ML20137U674
ML20137U674 | |
Person / Time | |
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Site: | Vogtle ![]() |
Issue date: | 03/27/1997 |
From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
To: | |
Shared Package | |
ML20137U658 | List: |
References | |
50-424-97-02, 50-424-97-2, 50-425-97-02, 50-425-97-2, NUDOCS 9704170065 | |
Download: ML20137U674 (17) | |
See also: IR 05000424/1997002
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U.S. NUCLEAR REGULATORY COMMISSION
REGION II
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Docket Nos.:
50 424, 50 425-
License Nos:
NPF 68, NPF 81
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Report Nos:
50 424/97 02, 425/97-02
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Licensee:
Southern Nuclear Operating Company, Inc.
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Facility:
Vogtle Electric Generating Plant
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. Location:
8805 River Road
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Waynesboro GA 30830
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Dates:
February 10-14, and March 3 7, 1997
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Inspectors:
George B. Kuzo. Senior Radiation Specialist
J. Kreh, Radiation Specialist
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Approved by:
K. Barr, Chief. Plant Support Branch
Division of Reactor Safety
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9704170065 97
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EXECUTIVE SUMMARY
Vogtle Electric Generating Plant
NRC Inspection Report Nos. 50 424/97 02, 425/97 02
This routine announced inspection reviewed and evaluated occupational
radiation protection, radioactive liquid and gaseous waste, solid radioactive
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waste (radwaste), and radioactive material transportation programs.
Specific
program areas reviewed and evaluated by the inspectors included general
employee training; radiation monitoring system (RMS) equipment operability;
radioactive effluent processing and release; meteorological station
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operations; and status of radwaste processing equipment and storage
facilities. The adequacy of associated procedures and radiological controls,
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and staff proficiency for radioactive waste processing and transportation
rogram activities were evaluated.
In addition, selected Safety Audit and
Engineering Review (SAER) audit findings and corrective actions were discussed
and evaluated. Conclusions included the following:
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In general, controls for low-level radioactive waste (radwaste) and
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material processing and storage met Technical Specification (TS) and
10 CFR Part 20 requirements.
Labels for containers of radioactive
naterials and waste and radiological controls for high radiation and
locked high radiation areas were in accordance with 10 CFR Part 20 and
TS requirements. Housekeeping was acceptable within the auxiliary
buildings, radwaste processing and storage facilities. An example of a
non cited violation (NCV) of TS 5.4.1(a) was identified for failure to
follow radiation protection procedures for maintaining radioactive waste
processing facility dose rates within procedural limits. (Section R1.1).
In general, RMS and meteorological equipment and systems were operable
and calibrated ap3ropriately. Corrective actions to address RMS
equipment or cali) ration issues identified in a recent SAER audit, e.g.,
inadequate electronic calibrations requirements, were adequate. An
unresolved item (URI) was opened regarding adequacy of the containment
high radiation monitor sensitivity to meet criteria detailed in
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NUREG 0737, Clarification of Three Mile Island (THI) Action Plan
Requirements, Item II.F.1-3 (Section R1.3).
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Excluding an October 23, 1996, radioactive material shipment,
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transportation and packaging activities for radwaste or radioactive
material shipments reviewed were implemented appropriately and met
10 CFR 71 and 49 CFR requirements.
For the October 23, 1996 shipment,
an apparent violation of 49 CFR 173.475 requirements was identified for
the failure to prepare a radioactive material package for transport such
that, under conditions normally incident to transportation, radiation
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levels on the external surface of the package exceeded 10 millisievert
per hour (mSv/hr) [1000 millirem / hour (mrem /hr)] (Section R1.2). The
licensee provided appropriate hazardous material (hazmat) training and
implemented, as required, revised Department of Transportation (DOT)
guidance (Section R5.2).
In general, sampling, analyses and processing of a liquid radioactive
waste tank for release was conducted in accordance with Operations and
Chemistry procedures, and Offsite Dose Calculation Manual (0DCM)
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methodology.
During observation of the liquid radioactive waste tank
release, a second example of a NCV of TS 5.4.1(a) was identified for
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failure to follow procedures regarding analysis of liquid waste tank
radionuclide concentrations. (Section R1.3).
The Engineered Safety Feature (ESF) ventilation systems were maintained
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appropriately and tested in accordance with TS requirements. Test
results for selected ESF systems were acceptable (Section R2.3).
General employee training was conducted in accordance with established
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commitments and procedures (Section R5.1)
Counting room quality control (QC) activities associated with effluent
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measurements were technically adequate (Section R7.1). Audits of
radioactive waste, effluent and transportation program activities were
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thorough and met TS, ODCM, 10 CFR Parts 20 and 71 requirements
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(Section R7.2).
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Report Details
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IV. Plant Support
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R1
Radiological Protection and Chemistry Controls
R1.1 Radioloaical Controls
a.
Inspection Scope (84750. 867501
During numerous tours of site radiologically controlled areas (RCAs),
the inspectors reviewed radir, logical controls associated with liquid and
gaseous waste processing Scilities, and with radioactive material and
waste storage areas and observed general housekeeping and cleanliness.
The toured areas included the auxiliary building and remote radioactive
material storage and processing buildings located within the Owner
Controlled Area. Radiation dose rate and contamination surveys were
conducted for selected areas and storage equipment.
Established controls were compared against procedural requirements,
Technical Specification (TS) 5.7 and 10 CFR Part 20 Subpart J
requirements, as applicable.
b.
Observations and Findinas
Postings and physical controls to limit personnel exposure from external
sources in restricted areas were in accordance with TS and 10 CFR Part 20 requirements. Administrative and physical controls for high
radiation and locked high radiation areas were in accordance with TS
requirements.
Label information for containers of radioactive materials
and waste met 10 CFR Part 20 and procedural requirements.
Excluding the radwaste processing building, dose rate and contamination
surveys conducted verified proper radiological controls were implemented
and corroborated current survey results. On February 11, 1997,
following tours of the dry active waste (DAW) processing facility,
surveys conducted on the outside surface of the facility walls
identified dose rates of approximately 2 millirem >er hour (mrem /hr),
which exceeded the procedural limit of 0.250 mrem /1r specified in
arocedure, 46102 C, Operation of the Support Systems in the Dry Active
Waste Processing and Storage Facilities, Rev. 2, approved March 28,
1994. The procedure requires that material stored inside the DAW
storage and processing facilities must be arranged so that the dose rate
on contact with facility walls is less than 0.25 mrem /hr. The inspector
noted that the failure to follow procedures for maintaining dose rates
within the established limits was a violation of TS 5.4.1(a).
Licensee
followup and root cause investigation determined that boxes of outage
equipment in close proximity to the inside walls resulted in the outside
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wall dose rate limits being exceeded. The infrequent use of the
facility and reduction in survey frequency resulted in misinterpretation
by Health Physics (HP) technicians regarding completion and
documentation of the required surveys.
Immediate corrective actions
documented in Radiological Incident Report (RIR) No. 97 003, included
rearrangement of the boxes and verification that dose rates were within
limits and posting of storage requirements for the facilities.
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addition, signs were posted at both DAW processing and storage
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facilities specifying surveys re
map documents were revised and ) quired, and dose rate survey and survey
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reprinted with established limits. The
RIR was to be included in all s11ft briefings.
In addition,
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improvements were made to survey maps and documentation for Spent Fuel
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Rack storage area and the Alternate Radwaste building.
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c.
Conclusions
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In general, controls for low level radioactive waste (radwaste) and
material processing and storage met TS and 10 CFR Part 20 requirements.
Labels for containers of radioactive materials and waste were in
accordance with 10 CFR Part 20 requirements. Consistent with Section IV
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of the Enforcement Policy based on corrective actions taken prior to the
end of the inspection, the DAW Processing dose rates exceeding
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procedural limits was identified as a non cited violation (NCV) 50 424,
50 425/97-02-01:
Failure to follow radiation protection procedures for
.a DAW processing facility dose rate limits.
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R1.2 Radioactive Waste and Material Transoortation Activities
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Inspection ScoDe (86750. TI2515/1331
The inspectors evaluated and discussed the licensee's current guidance
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for radioactive material and waste packaging and transportation program
activities. The following procedures were reviewed and evaluated
against recently revised 10 CFR Part 20, 49 CFR Parts 100 179 and
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10 CFR Part 71 regulations.
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46004 C, Shipment of Radioactive Material, Rev.13, approved
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April 4, 1996.
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46100 C,10 CFR61 Waste Classification Sampling Program, Rev. 2,
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approved April 4, 1996.
46102 C, Operation of the Support Systems in the Dry Active Waste
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Processing and Storage Facilities, Rev. 2 approved March 28,
1994.
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46110 C, Shipment of Radioactive Waste, Rev. 6 Approved April 4,
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Records associated with packaging and shipping of radioactive material
and waste to either vendor processing facilities or directly tc a
licensed burial facility were reviewed and discussed with responsible
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personnel. The following shipment records were reviewed in detail.
Radioactive Waste Shipment (RWS) Number (No) 96 001, Radioactive
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Material, Low Specific Activity, NOS. 7, UN 2912, Reportable
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Quantity (RQ) Fissile Exempt, containing dewatered Ion Exchange
Resin (Bead) from Plant Demineralizer System, shipped March 16,
1996.
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RWS No. 96 004 Radioactive Material, Low Specific Activity,
NOS. 7. UN 2912, Reportable Quantity (RQ) Fissile Excepted,
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containing dewatered Ion Exchange Resin (Bead and Powdex) from
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Plant Demineralizer System, shipped September 27, 1996.
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RWS No. 96 005, Radioactive Material, Low Specific Activity,
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NOS. 7. UN 2912, Reportable. Quantity (RQ) Fissile Excepted,
containing dewatered Ion Exchange Resin (Bead and Powdex) from
Plant Demineralizer System, shipped December 13, 1996.
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96 10 002, Radioactive Material Shipment, Surface Contaminated
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Object 2, NOS. 7. UN 2913, containing a-reactor coolant pump,
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dated October 13, 1996.
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96 10 005, Radioactive Material Shipment, Surface Contaminated
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Object 2, NOS. 7, UN 2913, containing seven strong tight packages
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of Fuel Sipping, UT, Fuel Reconstitution and RCCA equipment
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returned to a vendor dated October 23, 1996.
b.
Observations and Findinas
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The insmetors verified that changes to 49 CFR Parts 100 179 and
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10 CFR ' art 71 regulations were incorporated into the current procedures
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and were implemented as rethe inspectors' quired.
Excluding material shipment No.
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96 10 006,
reviews of shipping paper. documentation
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. verified that ' applicable regulatory requirements were met. ~ As
applicable, the inswetors verified the licensee was a registered user
of the shipping casts'and that the appropriate Certificates of
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Compliance and associated documents were maintained at the facility.
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Notifications and licensee followup regarding package surface dose rate
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concerns for'an October 23, 1996, shipment were reviewed and discussed
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in detail. On November 6,.1996, an NRC Region II (RII) Radiation
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Specialist was informed by NRC Region I staff of an October 23, 1996
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Vogtle shipment received at the Westinghouse Waltz Mill,. PA. vendor
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facility which contained a package having maximum surface dose rates of
approximately 20 millisievert per hour (20 mSv/hr) (2000 millirem per
hour [ mrem /hr]), which exceeded.NRC/ DOT 49 CFR 173.475 package surface
limits of 10 mSv/hr [1000 mrem /hr]. Subsequent discussions during
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November 7, 1996, teleconferences between RII and the Vogtle Radiation
Protection supervisors, indicated that plant management was unaware of
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the identified issue.
Preliminary review of Vogtle survey records for
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the out-going shipment documented a maximum surface dose rate of
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approximately 120 mrem /hr.for the subject package. Review of additional
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survey records for the individual pieces of equipment and the transport
vehicle' indicated that the package contained fuel reconstitution
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equipment having an initial maximum dose rate of 800 mrem /hr associated
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with a camera stand,-a one inch diameter hollow tube which had been
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stored in the sgnt fuel p l.
Recorded dose' rates associated with the
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transport vehicle were within regulatory limits. The package was part
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of. an exclusive use, closed transport vehicle shipment. Upon arrival at
the vendor facility on October 25, 1996, measured dose rates for the
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transport driver's position and on the outside of the transport vehicle
were within regulatory limits..
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On November 11, 1996, a Vogtle transportation specialist was dispatched
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to the Waltz Mill, PA site to review and evaluate the identified issue.
The Vogtle specialist independently verified that within a localized
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area on the package surface, dose rates exceeded regulatory limits of 10
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mSv/hr (1000 mrem /hr) specified in 49 CFR 173.475.
Further review,
determined that the equipment was Jackaged appro)riately and that
elevated dose rates, up to 50 mSv/1r (5000 mrem /1r), were measured at
one end of the camera stand. The investigation determined that the
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elevated dose rates resulted from a small, microscopic piece of crud
material which had not been identified during the original surveys and
was located at one end of the camera stand at the time of survey.
The licensee's root cause determination documented insufficient
preparation time allocated for the task and less than adequate task
distribution for work organization methods.
In addition, the lack of
standard policy or administrative controls were documented as an
additional root cause of the event.
Documented corrective actions
included ensuring all Health Physics (HP) and Decontamination
Technicians are trained in the requirements for removing material from
the spent fuel pool and subsequently shipping the objects offsite. The
vendor was also to complete cleanliness procedures regarding materials
and equipment stored in the spent fuel pool. Also, the need for
additional shielding of boxes was to be reviewed by the vendor.
c.
Conclusions
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In general, transportation and packaging activities for radioactive
waste or material shipments met 10 CFR Part 20, 71.5 and 49 CFR 100179
requirements. The licensee was implementing, as required, revised DOT
guidance. The failure to prepare a radioactive material packaga for
transport on October 23, 1996, such that, under conditions normally
incident to transportation, radiation levels on the external surface of
the package did not exceed 10 millisievert per hour (mSv/hr) [1000
millirem / hour (mrem /hr)] was identified as an apparent violation of
49 CFR 173.475 requirements: Escalated Enforcement Item (EEI) 50 424,
425/97 02 02: Failure to meet 49 CFR 173.475 package dose rate limits.
R1.3 Padioactive Waste Analysis. Processina and Release
a.
Insoection Scooe (84750)
During the onsite inspection, liquid radioactive waste analysis,
processing and release activities were reviewed.
Evaluated program
areas included equipment operability, procedural adequacy and staff
proficiency.
On March 4,1997, the ins)ectors directly observed and evaluated
activities associated wit 1 a Unit 2 (U2), No. 9 Waste Monitor Tank
(WMT 9) liquid effluent release. The review included pre-release sample
collection and radiological analyses, determination of the liquid
effluent radiation monitor (RE 18) setpoints, and operations associated
with subsequent release to the environment.
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-The following procedures were reviewed and evaluated during observation
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of the WMT processing and release:
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. Procedure 33035 C, Gamma Spectroscopy for Radiochemistry, Rev. 17.
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Procedure 34311 C, Operation of Digital Radiation Monitoring
System-(DRMS) Liquid Release Monitors 1(2)RE 0018)
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-Procedure 34331 C, Management of DRMS Status and Parameter,
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Rev. 9.-
Procedure 35420 C, Monitoring of Radioactive Liquid Waste
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Management System Rev. 15.
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Procedure 36015 C, Radioactive Liquid Effluent Release Permit
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Generation and Data Control, Rev.14.
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Procedure 37000 C, Sample Handling, Rev. 2.
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Personnel observed and interviewed regarding the liquid radwaste
processing and release evolutions included Operations and Chemistry
staff.
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b.
Observations and Findinas
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The inspectors noted that the current procedures were adequate for
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sample collection preparation, analysis, set point determination and
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final release documented in liquid release permit No. 97 0018 L.
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During observation of the release, concerns were noted for procedural
adherence and for the material condition of a radiation flow indicator-
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(FI) used to. verify the liquid waste monitor' flow rates.
Procedure
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33035 C, Rev.17 required a 1000 milliliter sample to be analyzed by
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gamma spectroscopy analysis to quantify the release radionuclide
concentrations. The inspectors noted that the technician failed to
accurately determine the exact volume. Evaluation of the sample used
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for the analysis verified that the volume did not meet the procedural
requirements. Review of the gamma spectroscopy data verified that the
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error was conservative and for this specific case, did not affect the
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final set points nor release data and dose estimates.
Licensee
management reviewed procedural guidance, training provided and stated
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that all responsible personnel would be instructed in the need for
accurate sample volume for completing the quantitative analyses.
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addition, the visibility of the float within Flow Indicator 2 FI 0018,
was marginal.
Licensee representatives stated that corrective actions
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regarding the flow indicator were initiated prior to the end of the
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onsite inspection.
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.c.
. Conclusions
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In. general, release of a Unit 2 WMT 9 was conducted in accordance with
0perations and Chemistry. procedures, and Off site Dose Calculation Manual
. methodology. Consistent with Section IV of the Enforcement Policy based
on corrective actions taken prior to the end of the inspection, the
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failure.to analyze a 1000 milliliter sample for gamma spectroscopy
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analysis was identified as a second example of NCV 50 424, 425/97-02 01:
Failure to follow procedures in accordance with TS 5.4.1(a) for a liquid
waste processing gamma spectroscopy analyses.
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R2
Status of Radiation Protection and Chemistry Equipment and Facilities
R2.1 Meteoroloaical Monitorina Proaram and Instrumentation
a.
Inspection Scope (84750)
The inspectors reviewed the licensee's meteorological program and
equipment against specifications detailed in FSAR Section 2.3.3, Onsite
Meteorological Measurements Program.
In addition, instrumentation and
equipment operability, calibration and maintenance were verified,
b.
Observations and Findinas
On February 13, 1997, the inspectors observed licensee personnel
performing daily checks of the meteorological data collection center
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(MDCC) instrumentation, located at the base of the 60-meter tower at the
periphery of the plant site. The observed checks were conducted in
accordance with procedure 36030 C, Meteorological Tower Monitoring and
Data Control, Rev. 11. The individual conducting this surveillance was
systematic and thorough.
Selected records of calibrations and surveillances performed during the
past 18 months were reviewed. The records reviewed were for eight
procedures associated with the meteorological monitoring system. The
subject surveillances were performed correctly and within schedule, and
instrumentation calibrations were completed as required.
c.
Conclusions
The licensee was maintaining the meteorological equipment appropriately,
and implementing the meteorological monitoring program in accordance
with established procedures and FSAR commitments.
R2.2 Radiation Monitor System Installation and Calibration
a.
Inspection Scope (84750)
The inspectors reviewed and evaluated the adequacy of installed process
and effluent Radiation Monitoring System (RMS) detectors, particulate
and iodine samplers, electronics, sampling lines and flow meters, as
applicable, to meet FSAR commitments and to implement Offsite Dose
Calculation Manual (ODCM) and 10 CFR Part 20 requirements. The
evaluation included, as applicable. RMS equipment walk-downs with
comparisons against configuration control documents, design changes and
vendor design specifications, as appropriate.
Further, the installed
sample line bend radii and piping specifications were evaluated against
recommendations detailed in American National Standards Institute (ANSI)
N13.1-1969, American National Standard Guide to Sampling Airborne
Radioactive Materials in Nuclear Facilities. General comparisons were
made between radiation monitor local and remote readout data, where
possible.
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Approved guidance and resultant data for selected RMS detector
calibrations were reviewed and discussed.
For each detector reviewed,
source calibration packages were reviewed, evaluated and discussed with
licensee representatives. The following RMS detectors and associated
electronics were included in the review: Unit 1 Waste Liquid Effluent
Monitor (RE-18): Steam Generator Sample Liquid Process Monitor (RE-19),
Steam Generator Sample Liquid Process Monitor (RE 21) Plant Vent Wide
Range Monitor (1RE12444C); Containment High Range Monitor (RE-0005).
The RMS source calibration guidance and results were evaluated against
applicable sections of the FSAR, Technical Specification (TS) and ODCM
requirements.
In addition, guidance for the containment high range
monitor was compared against special calibration requirements s)ecified
in NUREG 0737, Clarification of Three Mile Island (TMI) Action
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Requirements, Table II.F.13 Containment High Range Monitors (CHRMs).
b.
Observations and Findinas
For the RMS equipment reviewed, no significant issues regarding design
specifications, installed system equipment and sample line
ex, figurations, and operating parameters were identified. Housekeeping
practices associated with RMS equipment skids, cabinets and general
areas were appropriate.
From direct observation of RMS equipment and discussiens with
responsible personnel, operability of selected monitors was verified.
Comparison of selected monitor remote and local readouts did not
identify any significant discrepancies.
Sample flow rates were within
limits specified within vendor manuals.
In addition, the inspectors
verified that corrective actions were in progress regarding degraded
heat trace conditions identified in Safety Audit and Engineering Review
(SAER) audit No. OP05-97/03 dated February 7, 1997.
Excluding the containment high range monitor, no calibration concerns
were identified.
Surveillances were conducted at the required
frequencies and the reported results were acceptable.
For the
containment high range monitor calibration data, the inspectors noted
that the source strength, approximately 17 Roentgens per hour (R/hr)
used to conduct the in situ calibration exceeded the 1 10 R/hr range
specified in NUREG 0737, Table II.F.13.
Initial review of FSAR
commitments and discussions with licensee representatives indicated that
no exception was taken from meeting the requirements of NUREG 0737,
Table II.F.1-3, nor was an analysis of changes from the FSAR commitments
available. Review of vendor calculations and documents indicated that
the stronger source strength was selected to minimize interference from
system noise. The inspectors noted that additional NRC review of vendor
calculations and data for the installed containment high range monitors
to meet sensitivity requirements would be conducted.
c.
Conclusions
The RMS equipment was designed, installed, and operated appropriately.
Maintenance issues regarding heat tracing degradation identified during
a recent SAER audit were being tracked and corrected by the licensee.
The adequacy of the containment high range monitor to meet FSAR
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requirements based on the calibration source concerns was identified as
an unresolved item (URI) 50 424, 425/97 02 03:
NRC review and evaluate
17 R/hr source strength and installed containment high range monitor to
meet NUREG 0737 sensitivity and calibration requirements.
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R2.3 Enaineered Safety Feature (ESF) Ventilation System
a.
Insoection Scope (84750)
The inspectors verified implementation of Engineered Safety Feature
(ESF) ventilation systems filter testing surveillances recuired by TS
3.7 in accordance with TS 5.5.11 test requirements.
In acdition, the
material condition of selected ESF filter ventilation systems was
observed during system walk downs. Equipment walk downs were conducted
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and the most recent surveillance results reviewed for the following ESF
ventilation systems.
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Unit 1. Piping Penetration Area Filtration and Exhaust
Train A
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Unit 1 Control Room Emergency Filtration Room
Trains A & B
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Unit 2 Control Room Emergency Filtration Room
Trains A & B
b.
Observations and Findinas
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The material condition of the ESF equipment and cleanliness of
associated areas was adequate.
From reviews of maintenance work orders,
the inspectors verified that selected equipment issues noted during
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licensee walk downs or routine surveillances were identified, tracked
and completed in a timely manner.
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From review of licensee records, the inspectors verified that ESF 18-
month surveillances were conducted at the specified frequency. Test
results for the high efficiency particulate air (HEPA) filter, in place
charcoal adsorber and laboratory analysis of charcoal adsorber material
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met established TS acceptance criteria.
c.
Conclusions
The ESF ventilation systems were maintained appropriately and tested in
accordance with TS requirements. Test results for selected ESF systems
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were acceptable.
R5
Training and Qualifications in Radiation Protection and Transportation
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R5.1 Gen (ral Employee Trainina
a.
Inspe: tion Scope (83750)
The inspectors reviewed the licensee's program for providing General
Employee Training (GET), also known as Badge Training, for personnel
permanently or temporarily employed at the Vogtle facility.
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b.
Observations and Findinas
The licensee's program for GET was addressed in plant procedure 00700 C,
General Employee Training, Rev. 18, approved January 17, 1997. To
obtain unescorted access and dosimetry for the Protected Area and
selected Vital Areas, employees were required to receive formal initial
training and annual retraining in plant overview, emergency
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preparedness, fire protection, industrial safety, quality assurance,
security, and radiation protection, and to pass a written examination
covering those various areas.
Approximately 55 percent of the 110-page
initial GET handbook provided to employees was devoted to radiation
protection matters. A different handbook (73 pages) was provided for
employee study as part of annual retraining, with approximately
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50 percent of its content dedicated to the subject of radiation
protection.
Prior to Revision 17 of procedure 00700 C in October 1995,
each em)loyee was required (on a triennial basis) to read the GET
handboo(, and to so certify in writing, before attending the retraining.
Employees were encouraged and expected (but not required) to review the
GET retraining handbook before attending the annual training.
The inspectors discussed the GET program with cognizant licensee
mannement, and reviewed selected lessons plans, particularly with
regard to the area of Radiation Protection (RP). The inspectors
scrutinized training records for a total of 12 employees selected from
Operations, HP/ Chemistry, and Security. No discrepancies were
identified in the GET records of these licensee personnel for the period
1987-1996.
c.
Conclusions
The licensee was effectively administering its GET program in accordance
with established commitments and procedures.
R5.2 Hazardous Material Training
a.
Insoection Scooe (86750. TI 2515/133)
The training provided to meet the recuirements of 49 CFR Part 172
Subpart H were reviewed and discussec with licensee representatives.
Further, training details provided to staff regarding implementation of
recent Department of Transportation (DOT) changes to 49 CFR Parts 100-
179 were evaluated.
From discussion with responsible staff members, the inspector evaluated
the training effectiveness regarding recent D0T changes implemented for
49 CFR Parts 100 179.
b.
Observations and Findinos
From review of training records, the inspectors verified that staff
members involved in handling and packaging of radioactive materials were
receiving hazardous material (hazmat) training at the required
frequencies.
From review of training material presented to staff in
November 1996, the inspectors verified that recent DOT changes to
shipping and packaging requirements were provided to responsible
,
.
i
.
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10
)
'
personnel. Applicable hazmat training was provided to HP staff in June
1995 and in December 1996.
The most recent training was conducted as
part of the corrective actions for the issue involving improper
packaging and transportation of surface contaminated object (SCO)
material detailed in Section R1.2. Additional training regarding the
i
new DOT regulations is scheduled.
From discussion of shipping
procedures and records, the inspectors determined that resaonsible
licensee representatives were knowledgeable of the recent X)T changes.
j
c.
Conclusions
Hazmat training provided to personnel handling radioactive materials was
conducted at the ap)ropriate frequency, and included recent changes to
DOT regulations.
T1e training provided wa.s effective.
R7
Quality Assurance (QA) in Radiation Protection and Chemistry Activities
R7.1 Radioloaical Measurement Quality Control
a.
Insoection Scope (84750)
]
The inspectors reviewed implementation of the counting room cuality
control (QC) activities to meet the intent of Regulatory Guice (RG)
4.15, Quality Assurance for Radiological Monitoring Programs (Normal
l
Operations) - Effluent Streams and the Environment.
Specifically, the
l
results of the 1995 and 1996 inter laboratory cross check radiological
analyses were reviewed and discussed with cognizant licensee
i
representatives. Also, the inspectors reviewed and discussed composite
sample preservation to maintain sample representativeness.
b.
Observations and Findinas
All individual inter laboratory analyses were within the established
acceptance criteria.
No regulatory concerns nor negative trends were
identified from review of the counting room tritium and gamma-
spectroscopy QC performance data.
Licensee methods for preservation of
composite samples were appropriate.
c.
Conclusions
.
'
Sample ) reservation, and gamma spectroscopy and tritium inter laboratory
cross cleck QC activities were implemented appropriately and met the
intent of RG 4.15.
R7.2 Licensee Self Assessment Activities (84750. 86750)
a.
Inspection Scope (84750. 867FM
During the inspection period, the following Safety Audit and Engineering
Review (SAER) audit repor;s and associated checklists were reviewed and
discussed with cognizant licensee representatives. Specific radioactive
_.._ _
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_._ _ ...___ _ _. _
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11
waste, transportation. effluent monitoring and chemistry, Radiation
Control (RC): and radioactive waste (radwaste) processing, packaging and
trans>ortation program activities required by TS, 10 CFR Part 20, and
,
10 CFR Part 71 were reviewed and discussed with licensee
'
'
representatives.
i
.o
SAER Audit of Radioactive Waste Control
OP05 97/03, dated
February 7, 1997,
- .
e
SAER Audit of Radioactive Waste Control
OP05 96/17, dated
i
July 1,1996.
e
SAER Audit of Plant Chemistry
OP04 96/28, dated September 12,
1996.
!
e
SAER Audit of Plant Chemistry
OP04 95/25, dated November 20,
1995.
e
SAER Audit of Radioactive Waste - OP05 95/16, dated October 23,.
1995,
b.
Observations and Findinas
^
The audits met recuired frequencies and addressed 00CM, effluent,
.
Chemistry, RC, racwaste and trans>ortation program guidance and
!.
implementation.
Both compliance-)ased and performance based audit
' techniques were used to identify documented strengths, issues,
"
weaknesses and recommendations. The inspectors verified from review of
i
audit checklists and discussions with responsible personnel that the
i
'
audits included review and followup of previously identified items.
I
From review of audit team participants and discussions with licensee
i
management, the inspectors determined that audit teams included
j
experienced individuals from outside of the Vogtle facility.
.
c.
Conclusions
Audits of the radioactive waste, effluents and transprtation program
i
activities were thorough and comprehensive, and met TS, 10 CFR Part 20,
and 10 CFR Part 71 requirements.
,
k
.
I
,
,
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.
.
.
,
12
,
VI. Manaaement Meetinas
X1
Exit Meeting Sumary
The inspectors presented the inspection results to members of the licensee
i
management on February 14, 1997. The licensee acknowledged the findings
j
presented.
The inspectors noted that no proprietary information would be contained in the
.
report.
PARTIAL LIST OF PERSONS CONTACTED
Licensee
.
B. Beasley, General Manager
R. Brown, Training and Emergency Preparedness Manager
R. Carter, Su mrvisor Safety Audit and Engineering Review
S. Chestnut, ianager, Operations
1
J. Gasser, Assistant General Manager Operations
a
K. Holmes, Maintenance Manager
I. Kochery, Health Physics Superintendent
M. Kurtzman, Supervisor, Health Physics and Chemistry Training
A. Parton. Chemistry Superintendent
M. Sheibani, Supervisor, Nuclear Safety and Compliance
C. Tippins, Jr., Nuclear Specialist
INSPECTION PROCEDURES USED
IP 83750:
Occupational Radiation Exposure
IP 84750:
Radioactive Waste Treatment, and Effluent and Environmental
Monitoring
IP 86750:
Solid Radioactive Waste Manags ,t and Transportation of
Radioactive Materials
TI 2515/133:
Implementation of Revised 49 CFR Parts 100-170 and 10 CFR Part 71
ITEMS OPENED, CLOSED, AND DISCUSSED
i
Ooened
!
50 424, 425/97 02 01
Failure to follow radiation protection
procedures for a DAW processing facility dose
note limits (Section R1.1) and for a liquid
"
waste processing gamma spectroscopy analyses
(Section R1.3)
50 424, 425/97 02-02
Failure to meet 49 CFR 173.475 package dose rate
l
limits (Section R1.2).
<
._
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._
._ _ . ._
. _ . . .
. . _ . - _ . .
._
_ . _ . _ _ _ _ _ . _ _ _ _ _
.
.
<
,
,
1
13
,
50 424, 425/97 02 03
NRC review'and evaluate 17 R/hr source strength
and installed containment high range monitor to
meet NUREG 0737 sensitivity and calibration
,
i
requirements (Section R2.2).
Closed
50-424, 425/97-02 01
Failure to follow radiition protection
procedures for a DAW processing facility dose
rate limits (Section R1.1) and for a liquid
waste processing gamma spectroscopy analyses
.
(Section R1.3)
l-
1
f
i
n
4
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4
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14
LIST OF ACRONYMS USED
.
ANSI
American National Standards Institute
Dry Active Waste
Department of Transportation
1
Engineered Safety Feature
Final Safety Analysis Report
Hazmat
Hazardous Material
Health Physics
MDCC
Meteorological Data Collection Center
mrem /hr
millirem per hour
mSv/hr
millisieverts M r hour
'
Non cited Violation
Offsite Dose Calculation Manual
1
Quality Assurance
j
Quality Control
'
R/hr
Roentgens per hour
radwaste
Radioactive Waste
i
Radiologically Controlled Area
]
Regulatory Guide
i
Radiological Incident Report
Radiation Monitoring System
i
Radiation Protection
Reportable Quantities
Radiation Work Permit
RWS
Radioactive Waste Shipment
-SAER
Safety Audit and Engineering Review
Three Mile Island
i
TS
Technical Specification
]
Unresolved Item
Very High Radiation Area
Violation
.