ML20011E114
| ML20011E114 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 01/24/1990 |
| From: | Dragoun T, Amy Hull, Kottan J, Pasciak W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20011E112 | List: |
| References | |
| 50-353-89-32, NUDOCS 9002080006 | |
| Download: ML20011E114 (20) | |
See also: IR 05000353/1989032
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U. S. NOC1FAR RB3UIATORY 03MISSICH
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Report No.
89-32
Docket No.
50-353
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License No.
CPRt-107
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Licensee:
miladel;ttia Electric Otmpany
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230:. Mar'cet St M
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m i: adaltttia,
%nnsylvania
Facility Name Limerick Generating Station, Unit 2
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Inspection At: Limerick, Pennsylvania
Inspection Conducted:
Decenbar 11-15, 1989
Inspectors
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Seni r Radiation W inlist
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J. Kottan, IAboratory Specialist
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Pacilities Radiation
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Inspection Sunanary: Inspection on December 11-15, 1989 (Report No. 50-353/89-32)
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Arsas Inspected: Special, announced team inspection of the licensee's
implementation of the followirg task actions identified in NUREG-0737:
post-accident sanpling of reactor coolant arx1 contmiment atmosphere;
post-accident effluent monitorirg; contalment high range radiation monitoring;
and post-accident in-plant iodine monitoring. The inspection was conducted by
two region MW inspectors and one contractor from INL.
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Results: One violation regarding accountability of Special Nuclear Material was
observed. Post-accident sanplirg capability is adequate but sczne inprovenent
itans were identified.
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DETAILS
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Persons Contacted
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Philadelphia Electric Company
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- M. McCormick, Plant Manager
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- R. Dubiel, Superintendent of Services
J. Bilyeu, Chemical Engineer
M. Boyda,
R. Canzanai, I&C Systems Engineer
J. Dougherty, Senior Chemistry Technician
E. Frick, Radiochemist
K. Gordon, Technical Assistant (ST/RT)
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B. Graber
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C. Hetrick, Effluent PhysicistSupport Supervisory Chemist
C. Hoffman, Reactor Engineer
K. Hunt, Senior Engineer - Radwaste
- T. Jackson, Senior Chemist
D. Kelsey, Chemistry Technician
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K. Lall
W. Lee,yfuelManagementSection
C. Mcdonald, Instructor, Training Dept.
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N. McKenny, Chemistru Instrumentation Tech.
- G. Murphy, Sr. Healtfi Physicist
D. Musselman, Sr. Technical Assistant
M. Paulk, Chemistry Technician
1.2 NRC Personnel
- L.Scholl,SeniorResidentInspector
T. Kenny
Resident In3pector
- M. Evans, Resident inspector
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- Denotes attendance at the exit interview on December 15, 1989
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2.
Purpose
The purpose of this inspection was to verify and validate the adequacy of
the licensee's implementation of the following task actions identified in
NUREG-0737, Clarification of TMI Action Plan Requirements:
Task No.
Title
II.B.3
Post Acciden r5an ling Capability
II.F.1-1
Noble Gas Effluenf Monitors
II.F.1-2
Sampling and Analysis of Plant Effluents
II.F.1-3
Containment High-Range Radiation Monitor
III.D.3.3
-Improved in plant Iodine Instrumentation
under Accident Conditions
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As part of the inspection record a review was performed to verify and
validate the adequacy of the licensee's design and quality assurance
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program for the design and installation of the Post-Accident Sampling
System (PASS).
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3.
THI Action Plan Generic Criteria and Commitments
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The licensee's implementation of the task actions specified in Section 2.
was reviewed against criteria and commitments contained in the following
documents:
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NUREG-0737, Clarification of TMI Action Plan Requirements
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Generic Letter 82-05, letter from Darrell G. Eisenhut, Operating
Director,
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Division of Licensing (DOL), NRC, to all Licensees of
Power Reactors, dated March 14, 1982.
NUREG-0578, Recommendations, dated July G79
TMI-2 Lessons Learned Task Force Status Report and
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Short-Term
Letter from Darrel G. Eisenhut Acting Director, Division of
Operating Reactors, NRC, to all Operating Power Plants, dated
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October 30, 1979.
Letter from Darrell G. Eisenhut Director Division of Licensing
NRR to Regional Administrators Proposed duidelines for Calibration
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and Surveillance Requirements for Equipment Provided to Meet item
II.F.1, Attachments 1, 2, and 3, NUREG 0737" dated August 16, 1982.
Regulatory Guide 1.3. " Assumptions Used for Evaluating Radiological
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Consequences of a Loss of Coolant Accident for Boiling Water
Reactors."
Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-
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Cooled Nuclear Power Plants to Assess Plant and Environs Conditions
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During and following an Accident.'
Regulatory Guide 8.8, Revision 3, "Information Relevant to Ensurino
that Occupational Radiation Exposure at Nuclear Power Stations will
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be As low As Reasonable Achievable."
Final Safety Analysis Rcport (FSAR) for the Limerick Generating
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Stations Units 1 and 2, Philadelphia Electric Company.
Technical Specification 6.8.4, Procedures and Programs.
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NUREG 0991, " Safety Evaluation Report for the Limerick Generating
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Stations Units 1 and 2."
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4.
Post Accident Sampling System, Item II.B.3
4.1 Position
NUREG-0737, item II.B.3, specifies that licensees shall have the
capability to promptly collect, handle,ditions existing in theand analyze post-a
samples which are representative of con
reactor coolant and containment atmosphere. Specific criteria are
denoted in commitments to the NRC relative to the specifications
contained in NUREG-0737.
4.2 Documents Reviewed
The implementation adequacy and status of the licensee's
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post-accidentsamplingandmonitoringsystemswerereviewedagainst
the criteria identified in Section 3.0 of this report and in regard
to licensee letters, memoranda, Inspection Report. drawings and station pro
listed in Attachment 1 of this
The licensee's performance relative to these criteria was determined
by interviewing principal personnel associated with post-accident
sampling, reviewing associated procedures and documentation, and
conducting a performance test to verify hardware, procedures and
personnel capabilities.
4.3 System Description and Capability
The licensee has installed a post-accident sampling system which is
a standard General Electric Co. design. It has the ability to obtain
unpressurized undiluted and diluted samples of reactor water from
samples can be obtained from
the jet pump and the RHR system. Also, dary containment atmospheres.
the drywell, suppression pool and secon
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Analyses for chloride, boron, pH, and hydrogen are conducted in the
laboratory using an ion chromatograph, directl
l d plasma
ciromatograph,pH meter with a microelectrode, y coup e
s)ectrometer,
and a gas
respectively. Radioactivity analyses are performed in
the licensee s counting room using a computer based gamma
spectrometer Chloride analysis can also be performed by an off-site
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laboratory.
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4.4 PASS Performance Testing
Grab samples of reactor water and the secondary containment
atmosphere were collected during an operational test of the PASS
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system on December 13 and 14, 1989. During this test licensee
personnel demonstrated the integrated ability to collect and analyze
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samples within the constraints specified in NUREG-0737, Item II.B.3.
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4.4.1 Reactor Coolant Sampling
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The reactor coolant sampling system is designed to obtain samples of
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liquids and gases during all modes of operation. During this
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operational test, samples were collected from the jet pump sampling
line. Although the reactor was shut down during th s operational
test sufficient reactor water level was maintained to permit
'sampiing from this point. An undiluted liquid sample and a dissolved
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gas sample were obtained from the prescribed sampling point.
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4.4.2 Containment Air Sampling
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Air samples can be obtained from two suppression pool sample
locations, two drywell locations, and secondary containment. During
this operational test, secondary containment samples were taken. The
samples included a gas sample, an airborne particulate (le from) the
filter
sample and an airborne iodine (charcoal cartridge) samp
prescrlbedsamplingpoint.
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4.5
Recommendations (Sampling)
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Although the licensee demonstrated the ability to collect and
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analyze liquid and atmospheric samples as required, the following
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improvement items were d scussed w th the licensee.
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4.5.1 Procedure EP-231," Operation of Post-Accident Sampling System
(PASSl", does not specify dose rate limits for PASS sam)1es.
SpeciPic numerical guidance is not given in Procedure E)-231 but
rather statements such as ' acceptable dose rates' are used
throughout the procedure.
4.5.2 The samples taken during this operational test were taken with the
reactor shut down. Limitations on plant operations during start-up
testing have prevented taking samples form the reactor coolant
system at o)erating temperature and pressure and comparing the
results wit 1 samples obtained from normal system sampling points.
The licensee responded to the above items by stating that Procedure
EP-231 would be reviewed and the appropriate changes would be made. The
licensee also stated that a reactor coolant PASS sample would be taken
and compared with a routine coolant sample after operating at sufficient
power level for a sufficient time so that adequate amounts of
radioactivity for comparison would be present in the samples.
4.6 Analytical Capability
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The licensee's commitments relative to range uncertainty, and
Report (FSAR)pability are contained in the Final Safety) Analys
analytical ca
. The Safety Evaluation Report (NUREG-0991
specifies
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that the accuracy, range, and sensitivity of the PASS instruments
and analytical procedures are consistent with Regulatory Guide 1.97
ard NUREG 0737.
4.6.1 Chloride
The licensee's method of chloride analysis is ion chromatography.
The -ion chromatograph (IC) is also used for routine sample analysis
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with the exception that PASS samples are analyzed using a different
eluent because of boron which may be present in the PASS sample.
Chloride standards at three concentrations were submitted to the
licensee for analysis. The standards were prepared by Brookhaven
for the NRC. The licensee's analytical
National Laboratory (BNL)The analytical results are listed in
results were acceptable.
Attachment 2. The licensee has also contracted with an off-site
laboratory to perform chloride analysis on an undiluted sample
should the radiation level on the sample be too high for analysis in
the licensee's laboratory.
4.6.2 Boron
Boron analysis is performed using a directly coupled plasma
spectro.neter (DCP) lyses as well as PASS sample analyses. Boron. The D
routine sample ana
standards prepared by BNL for NRC were submitted to the licensee for
analysis. The licensee's results were acceptable and are listed in
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Attachment 2.
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4.6.3 pH
Analysis for pH is performed in the licensee's laboratory using a
microelectrode on an undiluted 0.5 mi sample. The analytical
instrumentation is located in a fume hood. The licensee performed a
pH analysis on the PASS sample obtained during this inspection and
demonstrated the ability to perfo:,n a pH analysis on an undiluted
PASS sample.
4.6.4 Radioactivity Analyses
Gamma isotopic analyses of both liquid and gaseous PASS samples are
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performed using the licensee's routine gamma spectrometry system.
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The gamma spectrometry system is located in the licensee s counting
room which is adjacent to the chemistry laboratory. The licensee has
a specially configured shield which can be sealed and purged of any
radioactive noble gases prior to sample counting. During this
operational test the licensee simulated analyses of PASS samples to
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demonstrate the adequacy of procedures and techniques for performing
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gamma isotopic analyses of PASS samples. The inspector noted that
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the procedures and techniques appeared adequate to perform the
required analyses.
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4.6.5 Hydrogen and Dissolved Gas
Hydro $en analyses of PASS liquid dissolved gas samples and PASS
conta nment gas samples are performed using a gas chromatograph
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the)GC is vented into the fume hood. The licensee simulated the(GC . The chrl
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analysis of a PASS sample by analyzing the liquid dissolved gas
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sample on the GC. The inspector observed this analysis and noted
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that the licensee's procedures and techniques were acceptable for
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performing a hydrogen analysis of a PASS sample.
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4.7 Recommendations (Chemical Analysis)
Although the licensee's analytical capability satisfied the
accuracy, range, and sensitivity requirements, the following
improvement items were discussed with the licensee.
4.7.1Thelicensee's$roceduresforsam$onandHandlingofecifically
le pr paration, s
Procedures EP-24 " Sample Preparat
Highly
Handling of Highly Radioactive Gas Samples"ple Preparation and, do not
Radioactive Liquid Samples" and EP-243 " Sam
criteria for sample analysis for either chemical or gamma isotopic
analyses. No specific guidance is given to ensure that the diluted
samples will be within the calibration range of the analytical
instrumentation or contain radioactivity concentrations which will
not exceed dead time limitations for the gamma spectrometer.
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4.7.2 The volume of the PASS liquid dilution valve has not been
incorporated into the licensee's dilution procedure. The measured
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volume of this dilution valve is 0.08 ml versus the designed volume
of 0.10ml.
4.7.3 The licensee has performed calibrations of the gamma spectrometer
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at source-to-detector distances of up to approximately 36 inches.
This requires counting samples with the shield lid open. The
licensee's assessment of radiation levels during accident conditions
indicates that under certain situations the counting room will
experience a exposure rate of 8-10 mR/hr from noble gas. The
licensee stated that samples could not be counted under these
conditions. However, the licensee's procedures do not provide
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specific limits on sample exposure rates so that the samples can be
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counted in a shield with the lid closed after purging the
radioactive noble gases from the shield. This applies in particular
to charcoal cartridge (or silver zeolite) samples which cannot be
diluted.
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4.7.4 The licensee's noble gas gamma isotopic results from a containment
atmosphere simple are reported at conditions of standard temperature
and pressure (STP). However, Procedure EP-C-326, "" Procedures for
Estimating Core Damage During Accident Conditions , requires the
actual sample vial temperature and pressure be reported so that the
noble gas activity result can be corrected to containment
temperature and pressure conditions. A procedure change should be
made so that the reported sample results are in the correct form to
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be used in Procedure EP-C-326 to assess core damage.
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The licensee stated these items would be reviewed and appropriate
procedure changes made as necessary.
5.
Noble Gas Effluent Monitor Item II.F.1-1
5.1 Position
NUREG-0737 Item II.F.1-1 requires the installation of noble gas
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monitors with an extended rang designed to function during normal
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and accident conditions. The criteria including the design basis
calke of monitors for individual releas,e pathways, power supply
ran bration and other design considerations are set forth in Table
II.F.1-1 of NUREG-0737,
5.1.1 Documents Reviewed
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The implementation, adequacy, and status of the licensee's
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monitoring systems were reviewed against the criteria identified in
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Section 3.0 of this report and in regard to documents listed in
Attachment I.
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The licensee's performance relative to these criteria was determined
by interviewing the principal persons associated with the design,
testing installation and surveillance of the high range gas
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monitoring systems,ining personnel qualifications, and direct
reviewing associated procedures and
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documentation exam
observationofthesystems.
5.2 Findings
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Within the scope of this review, the following was identified:
5.2.1 Description and Capability
The system as reviewed meets the guidance issued by the NRC in
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the north
The station has three kossible airborne release pathways:
stack, the Unit #1 sou h stack, and the Unit #2 south stack.
Both
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the Unit #1 and #2 south stacks are isolated under high radiation
airborne concentrations. The north stack is the only pathway for.
airborne release of radioactive effluent under accident conditions.
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All three stacks are monitored for routine releases of particulates,
iodine and noble gas b a General Atomic qGAl RD-60 articulate,
.I.G.) monitors. To Pulfill
UREG-0737
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iodine and noble gas (h stack is also equipped with a GA Wide Range
requirements the nort
Gas Monitor WRGM). All airborne effluent sampling points are
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equipped wit isokinetic sample nozzles, heat-traced and insulated
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sample lines, sensors to measure vent and sample flow rates using
mass flow techniques and with redundant computer polling for
collection of data, to minimize sample line lengths, the monitors
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have been located in enclosures on top of the Turbine Building
immediately adjacent to the north and south stacks.
The WRGM contains two sample conditioning modules, one for the
low-range detector flow path and one for the mid- and high-range
detector flow path. These modules consist of particulate filter and
iodine pre-filters and one particulate and iodine grab sample
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location. The gas sample evaluation modules are located
concentrations (10"feetawayonageparateskid.obtain
approximately four
Low range
uti/cc to 10' uti/cc of airborne effluent are
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component is monitored by a beta scintillation detector which views
a 350 cc volume in a lead shield. At mid- and high-range
concentrations, the sample stream flows through a 1/4 inch line at
0.06 CFM. The noble gas component is monitored by small volume -
x2mmxSmm)CadmiumTelluridedgjectorsghichviewa30ccvolum(2m
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for mid-range concentrations (10
to 10
uQ/cc)anda3ccvolume
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for the high range concentration 10'3 to 10
uC/cc).
The mid- and
high-range sample collection and a(nalysis system operates only if
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offluent air concentrations exceed set point values established by
the licensee.
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The selection of an air stream path can be controlled locally or
remotely from the control room. An automatic room air purge is
provided following automatic changeover from the high-flow to the
low-flow air pathway.
Sample collection parameters such as vent
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flow rate, sample flow rate and grab sam)le collection time plus
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uCi/cc or uti/second
estimates of radioactivity in units of C H, Air concentrations for
have local and remote readout capability.
all three WRGM detectors are recorded in uCi/cc on chart paper in
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the control room and stored in the WRGM computer database for trend
evaluation. The licensee also has the capability to collect a srab
sample from the air return line of the WRGM sample analysis skic!.
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The noble gas isotopic mixture, as evaluated from this sample point
is used to determine the instrument response calibration factor for
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the calculation of noble gas release rates.
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At this time the licensee has accepted the GA calibration of the
radiation detection instruments.
The original detectors for the
WRGM, which were installed for Unit 1 operations (see Inspection
Report 50 352/84 66? are still in place and do not show any-
appreciable change form the tertiary calibration provided by GA.
and efficiency calibration is
However,acompletelinearity,energykGMgasmonitoringwouldbe
under consideration. Backup for the W
provided by obtaining and analyzing grab samples of gas from the
WRGM or PIG lines. Taps and appropriate valves in the lines are
provided for this purpose.
5.3 Recommendations
Based on the above findings, the following measures are recommended.
5.3.1 Since the monitoring enclosure is located immediately adjacent
to the top of the north stack, it is possible that under some
post-accident meteorological conditions t'ne enclosure could be
permeated by radiogases. Their intake during a system pu ge would
defeat the purpose of the purge by filling the idled pip ng. This
will result in false indications when the system resets or normal
operation, it is therefore reccmmended that a supply of clean air or
inert gas be supplied to purge pathways.
5.3.2 Due to the close proximity of the monitoring enclosure to the
north stack, a high radiation field is expected under post-accident
conditions. Doses to technicians obtaining a backup gas sample could
approach the GDC-19 criteria. The substitution of a procedure to
calculate gas concentrations in the duct from survey readings near
the duct would materially lower the associated doses.
6.
Sampling and Analyses of Plant Effluents, Item II.F.1-2
6.1
Position
Item II.F.1-2, requires the provision of a capability
for the collection,iva iodines and particulates that may accompa
transport, and measurement of representative
samples of radioact
gaseous effluents following an accident.
It must be performable
within specified dose limits to the individuals involved.
media, sampling considerations,gn basis shielding envelope, sampling
The criteria including the desi
and analysis considerations are set
forth in Table II.F.1-2 of NUREG-0737,
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6.1.1 Documents Reviewed
The implementation, adequacy and status of the licensee's sampling
and analytical system and procedures were reviewed against the
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criteria identified in Section 3 of this report and in regard to
licensee correspondence,1 memoranda, drawings and station procedures
as listed in Attachment
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The licensee's performance relative to these criteria was determined
by interviewina the principal persons associated with the design
installation,andsurveillanceofthesystemsforsampilng
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testinglysis of high activity radioiodine and particulate effluents,
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by examining
by reviewing associated procedures and documentation,f the systems.
personnel qualifications, and by direct observation o
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6.2 Findings
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6.2.1 Description and Capabilities
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ThesstemasreviewedmeetsthekuidanceissuedbyNRCin
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NUREG 0737. As indicated above{ive material under accidentt e north stack is de
sole release >oint for radioac
conditions. Tie sample conditioning portion of the WRGM permits the
licensee to collect a tined grab sample of the effluent air stream
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for both the low range and the mid to hiah range flow paths. The
sample is routed through a particulate filter and silver zeolite
canister for the collection of particulates and iodines. Mid and
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high range pre-filters and grab sample collection assemblies are
lead shielded while the low range assemblies are not. Both the low
sample /high effluent samp(ling points are equipped with isokinetic
and midnozzles, heat trace and insulated sample lines, vent and
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sample mass flow rate meast ement techniques and redundant computer
polling for collection of cata. The low range sample system operates
under routine operating conditions and during accident conditions up
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to set point concentrations which have been defined by the licensee.
At this point sampling begins to occur by the mid/high range system.
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The low range system ceases to operate at an upper level air
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concentration defined by the licensee.
Sample collection parameters such as vent flow rate, sample flow
rate, and grab sample collection have local and remote readout
capability. Collection of the grab sample can be initiated locally
or remotely.The licensee's procedure (EP 237) calls for the
collection of a grab sam)1e timed to limit the activity to 400 uCi.
However, the basis for t11s value is unclear. In addition, the
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procedures do not indicatn the disposition of samples other than
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grab samples or how the total release of particulates and iodines
would be evaluated.
The licensee's analysis indicates that under >ost-accident
conditions the direct access path to the norti stack sampling
enclosure would be via a stairway immediately adjacent to the north
vent duct. A peak radiation level of 28 R/hr is expected at t-+4
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hours post accident along this path. An alternate path via the south
vent is available but involves a climb and descent of three steel.
ladders from 15' to 75' long. Cranes are provided to hoist the 50
pound shield cask used to transport the particulate and iodine
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samples from the WRGM. Personnel must carry the cask through the
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fuel handling building to the analytical laboratory. The analysis
also projects a peak post-accident dose rate of 3.9 R/HR inside the
sampling enclosure. During a demonstration, the time required to
complete preparations, obtain a sample, and deliver it to tile
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analytical laboratory was in excess of one hour.
Although the sampling lines from the isokinetic probes in the north
vent to the WRGM are relatively short (the licensee has notabout 35' for the low rang
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path and 25' for the high range path
quantified the transmission losses in),them.
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6.3 Recommendations
Based on the above findings the licensee should resolve the
following.
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6.3.1 The procedure for the limitation of grab sample activity should be
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clarified. An alternative means of evaluating the amount of activity
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collected should be devised in case the activity exceeds the
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capability of the count room equipment.
6.3.2 Procedures for the determination of the activity collected on other
than grab samples should be provided so as to establish the total
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activity released during a prolonged post accident release.
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6.3.3 A small hand truck should be provided to facilitate the transport
of the shield cask and activity samples through the level portions
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of the building leading to the chemistry laboratory.
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6.3.4 It a) pears that the location of the WRGM was chosen to minimize the
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lengt1 of sampling lines in accordance with the guidance of ANSI
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N13.1-1969. Since the promulgation of this guidance, it has been
demonstrated that long (100' to 200') sampling lines with a diameter
particulates and elemental iodines. provide for high transmission of
of 1" to 2" at flows of 1 to 2 CFM
However, long sampling lines of
1/4" piping with a flow at 0.06 CFM such as the WRGM high range flow -
3ath provide very low and uncertain transmission. Some licensees
lave located the WRGM in a readily accessible area with low
background at some distance from the plant stack and provided for
the continuous operation of the high volume pump". A flow splitter is
then installed close to the WRGM and feeds a 1/4 low flow line to
the mid/high range sample path.
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Movement of the WRGM to a location more remote from the stack would
significantly reduce the climbing hazards and transit exposures to
personnel. This is also recommended in view of the time and dose
constraints that are imposed on the frequency of obtaining samples
from the WRGM at its present location.
7.
In-Containment High Radiation Monitors, Item II.F.1-3
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7.1 Position
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NUREG-0737, Item II.F.1-3, lled. specifies that high range containment
radiation monitors be insta
The specific requirements are set
forth in Table II.F.1-3.
7.1.1 Documents Reviewed
The implementation, adequacy, and status of the licensee's
monitoring systems were reviewed against the specified requirements.
The licensee's performance relative to the requirements was
determined by interviewing the principal persons associated with the
design, testing, installation and surveillance of the containment
high range monitoring systems and by reviewing associated procedures
and documentation.
7.2 Findings
Within the scope of this review the system meets the guidance issued
J
by NRC in NUREG-0737. Four0 energl Atomic (GA) ion chamber detectors
G
with extended ranges of 10 - 10 R/hr have been installed with
appropriate separation in containment.
This exceeds the minimum
requirement of two detectors. Calibration and functional tests have
been performed. The installation is similar to that followed in Unit
e
withstand a harsh environment in the(EO) of the equipment todrywell during a
1. The Environmental Qualification
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reviewed. The results of the EQ review will be reported in
forthcoming Resident Inspection Report 50-353/90-02. Problems
regarding the calibration procedure previously reported in
!
Inspection Report 50-353/83-20 have been corrected by the licensee.
'
The detectors readout on panel RAD 200-600 in the control room in
full view of the operators. A secondary readout is provided via a
computer system with terminals in several on-site locations. An
on-site radiological engineer has been assigned as the ' system
engineer' responsible for this equipment.
7.3. Recommendations
None.
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8.
Improved In-Plant Iodine Instrumentation Under Accident Conditions.~
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Ilem III.D.3.3
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8.1 Position
NUREG 0737. d associated training and procedures for accuratelyItem I!!.D
,
equipment an
i
determining the airborne iodine concentration in areas within the
j
facility where plant personnel may be present during an accident.
adequacy, and status of the licensee's in plant
The implementation, der accident conditions was reviewed against the
iodine monitoring un
f
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referenced requirements. The licensee's performance was evaluated
.
by interviews with cognizant licensee personnel, review of
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applicable procedures, direct observation during a walk through, and
verification of equipment availability and storage.
8.2 Findings
Within the scope of this review, the following was observed.
The licensee demonstrated satisfactory monitoring and measurement
capabilities for accurately quantifying airborne radiciodine
concentrations in areas where plant personnel may be present during
an accident. The vital areas observed were the Technical Support
I
Center, Operations Support Center, and Control Room. Examination of
'
available radiciodine monitoring components and instrumentation
j
(portable low volume air samplers, Eberline PING, activated charcoal
,
indicated the
and silver zeolite radiciodine sampling cartridges)t to effectively
licensee has sufficient sampling media and equipmen
monitor airborne radiciodine levels within these areas or support
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emergency operations in the plant.
The majority of the procedures associated with operation and
1
calibration of radioiodine monitoring equipment were in place.
Sample cartridges will be analyzed by chemistry technicians on the
gamma spectrometer which allows discrimination of interfering
isotopes. A flushing apparatus is available to purge cartridges with
nitrogen gas to minimize the high background count caused by
adsorbed noble gases.
8.3 Recommendations
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Based on this review the following improvement item is recommended.
8.3.1 The policies and procedures regarding post-accident iodine
monitoring need to be improved by the addition of guidance for
supervisors and technicians. Current procedures rely extensively on
discretion and knowledge of the HP personnel. However, most
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personnel are only familiar with the hazards involved in routine
operations but not with the severe conditions that may occur in
plant after an accident. Information such as recommended air sample
size when extremely high activity is suspected cartridge
guidelines, and selection of cartridge type (cbarcoal vs. purge
zeolite) should be provided. The exposure values in General Design
Criterion 19 should also be provided in the procedures as guidance
in decision making.
9.
Control of Special Nuclear Material (SNM)
'
During this inspection the licensee reported to the inspector that a loss
of accountability of SNH had occurred. In May 1989, two Dunking Chambers
containing approximately 5 grams of Uranium-235 were received on site.
Receipt of this material was not acknowledged to the shipper
(Reuter-Stokes Co. nor entered into the licensee's accounting system for
SNM. The devices we)re not used and subsequently shipped to the Hope Creek
Station in September 1989 at the request of the Limerick start-up
contractor (General Electric Co.). No SNM accountability was transferred
since the devices were not recognized as containin SNM. A DOE /NRC
routine SNfi accounting audit discovered the oversi ht in October 1989 and
initiated corrective action. The licensee acknowle ged the oversight,
corrected the documentation, and informed the Hope Creek Station of the
SNM transfer.
The licensee has experienced previous problems with accountability of
items containing SNM. In December 1987 Dunking Chambers were left out of
a package sent fo LaSalle Station. In September 1989'a package containing
Source Range Monitors was not properly labeled during a return to the
vendor.
The recent loss of accountability of SNM constitutes an app (arent
violation of NRC requirements contained in 10 CFR 70.51(b) 1),
(1), and 74.15
74.13(a) ding in discuss (a) (50-353/89-32-01). The licensee acknowledged
this fin
ions with NRC Region I during the week of December
18 1989. The licensee stated that a root cause analysis of this event
willbeinstituted.This.matterwillbereviewedinafutureinspection.
10. Exit Interview
The inspection team met with the licensee personnel denoted in Section 1
at the conclusion of the inspection on December 15, 1989. The findings of
the inspection were presented at that time.
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Attachment 1
I
Documents Reviewed
Limerick Generatire Station Procedures
Ememency Procudures
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-EP-230,
" Chemistry Sanpling and Analysis Team"
-EP-231,
" Operation of Post-Accident Sanpling System (PASS)"
-EP-333 Rev1, " Adjustment of Wide Range Gas Monitor conversion Factors"
-EP-C-326
" Procedures for Estimating Core th9 Ikaring Accident
Conditions"
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-EP-237 Rev3 " Obtaining the Iodine / Particulate ard/or Gas Sanples frun the
North Vent Wide Range Gas Monitor (WR24)"
-EP-241 Rev3 "Sanple Preparation aM Handling of Highly Radioactive Liquid
Sanples"
-EP-242 Rev3 "Sanple Preparation and Handling of Hi
y Radioactive
Particulate Filters and Iodine Cartri
"
-EP-243 Rev4 "Sanple Preparation and Handling of Highly Radioactive Gas
,
Sanples"
-EP-244 Rev0 "Off-Site Analysis of High Activity Sanples"
-EP-250 Revi t"Fersonnel Safety Team Activation"
-EP-292
" Chemistry Sanpling and Analysis 'Ibam Ihone List"
Other Procedure _s
-IM-413
" Loading ard Closing of the PAS-1 Cask"
l
-HP-213 Rev5 " Airborne Activity Survey 'Ibchniques"
"Efficiercy Calibration of Canberra Gama Specttuneter Using
l
Series 90 Multichannel Analyzer (Stardalone)"--01-152.10
" Efficiency Calibration of the Canberra Gansna SWLuneter Using
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Apogee Ccmputer Prograns"
" Determination of Metals by DCP"
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--CH-222
" Operation and Calibration of the Perkin-Elmer Sigma 300 Gas
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Chrcanatograph with 'Ihermal Corductivity Detector"
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Attachment 1
2
]
1
--G-251
"Detamination of Gama Isotopic Activity"--01-903
"Deteminaticn of pH in Iow Volume Water Sanples turing Post
Accident conditions"--01-905 Rev1 "Detamination of Gama Isotcpic Activity IMring Post Accident
conditions"
Surveillance /Itst Procedums
-RP-5-030-350-0
"Ibst Accident Sanpling Station Operational Readiness Check"
--RP-5-030-573-2
" Routine Secondary Containment Atmosphen Sanpling frm
PASS"
,
-RP-5-030-576-2
" Routine Jet Punp Small Volume Ldquid Sanpling frm PASS"
-RP-5-030-577-2
" Routine RHR Small Volume Liquid Sanpling frun PASS"
,
-RP-5-030-578-2
" Routine Jet Punp Iarge Volume Liquid Sanpling frun PASS"
-RP-5-030-579-2
" Routine RHR large Volume Liquid Sanpling fra PASS"
-RP-5-020-620-0
" Post-Accident Sanpling Check of the Wide Range Gas Monitor
(WR34)"
-RP-5-000-930-0
" Routine Inventory Check of the Chemistry Emergency Cabinet
ard the Post Accident Sanple Preparation Station"
-ST-2-026-626-0
" Accident Monitoring-North Stack Accident Monitor" Function
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Test.
-9P-2-026-407-2
" Accident Monitoring-Primary containment Post-IDCA
Radiation" Division III, Calibration.
-ST-2-026-438-0
" Accident Monitorina-North Stack Wide Range Accident
Monitor Calibration" Functional Test.
~ZP -76. 2
" Post Accident Sanpling System" (Pre-Operational Test)
Training Lesson Plans
--OICT-8803
"IES Post Accident Sanpling Systen Operation and Analysis of Post
Accident Sanples"
--OICT-8807
" Gaseous Radwaste System and Gaseous Process and Effluent
Monitoring System"
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Attachment 1
3
4
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1
--OICT-8909
" Training Guide, PASS Sanpling and Dilution Hood Procedures"
1
1
Drawires
--M-26, Sheet 5 of 9, " Plant Process Radiation Monitoring (Casanon)"
--M-30, " Post Accident Sanpling (Unit 2)"
Maintenance Documents
Maintenance Request Fonns: #8403715, #8404275, #8403822, #8507770, #8507882,
- 8502295, #8505243, #8508704, #8503472, #8508643, #8602791, #8605451, #8606847,
+
- 8603888, #8603203, #8604337, #8605627, #8601209, #8606851, #8705925, #8803347,
- 8706806,
- 8707409, #8703118, #8706617, #8704758, #8705351, #8707144, #8802015,
- 8802030, #8805973, #8801331, #8801327, #8801460, #8801573, #8801694, #8801705,
- 8801898, #8804274, #8900595.
I&C Equipnent History Trending Irg
I&C Maintenance History Trundirg Irg
Colt & w derce
letter fran A. P. Hull, IHL, to M. Miller, Region 1 dated April 7,1987.
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Letter fIttu W. C. McDaniel Bechtel Power
to R. A. Halford, PECO,
"DesignReviewofPlantShielding" dated
10, 1984.
Vendor Manuals
GA Technology
-E-115-865 Rev3
" Wide Range Gas Monitor Equipnent Manue.1", March 1986
s
-E-115-647 Rev5
" Calibration Report for Model RD-52 off-Line Beta Detector"
-E-255-961 Rev2
" Calibration Report RD-72 Wide Range Gas Monitor High and
Mid-Range Detectors"
-E-115-791
" Calibration Report for Model RD-60 Partic11 ate, Iodine, and
Gas Detector Systan"
-E-255-978
" Energy Response 'Ibst and Ebse Rate Calibration of Model
"RD-23 High Rarge Radiation Monitor Detector"
-E-115-1070
" Digital High Range Radiation Monitor System- Equipnent
Manual"
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Attachment 1
4
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Literature
P. J. Unrein et al., " Transmission of Radiciodine Throu#1 Sanpling Lines",18th
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DOE Nuclear Airborne Waste Management ard Air Cleaning Conferunoe, OJNF-840806,
pgs.116-126, Mardt, 1985.
A. L. Wright et al., "Ihe Chemistry Behavior of Iodine Vapor Species in Nuclear
Plant Air Manitoring Sanple Lines", 20th DOF/NRC Air Cleaning Conference,
i
NUREG/G0098, pgs. 824-833, May,1989.
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Attachment 2
Comparison of Chemical Test Results
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NRC Known
licensee
licensee Accuracy
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Chemical
Concentration.
Measured
Commitment in
Parameter
(after dilution)
Value
FSAR Section 11
1
1.03+/- 0.02 ppm (1)
0.97 ppm
+/- 10%
,
300+/-40 ppb (2)
295 ppb
"
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510+/-10 ppb (2)
508 ppb
"
30+/-2 ppb (1)
31 ppb
+/-10%
31+/-2 ppb (3)
29 ppb
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"
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9.5+/-0.5 ppb (3)
9 ppb
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Notes:
(1) Original standard diluted 1:1000
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(2) Original standard diluted 1:10,000
(3) Original standard diluted 1:2000
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