IR 05000244/1987005
| ML20205G652 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/24/1987 |
| From: | Cioffi J, Mcfadden J, Shanbaky M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20205G622 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 50-244-87-05, 50-244-87-5, NUDOCS 8703310512 | |
| Download: ML20205G652 (17) | |
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U.S. NUCLEAR REdJi.A ORY COMNIGSION s REGION I ,.
Report N /87-05 , )
Docket No. 50-244 License No. DPR-18 Catepry ( ,
Licensee: Rochester Gas and Electric Corporation .
49 East Avenue ~ ~' ~
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Rochester, New Yorf. 14649 _ __
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Facility Name: Ginna Nuclear Dower Plant ,
Inspection At: Ontario, New York', s
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Inspectfon Conducted: Februpy 2bf(1987
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Inspector - bW
/ J. j.!sfalden, ' Senior 4R4di , tion N.e-ialist
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!ldathM/67 J.pioffi, Radiation fec411st . .
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Approved by: _ _ . fn .% h _ . 3/29/8'7 -
M. Shanbaky, Chief, Facilitres Raat.mion
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_ Protection Section .o .
'Irtsydction Summary: Inspection on February 24-27,1987 (Inspect W keport ~
No. l',5~744M7-05 )
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Areas Insper.tril: Routine unannounced inspection- of the occupativni radiation protectioniirodramduringanoutage,irck'<1tng:
fled items, organization and manageme:it controls, { tate:s training andof(;.:alification previously. inti- of . ,
new personnel l. external exposure ccatrol, internal exposuro contvol, c6ntrol of radioactive material and contamination, surveys and e nitoring and ALARA. Two regionally-based inspectors were on-site for Jbe inscectio Results: k violations were identified, s
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n 8703310512 DR 870324 AJOCK 03000244
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, . DETAILS t l .0 : Persons Contacted
[ Quring the course of th'is inspection, the following personnel were o contacted or interviewed.
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l 1.1 Licensee Personnel
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. G. Daniels, Manager-Electrical Engineering
- C . ,Filkhs, Manage'r-HP and Che:ni stry
- D. Filior, Radiochemist
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, S. Ersteniuk, Instructor-Training 1 3 ,. C. Gibbs, Dosimetry Technician
! *W. Goodman, HP Foreman
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K. Gould, HP M hnician M. Harrison, li Technician -
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C.1@nkins, be Technician (contracted)
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- N. Afejrowski, Training Coordinator-HP and Chemistry J. Krcrr, Dosimetry Supervisor /ALARA Coordinator (acting)
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C. Mcabretti, Senior Nuclear Engineer F. Mi$, Health Physicist
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C. Mitbano, S/G Project Coordinator
- B. Quins,? Corporate Health Physicist
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'* *B. Snow,' Superjntendent - Nuclear Production
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P. Spacher, Hdalth Physicist
- S. Spector, Superintendent - Ginna Production i J. St. Martin, Liaison Coordinator i '
- S. Warren, Health Physicist .
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R. Watts, President - Richard Watts, In .'
c 1 W. Zak,'dP Technician (contracted)
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- Attended the exit meeting on February 27, 1987.
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Add,itional licensee personnel were contacted or interviewed during
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- this inspectio .2 NRC Personnr1 Attending the Exit Interview i ,T..Polich, Resident Inspector 2.0 Purpose
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The purpose of this inspection was to review the licensee's occupational
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Status of previously identified items
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Organization and management controls
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Training and qualification of personnel i
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External occupational exposure control and personal dosimetry l ,
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Internal exposure control and assessment
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Control of radioactive materials and contamination, surveys, and monitoring
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Maintaining occupational exposures ALARA 3.0 Status of Previously-Identified Items 3.1 Routine Inspection Items 3.1.1- (Closed) 86-04-01 (Follow-up Item)
Review licensee commitment to provide documented guidance in their Health Physf cs Work Permit proceoere both for the generation of special radiation work permits (SWPs) (to those authorized to write SWPs) and for what type of surycy documentation methods are required to comply with a SWP as written (to radiation protection technicians).
During this inspection, the inspector reviewed Procedure No. HP-4.3, " Health Physics Work Permit Use" (revision No.
l 27; effective date of February 7,1987). This revised I procedure contained detailed instructions on the mechanics I and radiological safety considerations for generating a l
SW It also detailed when surveys should be performed by radiation protection technicians and where survey results may be documented (on the SWP or on a "SWP/RWP Survey Record Attachment Form" or on a survey map form). This ,
revised procedure is an adequate response to their specific
! commitmen Although this procedure adequately addresses w f- the commitment to pr. ovide documented guidance for what type of survey documentation methods are required to comply with a SWP as written (to radiation protection technicians), the
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I liberal use of wording such as "should", "may", and "can" versus the use of the word."must" or equivalent appears disproportionate. Based on the findings, this item is close .2 Status of Items Related to NUREG-0737 The licensee made significant progress in resolving the concerns addressed in NRC Inspection Report No. 84-04, which inspection was conducted on March 20-22, 1984, and in NRC Inspection Report N , which inspection was conducted on June 10-14, 1985. Two items, which were opened in NRC Report No. 84-04, were closed. Of the fifteen items to resolve in NRC Inspection Report No. 85-08, eight of the items are closed. Seven remain open pending further information or licensee action. The status and resolution of each item is presented belo .
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3. (Closed) 84-04-01 (Inspector Follow-up). Review PASS Functional Testin The inspector reviewed licensee procedures and test results SM-2606.6 and SM-2606.20 to verify functional testing of the PASS. The item is close . (Closed) 84-04-02 (Inspector Follow-up). Review
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representative sampling of PAS .
The inspector reviewed licensee procedures and test results SM-2606.20 series to verify the capability of the PASS to
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obtain representative samples. This item is close ?. (0 pen) 85-08-01 (Inspector Follow-up). Demonstrate that a representative sample can be collected at low RCS pressur The licensee furnished information in a letter dated October 22, 1985, based on a study performed by NUS on minimum reactor coolant system pressures which would be encountered in a design-basis accident. The licensee con-ducted further tests on their own to determine the minimum pressures available from the reactor coolant system to the PASS reactor coolant system (RCS) sample lines. The lic-ensee determined that a minimum flow of 17.2 psi would be available when the reactor coolant pumps were shut of The licensee stated that testing was conducted to verify that representative reactor coolant samples could be ob-tained at this minimum system pressure. However, at the time of the inspection, key personnel were unavailable to furnish the data to verify that representative RCS samples could be obtained under reduced system pressur A licensee representative stated, in a phone conversation, to the inspector on March 3, 1987, that the test data 4 verifying the ability of the PASS to obtain a representa-tive sample under reduced RCS system pressure would be sent to the Region I office. This item remains open pending NRC review, evaluation and acceptance of the results of the test . (0 pen) 85-08-02 (Inspector Follow-up).
Demonstrate that the capability of analytical instrumenta-tion and techniques can be achieved in the presence of the Standard Test Matrix solutio The licensee stated, in a letter to the NRC, Region I dated October 22, 1985, that the manufacturer of the Post Acci-dent Sampling Panel (NUS) had performed tests to demons-trate the capability of the analytical instrumentation and
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l techniques in the presence of the standard test matri However, at the time of the inspection, this material was unavailable. This test and the supporting data will be sent to.the Region I office for review. This item remains open pending NRC review and evaluation for acceptabilit . (Closed) 85-08-03 (Inspection Follow-up).
- Streamline or consolidate processes in the system oper-
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ating procedure to assure that samples can be collected and analyzed within 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> The inspector reviewed licensee procedure PC-25.7.11, re-
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vision 5, " Post-Accident Sampling at the PASS, to determine the adequacy and clarity of directions for following the action sequence in obtaining samples at the PASS. The pro-cedure, while lengthy, was clearly detailed and easy to follow. The inspector also reviewed the lesson plan and drill critique for the Health Physics Semi-Annual drill, conducted April 4, 1986, to time the operation of obtaining samples under accident conditions. This exercise demons-trated that a PASS sample could be obtained and analyzed within the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time constraint specified in NUREG-0737, section II.B.3.
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The licensee conducted another drill on September 10, 1986, and determined that PASS samples were obtained and analyzed
, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- Make appropriate corrections to equations in the core damage assessment procedure.
, Procedure PC-25.4, revision 8, " Guidelines for Interpreting Post-Accident Sampling Results to Estimate Reactor Core Damage" was compared to the earlier revision (revision 4)
which was in place at the time of NRC Inspection No. 85-0 The licensee had appropriately corrected the pressure correction factor equation and the power correction factor equatio . Consider the use of the boron pH probe as a back-up to the primary in-line pH analysis instrumentatio The inspector reviewed licensee procedure PC-25.7.12, re-
, vision 0, " Measuring pH at the PASS using the Boron Analy-zer". This procedure provided guidance and direction on the use of the boron pH probe in the event of failure of the primary in line pH analysis instrumentatio . Indicate the actual method used to isotopically analyze samples in procedures, and develop a consolida-ted library for post accident analysi I a
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The inspector reviewed licensee procedure PC-25.7.10, re-vision 2, " Analysis of Samples Taken at the PASS for Iso-topic Analysis" and a print-out of the isotope library associated with the counting equipment. The inspector found clear instructions on sample preparation and analy-sis. The procedure also provided equations to determine sample activity by hand for all radioisotopes in the librar These four concerns have been resolve .2.6 (Closed) 85-08-04 (Inspector Follow-up). Provide assurances that representative containment air samples can be obtained at the flow indicator's activation threshol The inspector reviewed modification procedure SM-2606.14, in which flow indicator, FS-5, was installed to provide minimum flow indication through the containment air sampl-ing system to ensure representative samplin . Repair the system valves and correct the flow indica-tor designation The inspector reviewed licensee procedure PT-2.5.1, revis-ion 9, " Air Operated Valves, Quarterly Surveillance (Con-tainment)", and noted that valve 955 had been tested for operability on June 12, 1986, and found satisfactor Additionally, all system flow indicator designations were correct, and the correct sequences were incorporated into procedure PC-25.7, revision 9, " Operation of Post Accident Conditions - Master Procedure." Include the system spare parts in the inventory syste The licensee provided the inspector with copies of the plant Spare Parts Lists for the PASS system. Inspector review of these documents determined that the required sys-tem parts, components, and supplies appeared to be include The above items are close .2.7 (Closed) 85-08-05 (Inspector Follow-up).
Calibration corrections are needed for the SPING -4 moni-toring systems to correct for the reduced p essure in the l noble gas detector volume l
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The inspector reviewed procedure change notice number 87-
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T-302 to procedure CP 223, revision 18, and verified that appropriate pressure corrections had been made to calculate
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noble ga.s activity. This item is close . (0 pen) 85-08-06 (Inspector Follow-up).
Empirical checks should be made to verify the calculated
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sensitivities of the shielded detectors used for monitoring steam lines A and The licensee provided the inspector with a copy of a pro-posal from TEC, Incorporated to perform the necessary mea-surements through system mock-ups, in order to evaluate and verify the capabilities of the steam line monitor The licensee is currently under budgetary restrictions which have prevented the study from being performed. This item remains ope '
3. (Closed) 85-08-07 (Inspector Follow-up).
An analysis should be made of the potential radiation dam-age to the micro-computer in the SPING-4, and of the ability of its low and mid-range detectors to withstand high concentrations of noble gases.
Through a review of licensee responses to NRC Inspection
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No. 85-08, additional documentation on-site, and discuss-ions with licensee and contractor personnel, the inspector determined that the SPING-4 microprocessor would not suffer from radiation damage. In a worst case situation,
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occur in approximately 3 minutes, with an integrated dose to the SPING-4 microprocessor of 660 rads. The licensee derived this information from the Westinghouse Core Damage Assessment inventory and used EPA 520 documentation to determine dose rate The licensee also evaluated the potential conditions of the low and mid-range detectors under accident conditions and determined that the low and mid-range detectors would not
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lea _e because the 2 detectors are located in a shielded configuration with charcoal and particulates pre-filtering prior to reaching the detectors.
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This item is closed.
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3.2.10 (Closed) 85-08-08 (Inspector Follow-up).
Upper and lower set points for triggering the sample flow rate alarms on the effluent monitors should be set at points reasonably close to normal sampling rates. I&C cali-bration procedures should include the restoration of flow limits to appropriate values after testin The inspector reviewed procedure CP-223, revision 18, "Cali-bration and/or Maintenance of SPING-4 Radiation Mcnitors for R-12A, R-14A, and R-15A Radiation Monitoring Channels,"
and determined that I&C does calibrate the SPING flowmeter during the calibration and/or maintenance of the SPING-4 monitors. Flow rates and alarm set points are performed by Health Physics personnel using procedure HP-14.11, revision 1, "SPING Sample Flow Rates and Alarm Set Points," using ANSI N 13.1-1969 information and empirical dat This item is close .2.11 (Closed) 85-08-09 (Inspector Follow-up).
Verify that monitor calibration procedures include the appropriate tests of GM plateaus and detector response The inspector reviewed procedure CP-223, revision 18, "Cali-bration and/or maintenance of SPING-4 Radiation Monitors for R-12A, R-14A, and R-15A Radiation Monitoring Channels."
The procedure contained directions for plotting GM plateaus and checking detector responses. Procedure CP-225, revision 4, " Calibration and/or Maintenance of DAM-3 Steam Line Radiation Monitors R-31 or R-32" contained directions for calibration that referenced the manufacturer's technical manua This item is close .2.12 (0 pen) 85-08-10 (Inspector Follow-up).
Ensure that particulate and iodine filters and grab samples can be changed, transported, and analyzed within the GDC limits of 5 rems to the whole body and 75 rems to the extremitie The licensee performed calculations to determine the dose rates to individuals while obtaining the iodine and parti-culate filters, and grab samples. However the procedure for changing and transporting the filters had not been tested to ensure that these activities could be performed I
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within a safe time limit to ensure doses will not be exceeded. Also the inspector identified some confusing steps in the procedure that addressed sample analysis and then referred back to obtaining another sample. The licensee stated that the procedure would be revised and demonstrated by actual procedure walk-through that the required sample acquisition, filter changes and analyses could be performed within the GDC criteri This item remains ope .2.13 (Open) 85-08-11 (Inspector Follow-up).
Determine line losses through sample lines to ensure representative samplin The licensee plans to participate in a multi-utility study on sample line losses which will be conducted by Science Application International Corporation. This item remains open pending further licensee action in this matte .2.14 (Closed) 85-08-12 (Inspector Follow-up).
Degradation of particulate and iodine filters for contain-ment vent monitors due to high humidity /high temperature of accident condition The inspector reviewed licensee modification packages SM-2504.3 and SM-2504.4 and visually inspected the containment ventilation and containment purge to verify the installa-tion of blind flanges. These flanges can only be manually removed at cold shutdow Therefore the SPING-4 system that monitored the containment vent will no longer be of use. The licensee installed a containment mini purge sys-tem which vents into the auxiliary building ventilation system. The auxiliary building ventilation exhausts into the plant vent which is monitored by the no'rmal effluent monitors as well as by the SPING-4 monitoring syste This item is close .2.15 (0 pen) 85-08-13 (Inspector Follow-up).
Verify the positioning of the high-range containment moni-tors inside containmen The inspector reviewed "as-built" Prints and Instrumenta-tion Diagrams (P&ID's) for the placement of the High Range
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Containment Radiation Monitors, a report by Technology for Energy Corporation on the calculational relationship bet-ween the containment high range monitors and a radioactive
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material release and toured the containment to verify the placement of the monitors. The inspector noted that moni-tor number 2 was placed so that it " viewed" a large un-obstructed fraction of containment. However, monitor number 1 was located behind the shield wall of steam genera-tor A, and did not appear to fully meet the requirement of item II.F.1, attachment 3 of NUREG-0737 for viewing a large, unobstructed fraction of the containment volum This item will be reviewed further by the NR .2.16 (Closed) 85-08-14 (Unresolved).
Review the environmental qualifications of the high range containment monitor's installed cable assembl This concern has been addressed in an environmental qualiff-cation inspection, NRC Inspection Report No. 87-0 This item is close .2.17 (0 pen) 85-08-15 (Inspector Follow-up). The operating procedure for the back-up counting equip-ment in the environmental lab was incomplete and difficult to follo The licensee planned to replace all of their counting equipment computers with a new system. At the time of the inspection, the old equipment was still in operation. This part of item 85-08-15 remains ope . Counting efficiency factors for the silver zeolite cartridge configuration were not available for the full range of anticipated concentration The licensee performed the necessary measurements to determine counting efficiency factors for all detector geometries and distances from the detector. This con-cern is resolve . Only a small number of technicians were trained in the use of the equipmen The licensee stated that all technicians were trained on the primary counting equipment and five technicians were trained to use the back-up counting equipment in the environmental la This concern is resolve . There was no requirement to purge a sample cartridge with clean air to remove noble gas prior to analysi .
The inspector reviewed procedure HP-11.2, revision 22,
" Iodine in Air - Charcoal and Silver Zeolite Cartridge Method" and found steps in the procedure and precau-tions for purging all cartridges prior to counting for iodine. This concern is resolve .0 Organization and Management Controls The licer.see's organization and management controls for the radiological protection function were reviewed against criteria contained in:
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10 CFR 50, Appendix B, QA Criteria for Nuclear Power Plants
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Licensee Technical Specification 6.0, Administrative Controls
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Regulatory Guide 8.8, Information Relevant to Ensuring that Occupa-
, tional Radiation Exposures at Nuclear Power Stations will be as low
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as reasonably achievable
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Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), November 1972
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ANSI /ANS-3.2-1982, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants The licensee's organization and management controls for the radiological protection function were evaluated against these criteria by the following:
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discussions with licensee radiological protection personnel
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review of recently revised radiological protection procedures such as:
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Procedure No. HP-4.3, Health Physics Work Permit Use (revision no. 27, effective date of February 7, 1987)
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Procedure No. HP-6.3, Personnel monitoring, Decontamination and Dose Assessment (revision no. 6 effective date of February 7, 1987)
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review of licensee records such,as:
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controlled area incident log (CAIL)
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personnel contamination incident log (PCIL)
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HP shift turnover log Within the scope of this review, no violations were identified.
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The inspector noted that for the outage, the licensee had made a success-ful effort to bring back contracted radiation protection technicians who had performed well in the past.
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The inspector made the following observations during the inspection, discussed them with the licensee:
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there was not an approved procedure provided to assure that the Eberline SRM-100 scaler with the 100 cm2 probe was properly control-led, calibrated, and adjusted at specified periods to maintain accuracy;
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the log entry for the highest level (in counts per minute-hours
. -(cpm-brs)) personnel contamination incident in the PCIL from February 6,1987 through February 24, 1987, which by procedure should have led to a skin dose evaluation, was recorded as not requiring a skin dose evaluation; this entry was not identified as requiring followup until well past the next HP shift turnover after the entry;
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there was a large handwritten flowchart on a message board at the frisking-out area; this flowchart did not accurately reflect Procedure No. HP-6.3 in that:
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the procedure did not mentioned a 100 cm2 probe while the flow-chart did, and
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the flow chart indicated that, if a frisk of a potentially con-taminated areas with the SRM-100 and 100 cm2 probe resulted in a reading of less than a stated level, the individual could be immediately released;
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the PCIL contained handwritten headings for information to be entered, that had changed during the outage and still did not provide all information required by a reviewer in that:
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entry headings need to be more explicit to differentiate between clothing contamination only, skin contamination only, or cloth-ing and skin contamination for each entry
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the identity of the person making the entry was not always avail-able. Most entries made during February 6-8, 1987 were initial-ed; some entries during February 9-11, 1987 were initialed; hardly any entries during February 11-26, 1987 were initiale The licensee stated that these previously cited items would be reviewe During the discussion of these observations, the licensee stated that the SRM-100 with the 100 cm2 probe was undergoing evaluation, that it had been calibrated, and that a calibration and operability check procedure was in j draft form. The calibration records were reviewed by the inspecto A similar situation occurred during NRC Inspection No. 50-244/86-04 when l the licensee was using an automated frisker in an evaluation mode. In I
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both these cases, the controls to ensure that the instrumentation was used in the evaluation mode (i.e. dual measurements with the instrument under I
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evaluation and with an instrument for which an approved calibration and operability check procedure existed) were not apparent. The licensee stated that, effective immediately, if the SRM-100 meter with 100 cm2 probe was used, its use would be followed by the use of an instrument with an approved procedure before a decision to release as uncontaminated is mad l
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For the case where a personnel contamination incident entry of 300,000 cpm-brs did not result in a skin dose evaluation form being generated, the licensee determined that a scale reading error had occurred and that the incident involved 30,000 cpm-hrs which still should have resulted in a skin dose evaluation form being generated. A subsequent skin dose evalua-tion by the licensee determined the assigned dose to be 630 mra The licensee stated that the previously-mentioned flow chart and personnel contamination incident log were in the process of being refined on a continuing basis to accurately reflect the intent of the recently revised Procedure No. HP-6.3. During review of recently revised procedures, no HP-4.3 and HP-6.3, the inspector noted that there was no method used to highlight additions, deletions and changes. New/ revised HP procedures are routed to HP personnel for reading and initiuling via a training depart-ment form. The mechanism used to verify that personnel understand what changes have been made in revised procedures will be reviewed in a future inspectio .0 Training and Qualification of Personnel The licensee's program for training and qualification of personnel was reviewed against criteria contained in:
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10 CFR 20.206, Instruction of Personrel
Licensee Technical Specification 6.0, Administrative Controls
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The licensee's performance relative to these criteria was determined by:
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discussions with licensee and contracted radiological protection personnel
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review of the following records for contracted radiological j protection personnel:
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experience summaries
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HP qualification cards and tests
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general employee and radiation worker training
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review of steam generator (S/G) mock-up training records for RG&E steam generator workers i Within the scope of this review, no violations were identified.
l The inspector's review of the S/G Crew Leader Training Program Manual and
two-week training agenda indicated a strong licensee commitment to this l
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type of preparation for outage work. The licensee stated that other RG&E personnel working with the crew leaders also received hands-on training utilizing the licensee's two S/G mock-up .0 External Occupational Exposure Control and Personnel Dosimetry The licensee's program for external occupational exposure control and personnel dosimetry was reviewed against criteria contained in:
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10 CFR 20, Standards for Protection Against Radiation
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Licensee Technical Specification 6.0, Administrative Controls
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Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), November 1972
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Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as Reasonably Achievable The licensee's performance relative to these criteria was determined by review of the following:
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individual cumulative dose records for 1987 (whole body, skin, and extremities) (based on TLDs & SRDs)
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skin dose determinations based on calculations and on skin contamination levels
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initial S/G doserate survey records for bowls and handholes
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multiple whole body and extremity badging protocols for S/G work
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recently revised Procedure No. HP-4.3, Health Physics Work Permit Use (revision no. 27; effective date of February 7,1987)
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selected active and completed special radiation work permits (SWPs)
The licensee's performance was also determined through discussions with licensee and contracted radiological protection personnel and observations by the inspecto Within the scope of this review, no violations were identifie The licensee was updating cumulative individual doses on a daily basis for each of two shifts. Copies of these updates were available in the dosi-metry office for review by work group supervisors. Adequate exposure control on high dose rate jobs was being provided as demonstrated by the quality of initial dose rate surveys on S/Gs and by the experienced radia-tion protection technicians used at the S/G control point The inspector noticed discarded self-reading dosimeters (SRDs) in the anti-C clothing removal area on several occasions. SRDs are supposed to be read and left at the HP desk at the RCA access point after the user records his reading. The licenseo stated that, since a user's name and the SRO serial number are recorded at SRO issuance, the users, leaving SR0s in the anti-C clothing removal area, can be and are tracked down and are reinstructed in the proper method of returning SR0 .
The revised SWP procedure was reviewed by the inspector and was discussed in section 3.1.1 of this repor .0 Internal Exposure Control and Assessment The licensee's program for internal occupational exposure control and assessment was reviewed against criteria contained in:
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10 CFR 20, Standards for Protection Against Radiation
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Licensee Technical Specification 6.0, Administrative Controls
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Regulatory Guide 1.33, Quality Assurance Program Requirements (0peration), November 1972
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Regulatory Guide 8.8, Information Relevant to Ensuring that Occupa-tional Radiation Exposures at Nuclear Power Stations will be As Low As Reasonably Achievable The licensee's performance relative to these criteria was determined by review of:
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whole body counting results
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daily MPC-hours tracking records
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selected respirator use sheets
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air sampling requirements on selected SWPs
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discussions with licensee and contracted radiation protection personnel
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observations by the inspector Within the scope of this review, no violations were identifie Based on whole body counting results (none greater than 3 percent maximum permissible organ burden) and on daily MPC-hours (none greater than one MPC-hours per day) during the outage, the licensee was providing adequate internal exposure control and assessment. However, there were two items which indicated that greater attention to detail might be require First, a tool decontamination tent in containment which in past outages had been equipped with a HEPA-filtered portable ventilation unit was in use without a unit since there was none available (one out of service for repair and one being used for asbestos removal work). Second, a licensee memo, titled " Tritium MPC-hours during 1986 Annual Outage" and dated June 18, 1986, stated that the dose commitment for an intake of 40 MPC-hours of tritium was approximately 4 millirem. Such an acute intake would actually result in a dose of an order of magnitude greater than that stated in the mem This error in documentation was not detected during initial review by the license The licensee stated during this inspect-ion that the error in the memo would be correcte The results of airborne tritium monitoring in containment for this outage were reviewed by the inspector. The readings for February 10-15, 1987 averaged approximately fif teen percent of MP r
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8.0 Control of Radioactive Materials and Contamination, Surveys, and Monitoring The licensee's program for control of radioactive materials and contamina-tion, surveys, and monitoring was reviewed against criteria contained in:
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10 CFR 20, Standards for Protection Against Radiation
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Licensee Technical Specification 6.0, Administrative Controls
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Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), November 1972 The licensee's performance relative to these criteria was determined by:
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review of:
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recently revised Procedure No~ HP-6.3, Personnel Monitoring,
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Decontamination and Dose Assessment (revision no. 6; effective date of February 7, 1987)
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personnel contamination incident log
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discussions with licensee radiation protection personnel
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observations by the inspector Within the scope of this review, no violations were identifie The following items were discussed with the licensee:
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initial contamination levels for five individuals involved in a con-tamination incident were not recorded in the incident log; the lic-ensee followed up on this matter and determined that the initial contamination levels did not require skin dose evaluations
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rubber boots were being reused at the containment access control point, and on one occasion, the inspector observed that the drum con-taining reused rubber boots was overflowing and was positioned bet-ween the clean step-off pad and the container for used gloves; upon exit, the piled up boots represented a contact contamination hazard since one had to lean across the boots in order to deposit the used gloves
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accumulation of used anti-C clothing immediately ad,jacent to frisking area was apparently causing false positives during frisking
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a bag of radioactively contaminated material was transferred from its
. point of generation to the RCA exit while unlabeled
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one worker was observed in the RCA with his lab coat open
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The licensee stated that the need for shielding for the friskers in close proximity to potentially contaminated anti-C clothing had been identified but that e.he ordered shielding material had arrived too late to be in-stalled for this outage.. The license also stated that the other previously-cited concerns would be reviewe .0 Maintaining Occupational Exposures ALARA The licensee's program for maintaining occupational exposures ALARA was reviewed against criteria contained in:
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10 CFR 20.1, Purpose
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Licensee Technical Specification 6.0, Administrative Controls
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Regulatory Guide 1.33, Quality Assurance Program Requirements (Operations), November 1972
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Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be As Low As Reasonably Achievable
The licensee's performance relative to these criteria was determined by:
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discussions with licensee radiation protection personnel
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observations by the inspector Within the scope of this review, no violations were identifie The inspector noted that accumulated person-rem for dose-significant jobs was being tracked. The accumulated person-rem for steam generator work was greater than that predicte The licensee stated that this fact was i due to the actual number of tube defects found being greater than the j preoutage estimat In early January 1987, a draft procedure, Procedure No. QE325, "ALARA/
- Radiation Safety Design Review," to be included in the Engineering Proce-
- dures Manual, was available. This procedure included an ALARA checklist l for any modifications, additions, or repairs which trigger certain radia-tion protection considerations. Although a number of engineering work
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requests (EWRs) were being implemented during this outage, the licensee
, stated that a formal ALARA design review mechanism had not been fully
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implemented.
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10.0 Exit Interview The inspector met with the personnel denoted in section 1.0 at the con-clusion of the inspection on February 27, 1987. The scope and findings of the inspection were discussed at that time.
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