IR 05000324/1987007
| ML20209E493 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/14/1987 |
| From: | Adamovitz S, Gloersen W, Kahle J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20209E447 | List: |
| References | |
| 50-324-87-07, 50-324-87-7, 50-325-87-07, 50-325-87-7, NUDOCS 8704300062 | |
| Download: ML20209E493 (14) | |
Text
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UNITED STATES o
NUCLEAR REGULATORY COMMisslON o
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REGION li p
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101 MARIETTA STREET N.W.
ATLANTA, GEORGI A 30323
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APR 17 EG7 Report Nos. 50-325/87-07 and 50-324/87-07 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos.: 50-325, 50-324 License Nos.: DPR-71, DPR-62 Facility Name: Brunswick Steam Electric Plant 1 and.2
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Inspection Conducted: March 30-April 3,1987 Inspectors:
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W."B. Gloersen '
Date Signed
/MM lff/297//2:Wh F//Fb7 '
5. 5. Adamovitz v
D6te 5fgned Accompanying Per el:
G. L. Froemsdorf Approved by:
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N y//V/T7
.'Kahle, Section Chief Date Signed
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Di ision of Radiation Safety and Safeguards SUMMARY Scope:
This routine unannounced inspection involved an onsite examination in the areas of quality control and confirmatory measurements including a review of the laboratory quality control program; review of procedures and instructions; review of quality control records and logs; review of the counting room and chemistry laboratory facilities; and results of split samples analyzed by the licensee and the NRC Region II mobile laboratory.
Results: No violations or deviations were identified.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- P. W. Howe, Vice President-Brunswick Nuclear Plant
- C. R. Dietz, General Manager
- S. Strickland, Shift Operations Supervisor
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- E. R. Eckstein, Manager, Technical Support
- A. G. Cheatham, Manager, Environmental and Radiation Control (E&RC)
- C. E. Robertson, Supervisor, E&RC
- R. Queeder, Principal Engineer
- J. Davis, Principal Engineer
- K. E. Enzor, Director, Regulatory Compliance
- P. M. Paulk, Senior Specialist
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M. L. Millinor, Foreman, E&RC I
S. E. Fitzpatrick, Chemistry Technician C. A. Harris, Chemistry Technician T. Roeder, Chemistry Technician S. Watson, Senior Specialist, E&RC B. White, Senior Specialist, E&RC D. N. Allen, Project Engineer (0NS)
W. J. Dorman, QA Supervisor S. E. Thorndyke, Project Specialist NRC Resident Inspectors
- W. Ruland
- L. W. Garner
- Attended exit interview 2.
Exit Interview The inspection scope and findings were sumarized on April 3,1987, with those persons indicated in Paragraph 1 above. The inspector described the areas examined and discussed in detail the inspection findings listed below.
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.
3.
Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspection.
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4.
EffluentReports(84723,84724)
Technical Specification (TS) 6.9.1.8 requires the licensee to submit a Semiannual Radiological Effluent Release Report within the time periods specified in TS 6.9.1.9 and 6.9.1.12 cnvering the operation of the facility during the previous six months of operation.
The inspectors reviewed the Semiannual Radiological Effluent Release Report for the period July 1,1986 through December 31, 1986.
The review included an examination of any unplanned releases, the liquid and gaseous effluent release data trends, omissions, and obvious mistakes.
The data are summarized below for liquids and gases for calendar years 1985 and 1986:
Effluent Summary:
Brunswick Plant Gaseous Releases - Sunnation of all releases for Units 1 and 2 1985 1986 Fission and Activation gases (curies)
1.75 E+4 4.51 E+4 Iodines (curies)
4.01 E-2 1.46 E-2 Particulates (curies)
2.32 E-2 3.23 E-2 Gross Alpha (curies)
2.61 E-4 2.72 E-4 Tritium (curies)
3.82 E O 7.07 E 0 Abnormal Releases, nos./ quantities 2/2.67 E-2 1/1.20 E-5 Total plant releases of fission gas activity to the atmosphere for 1986 were approximately 2.26 E+4 curies per unit.
By comparison, approximately 2.5 tines less activity due to fission and activation gases per unit was released in 1985.
The comparable value for the average of 21 operating U.S. BWRs of greater than 500 MWe in 1982 was 4.85 E+4 curies per unit.
Liquid Releases - Summation of all releases for Units 1 and 2 1985 1986 Fission and Activation products (curies)
1.51 E-1 1.26 E-1 Tritium (curies)
9.88 E 0 5.78 E 0 Gross alpha (curies)
4.97 E-3 2.70 E-3 l
Abnormal Releases, nos./ quantities 1/2.10 E-4 none There was no significant change noted in liquid effluent releases in 1986 in that the total activity in liquid effluent due to mixed fission and activation products for the report period was 0.126 curies as compared to 0.151 curies in 1985.
Additionally, only 5.78 curies of tritium were released as compared to 9.88 curies in 1985.
By comparison, the annual releases from 21 operating U.S. BWRs of greater than 500 MWe for calendar year 1982 (last year for which summary data were available) showed liquid releases of 3.56 curies of mixed fission and activation products and 13.2 curies of tritium per yea'r.
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There was one abnormal gaseous release during 1986 which totaled 1.20 E-5 curies involving a failure of a clothes dryer thermostat causing 14 coveralls to catch fire.
The release estimate was based on assumptions that the coveralls were contaminated and that 10% of the contamination became airborne as Co-60.
The annual dose assessment for calendar year 1986 was provided in the July-December 1986 Semiannual Effluent Release Report.
All dose estimates were below the limits of 40 CFR 190.10
" Environmental Radiation Protection Standard for Nuclear Power Operations."
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No violations or deviations were identified.
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Audits (84725)
Technical Specification 6.5.5.2 requires audits of facility activities to be performed by the Performance Evaluation Unit (PEU) encompassing performance of facility operation to provisions contained within the
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Technical Specifications and applicable license conditions at least once per 12 months and the performance of activities required by the Quality Assurance Program to meet the provisions of Regulatory Guide 1.21, Revision 1, June 1974 and Regulatory Guide 4.1, Revision 1, April 1975 at least once per 12 months.
The inspectors reviewed three different types of self-appraisals p(erformed by the licensee.
Briefly, these self-appraisals included:
1)
audits-which were performed by the PEU as required by the Technical Specifications; (2) surveillances-which were performed by the Operations QA/QC onsite staff; and (3)
corporate health physics (HP)
assessments-which were performed periodically by the Corporate HP staff.
The inspectors reviewed the following reports:
QAA/0021-86-06, " Quality Assurance Audit of Brunswick Steam Electric Plant," August 19, 1986 QASR 86-026, " Surveillance Report - Reactor Coolant Chemistry,"
March 27, 1986 QASR 86-035, " Surveillance Report - Implementation of RETS," June 11, 1986 QASR 86-039, " Surveillance Report - Prestartup/Startup Chemistry,"
July 2, 1986 CHP-87-33, " Radiological Effluent Monitoring Assessment," March 27, 1987 The inspectors discussed the audits with licensee personnel and reviewed their corrective actions.
The inspectors noted that corrective actions were taken or being taken to resolve items of concern.
No violations or deviations were identified.
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6.
Procedures (84725)
Technical Specification 6.8.1 requires written procedures to be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, November 1972, and the Quality Assurance Program for effluent and environmental monitoring.
The inspectors reviewed selected portions of Environmental and Radiation Control (E&RC) procedures concerning radioactive sampling and analysis, standards preparation, equipment operation, calibration and performance checks, and determination of the lower limits of detection.
In reviewing Procedure E&RC 2205, " Operation of the Packard 4530 Scintillation Counter," Rev. 2, September 13, 1985, the inspectors noted that the form, Appendix D,
"Packard Liquid Scintillation Counter Efficiency Determination," did not include a date which made it difficult to track the instrument calibrations.
The procedure specified the frequency of the efficiency determinations, and the form, Appendix D, contained spaces for appropriate sign-offs.
The inspectors discussed the procedure with licensee representatives and determined that laboratory personnel were aware of the omission and that the procedure was currently under revision.
The inspectors also noted Procedure E&RC-2206, Radioactive Standards Preparation for Calibration of the ND6600 Multichannel Analyzer," Rev. 5, January 31, 1986, did not require comparison of newly determined efficiencies to previously determinated efficiencies for the same detector. Licensee representatives indicated the data were being compared and they would evaluate procedural changes to include the comparison.
No violations or deviations were identified.
7.
Records (84725)
The inspectors reviewed selected portions of the following records:
a.
Gamma Spectroscopy System Detector Nos. 1216, 1267, and 49RB, Quality Control Checks for March 1987 including:
(1) Energy Calibrations (2) Separate Source Verification Worksheet (3) FWHM Determinations (4 Reliability Checks (5 System Resolution (6 Efficiency Checks
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Tennelec LB5100 Proportional Counter Quality Control Records:
(1 Daily Quality Control Sheets, November 1986-March 1987 (2 Chi-square Determinations, December 1986-April 1987 (3 Alpha Absorption Curves, November 4,1986; January 8,1987 (4 Efficiency Checks, September 17, 1986; January 5, 1987; March 20, 1987 (5) Plateau Determinations, September 17, 1986; January 5,1987 c.
Packard Tri-Carb 4530 Quality Control Records for 1987 including:
(1) C-14 Daily Quality Control Sheets (2) H-3 Daily Quality Control Sheets (3) Chi-square Determinations (4) H-3 Efficiency Determinations d.
Annual (1986-1987) Gamma Spectroscopy Detector System Nos. 1216, 1267 and 49RB, Efficiency Calibrations for the following geometries:
(1) Charcoal Cartridge (2) Liquid Vial 43 mm Particulate Filter 500 milliter Liquid Marinelli One Liter Gas Marinelli (6) Gas Vial e.
Analytics Cross Check Program Quarterly Tests for 1986 including:
(1) Charcoal Cartridge - I-131 2)
14cc Gas Vial - Kr-85, Xe-133 3) Gross Beta
Gross Alpha Tritium Sr-89, Sr-90 and Fe-55 Particulate Filter - Mixed Gamma (8) Liquid Marinelli - Mixed Gamma f.
Semiannual Intralaboratory Quality Control, per Procedure E&RC 1700, Rev.
5, Appendix C Nuclear Counting Spiked Samples for three laboratory technicians during 1985 and 1986 including the following determinations:
(1) Tritium (2) Gross Alpha (3) Gamma Isotopic - Liquid, Particulate Filter and Charcoal Cartridge g.
Semiannual Round Robin Radiochemistry Cross Check Program for 1986.
Results of the record review were discussed with cognizant licensee representative.
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No violations or deviations were identified.
8.
Laboratory Quality Assurance Program (84725)
The inspectors reviewed the licensee's quality assurance program for radiochemical measurements using the guidance contained in Regulatory Guide 4.15 (Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment - February 1979).
The chemistry organization consisted of four sections including: Units 1 and 2 Wet Chemistry, Environmental Chemistry, and Count Room Chemistry.
Each section was supervised by a Foreman and the four sections' staff totalled 28 Chemistry Technicians.
Section Foremen were supervised by a Chemistry Supervisor who reported to the E&RC Manager.
The E8RC Manager reported directly to the Plant Manager. The various position descriptions were detailed in administratively controlled documents.
The inspectors noted that the training program for the Chemistry Technicians had been evaluated by the Institute of Nuclear Power Operations (INP0).
Quality control records for laboratory counting systems included the results from measurements of radioactive check sources, calibration sources, backgrounds, and blanks. Daily QC logs and plots were maintained and kept near the appropriate laboratory counting systems. The inspectors noted that records were well organized and easily accessible.
The licensee's count room quality control program included the use of NBS traceable reference standards to determine counting efficiencies for specific radionuclides and to determine counting efficiency as a function of gamma-ray energy for gamma-ray spectrometry systems.
The inspectors also noted that the licensee routinely analyzed inter-and intralaboratory crosscheck samples. The interlaboratory program included participation in a vendor Crosscheck Program and a Round-Robin Radiochemistry Crosscheck Program.
The vendor program consisted of a set of eight spiked unknowns including various geometries and isotopee issued quarterly.
The Round Robin test was a spiked liquid issued semi-annually and the results from the Brunswick, Harris, and Robinson facilities were compared.
The licensee had also established an intralaboratory comparison program for the chemistry technicians.
Various spiked unkowns were analyzed semi-annually by the technicians.
The results were reviewed and subsequent followup actions were defined if the results were not within specified limits.
No violations or deviations were identified.
9.
Confirmatory Measurements (84725)
During the inspection, reactor coolant and selected liquid and gaseous plant effluent process streams were sampled and the resultant sample matrices were analyzed for radionuclide concentrations using licensee and NRC Region II Mobile Laboratory gamma-ray spectroscopy systems.
The purpose of these comparative measurements was te verify the licensee's
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capability to measure radionuclides accurately in various plant systems.
Analyses were conducted utilizing as many of the licensee's gamma spectroscopy systems as practicable. Sample types and counting geometries included the following:
drywell gas - 1.26 liter gas Marinelli, pretreatment offgas - 14cc vial, reactor coolant - 14cc liquid vial, and fuel pool coolant - 1 liter liquid Marinelli.
Spiked particulate filter and charcoal cartridge samples were provided for analysis in lieu of licensee samples which did not have sufficient levels of activity for analysis.
Comparison of licensee and NRC results are listed in Table 1 with the acceptance criteria listed in Attachment 1.
Results were in agreement for all sample types analyzed.
No violations or deviations were identified.
10. Utility Services for the NRC Mobile Laboratory during Emergencies (92705)
The inspectors discussed utility service specifications and parking arrangements for the NRC Region II Mobile Laboratory that would be used during an emergency situation at the Brunswick Station.
Since the licensee utilizes on onsite Emergency Operations Facility, it would not be feasible to locate the mobile laboratory at this facility during an emergency involving radiological releases.
It was noted that the licensee and the State of North Carolina utilize the power supply at the Brunswick County Government Center in Bolivia, NC for their mobile laboratories during emergency situations.
The inspectors discussed with the licensee the need to inquire about arrangements that would be necessary for the NRC mobile laboratory.
The specifications for electrical power to the mobile laboratory were three single phase lines with separate circuit breakers at 120 volts and 30 amps each within 25 feet of the mobile laboratory. The receptacles should be compatible with 120 volt, 30 amp three prong twist lock plugs. The parking area should accommodate a vehicle with dimensions of 30 feet long, 8 feet wide, 14 feet high.
The site should be surfaced and level.
Additionally, one standard telephone outlet within 25 feet of the mobile laboratory should be provided.
11. Environmental Monitoring (80721)
During an inspection conducted September 15-19,1986, (IE Report Nos.
50-325/86-26 and 50-324/86-27), the licensee was requested to collect, split, and send a liquid environmental sample to the NRC Region II office.
It was agreed that this sample would be collected during the monthly collection period when the samples are normally split with the State of North Carolina at the normal sample location downstream of the plant discharge.
The licensee transmitted the results of the isotopic analysis of the water sample to the NRC Region II office in a letter dated January 22, 1987, for comparison purposes.
Basically, the NRC and licensee analytical results were in agreement in that no detectable radioactivity due to plant effluents were noted in the liquid sample. The inspectors had no further questions.
No violations or deviations were identified.
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12.
Information Notices (92703)
l The inspectors reviewed IE Information Notices (IEN) 86-42, " Improper Maintenance of Radiation Monitoring Systems" and 86-76, " Problems Noted in Control Room Emergency Ventilation Systems."
It was determined that the IENs had been received, reviewed, distributed to the cognizant engineer, and that actions were taken or being planned.
TEN 86-42 was issued to alert licensees to the potential for defeating the safety function associated with radiation monitoring systems b' not properly adhering to established surveillance and maintenance procedures. The licensee has had in place administrative controls and programs which addressed the corrective actions listed in the IEN.
'.EN 86-76 was issued to alert licensees to problems noted in the operation of Control Room emergency ventilation systems during an onsite NRC Control Room habitability team appraisal.
The licensee has designated an individual to be responsible for Control Room habitability and ventilation systems.
This individual was resolving the various problems that have been identified.
The inspectors had no further questions in this area.
13. Licensee Action on Previously Identified Inspector Followup Items (92701)
(0 pen) 50-324/85-12-01 and 50-325/85-12-01:
Inoperable condition of
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hydrogen gas monitoring instruments in Augmented Offgas (A0G) system to be
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corrected and instruments returned to service.
At the time of this inspection, the hydrogen monitoring instruments in the A0G system were
still inoperable. The inspectors noted that the licensee had modified the Unit 2 hydrogen monitoring system by rerouting the tubing and installing
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i new filters to minimize the water moisture in the system which had l
l resulted in plugging up the flow switch.
Although it appeared that the
new tubing arrangements and filters allowed the monitors to function properly, they were still considered inoperable due to electrical i
l problems. The licensee had established an April 30, 1988, completion date
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with its Regulatory Compliance Department for this monitoring system.
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should be noted that these hydrogen gas monitoring system modifications
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have been tied in with the hydrogen water chemistry modification.
This i
item remains open.
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(Closed) 50-324/86-27-01 and 50-325/86-26-01:
Revise E&RC Procedure 270 l
to include provisions for testing fixed HEPA filtration units.
This procedure was revised so that the reference to limit the procedure to i
l portable HEPA filtration units was eliminated. Consequently, the revised procedure could be used to test the various types of HEPA filter units l
used by'the Health Physict Department. This item is considered closed.
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(Closed) 50-324/86-27-02 and 50-325/86-26-02:
Review charcoal adsorber l
laboratory test criteria for the Control Room emergency filtration system.
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The licensee reviewed the criteria used for testing the charcoal samples i
from the Control Room emergency filtration system as it related to l
ASTM D3803, Regulatory Guide 1.52, and the Technical Specifications.
l Presently, the licensee's Technical Specifications require the charcoal l
samples to be tested at 80 C and 70% relative humidity.
Although the l
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retention capabilities of the carbon bed decreases as moisture increases and temperature decreases, the licensee did not indicate a need to test the carbon samples under these conditions.
According to Regulatory Guide 1.52 typical accident conditions for secondary systems such as the Control Room emergency air filtration system could experience maximum influent temperatures of 180 F (82 C) and relative humidities of 100%.
The licensee indicated that the Control Room influent temperatures under accident conditions would be lower than 180 F.
It should be noted that Regulatory Guide 1.52 provides no acceptance criteria for the radioiodine retention test conducted at lower temperatures and higher relative humidities (e.g., 30 C, 95% relative humidity).
The licensee's response appeared to be adequate and the inspectors had no further questions. This
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i item is considered closed.
(Closed) 50-324/86-27-03 and 50-325/86-26-03:
Evaluate the Offsite Dose Calculation Manual (0DCM) to consider a revision to change the receptor location for the hypothetical cow-milk pathway.
The inspectors reviewed the licensee's response and noted that the licensee had discussed this issue with the Office of Nuclear Reactor Regulation (NRR).
Briefly, NRR reviewed the Brunswick ODCM, Revision 6 and noted that the document used approved methods that were consistent with the methodology and guidelines in NUREG-0133. NRR concluded that the ODCM, Revision 6, was an acceptable reference to the Brunswick Technical Specifications.
The NRR evaluation of the Brunswick ODCM, Revision 6 was transmitted to the licensee in a letter dated March 10, 1987.
The inspectors had no further questions.
This item is considered closed.
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TABLE 1
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RESULTS OF GAMMA SPECTROSCOPY CONFIRMATORY MEASUREMENTS AT BRUNSWICK STEAM ELECTRIC PLANT, MARCH 30-APRIL 3, 1987 r
SAMPLE ISOTOPE C0_NCENTRATION fuCi/ uni Q RESOLUTION RAT 10 COM PAR 1 SO'A (Geometry)
Licensee
.N, R.Q Licensee /NRC
Cha rcoa l Cartridge Spike Co-60 4.94 E-2 5.10 i.06 E-2
.97 Ag reement a.
Detector 1267 Hg-203 7.71 E-2 7.94 i.75 E-2
.97 Ag reement Co-57 3.86 E-2 3.91
.03 E-2 130
.99 Ag reement Y-88 1.00 E-1 1.04 i.03 E-1
.96 Ag reemen t Cs-137 6.04 E-2 5.95 i.05 E-2 119 1.02 Ag reement Cd-109 1.83 E O 1.86 i.01 EO 186
.98 Ag reement Ce-139 3.80 E-2 3.83 i.06 E-2
.99 Ag reement Sn-113 7.70 E-2 7.89 1.16 E-2
.98 Ag reement b.
Detector 49RB Co-60 4.95 E-2 5.10 i.06 E-2
.97 Ag reement Hg-203 8.25 E-2 7.94 i.75 E-2
1.04 Ag reement Co-57 3.88 E-2 3.91 i.03 E-2 130
.99 Ag reement Y-88 9.83 E-2 1.04 i.03 E-1
.95 Ag reemen t Cs-137 5.95 E-2 5.95 i.05 E-2 119 1.00 Ag reement Cd-109 1.82 E O 1.86 i.01 E-2 186
.98 Ag reement Ce-139 3.73 E-2 3.83 i.06 E-2
.97 Agreement SN-113 7.78 E-2 7.89 1.16 E-2
.99 Ag reement c.
Detector 1216 Co-60 5.06 E-2 5.10 i.06 E-2
.99 Ag reement Hg-203 8.34 E-2 7.94 i.75 E-2
1.05 Ag reemen t Co-57 3.98 E-2 3.91 i.03 E-2 130 1.02 Ag reemen t Y-88 1.01 E-1 1.04 i.03 E-1
.97 Ag reemen t Cs-137 6.03 E-2 5.95 i.05 E-2 119 1.01 Ag reemen t Cd-109 1.84 E O 1.86 i.01 E-2 186
.99 Agreement Ce-139 3.79 E-2 3.83 i.06 E-2
.99 Ag reement Sn-113 7.70 E-2 7.89 i.16 E-2
.98 Ag reemen t 2.
Particulate filter Spike Co-60 8.35 E-3 8.72 i.14 E-3
.96 Ag reemen t a.
Detector 1216 Cs-137 1.16 E-2 1.19 i.11 E-2 119
.97 Ag reement b.
Detector 49RB Co-60 8.43 E-3 8.72 i.14 E-3
.97 A9 rcemen t Cs-137 1.18 E-2 1.19 i.11 E-2 119
.99 Ag reement c.
Detector 1267 Co-60 8.64 E-3 8.72 i.14 E-3
.99 Ag reemen t Cs-137 1.18 E-2 1.19 i.11 E-2 119
.99 Ag reement 3.
Orywel8 (1 Iiter gas ma ri ne l l i )
Xe-133 4.55 E-7 4.73 i.64 E-7
.96 Ag reemen t a.
Detector 1267 Xe-135 2.11 E-6 2.32 i.06 E-6
.91 Ag reement b.
Detector 1216 Xe-133 3.90 E-7 4.73 i.64 E-7
.82 Ag reemen t Xe-135 2.07 E-6 E 12 i.06 E-6
.90 Ag reement c.
Detector 49RB Xe-133 4.78 E-7 4.7J i.64 E-7
1.01 Ag reement Xe-135 2.06 E-6 2.J2 i.06 E-7
.89 Ag reemen t
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TABLE 1 (cont'd)
RESULTS OF CAMMA SPECTROSCOPY CONFIRMATORY MEASUREMENTS AT BRUNSWICK STEAM ELECTRIC PLANT, MARCH 30-APRIL 3, 1987 SAMPtE ISOTOPE CONCENTRATION fuCi/unill RESOLUT10N RAT 10 COMPARISON (Ceometry)
Licensee MRfC Licensee /NRC 4.
Offgas Pretreatment (14cc vial)
Xe-135 4.94 E-2 5.57 i.05 E-2 111
.89 Ag reement a.
Detector 1267 Xe-135m 2.28 E-1 2.49 i.06 E-1
.92 Ag reement Xe-138 8.99 E-1 1.03 i.02 E O
.87 Agreement Kr-85m 8.82 E-3 9.54 i.23 E-3
.92 Ag reement Kr-88 2.97 E-2 3.84 i.82 E-2
.89 Ag reement Kr-87 6.47 E-2 6.04 i.12 E-2
1.07 Ag reement b.
Detector 49RB Xe-135 4.84 E-2 5.57 i.05 E-2 111
.87 Ag reemen t Xe-135m 2.32 E-1 2.49 i.06 E-1
.93 Ag reement Xe-138 9.02 E-1 1.03 i.02 E O
.88 Ag reement K r-85m 7.68 E-3 9.54 i.23 E-3
.81 Ag reement K r-88 2.98 E-2 3.34 i.82 E-2
.89 Ag reemen t Kr-87 5.77 E-2 6.04 i.12 E-2
.96 Ag reement c.
Detector 1216 Xe-135 4.94 E-2 5.57 i.05 E-2 111
.89 Ag reement Xe-135m 2.41 E-1 2.49 i.06 E-1
.97 Ag reement Xe-138 9.04 E-1 1.03 i.02 E O
.88 Ag reement K r-85m 8.67 E-3 9.54 i.23 E-3
.91 Ag reement Kr-88 2.93 E-2 3.34 i.82 E-2
.88 Agreement Kr-87 5.15 E-2 6.04 i.12 E-2
.85 Ag reement 5.
Reactor Coolant (14cc liquid vial)
1-131 3.11 E-4 2.58 i.25 E-4
1.21 Ag reement a.
Detector 1267 l-132 3.12 E-3 2.96 i.05 E-3
1.05 Ag reement 1-133 3.62 E-3 3.50 i.04 E-3
1.03 Ag reemen t 1-135 6.84 E-3 6.53 i.16 E-3
1.05 Agreement Sr-91 1,43 L-3 1,42 i.11 E-3
1.01 Ag reement -
S r-92 1.51 E-3 1.49 i.06 E-3
1.01 Ag reement Tc-99m 1.44 E-2 1,32 i.01 E-2 132
'1.09 Ag reement Cr-51 1.01 E-2 9.52 i.19 E-3
1.06 Ag reement As-76 9.87 E-4 8.84 i.43 E-4
1.12 Ag reement b.
Detector 1216 l-131 2.76 E-4 2.58 i.25 E-4
1.07 Ag reement 1-132 3.02 E-3 2.96 i.05 E-3
1.02 Ag reement 1-133 3.62 E-3 3.50 t.04 E-3
1.03 Ag reement 1-135 6.87 E-3 6.53 i.16 E-3
1.05 Ag reemen t S r-91 1.59 E-3 1.42 i.11 E-3
1.12 Ag reement S r-92 1.63 E-3 1,49 i.06 E-3
1.09 Ag reement Tc-99m 1.38 E-2 1.32 i.01 E-2 132 1.05 Ag reement C r-51 1.01 E-2 9.52 i.19 E-3
1.06 Ag reement A3-76 9.03 E-4 8.84 i.43 E-4
1.02 Ag reement
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TABLE 1 (cont'd)
RESULTS OF GAMMA SPECTROSCOPY CONFlRMATORY MEASUREMENTS AT BRUNSWICK STEAM ELECTRIC PLANT, MARCH 30-APRIL 3, 1987 SAMPLE ISOTOPE CONCENTRATION fuCi/ unit 1 RESOLUTION RATlO COMPARISON (Geometry)
Licensee E
Licensee /NRC c.
Detector 49RB l-131 2.74 E-4 2.58 i.25 E-4
1.06 Ag reement I-132 3.04 E-3 2.96 i.05 E-3
1.03 Ag reement I-133 3.54 E-3 3.50 i.04 E-3
1.01 Ag reemen t 1-135 6.77 E-3 6.53 i.16 E-3
1.04 Ag reement S r-91 1.60 E-3 1.42 i.11 E-3
1.13 Ag reement Sr-92 1.56 E-3 1.49 i.06 E-3
1.05 Ag reement Tc-99m 1.38 E-2 1.32 i.01 E-2 132 1.04 Ag reement Cr-51 9.31 E-3 9.52 i.19 E-3
.98 Agreement As-76 9.21 E-4 8.84 i.43 E-4
1.04 Ag reement 6.
Fuel Pool Coolant (1 liter liquid ma rinelli)
Mn-54 7.24 E-5 7.50 i.05 E-5 150
.97 Ag reement a.
Detector 49RB Fe-59 1.15 E-5 1.11 i.07 E-5
1.04 Ag reemen t Co-58 2.43 E-5 2.39 i.04 E-5
1.02 Ag reemen t Co-60 1.77 E-4 1.84 i.01 E-4 184
.96 Ag reemen t Cs-137 6.82 E-6 6.98 i.25 E-6
.98 Ag reement b.
Detector 1216 Mn-54 7.33 E-5 7.50 i.05 E-5 150
.98 Ag reement Fe-59 1.24 E-5 1.11 i.07 E-5
1.12 Ag reement Co-58 2.34 E-5 2.39 i.04 E-5
.98 Ag reement Co-60 1.78 E-4 1.84 i.01 E-4 184
.97 Ag reement Cs-137 7.13 E-6 6.98 i.25 E-6
1.02 Ag reement c.
Detector 1267 Mn-54 7.69 E-5 7.50 i.05 E-5 150 1.03 Ag reemen t Fe-59 1.14 E-5 1.11 i.07 E-5
1.03 Ag reemen t Co-58 2.49 E-5 2.39 i.04 E-5
1.04 Ag reement Co-60 1.85 E-4 1.84 i.01 E-4 184 1.01 Ag reement Cs-137 6.99 E-6 6.98 i.25 E-6
1.00 Ag reement
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ATTACHMENT 1 CRITERIA FOR COMPARING ANALYTICAL MEASUREMENTS
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s This enclosure 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.
In these 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 dif ferences between the NRC and licensee values should be more ra trictive. Conversely, poorer agreement between NRC and licensee values must be considered acceptable as the resolution decreases.
2 of the licensee value to the NRC value for each For comparison purposes, a ratio individual nuclide is computed. This ratio is then evaluated for agreement based on the calculated resolution. The corresponding resolution and calculated ratios which denote agreement are listed in Table 1 below.
Values - outside of the agreement ratios for a selected nuclide are considered in disagreement.
NRC Reference Value for a Particular Nuclide 8 Resolution =
Associated Uncertainty for the Value
'
..
,
Licensee Value
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8 Comparison Ratio = NRC' Reference Value
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TABLE 1 - Confirmatory Measurements Acceptance Criteria Resolutions vs. Comparison Ratio Comparison Ratio for Resolution Agreement
<4 0.4 - 2.5 4-7 0.5 - 2.0 8 - 15 0.6 - 1.66 16 - 50 0.75 - 1.33 51 - 200 0.80 - 1.25
>200 0.85 - 1.18 z=-