ML20137R373
| ML20137R373 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/16/1985 |
| From: | Cline W, Gloersen W, Golersen W, Harris J, Krzo G, Kuzo G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20137R365 | List: |
| References | |
| 50-413-85-50, 50-414-85-60, NUDOCS 8602130094 | |
| Download: ML20137R373 (16) | |
See also: IR 05000413/1985050
Text
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET N.W..
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ATLANTA, GEORGI A 30323
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JAN 311986 ~
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Report Nos.:
50-413/85-50 and 50-414/85-60
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Licensee: Duke Power Company. _
422 South Church Street
Charlotte, NC -28242 .
Docket Nos.:
50-413 and 50-414
License Nos.:i NPF-35 and CPPR-117
Facility Name:
Catawba.1 and 2
Inspection Conducted:
December 9-13, 1985
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Inspectors:
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J. D. Harris
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Accompanying Per
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Approved by:
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K E. Cline,LSection Chief
Date Sidned
Emergency Preparedness and Radiological
Protection Branch
Division of Radiation Safety and Safeguards
SUMARY
Scope:
This routine, unannounced inspection iiivolved 115 inspector-hours onsite
in the areas of quality control-and confirmatory measurements including review of
the laboratory quality control program; review ~of procedures and instructions;
review of quality control records and logs; review of counting room and chemistry
laboratory facilities; results of split samples analyzed.by the licensee and the
NRC Region II Mobile ' Laboratory;-.and whole-body counter measurements using a
. fission product phantom.
Results:
One violation was identified - failure to have adequate procedures to
implement the Quality Assurance Program for effluent measurements.
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8602130094 060131
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REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- H. B. Barron, Superintendent of Operations
W. P. Deal, Station Health Physicist
- G. T. Mode, Health Physics Coordinator
- C. V. Wray, Health Physics Coordinator
G. L. Courtney, Staff Health Physicist
- M. J. Geer, Assistant Engineer, Corporate
- B. Chundtlik, Health Physics Supervisor
- C. L. Hartzell, Licensing and Project Engineer
- P. G. LeRoy, Licensing Engineer
Other licensee employees contacted included engineers, technicians, and
office personnel.
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NRC Resident Inspector
- P. K. Vanuocrn
" Attended exit interview
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2.
Exit Interview
The inspection scope and findings were summarized on December 13, 1985, with
those persons indicated in Paragraph 1 above. The inspector noted that the
temporary power supply provided by the licensee for th'e NRC Region II Mobile
was inadequate and discussed the need for a permanent power supply to be
established in a timely manner.
The inspector discussed the following
inspector followup items:
updating of counting room procedures
(Paragraph 6.b); evaluation of alpha / beta and beta / gamma analysis
instrumentation performance biases (Paragraph 7.b); resolution of Fe-55
analysis methodology and verification analyses (Paragraph 8.b); and review
of calibration and quality control data for the whole-body counting
instrumentation .(Paragraph 9.b).
One violation concerning inadequate QA
procedures (Paragraph 6.c) for effluent measurement instrumentation was
discussed.
Licensee ~ representatives acknowledged the inspectors' comments
and committed to completing resolution of Fe-55 analysis concerns and
requesting an additional spiked sample from the NRC Region II Office by
March 31, 1986, to demonstrate Fe-55 analysis capability.
The licensee did
not _ identify as proprietary any of the materials provided to or reviewed by
the-inspectors during this inspection.
3.
Licensee Action on Previous Enforcement Matters
This subject was not addressed in the inspection.
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4.
Laboratory Quality Control Program (84725)
The inspectors noted that the licensee's Quality Assurance (QA) program for
effluent measurements had not changed since previous inspections
(50-413/84-42, 50-413/84-64, .50-413/84-76). _The inspectors reviewed and
discussed with cognizant licensee representatives concerns regarding quality
control (QC) procedures and records, and results of interlaboratory
comparisons. for radiological effluent measurements conducted by the Catawba
Nuclear Station (Paragraphs 6 and 7).
The inspectors further noted that
results of the split sample analyses between the licensee and the.NRC
Region II Mobile Laboratory gamma spectroscopy systems (Paragraph 8.a)
identified concerns regarding the adequacy of the effluent measurements QA
program to ensure accurate effluent measurement capability.
Details of
these concerns are discussed in Paragraphs 6.b, 6.c, and 8.
No violations or deviations were identified.
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5.
Audits and Reviews (84725)
a.
Technical Specification (TS) 6.5.2.9 states that audits of unit
activities shall be performed under the cognizance of the Nuclear
Safety Review Board (NSRB) encompassing:
the conformance of unit
operation to provisions contained within the Technical Specifications
and applicable license conditions at least once per 12 months; the
radiological environmental monitoring program and the results thereof
at least once per 12 months; the-Offsite Dose Calculation Manual and
implementing procedures at least once per 12 months; and the
performance of activities required by the Quality Assurance Program to
meet the criteria of Regulatory Guide 4.15, December 1977 at least once
per 12 months.
The inspectors reviewed the following audit report:
QA Audit NP-85-89(CN) Technical Services and Operations Activities,
April 15 - May 1, 1985.
The inspectors noted that the counting room and body burden analysis
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programs were audited against- Technical Specifications, Regulatory
Guide 4.15, and Catawba Station Health Physics Manual.
One concern
regarding the inability to verify algorithms used in the Ortec ADCM
Analysis Systems was identified.
The inspectors discussed and reviewed
licensee action regarding this item.
Cognizant licensee personnel
presently are developing computer software which will allow review and
verification of algorithms utilized for each separate system.
b.
The inspectors discussed the following technical reviews completed by
licensee personnel:
(1) . Quality Assurance, Teledyne Isotopes Technical Review, GS-750.00,
110.70, 10/24/85.
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(2) Catawba Nuclear Station Radioanalysis Program Review, GS-750-05,
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768, 11/26/85.
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The inspectors noted that several technical items concerning
radiological measurement capability and QA were identified in the
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reviews.
Identified items included inaccurate Fe-55 measurem mts,
improvement of instrument performance acceptance criteria, initiation -
of intralaboratory crosscheck program, Lower Limit of Detection (LLD)
capabilities, updating of procedures, improvement in gamma spectroscopy
efficiency graphs, and quality control performance analyses.
The
inspectors noted that licensee action regarding Fe-55 analysis concerns
was being finalized (Paragraph 8.b) and responses to and/or action
concerning the other identified items were in progress.
No violations or deviations were identified.
6.
Procedures (84725)
a.
Technical Specification (TS) 6.8.1 states written procedures shall be
established, implemented and maintained covering the applicable
procedures recommended in Appendix A of Regulatory Guide 1.33,
Revision 2,
February 1978,
Offsite
Dose Calculation Manual
implementation, and the Quality Assurance Program for effluent and
environmental monitoring.
The following selected procedures were
reviewed by the inspector.
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(1) HP/0/8/1001/07 Operation and Calibration of Liquid Scintillation
Counters, Rev. 6, 11/21/84.
(2) HP/0/8/1001/09 Operating / Calibration Procedure for Body Burden
Analyzer, Rev. 4, 3.16.84.
(3) HP/0/8/1001/12 Technical Specifications Gaseous Waste Sampling and
Analysis, Rev. 11, 12/3/85.
(4) HP/0/8/1001/15 Preparation of Sources, Rev. 2, 8/24/84.
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(5) HP/0/8/1001/16 Operation and Calibration:
ORTEC ADCAM GAMMA
ANALYSIS System, Rev. 2, 7/5/84.
(6) HP/0/8/1001/21 Operation ' and Calibration:
Canberra Fastscan
- Body Burden Analysis, Rev. O, 10/22/85.
The inspector discussed results of the procedure review with cognizant
licensee representatives as detailed in Paragraphs 6.b.
.c.
b.
The inspectors discussed the need for improved review and update of
radiochemical analytical procedures.
For example, selected procedures
(HP/0/8/1001/12, HP/0/8/1001/15) detailed a volume of 4400 cc for the
marinelli beaker geometry utilized to monitor gaseous effluents whereas
the actual volume is 4600 cc.
The inspectors verified that the correct
volume was used for effluent sample measurements.
The inspectors noted
that all procedures should be periodically reviewed and updated by
cognizant licensee personnel as part of the QA program.
The inspectors
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informed licensee representatives that a review of radiological
analytical procedures for effluent measurement methodology and changes
in QA procedures to ensure adequate analytical nuclide measurement
capabilities would be considered an open item to be reviewed during a
subsequent inspection (50-413/85-50-01, 50-414/85-60-01).
The inspectors noted that QC procedures for the gamma spectroscopy
c.
. systems required review of daily background, performance checks, energy
calibration, and Full Width Half-Maximum (FWHM) results by cognizant
count room personnel.
Procedures required only performance checks to
be graphed using trend charts.
The inspectors discussed with cognizant
licensee representatives the need for improved review and trending of
the QC data collected.
Furthermore, the inspectors noted that QA
procedures to maintain consistency and accuracy of analytical
measurements and minimize differences observed between detectors, i.e.,
standardization of plotted efficiency graphs for all geometries for
each detector system, verification of efficiency calibration algorithms
for all detectors, and the implementation of intralaboratory. and
interlaboratory crosscheck programs were not adequate.
The inspectors
informed licensee representatives that the inaccurate licensee
measurements noted for the confirmatory measurements split sample
analyses (Paragraph 8.a) may have been identified by the licensee with
an adequate QA program. An adequate program should include a tnorough
review and evaluation of all analytical QC trend data, and proper
technical review of intralaboratory and interlaboratory crosscheck
results for trends and anomalous data.
Furthermore, criteria for
initiating action regarding trends and anomalous data should be
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established.
The inspectors informed licensee representatives that
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10 CFR 20.201(b) requires the licensee to make surveys that are
reasonable under the circumstance to evaluate radiation hazards that
may be present and TS 6.0.1.g requires procedures to be established,
implemented and maintained for QA Program implementation for effluent
and environmental monitoring.
The inspectors informed licensee
representatives that failure to have adequate procedures for QA Program
implementation to ensure accurate effluent measurements as demonstrated
by the inaccuracies noted in the licensee's gamma spectroscopy
measurements
(Paragraph 8.a)
was
considered
a
violation
(50-413/85-50-02),50-414/85-60-02).
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Open Item (50-413/85-50-01, 50-414/85-60-01) - review of effluent
measurement procedures for inconsistencies, improved methodology, and
improvement in QA Programs to ensure adequate effluent measurements.
Violation (50-413/85-50-02, 50-414/85-60-02) - failure to have adequate
QA procedures to implement QA programs for effluent measurements.
7.
Review of Records and Logs (84725)
a.
The inspector reviewed selected portions of the following records and
logs:
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(1) LS 1800 Liquid Scintillation Counter Nos. CHP 1127-CE and
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CHP 1128-CE Instrument Quality Control
Logs for August
- December 1985 including:
~(a) Daily Background & Performance (H-3, Gross Beta) Log Sheets
(b) Monthly Performance Data & Charts
(c) Weekly H-3 Alignment Log Sheet
(di Monthly Background Log Sheet
(e) Monthly Performance Data Sheet
(f) Quarterly Quenched H-3 Standard Preparation Data
'(2) Alpha / Beta Proportional Counter Nos. CHP-1131-CE and CHP1133-CE
Quality Control Logs for January - December 1985 including:-
(a) Daily Background and Alpha / Beta Performance Check
-(b)- Alpha / Beta Check Source Data Sheet
(c) Efficiency & Monthly Performance Worksheet
(d) Efficiency & Monthly Performance Value Determination
(3) Manual APC Beta / Gamma Counter Nos. CHP-1116, CHP-1117 CHP-1118 and
CHP-1119 Quality Control Logs for January - December 1985
including:
(a) Daily Performance and Background Data Sheets and Graphs
(b) Check Source Data Sheets
(c) Efficiency and Monthly Performance Worksheet
(4) Ge(Li) Detector System Nos. 24-N-1096, 24-P-94VC, . 24-P-92-VA and
24-P-04TC Quality Control Logs- for August - December 1985
including:
(a) Daily Performance Checks
(b) . Performance Graphs
(c) Daily Background Data
(d) Daily Energy Calibration & FWHM Data
(5) Annual (1985) Gamma Spectroscopy Efficiency Calibration data for
Detector Nos. 24-N-1096, 24-P-94VC, 24-P-92VA, and 24-P-04TC
including the following geometries:
50 ml bottle, 1000 & 3500 ml
liquid marinelli beaker, 2" filter paper, face-loaded charcoal
cartridge, 12 cc gas vial, 100 cc gas bomb, and 4600 gas marinelli
beaker.
(6) Interlaboratory
Crosscheck
Results,
December 1984
- September 1985.
(7) Daily Quality Control Checks and Performance Plots for the
Canberra Fastscan Whole-body Counter, October - December 1985.
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(8) Annual (1985) Efficiency Checks and Performance Plots for Canberra
Fastscan Whole-body Counter, August - Septen;ber 1985 for the
following geometries:
lungs, GI tract, and thyroid.
Results of the record review are discussed in Paragraphs 7.b
.c.
b.
The inspectors discussed systematic biases noted for performance checks
of all alpha / beta and beta / gamma counting systems.
The inspectors
noted that the observed biases were not correlated with the following:
changeout of counting gas, contamination problems, location (systems
are located in different rooms), and technician errors.
The inspectors
noted that biases for background counts were not observed.
Cognizant
licensee representatives presently are investigating the performance
data and have detailed their concerns to an approved vendor.
The
inspectors informed licensee representatives that resolution- of this
problem would be considered an inspector followup item and would be~
reviewed
during
a
subsequent
inspection
(50-413/85-50-03,
50-414/85-60-03).
c.
The results of the 1985 interlaboratory crosscheck program were
discussed with cognizant licensee representatives.
The inspectors
noted that gamma spectroscopy results for liquid samples met
established licensee acceptance criteria and were~ consistent among the
station's gamma spectroscopy systems.
However, the inspectors noted
that results for a July 1985, charcoal cartridge crosscheck met the
licensee's acceptance criteria but showed a maximum difference of 26%
between detectors.
The inspectors noted that this large difference was
not observed for a January 1985 charcoal cartridge crosscheck sample
and informed licensee representatives that the July 1985 difference
among detectors indicated improper efficiency calibrations.
The
inspectors informed licensee representatives that the failure to
identify and resolve the noted differences was indicative of poor
Quality Control (QC) review for effluent measurement instrumentation.
Furthermore resolution of these differences may have identified
licensee measurement problems noted for the confirmatory measurement
samples (Paragraph 8.a).
Licensee representatives informed the
inspectors that procedures for an improved QA program for effluent
measurements capability, including an intraloboratory crosscheck
program were in progress.
The inspector informed licensee
reprer 'tatives that these procedures would be reviewed as part of the
inspe _
followup item regarding procedure review discussed in
Paragraph 6.b.
Open Item (50-413/85-50-03, 50-414/85-60-03) - review of licensee's
evaluation of alpha / beta and beta / gamma performance check biases.
No violations or deviations were identified.
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8.
Confirmatory Measurements (84725)
a.
During the inspection, reactor coolant and selected liquid and gaseous
plant effluent process streams were sampled and the resultant sample
matrices analyzed for radionuclide concentrations using licensee and
NRC Region II Laboratory gamma-ray spectroscopy systems.
The purpose
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of these comparative measurements was to verify the licensee's
capability to measure radionuclides accurately in various plant
systems.
Analyses were conducted utilizing as many of the licensee's
gamaa spectroscopy systems as practicable.
Sample types and counting
geometries included the following:
Reactor Coolant System (RCS) sample
- 50 ml polybottle; Liquid Waste - 3500 ml marinelli beaker; RCS Gas
sample - 12 cc vial; Waste Gas Decay Tank Sample - 100 cc bomb; and
Effluent Gas Grab Sample 4600 cc marinelli beaker.
Spiked particulate
filter and charcoal cartridge sample types were provided for analyses
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 and no significant trends in comparisons were
noted in liquid waste, particulate filter and the 100 cc gas bomb
geometries.
Concerns regarding the RCS, charcoal filter; and selected
gas sample geometries were discussed with licensee representatives as
noted below.
(1) For the reactor coolant sample, Co-58, a major plant contaminant,
was in disagreement for all detectors, with concentrations
underestimated by 27 to 39%.
Also, I-131 was in disagreement (29%
underestimate) for Detector No. 24-P-94VC.
The inspectors noted
that these differences were not identified for the liquid waste
geometry which also contained these isotopes, ' thus the observed
differences most likely resulted from inaccurate efficiency
calibration of this geometry.
Furthermore, the inspectors noted
that the 36% difference between the licensee's detectors for RCS
I-131 analyses was considered inadequate for the quantitative
measurements conducted with these detector systems.
(2) For the charcoal cartridge sample geometry, Ba-133 results.were in
disagreement for Detector No. 24-P-92-VA and overall results were
biased high, approximately 5 to 26%.
A reanalysis of an
additional NRC spiked sample (multiple isotopes) and the
licensee's original calibration source confirmed that the noted
bias resulted from inaccurate calibrations.
(3) Xe-135 analysis for the 12 cc gas vial geometry was in
disagreement.
All other gas results for this geometry were biased
low.
For this geometry, efficiency calibrations were suspect in
that the 100 cc gas bomb analyzed using this detector were in
agreement, thus eliminating sof tware nuclide identification and
quantification problems.
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(4) For the 4600 cc gas marinelli geometry analyzed using Detector
No. 24-P-04TC5 results were biased low for all nuclides and in
disagreement for .Xe-133.
Because results for this geometry on
Detector 24-N-1096 were in agreement, the inspectors again
suspected inaccurate calibration resulted in the observed
differences.
The inspectors noted that the above detailed differences demonstrated
and supported the violation regarding inadequacy of the QA Program
procedures for effluent measurements detailed in Paragraph 6.c.
Furthermore, the inspectors informed cognizant licensee representatives
.that large
inconsistencies
for
nuclide
identification and
quantification among detectors were identified in a previous inspection
(50-413/84-64-02) and thus, initial licensee _ actions regarding that
item are now considered inadequate.
Licensee representatives
acknowledged the. concerns and stated that an evaluation of all
geometries to ensure accurate effluent measurements woul_d be
-implemented in a timely manner.
b.
The inspectors noted that the licensee was provided with a simulated
liquid waste sample by the NRC Contract Laboratory, July 1985, and was
requested to complete radiochemical analyses for H-3, Fe-55, Sr-89, and
Results of these comparisons were issued to the licensee in a
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. letter from Roger D. Walker, Director, Division 'of Reactor Projects,
NRC R.cgion II, dated November 29, 1985.
Results were in agreement for
all nuclides.
However, the inspectors noted that a significant bias,
overestimation of. Fe-55 concentrations by 24 to 45%, has been
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identified in three NRC samples analyzed since June 1984.
The~
inspectors discussed licensee actions regarding this systematic bias
with cognizant licensee representatives.
Licensee actions have
included a technical audit of the vendor laboratory, suggested
technical changes to established procedures, and providing the vendor
with spiked liquid samples for Fe-55 analysis capability verification.
Iicensec representatives committed to completing verification of Fe-55
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analysis method 31ogy and stated that they would request by March 31,
1986, additional spiked samples from the Region II Office to
demonstrate Fe-55 analysis capability.
The inspectors informed
licensee representatives that review of the licensee's Fe-55 results
for the NRC spiked sample would be considered an open item and would be
reviewed
during
a
subsequent
inspection
(50-413/85-50-04,
50-414/65-60-04).
No violations or deviations were identified.'
Open Item (50-413/85-50-04, 50-414/85-60-04) - Review of licensee's
Fe-55 analysis of NRC spiked samples.
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9.
Use of Fission Product Phantom for Checking Whole-body Counter Measurements
(84725)
During the inspection, the inspectors verified the licensee's capability to
perform radiological bioassays using their whole-body counter system.
A
fission product phantom containing radioactive sources was provided to the
licensee.
The phantom duplicated nuclide and organ burdens that the
licensee might encounter during normal operations. The phantom was analyzed
using the licensee's normal methods and equipment.
The licensee had one whole-body counting system which had recently been
placed into service.
This system was a " stand-up" linear geometry counter
referred to as "FASTSCAN" manufactured by Canberra.
The FASTSCAN
incaporated two large NaI(TL) detectors, configured in a linear array on a
common vertical axis.
The dimensions of the crystals were 4" x 4" x 16" and
each was viewed by a single photomultiplier tube.
The subject stands inside
the shield on an axis parallel with the detectors.
The inspectors reviewed
the licensee's procedures for operating and calibrating the whole-body
counting system.
In addition, the Quality Assurance program was reviewed.
Calibrations were conducted using a vendor supplied block phantom containing
an NBS traceable mixed gamma source.
The mixed gamma source was contained
in a cylindrical solid matrix and consisted of Cd-109, Cc-57, Ce-139,
Hg-203, Sn-113, Sr-85, Cs-137, Y-88, and Co-60.
The licensee calibrated the
whole body counting system for three geometries:
lung, thyroid, and
gastrointestinal (GI) tract.
The inspectors noted that the licensee's
calibraticn phantom was a Radiation Management Corporation (RMC) REMCAL
Transfer phantom which was designed as an analog to the Alderson REMCAL
Phantom.
The REMCAL phantom assembly consisted of the following separate
cavities: GI, lung, and thyroid.
The calibration source was placed
approximately in the center of the air space in each cavity.
The inspectors
questioned whether or not the GI cavity should be filled with water and the
lung cavity. with a lung tissue equivalent matrix.
The inspectors reviewed
QA records including daily source checks.
Daily source checks were plotted
and tracked on control charts.
The inspectors noted that daily background
checks were performed and the analytical results compared to previous
background analysis results.
The computer system prints the " minimum
detectable activity" (MDA) amounts as the background.
The inspectors and
the licensee discussed plotting the gross background counts from each
detector to note' any significant change in detector background
characteristics.
Subsequent to discussions with NRC inspectors, cognizant licensee
representatives agreed to evaluate the following areas as related to review
of HP/0/8/1001/21:
Trend gross background counts from each detector to note any
significant change in background or detector characteristics.
Provide methodology to plot and tabularize efficiency versus energy
data for each geometry used on the FASTSCAN.
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Evaluate 'the use of ' the REMCAL Transfer Phantom with air-filled
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cavities'and 'a point calibration source.
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The inspectors informed licensee representatives that .these items will be
considered an open. item and will be reviewed during a future inspection
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(50-413/85-50-05,.50-414/85-60-05).
Results of the'intercomparisons are presented.in' Table 2.
The results were
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based on an average of five measurements, each measurement counted for 60
seconds.
All licensee results were higher than; the know values, with
-measurements ranging from 23 to 54% above the know values.
The inspectors
noted that longer counting times, approximately 180 seconds, did not improve
. comparison results.
~No violations or deviations were. identified.
'Open Item (50-413/85-50-05, 50-414/85-60-05) - Review of Calibration
and QA Data Required for Operation of Whole-body "Fastscan" BBA System.
10.
Review of Counting Room Procedures for Unit 2 (84525)
The inspectors reviewed procedures and discussed with cognizant licensee
- representatives . radiological analytical capabilities for Unit 2 ~ effluent
measurements.
The inspectors noted that the counting room is a shared
- facility between Units 1 and 2' and . thus all . analytical procedures are -
identical.
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TABLE 1
RESULTS OF CONFlRMATORY MEASUREMENTS AT CATAWBA NUCLEAR PLANT - DECEMBER 9-13, 1985
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CONCENTRATION fuCl/ Unit)
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~ RATIO
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SAMPLE
ISOTOPE
LICENSEE
HRQ
. RESOLUTION
LICENSEE /NRC
. COMPARISON-
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(1) Reactor Coolant
Co-58
7.23 E-3
1.18 0.03 E-2
39
0.61
Disag reement
(50 ml Bottle)
1-131
~7.13 E-3
8.23 0.27 E-3
30
0.87
Ag reement
1-133
1.28 E-2
1.48 .0.04 E-2
37
0.86
Ag reement
I-135
1.13 E-2
1.18
0.11 E-2
'11
0.96
Ag reemen t '
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(2) Reactor Coolant
Co-58
'6.89 E-3
1.18 0.03 E-2
39
0.58
Disagreement
(50 ml Bottle)
1-131
5.84 E-3
8.23
0.27.E-3
30
0.71
Di sag reement
1-133
1.27 E-2
1,48
0.04.E-2
37
'O.86
Ag reement
I-135
1.10 E-2
1.18 0.11 E-2
11
0.93,
Ag reement
(3) Reactor Coolant
Co-58
8.09 E-3-
1.18 0.03 E-2
39
0.68
Disagreement
(50 ml Bottle)
1-131
8.81 E-3-
8.23 0.27 E-3
30 -
1. 0 7.'
Ag reement
1-133
1.48 E-3
1.48' O.04 E-2
.37
1.00
Ag reement
l'135
1.40.E-2
1.18 0.11 E-2
~ 11
1.19
Ag reement -
(4) Reactor Coolant
Co-58-
8.67 E-3
1.18
0.03.E-2'
39
0.73
Disagreement
(50 ml Bottle)
1-131
7.67 E-3
8.23 0.27 E-3
30
0.93
Ag reement
1-133
1.36 E-2
1.48 0.04 E-2
37
0.92
Ag reement
1-135
1.36 E-2
1.18 0.11 E-2
11
'1.15
Ag reement
(1) Liquid Waste
9.10 E-7-
1.28 0.18 E-6
7
0.71
Ag reement
(3500 ml Marinell:)
Co-58
3.04 E-5
2.96 0.06 E-5
49
1.03
Ag reement
Fe-59
2.13 E-6
2.06 0.27 E-6
8
1.03
Agreement
2.28.E-6
1.82. 0.24 E-6
8
1.25
Ag reement .
1-131
2.58 E-6
2.48 0.24 E-6
10
1.04
Ag reement
1-133
9.97 E-7
1.02 0.37 E-6
3 c
Cs-134
4.04 E-6
3.27 0.29 E-6
11
~
0.98
Ag reemen t
1.24
Agreement
7.42 E-6
6.82
0.39.E-6
17
1.09
. Agreement
(2) Liquid Waste
9.28 E-7
-1.28
0.18 E-6
7
0.72
Ag reement
(3500 mi Ma rine l l : )
Co-58
3.03 E-5
2.96 0.06 E-5
49
1.02.
' Ag reemen t .
Fe-59
1.76 E-6
2.06 0.27 E-6
8
0.85
Ag reement
1.93 E-6
1.82: 0.24 E-6
8
1.06
Ag reement
1-131
2.67 E-6
2.48 0.24 E-6
10
1.08
Ag reement
I-133
7.88 E-7
1.02 0.37 E-6-
3
0.77
Ag reement
Cs-134
3.79 E-6
3.27 0.29 E-6
11
1.16
Ag reement
7.64 E-6
6.82 0.39 E-6
17
1.12
Ag reement
(3) Liquid Waste
9.86 E-7
1.28 0.18 E'-6
'7
0.77
Ag reement
(3500 ml Ma rine l l: )
Co-58
3.08.E-5
2.96 0.06 E-5
49
1.04
Ag reement
re-59
1.86 E-6
2.06 0.27 E-6
8
0.90
Ag reement
2.13 E-6
1.82 'O.24 E-6
8
'1.17-
Ag reemen t
1-131
2.72 E-6
2.48 0.24 E-6
10
1.10
Ag reement
1-133
1.12 E-6
1.02 0.37 E-6
.-3
1.10
Ag reemnnt
Cs-134-
_3.68 E-6
'3.27
0.29 E-6
'11
1.12
Ag reement
7.59 E-6
6.82 0.39 E-6-
17
1.11
Ag reement
+~
,
'
<
1
'
~
,
I .
-
TABLE 1-(Cont'd)
,
f
'
C95GENTRAT10N fuCl/ Unit 1'
RATIO'
SAMPLE
ISOTOPE
LICENSEE
HRQ
RESOLUTION'
LICENSEE /NRC
, COMPARISON
(4) Liquid Waste
<Mn-54
9.40 E-7:
1.28 0.18 E-6
7-
0.73-
Ag reemen t .-
( 3500 ml . Ma ri ne i I : )
Co-58
3.07 E-5
2.96 0.06 E-5
49-
1.04
Agreement
Fe-59
2.20 E-6
2,06~ 0.27 E-6-
8
.1;07:
- Ag reement
-
2.17 E-6
s1,82
0.24.E-6
8'
1.' 19 _
' Ag reement
'
>
1-131
2.86 E-6
2.48' O.24 E-6
.10
1'15 -
~
Ag reement -
.
<
I-133
1.21 E-6_
'1.02
0.37 E-6'
.-3-
1.19
Agreement
Cs-134
3.83 E-6
3.27 0.29 E-6
11
1.17
'.- Ag reement
7.56 E-6
6.82 0.39 E-6
17
1.11-
Ag reement
(1) Particulate Filter
8.23 E-3
.9.40
0.28.E-3-
34
0.88
' Agreement
(47 mm Filter Paper)
1.13 E-2
1,32 0.03 E-2'
44
0.86
Agreement-
(2) Particulate Filter
3.32 E-3
.3.62
0.21 E-3
17
0.92
Agreement.
(47 mm filter Paper)
. 1.82 E-2
2.02. 0.04 E-2
50
0.90
Ag reement '
Cs-137~
1.36 E-2
1.34 0.03 E-2
45
1.01
Agreement.
4
Ce-144
5.92 E-3.
6.64. 0.42 E-3
16
0.89
. Ag reement L
1
f-
(3) Particulate Filter
3.34 E-3
3.62 0.21 E-3
17
~0.92
Agreement
(47 mm Filter Paper)
1.84 E-2
2.02 0.04 E-2
50
-0.91
Ag reement
4
1.38 E-2
1.34 0.03 E-2
45
1.03
Ag reement
j-
Ce-144
6.53 E-3
6'64 0.42 E-3
16
-0.98
Ag reement
.
.(4) Pa rticulate Filter
9.55 E-3
9.40 0.28 E-3
34
1.02
Ag reemen t .'
(47 mm Filter Paper)
1.38 E-2
1.32 0.03 E-2
44
1.04
Ag reement
I
(2) Charcoal Cartridge
Ba-133
5.01'E-2'
4.08 0.03 E-2
136.
1.23
'Ag reement
(Face Loaded)
4
( 3 ) Cha rcoa l Ca rt ridge
Ba-133
-5.16 E-2
4.08 0.03 E-2
136
1.26
Di sag reement
(Face Loaded)
,,
(4) Cha rcoa l Ca rtridge
Ba-133
4.60 E-2
4.08 0.03 E-2
136
1.13
Ag reement
( Face Loaded)
(1 ) Cha rcoa l Ca rtridge
Ba-133
4.30 E-2
4.08 'O.03.E-2
136
1.05
Ag reement
( Face Loaded)
(1) Cas
K r-85m
4.94 E-3
6.21
0.11 E-3
56
0.80
(12cc Vial)
Kr-88
1.20 E-2
1.40 0.05 E-2-
28
0.86
. Ag reement
Ag reemen t ~
3.70'E-2
4.02 0.02 E-2
201
0.92
Ag reement
Xe-135
2.18 E-2
2.81
0.02 E-2
140
0.78
Di sag reement
.
( 3 ) Ca s
.
Xe-133m
~1.11 E-5
1.00
0.06.E-5
17
1.11
' Ag reement
(4600cc Ma ri ne l l i )
6.88 E-4
6.30 0.02 E-4
315
1.09
Agreement
Xe-135
1.09 E-5
9.86 0.15 E-6
66
1.10
Ag reement
(4) Gas
.
Xe-133m
9.24 E-6
1.00 0.06 E-5
17-
0.92
Agreement
( 4600cc - Ma rine l l i )
4.47 E-4
.6.30.'O.02 E-4
315
0.71
Di sag reement
Xe-135
.8.52 E-6
9.86 'O.02 E-6
66
0.86
'Ag reement
U
-.
_
=
-
.
,
. .
_.
-
'(
l
-
TABLE 1 (Cont'd) .
C0fLCERTRATION f uci/Uniti-
..
RATIO
.
.
>
SAMPLE-
ISOTOPE-
LICENSEE
![RC
' RESOLUTION
LICENSEE /NRC
. COMPARISON
'
(1) Cas
Xe-131m
.1.48 E-3
1.38 0.12 E-3
12 ~
'1.07
Ag reement
- ( 100cc Ca s Bomb )
Xe-133m
.1.04'E-3'
1.10
0.03.E-3
37-
0.94
Ag reement
7,77 E-2
7.82' O.01 E-2
782
0.99
Ag reement
Xe-135
4.69 E-4-
4.78 0.07 E .4
68
0.98
Ag reement -
(2) Gas-
Xe-131m
1.26 E-3
1.38 0.12 E-3
12
0.91
Ag reement
(100cc Gas Bomb)
Xe-133m
9.78.E-4
1.10 .0.03 E-3
37
0.89
Ag reement
7.29 E-2
7.82 0.01 E-2'
782
0.93
Ag reement
,
Xe-135
4.26 E-4
4.78 0.07 E-4
68
0.89,
Ag reement
ND Not Detected
NC Not Compared
(1) Analyzed using Gamma Spectroscopy System No. SN.24-N-1096
(2) Analyzed Using Camma Spectroscopy System No. SN 24-P-94VC
( 3 ) Ana lyzed Using Gamma Spectroscopy System No. . SN 24-P-92VA
(4) Analyzed Using Gamma Spectroscopy. System No. SN 24-P-04TC
,
V
4
_-
.:
-
, -
,
.
TABLE 2
,,
RESULTS OF WHOLE BODY. COUNTER MEASUREMENTS
USINC A COMMERCIALLY AVAILABLE FISSION PRODUCT PHANTOM AT
CATAWBA NUCLEAR PLANT, DECEMBER 11, 1985
Licensee (1)
NRC
Ratio
Nuclide
O rga n
(nCl)
(nCl)
( Licensee /NRC)
Lungs
30.9
20.1
1. 54 ..
Lungs
51.9
33.7
1.54
Lungs
211.5
172.4
1.23
-Cs-137
Lungs
120.3
87.8
1.37
1.
Licensee value represents the arithmetic mean of five measurements, each measurement was counted for 60 seconds.
NOTE: Licensee's calibration procedure used a point source.
r-
.
ATTACHMENT 1
CRITERIA FOR COMPARING ANALYTICAL MEASUREMENTS
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 differences between the NRC and licensee
values should be more restrictive. Conversely, poorer agreement between NRC and
licensee values must be considered acceptable as the resolution decreases.
2
For comparison purposes,-a ratio of the iirensee value to the NRC value for each
individual nuclide is computed. This rath- ls then. evaluated for agreement based
on the calculated resolution. The corr...anding resolution and calculated ratios
which denote agreement are listed in Table 1 below.
Values outside of the
agreement ratios for a selected nuclides are considered in disagreement.
NRC Reference Value for a Particular Nuclide
1 Resolution = Associated Uncertainty for the.Value
Licensee Value
2 Comparison Ratio = NRC Reference Value
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
,
,
L