ML18152A334
| ML18152A334 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/12/1993 |
| From: | Decker T, Gloersen W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18152A335 | List: |
| References | |
| 50-280-93-19, 50-281-93-19, NUDOCS 9308300116 | |
| Download: ML18152A334 (20) | |
See also: IR 05000280/1993019
Text
Report Nos.:
UNITED STATES
N~CLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
AUG I 3 1993
50-280/93-19 and 50-281/93-19
Licensee:
Virginia Electric and Power Company
Glen Allen, VA 23060
Docket Nos.:
50-280 and 50-281
F ac il ity Name: -surry 1 and 2
License Nos.:
Inspection
. Accompanying Per/~yel: *:* R,', Vo/
Approved by: ~IV-11~< /? i.ltz(~~'--
T. R. Decker, Chief
Radiological Effluents and Chemistry Section
Radiological Protection and Emergency Preparedness
Division of Radiation Safety and Safeguards
ate Signed
SUMMARY
Scope:
Branch
This routine, unannounced inspection was conducted in the areas of audits,
confirmatory measurements with the Region II Mobile Laboratory, post accident
sampling capability of highly radioactive coolant and containment atmosphere
samples, and the status of previously identified inspection findings.
Results:
The licensee's audits and activities in the areas of radioactive waste
treatment, and effluent and environmental monitoring were technically sound,
thorough, detailed, and well documented.
The licensee's program for identifying and tracking the relatively large _
number of maintenance problems associated with the High Range Sampling System
(HRSS) was effective in that work requests and repairs were generally timely.
The confirmatory measurement results for gamma emitters were reviewed, and, in
general, the licensee's results were in agreement with the NRC results
97 percent of the time (Paragraph 6.b). The disagreements which did occur
could be attributed to photopeak and background interference and differences
between the NRC's and licensee's software for gamma ray spectroscopy.
~30830011, 930813
'
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ADOC~. 05000280
.'
One inspector followup item (IFI) was identified concerning the methodology
for measuring gross alpha particle radioactivity in liquid effluent and water
samples (Paragraph 6.c) .
1.
REPORT DETAILS
Persons Contacted
Licensee Employees
- J. Artigas, Supervisor, Quality
- W. Benthall, Supervisor, Licensing
- R. Bilyeu, Licensing Engineer
- M. Biron, Supervisor, Radiological Engineering
R. Cox, Senior Chemistry Technician
- B. Dorsey, Radiological Protection
- D. Erickson, Superintendent, Radiological Protection
P. Harris1 Health Physics Technician
- D. Hart, Supervisor, Quality
- M. Kansler, Station Manager
R. Lasalle, Supervisor, Radiological Analysis
C. Mehalic, Senior Instructor, Chemistry.
- J. Price, Assistant Station Manager
- S. Swientoniewski, Supervisor, Station Nuclear Safety
T. Swindell, Supervisor, Chemistry
Other licensee employees contacted included engineers, technicians, and
office personnel.
Nuclear Regulatory Commission
- M. Branch, Senior Resident Inspector
S. Tingen, Resident Inspector
- Attended exit meeting on July, 16,* 1993
2.
. Status on Previously-Identified Inspection Findings (92701)
(Closed) Inspector Followup Item (IFI) 50-280, 281/~3-10-02:
Review the
licensee's quality*control program for shipping paper documents.
During the inspection documented in Inspection Report (IR) 50-280, 50-
281/93-10, the inspector noted that disposal facility waste manifest
shipment number 893-1 incorrectly specified the model number of the high
integrity container (HIC) as PL8-120FR.
Shipment 893-1 was a Class B
waste shipment of dewatered resin.
The correct HIC model number, as
stated on the Bill of Lading, was PL6-80FR.
The inspector discussed
this concern with Radioictive Material Control representatives, and the
problem was corrected immediately and the disposal site was notified of
the error on April 15, 1993.
Although this problem appeared to be an
isolated case of clerical error, the inspector and licensee
representatives discussed the need to review the licensee's quality
control of shipping paper documents .
.....,
------ -----.,-
2
During this inspection, the inspector verified that the licensee had
revised and initiated Procedure Action Requests (PARs) for the following
radioactive waste and materials shipping procedijres:
0
HP-7;1.40, Packaging and Shipment of Radioactive Material
(Procedure Change Number (PCN) 93-31, dated July 12, 1993)
0
HP-7.2.40, Disposing of Radioactive Waste Using the Barnwell
Disposal Facility (PCN 93-32, dated July 12, 1993)
0
HP-7.2.50, Transfer of Radioactive Waste to Licensed Waste
Processors (PCN not specified, dated July 15, 1993)
The revisions in the procedures listed above basically required that
shipping papers will be reviewed by two individuals prior to each
shipment.
Specifically, the revisions included a requirement for an
independent review by a qualified Radioactive Material Control
Technician prior to the.final review performed by the Radioactive
Material Control Supervisor. After reviewing the PARs, the inspector
considered this item closed.
3.
Audits (84750)
Technical Specification (TS) 6.1.C.2.h req~ires that audits of ~nit
activities be performed under the cognizance of the Management Safety
Review Committee (MSRC) in the following areas: (1) the conformance of
facility operation to provi_sions contained within the TSs and applicable
license conditions at least once per 12 months; (2) the radiological
environmental monitoring program at least once per 12 months; (3) the
OFFSITE DOSE CALCULATION MANUAL (ODCM) and implementing procedures at
least once per 12 months; and (4) the PROCESS CONTROL PROGRAM (PCP) and
implementing procedures for processing and packaging of radioactive
wastes at least once per 12 months.
The inspector reviewed the following audit reports and assessments:*
0
QA Audit 93-01:
Radiological Environmental Monitoring and
Environmental Protection Plan Audit, dated Maren 16, 1993
0
QA Audit 93-02:
Offsite Dose Calculation Manual and Process
Control Program, dated March 16, 1993
The above audits assessed the adequacy and effectiveness of'the
radiological effluent monitoring program, radiological environmental_
monitoring program, the ODCM, and the PCP.
The audits covered the areas
. specified in TS 6.1.C.2.h.
In general, the audits were thorough,
detailed, .and well documented.
There were no significant findings in
Audit 93-01.
In Audit 93-02, the licensee identified a concern for
increased emphasis *on compensating for differential pressures affecting
3
radioactive gaseous effluent monitoring instrumentation. Similar
concerns had been identified in NRC Information Notice (IN) 82-49,
Correction for Sample Conditions for Air and Gas Monitoring.
Licensee
management made adequate commitments to correct the problem noted above.
No violations or deviations were identified.
4.
Program Changes (84750)
The inspector and the licensee discussed any changes in the radwaste and
radiological environmental monitoring organizations or programs since
the last inspection. Although there were no significant program
changes, the inspector did note that the Supervisor. Health Physics had
been reassigned to .work in the Radiological Engineering Section and
placed in charge of the preoperational testing and calibration program
of the licensee's new Canberra Series 95 gamma ray spectroscopy system.
The licensee expected the new system to be fully operational by the end
of 1993.
No violations or deviations were identified.
5.
Post Accident Sampling System (84750)
NUREG-0737,Section II.B.3 specified the requirements for the Post
Accident Sampling System (PASS).
These requirements specified the types
of samples to be taken, sampling times, accuracies, and sensitivities of
sample analyses, limits of radiation dose to operators, and design
considerations.
The initial inspection of the Surry Units 1 and 2 PASS was conducted by
NRC contractors from January 9 through January 13, 1984.
During this
inspection, the inspector verified through a combination of interviews,
direct observation, and record review that the licensee maintained the
capability to safely obtain and accurately analyze highly radioactive
reactor coolant and containment atmosphere samples rnder accident
conditions.
The details of this assessment are noted below.
a.
System Description
The licensee's PASS or High Radiation Sampling System (HRSS) was a
semiautomated system Model A series built by the Sentry Equipment
Corporation.
The system was installed in 1981 and became
operational in 1984.
The HRSS consisted of two panels which were
remote from each other and were shared between Units 1 and 2.
Jhe
main control panel was in the Turbine Building and contained
controls and instrument readouts for most of the equipment. The
second (sampling) panel was in the Auxiliary Building and housed
all of the analysis equipment. This panel had a large number of
hand-operated valves and required an operator to be present during
4
the sampling process.
The panel was heavily shielded to maintain
the operator's exposure below the maximum dose levels specified in
The operator at the sample p~nel must be in voice
contact with the control panel operator. This was accomplished
with a dedicated headphone system.
The HRSS had the capability to collect liquid samples from the
following five sample points: (1) reactor coolant hot leg;
(2) reactor coolant cold leg; (3) containment sump; (4) residual
heat removal system (RHR); and (5) chemical volume control system
(CVCS) demineralizer outlet.
The normal lineup would be from the
reactor coolant hot leg. The liquid sample could be directed to a
diluted or undiluted.sampler and then directly delivered to a
sample vial which was then lowered into a shielded cask and.
prepared for shipment or analysis.
The containment air sample was drawn through a heat traced sample
delivery line. After the heat traced line was at full operating
temperature, *the particulate-iodine-gas sampler directed small
portions of the gas stream through a particulate filter, a silver
zeolite cartridge, and into an evacuated vial for noble gas
analysis. This gaseous stream could also be directed to an
undiluted grab sampler or the gas chromatograph for hydrogen
analysis. All waste from the system was pumped from the waste
tank back into the containment of the affected unit.
b.
Maintenance
The inspector reviewed the maintenance history of the HRSS during
the last 18 months.
It was noted that the licensee had assigned a
Senior Chemistry Technician to track maintenance problems and
determine monthly HRSS availability status. The assignment of an
individual to monitor this system has been beneficial considering
the maintenance required to maintain its operation .. In general,
it was noted that maintenance repairs were timely and normally
completed within less than 30 days.
The.following summarizes the
maintenance problems during the last 12 months:
Recurring problems with low level indication on liquid waste
tank (17 gallon tank) in Auxiliary Building basement:
(1) check valve leaking problems; (2) recirculation valve
closure problems; (3) solenoid control valve problems; and
(4) heat trace panel controller/alarm problems.
Gas chromatograph column temperature display problems.
Gas chromatograph attenuation selector problems.
Clogging of containment sump sampling lines and coolers due
to suction of dirt and grit into the lines. The design
change package included the installation of grit screens to
prevent large material from obstructing the sampling lines.
5
~
Flow restriction problems between sample cooler and HRSS
panel.
0
Reactor coolant valve #2 leakage problems.
The inspector noted that the licensee utilized a worksheet to
determine overall HRSS availability on a monthly basis.
Basically, the licensee determined the availability of the
following three systems: (1) reactor coolant sampling module,
(2) containment air sampling system, and (3) containment sump
sampling module.
The licensee assigned a percent weight
availability to each component of the systems noted above as shown
in Table 1.
Table 1
HRSS Percent Weight Availability
HRSS Module
% Weight-availability
1. Reactor Coolant Sampling Module
Liquid Sample Panel
50
pH Analyzer
10
Gas Chromatograph
10
Born Analyzer
10
Oxygen Analyzer
10
Diluter Valve for Liquid Isotopic
10
2. Containment Air Sampling System
Automatic Sampler
100
3. Containment Sump Sampling Module
Radwaste Liquid Sample Panel
33.3
U-1 Sample Sump
33.3
U-2 Sample Sump
33.3
4. Overall Availability
Reactor Coolant Sampling Module
50
Contai~ment Air Sampling System
25
Containment Sump Sampling Module
25
The inspector reviewed the overall availability of the HRSS from
January 1992 through June 1993 and noted the following:
I
6
Table 2
HRSS Availability.
Calendar Quarter
i
8'l92
2n /92
3rd/92
4th/92
l
8jt93
2n /93
Percent Availability
98
95
96
95
81
100
- It should be noted that the percent availability of the HRSS
components indicated above did-not necessarily indicate periods of
inoperability.
In summary, HRSS problems were readily identified,
initiation of work orders was prompt, and repairs were generally
timely.
c.
HRSS Testing Program
The inspector reviewed the performance testing program for the
HRSS.
This review included the observation of an HRSS reactor
coolant system (RCS) hot leg sample collection on July 14, 1993; a
review of performance testing procedures; and a review of the
performance testing results for the last 12 months.
The inspector reviewed the following HRSS sample collection and
performance testing procedures:
0
O-CSP-HRS-001, High Range Sampling System Chemistry
Instrumentation Calibration, Revision (Rev.) 6, dated
March 25, 1993
0
1, 2-CSP-HRS-002, High Radiation Sampling System Chemistry
Test and Operator Training, Rev. 3, dated November 24, 1992
0
1-CSP-HRS-003, High Radiation Sampling System: Sampling
Containment Air Chemistry Test and Operator Training,
Rev. 2, dated April 1, 1993
0
PT-38.62, High Radiation Sampling System Waste Tank Valve
Test for Post Accident Conditions, Rev. 1, dated May 7, 1992
The inspector reviewed the HRSS performance test results from
August 1992 to July 1993.
Basically, the performance test*
involved a comparison of analytical results of samples taken from
the HRSS panel and "normal" sample locations.
In the case of RCS,
HRSS samples were collected from either the hot or cold legs of
the reactor coolant system and normal RCS samples were collected
from the demineralizer influent or letdown lines.
An acceptable
HRSS sample result was when the HRSS sample result was within the
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7
range of 0.5 to 2.0 of the normal RCS sample result. The HRSS
. liquid sample results reviewed for the time period noted above
were all within the acceptance range.
The licensee was unable to
compare containment atmosphere HRSS sample results since activity
levels were generally below detection limits.
The inspector also observed the collection of a Unit 1 RCS Hot Leg
sample on July 14, 1993. The licensee followed appropriate
procedures and the liquid sample was collected, prepared,
transported to the counting laboratory, and analyzed within three
hours. A confirmatory measurements comparison with the NRC Mobile
Laboratory was unsuccessful due to the low concentrations of
radioactive material in the diluted HRSS sample.
d.
Training
The inspector reviewed the HRSS training program for chemistry
technicians, which included a review of job performance measures
(JPMs), self-study guides, training and examin~tion records, and
HRSS performance testing records.
The following JPMs and self-
study guides were reviewed:
0
Self-Study Guide 6.01 for JPM 6.01, Rev. 0.1, dated May 19,
1993, Operate the High Range Sampling System (HRSS)
0
-
Self-Study Guide 6.02 for JPM 6.02, Rev. 0.1, dated May 19,
1993, Calibrate/Standardize the HRSS 02 Monitor
0
Self-Study Guide 6.03 for JPM 6.03, Rev. 0.1, dated May 19,
1993, Calibrate/Standardize the HRSS pH Meter
0
Self-Study Guide 6.04 for JPM 6.04, Rev. 0.1, dated May 19,
1993, Calibrate/Standardize the HRSS Gas Chromatograph
0
Self-Study Guide 6.05 for JPM 6.05, Rev. O~l, dated May 19, _
1993, Calibrate/Standardize the HRSS Auto Titrator
The inspector reviewed the initial and requalifi~ation training
records, including the questions and results of the written
examination for six chemistry technicians. It was noted that the
chemistry technicians receive requalification training on the HRSS
annually.
For the six technicians noted above, requalification
training was provided in either January or February 1993, and all
of the individuals passed the examination with at least a grade of
80 percent.
In addition, the inspector reviewed the HRSS
performance test records for the last 12 months, which was
performed once per month on the alternating units, and noted that
15 out of 17 qualified chemistry technicians had participated in
the HRSS performance test .
No violations or deviations were identified.
- ' 1 ** :~::,,:
- '
8
6.
Confirmatory Measurements (84750)
10 CFR 20.20l(a) defines a "survey" as an evaluation of the radiation
hazards incident to the production, use, release, disposal, or presence
of radioactive materials or other sources of radiation under a specific
set of conditions.
10 CFR 20.20l(b) requires that each licensee shall make or cause to be
made such surveys as:
(1) may be necessary for the licensee to comply
with the regulations and, (2) are reasonable under the circumstances to
evaluate the extent of radioactive hazards that may be present.
a.
Laboratory Equipment
b.
The inspector examined the licensee's facilities for performing
required radioanalytical measurements of radioactive material in
liquid, gaseous, and particulate mediums.
This examination
included a review*of the licensee's analytical equipment.
The
licensee had a total of six intrinsic germanium detectors onsite.
Four of the detectors were located in the Health Physics (HP)
count room; one detector was located in the Surry Radwaste *
Facility count room; and the other dete~tor was located in the
licensee's training facility. The four intrinsic germanium
detectors in the HP count room were interfaced with a Nuclear Data
6600 series spectrum analyzer.
The remaining two detectors were
interfaced with a Canberra Series 95 multichannel analyzer and
Apogee Version 2.0 software.
The licensee had p~rchased similar
Canberra equipment to replace the Nuclear Data equipment in the HP
count room.
The projected implementation of the new equipment was
Fall 1993.
During this inspection, the licensee had apparently
experienced both detector and preamplifier equipment failures on
two detector syst~ms in the HP count room.
This equipment had
been shipped to the vendor for repair.
In addition, the licensee's other radioanalytical equipment for
effluent sample measurements included a Beckman Model LSC-100
liquid scintillation counter and a Gamma Products Series 5000
alpha/beta proportional counter.
Representative Sampling and Confirmatory Measurements
During this inspection, samples of reactor coolant, waste gas, and
pre-treated liquid waste were collected and the resultant sample
matrices were analyzed for radionuclide concentrations using the
gamma-ray spectroscopy systems of the licensee's counting
laboratories and the NRC RII mobile laboratory. The purpose of
these comparative measurements was to verify the licensee's
capability to measure quantities of radionuclides accurately in
the various plant processes and waste streams. Analyses were
conducted using two of the four gamma sp~ctroscopy systems located
in the radiochemistry count room and the gamma spectroscopy system
in the new Surry Radwaste Facility.
Two of the four count room
c.
9
detectors (detector numbers 1 and 4) were out of service and in
the process of being repaired. Sample types and counting
geometries included the following: * (1) reactor coolant- 100cc
cup; (2) liqui~ waste monitor tank- one lfter Marinelli container;
and (3) Unit I containment atmosphere- one liter Marinelli
container and a charcoal cartridge. A spiked particulate filter
was provided to the licensee for analysis in addition to the
licensee's samples.
A comparison of the licensee's and NRC's results is listed in
Tables 3 through 7, with NRC's acceptance criteria listed in Table-
8.
In general, the licensee results were in agreement with the
NRC's results 97 percent of the time.
The results showed *
agreement for the one liter gas Marinelli, charcoal cartridge,* and
particulate filter geometries (see Tables 3 through 5).
Comparisons of pretreated liquid radwaste sample results (see
Table 6) were all in agreement.
Comparisons of reactor coolant
sample results (see Table 7) showed disagreements in the SRF
detector.for I-132 and detector number 3 in the count room for Tc-
99m.
The disagreement for I-132 was primarily due to a
combination of interference problems due to high sampl*e background
and software differences between the NRC and licensee. Normally,
the licensee waits for several hours from the time of reactor
coolant collection to analysis to minimize the background effects.
In the case noted above, approximately five hours had elapsed
between sample collection and analysis. It should be noted that
the SRF detector was intended to be a backup detector to the four
detectors in the count room, and thus not normally used to analyze
reactor coolant samples.
The inspector observed the collection of the containment
atmosphere sample.
The inspector noted that approved procedures
were being utilized for representative sample collection and that
the technician was knowledgeable of the sample collection process.
Quality Control and Quality Assurance
Regulatory Guide 4.15, Quality Assurance for Radiological
Monitoring Programs (Normal Operations)- Effluent Streams and the
Environment, Revision I, February 1979 specifies a program
acceptable to the NRC to assure the quality of results of
measurements of radioactive materials in the effluents and the
environment outside nuclear facilities during normal operations.
The inspector reviewed the licensee's quality assurance (QA)
program with respect to Regulatory Guide 4.15. This inspection
included a review of quality control (QC) and calibration
procedures for gamma-ray spectroscopy and alpha/beta proportional
counting equipment; calibration records; and quality control data.
The following procedures were reviewed:
0
HP-7.3A.21, .Radioactive Liquid Waste Sampling and Analysis,
Rev. 3, dated July 23, 1993
0
10
HP-9.0.321, Calibration of Gamma Products G-5000, Rev. 3,
dated June 25, 1992
0
HP-9.0.221, Operation of the Gamma Products G-5000,
Rev. 2,
dated April 16, 1992
0
HP-9.0.121, Performance Checks of the Gamma Products G-5000,
Rev. 2, dated June 25, 1992
0
HP-9.0~345, Calibration of Eberline Alpha Counter SAC-4,
Rev. 0, dated September 17, 1990
0
HP-9.0.245, Calibration of Eberline Alpha Counter SAC-4,
Rev. O, dated September 18, 1990
0
HP-9.0.145, Performance Checks of Eberline Alpha Counter
SAC-4, Rev. O, dated September 18, 1990
0
HP-9.0.30, Determination of Lower Limit of Detection, dated
August 29, 1988
0
HP-9.0.301, Calibration of the Canberra Series-95 MCA
System, Rev. 1, dated April 16, 1992
0
HP-9.0.201, Operation of the Canberra Series-95 MCA System,
Rev. 0, dated July 18, 1991
0
HP-9.0.101, Performance Checks of the Canberra Series-95 MCA
System, Rev. 0, dated July 18, 1991
In general, the procedures were acceptable and readily available
for counting room analysts.
The inspector noted a concern with
the procedure for the measurement of alpha particle activity in
liquid samples.
It was noted that there was no apparent guidance
or methodology described in the licensee's procedures for
correcting for interferences due to solids content in liquid
samples containing alpha emitters. According to ASTM Dl943-81,
Standard Test Method for Alpha Particle Radioactivity of Water, _ -
solids content in the sample containing the alpha emitter produces
significant losses in sample counting rates of about 10 to
15 percent loss at absorber density thicknesses of I mg/cm2 *
Liquid samples must be evaporated to dryness onto sample dishes
that allow the sample to be "seen" directly by the detector.
In
effect, the licensee would underestimate the amount of alpha _
radioactivity in a liquid sample with solids content without the
use of an,alpha self-absorption correction factor.
The inspector
reviewed the gross alpha results for liquid -effluent samples
. during the last four quarters and noted that the gross alpha
activity was essentially zero. The inspector also reviewed the
quarterly QC cross check results for the third and fourth qtiarter
1992, and noted a low bias for gross alpha measurements.
At the
11
time of this ins*pection, it appeared that the licensee's method
for measurements of gross alpha particle activity in water was
inadequate, and thus this issue was* identified as an unresolved
item (URI).
On July 27, 1993, in a telephone conversation
initiated by the licensee, licensee and NRC RII representatives
discussed the concern regarding the gross alpha analysis procedure
for liquid effluent compoiites. The discussion included a review
of a faxed copy of revised procedure HP-7.3A.21, Radioactive
Liquid Waste Sampling and Analysis.
The revised procedure
provided enhanced guidance for sample preparation. The guidance
indicated that if hi~h solids were present in the sample (eg.,
greater than 1 mg/cm in a planchet}, a smaller aliquot may be
required to avoid excessive self attenuation of alpha particles.
If less than 50 ml were used, then the licensee would oe required
to recalculate the lower limit of detection (LLD).
The procedural
guidance further stated that if high solids were present, then
select an aliquot such that the sample to be evaporated contains
less than 20 mg of solids. This guidance was based on the
information provided in ASTM D1943-81 (noted above) indicating
that significant losses in sample counting rates of about 10 to
15 percent loss occur at absorber density thicknesses of 1 mg/cm2 *
The sampling planchets used at Surry were 20 cm2 *
Therefore, the
20 mg limit on the weight of the evaporated sample was added to
the procedure.
The licensee further indicated that liquid waste
was processed using a thin film evaporator technology.
The
resulting distillate was the effluent and, according to the
licensee, was essentially demineralized water.
The licensee
indicated that the distillate ~amples had produced no build-up on
the sample planchet and therefore alpha self attenuation had not
been a concern. Althogh it appeared that the additional
information provided by the licensee would resolve the issue noted
above, the inspector indicated to the licensee that the
methodology for measuring gross alpha particle radioactivity in
liquid effluent and water samples including a review of
calibration methods and QC cross check results would be reviewed
during a subsequent inspection and would be tracked as an IFI
(IFI: 50-280, 50-281/93-19-01).
The inspector also reviewed QA/QC data for the four intrinsic
germanium detectors, including daily energy calibrations and full
width at half maximum (FWHM) checks; and monthly background checks
and lower limit of detection (LLD) verification~ for selected
geometries. 'In addition, the inspector reviewed the annual
efficiency determinations for the four detectors which were
performed during the period from July 1991 to November 1991.
Where applicable, the QA/QC data were plotted daily and trended by
laboratory personnel. Acceptance criteria were established at
+/- 2u and 3u.
The inspector noted that the data were readily
12
available for review, well maintained, and organized.
In general,
the laboratory was well organized and personnel were knowledgeable
of their specific duties and theor~tical principals involving
gamma ray spectroscopy.
One inspector followup item was identified.
7.
Exit Meeting
The inspector met with licensee representatives indicated in Paragraph 1
at the conclusion of the inspection on July 16, 1993.
The inspector
summarized the scope and findings of the inspection, including the
unresolved item. During a telephone conversation on July 27, 1993, a
licensee provided additional information regarding the methodology for
measuring gross alpha particle radioactivity in water samples. After
reviewing the information, the inspector informed licensee
representatives that the unresolved item would be tracked as an IFI.
The inspector also discussed the likely informational content of the
inspection report with regard to*documents or processes reviewed by the
inspector during the inspection.
The licensee did not identify any
proprietary documents or processes during this inspection. Dissenting
comments were not received from the licensee.
Item Number
50-280, 281/93-19-01
Description and Reference
IFI - Review the methodology for measuring gross
alpha particle radioactivity in liquid effluent
and water samples (Paragraph 6.c).
13
TABLE 3
Surry Power Station: July 12 - 16, 1993
Sample/Geometry: Containment Atmosphere Unit I/One Liter Gas Marinelli
Detector: #1 (S/N 6912413) Surry Radwaste Facility
Concentration (uCi/unit)
Nuclide
Licensee
NRC
Resolution
Ar-41
8.92E-7
8.91 +- 1.35 E-7
7
1. 55E-5
1.65 +- 0.11 E-5
15
Xe-135
1.54E-6
1.64 +- 0.09 E-6
18
Detector: #2 (S/N 22-P32XB) Count Room
Concentration (uCi/unit)
Nuclide
Licensee
NRC
Resolution
Ar-41
8.53E-7
8.91 +- 1.35 E-7
7
l.96E-5
1.65 +- 0.11 E-5
15
Xe-135
1. 58E-6
1.64 +- 0.09 E-6
18
Detector: #3 (S/N 22-P-959C) Count Room
Concentration (uCi/unit)
Nuclide
Licensee
NRC
Resolution
Ar-41
8.94E-7
8.91 +- 1.35 E-7
7
1. 67E-5
1.65 +- 0.11 E-5
15
Xe-135
l.6IE-6
1.64 +- 0.09 E-6
18
Ratio
1.00
0.94
0.94
Ratio
0.96
1.19
0.96
Ratio
1.00
1.01
0.98
Comparison
. Agreement
Agreement
Agreement
Comparison
Agreement
Agreement
Agreement *
Comparison
Agreement
Agreement
Agreement
- .
. '~;, . ***:.
14
TABLE
4
Surry Power Station: July 12 - 16, 1993
Sample/Geometry: Containment Atmosphere Unit 1/F&J Charcoal Cartridge
Detector: #1 (S/N 6912413) Surry Radwaste Facility
Nuclide
1-133
1-135
Concentration (uCi/unit)
Licensee
NRC
l.13E-9
8.73E-10
1.04 +- 0.11 E-9
I. 08 +- 0 .. 21 E-9
Dete~tor: #2 (S/N 22-P32XB) Count Room
Nuclide
1-133
1-135
Concentration (uCi/unit)
Licensee
NRC
l.35E-9
l.33E-9
1.04 +- 0.11 E-9
1.08 +- 0.21 E-9
. Detector: #3 (S/N 22-P-959C) Count Room
Nuclide
1-133
1-135
Concentration (uCi/unit)
Licensee
NRC
l.39E-9
l.42E-9
1.04 +- 0.11 E-9
1.08 +- 0.21 E-9
Resolution Ratio
9
5
1.09
0.81
Resolution Ratio
9
5
1.30
1.23
Resolution Ratio
9
5
1.33
1.31
Comparison
Agreement
Agreement
Comparison
Agreement
Agreement
Comparison
Agreement
Agreement
15
TABLE
5
Surry Power Station~ July 12 - 16, 1993
Sample/Geometry: Particulate Filter, NRC spike
Detector: #1 {S/N 6912413) Surry Radwaste Facility_
Concentration {uCi/unit)
Nuclide
Licensee
NRC
Resolution Ratio
Comparison
4.59E-2
5.51 +- 0.25 E-2
22
0.83
Agreement
9.03E-4
8.68 +- 0.67 E-4
13
1.04
Agreement
2.30E-2
2.22 +- 0.08 E-2
28
1.04
Agreement
2.51E-2
2.38 +- 0.09 E-2
26
1.05
Agreement
Detector: #2 {S/N 22-P32XB) Count Room
Concentration {uCi/unit)
Nuclide
Licensee
NRC
Resolution
Rat'.)
Comparison
N.D.
5.51 +- 0.25 E-2
22
None
9.02E-4
8.68 +- 0.67 E-4
13
1.04
Agreement
2.32E-2
2.22 +- 0.08 E-2
28
1.04
Agreement
2.62E-2
2.38 +- 0.09 E-2
26
1.10
Agreement
Detector: #3 {S/N 22-P-959C) Count Room
Concentration {uCi/unit)
Nuclide
Licensee
NRC
Resolution Ratio
Comparison
N.D.
5.51 +- 0.25 E-2
22
None
9.12E-4
8.68 +- 0.67 E-4
13
1.05
Agreement
2.36E-2
2.22 +- 0.08 E-2
28
1.06
Agreement
2.62E-2
2.38 +- 0.09 E-2
26
1.10
Agreement
16
TABLE
6
Surry Power Station: July 12 - 16, 1993
Sample/Geometry: Pretreated Liquid Waste/One Liter Marine 11 i
Detector: .#1 (S/N 6912413) Surry Radwaste Facility
Concentration (uCi/unit)
Nuclide
Licensee
NRC
Resolution Ratio
Comparison
Co-58
5.54E-5
5. 71 +- 0.21 E-5
27
0.97
Agreement
l.29E-4
1.37 +- 0.04 E-4
34
0.94
Agreement
Cs-134
9.71E-6
1. 09 +- 0.09 E-5
12
0.89
Agreement
l.39E-4
1. 42 +- 0.05 E-4
28
0.98
Agreement
3.91E-:6
5.22 +- 0.68 E-6
8
0.74
Agreement
Na-24
2.94E-6
3.41 +- 0.33 E-6
10
0.86
Agreement
Nb-95
2.04E-5
1.99 +- 0.15 E-5
13
1.03
Agreement
Zr-95
l.llE-5
1.23 +- 0.11 E-5
11
0.90
Agreement
Detector: #2 (S/N 22-P32XB)' Count Room
Concentration (uCi/unit)
Nuclide
Licensee
NRC
Resolution Ratio
Comparison
Co-58
5.99E-5
5.71 +- 0.21 E-5
27
1. 05
Agreement
1. 40E-4
1.37 +- 0.04 E-4
34
1.02
Agreement
Cs-134
9.66E-6
1.09 +- 0.09 E-5
12
0.89
Agreement
l.34E-4
1.42 +- 0.05 E-4
28
0.94
Agreement
4.76E-6
5.22 +- 0.68 E-6
8
0.91
Agreement
Na-24
3.42E-6
3.41 +- 0.33 E-6
10
1.U
Agreement
Nb-95
2.75E-5
1.99 +- 0.15 E-6
13
1.38
Agreement
Zr-95
l.38E-5
1.23 +- 0.11 E-5
11
1.12
Agreement
Detector: #3 (S/N 22-P-959C) Count Room
Concentration (uCi/unit)
Nuclide
Licensee
NRC
Resolution Ratio
Comparison
Co-58
5.58E-5
5.71 +- 0.21 E-5
27
0.98
Agreement
1.34E-4
1.37 +- 0.04 E-4
34
0.98
Agreement
Cs-134
9.lOE-6
1.09 +- 0.09 E-5
12
0.83
Agreement
l.31E-4
1.42 +- 0.05 E-4
28
0.92
Agreement
4.00E-6
5.22 +- 0.68 E-6
8
0.77
Agreement
Na-24
3.74E-6
3.41 +- 0.33 E-6
10
1.10
Agreement
Nb-95
2.53E-5
1.99 +- 0.15 E-6
13
1.27
Agreement
Zr-95
1. 24E-5
1.23 +- 0.11 E-5
11
1.01
Agreement
17
TABLE 7
Surry Power Station: July 12 - 16, 1993
Sample/Geometry: Reactor Coolant/lOOml Cup (licensee) & 50ml Vi al (NRC)
Detector: #1 (S/N 6912413) Surry Radwaste Facility
Concentration (uCi/unit)
Nuclide
Licensee
NRC
Resolution Rat*io
Comparison
1-131
7.97E-4
6.42 +- 1.17 E-4
5
1.24
Agreement
1-132
4.75E-3.
2.53 +- 0.12 E-2
21
1.88
Disagreement
1-133
l.15E-2
1.19 +- 0.07 E-2
17
0.97
Agreement
1-135
2.36E-2
2.38 +- 0.11 E-2
22
0.99
Agreement
Na-24
l.56E-2
1.55 +- 0.05 E-2
31
1.01
Agreement
Tc-99m
3.90E~4
4.65 +- 0.86 E-4
5
0.84
Agreement
Detector: #2 (S/N 22-P32XB) Count Room
Concentration (uCi/unit)
Nuclide
Licensee
NRC
Resolution Ratio
Comparison
1-131
8.62E-4
6.42 +- 1.17 E-4
5
1.34
. Agreement
1-132
2.38E-2
2.53 +- 0.12 E-2
21
0.94
Agreement
1-133
1. 09E-2
1.19 +- 0.07 E-2
17
0.92
Agreement
1-135
2.42E-2
2.38 +- 0.11 E-2
22
1.02
Agreement*
Na-24
l.54E-2
1.55 +- 0.05 E-2
31
0.99
Agreement
Tc-99m
6. lOE-4
4.65 +- 0.86 E-4
5
1.31
Agreement
Detector: #3 (S/N 22-P-959C) Count Room
Concentration (uCi/unit)
Nuclide
Licensee
NRC.
Resolution Ratio
Comparison
1-131
7.98E-4
6.42 +- 1.17 E-4
5
1. 24 .
Agreement
1-132
2.29E-2
2.53 +- 0.12 E-2
21
0.91
Agreement
1-133
1. 06E-2
1.19 +- 0.07 E-2
17
0.89
Agreement
.1-135
2.46E-2
2.38 +- 0.11 E-2
22
1.03
Agreement
Na-24
l.51E-2
1.55 +- 0.05 E-2
31
0.97
Agreement
Tc-99m
8.55E-4
4.65 +- 0.86 E-4
5
1.84
Disagreement
18
TABLE 8
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 this criteria, the judgement limits denoting agreement or disagreement
between licensee and NRC results are variable. This varirbility is a function
of the NRC's value to its associated uncertainty.
As the ratio _of the NRC
value to its uncertainty, referred to in this program as the resolution1
increases, the range of acceptable differences between the NRC ahd licensee
values should be more restrictive. Conversely, poorer agreement between NRC
and licensee values must be considered acceptable as the resolution decreases.
For comparison purposes, a comparison ratio 2 of the licensee value to the NRC
value for each 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 ratio for a particular nuclide are
considered in disagreement.
TABLE 1
Confirmatory Measurements Acceptance Criteria
Resolutions vs. Comparison Ratio
Resolution .
< 4
Comparison Ratio for Agreement
0.40 - 2.5
4 - 7
8 - 15
16 - 50
51 - 200
> 200
0.50 - 2.0
0.60 - 1.66
0. 75 - 1.33
0;80 - 1.25
0.85 - 1.18
1 Resolution= NRC Reference Value for a Particular Nuclide
Associated Uncertainty for the Value
. 2 Comparison Ratio= Licensee 'Value
NRC Reference Value