IR 05000245/1987031
| ML20149D444 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/28/1987 |
| From: | Shanbaky M, Thomas W, Weadock A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20149D436 | List: |
| References | |
| 50-336-87-26, 50-423-87-27, 52-245-87-31, NUDOCS 8801120367 | |
| Download: ML20149D444 (9) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION 1 Report Nos. 87-31 87-26 87-27 Docket Nos. 50-245 50-336 50-423 License Nos.
DPR-21 Category C
DPR-65 NPF-49
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Licensee: Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06101 Facility Name: Millstone Nuclear Generating Station Inspection At: Waterford, Connecticut Inspection Conducted:
November 16-20, 1987 Inspectors:
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( ? \\ 7 4lC A. Weadock, Radiat%n Specialist, Date-FRPS, EP&RP Branch PLTd M [n lL{ t ilM W. Thomas, Radiati%n Specialist, Date
FRPS, EP&RP Branch Approved by:
.441 5 4 1.J#
nh887 M. Shanbaky, 'CMef,
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Date Facilities Radiation Protection Section, EPRPB Inspection Summary: Inspection on November 16-20, 1987 (Report Nos.
50-245/87-31, 50-336/87-26, 50 423/87-27)
Areas Inspected: Routine, unannounced inspection to review radiation protection activities associated with the Unit 3 outage. Areas reviewed included internal and external exposure control, personnel selection and training, and ALARA.
The inspection also included a review of Unit I drywell personnel access controls during refueling.
Results: No violations were identified.
The radiation protection program was being effectively implemented during the outage.
Several specific areas of program strength were noted. A weakness was icentified with the qualification of respiratory protection support personnel and is described in section 4.1.
8801120367 880104 ADOCK0500g5 DR
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DETAILS 1.0 Persons Contacted
- M. Brennan, Unit 1 Radiation Protection Supervisor (RPS)
- C. Clement, Unit 3 Superintendent R. Crandall, Supervisor, RES - NUSCO
- T. Cummins, Unit 3 Radiation Protection Technician
- B. Granados, Station Health Physicist
- H. Haynes, Station Services Superintendent
- J. Laine, Unit 2 RPS
- R. Sachatello, Unit 3 RPS
- P. Simmons, RPS, Services
- Attended the exit meeting on November 20, 1987.
Other licensee personnel were also contacted during the course of this inspection.
2.0 P_urpose The purpose of this routine, unannounced inspection was to review the implementation of the Radiation Protection Program during the Unit 3 outage.
The following areas were included in this review:
- Posting and Labeling,
- Personnel Selection and Training,
- Internal Exposure Controls,
- External Exposure Controls,
- ALARA.
The following areas were also reviewed during this insp<
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- Status of Previously Identified Items, and,
- Unit 1 Drywell Personnel Access Controls during Refueling.
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3.0 Status of Previously Identified Item; 3.1 (Closed) 78-BU-08 (Bulletin): Radiation levels from fuel element transfer tubes.
The licensee performed an extensive evaluation of the adequacy of the Unit 3 fuel transfer tube shielding during the current outage.
Real time measurements, using portable survey instruments and radiation monitors, were made to evaluate dose rates in areas adjacent to the transfer tube during element transfer.
ine walkway area adjacent to the Fuel Building / Containment Building interface (see NRC Report 423/85-56) was appropriately locked, posted and controlled by the licensee.
The licensee also placed numerous TLDs in areas adjacent to the fuel transfer tube (inside containment, the enclosure and auxiliary buildings, etc.) to evaluate integrated dose due to multiple fuel movements.
Results of the surveys indicated unexpected dosa rates in only one area, the Boron Injection Tank (BIT) cubicle in the auxiliary building.
Transient dose rates of approximately 30 R/hr were measured 20 feet off the floor of the cubicle; dose rates of approximately 2 R/hr were measured at head level.
The b!T cubicle was already located in a posted, locked High Radiation Area previous to any fuel movements. The inspector verified that the licensee was taking additional actions to control access to the BIT cubicle; these included erecting a plywood barrier at the cubicle entrance to prevent access.
The licensee indicated additional long-term control measures would be evaluated.
The inspector concluded the licensee carried out a comprehensive survey program to identify potential shielding deficiencies and implemented appropriate control measures as necessary. This item is closed.
3.2 (Closed) 423/86-17-02 (Follow-Up Item): Shield CHSRE69 letdown line and add delay coils to primary sampling lines.
The licensee has installed temporary shielding around CHSRE69 to reduce area dose rates. The inspector observed that the letdown line is appropriately posted as a High Radiation Area (HRA).
An engineering request has been submitted to add delay coils to the primary sampling lines to reduce dose rates due to N-16 during primary sampling.
The inspector noted that in the interim, the sampling cubicle is kept locked and posted as a HRA. As the licensee is employing adequate control measures, this item is closed.
3.3 (Closed) 423/86-17-02 (Follow-Up Item): Licensee to complete procedures and fit-test methodology for Biopak 60 prior to claiming protection factors. The inspector reviewed the status of the Biopak during the current inspection.
The licensee has incorporated the Biopak into their respiratory equipment maintenance and QA L
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procedure, however a fit-test ethodology has not been developed and the licensee indicated no immediate plans to take protection factors for the Biopak.
Consequently this item is closed. Any future incorporation of the Biopak into the licensee's radiological respiratory protection program will be reviewed as part of ongoing NRC inspections in this area.
4.0 Personnel Selection, Qualification and Training 4.1 Radiation Protection Personnel The selection, qualification and training of contractor radiation protection technicians and support personnel was reviewed against the following criteria:
ANSI N18.1,1971, "Selection and Training of Nuclear Power Plant
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Personnel,"
Procedure SHP 4920, "Contracted Health Physics Personnel Training
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Program,"
Procedure HP 911/2911/3911A, "Health Physics Department Services
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Training Program."
Licensee performance in this area was determined by interview of technician-level and supervisory-level health physics (HP) personnel and review of selected technician training records and resumes.
The licensee increased the normal HP technician staffing level by
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approximately 75 contractor technicians to support outage activities.
Review of technician resumes and training records identified that the licensee was complying with the requirements of SHP 4920.
Permanent HP technicians were upgraded to supervisors and assigned to specific work areas to oversee and provide guidance for contractor techniciar,s.
The inspector observed and interviewed permanent technicians assigned to sten generator (S/G) eddy-current testing (ECT) and noted they were faailiar with ongoing activities and provided effective controls.
Upper-1 vel HP supervision was noted frequently performing tours of
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the work areas to provide additional oversight.
The qualifications of personnel assigned to respiratory protection equipment issue were reviewed against the requirements of procedure HP 911/2911/3911A.
This procedure requires technicians performing HP support activities to qualify by receiving specific training, passing a test and signing off on applicable procedures.
Specific lesson plans included in the procedure cover the following areas: respirator inspection, respirator repair, filter testing, fit-booth operation, and whole body counting...
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The inspector reviewed training records and interviewed several technicians issuing respirators to plant workers and identified the following concerns:
Procedure HP 911/2911/3911A did not include any specific training
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requirements for personnel performing respirator issue nor indicate how they should be qualified. HP supervision indicated respi.ator issue personnel routinely qualified by fulfilling the requirements for respirator inspectors, however this was not specified by the procedure.
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Four individuals performing respirator issue and inspection had not been given the examination. One individual had not signed-off on required procedures.
The majority of respirator issue personnel interviewed indicated
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it was acceptable to issue more than one respirator at a time to a worker. HP supervision stated the intended policy was that only one respirator be issued at a time to a worker.
The licensee was able to demonstrate that the four individuals referenced above had performed all respirator inspection activities under the supervision of a qualified individual. The inspector also identified, by interview of respirator issue personnel, that ;he issuers were aware of respirator user fit-test, training and whole body count requirements. Review of respirator issue logs indicated that respirators were issued to qualified workers.
The licensee immediately completed the qualification process (i.e.,
administered the examination and completed the procedure review ) for the individuals noted above. This was completed during the week of the inspection. The licensee also stated procedure SHP 911/2911/3911A would bc reviewed and revised to specify training and qualification regt.irements for respirator issue personnel.
Lesson plans will be revised to clearly indicate the one-person, one-respirator issue I
policy.
This will be reviewed during a subsequent inspection
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(245/87-31-01,336/87-26-01,423/87-27-01).
4.2 Radiation Worker Training The inspector reviewed training and dosimetry records for selected
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contract workers signed in on radiation work permits (RWPs)
controlling S/G ECT and resistance temperature detector (RTD) system work. The inspector verified workers had received required training and that reviewed exposure histories were complete.
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5.0 Internal Exposure Controls The licensee's program for the assessment and control of internal exposure was reviewed against the following criteria:
- 10 CFR 20.103, "Exposure of individuals to concentrations of radioactive material in air in restricted areas,"
- NUREG-0041, "Manual of Respiratory Protection Against Airborne Radioactive Materials,"
- Procedure SHP 4905, "Radiologic &l Surveys."
Evaluation of licensee performance was based on the following activities:
- discussion with cognizant personnel,
- review of respirator issue logs and respirator user qualifications,
- review of selected respirator fit-test records,
- review of air-sample records and isotope analysis sheets,
- cross-check of radiation work permit (RWP) sign-in times against air-sample times.
Within the scope of the above review, no violations were identified.
The licensee was taking sufficient air samples to adequately assess airborne
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radioactivity. Review of a large number of analysis records indicated that the majority of samples showed negligible activity.
6.0 External Exposure Controls The licensee's program for evaluating radiological conditions and providing
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i, controls over worker exposure was reviewed by the following methods:
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- tours of radiological work areas and observation of
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ongoing work activities,
- inventory of the Unit 3 high radiation area key locker,
- review of the following documentation:
o procedure SHP 4902, "External Radiation Exposure Control and Dosimetry Issue,"
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o procedure SHP 4912, "Radiation Work Permit Completion and Flow Control,"
o various RWPs and applicable ALARA control sheets associated with the following activities:
RTD cut out, S/G ECT and sludge lancing, insulation removal, valve repair, etc.
Within the scope of the above review, no violations were identified. The licensee was noted to be implementing an effective external exposure controls program.
Pre-work surveys were extensive; RWPS included comprehensive controls and were noted to address non-radiological (hydrogen sampling, etc.) as well as radiological safety concerns. The licensee was noted to be using an extensive catalog of plant and system pictures to aid in pre-job work briefings.
Prior to the outage model RWPs were completed for the significant work evolutions that were scheduled. This was noted to be a licensee strength and ensured all necessary requirements were included and not overlooked.
7.0 Posting and Labeling The licensee's program for the survey, posting and control of radioactive materials and radiological areas was reviewed against applicable criteria in 10 CFR 20.203, Technical Specification 6.12, and licensee procedure SHP 4906, "Posting of Radiological Controlled Areas."
Licensee performance in this area was evaluated by review of selected routine radiological surveys and tour of various radiological areas.
Within the scope of this review, no violations were identified.
Radiological posting was satisfactory and housekeeping had not deteriorated significantly despite the outage.
Area decontamination was being performed during slack periods to keep contamination levels down.
8.0 ALARA The inspector reviewed the implementation of the licensee's ALARA program by the following methods:
- discussion with the Unit 3 ALARA coordinator,
- review of outage job exposure estimates and tracking,
- review of ALARA reviews and exposure controls for the RTO and S/G jobs,
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- review ef procedure ACP 6.32, "Maintenance of Occupational Radiation
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Exposures ALARA."
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The licensee has estimated an overall outage exposure of 360 person-rem, with a tot 21 year estimate of 444 person-rem.
The outage modification with the highest estimated exposure is the RTD bypass loop elimination which the licensee indicated should,once completed, substantially reduce future exposure.
Several examples of positive performance were noted in this area and
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include the following:
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Exposure estimates for each job appeared reasonable, despite the lack of a work history file to draw from. Accruing exposure and man-hours for each job were being tracked daily.
The ALARA staff was photographing and videotaping all major evolutions
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for future reference and training.
"ALARA area" signs were placed inside containment to notify workers of
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low dose waiting areas.
Overall S/G ECT exposures were noted to be quite low. At the time of
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the inspection all initial S/G jumps had been completed, with an average jumper ex?osure of approximately 300 millirem.
The ALARA staff was noted to be both familiar with and actively
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following day to day job progress at M.e worksite.
9.0 Unit 1 Drywell Access Ontrols The inspector reviewed licensee restrictions and controls over Unit I drywell access during refueli.'q and fuel movement.
The need for such controls was discussed in two GC generic information letters, sent to licensees in 1973 and 1980. These letters communicated information
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concerning the potential for extreme'y high dose rates to be generated in BWR drywells in certain refueling sitcations.
These situations include a postulated fuel bundle drop incident o.-
evolutions in which a fuel bundle is posi~ioned next to the inner reactor vessel wall.
Effort during this inspection was directed towards ensuring that the licensee had received the GE information letters, that appropriate access j
controls were in place, and review of shielding calculations.
NRC review consisted of discussion with cognizant personnel and review of the l
following:
licensee dose calculations and accompanying memos generated in 1981,
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1984 and 1985 concerning potential fuel drop accide;1ts, procedure 0P 328B, "Fuel Loading / Unloading / Shuffling,"
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procedure OP 328G, "Upper Level Drywell Access Control During
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Irradiated Component Transfer."
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The inspector verified the licensee was aware of the GE informational material and the potential for high dose rates in the drywell during certain accident situations. During refueling, personnel access in the Unit 1 drywell is limited to elevations below the 54' elevation, well below the bioshield wall (which extends to approximately the 71'). All RWPs during the refueling period are stamped in red ink with the upper elevation access limitations; ladders to elevations above 54' are locked and posted
"no-entry."
The licensee utilizes a shielded "cattle chute" in the fuel transfer canal during refueling to reduce dose rates in the drywell during a fuel drop accident.
Licensee projections assume 6 inches of lead equivalent in the cattle chute and indicate dose rates in the drywell due to a fuel drop
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incident in the transfer canal would be negligible.
The licensee's
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evaluations did identify, however, that dose rates above the bioshield wall would be prohibitive in the scenario in which a fuel bundle fell and came to rest against the inner reactor vessel wall, centerline with a main steam line penetration.
The inspector noted the licensee's ar.alysis went beyond the scope of the GE letters, in that it considered the effect of penetrations on dose rates, in addition to access limitations, the licensee has also placed radiation monitors in the upper drywell, with alarms at the drywell access point, to alert personnel to any change in dose rates.
The inspector noted that licensee memos variously indicate that the cattle chute shield contains from 6 to 8.25 inches of lead shielding.
The licensee was unable to produce & drawing during the inspection which
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verified the amount of shielding in the cattle chute.
The licensee indicated suitable documentation would be collected and provided to the
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inspectors.
The following areas will be reviewed during a subsequent inspection:
verification of the thickness and adequacy of the shielding in the
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cattle chute, procedures and training applicable to this area.
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i 10.0 Exit Meeting i
The inspector met with the licensee personnel denoted in Section 1.0 of this report on November 20, 1987. At that time the inspector summarized the purpose, scope and findings of this inspection.
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