IR 05000400/1990012
| ML18009A591 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/03/1990 |
| From: | Decker T, Seymour D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18009A590 | List: |
| References | |
| 50-400-90-12, NUDOCS 9007170094 | |
| Download: ML18009A591 (24) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEO RG IA 30323 J'JL 0,5 SX Report No.:
50-400/90-12 Licensee:
Carolina Power and Light Company P. 0.
Box 1551 Raleigh, NC 27602 Docket Nos.:
50-400 License No.:
NPF-63 Facility Name:
Shearon Harris Nuclear Power Plant Inspection Conducted:
June 11-15, 1990 Inspector:
</A~i~.-.~.: 6 /!c'/v"'c eymour Approved by:
<7 us v-P~'(iivv;:i.c
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R.
Dec er, Chief Radiological Effluents and Chemistry Section Emergency Preparedness and Radiological Protection Branch Division of Radiation Safety and Safeguards Date Signed Date Signed SUMMARY Scope This routine, unannounced inspection was conducted in the areas of confirmatory measurements, count room quality control, the Semiannual Radiological Effluent Release Report, the Annual Radiological Environmental Operation Report, Spent Fuel Pool Water guality, and the Post Accident Sampling System (PASS).
Results:
One Non-Cited Violation (NCV) was identified: failure to prevent the potential release of radioactive materials at concentrations which would exceed
CFR 20 limits.
The licensee's confirmatory measurement results were in agreement for all sample types analyzed.
No significant radiological consequences attributable to the operation of the plant in 1989 were noted from airborne, waterborne, aquatic, ingestion, or direct exposure pathways.
For 1989, liquid, gaseous, and particulate effluents were well within Technical Specification,
CFR 20, and
CFR 50 effluent limits.
The radiation doses to the maximum exposed member of the public were a small fraction of the limits allowed by 40 CFR 190 and
CFR 20.105(c).
9007j.70094 9007CI5 PDR AQI:ICK 0 000400
Harris is formulating plans on dealing with iron oxide deposits in their spent fuel pools and transfer canals.
The PASS is continuing to experience operability problems and is not operating at optimum levels.
In the areas inspected, violations or deviations were not identifie REPORT DETAILS Persons Contacted Licensee Employees S.
Buch, Environmental and Chemistry Technician D. Elkins, Shipment Director, Spent Nuclear Fuel
- G. Forehand, Manager, Quality Assurance/Quality Control (QA/QC)
P. Gastor, Nuclear Operations Specialist
- C. Gibson, Director, Programs and Procedures
- C. Hinnant, Plant General Manager
- S. Johnson, Chemistry Foreman
- G. Nathan, Senior Specialist, Chemistry
- A. Poland, Project Specialist, Environmental and Radiation Control (ESRC)
- J. Sipp, Yanager, ESRC
- H. Smith, Manager, Radwaste
- M. Station, Power Agency
- D. Tibbitts, Manager, Regulatory Compliance
- M. Wallace, Senior Specialist, Regulatory Compliance
- L. Woods, Engineering Supervisor, Technical Support Other licensee employees contacted during this inspection included engineers, operators, technicians, and administrative personnel.
NRC Resident Inspector
<<J.
Tedrow, Senior Resident Inspector
- Attended exit interview Acronyms and Initialisms used throughout this report are listed in the last paragraph.
Licensee Action on Previously Identified Inspector Follow-up Items (IFIs)
(92701)
(Closed)
IFI 50-400/88-24-02:
Review Licensee Resolution of Flow Measurement Device Operability in Plant Stacks.
As discussed in Inspection Report No.
88-25, turbulent flow problems in three out of four plant gaseous effluent stacks (Nos.
1, 5, and 5A) caused by the relatively short and wide design of these stacks had prevented the licensee from accurately measuring the flow rates out of these stacks.
This had placed the licensee in continuous Action Statements of the Technical Specifications (TSs)
'equiring periodic flow rate estimations.
To resolve this issue, the licensee
. planned on using a microprocessor to summarize the individual
'nfluents into each stack by sensing the operation of the fan motors under various operating fan configurations.
The microprocessor flow values will be determined from in situ flow measurements made under these fan configurations.
Plans were made to have the system installed, tested and operational by September 1989.
The modifications were approved, but due to delays caused by a refueling outage, installation did not commence until January 1990.
At the time of this inspection, the installation was completed on stacks No.
5 and 5A.
A delay was incurred with stack No.
1 in that this stack was rated as being safety-related.
This rating required verifiable separation of safety and non-safety cabling and electrical inputs.
The inspector determined through conversations with the licensee that a
package=
was being prepared in coordination with the Nuclear Engineering Department which would downgrade this stack to a non-safety rating.
This package was expected on site by July 6, 1990.
Installation is expected to be completed by September 1,
1990.
Based on this licensee commitment, and on discussions with the licensee, this item, is considered closed.
Licensee Event Report (LER)
(92700)
The inspector reviewed LER 90-014-00 and discussed the corrective actions that had been implemented with licensee personnel.
This LER dealt with two liquid radwaste releases which had been made by two separate operators at release rates greater than that specified on the release permits.
The licensee identified this situation on Yiay 16, 1990.
The two release permits were issued on Nay 13, 1990.
The release permits specified release rates of 14.6 gallons per minute (gpm)
and carried a warning that this release rate was less than the maximum pump capacity of 35 gpm.
The releases were being made from the Waste Evaporator Condensate Tanks (WECTs),
an uncommon release point.
The Waste Monitor Tanks, which are typically released at full pump'apacity, are frequently released, and share the same release path with the WECT.
Both operators initialed the step in the release procedure requiring that the release flow be slowly increased until it is at 90 percent of the flow rate specified on the Pre-release Permit.
The licensee determined that the cause of the event was personnel error.
The corrective actions included requiring the operators to transcribe the maximum release rate from the discharge permit to the discharge log, requiring the cognizant Radwaste Shift Foreman to verify the maximum release rate prior to discharge and the counseling of appropriate Radwaste personnel on the need for attention to detail and correct implementation of written procedure f
Post-release calculations by the licensee showed that
CFR
or
CFR 50 limits were not exceeded because the Cooling Tower dilution flow was high enough to maintain the required dilution factor.
The release permits specify a
minimum Cooling Tower blowdown rate of 4,000 gpm.
Releases are terminated automatically if the flowrate drops below 4,000 gpm.
However, in these two cases, if the Cooling Tower blowdown flow rate had been reduced, but still at a point above the minimum 4,000 gpm cutoff, 10 CFR 20 limits would have been exceeded.
The inspector reviewed selected pre-and post-release permits for the WECTs for 1989 and 1990 to date and determined that these two releases were isolated incidents for releases from WECTs.
The inspector also reviewed SHNPP Significant Operational Occurrence Report No.90-068, which detailed the licensee identification and analysis of this situation.
The inspector
'lso discussed with cognizant licensee personnel the fact that the flow instrumentation for this release path was inoperable, requiring the operators to use the release time and change in tank volume to determine an average release rate.
CFR 20 limits for releases are based on instantaneous release rates (instantaneous concentrations).
The inspector verified that an outstanding Plant Change Request (No.
4746) existed to replace the flow orifices on this instrumentation, returning it to operability.
TS 3.3.3. 10 requires the alarm/trip setpoints-of the radioactive liquid effluent monitoring instrumentation channels to be set to ensure that the
'nstantaneous limits as specified in
CFR 20, Appendix B, Table II, Column 2 are not exceeded.
The inspector informed licensee management that failure to prevent the potential release of radioactive materials at concentrations which would exceed the
CFR 20 instantaneous limits would be considered an apparent violation of TS 3.3.3. 10.
Because the licensee had identified this problem and implemented appropriate corrective actions, a licensee-identified Non-Cited Violation (NCV) would be identified.
The criteria specified in Section V.G. 1 of the NRC Enforcement Policy were satisfied and additionally, no response would be required (NC('0-400/90-.,12-01).
One NCV was identified.
Confirmatory h1easurements (84750)
Pursuant to
CFR 20.201(b)
this area was inspected to verify the licensee's ability to conduct precise and accurate measurements.
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-ray spectroscopy
'ystem.
The purpose of these comparative measurements was to verify the licensee's capability to measure quantities of radionuclides accurately in various plant systems.
Analyses were conducted using the licensee's four intrinsic germanium gamma spectroscopy systems.
Sample types and counting geometries included the following:
reactor coolant, 50-milliliter bottle; liquid waste, one-liter marinelli; Waste Gas Decay Tank atmosphere, 33-cc gas bulb; and a charcoal cartridge.
A particulate filter sample was generated for analysis by the filtration of 500 milliliters of reactor coolant.
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 analyzed.
The inspector observed the licensee obtain the Waste Gas Decay Tank atmosphere sample and the Unit ¹1 reactor coolant sample.
Proper sampling techniques and health physics practices were observed.
The inspector reviewed selected portions of Procedure Nos.,CRC-100, Revision 7, entitled,
"Reactor Coolant System Chemistry Control,"
dated May 18, 1990; and CRC-255, Revision 6, entitled,
"Waste Gas Decay Tank Sampling,"
dated April 19, 1990.
The portions reviewed were adequate for the intended purpose.
No violations or deviations were identified.
Radiological Environmental Monitoring (84750)
TS 6.9. 1.3 requires the submittal of a routine Radiological Environmental Operating Report.
Pursuant to this requirement, the inspector reviewed the r'eport for 1989.
The following observations were made:
1-131 was not detected in air samples in 1987, 1988, or 1989.
Gross beta activities on quarterly composite air filter samples in 1989 were unchanged from 1988 arid were at preoperational levels (approximately 1.42E-02 pCi/cubic meter).
I-131 in drinking water remained at less.,than detectable levels in 1989.
Gross beta activities were unchanged from 1988 and were similar to the control locations.
Gamma emitting radionuclides analyses for drinking water indicated concentrations were less than the lower limit of detection (LLD).
Tritium activity in SHNPP drinking water was detected at an average level of 2250 pCi/liter.
Tritium activity in Harris Lake has increased from preoperational levels of 1.23E+03 pCi/liter to 5.5E+03 pCi/liter in 1989.
The levels reported for 1989 are well below the reporting level of 20,000 pCi/liter for drinking water (TS Table 3.12-2).
No detectable gamma activity was revealed in analyses of fish from the indicator and control locations in 198 No tritium or gamma activity was observed in groundwater in 1989.
Two milk samples showed detectable concentrations of I-131 in 1989, one of these from a control station.
The licensee determined that these results were most likely statistical artifacts of the measurement.
No other man-made gamma emitting radionuclides were detected in the milk samples in 1989.
Sediment samples showed a
decrease in Cs-137 and other fission and activation products in 1989.
No increase was seen in shoreline activity in 1989, relative to 1988, as a result of plant activity.
Cs-137 was detected in 1 of 20 samples in 1989 at 1.9E-02 pCi/liter.
In 1988 and 1987, Cs-137 had been detected in control location samples.
The single value for Cs-137 in 1989 was consistent with the values for the control.location.
The thermoluminscent dosimeter (TLD) results for 1989 were consistent with the results obtained in 1988 and 1987.
The doses were similar to preoperational levels.
In summary, no significant radiological consequences attributable to the operation of Harris in 1989 were noted from airborne, waterborne, aquatic, ingestion, or direct exposure pathway.
No violations or deviations were identified.
i 6.
Semiannual Radioactive Effluent Release Reports (84750)
TS 6.9. 1.4 requires the licensee to submit a
Semi-Annual Radiological Effluent Release Report within the time periods specified in TS 6.9.1.4 covering the operation of the facility during the previous six months of operation.
The inspector reviewed the Semiannual Radioactive Effluent Release Reports for 1989.
This review included an examination of the liquid and gaseous effluents for 1989 as compared to those of 1988 and 1987.
This data is summarized below.
One abnormal release was reported in 1989.
Information concerning this release was detailed in Inspection Report No. 89-24.
This release was determined by the licensee to be fully monitored and well below the offsite dose limitations required by the TSs.
Liquid and gaseous effluents showed no significant trends between 1987 and 1989.
For 1989, Harris liquid, gaseous, and particulate effluents were well within TSs,
CFR 20, and
CFR 50 effluent limit Harris Radioactive Effluent Release Summary No. of Abnormal Releases 1987 1988 1989 a.
Liquid b.
Gaseous Activity Released (curies)
a.
Liquid 1.
Fission and Activation 9,08E-01 Products 2.
Tritium 2.48E+02 3.
Gross Alpha 2.73E-04 8.04E-02 4.01E+02 2.55E-06 2. 42E-01 4.58E+02 O.OOE+00 b.
Gaseous 1.
Fission and Activation 1.71E+03 Products 2.
Iodines 0.00E+00 3.
Particulates 4.43E-06 4.
Tritium O.OOE+00 2.25E+03 0.00E+00 4.59E-05 O.OOE+00 1.15E+02 9.47E-07 6.56E-07 0.00E+00 TS 6.9.1.4 requires the Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year and to include an assessment of the radiation doses to the maximum exposed member of the public due to radioactive liquid and gaseous effluents released from the site during the previous calendar year.
Pursuant to these requirements, the inspector reviewed the 1989 annual doses.
For the airborne pathway, the
"Yiaximum Hypothetical Individual" was considered to reside at the site boundary in the sector of highest plume exposure, and had a whole body dose of 1.25E-Ol mrem per year.
This included exposure from noble gases, particulates, iodines, and tritium.
This corresponds to,0.02 percent of the applicable dose limit, slightly less than in 1988.
The skin dose for this individual due to gaseous effluents was 2.65E-01 mrem per year.
For the liquid pathway, the doses to the "Likely Most Exposed Individual" included consuming fish, drinking, swimming, boating, and shoreline exposure.
This individual was defined as an adult, with the liver as the critical organ.
The calculated dose was 2.94E-02 mrem per year, which corresponds to 0.29 percent of the applicable dose limit.
The whole body dose for this individual due to liquid effluents was 2.71E-02 mrem per year, corresponding to 0.90 percent. of the applicable dose limit.
The doses reported were a small fraction of the limits allowed by 40 CFR 190 and
CFR 20.105(c).
No violations or deviations were identifie,
Counting Instrumentation (84750)
N The inspector discussed a recent incident at a waste facility where an incorrect detector efficiency had been used 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 two.
The inspector determined that the in-plant count rooms did not use Sr-90 as a calibration or check source.
The inspector also discussed this information with cognizant personnel at the CPKL/Harris Energy Center and determined that the Center's counting procedures did take into account the Y-90 activity contribution in their efficiency calculations, when applicable.
The licensee noted this information for possible future use.
Ho violations or deviations were identified.
8.
Count Room guality Assurance (84750)
TS 6.8. 1 requires written procedures to be established, implemented and maintained for the equality Assurance (gA)
Program for effluent and environmental monitoring.
The inspector reviewed the licensee's gA program pursuant to these requirements, and to ensure compliance with selected and applicable portions of Regulatory Guide 4.15, equality Assurance for Radiochemical Nonitoring Programs (Normal Operations)
Effluent Streams and the Environment, Revision 1,
February 1978.
The following observations were made:
a.
Daily intensity, gain, resolution, and background quality control charts for the four intrinsic germanium detectors for April 1990 through June 8,
1990, were generally within specified limits, indicating stable detector performance.
b.
Daily background and reliability (efficiency) checks for gross alpha and gross beta analyses for the two LB5100 gas flow proportional counters were within specified control limits for the period January through Narch 1990.
~ cd Daily efficiency and background checks for the Packard 4530 liquid scintillation counter used for tritium analyses were within specified control limits for the period January through Narch 1990.
d.
quarterly, the licensee participated in an extensive split gamma spectroscopic, tritium, gross alpha, and gross beta analyses program with an outside vendor.
The inspector reviewed the results of this cross-check program for the third and fourth quarters of 1989, and for the first quarter of 1990.
The licensee and the vendor were in agreement for all isotope No violations or deviations were identified.
Spent Fuel Pool (84750)
Harris'pent Fuel Pool (SFP) is currently being used to store spent fuel from Brunswick and Robinson.
The licensee began receiving spent fuel from Brunswick Steam Electric Plant (BSEP) in July 1989.
It was noted by the licensee and by the Resident Inspectors that a large amount of corrosion products was adhering to the outside surfaces of this fuel.
This material was dislodged from the fuel bundles when they were put into the transfer canals or into the fuel pool.. It has been hypothesized that the boric acid in the SFP water acted as a chemical cleaner for the fuel.
At the time of this inspection, the inspector noted that this material was deposited on various surfaces and bottom of the
"B" SFP.
A licensee representative estimated that the deposits were approximately 0.25 to 0.50 inches thick at the bottom of the SFP.
Licensee representatives stated that the material was an iron oxide and that this oxide was a very fine, dense, particulate, which settled out rapidly in the water.
The SFP cleanup systems were ineffective in removing these deposits because these systems take suction off the surface of the SFP.
Conventional filtration techniques were impractical due to the small size of the particulate, which would quickly clog filters, requiring frequent replacement, causing an ALARA concern.
The inspector noted to the licensee that the fine particulate might cause serious airborne problems during modifications or operations in the SFP.
The inspector was also concerned that the potential existed for this material to be transported into the primary system during a refueling operation at Harris.
The inspector determined, via discussions with the licensee, that studies performed by GE and other outside contractors indicated that this deposit did not exacerbate corrosion, was benign to zircaloy, and had not caused problems at Brunswick with the fuel racks or vessels.
The licensee also stated that they did not believe that significant amounts of this material would migrate into their primary system.
The inspector reviewed a sample analysis report for the particulate.
This analysis, which was performed by an outside vendor, indicated that the primary contributors to the activity of the particulate were Fe-55, Co-60, Ni-63, Cr-51, and Zn-65.
The inspector determined, through discussions with the licensee, that the licensee was making an effort to resolve this issue.
The licensee had considered several methods for cleaning up the deposit, including using an outside contractor to bring in filters in HICs on trucks, and submersible linings for the walls of the SFP.
BSEP has also started to flush the spent fuel assemblies prior to shipment to Harris, removing large amounts of the deposits.
This situation will be monitored by regional inspectors during subsequent inspections.
No violations or deviations were identifie Post Accident Sampling System (PASS)
(84750)
NUREG-0737, Criterion 2a provides specifications for the establishment of onsite radiological analysis capabilities to provide quantification of noble gas'es, iodines, and non-volatile radionuclides in the reactor coolant and containment atmosphere.
TS 6.8.4.e requires that a
program be established,'implemented and maintained to ensure the capability to obtain and analyze, under accident conditions, reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples.
Pursuant to these specifications, the inspector reviewed portions of selected procedures for the operation, maintenance, and testing of the licensee's PASS, and examined components of the PASS facility.
The inspector also discussed system operation, performance testing, and analytical capabilities of the PASS with the licensee.
Discussions with the licensee and the review of records indicated that operability of the PASS had been an on-going problem in the last few years.
As part of the review, the inspector examined PASS quarterly Test Results for 1987, 1988, 1989, and 1990 to date.
These records summarized the resul ts of the quarterly tests in terms of passing of failing the comparisons between PASS analyses and routine RCS sampling, as detailed in NUREG-0737 Criterion 10 and Attachment No.
1 to the Generic Letter.
These analyses included:
boron, isotopic activity, and chlorides for diluted RCS.
liquid; dissolved hydrogen and isotopic activity for stripped RCS gas; in-line pH, in-line dissolved oxygen, and in-line hydrogen; and gas activity and iodine activity for containment air.
These records indicated a continuing problem with the stripped gas results and with the in-line pH, dissolved oxygen and hydrogen.
Discussions with the licensee indicated that maintenance work was on-going with the PASS, including hydrogen leak testing and overall balancing of the system.
This system has received attention from the system enginee'rs and from the Technical Support organization.
Discussions with the system engineer for the PASS indicated that the system has received support, manpower and resources from plant management.
The inspector also reviewed selected portions of three procedures dealing with the operation and maintenance of the PASS.
These procedures were No. CRC-821, entitled "Postaccident RCS/RHR Sampling,"
dated May 9, 1989; No. CRC-830, entitled "Periodic Maintenance and Operability Verification of the PASS,"
dated Yiay 20, 1988; and No.
CRC-823, entitled "Postaccident Containment Air Sampling,"
dated January 18, 1990.
The portions reviewed were adequate for their intended purpose, except for one possible change as detailed in Paragraph 11.
The inspector discussed, with the PASS system engineer, and with licensee management, the importance of maintaining the PASS in a fully operational conditio No violations or deviations were identified.
Emergency Response Drill (84750)
The inspector observed portions of an in-house emergency response drill that was conducted by the licensee.
Among the objectives of this drill, the inspector was most concerned about the ability of the PASS to demonstrate:
the capability to draw samples from the PASS under simulated elevated radiological conditions; the proper procedures for on-site collection and analysis of samples from radiological monitoring,and sampling stations withir the plant; and proper procedures and use of protective measures for entering a Radiologically Controlled Area under abnormal radiological conditions.
The inspector accompanied the PASS team and the PASS team drill evaluator while they:
were directed to obtain a
PASS sample; received the appropriate instructions for dressing out and HP coverage; obtained the sample; and transported the sample to the count room laboratory for counting and analysis.
The inspector also observed portions of the analysis.
Post drill, the inspector shared comments/concerns with the PASS team drill evaluator.
These comments included improving the communication ability of the PASS team, improving familiarity of potential PASS team personnel with the system and procedures; and simplification and/or clarification of 'one portion of the procedure utilized for the PASS sampling (No.
CRG 821).
Overall the inspector considered the evaluation of the PASS drill and the HP coverage of the PASS drill to be thorough, and that the three objectives centering around the PASS to have been adequately
.
met.
No violations or deviations were identified.
Exit Interview The inspection scope and results were summarized on June 15, 1990, with those persons indicated in Paragraph 1.
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 report.
Dissenting comments were not received from the licensee.
Acronyms and Initialisms ALARA - As Low As Reasonably Achievable BSEP - Brunswick Steam Electric Plant cc - cubic centimeter CFR - Code of Federal Regulation E 5 RC - Environmental and Radiation Control GE - General Electric Company gpm - gallons per minute HIC - High Integrity Container HP - Health Physics IFI - Inspector Follow-up Item LER - Licensee Event Report
LLD - Lower Limit of Detection mrem - millirem NCV - Non-Cited Violation No. - Number
- NRC - Nuclear Regulatory Commission PASS - Post Accident Sampling System pCi - picocurie gA - guality Assurance gC - guality Control RCS - Reactor Coolant System SFP - Spent Fuel Pool SHNPP - Shearon Harris Nuclear Power PLant TLD - Thermoluminesent Dosimeter TS - Technical Specification
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3.03C-3 1.07E-2 6.20E-3 1.59C"2 5. I IE-3 3. 17E"3 1.06E-2 6.07<.-3 1.74E-2 5.6 I <O. 19C-3 3. 18!.O. 10E" 3 I. 10 <.O. 05E-2 5.72<0.<<2E-3 1.76%0.07C-2 5. 61 <0. 19E'-3 3. 18
< O. IOE-3 1. 10<0.05E-2 j".72<0.42E-3 1.76+0.07E-2 5 61 <0. 19E-3 3. 18 <0. 10E-3 1. 10>0. 05E-2 5.72+0.42E-3 I. 76<0.07E-2 5. 61iO. 19E-3 3. 18+0. 10E-3 1. 10<0.05E-2 5.72<0.42E-3 1.7640.07E-2
32
14
30
22
25
32
14
30
22
25 0.96 1.00 1.0<i 1.00 0.89 0.89 1.02 0.93 1.06 0. 92 0.90 0.95 0.97 1. 08 0.90 0. 91 1. Of)
0.96 1.06 0.99 Agrecmcnt Agrrement Agrecmcnt Agrccmont
"Agreement Agreement Agrcemcn<.
Agreemont Agrcemcnt Agreement Agreemcnt Agrcernent A<grecment Agreement Agrocment Agreement Agreement Agreement Agreement Agreement Charcoal Cartridge Analytics Spike Detector gl Detector //2 Detector //3 Detector
//l<
I-131 I-131 I-131 1-131 2. 72E-1 2. 75E-I 2. 79E" 1 2.68E-l 2.74+0.03E-1 2.71<+0.03E-1 2.74<0.03C-1 2.74+0.03E-1
91
0.99 1.00 1.02 Agreement Ag rccment Ag rocment Agreement Waste Gas Decay Tank Atmosphere 33 ml gas bulb Detector //I Detector //2 Detector //3 Detector //4 Xe-133 Xe-133 Xe-133 Xe-133 4. 15E-<<
4.03C-<<
3.93E-<<
3.89C-<<
3.27+0.07E-<<
3.27<0.07E-<<
3.27X0.07C-<<
3.27<0.07E-4
47 1<7
1.27 1.23 I. 20 l. 19 Agreement Agreement Agrccment Agrccmcrll,
ATTACHMENT 2 CRITERIA FOR COMPARISONS OF ANALYTICALMEASUREMENTS This attachment provides criteria for the comparison of results of analytical radioactivity measurements.
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 program.
In these criteria, the
"Comparison Ratio Limits" denoting agreement or disagreement between licensee and NRC results are variable.
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"~.
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 below, 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 agreement.
TABLE NRC Confirmatory Measurements Acceptance Criteria Resolution vs.
Comparison Ratio Limits Resolution Comparison Ratio Limits for A reement 4" 7 8-15 16 - 50 51 " 200
>200 0.4 - 2.5 0.5 - 2.0 0.6 - 1.66 0.75 - 1.33 0.80 - 1.25 0. 85 - 1. 18
'Comparison Ratio = Licensee Value NRC Reference Value
~Resolution
=
NRC Reference Value Associated Uncertainty