ML20235A099
| ML20235A099 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/09/1987 |
| From: | Mcfadden J, Mark Miller, Shanbaky M, Thomas W, Woodard C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235A043 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-3.D.3.3, TASK-TM 50-412-87-50, GL-82-05, GL-82-5, NUDOCS 8709230207 | |
| Download: ML20235A099 (25) | |
See also: IR 05000412/1987050
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 87-50
Docket No.
50-412
License No. NPF-64
Category C
Licensee: Duquesne Light Company
P. O. Box 4
Shippingport, PA 15077
Facility Name:
Beaver Valley Power Station Unit 2
Inspection At:
Shippingport, Pennsylvania
Inspection Conduct
June 29-July 2 and July 7-9, 1987
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Inspector :
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M.McFad6en,SgrRadiation> Specialist
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M. M11eN Senior Radiation Specialist
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W. Thomas, Emerg(ncy Prepared ess Specialist
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C. Woodard, Reactor 9 Engineer
date
Approved by:
b L .MM
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et\\etk M
M. Shanbaky, Chief, Facilitle.)s Radiation
date
Protection Section
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Inspection Summary:
Inspection conducted on June 29-July 2 and July 7-9, 1987
(Inspection Report No. 50-412/87-50)
Areas Inspected:
Special, announced safety inspection of the licensee's
implementation and status of the following task actions identified in
NUREG-0737:
Post-accident sampling of reactor coolant and containment
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atmosphere; increased range of radiation monitors; post-accident effluent
monitoring; containment radiation monitoring; in plant radiciodine
measurements; and design and qualification,
Results: There were no violations or deviations identified during this review.
However, several areas were identified which require improvements.
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DETAILS
1.
Personnel Contacted
1.1 Licensee Personnel
- D. Clardige, Lead Compliance Engineer
- E. Cohen, Director-Rad Ops (Unit 2)
- A. Dulick, Chemistry Supervisor
E. Eilman, Environmental Qualifications (EQ) Group Leader
R. Freund, Sr. H.P. Specialist
- D. Girdwood, Directors-Rad Ops (Unit 1)
- J. Godleski, TCT Engineer
J. Johns, Surveillance Supervisor-QC
- J. Kosmal, Manager-Rad Control
- S. LaVie, Sr. H.P. Specialist
- V. Linnenbom, Director-Plant Chemistry
- F. Lipchick, Sr. Licensing Supervisor
- F. Liptak, Count Room Coordinator
- A. Lombardo, Nuclear Chemistry Specialist
- H. Nickl, Principal Engineer
- T. Noonan, Assistant Plant Manager
- S. Palian, Sr. Chemist (Unit 2)
W. Rutherford, EQ Engineer
R. Sattler, Startup Test Engineer
D. Szycs, EQ Engineer
- R. Vento, Director-Rad Engineering
G. Wargo, Assistant Director-QC
- K. Winter, Sr. H.P. Specialist
K. Woessner, EQ Supervisor
- J. Wolfe, Coordinator-QC Commitments
- M. Zaki, Principal Engineer
1.2 NRC Personnel
A. Asars, Resident Inspector (Visiting)
J. Beall, Senior Resident Inspector
- L. Prividy, Resident Inspector
- Denotes attendance at the Exit Interview conducted on July 9, 1987
Other members of the licensee's staff were also contacted.
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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-Accident Sampling Capability.
II.F.1-1
Noble Gas Effluent 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
3.
TMI Action Plan Generic Criteria and Commitme'nts
The licensee's implementation of the task. actions specified in Section 2.0
-were reviewed against criteria and commitments contained in the following-
documents:
NUREG-0737, Clarification of TMI Action Plan Requirements, dated
November 1980.
Generic Letter 82-05, letter from Darrell G. Eisenhut, Director,
Division of Licensing (DOL), NRC, to all Licensees of Operating Power
Reactors, dated March 14, 1982.
NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short-
Term Recommendations, dated July 1979.
Letter from Darrell G. Eisenhut, Acting Director, Division of
Operating Reactors, NRC, to all Operating Power Plants, dated October
30, 1979.
Letter from Darrell G. Eisenhut, Director, Division of Licensing, NRR
to Regional Administrators, " Proposed Guidelines for Calibration and
Surveillance Requirements for Equipment Provided to Meet Item II.F.1,
Attachments 1, 2 and 3, NUREG-0737" dated August 16, 1982.
Safety Evaluation Report related to the operation of Beaver Valley
Power Station, Unit No. 2, Docket No. 50-412, dated October 1985.
Supplementary Safety Evaluation Report related to the operation of
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Beaver Valley Power Station, Unit No. 2, Docket No. 50-412,
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Supplement No. 1, dated May 1986.
Regulatory Guide 1.4, " Assumptions Used for Evaluating Radiological
Consequences of a Loss of Coolant Accident for Pressurized Water
Reactors".
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Regulatory Guide 1.97, Revision 3, " Instrumentation for
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Light-Water-Cooled Nuclear Power Plants to Assess Plant' and Environs-
Conditions During and Following an Accident".
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Regulatory Guide' 8.8, Revision 3, "Information Relevant to Ensuring
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.that' Occupational Radiation Exposure at Nuclear Power Stations will be
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As low As Reasonably' Achievable".
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4.
Post-Accident Sampling' System, Item II.B.3
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4.1 Fosition
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NUREG-0737, Item II.B.3, requires that-licensees shall have the
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capability to promptly collect, handle, and analyze post-accident
samples which are representative of conditions existing in the
reactor coolant and containment atmosphere.
Specific criteria are
denoted in commitments to the NRC relative to the specifications
contained in NUREG-0737,
4,2 Doruments Reviewed
The implementation, adequacy and status of the licensee's
post-accident sampling and monitoring system were reviewed against
the criteria identified in Section 3.0 and in regard to licensee
letters, memoranda, drawings and station procedures as listed in
Attachment 1 of this Inspection Report.
The licensee's performance relative to these criteria was determined
from interviews with the principal personnel associated with
post-accident sampling, reviews of associated procedures and
documentation, and the conduct of 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 (PASS)
which is a standard Sentry System modified by NUS Corporation. The
PASS has the capability to obtain pressurized and unpressurized
containment atmosphere and reactor coolant samples.
Liquid samples
can be obtained from the primary coolant hot legs, residual heat
removal system, reactor containment sump, the safeguards building
sumps, auxiliary building sump,' pipe tunnel sump, recirculation spray
pump discharges, and the reactor containment atmosphere.
The PASS
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sample cubicle containing the sampling cabinet and control panels is
located on the 718 -6" Level of the Primary Auxiliary Building.
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The PASS consists of the PASS control panel and all interconnecting
piping and valving necessary to selectively route the required
samples to the control panel for acquisition and/or on-line analysis.
Samples not analyzed on-line in the PASS panel can be analyzed via
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grab samples either at the primary chemistry laboratory if radiation
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levels permit or at the emergency response facility laboratory.
Backup sampling capability, via grab samples, exists for all on-line
analyses.
During normal plant operations the primary drains transfer
tank, volume control tank, or the liquid waste drains tanks are used
for sample return points.
During post-accident operations the
pressurizer relief tank or the reactor containment sump are the
preferred sample return points.
4.4 Findings
4.4.1
Performance Test
Grab samples of reactor coolant and containment atmosphere were
obtained during sampling system performance tests.
During these
tests licensee personnel exhibited the integrated ability to collect
and analyze samples within the constraints of NUREG-0737, II.B.3.
However, during the week of July 6, equipment malfunctions resulted
in a complete depressurization of the Unit-2 system.
Depressurization
resulted in the licensee being unable to demonstrate that the PASS
could acquire all of the designated samples at reactor operating
temperatures and pressures. The residual heat removal system pumps
were operational and samples from this system (at 60 psig) were
obtained and analyzed.
4.4.2
Sampling - General Observations
Pre-operational, surveillance, and training programs had been
conducted in accordance with the commitments documented in the
Unit 2 FSAR.
Licensee personnel exhibited very good under-
standing of the PASS and the type of health physics coverage
that would be needed during sample handlir,g.
However, the
following items were identified:
The licensee had identified and wcs tracking the need to
complete heat trace installation on the containment sampling
line.
The procedures for collection of reactor coolant and
containment atmosphere samples (BVPS-1/BVPS-2 CM Chapter 6
PASS Unit 2) contained in the Chemistry Manual (C.M. 2-6.1
through 2-6.12) had not been finalized.
In addition, numerous
procedure changes, to ensure sample collection and analysis
within the three hour time limit, have yet to be made.
The
planned procedure changes were considered when evaluating the
licensee's ability to collect samples.
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The licensee had not documented that the PASS piping and valving
system was capable of functioning at design " temperatures and
pressures.
The licensee had not verified the gas dilution factor.
4.4.3 Analytical Capability
The PASS provides an in-line analysis capability for total dissolved
gas and oxygen, pH, chloride and boron concentrations, and gross
radioactivity.
Radionuclides gamma spectrum analysis is performed via
grab samples at the primary chemistry laboratory or at the on-site
emergency response facility laboratory.
The required chemical analyses utilize the following principal
methods and/or equipment:
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Item
In-Line
Grab
Mannitol / Titration
Carminic Acid / Spectrophotometry
Ion Chromatography
Ion Chromatography
Dissolved
Gas
Gas Chromatography
Gas Chromatography
Gas Chromatography
Gas Chromatography
Gamma
Analysis
Gamma Spectroscopy
Gamma Spectroscopy
Verification of the licensee's ability to fulfill the chemical
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analytical requirements of NUREG-0737 II.B.3 was made in part by
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licensee analysis of spiked samples for boron and chloride.
The
licensee demonstrated that their grab boron analytical method and
their in-line chloride system meet the licensee's documented ranges
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and sensitivities. A comparison of the routine reactor coolant
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sampling system and the PASS will be performed and documented by the
licensee when sufficient radioactivity is present in reactor coolant.
The overall performance of the licensee's analytical capabilities
was acceptable with the following exceptions:
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The in-line boron analyzer was not operational and the in-line
gas chromatograph functioned improperly,
i.e.,
it could not
maintain baseline and the standard check was out of specification.
The gamma isotopic detector was calibrated without having an
approved procedure.
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4.5 Items For Improvement
Pass Sampling and Analysis
The licensee indicated the following matters will be reviewed and
clarification / improvements will be considered, as appropriate:
(50-412/87-50-01)
The procedures for performance of reactor coolant and
containment atmosphere sampling operations have yet to be
finalized and implemented, including revisions to the sequence
for collecting samples, verification of the gas dilution factor
and reference to appropriate procedures for handling high
activity samples.
Document that the PASS piaing and valving systems are capable
of functioning as designed at applicable reactor coolant
temperatures and pressures.
Perform a comparison of the routine reactor coolant system and
PASS sample analyses results when sufficient activity is
present in reactor coolant.
Ensure that the in-line boron analyzer and gas chromatograph are
operational.
Develop and implement a procedure for gamms isotopic detector
calibration.
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
monitors with an extended range designed to function during normal
operating and accident conditions. The criteria, including the
design basis range of monitors for individual release pathways, power
supply, calibration and other design considerations are set forth in
Table II.F.1-1 of NUREG-0737.
5.2 Documents Reviewed
The implementation, adequacy, and status of the licensee's monitoring
systems were reviewed against the criteria identified in Section 3.0
and in regard to licensee letters, memoranda, drawings and station
procedures as listed in Attachment 2.
The licensee's performance relative to these criteria was determined
by interviews with the principal persons associated with the design,
testing, operation, installation and surveillance of the high-range
gas monitoring systems; a review of the associated procedures and
documentation; and direct observation of the systems.
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5.3 System Description
Beaver Valley Unit 2 has two accident gaseous release pathways; the
supplementary leak collection and release system (SLCRS) vent and the
main steam discharge lines.
The SLCRS vent is monitored with a
General Atomics (GA) Wide Range Gas Monitor (WRGM) which provides
monitoring in concentrations ranging from 1.0E-7 to 1.0E+5 uCi/cc
Xe-133 equivalent.
The WRGM contains two sampling flowpaths.
The
effluent sampling points are equipped with isokinetic sample
nozzles, sensors to measure vent and sample flow rates using mass
flow techniques and redundant readout locally and in the control
The main steam monitor assembly consists of three NaI(TL)
room.
scintillation detectors adjacent to each steam line.
A steam flow
sensor is iastalled in each sample line.
These accident pathway
monitors are part of a digital, microprocessor-based radiation
monitoring system, which will detect, indicate, annunciate, store
and record effluent radioactivity data.
Phase I calibrations and pre-operational functional testing of the
WRGM and Main Steam Line Monitors had been completed.
Surveillance
procedures were also finalized.
The licensee, at the time of the
inspection, had not approved the test results for procedure P.O.
2.43.01, Sections N and 0, which relate to these monitors.
No test exceptions were noted.
Vendor supplied calibration reports
were available on-site and were used to develop monitor isotopic
efficiencies.
The licensee also adequately addressed the requirements to convert
the monitor reading to a time dependent release rate for dose
assessment purposes.
5.4 Findings
The installed system meets the requirements for high range noble gas
monitoring as discussed in NUREG-0737, Attachment II.F.1-1.
5.5 Recommendations for Improvement
The licensee indicated the following matters will be reviewed and
clarification / improvements will be considered, as appropriate:
(50-412/87-50-02)
Approve Preoperational Procedure P.O. 2.43.01, Sections N and 0,
" Radiation Monitoring System."
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6.
Sampling and Analyses of Plant Effluents, Item II.F.1-2
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6.1 Position
NUREG-0737, Item II.F.1-2, requires the provision of a capability for~
the collection, transport, and measurement of representative
samples of radioactive iodines and particulate that may accompany
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gaseous effluents following an accident.
These activities must be
performable without exceeding the GDC-19 dose limits to the indivi-
duals involved.
The criteria including the design basis shielding
envelope, sampling media, sampling considerations, and analysis
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considerations are set forth in Table II.F.1-2 of NUREG-0737.
6.2 Documents Reviewed
The implementation, adequacy and status of the licensee's sampling,
analysis system and procedures were reviewed against the criteria
identified in Section 3.0 and in regard to licensee letters,
memoranda, drawings and station procedures as listed in Attachment
3.
The licensee's performance relative to these criteria was
determined by interviewing the principal persons associated with
the design, testing, operation, installation and surveillance of the
systems, by reviewing associated procedures and documentation, by
examining personnel qualifications, and by direct observation of the
systems.
6.3 System Description
The licensee will monitor potential releases of airborne radioactive
particulate and iodines during an accident by an off-line gas and
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particulate monitor.
This monitor is part of the wide range gas
monitor (WRGM) assembly located on the 773'-6" elevation of the
auxiliary building.
This assembly is in turn part of the process
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and effluent radiation monitoring system.
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The elevated release particulate and iodine monitor includes two
off-line sampling paths from the 54"x48" exhaust duct.
Both paths
include isokinetic nozzles, sampling lines, pump, flow control valve,
particulate filter paper followed by a silver-zeolite filter cartridge,
and automatic isokinetic flow control.
The low activity /high volume
flow (1.67 CFM) path and the high activity / low volume flow (0.06 CFM)
path each are provided with three collection assemblies in parallel.
The collection assemblies in the high activity flow path are located
in individual shielded chambers.
The high activity flow path is
designed to come into service when the noble gas monitor indicates
greater than 1.0 E-2 microcuries per cubic centimeter.
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'6.4
Findings
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Within the scope of this. review, the following items were
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identified:
Appropriate sampling media'were being utilized
The licensee and.their contractor, Stone and Webster Engineering
Corporation (S&WEC), statv that ANSI N13.1-1969 recommendations.
had been followed in the design and installation of this
monitoring system and its sampling lines and that an evaluation
indicated that heat tracing and insulation were not required.
However, the inspector noted that the sample filter collection
assemblies were approximately ten to fifteen feet'from the point
of collection in the process stream.
The quarter inch 0.D.
sampling line traversed approximately thirty feet with numerous
abrupt changes in direction to connect these two points.
The
licensee has not determined line loss or line deposition
correction factors for particulate and radiciodines, as may be
needed to obtain results which can be. considered conservative
approximations'of the actual concentration of particulate and-
radioiodines in plant effluents under accident conditions.
The
licensee committed to developing such correction factors by mid
October 1987 by using the data base being accumulated by the
General Atomics (GA) Radiation Monitor System User's Group
(selecting data from systems similar in sample tubing length,
diameter, number of and radius-of-curvature of bends, and
degree of heat tracing and insulation).
The isokinetic rozzle calculations for the high activity / low
volume flow rate sample line used a nominal process flow rate
of'32,500 CFM; while the nominal process flow rate currently
determined by the licensee to be applicable is 59,000 CFM.
The
licensee stated that a request for information (RI) to the
contractor would be generated to determine what change if any,
needed to be made to the isokinetic control ratio for this
sampling path in the RM-80 data base log.
The radioactive particulate and iodine monitoring system is
designed to collect samples continuously.
The licensee's time
and motion study for obtaining a high activity particle and
iodine shielded collection assembly included travel time to the
WRGM skid, removal of a shielded chamber containing a collection
assembly (with dedicated tools), transport of the shielded
assembly to a location for analysis / storage (dedicated transport
device), and measurement of the gamma dose rate at one inch
above the shield assembly (dose rate v:, microcurie per cubic
centimeter (iodine) available in procedure REOP 4.1).
The
additional contribution to the dose incurred by sample collection
personnel while at the WRGM skid from the source term in the
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elevated release duct directly overhead would amount to less than
ten percent as stated by the licensee.
The licensee committed to
document the assumptions and evaluation used to support that
statement.
The licensee has determined that cartridges will not be counted
on a gamma spectrometer until they read less than or equal to
twenty-five mrem /hr on contact.
Gamma spectrometer counting
facilities with trained personnel and approved procedures are
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currently available at Unit 1 and at the Emergency Response
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Facility. At the time of this inspection, the gamma spectro-
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meter in the Unit 2 radiochemical laboratory was not yet fully
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operational, and none of the gamma spectrometers had as yet been
calibrated for counting the silver-zeolite cartridge used in the
WRGM. The licensee committed to having calibration records for
counting this cartridge before exceeding five percent power and
to having trained radiochemical personnel and approved procedures
for manual operation of the gamma spectrometer in the Unit 2
laboratory by the end of July 1987.
6.5 Acceptability
The installed system and capability meets the requirements of
NUREG-0737 Attachment II.F.1-2, except for the following:
Perform gamma spectrometer calibration for counting the silver
zeolite cartridges.
6.6 Recommendations for Improvement
The licensee indicated the following matters will be reviewed
and clarification / improvements will be considered, as
appropriate:
(50-412/87-50-03)
Determine conservative correction factors for line loss or line
deposition for particulate and radiciodines in the high
activity / low volume flow rate (HA/LVFR) effluent sampling path.
Evaluate need to change the isokinetic control ratio for the
HA/LVFR effluent sampling path and document assumptions and
resultant dose to sample collection personnel from the source
term in the elevated release duct above the monitoring skid;
have approved procedures for manual operation of the gamma
spectrometer in the unit 2 radiochemical laboratory and radio-
chemical personnel trained in those procedures.
Provide calibration and records for counting silver zeolite
cartridges on the gamma spectrometer.
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7.
Containment High-Range Radiation Monitor, Item II.F.1-3
7.1 Position
NUREG-0737, Item II.F.1-3, requires the installation of two in-
containment radiation monitors with a range of 1 rad /hr to 10' rad /hr
(beta and gamma) or alternatively 1 R/hr to 10' R/hr (gamma only).
The monitors shall be physically separated to view a large portion of
containment and developed and qualified to function in an accident
environment. The monitors are also required to have an energy
response as specified in NUREG-0737, Table II.F.1.3.
Table II.F.1-3
of NUREG-0737 also outlines specific high-range monitor calibration
criteria.
7.2 Documents Reviewed
The implementation, adequacy, and status of the installed
in-containment high range monitors were reviewed against the criteria
set forth in Section 3.0 of this report and in regard to. interviews
with cognizant licensee personnel, licensee letters, station
procedures, as-built prints and drawings as listed in Attachment 4
to this Inspection Report, and by direct observation.
7.3 System Description
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The licensee has installed two General Atomics high-range monitors
above the 767'-10" elevation of containment. These monitors are part
of the Digital Radiation Monitoring System (DRMS).
The system is
based on distributed microprocessors and redundant central processor
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units. The microprocessor will provide controls, alarms, indication,
and data processing.
7.4 Findings
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Within the scope of this review, the following was identified:
Licensee and vendor documentation, procedures, and data, discussions
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with cognizant personnel, and direct observation indicated that the
following criteria were met: detector range (R/hr), linearity of
energy response, redundancy (widely separated monitors {approximately
180 degrees) with unobstructed views of widely separated spaces
within containment), in-situ calibration, and vendor calibration.
The design and qualification criteria are discussed in section 9.0.
7.5 Acceptability
The installed monitors and capabilities meet the requirements of
NUREG-0737, Attachment II.F.1-3.
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8.
Improved -In-Plant' Iodine Instrumentation Under Accident' Conditions - Item
II.D.3.3
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8.1 -position
NUREG-0737, Item III.D.3.3 requires that each licensee shall provide
equipment and associated training and procedures for' accurately
determining the airborne iodine concentration in areas within the
facility.where plant personnel may be present during an accident.
8.2 Documents Reviewed
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The implementation, adequacy an'd status of the licensee's in plant
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iodine' monitoring under accident conditions were reviewed against
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the criteria'in Section 3.0 and in regard to the documents stated in
Attachment 5.
The licensee's' performance relative to these criteria
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was determined by:
interviews with cognizant personnel; review of
applicable procedures for in plant survey team emergency operations;
'and verification of equipment availability and storage.
_8,3
System Description
The licensee's-capability of obtaining samples was reviewed.
The
licensee;had assembled adequate. numbers of dedicated equipment and
supplies, including air samplers, silver zeolite cartridges, and
portable radiation detectors (HP-210_ probe connected to count rate
meter) to sample in plant for radioiodines. Wheeled carts containing
the previously. mentioned equipment and supplies were observed in
locked ' storage locations (2 each on the 730' elevation of the unit _2
South Offices and Services Building { west stairwell) and 2 each on
the Unit I turbine deck).
The. licensee's capability of analyzing samples was reviewed.
The
licensee has a procedure containing a graphical method to convert net
counts per minute (cpm) to an estimated microcuries (I-131) per cubic
centimeter value.
Gamma spectrometry of samples can take place at
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the Unit i radiochemistry laboratory, the Unit 1 HP counting room,
and the Emergency Response Facility (ERF).
The licensee has
committed to have this gamma spectrometer capability at the Unit 2
radiochemistry lab in the immediate future and eventually at the Unit
2 HP counting facility.
Calibrated SAM-2s were available in the Unit
1 instrument issue area and at the ERF.
8.4 Acceptability
The licensee meets the requirements specified in NUREG-0737, Item III
D.3.3.
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9.
QualityAssurancea5dDesignReview
9.1 Position
NUREG-0737, Items II.B.3 and II.F.1 Attachment 3 specify the design
and qualification criteria. for the Post Accident' Sampling System
-(PASS). and for the Containment High Range Radiation Monitoring System.
(CHRRM).
Specific criteria are denoted in commitments to the NRC
relative to.the specifications contained in NUREG-0737 and in
licensee commitments listed in Section 3.
9.2 Documents Reviewed
The inspector reviewed pertinent work and quality assurance records
-for the design,. procurement, qualification,' construction and
installation of the CHRRM and PASS to ascertain whether the records
reflect work accomplishments consistent with NRC requirements and -
licensee commitments.
Documents reviewed for this determination
include the basic criteria / commitment documents listed in Section 3
and Attachments 1 and 2.
9.3 Containment High Radiation Monitor (CHRRM)
9.3.1 'CHRRM System Design
NUREG-0737 provides the design and performance requirements for the
CHRRM.
It requires that there be two separate monitoring channels
which are independent from each other from detector to
instrumentation output indication and that each be powered from
separate Class 1E power sources.
Inspectior, was made by a review of
pertinent documentation cited in Attachment 2 and by visual
observations of the installation.
A walkdown inspection of the CHRRM outside containment was made to
inspect for proper equipment locations, cable and conduit routing,
,
identification marking, electrical separation, and terminations.
!
The documentation review and walkdown inspection of the CHRRM did not
disclose any areas of discrepancy between the design criteria and the
system as installed.
9.3.2
CHRRM Environmental Qualification (EQ)
Inspection was made to determine licensee environmental qualification
compliance with NUREG-0737, Appendix B, Criterion (1).
The inspector reviewed the EQ files for portions of the system
located within the containment harsh environment and also for
portions of the system located outside the containment in a mild
environment.
Pertinent documents reviewed include those listed in
Attachment 2.
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Within the scope of this' inspection, no violations or unresolved
items were identified.
9.3.3~ CHRRM Quality Assurance Review
Inspection of portions of the CHRRM System installation was made to
assure licensee compliance with Duquesne Light' Company' Site Quality
Control Procedures.-
Review of the procurement, receipt, and installation documentation
indicated proper QC inspection.
The surveillance and test program
under Procedure 2.43.01 is the pre-operational. test of the CHRRM
system.
Review of this documentation reflected active QC
involvement both in' tests completed and current tests in progress
with appropriate QC. witness and hold points in the program and
proper sign-off of QC documentation.
Within the scope of this inspection, no violations or unresolved
items were identified.
9.4 Post Accident Monitoring System
9.4.1
Installation
The physical installation of the PASS was inspected by walkdown of a
portion of the system outside containment in order to verify
compliance with the' licensee's design and to assess the quality of
work performed.
Items inspected included the containment
penetrations, sample tubing, cable trays, junction boxes, control
panels, control devices, wire terminations, identifications and
electrical. separations. The inspection also incl'uded a review of the
installation documents listed in Attachment 1.
Current calibration dates were observed on calibration stickers on
instrumentation and control devices both inside and outside of the
PASS local control panel. Within the scope of this inspection and
review no deficiencies were discovered; however there was a concern
for the design of the electrical system power feeds to the PASS
system as discussed in 9.4.2.
9.4.2 Power Sources to the PASS System
Criterion 3 of Appendix B to NUREG-0737 states that "The
instrumentation should be energized from station Class IE power
sources".
Review by the inspector diselosed that the PASS is energized from
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both Class 1E and Non IE sources as follows:
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- Class IE 125 volt d-c Power from UPS source orange Bus 2-1 powers the
in-containment solenoid-operated containment sample line isolation
valves.
- Class 1E 125 Volt d-c power from UPS source Bus 2-2 powers the
outside containment solenoid operated containment sample line
isolation valves.
- Non Class 1E 125 Volt d-c power from UPS source Bus 2-5 powers the
in-containment solenoid-operated sample line selection valves.
- Non Class 1E 125 Volt d-c pcwer from UPS source Bus 2-6 powers the
outside containment solenoid-operated sample line
control / selection / return valves.
- Non Class 1E AC Power from MCC-2-23.
This sources provides AC
power to PASS motor-operated valves, air-operated valves, pumps,
compressors, and instrumentation.
All of the above power supplies are considered to t e capable of
providing reliable power to the segment / portion of the PASS system
they feed.
However, the inspector identified the large number of
power supplies which are required to be operable in order for the
PASS system to operate add to the complexity of the system. As an
example, in order to take any sample all of the five power feeders
must provide power to a position of the system - a single failure in
any of the feeders disables the entire PASS sampling capability.
DLC0 committed to review their station operating procedures in order
to assure that a power feeder failure can be easily detected / traced
and overcome to permit PASS sampling within the three hour PASS
system sampling time design requirement.
This matter is considered
unresolved (50-412/87-50-04).
9.4.3
PASS Environmental Qualification (E21
Inspection was made to review licensee environmental qualification
compliance with NUREG-0737, Appendix B, Criteria (1) and with
licensee commitments to Regulatory Guide 1.97.
The inspector reviewed the EQ files for portions of the system
located within the containment harsh environment and also for
portions located outside the containment in a mild environment.
Documents reviewed included those listed in Attachment 1.
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The review of the system design,1 function and environmental
qualification also included a review of.NRC Supplemental Safety
Evaluation Report 5 which was transmitted to the licensee on June 11,
1987.
W'ithin the scope of_the review there were no discrepancies discovered
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between the items qualified-and those specified by.the. documents as
y
- requiring qualification.
However, there does appear to be a potential-
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problem with certain devices that do not require environmental
qualification, and yet must' operate under a harsh containment
environment ~for the PASS system to operate. 'The devices in question
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include the PASS in-containment scienoid-operated sample selection
valves and their connecting cables.
If these valves are not
satisfactory for operation within the harsh containment accident
environment and fail to provide the PASS system with samples, then
the sampling requirements of NUREG-0737 are.not satisfied.
The licensee indicated the following matters will be reviewed and
!
clarification / improvements will be considered. as appropriate:
(50-412/87-50-05),
This item is considered unresolved: ~
Verify acceptability of PASS in-containment valves which are
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currently not considered by'the licensee to require
environmental qualification but may be required to' operate under
accident conditions for purposes of sample. collection.
10.0. Unresolved Items
Unresolsed items are items for which additional information is needed.to
determine their-acceptability.
Two unresolved items are-discussed in
Section 9.
11.0 Exit Meeting
The NRC team met with the licensee representatives listed in Section 1.1
of this report at the end of the inspection.
The team leader summarized
the observations made during the inspection.
At no time during this
inspection did the inspectors provide any written information to the
licensee.
Licensee management acknowledged the findings and indicated
that appropriate action would be taken regarding the findings.
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Attachment 1
Documentation For NUREG-0737, II.B.3
Beaver Valley. Power Station Unit 2 FSAR, Sections 9.3.2 and 9.3.11.
Beaver Valley Technical Specifications, Section 6.8.
Chemistry Manual, Chapter 6, " PASS" (Draft)
Chemistry Manual, Chapter 4, " Analytical Procedure, Part 1" Issue 1;
Radiological Controls Manual, Chapter 5, Radcon Emergency
Operating Procedures, Attachment 2, Issue 1.
PASS Time and Motion Study, dated July 7, 1987.
P.O. - 2.14 A.02, " PASS Test".
Specifications
Specification No. 2BVS-114A Post Accident Sampling System, Addendum No. 4
May 22, 1984.
Equipment and Installation Drawings
,
10080-RM-99A,10080-RM-99B and 10080-RM-99E' Flow Diagrams, Post Accident
Sampling Piping.
AA10080-RE-1AM 120VAC One Line Diagram Sheet 5, Rev.10,1/87
AA10080-RE-1AR 125VOC One Line Diagram Sheet 1, Rev. 9, 5/87
AA10080-RE-1AR 125VDC One Line Diagram Sheet 2, Rev. 14, 3/87
AA10080-RE-1AR 125VDC One Line Diagram Sheet 3, Rev. 12, 5/87
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AA10080-RE-1Y 480VMCC One Line Diagram, Sheet 14, Rev. 12, 12/86
AA10080-RE-1J 480VUSS One Line Diagram, Sheet 3, Rev. 7, 12/86
AA10080-RE-1AA Standby Diesel 480 V MCC One Line Diagram, Rev. 8.
3/86
AA10080-RE-1AB One Line Diagram, Standby Diesel, 480V-Substa 2-5, Rev. 5,
3/87
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Environmental Qualification
BV-2417-Electrical Equipment Qualification Master List, Rev. 8 6/11/85
"
2BV-731-Environmental Qualification Documentation Package SDDF 2701-600-
731-092, Rev. 1 Control Panels - Systems Control.
2BV-555-Environmental Qualification Documentation Package SDDF2701.550-
555-019 Rev. 2, Tnermal Heat Tracing System.
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'2BV-816-F. environmental Qualification Documentation Package SDDF2701.J70-
816-346, Rev. 2, 600v Control Cable - EPR/FMR Insulation TypeT 600v)
. Control Cable - TEFZEL Insulation Type.
2BV-719 Environmental Qualif_fcation Documentation Package SDDF 2701-650-
799-188, Rev.
2'.
In'Line Solenoid Operated Valves.
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Attachment 2
Documentation for NUREG-0737, II.F.1-1
Beaver Valley Power Station Unit 2 FSAR, Chapters 11.5.1 and 11.5.2.4.2
Procedures
R.C.M. Chapter 5, REOP 4.1, " Emergency Operation of WRGM Assembly, "(Draft)
R.C.M. Chapter 4, Radcon Instrument Procedure 2.21, "DRMS, Effluent
Monitoring Subsystem," Issue 1.
R.C.M. Chapter 4, Radcon Instrument Procedure 2.18, "DRMS, RM-11 Descrip-
tion and System Operation, " Issue 1.
EPP/IP 2.6.1, " Dose Projection - Backup Methods," Issue 8.
Maintenance Surveillance Procedure 2 MSP-43.33-I, " Elevated Release Gas
Radiation Monitor 2HVS*RQ11098,C Calibration Operating Manual 16,
Supplemental Leak Collection and Release System."
System Testing and Calibration
GA Calibration Report E-255-961, Revision 2, "RD-72 WRGM," dated
January 1983.
P.O. 2.43.01, " Radiatio.1 Monitoring Systems Test CAT 1," dated May 1,1987.
IP.P. 2T-RMS-43-2.70, " Elevated Release Effluent Radiation Monitor Test",
dated March 17, 1987.
Calculation Packages
ERS-SFL-86-031 " Emergency Dose Assessment Source Terms for BVPS-2," dated
February 9, 1987.
ERS-SFL-86-033 " Conversion Factors for EPP/IP-2-6.1 for BVPS-2," dated
March 16, 1987.
ERS-SFL-86-026, " Unit 2 DRMS Isotopic Efficiencies," dated
February 9,1987.
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Equipment and Installation Drawings
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AA10080-RK-15A- Instrument Piping, Radiation Monitoring Rev. 6,
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January 15, 1987.
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AA10080-RE-428- Conduit Plan, Service Building, Purple Switchgear
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Room, Elev. 730'6", Rev. 12, March 1987.
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A410080-RE-42A- Conduit Plan, Service Building, Orange Switchgear
Room, Elev. 730'6", Rev. 12, March 1987.
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AA10080-RE-57AJ- Conduit Plan, Rod Control Building Elev. 773'6",
Rev. 5, April 1986.
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AA10080-RE-6CM- Wiring Diagram, Radiation Monitoring System, Loop 3,
Sheet 4, Rev. 3, March 1987.
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AA10080-RE-6CL- Wiring Diagram, Radiation Monitoring System, Loop 3,
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Sheet 3, Rev. 3, March 1987.
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AA10080-RE-1AM- 120 VAC One Line Diagram, Sheet 5, Rev. 10, January
1987.
2RMR*DAV206 - Digital Radiation Monitor Panel Mounting and Electrical
Hook-up at Elevation 730' Service Building.
2RMR*DAV207 - Digital Radiation Monitor Panel Mountirg and Electrical
Hook-up at Elevation 730' Service Building.
2BV-509A - Environmental Qualification Documentation Package SDDF
2702.890-509-097, Rev.1, . In-Containment iiigh Range Radiation Monitor.
2BV-731 - Environmental Qualification Documentation Package
SDOF 2701.600-731-093, Rev. 1, Control Panels - Systems Control.
BV-2417 - Electrical Equipment Qualification Master List, Rev. 8,
June 11, 1985.
Equipment Procurement and Receipt
E-6-8721 - Digital Radiation Monitoring System Components
E-6-9865 - Digital Radiation Monitoring System Components
Surveillance and Tests
Surveillance and Test Procedure Series 2.43.01
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Attachment 3
Documentation for NUREG-0737, II.F.1-2
Licensee Drawings
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S&WEC Dwg. No.12241-RK-328P-1-3, (Sh.1 of 3), WRGM Sampling Lines,
2HVS*RQIl09B(-), Auxiliary Bldg., EL. 773'-6"
Licensee Procedures
REOP No. 4.1, Emergency Operation of the WRGM Assembly.
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2MSP-43.33-I, Elevated Release Gas Monitor 2 HVS*RQIl09B, C
Calibration.
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RIP 2.18, DRMS RM-11 Description and System Operation.
Licensee Correspondence
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DLC to SWEC, (2DLC7676) Representative Sampling of Effluents, dated
November 8, 1984.
SWEC to DLC (2DLC-23770) Representative Sampling of Effluents, dated
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December 13, 1984.
Licensee Documentation
Personnel training records for R0 PCT Module 22, 42 DRMS
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S&WEC Calculation No. UR(B)-461, Effect of Post-LOCA Background
Radiation Level on the Elevated Release Monitor (2 HVS*RQ109A, B)
Setpoint.
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SWEC Calculation No. UR(B)-148-0, Auxiliary Building Post-LOCA Six
Month Integrated Doses for 2BVM-119 Equipment Qualification
Confirmation, pg. 68 Fig.12.
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Bid Specification for DRMS, Addendum No. 1, dated April 2, 1984.
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RM-80 Data Base Log, WRGM, Dwg. No. 0390-1010, Sh. 36-41.
SWEC Calculation Sheets for Isokinectic Nozzles (WRGM),
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2702.890-509-115A, Pgs. 156.
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DLC WRGM Time / Motion Study, June 16, 1987.
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Meeting Minutes of the WRGM Technical Subcommittee (GA's Radiation
Monitor System User's Group), February 1-3, 1987.
S&WEC Calculation No. UR(B)-441, Dose Received when Accessing the
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Elevated Release Monitor (2HVS*RQI 109B) Post-LOCA (Vital Access).
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Attachment 4
Documentation for NUREG-0737, II.F.1-3
Licensee Drawings
Survey Map 203401, Reactor Containment, EL. 767'10"
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Licensee Procedures
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RIP 2.19, DRMS-Area Monitoring Suasystem
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2 MSP-43.40-I, In-Containment Area High Range Radiation Monitor, 2RMR*DAU
206, Calibration.
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2MSP-43.41-I, In-Containment Area High Range Radiation Monitor, 2RMRDAU
207, Calibration.
Licensee Documentation
.Sorrento Electronics Energy Response Test and Dose Rate Calibration of
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Model RD-23, High-Range Radiation Monitor Detector, E-255-978 (Rev. 2)
June 1986.
General Atomics Transfer Calibration Procedure-lon Chamber Area Monitor.
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General Atomics Bench Alignment and Calibration of Log Picoammeter and ADC
Circuit Board, 0357-2170-01&O2
Test Procedure No. 2T-RMS-43-2.17 (Rev.1), Startup Proof Test, 2
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RMR*RQ206, In-Containment High Range Area Radiation Detector
P.O.-2.43.01-VII.B, Radiation Monitoring Systems Test.
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SWEC Calculation No. UR(B)-397-0, In-Containment High Range Area Radiation
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Monitor Limits, 2RMR*RQ206207.
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Attachment 5
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Documentation for NUREG-0737, III.D.3.3
Licensee Procedures
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EPP/IP 2.2, Dnsite Monitoring for Airborne Release.
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RP7.3, Air Sampling, Field Evaluation and Sample Assessment.
RIP 5.13, SAM-2/RD-22 I-131 Counting System.
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RIP 5.15, ND-6650 Counting System.
RIP 5.16, ND-668 Counting System.
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