IR 05000261/1990018
| ML14176A906 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 09/06/1990 |
| From: | Decker T, David Jones, Seymour D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14176A905 | List: |
| References | |
| 50-261-90-18, NUDOCS 9009240287 | |
| Download: ML14176A906 (18) | |
Text
01-t, REG "lUNITED STATES o
NUCLEAR REGULATORY COMMISSION
REGION II
<101 MARIETTA STREET, ATLANTA, GEORGIA 30323 SEP 10 33O Report No.:
50-261/90-18 Licensee:
Carolina Power and Light Company P. 0. Box 1551 Raleigh, NC 27602 Docket No.:
50-261 License No.:
DPR-23 Facility Name: H. B. Robinson Inspection Conducted: A ust 13-17, 1990 Inspectors:
--
iz D. A. Seymour/
Date gned D. W. Jones Date Sioned Approved by:
/
/
T. R. Decker, Chief Date Siqned Radiological Effluents and Chemistry Section Emergency Preparedness and Radiological Protection Branch Division of Radiation Safety and Safeguards SUMMARY Scope:
This routine, unannounced inspection was conducted in the areas of radiological effluents, plant chemistry, confirmatory measurements, the Post Accident Sampling System (PASS), and radiation monitorin Results:
A review of the chemistry counting room quality control indicated that this program was adequate to ensure accurate and reliable analytical results. This was supported by good agreement between the licensee's chemistry and health physics count rooms and the NRC mobile laboratory gamma spectroscopic measurement There were no significant radiological consequences attributable to the operation of Robinson in 1989 noted from airborne, waterborne, aquatic, ingestion or direct exposure pathways (Paragraph 4).
Primary chemistry parameters of Dose-Equivalent Iodine-131, chlorides and oxygen were all maintained well below Technical Specification limit There were no indications of fuel leakers during this fuel cycl PDR ADOCK 05000261 G
PNU
The PASS had exhibited adequate operability and had achieved good comparisons between normal reactor coolant sample analyses and PASS sample analyse Robinson had developed a plant modification pac.kage for the upgrade of their radiation monitors during the upcoming outage (Fall 1990).
One Non-Cited Violation was identified involving improper contamination control practices by an individual using a fume hood. This was considered an isolated incident with adequate licensee corrective actions taken prior to the close of this inspectio REPORT DETAILS 1. Persons Contacted Licensee Employees
- S. Billings, Technical Aid, Regulatory Compliance
- B. Christensen, Foreman, Environmental and Radiation Control
- Corley, Manager, Environmental and Radiation Control
- R. Crook, Senior Specialist, Regulatory Compliance
- Eaddy, Jr., Support Supervisor, Environmental and Radiation Control
- E. Harris, Jr., Manager, Onsite Nuclear Safety
- J. Hill, Technician I, Environmental and Radiation Control
- J. Kloosterman, Director, Regulatory Compliance
- R. Morgan, Plant General Manager
- R. Smith, Manager, Maintenance
- D. Stadler, Onsite Licensing Engineering and Nuclear Licensing
- H. Young, Manager, Quality Assurance and Quality Control Other licensee employees contacted during this inspection included craftsmen, engineers, operators, mechanics, security force members, technicians, and administrative personne Nuclear Regulatory Commission
- K. Jury, Resident Inspector
- Attended exit interview
- Attended entrance and exit interview
- Attended entrance interview 2. Confirmatory Measurements (84750)
Pursuant to 10 CFR 20.201(b)
this area was inspected to verify the licensee's ability to conduct precise and accurate measurement During this inspection, samples of reactor coolant and selected liquid and gaseous process streams were collected and the resultant sample matrices were analyzed for radionuclide concentrations using-the licensee's counting laboratory and the NRC Region II mobile laboratory gamma spectroscopy system., The purpose of these comparative measurements was to verify the licensee's capability to accurately measure quantities of radionuclides in various plant systems. Analyses were conducted using the licensee's three intrinsic germanium gamma spectroscopy systems in the
count room; and for one sample, the health physics intrinsic germanium gamma spectroscopy system. Sample types and counting geometries included the following:
reactor coolant, 50-milliliter bottle; liquid waste, one liter marinelli; containment atmosphere, 1260-milliliter marinelli; an NRC spiked charcoal cartridge, and a particulate filter generated by filteration of 500 milliliters of reactor coolan Comparison of licensee and NRC results are listed in Attachment 1, Table 1 with the acceptance criteria listed in Attachment 2. The results were in agreement for all sample types analyze The inspectors observed the licensee obtaining the reactor coolant, liquid waste, and containment atmosphere samples. Proper sampling techniques and health physics practices were observed. The inspector reviewed selected portions of Procedure No. EMP-025, titled Gaseous Effluent Sampling and Analysis Requirement The portions reviewed were adequate for the intended purpos As of January 1990, the licensee began participation in an extensive split gamma spectroscopic, tritium, charcoal cartridge, gross alpha, and gross beta analyses program with an outside vendor. The inspectors reviewed the results of this cross-check program for the first and second quarters of 1990. The licensee and the vendor were in agreement for all isotope Prior to 1990, the license participated in a semi-annual cross-check program for gamma spectroscopic analysis of mixed gamma emitting liquid samples. The inspectors reviewed the results from this program for the year 1989, and found that the licensee was in agreement with the outside vendor for all isotope. Counting Room (84750)
a. The inspectors toured the radiochemistry counting laborator Instrumentation in the laboratory included the following:
(1) Two Nuclear Data computer based gamma spectroscopy counting systems with three operational Intrinsic Germanium detector (2) Two Tennelec LB1000 Proportional Counters used for gross alpha and gross beta determination (3) Packard 460 Liquid Scintillation Counting system used for tritium (H-3) determinatio b. The licensee's quality assurance (QA) program for the above equipment was reviewed to ensure compliance with selected and applicable portions of Regulatory Guide 4.15, Quality Assurance for
3 Radiochemical Monitoring Programs (normal operations)
- Effluent Streams and the Environment, Revision 1, dated February 197 The following observations were made:
(1) Daily energy calibration checks, and system resolution checks were performed on the gamma spectroscopy system (3 detectors)
using a mixed gamma ray sourc The values obtained from the calibration checks were recorded and trended on control charts with specified predetermined limits in order to confirm detector stabilities and system operabilit The control chart limits were reevaluated at least annually for each detector. A review of the control charts showed all values to be within the established limits, indicating stable detector performance. The periods reviewed varied for each detector beginning on March 6, 1990, April 4, 1990, and June 1, 1990, and ending on August 15, 199 The licensee indicated that daily background determinations were performed for each detecto (2) Daily background and response checks were performed on the LB1000 proportional counter A review of the control charts for these instruments over the period May - August 1990, revealed that the performance of the detectors had been stabl (3) The daily response checks performed during the period May 5 through August 15, 1990, for the Packard 460 liquid scintillation counter, were verified by the inspector to be within predetermined control limits, indicating stable instrument performanc The licensee indicated that daily background checks were also performed on this instrumen. Reactor Coolant Chemistry (84750)
Technical Specification (TS) 3.1.4 specifies the reactor coolant specific activity limits and TS 3.1.6 specifies the reactor coolant oxygen and chloride concentration limit Table 4.1-2 of TS 4.1 specifies -the sampling frequencies for these parameter These parameters are related to corrosion resistance and fuel integrit Pursuant to these requirements, the inspector reviewed TS Surveillance Results forms for the period August 1-14, 199 The results for the following parameters were examined: dose equivalent 1-131, 100/E (E =
average beta and gamma energy per disintegration),
ppb oxygen and ppb chlorid The reactor coolant specific activity and the oxygen and chloride concentrations were within their specified limit During the inspection, the licensee also produced graphs of dose equivalent 1-131 (DEI), and ppb chloride for the period January 1, 1990 through August 14, 1990. The average DEI for that period was 6.62 E-4 microcuries per gram and the maximum value observed was 1.1 E-3 microcuries per gra The specified limit was 1 microcurie per gra The graph for chloride concentration indicated that this parameter had remained at or below 20 ppb for the entire period. The specified limit was 150 ppb chlorid The licensee indicated that the oxygen concentration for that period remained below 1 ppb and the specified limit was 100 pp The inspectors determined that these parameters were maintained well below TS limit No violations or deviations were identifie. Radiation Monitoring System (RMS) (84750)
TSs 3.5.2 and 3.5.3 define the operating requirements for the radioactive liquid and gaseous effluent instrumentation system These instruments monitor and control the releases of radioactive materials in effluents during actual and potential release Pursuant to these requirements, the inspectors discussed the status of the radiation monitoring system with licensee representative The licensee indicated that a program had been developed to upgrade the existing RMS at Robinson. The inspectors reviewed pertinent documentation concerning this upgrad Problems with the current RMS included antiquated analog equipment which required frequent maintenance and difficulty in finding replacement parts. Also the original monitors did not have a built in check source, requiring.technicians to hand carry a source to the monitors, an ALARA concer The primary objective of the upgrade will be to improve performance and reliability of the RMS. The new equipment will all be from the same manufacturer and will directly replace the old analog equipmen The proposed system will consist of state of the art microprocessor based digital equipmen This upgrade will include selected components for 10 process monitors and 13 area monitor A separate modification package has also been developed to replace the main stack monitor which has been inoperable since September 22, 1989 (Hurricane Hugo).
A "System Team" has been meeting through 1989 and 1990 to develop and plan this modification. This team consisted of 12 individuals with experience in different areas, in order to more effectively meet the needs and requirements of the site organizations that work with the RM *
This project requires that the RMS be removed from service, and therefore can only be performed during an outage. These projects are scheduled to be performed during the outage starting in September 1990. The upgrades will be scheduled around significant outage activities, such as refuelin The inspector discussed the importance of RMS operability with licensee management and technical personne The progress of the two modification packages for the RMS will be reviewed by regional inspectors during subsequent inspection No violations or deviations were identifie. Semiannual Radioactive Effluent Release Report (84750)
TS 6.9.d requires the licensee to submit a Semiannual Radioactive Release Report within 60 days after January 1 and July 1 of each year covering the operation of the facility during the previous six months of operatio These reports summarize the amounts of liquid and gaseous effluents released from the site and assess the dose to offsite populations from these effluent Pursuant to these requirements, the inspectors reviewed the Semiannual Radioactive Effluent Release Reports for January - June 1989 and for July - December 1989, and discussed the reports with licensee representative The effluent information presented in Table A was obtained from current and previous effluent report In general, the 1989 gaseous effluents decreased as compared to 1987 and 1988 values. The licensee attributed this to improved processin In particular the licensee had switched the liquid radwaste from an evaporator system to demineralizer This improved both the liquid and the gaseous fission and activation products, because a vacuum was pulled on the evaporator system which stripped gases from the liquid, increasing the gaseous fission and activation gases being sent out the plant stac There was an increase in the 1989 gaseous particulate The licensee attributed this to ventilation problems in the Residual Heat Removal pi A plant modification to correct this has been performe Liquid and gaseous effluents and the resulting doses were well within TSs, 10 CFR 20, 40 CFR 190, and 10 CFR 50 Appendix I effluent limit The inspectors concluded that the TS reporting requirements regarding the effluent reports had been met and that the releases were well within TS and regulatory requirement TABLE A Effluent Release and Dose Summary for Robinson Activity Released (curies)
1987 1988 1989 a. Gaseous Effluents 1. Fission and 7.70E+02 1.04E+03 2.78E+01 Activation Products 2. Iodines 2.08E-02 1.07E-03 3.17E-06 3. Particulates 1.56E-05 2.76E-05 1.38E-04 4. Tritium 1.39E-01 7.66E+00 4.18E+00 Liquid Effluents 1. Fission and Activation Products 9.36-01 9.64E-01 2.81E-01 2. Tritium 2.47E+02 5.36E+02 1.63E+02 3. Gross Alpha
§LLD
§LLD
§LLD Dose Estimates (mrem)
1. Gaseous Effluents Whole Body 8.37E-02 2.61E-01 3.49E-02 Ski.02E-01 7.39E-01 5.73E-02 2. Liquid Effluents Liver 1.57E-01 5.12E-02 4.68E-02 7. Radiological Environmental Monitoring (84750)
TS 6.9.e requires the submittal of a routine Radiological Environmental Operating Report. This report summarizes the results of the Radiological Environmental Surveillance Program, which measures accumulation of radioactivity in the environment, and determines whether the radioactivity detected is due to the operation of the Robinson Plant. This program also assesses the dose to off-site populations from plant effluent Pursuant to these requirements, the inspector reviewed the report for 198 The average gross beta activity for 404 air samples was 1.60 E-2 picoCuries per meter cubed (pCi/m)
for the indicator samples, and 1.58E-2 pCi/m for the control sample This is
7 consistent with preoperational average of 0.14 pCi// The quarterly composite gamma analyses for the air particulate samples did not show any radionuclides typical of plant effluents. All 404 air cartridge samples from the indicator and from the control locations had less than detectable (LLD) activities for iodine-13 Cesium-137 (CS-137)
was the only nuclide detected in the broad leaf vegetation sample Eight out of 19 control location samples indicated Cs-137 with an average concentration of 4.68E-01 pCi/gram, while 32 out of 36 indicator locations showed 1.56E-01 pCi/gram. The report stated that this was due to world-wide fallout in the biosphere, and that any contribution of Cs-137 from Robinson was not discernabl Gamma isotopic analyses were performed on the edible portions of fre swimming and bottom feeding fish from Bee Lake and May Lake (control locations) and Robinson Lake and Prestwood Lake (indicator locations). Statistically there were no discernable increases in the concentration of Cs-137 due to plant operations if the average concentration of Cs-137 in the free swimming indicator fish was 2.13E-1 pCi/gram. Regulatory Guide 1.109 methodology yielded a dose to the -adult liver of 0.49 millirem per year (mrem/y).
The average concentration of Cs-137 in the free swimming control fish-for Bee Lake was 2.8E-01 pCi/gra Cs-134 was identified in one fish from Prestwood Lake at a concentration of 5.3E-2 pCi/gra Regulatory Guide 1.109 methodology, yielded a dose to the liver of 0.165 mrem/ The monthly ground water analyses for. tritium and gamma emitting radionuclides were all less than LL One of 26 milk samples revealed Cs-137 at 4.121.51 pCi/lite The sampling location was 10.1 miles east of the plant (no milk animals are located within a radius of five miles from the plant). The dose to an infants liver would be 0.83 mrem assuming consumption of milk with this concentration for one yea Since t his occurred in only one sample out of 26, the dose would probably be much les No isotopes due to plant effluents were detected in food crops of collards, peaches, and cor Shoreline sediments are collected twice a year at two locations. The first location had one sample indicate 1.02E-2 pCi/grams of Cs-137, yielding a dose of 1E-04 mrem to a teenager, assuming 67 hour7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br />s/year'
exposure (recreational exposure).
The second location for shoreline sediment was at an ash pond which had limited public accessibility -and had received slightly contaminated soil under approval by the NRC and the state of South Carolin Analysis indicated Co-60 (4.54E-2 pCi/g) and Cs-137 (4.53 E-2 pCi/g).
These would yield an annual dose of less than 0.002 mrem/ o All surface water samples indicated less than LLD for the monthly gamma-emitting analyses. Tritium was detected in four out of 12 samples at the indicator station for evaluating routine releases at an average concentration of 1.86E+3 pCi/lite Regulatory Guide 1.109 methodology yielded a dose of 3.69E-3 mrem/year to the maximum individual (due to the ingestion of fish) for this concentratio Tritium was detected at the ash pond with an average concentration of 1.68E+3 pCi/liter. (See paragraph above for details concerning the ash pond.)
o The average dose rate for all indicator thermoluminescent dosimetry (TLD) locations was 1.09 mrem/wk. The control location average was 1.12 mrem/wk. There was no measurable effect detected from plant operation In summary, no significant radiological consequences to the environment were attributable to the operation of Robinson in 1989 from airborne, waterborne, aquatic, ingestion, or direct exposure pathway No violations or deviations were identifie.
Post Accident Sampling System (84750)
NUREG-0737, Criterion 2a requires the licensee to establish an onsite radiological analysis capability to provide quantification of noble gases, iodines, and non-violatile radionuclides in the reactor coolant and containment atmospher Pursuant to these requirements, the inspectors examined Robinson's Post-Accident Sampling System (PASS)
for reactor coolant, gaseous effluents, containment atmosphere, pH, boron, total gas, and hydrogen. The inspectors discussed PASS operation and maintenance experience with licensee personne The inspectors determined that the PASS was flushed monthly to help ensure system operability and to determine if any problems had develope The PASS was run semiannually for comparison with routinely obtained reactor coolant sampling data. The inspectors reviewed PASS Operational Test Data Sheets for the last four tests performed in November, 1988, June, 1989, November, 1989, and June, 199 These records indicated that a high percentage of the tests met the PASS Acceptance Criteria. In one case, a work request was written to determine why the measurement did not meet the criteria (November, 1988), and in another case, some measurements could not be performed due to the plant being in.an outage (November 1989).
All designated measurements met the PASS Acceptance Criteria for the test performed in June 199 The inspectors also reviewed records and discussed the licensee's methods to ensure continuing PASS operability, and to track required tests and procedure The inspectors also performed a brief review of selected portions of several procedures which covered different aspects of the
PASS operatio Within the scope of this review, the inspectors determined that the procedures were adequate for their intended purposes, and that PASS operability was being maintaine No violations or deviations were identifie. Radioactive Liquid Effluent-Treatment System (84750)
TS 3.16.1 defines the operating requirements for the liquid radwaste treatment syste TS 3.9.1 defines the concentration limits for radioactive material in liquid effluents released to unrestricted area These requirements provide assurance that releases of radioactive material in liquid effluents will be ALAR Pursuant to these requirements, the inspectors discussed operation of the liquid radwaste system with a licensee representative; and toured the plant, examining components of the system;, in order to gain familiarity with system operabilit The inspectors determined that Robinson was not using their waste evaporator The liquid radwaste system was being run by an outside contractor. The liquid waste entered the system through two prefilters, consisting of a 25 micron bag filter and a canister containing six 10 micron filter cartridges. Next, the liquid was processed through four deep bed demineralizers. Three of these were cation demineralizers, while the fourth contained a anionic macroreticular resin which acted as a submicron filter. The liquid then passes through a canister containing six 0.45 micron cartridge filters. Liquid into thesystem was pretreated with a coagulant to increase filter efficiency. Liquid out of the system was collected in the Waste Hold Up Tank or in one of the three Chemical Volume Control System Hold Up Tanks. The flow through the system was limited by the maximum flow through the demineralizers of 100 gallons per minute. The licensee representative stated that the order the waste water flowed through the four demineralizers could be adjusted based on need and resin conditions in the demineralizers. The resin was sluiced out of the demineralizers approximately every four months for dewatering and shipment for buria Each deep bed demineralizer held 20 cubic feet of resin. The prefilters and post filters were changed out approximately every othe month, depending on liquid radwaste condition The inspectors also reviewed selected 1990 liquid waste release permit The examined packages appeared complete per procedural requirement The inspectors determined, within the scope of the review, that the operability of the liquid radwaste system was adequat No violations or deviations were identifie *
1 Counting Room Instrumentation (84750)
The inspectors discussed a recent incident at a waste facility where an incorrect detector efficiency had been u.sed due to source strength,
assumption errors for the strontium-90 (Sr-90) radioactive standard. the certificate of activity supplied by the source manufacturer did not include the activity contribution from the strontium daughter product, yttrium-90 (Y-90), which was equal to that of the Sr-90. This resulted in an instrument beta efficiency that was twice the correct value and subsequently underestimated the beta activity by a factor of tw The inspector determined that inplant count rooms did not use Sr-90 as a calibration sourc The licensee noted this information for possible future us No violations or deviations were identifie.
Organization (84750)
TS 6.2.1 requires that the licensee establish an onsite and an offsite organization for unit operation and corporate management, and that this organization shall include the positions for activities affecting the safety of the nuclear power plan The inspectors reviewed the licensee's organization, staffing level, and lines of authority as they related to radiation protection and radioactive material control to verify that the licensee had not made organization changes which would adversely affect the ability to control radiation exposures or radioactive materia The inspectors determined that in August of 1989, Robinson underwent a reorganization. The licensee indicated that this reorganization had a minimal impact on Robinso One chemistry foreman and one technician position were eliminate Overall, approximately 13 positions were eliminated at Robinson. The inspectors determined, based on the scope of the review performed, that the reorganization did not adversely affect Robinson's ability to control radioactive exposures or radioactive material No violations or deviations were identifie.
Laboratory Radiation Protection Practices (84750)
TS 6.11 specifies that procedures for personnel radiation protection shall be adhered to for all operations involving personnel radiation exposur The licensee's procedure, PLP-016, Revision 9, Radiation Work Permit (RWP)
Program, describes the RWP program including the use of an RW Section 5.2.2 specified that all personnel shall read, understand, and follow the provisions set forth on their RW The instructions on RWP 90-0003 specify that a lab coat and gloves are required when working with contaminated material During a tour of the licensee's laboratory facilities, the inspectors observed a laboratory technician removing a particulate filter from inside a fume hood and placing it in a desiccator located on a laboratory bench adjacent to the fume hoo The technician was in the process of performing an analysis for suspended solids in a liquid sample taken from a waste hold-up tan The fume hood was posted with an
"Internal Contamination" sign. Although forceps were used to pickup the planchet containing the filter, the technician was not wearing gloves and a lab coat while transferring the filter from the hood to the desiccator. This is contrary to the instructions in RWP 90-000 After having been informed of the inspectors' observations, the licensee issued a "Radiation Safety Violation" to document the details of the violation and the corrective action to be take The corrective action consisted of retraining the laboratory personnel on the importance of complying with RWPs and radiation postings and was initiated before the end of the inspection. The NRC identified violation is not being cited because the criteria specified in Section V.A of the NRC Enforcement Policy were satisifed (NCV:
50-261/90-18-01).
One NCV was identifie O 1 Exit Interview The inspection scope and results were summarized on August 17, 1990, with those persons indicated in Paragraph The inspectors described the areas inspected and discussed in detail the inspection results as listed in the summary. Proprietary information is not contained in this repor Dissenting comments were not received from the license ATTACHMENT 1 TABLE 1 NRC-LICENSEE SAMPLE COMPARISON EVALUATION FOR ROBINSON STEAM ELECTRIC PLANT, AUGUST 13-17, 1990 Concentration (uCi/unit)
Ratio Sample Isotope Licensee NRC Resolution Licensee/NRC Comparison Reactor Coolant 50 ml Bottle Detector No. PGT 1452 Co-58 5.51 E-4 5.380.10 E-4
1.02 Agreement Co-60 2.62 E-5 2.640.33 E-5
0.99 Agreement 1-131 1.16 E-4 1.060.11 E-4
1.09 Agreement 1-132 2.92 E-3 2.580.02 E-3 129 1.13 Agreement 1-133 1.51 E-3 1.470.01 E-3 147 1.03 Agreement 1-134 4.92 E-3 4.130.16 E-3
1.19 Agreement 1-135 2.29 E-3 2.780.05 E-3
0.82 Agreement Detector No. APTEC 1299 CO-58 5.57 E-4 5.380.10 E-4
1.04 Agreement Co-60 3.68 E-5 2.640.33 E-5
1.39 Agreement 1-131 1.54 E-4 1.060.11 E-4
1.45 Agreement 1-132 2.96 E-3 2.580.02 E-3 129 1.15 Agreement 1-133 1.53 E-3 1.470.01 E-3 147 1.0i Agreement 1-134 5.28 E-3 4.130.16 E-3
1.28 Agreement 1-135 2.28 E-3 2.780.05 E-3
0.86 Agreement Detector No. ORTEC 1602A Co-58 6.03 E-4 5.380.10 E-4
T. 12 Agreement Co-60 2.91 E-5 2.640.33 E-5
1.10 Agreement 1-131 1.20 E-4 1.060.11 E-4
1.13 Agreement 1-132 2.93 E-3 2.580.02.E-3 129 1.14 Agreement 1-133 1.54 E-3 1.470.01 E-3 147 1.05 Agreement 1-134 4.93 E-3 4.130.16 E-3
1.19 Agreement 1-135 2.41 E-3 2.780.05 E-3
0.87 Agreement Liquid Waste One Liter Marinelli Detector No. APTEC 1299 Xe-51 3.29 E-7 3.430.90 E-7
0.96 Agreement Co-60 1.92 E-5 2.060.03.E-5
0.93 Agreement Sb-125 2.27 E-6 1.850.22 E-6
1.23 Agreement
- 0
Concentration (uCi/unit)
Ratio Sample Isotope Licensee NRC Resolution Licensee/NRC Comparison Liquid Waste One Liter Marinelli Detector No. PGT 1452 Xe-133 4.58 E-7 3.430.90 E-7
1.36 Agreement Co-60 2.09 E-5 2.060.03 E-5
1.01 Agreement Sb-125 1.90 E-6 1.850.22 E-6
1.03 Agreement Detector No. ORTEC 1602 Xe-133 4.23 E-7 3.430.90 E-7
1.23 Agreement Co-60 1.99 E-5 2.060.03 E-5
0.97 Agreement Sb-125 1.68 E-6 1.850.22 E-6
.8 0.91 Agreement Containment Gas, 1260 cc Gas Marinelli Detector No. APTEC 1299 Ar-41 1.24 E-6 1.060.07 E-6
1.17 Agreement Xe-133 2.18 E-6 1.960.06 E-6
1.11 Agreement Xe-135 1.21 E-7 1.160.15 E-7
1.04 Agreement Detector No. PGT 1452 Ar-41 1.09 E-6 1.060.07 E-6
1.03 Agreement Xe-133 2.08 E-6 2.120.14 E-6
1.06 Agreement Xe-135 8.55 E-8 1.160.15 E-7
0.74 Agreement Detector No. ORTEC 1602A Ar-41 1.29 E-6 1.060.07 E-6
1.22 Agreement Xe-133 2.08 E-6 1.960.06 E-6
1.08 Agreement Xe-135 1.09 E-7 1.160.15 E-7
0.94 Agreement Charcoal Cartridge Detector: No. ORTEC 1602 Co-60 5.10 E-2 5.170.05 E-2 103 0.99 Agreement Cd-109 3.90 E-1 3.490.03 E-1 116 1.12 Agreement Sn-1.13 3.13 E-3 2.740.16 E-3
1.14 Ag reement Ce-139 2.39 E-3 2.310.08 E-3
1.03 Agreement Co-57 5.23 E-3 4.880.08 E-3
1.07 Agreement Y-88 4.19 E-3 4.190.20 E-3
1.00 Agreement Cs-137 4.89 E-2 4.500.04 E-2 112 1.09 Agreement Co-60 5.17 E-2 5.170.05 E-2 103 1.00 Agreement Cd-109 3.68 E-1 3.490.03 E-1 116 1.05 Agreement Sn-113 2.89 E-3 2.740.16 E-3
1.05 Agreement Ce-139 2.39 E-3 2.310.08 E-3
1.03 Agreement Co-57 5.16 E-3 4.880.08 E-3
1.06 Agreement Y-88 4.39 E-3 4.190.20 E-3
1.05 Agreement Cs-137 4.74 E-2 4.500.04 E-2 112 1.05 Agreement
Concentration (uCi/unit)
Ratio Sample Isotope Licensee NRC Resolution Licensee/NRC Comparison Charcoal Cartridge Detector No. APTEC 1299 Co-60 5.57 E-2 5.170.05 E-2 103 1.08 Agreement Cd-109 3.94 E-1 3.490.03 E-1 116 1.13 Agreement Sn-113 3.37 E-3 2.740.16 E-3
1.23 Agreement Ce-139 2.46 E-3 2.310.08 E-3
1.06 Agreement Co-57 5.55 E-3 4.880.08 E-3
1.114 Agreement Y-88 4.82 E-3 4.190.20 E-3
1.15 Agreement Cs-137 5.25 E-2 4.500.04 E-2 112 1.17 Agreement Detector No. PGT 1452 Co-60 5.42 E-2 5.170.05 E-2 103 1.05 Agreement Cd-109 3.92 E-1 3.490.03 E-1 116 1.12 Agreement Sn-113 3.09 E-3 2.740.16 E-3
1.13 Agreement Ce-139 2.39 E-3 2.310.08 E-3
1.03 Agreement Co-57 5.50 E-3 4.880.08 E-3
1.13 Agreement Y-88 4.79 E-3 4.190.20 E-3
1.14 Agreement Cs-137 4.90 E-2 4.500.04 E-2 112 1.09 Agreement Particulate Filter Of Reactor Coolant Detector No. APTEC 1299 Cr-51 3.63 E-4 3.920.04 E-4
0.93 Agreement Mn-54 3.96 E-6 3.520.24 E-6
1.12 Agreement Co-58 4.89 E-4 4.940.02 E-4 9200 0.99 Agreement Co-60 2.59 E-5 2.580.05 E-5
0.99 Agreement Zr-95 6.53 E-6 7.320.55 E-6
0.89 Agreement Nb-95 7.41 E-6 8.060.38 E-6
0.92 Agreement Detector No. PGT 1452 Cr-51 3.71 E-4 3.920.04 E-4
0.95 Agreement Mn-54 3.25 E-6 3.520.24 E-6
0.92 Agreement Co-58 4.89 E-4 4.940.02 E-4 1200 0.99 Agreement Co-60 2.63 E-5 2.580.05 E-5
1.02 Agreement Zr-95 6.21 E-6 7.320.55 E-6
0.85 Agreement Nb-95 7.97 E-6 8.060.38 E-6
0.99 Agreement Detector No. ORTEC 1602A Cr-51 3.65 E-4 3.920.04 E-4
0.93 Agreement Mn-54 2.96 E-6 3.520..24 E-6
0.84 Agreement Co-58 4.76 E-4 4.940.02 E-4 1200 0.96 Agreement Co-60 2.54 E-5 2.580.05 E-5
0.98 Agreement Zr-95 7.53 E-6 7.320.55 E-6
1.03 Agreement Nb-95 7.48 E-6 8.060.38.E-6
0.93 Agreement
Concentration (uCi/unit)
Ratio Sample Isotope Licensee NRC Resolution Licensee/NRC Comparison Particulate Filter Of Reactor Coolant HP Detector:
QUANTUM Cr-51 3.94 E-4 3.920.04 E-4
1.00 Agreeent Mn-54 3.64 E-6 3.520.24 E-6
1.03 Agreement Co-58 5.16 E-4 4.940.02 E-4 11200 1.04 Agreement Co-60 2.59 E-5 2.580.05 E-5
1.00 Agreement Zr-95 7.81 E-6 7.320.55 E-6
1.07 Agreement Nb-95 7.73 E-6 8.080.38 E-6
0.96 Agreement
ATTACHMENT 2 CRITERIA FOR COMPARISONS OF ANALYTICAL MEASUREMENTS This attachment provides criteria for the comparison of results of analytical radioactivity measurement These criteria are based on empirical relationships which combine prior experience in comparing radioactivity analyses, the measurement of the statistically random process of radioactive emission, and the accuracy needs of this progra In these criteria, the "Comparison Ratio Limits"'
denoting agreement or disagreement between licensee and NRC results are variabl This variability is a function of the ratio of the NRC's analytical value relative to its,
associated statistical and analytical uncertainty, referred to in this program as "Resolution" 2.
For comparison purposes, a ratio between the licensee's analytical value and the NRC's analytical value is computed for each radionuclide present in a given sample. The computed ratios are then evaluated for agreement or disagreement based on "Resolution."
The corresponding values for "Resolution" and the
"Comparison Ratio Limits" are listed in the Table belo Ratio values which are either above or below the "Comparison Ratio Limits" are considered to be in disagreement, while ratio values within or encompassed by the "Comparison Ratio Limits" are considered to be in agreemen TABLE NRC Confirmatory Measurements Acceptance Criteria Resolution vs. Comparison Ratio Limits Comparison Ratio Limits Resolution for Agreement
<4 0.4 -.5 -.6 -
1.66 16 -
0.75 -
1.33 51 -
200 0.80 -
1.25
>200 0.85 - 1.18
'Comparison Ratio = Licensee Value NRC Reference Value 2Resolution = NRC Reference Value Associated Uncertainty