implementation of the Respiratory Protection Program and accompanying procedures will not be achieved until March 1,1981, depending upon the procurement of necessary equipment and facilities.
Several facilities are being evaluated at this time, which could result in implementation of the Respiratory Protection Program and accompanying procedures as early as January 1,1981."
Licensee Supplemental Response
" Delays have been experienced in equipment procurement and delivery, facility modifications and recruitment of experienced personnel each of which is an essential component utilized in development of the Respiratory Protection Program.
Modifications to the station air system are necessary to facilitate a fresh air supply network. The nature and scope of these design changes have been determined and equipment ordered.
Installation is dependent upon receipt of the equipment.
A drycleaning system for cleaning respirators, health physics test equipment and new respirators / face masks have been purchased and are onsite awaiting completion of the modifications to the dedicated facility.
At this junctere the Respiratory Program is approximately 80%
complete.
The procedures, which are an adjunct of the program, are approximately 50% complete.
Completion of the total program s
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is contingent upon completion of the essential components and the major contributor to the delay, the facility modifications, is now scheduled for completion during August,1981. A draft of the program and its implementing procedures will be available at that time.
Implementation of the final Respiratory Program will be achieved by September 1,1981."
Inspection Findings The licensee is implementing process and engineering controls to minimize airborne radioactivity.
Examples identified by the inspector during tours of the controlled area included the licensee's utilization of four portable ventilation units to minimize airborne radioactivity in the reactor cavity and the decontamination of the torus prior to work therein.
Inspector review of the licensee's ALARA procedure indicated they do provide consideration of decontamination, ventilation and containment to eliminate or minimize airborne radioactivity prior to the start of work. However, the only criteria provided in the ALARA procedures relative to the above was that areas or
rooms contaminated in excess of 200,000 dpm/100 cm and which are frequently entered for surveillance should be considered for decontamination.
Regarding implementation of the Respiratory Protection Program,.
this is discussed in the review of finding A.4 below.
Regarding the licensee's ALARA program, this is discussed in Section 4 of this report.
The licensee's program for maintaining contamination levels low will be reviewed during a subsequent inspection.
(50-293/81-21-12)
Significant Finding (A.4)
4.
Lack of adequate facilities for cleaning, inspecting and maintaining respiratory protection equipment.
Licensee Response
"A purchase order has been issued for the purchase of the Health Physics Systems drycleaning system, which will be used for cleaning respirators.
Expected delivery is prior to January 1, 1981.
The drycleaning system will be located in a dedicated area, as yet to be determined, for cleaning, maintaining and inspecting respiratory equipment.
As stated above, several facilities are in the process of being evaluated, which could results in full implementation of the
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Inspection Findings The licensee has established a dedicated respiratory protective equipment facility for maintaining respiratory protection equipment.
The inspector toured the facility on September 28, 1981 accompanied by the licensee's contractor responsible for establishment of the respiratory protection program. The inspection tour and discussions with the contractor indicated a water washing system, with backup drycleaning, (both in the process of being installed) would be used for cleaning of equipment; hot air dryers were in place for drying respirators.
Equipment for testing of inspected and repaired equipment was on hand. The respiratory protective equipment facility had a dedicated area for inspection and repair of equipment.
Because the licensee is not yet making allowance for the use of respiratory protective equipment, as permitted by 10 CFR 20.103(b)(2), an extensive review of the licensee's respiratory protection program was not performed.
However, based on the inspector tours of the licensee's facility, the facility appeared adequate for cleaning, and decontamination, inspection, repair and storage of respiratory protective equipment.
Licensee representatives indicated the formal respiratory protection program is to be in place by January 1, 1982.
The licensee's implementation of the respiratory protection program will be reviewed during a subsequent inspection (50-293/81-21-05)
Significant Finding (A.5)
5.
Lack of adequate training for contractor health physics technicians in the operation of the respirator fitting booth.
Licensee Response
"The factory representative for the respirator fitting booth provided onsite instruction at PNPS.
In addition, the factory representative recommended modifications to the existing procedure for the respiratory fitting booth, Procedure #6.7-105, ' Operation of Mask Fitting Booth'.
This procedure is being revised and should be completed by September 15, 1980.
Formal training on the respirator fitting booth will be given to all PNPS H.P. personnel and to selective contractor H.P.
personnel as part of the formal training / retraining program to be implemented by January 1, 1981, as stated in Section B, Item 1 below."
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21 Inspection Findings The inspector review of training records on September 28, 1981 indicated 6 of 16 licensee radiation protection technicians had
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received training in Procedure 6.7-105 during June of 1981.
The inspector discussed the training given to several contractor technicians _ operating a contractor supplied fitting booth near the training center.
The inspector determined that the licensee had reviewed the training given the contractor technicians and determined that the technicians had received adequate training by the vendor of the fitting booth.
I As a minimum, the licensee indicated contractor radiation protection technicians will be required to read, understand, and acknowledge all radiation protection procedures they will be implementing.
Regarding procedures for the respirator fitting booth, the inspector noted procedure No. 6.7-105, as revised, was approved on January 14, 1981. The procedure contains, among others, instructions for pre-fit test evaluation, qualitative and quantitative fit testing, and post-fit evaluation.
On September 28, 1981, the inspector was fitted for a respirator-by contractor technicians operating the contractor supplied Dynatech Model 260B Corn Oil Fitting Booth. The inspector review of the fitting indicated the technicians were knowledge-able in operation of the booth. The inspector noted the technician
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to be fitting personnel without approved procedures. This was brought to the licensee's attention who initiated a procedure change notice to approve procedures for operation of the booth.
Significant Finding (A.6)
6.
Lack of in plant surveillance to insure proper usage of respiratory equipment.
Licensee Response
"The Health Physics Staff has developed a one day training course on radiation protection. This training course, as with the General Employee Training (GET) class, includes the proper use of respiratory equipment.
Implementation of this course began in June and, to date, approximately one-half of the PNPS personnel have received this training.
Full implementation should be achieved by October 1, 19E0.
In addition, when the new procedures for the Respiratory Pr.tection Program are implemented, H.P. control point individuals will be held responsible for ensuring the proper use of respiratory equipment by personnel."
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Inspection Findings The licensee had developed a supplemental training course on radiation protection which included discussion of respiratory-protection equipment. Also included were such items as ALARA and general housekeeping.
The training course duration was approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and was given during the period _ June 1980 through May 1981.
The course was intended as a sincie training
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session with retraining to be performed through General Employee Training.
The licensee established Procedure No. 6.7-102, Respirator Equipment Issuance, Field Testing and Wearing, to provide guidance for proper wearing of respiratory protection equipment.
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The licensee will require Health Physics control point personnel t
to ensure proper use of equipment. Additionally, based on discussions with the licensee's Chief Radiological Engineer, a
surveillance sheet is to be established and used by the control
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point personnel for this purpose.
Significant Finding (A.7)
7.
Failure to have a technically knowledgeable individual assigned responsibility for maintaining cognizance of developments in respiratory protection use and equipment and evaluation of'the effectiveness of the respiratory protection program.
Licensee Response
" Manpower requirements are being evaluated at the present time.
Requests-for additional personnel will be made as a result of the evaluation.
The schedule for full implementation is not available at this time."
Inspection Findings The inspector review of this item indicated the licensee has assigned an individual the respon'.ibility for maintaining cognizance of developments and use of respiratory protective equipment and the evaluation of the respiratory protection program effectiveness. The' individual was assigned this_ function in May 1981.
b.
Personnel Selection and Training Program
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General Finding The level of formal training and education of most of the staff and the contractor personnel is limited.
Significant Finding (B.1)
1.
There is no formal training / retraining program that exists for members of the plant health physics staff. A check sheet (qualification) is used to documeqt that personnel have received new hire orientation and some on-the-job orientation.
Licensee Response
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"The Health Physics Staff is developing a formal training /
retraining program for all Health Physics personnel, including
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contractor H.P. personnel. The program is in the final draft
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stage at the present time.
The training / retraining program will be implemented by January 1,1981."
Inspection Findings The inspector review of this item and discussion with cognizant licensee representatives indicated the licensee has provided'
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training to the radiation protection group in order to increase the group's overall level of knowledge.
This training included
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two separate 100-hour training courses (given January and March of 1981). The course addressed approximately 40 radiation protection related topics and was given to all licensee health physics technicians and supervisors. The licensee's Waste Management Supervisor was unable to attend these courses.
In addition, the licensee provided approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of plant
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systems training. The training was provided at the end of 1980 and during April and August of 1981. The systems training included an oral examination during a facility tour and a written exam.
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Based on inspector review of training records on September 28, 1981, 5 of 16 licensee health physics technicians received the systems training, 7 of 16 technicians had received a systems exam and 3 of 16 had received an oral system exam during a plant tour.
Regarding contractor technician training, the licensee gave approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of health physics procedure and system training to approximately 5 of 9 contractor technicians onsite during August 1981.
In addition, the licensee obtained approximately 75 additional contractor technicians for use during the outage. Of the 75, approximately 40 technicians received the 30-hour training course during September 1981.
The remainder were hired later and were given oral and written
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exams in lieu of the 30-hour training course.
For those individuals responsible for operation of the respirator fitting booth, this training included training in appropriate procedures.
The licensee's training, qualification and retraining program for members of the health physics staff is described in Section 3.5 of the Pilgrim Nuclear Power Station Training Manual.
The licensee technician qualification is comprised of f_ive major sections.
These sections are:
plant systems, basic radiation protection, Pilgrim Station procedures, and health physics /ALARA technician practices, and training evaluation.
Upon completion of the five sections, the newly assigned licensee technician completes a written and/or oral examination.
Regarding formal contractor health physics technicians training requirements, the technicians are instructed in appropria.te health physics /ALARA procedures and a test is given to the technican.
In the event contractor technicians are urgently needed, an oral exam is to be given with a written' exam to follow.
These contractor exams are recommended by the procedure i
(i.e., they are not required).
The licensee's training manual also recommends that a previous employer of the technician be contacted. This is to determine-acceptability of the technician. The training manual further recommends that the Chief Radiological Engineer or his designee
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indicate, by signature, the approval of the technician to work.
Regarding radiological group retraining, Section 3.5.2.1 of the Pilgrim Nuclear Power Plant Training Manual indicates that retraining is conducted as considered necessary by the Chief Radiological Engineer and also indicates that such items as a
noted trend in improper contamination surveys might require a training period on the subject.
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Based on the inspector's review of the licensee's health physics technician training / retraining, the following was noted:
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No specified retraining program for health physics technicians was in place.
Such a program might include such items as frequency, scope and content, objectives and minimum performance standards.
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No specified training / qualification / retraining require-ments we: e in place for each position in the radiation protection program.
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No formal means were in place for assuring appropriate personnel were made knowledgeable in procedure changes and documenting this assurance.
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No training / retraining program for Health Physics Supervisory
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personnel was in place.
The inspector noted the above items were not consistent with the recommendations of ANSI-N18.1 (referenced above) which reccmmends in Section 5.5 that a training program be established which maintains the proficiency of the operating organization.
Changes in procedures are recommended as an item to be included in the retraining program.
The inspector noted the licensee's Technical Specification 6.4.a places the retraining program for the facility staff under the direction of the Pilgrim Station Manager.
The above items were brought to the licensee's attention. The inspector indicated the above matter would be reviewed during a subsequent inspection.
(50-293/81-21-06)
Significant Finding (B.2)
2.
There is no established retraining program in radiation safety for general employees.
Licensee Response
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" Contrary to the appraisal finding, radiation safety and. protection retraining has always been given to all general employees through regular reindoctrination of General Employee Training (SET) at PNPS."
Inspection Findings The inspector review of the licensee's retraining program, as described in the licensee's Pilgrim Station Training manual
indicated retraining is addressed.
(See Section 3.a of this report).
Significant Finding (B.3)
3.
There was minimal effort to determine the qualifications of the contractor supplied health physics personnel.
The program in place consisted of only a screening process. The training provided these individuals is lacking and the qualifications of many of the contractor health physics technicians used during the refueling outage were questionable.
Licensee Response
"As stated above, a formal training / retraining program is in the process of being developed. This training / retraining program will be used to determine the qualifications of the contractor H.P. personnel.
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The training / retraining program will be implemented by January
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1, 1981."
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Inspection Findings The inspector review of this item indicated the licensee was determining the qualifications of contractor supplied health physics personnel for the upcoming outage.
In accordance with the recommendation of the Pilgrim Nuclear Station Training Manual (Section 3.5.1.2), at least one previous employer of a contractor technician was contacted to determine acceptability of the technician, the technician was interviewed to assure
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-accuracy of resume, appropriate procedures ere being read by the technician, and a supervisor was signing off the qualification sheet to indicate approval of the technician to work.
7.
Radioactive and Contaminated Material Control The inspector toured the controlled areas and reviewed radioactive and contaminated material control with respect to the following:
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10 CFR 20.203, Caution Signs, Labels, Signals and Controls Pilgrim Nuclear Power Station Procedure No. 6.1-024, Revision 1,
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Radiological Posting of Areas of the Station, dated January 12, 1978.
The tours indicated radioactive and contaminated materials were controlled in accordance with the above.
No violations were identified.
8.
High Radiation Area Posting and Control The inspector toured the controlled area at various times during the inspection and made radiation intensity measurements to verify licensee compliance with the following:
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10 CFR 20.203, Caution Signs, Labels, Signals and Controls
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Technical Specification 6.13, High Radiation Area
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Pilgrim Nuclear Power Station Procedure No. 6.1-024, Revision 1, Radiological Posting of Areas of the Station, dated January 12, 1978
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Pilgrim Nuclear Power Station Procedure No. 6.1-012, Revision 6, Access to High Radiation Areas
Technical Specification 6.13 requires that each high radiation area in which the intensity of radiation is greater than 100 millirem /hr be barricaded and conspicuously posted as a High Radiation Area.
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During a tour of the controlled area on September 28, 1981, at approxi-
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mately 11:30 p.m., the inspector noted that neither the main access to the drywell was posted as a High Radiation Area nor was there any conspicuous posting in the area to indicate a High Radiation Area would be entered upon entry into the drywell. The area was barricaded.
A subsequent tour of the drywell and performance of radiation intensity measurements indicated the drywell general area radiation dose rates ranged from 100-300 millirem /hr.
Further, the inspector noted no posting in the drywell to indicate the general area dose rates. However, the inspector noted that radiation surveys of the drywell were posted at the access control point and the licensee was apparently briefing personnel on the radiation fields in the drywell prior to their entry.
The above was brought to the attention of a licensee radiation protection representative who subsequently posted the access point.
The inspector noted that failure to post a High Radiation Area in accordance with Technical Specification 6.13 was a violation.
(50-293/81-21-07)
9.
Review of High TLD Badge Reading The inspector reviewed the circumstances, licensee actions and evaluations relating to an apparent 9.67 rem exposure to the skin of the whole body of one individual during the second quarter of 1981.
The regulatory limit for exposure to the skin, as provided in 10 CFR 20.101, is 7.5 rem.
The inspector reviewed the event with respect to the following:
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10 CFR 20.403, Notification of Incidents
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10 CFR 20.405, Reports of Overexposures and Excessive Levels and Concentration
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10 CFR 20.409, Notification and Reports to Individuals
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10 CFR 50.72, Notification of Significant Events
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BEco Letter No.81-189, dated August 13, 1981, Transmittal with 10 CFR 20.405 Report
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Technical Specification 6.11, Radiation Protection Program i
a.
Circumstances and General Description The personnel monitoring system at Pilgrim Nuclear Power Station utilizes thermoluminescent dosimeters (TLD's) and pocket chambers for routine monitoring of personnel. The pocket chambers are read
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daily while the TLD's, unless unusual circumstances develop, are worn by the individuals for an entire calendar quartar and are read at the end of that quarter. The TLD provides beta and gamma personnel monitoring data while the pocket chamber provides only gamma monito.ing
. data.
On July 2,1981, during routine processing of TLD's worn by ' individuals-at the Pilgrim Nuclear Power Station for the second quarter of 1981,
'the licensee determined that an individual may have received an exposure to the skin of the whole body of 550 mR gamma.(penetrating)
and 9120 mrad (nonpenetrating) for a total skin exposure of-9670 mrem. As a result of the potential excessive exposure to.the skin, the licensee initiated an investigation.
b.
Reporting-The inspector reviewed the licensee's reporting of the event with respect to the reporting requirements referenced above.
No violations were identified, c.
Licensee' Evaluation The licensee's 10 CFR 20.405 report (referenced above)'provided details of the evaluation.
The inspector review of the evaluation indicated an attempt was'made to reconstruct the activities of the individual (Watch Engineer) who wore the badge during the quarter to determine if the-individual had entered any areas with a significant potential to cause a high sn:n dose, i.e., high beta radiation areas.
The licensee was unable to adequately reconstruct the individual's_ activities due to the individual's failure to sign-in on radiation work permits.
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Licensee representatives interviewed the individual during their investigation and determined that the individual'had not entered any areas that required a Radiation Work Permit (RWP) unless an RWP existed. The licensee conducted a review of all outstanding RWP's and, based on the review, determined that no apparent area existed, other than direct contact with equipment, where the' individual could have sustained an exposure with the indicated beta / gamma ratio (9120 mrad /550 mrem = 16/1).
In order to determine if possible anomalies existed with the TLD badge, the licensee checked the badge for_ contamination, recalibrated the badge, and processed a quality assurance TLD chip in the badge.
No unusual conditions were identified.
The licensee subsequently performed calculations to determine the
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amount of radioactive contamination which would cause the high skin
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dose reading. As a result of the licensee's review and calculations,
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the licensee believes the high skin dose reading to be the result of contamination of the badge.
However, the licensee will credit the individual with the apparent skin exposure.
d.
Inspector Evaluation The inspector interviewed the individual, performed reviews of selected radiation survey data to determine the extent of beta radiation fields at the facility, and reviewed the characteristics of the TLD badge to determine if radioactive contamination of the badge could cause such an exposure.
On September 2, 1981, the inspector interviewed the individual (Watch Engineer) that wore the TLD which read high. At the interview the individual indicated that he always wore his TLD and pocket chamber together. He stated that he had not lost his TLD at any time.
The individual acknowledged entering RWP controlled areas at times without signing in and out on the RWP, but the individual said he always wore the appropriate protective clothing and/or respirator.
as required. The individual told the inspector of several areas which would have high beta radiation readings.
Further, the individual said that he had, at times, entered areas containing primary steam leaks.
Based on the above the inspector selected and reviewed previous surveys of high beta radiation areas made by the licensee.
The areas reviewed included:
residual heat removal cubicles, the spent resin storage tank area, sample sinks, the sludge tank area, the clean waste tank area, the condensation make up rooms, the charcoal vaults and other areas. The inspector review of the previous surveys could not identify any areas where the individual, could have sustained the indicated exposure based on the location of high dose rate areas, stay time and beta / gamma ratios.
Further, because the individual indicated he had entered areas with primary steam leaks, the inspector estimated the possible whole body and skin dose rates assuming submersion in an air activity equivalent to the maximum gaseous activity anticipated, i.e.,
that at the outlet of the recombiner.
Utilizing a May 13, 1981, recombiner outlet sample, th?
inspector's calculations indicated an estimated beta / gamma ratio of 1/1 to 2/1 versus the 16/1 ratio indicated by the TLD badge readout.
Regarding the licensee's capability to monitor low energy beta doses, licensee representatives indicated that an evaluation had been performed of the badge's beta monitoring capability. The evaluation, using low energy beta sources, indicated the badges were capable of measuring beta radiation with acceptable accuracy.
The inspector review of the TLD badge characteristics indicated that the badge had small crack-like openings where radioactive contamination may collect. Consequently, the inspector noted that, based on the position of the TLD chips in the badge and the position of these
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openings, it appeared possible for contamination to cause a high reading of the skin dose monitoring TLD chip and a relatively low dose to the whole body penetrating dose monitoring TLD chip..The inspector also noted that since the licensee uses the TLD badge to monitor personnel for an entire quarter, rather than monthly, the probability that low level contamination could cause such a reading would be increased due to the contamination remaining close to the TLD chips for a longer period of time.
In an attempt to reconstruct movements of the individual during the second quarter, the inspector reviewed and compared radiation work-permit sign-in data with licensee key-card access data.
This review had also been performed by the licensee.
Technical Specification 6.11, Radiation Protection Program, required that procedures for personnel radiation protection be prepared, approved, maintained and adhered to for all operations involving personnel radiation exposure.
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Pilgrim Nuclear Power Station Procedure No. 6.1-022, Radiation Work Permit, Revision 7, requires in Section G, that,among other items, all personnel entering or leaving a work area under the control of an RWP must sign in on the access control sheet indicating name, social security number, time in and out, and pocket dosimeter readings upon entry and exit.
During the review and comparison of the various access control data,
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the inspector noted that the individual had, on June 1 and 2, 1981, entered the access to the drywell personnel airlock. These entires were made for inspection purposes. The inspector review of the Radiation Work Permit (No. 81E-1) providing radiological control of the area and the associated sign-in sheets for the airlock indicated the individual had not signed in or out on the sign-in sheet, or provided other required information, i.e., social security number, time in and out, and pocket dosimeter readings upon entry and exit.
Maximum radiation dose rates in the area were noted to be 100 millirem /hr with contamination levels ranging up to 23,000 dpm/100 cm (beta / gamma), according to survey map 131 dated June 4, 1981.
Failure to adhere to procedure 6.1-022 was a violation of Technical Specification 6.11.
(50-293/81-21-08)
10.
Procedures The inspector toured the controlled areas at various times during the inspection to review radiation worker and radiation protection personnel adherence to the requirements of selected licensee radiation protection procedures.
Technical Specification 6.11, Radiation Protection Program, requires that procedures for radiation protection be prepared consistent with the requirements of 10 CFR Part 20 and be approved, maintained and adhered to for all operations involving personnel radiation exposure.
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31 Pilgrim Nuclear Power Station Procedure 6.1-022, Revision 7, Radiation
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Work Permit (RWP), states in part in Section V.A. that, "it is the responsibility of the first line supervisor and the individuals working for the supervisor under the control of an RWP to follow all instructions on the RWP..."
On September 2, 1981 at 3:00 p.m., the inspector toured the 117' elevation of the Reactor Building to review work in that area. The inspector noted four individuals to be working under RWP No.81-963, Overhaul tools for Outage, dated July 17, 1981.
The review of RWP No.81-963 requirements indicated that cloth hoods were to be worn by the personnel working under the RWP. During the RWP work review the inspector noted that none of the individuals were wearing cloth hoods, while only one individual was wearing a small cloth " skull cap."
The individuals were also noted to be wearing faceshields which they moved up and down by hand. The material the individuals were working with was indicated by the health physics technician in the area to exhibit maximum removable contamination levels
of 10,000 dpm/100 cm (beta / gamma).
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The inspector discussed the above with licensee representatives and
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indicated that failure of the workers to adhere to the requirements of the RWP was a violation of Technical Specification 6.11 (50-293/81-21-13).
Licensee radiation protection representatives immediately required the individuals to don hoods.
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11.
Exit Interview The inspector met with licensee representatives (denoted in paragraph 1)
at the conclusion of the inspection on October 2, 1981. The inspector summarized the purpose, scope and findings of the inspection.
The licensee representatives indicated the following:
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A change to the condensate demineralizer system drawing to reflect the differences identified during a September 10, 1981 walk-down of the system would be initiated by the end of the day (October 2, 1981).
(Details Paragraph 3.a)
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A notice of violation dated June 8, 1981 and the licensee's response thereto, dated July 29, 1981 had not been posted in accordance with 10 CFR 19.11.
(Details Paragraph 4.b)
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A policy statement addressing ALARA would be issued by November 1, 1981 (Details Paragraph 5.b(4))
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The Pilgrim Nuclear Power Station ALARA program would be fully formalized by April 1,1982.
(Details Paragr oh 5.b(4))
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