IR 05000424/1987064
| ML20237A304 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/03/1987 |
| From: | Adamovitz S, Gloersen W, Kahle J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20237A273 | List: |
| References | |
| 50-424-87-64, NUDOCS 8712140392 | |
| Download: ML20237A304 (15) | |
Text
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UNITED STATES pRtop
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gD NUCLEAR REGULATORY COMMISSION
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REGION li h ()
101 MARIETTA STREET, N.W.
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C ATLANTA, GEORGIA 30323
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DEC 0 81987
Report No.:
50-424/87-64 Licensee:
Georgia Power Company
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l P. O. Box 4545 Atlanta, GA 30302
{
i Docket No.:
50-424 License No. :
NPF-68 l
I Facility Name:
Vogtle
Inspection Conducted:
November 16-20, 1987
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/#N/57 Inspector: [//.
s f B. Gloe'rse '
Cat:e Signed
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'Date Signed Accompanying Personnel:
C. A. Hughey Approved by:
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/ L/3/77
'. B/ Kahle, Section Chief Date Signed J
D sion of Radiation Safety and Safeguards SUMMARY-Scope:
This routine, announced inspection was conducted in the areas of quality assurance and confirmatory measurements for in plant radiochemical analysis and previously identified inspector followup items.
Results:
One violation was identified - failure to perform surveys for in plant noble gas concentration calculations.
8712140392 871208 PDR ADOCK 05000424 G
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- R. M. Bellamy, Plant Manager
- A. L. Mosbaugh, Assistant Plant Support Manager
- W. F. Kitchens, Operations Manager
- T. Greene, Plant Support Manager
- J. E. Swartzwelder, Manager-NSAC
- G. R. Frederick, QA Site Manager-Operations
- D. F. Hallman, Chemistry Superintendent
- A. E. Desrosiers, Health Physics Superintendent
- I. Kochery, Health Physics Superintendent ( Acting)
- R. M. Odom, Engineering Supervisor
- C. L. Cross, Senior Regulatory Specialist
- J. R. Petro, Senior Quality Assurance Field Representative
- R. Hand, Plant Chemist S. Ewald, Manager, Health Physics and Chemistry A. Stalker, Health Physicist (Corporate)
S. Sundaram, Senior Plant Chemist
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J. B. Sills, Lab Supervisor l
J. A. Carswell, Health Physics Foreman J. L. Willcox, Senior Quality Assurance Field Representative S. McCann, Consultant R. Cislo, Consultant Other licensee employees contacted included engineers, technicians, and office personnel.
NRC Resident Inspectors
- H. L. Livermore
- J.
F. Rogge
- C. W. Burger
- Attended exit interview 2.
Exit Interview
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l The inspection scope and findings were summarized on November 20, 1987, l.
with those persons indicated in Paragraph 1 above.
One violation (Paragraph 4) concerning the failure to perform adequate surveys for airborne radioactive material in areas within the plant was discussed.
The inspectors also discussed areas for improvement in the Health Physics and Chemistry laboratory quality assurance p r.og ram s.
No dissenting comments were received from the licensee.
The licensee did not identify as proprietary any of the material provided to or reviewed by the inspector during this inspection.
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3.
Quality Assurance Program (84725)
a.
Radiochemistry Laboratory The inspectors toured the radiological sample preparation and counting laboratories to examine the equipment that was used for radiological sample counting activities. Analytical equipment in the radiochemistry count room included one Nuclear Data Series 76 stand-alone multichannel analyzer, one Nuclear Data Series 76 terminal and four intrinsic germanium detectors with two detectors having nominal efficiencies of 10?s, a third detector with a nominal efficiency of 20?; and a fourth detector with an efficiency of approximately Sis.
Date processing and management for the gamma spectroscopy system was accomplished by a Microvax 11 central processing unit. This system also had a backup Microvax. Additional analytical equipment included a Beckman LS3801 liquid scintillation counter.
The inspectors reviewed the licensee's quality assurance program for the radiochemical counting laboratory.
The guidance contained in Regulatory Guide 4.15 (Quality Assurance for Radiological Monitoring Program (Normal Operations) - Effluent Streams and the Environment, February 1979) was used to evaluate the licensee's program.
The inspectors noted an overall improvement in the area of quality assurance since the last inspection (50-424/86-119). At the time of this inspection, the Chemistry and Health Physics groups had undergone a reorganization so that the two groups had formed one Department. The Chemistry and Health Physics Department was divided into three sections:
Health Physics, Chemistry, and Support Services.
The inspectors were informed that approximately ten chemistry specialists were qualified to operate the counting equipment.
The inspectors verified that written procedures were reviewed and approved for activities involved in in plant radiochemical analyses according to administrative control instructions.
The inspectors reviewed selected procedures for sample collection, sample logging, sample preparation and analysis and operation and calibration of radiological analytical eq ui pme nt.
The inspectors noted that the licensee had incorporated some of the comments made during inspection 50-424/86-119 into Procedures 3300-C, Preparation of Liquid Samples for Radiochemical Analysis, Rev. 4, October 20, 1987, and 33015-C, Obtaining Ventilation and Gaseous Samples for Radioactivity Analysis, Rev. 4, October 26, 1987.
Laboratory quality control included the use of NBS traceable reference standards to determine counting efficiencies for specific radionuclides and to determine the counting efficiency as a function of gamma ray energy for the gamma spectroscopy systems.
The inspectors observed that efficiency determinations for all four detectors were made for a variety of geometries at various shelf l
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heights.
The following efficiency data records were reviewed l
(calibrations were performed during November 1986):
l Detector 1 (SN 1731) - 10cc gas vial shelf 1; 1 liter bottle; 1 liter liquid Marinelli; 4 liter gas marinelli; 4 liter liquid Marinelli; charcoal cartridge shelves 0,1, 2, 3; particulate filter shelves 0, 1, 2, 3; reactor coolant and scintillation vial shelves 0, 1, 2, 3; 125cc gas container shelf 0.
Detector 2 (SN 1759) - 1 liter bottle; 1 liter liquid Marinelli; charcoal cartridge shelf 0; particulate filter shelf 0; reactor coolant scintillation vial shelf 0.
Detectors 3 (SN 37) and 4 (SN 1755) - 10cc gas vial shelves 1 and 2; 1 liter oottle; 1 liter liquid Marinelli; 4 liter gas Marinelli; charcoal cartridge shelves 0,1, 2, 3; particulate filter shelves 0,1, 2, 3; reactor coolant scintillation vial shelves 0, 1, 2, 3; 125cc gas container shelves 0, 1, 2.
The inspectors reviewed the 1987 daily performance checks for the gamma spectroscopy systems.
The licensee graphically tracked the detector efficiencies and resolutions for three radionuclides (Co-57, Co-60, and Cs-137) on all four detectors.
The baseline and the t 2 sigma, 3 sigma control limits were determined at the time the detectors were calibrated.
It was noted, however, that the daily QA/QC data was plotted on control charts with time durations of one month.
The inspectors discussed with the licensee the advantages of plotting the daily QA/QC data on control charts with time spans covering at least six months.
The licensee monitored daily performance checks to ensure that the measured values were within specified control limits.
Corrective actions and measurements outside the specified control limits were documented in equipment logbooks.
Although, the inspectors observed that copies of the calibration source certificates should be kept in the laboratory, all other records were generally well organized and easily accessible.
Last year, the licensee had participated in a cross-check program with a vendor who supplied the licensee with unknown spikes periodically.
During this inspection, it was apparent that the cross-check program was not given as much attention. The inspectors discussed ways to improve the cross-check program by participating, perhaps on a quarterly basis but only using one or two different geometries per quarter.
The licensee indicated that plans were developed to have a cross-check program with Plant Hatch, b.
Health Physics Laboratory The inspectors toured the Health Physics counting laboratory to examine the equipment that was used mainly for in plant radiological sample counting activities.
The inspectors were informed that the Health Physics gamma spectroscopy system could be used as a backup to the Chemistry laboratory in the event that the Chemistry lab would be
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l-inaccessible due to radiological concerns. Analytical equipment in
the counting room included two intrinsic germanium detectors and Nuclear Data Series 76 processing equipment, two NMC proportional counters (Model PCC 11T) which were not operational, one Eberline SAC-4 for alpha counting, and an Eberline HP-210 shielded pancake probe.
The inspectors reviewed the licensee's quality assurance program using the guidance contained in Regulatory Guide 4.15.
The inspectors and licensee representatives discussed 'various areas for improvement of the quality assurance program and the following list summarizes the discussion:
Assignment of a lab manager or supervisor to the count room to monitor the day-to-day operations.
Consideration of the participation in a cross-check program for gamma spectroscopy measurements.
Participation of lab technicians in a gamma spectroscopy classroom training program.
Trending of daily QA/QC data on graphs.
Establishment of control limits (
2, 3 sigma) associated with gamma spectroscopy system performance at each calibration cycle (typically annually).
Daily tracking of the resolution of at least two well-defined photopeaks in addition to tracking photopeak activities.
Keeping copies of QA/QC trend data for the current calibration cycle in the lab, preferably in notebooks.
Use serial numbers for identification of the two gamma detectors.
Keep copies of source certificates in the lab.
Establish a communication link between the Health Physics group and the Chemistry group, especially for technical assistance.
The licensee agreed to consider the items listed above.
The inspectors also reviewed selected Health Physics related counting room procedures.
It was noted that in 43813-C, Reliability Checks for Nuclear Data Gamma Spectroscopy System, Rev. 6, April 30, 1987, Step 4.3.5.2 for evaluating the energy difference between the known and measured values, the licensee should consider using a tolerance
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5 of 0.5 kev instead of i 1 kev. Additionally, the check source that is used for the daily QA/QC checks should be changed at each calibration cycle.
No violations or deviations were identified.
4.
Confirmatory Measurements (84725)
l During this inspection, samples of reactor coolant, containment l
atmosphere, and liquid and gaseous waste monitor tanks were collected and i
the resultant sample matrices were analyzed for radionuclides l
concentrations using the gamma-ray spectroscopy systems of the licensee's l
counting laboratories and the NRC Region II mobile laboratory.
The l
purpose of these comparative measurements was 'o verify the licensee's
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capability to measure quantities of radionuclides accurately in the various plant systems. Analyses were conducted using the licensee's six gamma spectroscopy systems (four detectors located in the chemistry count room and two detectors located in the health physics count room).
Sample types and counting geometries included:
(1) containment atmosphere -
charcoal cartridge; (2) reactor coolant system (RCS) dissolved gas - 10 cc l
l vial; (3) degassed reactor coolant - 20 cc vial; (4) liquid waste monitor 125 cc gas tank - one liter Marinelli; (5) waste gas decay tank
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I container and one liter gaseous Marinelli. A spiked charcoal cartridge and particulate filter were provided for analysis in addition to the licensee samples.
A comparison of licensee and NRC results is listed in Attachment 1, Table 1 with the NRC acceptance criteria listed in Attachment 2.
For the spiked particulate filter, Americium-241 was not detected by the Health Physics Detectors 1 (PGT 1851) and 2 (PGT 1768).
Discussions with count room personnel indicated that these two detectors were not calibrated for energies below 80 kev and therefore could not identify the 59.5 kev gamma for Am-241.
Initial analysis of the containment atmosphere charcoal cartridge by the four Chemistry detectors did not identify the isotope Br-82.
Licensee personnel indicated this isotope was not in the computer's isctopic identification, library.
The library was modified to include the isotope and subsequent re-analysis of the spectrums identified Br-82.
A dissolved gas sample from the reactor coolant was analyzed by three detectors in the Chemistry count room.
Detectors 1 (PGT 1731) and 3 (PGT 1737) showed agreement for all isotopes.
However, Detector 4 (PGT 1755) did not identify Ar-41 or Kr-87.
It was noted that Detector 4 (PGT 1755) was the least efficient of the four Chemistry detectors (5?;
efficiency as compared to 10'. to 20?J for the other detectors) but also that this detector was the primary system for counting reactor coolant liquid and gaseous samples.
The inspectors discussed with count room personnel the advantages of extending sample count times or utilizing a more efficient detector for analyzing RCS dissolved gas.
A degassed reactor coolant sample was counted on the four Chemistry detectors and i
showed agreement for Detectors 1 (pGT 1731) and 4 (PGT 1755). The isotope
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I-132 was not detected on Detector 3 (PGT 1737) and I-132 activities were in disagreement for Detector 2 (PGT 1759).
However, as noted above, Detector 4 (PGT 1755) was the primary system for analyzing reactor coolant
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samples and all isotopic values were in agreement for Detector 4.
A liquid waste monitor tank sample was analyzed by the Chemistry and Health Physics count rooms.
Results from the Health Physics Detectors 1 (PGT 1851) and 2 (PGT 1768) were biased low and generally in disagreement with values ranging from 36?; to 61?; below the NRC results. A waste gas decay tank was also sampled and counted using two different geometries; i.e., a 125 cc gas container and a gaseous Marinelli.
For the gaseous Marinelli geometry results from the Health Physics Detectors 1 and 2 were biased high (33?J to 55?4 above NRC values) and in disagreement for Xe-133 and Xe-135.
Discussions with Health Physics count room personnel indicated an erroneous volume of 1100 cc had been deter.nined for the gaseous Marinelli when the actual volume was approximately 1250 cc.
However, when factoring out the error due to using the wrong container volume, the licensee results would still be biased high by 17?; to 37?; and in disagreement for Xe-133.
The incorrect volume had been used for approximately one year to determine noble gas concentrations in restricted areas within the plant.
The inspectors indicated to licensee representatives that the use of the incorrect volume for the gaseous Marinelli was an apparent violation of 10 CFR 20.201(b)
and 10 CFR 20.103(a)(3) (50-424/87-64-01).
10 CFR 20.201(b) requires each licensee to make or cause to be made such surveys as (1) are necessary for the licensee to comply with the regulations in this part and (2) are reasonable under the circumstances to evaluate the extent of radiation hazards that may be present.
10 CFR 20.103(a)(3) requires for the purposes of compliance with the requirements of this section, that the licensee use measurements of concentrations of radioactive materials in air for determining and evaluating airborne radioactivity in restricted areas.
One violation was identified - failure to make adequate surveys.
5.
Process and Effluent Radiation Monitoring and Sampling System (84723)
The inspectors and licensee representatives discussed the spurious alarm problems with the Process and Effluent Radiation Monitoring and Sampling System (PERMSS). Chemistry Department personnel indicated that high level management attention had been given to resolve the apparent computer problems with the system vendor.
The inspectors discussed the need to strive for an early resolution and indicated that this area will be reviewed during subsequent inspections.
No violations or deviations were identified.
6.
Licensee Action on Previously Identified Inspector Followup Items (92701)
(Closed)
50-424/86-119-05:
Review gamma-ray spectroscopy system calibrations for various geometries at different shelf heights and efficiency determinations for gas geometries greater than 500 kev.
The inspectors reviewed the licensee's efficiency determination records which were completed in November-December 1986. The various geometries used for the four intrinsic germanium detectors are listed in Paragraph 3.
The
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i inspectors noted that the licensee had purchased gas calibration (mixed gamma) sources in which each geometry had been filled with styrofo'am spheres. These sources were going to be used during the next calibration cycle. During-the last calibration cycle, the licensee prepared the gas geometries using a material that approximated the density of air.
This item is considered closed.
(Closed) 50-424/86-119-06: Review results of the spiked simulated liquid radwaste sample containing Fe-55, H-3, Sr-89, and Sr-90. The comparison
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between the licensee and known results can be seen in Attachment 1, Table 2.
The licensee was in disagreement for Fe-55.
This disagreement was discussed with licensee representatives.
The licensee agreed to address this matter.
This item is considered closed.
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ATTACHMENT 2 Criteria for Comparing Analytical Measurements l
l This attachment provides criteria for comparing results of capability tests and verification measurements. The criteria are based on an empirical relationship which combines prior experience and the accuracy needs of this program, i
In this criteria, the judgement limits denoting agreement or disagreement between licensee and NRC results are variable.
This variability is a function of the NRC's value relative to its associated uncertainty. As the ratio of the NRC value to its associated uncertainty referred to in this program as " Resolution"2 increases, the range of acceptable differences between the NRC and licensee values should be more restrictive. Conversely, poorer agreement between NRC and licensee values must be considered acceptable as the resolution decreases.
For comparison purposes, a ratio 2 of the licensee value to the NRC value is computed.
This ratio is then evaluated for agreement based on the calculated resolution.
The corresponding resolution and calculated ratios which denote agr.eement are listed in Table 1 below.
Values outside of the agreement ratios are considered in disagreement.
2 Resolution =
NRC Reference Value Associated Uncerainty for the Value
' Comparison Ratio =
Licensee Value I
NRC Reference Value TABLE 1 Confirmatory Measurements Acceptance Criteria Resolutions vs. Comparison Ratio Comparison Ratio for Resolution Agreement
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