ML20126J258
| ML20126J258 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 04/19/1985 |
| From: | Chueng L, Mark Miller, Nimitz R, Pasciak W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20126J236 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-3.D.3.3, TASK-TM 50-322-85-04, 50-322-85-4, NUDOCS 8506100543 | |
| Download: ML20126J258 (38) | |
See also: IR 05000322/1985004
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-322/85-04
Docket No. 50-322
-License No. NPF-19 Priority -
Category C
Licensee: Long Island Lighting Company
P. O. Box 618
Shoreham Nuclear Power Station
Wading River, New York 11792
Facility Name: Shoreham Nuclear Power Station
Inspection At: Shoreham, New York
Inspection Conducted: February 11-15, 1985
Inspectors: k Lbld
R. L. Nimitz,1enior Radiation
3/29/85
'date
Specialist (Team Leader)
73 a r 711eA
M. P Miller, RadTstion Specialist
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L.pS. Chueng, Rbactor Engineer date
A. P. Hull, Brookhaven National Laboratory
S. V. M 'no, Brookhaven National Laboratory
W. H. nox, Brookhaven National Laboratory
Approved by: 6Y
.J J ascfak, Chi'Ef) BWR Radiation
/9 f[
dage
P atection Section
Inspection Summary:
Inspection on February 11-15, 1985 (Report No. 50-322/85-04)
Areas Inspected: Special, announced safety inspection of the licensee's
implementation and status of the following task action items identified in
NUREG-0737: II.B.3, Post Accident Sampling Capability; II.F.1-1, Noble Gas
Effluent Monitors; II.F.1-2, Sampling and Analyses of Plant Effluents;
II.F.1-3, Containment High-Range Radiation Monitor; III.D.3.3, Improved
Inplant Iodine Monitoring. The inspection involved 208 hours0.00241 days <br />0.0578 hours <br />3.439153e-4 weeks <br />7.9144e-5 months <br /> onsite by three
region-based inspectors and three contractors from Brookhaven National
Laboratory.
8506100543 850605-
PDR ADOCK 05000322
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Results: No violations were identified. Several areas requiring improvements
were identified,
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DETAILS
1.0 -Ir.dividuals Contacted
The. individuals contacted during this inspection are listed in Attach-
ment I to this inspection report.
2.0 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-A'ccident 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 Inplant Iodine Instrumentation
under Accident Conditions
As part of the inspection, a review was performed to verify and validate
the adequacy of.the licensee's design and quality assurance program for
the design and installation of the Post-Accident Sampling System (PASS).
3.0 TMI Action Plan Generic Criteria and Commitments
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.
- 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 Oper-
ating Reactors, NRC, to all Operating Power Plants, dated October 30,
'1979
"
- Letter from Darrell G. Eisenhut, Director, Division of Licensing,
NRR, to Regional Administrators, "Preposed 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.
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. Regblatory Guide 1.3, " Assumptions Used for Evaluating Radiological
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Consequences of a Loss of Coolant Accident for Boiling Water,
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Reactors"
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Regulatory Guide 1.97,' Revision 2, " Instrumentation for Light-Water-
Cooled Nuclear Powe'r Plants to Assess Plant and Environs Conditions
f' During.and Following an Accident" ,
.
' Regulatory Guide 8.8, Revision 3, " Info'rmation Relevant to Ensuring
that Occupational Rabiation Exposure at Nuclear Power Station will
, be As Low As Raasonably Achievable"
.
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NUREG:0420, " Safety Evaluation Report, Related To The Operation Of
Shorehah Nudlear Power Station, Unit No. I'," Supplement No. 1,
'E September 1981- -
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NUREG-04 0,' " Safety Evaluatior. Report,' Related To The Operation Of
Shoreham #cclear Power Statien, Unit No.1," Supplement No. 4,
September 1983.
Postl Ahcident Sampling System, Item II.B.3
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4.0
_ 4.1 Poil][lon ,
NUREG-0737,-Item II.B.3, specifies that licensees shall have the
~
. 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
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denoted in commitments to the NRC relative to the specifications
contained in NUREG-0737.
n
Documents Reviewed
'
The implementation, adequacy and status of the licensee's post-
accident sampling-and 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
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of this Inspection Report.
m
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.
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! 4.2 Findings j
Within the scope of the review, the following items were identified:
4.2.1 System Description and Capability
The licensee has installed a post-accident sampling and analysis
system which was designed and built by its architectural engi-
i neering firm, Stone & Webster. The system is situated in a
Post-Accident Sampling Facility (PASF) which is located in an
extension to the Reactor Secondary Containment Building. The
facility is provided with its own radiation monitoring, venti-
lation and effluent air filtration systems. The system is
designed so that it can obtain pressurized and unpressurized
samples of reactor coolant from the jet pump and the RHR System.
Atmosphere samples can be obtained from the drywell, suppression
pool and reactor building.
The system is designed to provide for the on-line analysis of
coolant chlorides, boron and pH. A capability for backup
analysis is available at the onsite radiochemistry laboratory
and at an off site laboratory, with the exception of hydrogen. l
Redundant containment hydrogen analyzers provide a capability '
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for analysis of containment hydrogen.
The system is designed to provide diluted grab samples for off-
line isotopic analyses.
The review of the design of the system and facility indicates
that the personnel exposure guidance of 10 CFR 50, General Design
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Criterion 19, would be met if the system was utilized properly.
4.2.2 PASS Performance Testing
Grab samples of reactor coolant and of the drywell atmosphere
were collected during an operational test on February 13 and
14, 1985. Due to system testing and on going modifications,
the on-line analysis capability could not be fully tested.
During the tests, it was verified that licensee personnel did
possess the ability to collect and analyze samples, using the
on-line system, within the time constraints of NUREG-0737,
II.B.3. This verification was based on a walkthrough of all
applicable procedure steps.
4.2.3 Sampling System
4.2.3.1 Reactor Coolant
The reactor coolant sampling system is designed to provide
samples of liquids and dissolved gases during all modes of
operation.
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The system operates in the following manner: A sample point
is selected from the sample panel and flow from the point is
initiated. The sample is cooled via a heat exchanger, then
directed to a radiation detector for initial dose rate
determination. The sample is then isolated and dilution,
as necessary, is initiated. As dilution is progressing, the
sample radiation dose rate is determined. Dilution continues
until a preselected radiation dose rate is obtained.
The review of the licensee's capabilities relative to reactor
coolant sampling, sample transport, and sample analysis iden-
tified the following matters requiring licensee attention.
The licensee should resolve these matters to demonstrate
conformance with the guidance contained in NUREG-0737, Item
II.B.3:
Representativeness of Samples Collected (50-322/85-04-01)
Evaluate and establish appropriate sample system purge
times to ensure a representative reactor coolant sample.
Place such purge times in appropriate procedures.
System purge times have not been determined on the basis
of an assessment of the "line plus sample" volume and
flow rate. The procedure relies on the establishment of
a constant reading from an installed on-line radiation
detector as an indication of the representativeness of a
sample. Under certain conditions, this reading may not
ensure a representative sample.
Evaluate and modify the system and applicable procedures
to provide for acceptable reactor coolant dissolved gas
quantification.
In the dissolved gas sampling procedure, the stripped
gas from a 300 cm3 reservoir is released into a closed
loop, which is initially evacuated to 5 psi. Based on
the system design and operation, the ability of the gas
to reach equilibrium throughout the sample loop is
questionable. Also, during a " feed and bleed" procedure,
using nitrogen, there is a change in system pressure which
is not accounted for in the calculations.
Revise reactor coolant sample collection procedures to
ensure samples of relatively low dose rates can be
collected consistent with sample dose rate limits speci-
fled in procedures.
The procedures for collection of a reactor coolant sample
for laboratory analysis require dilution of the sample
down to a dose rate 0 0..i mR/hr. Although installed
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instrumentation is capable of resolving such a low dose
rate, it is not clear that radiation sources in the PASS
Facility would allow such a low dose rate to be realized.
Consequently,-current procedures preclude collection of a ,
sample for onsite laboratory analysis, if such a dose I
rate is not realized. !
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System Operation (50-322/85-04-02)
- Evaluate adequacy of reactor coolant sample water sources.
Provide backup water sources as needed.
Water from the condensate storage tank (CST) is used for
sample dilution, system flush, and to cool the incoming
reactor coolant. During a large break LOCA, this source
of water might not be available.
Sampling (50-322/85-04-03)
- Provide procedure guidance for collection of undiluted
reactor coolant samples for onsite laboratory analysis.
The post-accident sample procedure does not clearly
indicate how the undiluted liquid grab sample would'be
collected for backup analysis onsite.
SampleTransport(50-322/85-04-041
Establish and approve procedures for transporting highly
radioactive samples to off-site analysis facility.
4.2.3.2 Containment Air
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Atmosphere samples can be obtained from the Drywell, Reactor
Building and the Suppression Pool. The system operates in
the following manner: ane of seven sample point locations
is selected from the sample panel. The sample passes through
heat traced lines to the sample panel. The incoming sample
dose rate is' determined. Based on dose rate, the sample is
diluted as necessary as the sample is recirculated through
heat traced lines. As dilution is progressing, the sample
dose rate is determined. Dilution continues until'a pre-
selected dose rate is obtained.
The review of the licensee's capabilities relative to con-
tainment atmosphere sampling identified the following matters
requiring licensee attention. The licensee should resolve
these matters to demonstrate conformance with the guidance
contained in NUREG-0737, Item II.B.3:
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Sample Representativeness (50-322/85-04-05)
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- Evaluate and establish appropriate sample system purge
times to ensure representative atmosphere samples. The
purge times have not been determined.
- Evaluate and modify the system to ensure acceptable
atmosphere sample dilution. During the dilution process,
it is not clear that samples will be properly evaluated
for dilution. The sample is recirculated during dilution;
however, it is not clear that all portions of the sample
are recirculated.
4.2.4 ' Analytical Capability
4.2.4.1 Chloride
The system provides for the on-line analysis of chlorides
by a Dionex Ion Chromatograph. At the time of this inspec-
tion, the system was being modified by the addition of a
pump to increase the pressure differential across the column.
Therefore, it was not tested.
Arrangements have been made for chloride analysis of a
grab sample at nearby facilities at Brookhaven National
Laboratory.
Components of the ion chromatograph have been installed by
hanging them on the rear panel shield wall of the system
where they are vulnerable to disruption during maintenance
or inspection of other systems.
The applicable system operating procedure, SP 73-720.14,
does not clearly indicate when the system can be effectively
operated. Paragraph 8.1.1.7 states, "If pressure does not
increase.to severallhundred psi, the pump must be primed."
The quantitative interpretation of "several hundred psi"
was not well understood by the technicians who operate the
system.
Based on the above, the licensee should perform the following
(50-322/85-04-06):
- After the system modifications are complete, the on-line
analyzer should be tested to demonstrate its ability to
perform chloride analysis within the specified accuracy.
- A cover should be placed over the plastic tubing com-
.ponents of the ion chromatograph to prevent damage to
them.
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- The PASS procedures should. quantitatively state action-
level criteria _(i.e., eliminate such terms as several
hundred psi).
~4.2.4.2 Boron
The system provides for the on-line analysis of boron by an
10rion Model 1610 Baron /pH analyzer. At the time of the
inspection, the boron analysis capability could not be demon-
strated due to equipment failure. The ability of the system
to determine boron concentration in the presence of multiple
acids and bases is questionable. The analyzer determines
the concentration of boron by the measurement of the pH and
the concentration of sodium. This could represent a source
of error,.since the acidity of the coolant may not be
- totally due to boric acid. Also, the base components of the
solution may be more than sodium hydroxide. No test data
were presented to demonstrate the performance of the ana-
lyzer in a multi-acid / base environment and in the presence
of elements in the Standard Test Matrix.
The licensee performance criterion for_ acceptance of the
system is in part based on millivolt readings, i.e., 75
10 mv. This could not be directly related to the commitment
to perform boron analysis within an accuracy of 5% as
there was not a direct correspondence between the millivolt
reading and boron concentration.
The reagent and solution containers associated with the
system were not labelled to indicate the type of solution
and its concentration. The system operators were not
familiar with the container contents.
A Fluoroborate Selective Ion Electrode is used for the boron
analysis of the grab sample. The results of the analysis'of
spiked samples are contained in Attachment 3 to this Inspec-
tion Report.
Based on the above, the licensee should perform the following
(50-322/85-04-07):
- The boron /pH analyzer should be tested to determine its
response to a multi-acid / base mixture which includes the
elements in the Standard Test Matrix. Also, the ability
of the system to meet the analysis acceptance criteria
commitment should be demonstrated.
- The reagent and solution containers should be clearly
identified.
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4.2.4.3 pH' Analyses
The system provides for the on-line analysis of pH with an
Orion Model 1610; Boron /pH analyzer. Since the boron and pH
-analyses are integrated, similar concerns exist for both'
types of' analyses. -(See Section 4.2.2.2).
Provisions have been made for pH analysis of an undiluted
grab sample with a flat surface electrode.
The licensee committed to an accuracy for pH to with 0.01
pH units throughout the range of 0-14 pH. This commitment
may be unnecessarily restrictive.
Based on the above, the licensee should perform the following
(50-322/85-04-08):
- The capability for the on-line analysis of pH should be
demonstrated. The commitment to measure concentrations
to within an accuracy of 0.01 pH units should be
reassessed.
- 4.2.4.4 Gross Gamma and Isotopic Analyses
.The isotopic analyses of a diluted grab sample is conducted
off-line by multichannel analyzer. In view of the start-up
-status of the reactor, there was no activity in the coolant.
The analyzer's performance could not be demonstrated, since
there_was no liquid nitrogen in the Dewar.
The following observations were made during a dry-run of the
analysis procedure:
a. There was no stated commitment for the accuracy of the
isotopic analysis.
b..The sample flask was placed directly in the detector
shield without first placing it in a plastic bag to
minimize internal contamination spread.
c. The detector shield was purged with service air which may
contain noble gases under accident conditions.
Based on the above, the licensee should perform the following
(50-322/84-04-09):
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- When the plant becomes operational and sufficient activ-
ity~has built up in the coolant, appropriate tests should
be conducted to demonstrate the capability of the system
to obtain. representative samples, based on a comparison
of isotopic analysis of normal and PASS samples. The
. accuracy of-the analysis should also be stated.
- Provisions should be made in the procedure to protect the
Ge-Li detector from contamination.
- Nitrogen should be used to purge the detector shield under
accident conditions.
4.2.4.5 Hydrogen and Dissolved Gas
The system is designed to establish the total volume of
dissolved gas in the coolant. The core damage procedure and
the form of the input data to it were not evaluated during
this inspection since the licensee had previously committed
to have this procedure available after the first refueling.
The analysis of hydrogen in the containment atmosphere is
provided for.by redundant on-line continuous. analyzers, as
required by item II.F.1-6. No provisions have been made for
the conduct of hydrogen analysis on grab samples. This
arrangement has been considered and found acceptable by NRR.
Based on the above, the license should perform the following
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(50-322/85-04-19):
- The core damage procedure-should be finalized before
completion.of the first refueling outage. An evaluation
should be conducted to assure that all necessary input
data is available and in the proper format.
- Obtain documented approval from NRR which allows the
licensee to solely use in-line hydrogen analysis metho-
dology to satisfy the requirements of NUREG-0737, IIB.3.
-(Clarification Paragraph 2).
4.2.5 Addi tional - Findings
The following additional findings were identified. The licensee
should review these findings and take appropriate action (as
necessary) to resolve the concerns identified (50-322/85-04-10):
- Evaluate the acceptability of using station supplied
breathing air.
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EPIP 2-12 and the System Design Description indicate that
service air would be the primary source of breathable air in
-the PASF under-accident situations. However, this air supply
may'not be readily available because of the following:
a. Forty-five minutes are required to test the quality of
this air supply.
b. The service' air system is non-safety related and isolates
automatically under certain emergency conditions.
c. As previously indicated, it may contain noble gases.
- Perform a " time and motion" study for collection of undiluted
reactor coolant samples to ensure the personnel dose accept-
ance criteria of General Design Criterion 19 are met.
During the collection of the grab sample, personnel would
be exposed to unshielded sample lines containing undiluted
coolant. A detailed " time and motion" study, which covers
all aspects'of the collection, transportation and analyses
-of samples has not been conducted.
- Tag all' appropriate valves in the PASS facility.
There are large numbers of valves in the system whose
positions have not been specified in the PASS operations
procedure. These valves may not be accessible, once sample
collection and analysis has been initiated. Provisions have
not been made'in the procedure to assure the proper position
or line-up of these valves before sample collection is
started. In addition, one valve, #A0V0BO, is.not listed in
the procedure. However, the System Description contains a
warning which states "Do not open A0V0B0 while operating or
prior to operating pump P-006."
Note:
Ouring the review, a list of all valves and their expected
positions was generated.
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- Ensure the installed oxygen analyzer can withstand full
reactor coolant system pressure. No documentation was
provided to demonstrate that the actual installed system
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would withstand RCS pressure (@ 1100 psi).
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- Approve calibration procedures for the installed PASS radia-
tion monitors.
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.The calibration procedures for the newly installed dilution
process radiation monitors and the CST water radiation
monitor were in a draft form.
- Consider movement of the heat trace temperature indicator
to the operating floor elevation of the PASS Facility. Dur-
ing accident conditions, technicians in breathing apparatus
may need to climb a circular stairway to obtain temperature
. readouts. Also, during the sample collection drill, the
technicians could not locate the indicator.
- Clarify valve position guidance in Procedure EPIP 2-11.
During a sample collection, " Realign" was misinterpreted as
' leave in original position. This resulted in a sample being
unintentionally flushed from the sample system.
- Evaluate the need for use of respiratory protection equip-
ment during the disconnecting of pressurized samples from
the. system. Respirators were not required during the
disconnection.
- Correct the incorrect reference in Procedure EPIP 2-11,
paragraph 5.4.4.16. The paragraph refers to the wrong
paragraph number for further guidance.
- Complete labeling of all readouts and monitors on the PASS
panel. A significant number of readouts were not labeled on
the panel.
- Establish several operating / sample collection procedures for
the PASS Facility. The current operation / sampling is con-
trolled by one procedure of 150 pages. The use of this
single procedure is cumbersome and difficult, as evidenced
by observation of licensee technicians attempting to use it.
The use of the procedure was further complicated by incorrect
references contained therein (see above).
- Clarify the sample analyses to be performed by Brookhaven
National Laboratory and make provision for periodical
updating of the agreement for these analyses.
" * Establish a designated area for storage of PASS samples.
- Review the training of technicians in use of portable oxygen
detectors. The technician using the detector to determine
habitability of the PASS Facility was uncertain of the
appropriate percent oxygen limit for normal, unassisted
breathing.
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5.0 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.
Documents Reviewed
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The implementation, adequacy, and status of the licensees 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 4 of this Inspection Report.
The licensee's performance relative to this criteria was determined
by interviews with the principal persons and consultants associated
with the design, testing, installation and surveillance of the high
range gas monitoring systems, a review of the associated procedures
and documentation, an examination of personnel qualifications and
direct observation of the systems.
5.2 Findings
Within the scope of this review, the following was identified:
5.2.1 Description and Capability
There are four plant effluent inputs into the station vent.
They are the turbine building, radwaste building, reactor build-
ing normal ventilation system (RBNVS), and the reactor building
standby ventilation system (RBSVS). The turbine and radwaste
buildings and the RBNVS input into the station vent below the
input of the RBSVS, following which the vent is sampled independ-
ently by the RE-126 high range post-accident monitor. The RBSVS
is sampled prior to release to the stack by the RE-134 high range
post-accident monitor.
These effluent monitors consist of two Kaman Instrument
Corporation High Range Effluent Monitors, Model KMG-HRH, and one
microcomputer, Model KEM-P. Each sampling skid contains a high
range noble gas sampler assembly.
The high range noble gas detector is a GM tube which is mounted
so that it views the inlet / outlet tubing to the sample chamber,
with the necessary shielding to cover the range of 101 to 10'
uCi/cc for Xe-133. The vendor has supplied National Bureau of
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Standards (NBS) traceable calibration data, using Xe-133 gas and
solutions of Ba-133, Cs-137, and Co-60 at single activity-
levels. .In addition, the_ vendor has supplied the licensee.with
NBS traceable isotopic sources for routine calibration checks.
In the overall vendor calibration of the high range noble gas
channel, a conversion based on' CPM /uCi of Xe-133 equivalent
20% is established. This information is communicated to a
Modcomp computer which collects' data for all the plant process
radiations monitors.
The computer readout of the accident monitors is rapid and triple
redundant. .The Modcomp system is double redundant, since there
is a second computer to sense a failure and back up the primary
one. The Kaman microcomputer makes the system triple redundant.
Since the Kaman computer is interfaced via an isolated analog
signal (the Modcomp digitizes Kaman signals), if both Modcomp
systems fail, the accident monitors have a " stand alone"~
capability. In this event, readout would be possible at two
racks located adjacent to the control room. Backup AC power is
provided to the entire Kaman hardware. Normally, signals from
the noble gas monitors are processed by the Modcomp software
during accident modes to calculate off-site doses automatically.
This provides the licensee with a rapid method of data reduction
and dose assessment. In the event of a total Modcomp computer
failure, a backup Hewlett Packard HP-85 based system exists to
perform the calculations after manual entry of release data.
Procedures call for the collection of a grab gas sample in a
Marinelli configured chamber in addition to the filter cartridge;
however, no calculations were provided to demonstrate that the
activity in the Marinelli would be low enough to allow it to be
handled, transported, and analyzed. (See Section 6 of this
inspection report for a discussion of effluent sample
. capability.)
5.2.2 Acceptability
The installed system meets the guidance for_high range noble gas
monitoring as contained in NUREG-0737, Attachment II.F.1.1.
However, the following matters should be reviewed and resolved
(50-322/85-04-11):
- The Operating and the EPIP procedures for the RE-126 and
RE-134 effluent monitors differ. The procedures differ
relative to their guidance for changing out filters. One
procedure says valve out the sample pump, whereas the other
procedure says to manually shut off the pump.
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Complete onsite flow calibration of ' sample flow paths. Flow
calibration should be implemented for the 650 cm'/ min sample
paths of RM-126 and RM-134.
Consider use of computer ass.isted/ generated decay corrections
for Modcomp software in order to accurately quantify the
source term. Currently, no decay correction is applied to
the nuclide library used by the Modcomp software. Modifica-
tion of the library to allow for radioactive decay will
reduce the analytical error. The correction could be made
by hand via incorporation of a gamma spectra. This would be
time consuming and prone to errors.
6.0 Sampling and Analyses of Plant Effluents, Item II.F.1-2
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 particulates that may accompany gaseous
effluents following an accident. It must be performable without
exceeding specified dose limits to the individuals involved. The
criteria including the design basis shielding envelope, sampling
media, sampling considerations, and' analysis considerations are set
forth in Table II.F.1-2.
Documents Reviewed
The implementation, adequacy and status of the licensee's sampling
and. analysis system and procedures were reviewed against the criteria
identified in Section 3.0 and in regard to licensee letters, mem-
oranda, drawings and station procedures as listed in Attachment 4 to
this Inspection Report.
The licensee's performance relative to these criteria was determined
by interviewing the principal persons and consultants associated with
the design, testing, installation, and surveillance of the systems
for sampling and analysis of high activity radioiodine and particu-
-late effluents, by reviewing associated procedures and documentation,
by examining personnel qualifications, and by direct observation of
the systems.
6.2 Findings
Within the scope of this review, the following was identified:
6.2.1 Description and Capability
The flow to both the RBNVS sampler (RE-126) and the RBSVS
sampler (RE-134) is provided with an isokinetic nozzle, and is
controlled on the basis of normal flow in the plant vent and
RBSVS respectively. The main sampling lines are 1" diameter
. - - . _ _ . -
. _.
--
..- 3
17
with smooth bends and long vertical drops. A pump located near
the sampling skid takes a second isokinetic sample from the main
sampling line and is delivered to the sampling skid at 650
cm'/ min. Although the system was provided with a vendor cali-
bration for'the isokinetic flow path, routine calibration of the
650 cm*/ min . flow path has not been implemented by the licensee
on either the RM-126 or RM-134 skids.
Both sampling skids contain three particulate and iodine sam-
pling channels. Each consists of a 2" diameter particulate
. filter paper followed by a 2.25" by 1" deep charcoal filter
cartridge. The collection assembly is located in a shielded
sample chamber to protect personnel during filter changes. Each
channel has a GM tube to sense radiation buildsp on the filter,
and alarm is set to notify the control room of high radiation at
the filter. Upon alarms, the micro computer will time a 30
minute sample and automatically switch to the next filter. If
no filter change occurs after 90 minutes, the system would con-
tinue to sample the third filter.
The collection assembly has a quick release and a remote hand-
ling tool for the removal and transfer of a potentially high
level sample. A portable shield is located near the sampling
skid. The procedure calls for its use when samples exceed 20
mR/hr; however, the licensee has not demonstrated calculations
of the dose rates which might arise-from the filter cartridges
under accident sampling conditions. The licensee has not
implemented any procedures for the analysis of a filter which
produces levels greater than 0.5 mR/hr, or for the handling and
transport of samples greater than 20 mR/hr.
The licensee has demonstrated calculations which show the
radiological accessibility for filter changes under accident
conditions.
Due to the current configuration of the radiation monitoring
alarms in the control room, this system is in continuous
operation. The licensee is therefore reluctant to employ silver
zeolite cartridges in its routine.
During the review, it was found that Channel C on the RM-126 had
an inoperative GM tube.
6.2.2 Acceptabili+v
The installed system can be considered to meet the guidance
specified in NUREG-0737, Attachment II.F.1-2 if the licensee can
satisfactorily resolve the following matters (50-322/85-04-12):
- Establish and implement procedures for analysis of highly
radioactive effluent samples. Currently, no procedures have
been established for analysis of such samples.
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- Perform a " time and motion" study, as necessary, to
ensure the personnel dose guidance specified in 10 CFR 50,-
Appendix A, General Design Criteria 19, would be met dur-
H ing effluent sample, collection, transport, handling, and
analysis. The time and motion study should use source term
guidance specified in NUREG-0737.
- Replace the inoperable detector for Channel C of Rm-126.
Establish surveillance procedures (as necessary) to ensure
prompt replacement of inoperable detectors.
7.0 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 maximum range of 1 rad /hr to
10' rad /hr (beta and gamma) or alternatively 1 R/hr to 107 R/hr
(gamma only). The monitors shall be ph9sically 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.
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 proce-
dures, as-built prints and drawings as listed in Attachment 5 to this
Inspection Report, and by direct observation.
7.2 Findings
Within the scope of this review, the following was identified:
Two Kaman ion chamber detectors have been installed inside the
drywell near the personnel and equipment hatches to provide separa-
tion and accessibility for routine mainte,ance. The detectors met
the following design criteria of NUREG-0737, namely, energy response,
range, calibration, and separate vital instrument power supplies.
However, the detector assemblics were not environmentally qualified.
The licensee informed the NRC of this deficiency and has proposed a
new configuration that is expected to be environmental _ly qualified
and installed by November 30, 1985.
Vendor calibration data certified that the detector model was type-
test calibrated over the prescribed exposure range (i.e., 1 R/hr to
10' R/hr) and individually calibrated for at least one point per
decade of ranges between 1 R/br and 10' R/hr. An internal source
supplies a 2 R/hr field as a one point calibration check below 10
R/hr.
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>
Procedure EPIP 1-11 provides a method to estimate core damage based
on the drywell monitor (s) data and plant specific core damage assess-
ment curves. Training records indicated annual requalification in
the emergency implementing procedures was required for Control Room
personnel, Shift Technical Advisors and Radiological Assessment
Coordinators.
The licensee should resolve the following matters to demonstrate
conformance with the guidance contained in NUREG-0737, Item II.F.1-3
(50-322/85-04-13):
- Install environmentally qualified high range detector assemblies
'by November 30, 1985.
8.0 Improved In-Plant Iodine Instrumentation Under Accident Conditions,
Item III.D.3.3
8.1 Position
NUREG-0737, Item III.D.3.3, requires that each licensee shall' provide
equipment and associated training and procedures for accurately deter-
mining the airborne iodine concentration in areas within the facility
where plant personnel may be present during an accident.
Review Criteria
The implementation, adequacy and status of the licensee's in plant
iodine monitoring under accident conditions were reviewed against the
criteria in Section 3.0 of this report and in regard to the documents
identified in Attachment 6 to this Inspection Report. The licensee's
performance relative to these criteria was determined by:
- Interviews with cognizant licensee personnel;
- Review of applicable operational and emergency plar. procedures;
- Review.of applicable lesson plans and training records;
Direct observation of performance during a walkthrough; and
- Verification of equipment availability and storage.
8.2 Findings
,
Within the scope of this review, the following was identified:
The licensee's program to sample radioiodine against a background of
highly radioactive noble gases was found to be generally acceptable.
Appropriate sample media such as silver-zeolite cartridges, analytical
instrumentation and sufficient personnel training / qualifications were
observed during the inspection.
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Within the scope of the review, the following matter was identi-
fied which should be reviewed and resolved by the licensee
(50-322/85-04-14):
- Review the adequacy of calibration of battery powered air
samplers (RADECO H809-C). These samplers are flow calibrated
with charcoal cartridges in place. However, during accident
situations, silver zeolite cartridges may be used. The flow
calibration may not be valid when the zeolite cartridges are
used.
9.0 Quality Assurance (QA) and Design Review
9.1 QA Documents Reviewed
The inspector reviewed pertinent work and quality assurance records
for the design, construction and installation of the Post-Accident
Sampling System to ascertain whether the records reflect work accom-
plishments consistent with NRC requirements in the areas of receipt
inspection, equipment qualification, installation and inspection.
The documents reviewed are listed in Attachment 7 to this Inspection
Report.
9.2 Findings
Within the scope of this review, no violations were identified.
9.3 Environmental Qualification of the PASS Containment Isolation Valves
9.3.1 The following containment isolation valves are required to be
operable for post-accident sample collection:
IT48*S0V-126A,B; 1T48*S0V-127A,B; 1T48*S0V-128A,8;
1T48*S0V-129A,B; 1T48*S0V-130; 1T48*S0V-131;
1821*S0V-313A,8; IE11*S0V-166A,B; 1E11*SOV-167A,B.
These are Valcor solenoid valves, Models V526-5295 and V526-5683.
The inspector reviewed the environmental qualification (EQ) data
file for these valves which contain the following:
- Valcor Qualification Test Reports QR52600-5940-2,
Revision C, and QR52600-515
- Wylie Lab's thermal aging analysis of organic materials used
in these solenoid valves
- Radiation aging analysis
- Radiation and temperature profiles after the LOCA and Help
Accidents
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- Stone and Webster's EQ Status Report, pages 74 through 83,
dated June 21, 1983
- Environmental. Qualification Report Evaluation Form (EQREF)
TR-S1554-2-01, Revision 2, dated May 28, 1982.
9.3.2 Findings
The solenoid valve models installed differed from the solenoid-
valve model tested. The licensee elected to qualify the
installed models by similarity analysis. Comment #1 of the EQREF
TR-S1554-2-01 states "Section 3.2 of the test report provides
justification that the components tested represent the components
being reviewed."
Proper justification of similarity was not contained in Section
3.2 of the test report, nor can it be found in other parts of
the EQ file. The licensee subsequently obtained and provided
a facsimile of Valcor similarity analysis SK 10702, dated
July 16, 1981, from Stone and Webster in Boston. The justifi-
cation was acceptable.
Since these solenoid valves are qualified by similarity analysis,
the Valcor similarity analysis should be in the EQ data file and
Comment #1 of EQREF TR-S1554-2-01 should be revised to reflect
the correct location of the similarity analysis.
This is an open item pending NRC verification that EQ data file
is updated accordingly (322/85-04-15).
9.4 Instrument Calibrations and loop Tests 9.4.1 The inspector examined the instruments of the PASS and reviewed
the calibration and loop check records of those instruments to
ascertain that the PASS instruments were properly maintained.
9.4.2 Documents reviewed for this determination.
--
Shoreham Preventive Maintenance Program, SP No. 12.015.01,
Revision 7, dated February 6,1985
--
Shoreham Instrument Loop Calibration, SP No. 41.005.01,
Revision 1, dated December 13, 1982
--
Shoreham Preventive Maintenance (Computer P'intout
Schedule), dated January 30, 1985, Pages 356, 361 and 362
--
Shoreham Scheduled Activity Worksheets (SAWS) for PASS
Process Instruments, No. 1071.200-6.413
--
Shoreham Test Loop Diagram IZ96 for PASS Process
Instruments
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Calibrat1on'and Loop Check Records of PASS Process .
. Instruments
,
9.4.3 Findings
9.4.3.1 Shoreham Technical Specification Paragraph 6.8.4 states in
part "The following programs shall be established, imple-
'
mented, and maintained ... Post-Accident Sampling - A
program which will ensure the capability to obtain and
'
. analyze reactor coolant. ... and containment atmosphere
samples under accident conditions. The program shall
include ... provision for maintenance of sampling and
analysis equipment."
Paragraph 8.3.1 of Shoreham " Preventive Maintenance Program
SP No. 12.015.01, Revision 7, dated February 6,1985, stated
" Preventive Maintenance Activities shall be performed in
accordance with approved procedures when required by the
Scheduled Activity Worksheet (SAWS)."
Listed below are the SAWS (loop check and calibration
schedules for 23 instrument loops (each loop consists of
one or more instruments):
Extension
Instrument Loop Due Date Date
1Z96-410CLR-014 09/28/84 11/09/84
1Z96-410E/I-018 05/11/84 09/14/84
1Z96-410E/I-019 05/11/84 09/14/84
1Z96-410E/I-021 05/11/84 09/14/84
1Z96-410 FIT-120A 03/20/84 07/24/84
1Z96-410 FIT-120B 03/20/84 07/24/84
1Z96-410 FIT-131 03/23/84 07/27/84
1Z96-410 PRS-503 06/29/84 08/10/84
1Z96-410PS-058 06/13/84 10/17/84
1Z96-410PS-079A 06/01/84 10/05/84
1Z96-410PS-0798 06/01/84 10/05/84
1Z96-410PS-124 05/22/84 09/25/84
_ __ _. _
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23
Extension
Instrument Loop Due Date Date
1Z96-410PT-015 09/14/84 01/18/85
1Z96-410PT-016 06/13/84 10/17/84
1Z96-410PT-017 05/15/84 09/18/84
1Z96-410PT-125V 06/01/84 10/05/84
1Z96-410 TIC-076B 05/18/84 09/21/84
1Z96-410 TIC-076C 05/22/84 09/25/84
1Z96-410TRS-076 06/29/84 08/10/84
1Z96-410TRS-502 06/29/84 08/10/84
IZ96-410TS-076X 03/09/84 07/13/84
1Z96-410TS-076Y 03/09/84 07/13/84
1Z96-410X7R-025 06/29/84 08/10/84
The calibrations and loop checks of the above instruments
have not yet been conducted as of February 14,1985(50-322/
85-04-16).
The above matter was discussed with licensee representatives.
Licensee representatives indicated their belief that the PASS
system need not be operable until exceeding 5% power. As a
result, the identified calibrations were placed on hold and
would be completed prior to placing the PASS in operation.
This matter is discussed in section 10 of this report.
9.4.3.2 The inspector examined the calibration stickers attached to
the PASS instruments, and noted that the "Next Calibration
Due Date" column of most stickers was not properly filled
in. Paragraph 8.10 of the licensee's calibration procedure
SP No. 41.005.01, Revision 1, does not clearly indicate that
this is a requirement. This information on the sticker is
important because it helps the licensee's QA or QC inspector
to identify calibration overdue instruments without searching
through the SAWS or other paper work.
The inspector also identified numerous instruments with
missing calibration stickers, for example, IZ96-PT016,
-PT125, -FT131, -I/P131, -TIO18.
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These items are unresolved pending NRC verification of
licensee corrective actions (322/85-04-17).
9.5 ' Alternate Power Supplies to PASS
9.5.1 The inspector reviewed pertinent documents concerning the
alternate power supplies to the PASS and its associated equipment
to ascertain whether the alternate power supplies can be secured
within 30 minutes after loss of off-site power.
Items reviewed in this determination include: Shoreham Supple-
montal Diesel Generators - EMO - Electrical Functional Test
Procedure TP No. 85.84042.3, Revision 1, dated May 29, 1984,
including Simplied One-Line Diagram Showing Circuit Breakers,
and functional test result dated July 2, 1984.
9.5.2 Findings
No violations were identified in this area.
10.0 Technical Specifications
The licensee's implementation of Post-Accident Sampling and Analysis
requirements was reviewed against criteria contained in the following:
- Technical Specification 6.8, " Procedures and Programs"
- Facility Operating License No NPF-19
The evaluation of the licensee's performance in the area was based on:
- discussions with cognizant licensee personnel;
- review of installed equipment;
- observations by the inspector; and
review of documentation.
Within the scope of this review, the following matters requiring licensee
attention was identified:
Technical Specification 6.8 requires that the licensee establish, imple-
ment and maintain a program which will ensure the capability to obtain
and analyze reactor coolant, radioactive iodines and particulates in
plant gaseous effluents, and containment atmosphere samples under accident
conditions. The program shall include training of personnel, procedures
for sampling and analysis and provisions for maintenance of sampling and
analysis equipment. ,
. _ _ - -
, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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25
The existing reactor coolant and containment atmosphere sampling, as
well as implementation procedures are adequate for obtaining and analyzing
samples in the event of an accident while the plant is in its current
condition of less than 5% power. However, the PASS system, designed for
accident conditions occurring whenever the reactor exceeds 5% power, is
not fully operable and requires upgrading as described below.
Training of Personnel
The licensee had provided training of applicable personnel in the area
of post-accident sampling and analysis. However, before the inspection
(November, 1984) the licensee revised the post-accident sampling pro-
cedure, but had not yet trained all personnel on the procedure. The
procedure was not yet effective (effective February 22,1985). The
licensee has yet to complete training of all personnel.
This matter remains open (50-322/85-04-20).
Procedures for Sampling and Analysis
The licensee has not established the following procedures:
procedures for analysis of highly radioactive effluent samples (i.e.,
particulates and iodines) (Details section 6.2.2)(50-322/85-04-12)
'
procedures for transportation of highly radioactive samples to the
licensee's offsite vendor for analysis (Details section 4.2.3.1)
(50-322/85-04-04)
These items remain open pending NRC verification that these procedures
have been established.
Provisions for Maintenance of Sampling and Analysis Equipment
The licensee has established a maintenance program to maintain plant
equipment. However as of the time of this inspection the licensee had not
implemented the maintenance program for the post-accident sampling facility
(Details section 9.4).
This item remains open until NRC has verified that the maintenance program
has been developed (50-322/85-04-18).
11.0 Exit Interview
The Post-Accident Sampling and Analysis Team met with licensee represent-
atives at the conclusion of the inspection on February 15, 1985. The Team
Leader summarized the purpose, scope, and findings of the inspection.
At no time during the inspection was written material provided to the
licensee.
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ATTACHMENT 1
TO INSPECTION REPORT
l
50-322/85-04
PERSONS CONTACTED
A. Licensee Personnel
T. Burns, Radiochemistry Health Physics Technician
3
G. Cubeta, Health Physics Foreman
- N. DiMascio, Health Physics Engineer
R. DuPrey, Computer Programmer
J. Etzweiler, Section Head, Equipment Qualification
R. Grunseich, Supervisor, Nuclear Licensing
M. Juliano, Radiochemistry Technician
- N. Morcos, Acting Radiochemistry Engineer
A. Nielson, Health Physics Technician
A. Parker, Nuclear Engineer, Environmental Qualification
- R. Petricek, Radiochemistry Support Supervisor
.D. Puckett, Consultant, Emergency Planning
S. Refaey, Consultant, PASS
- J. Schmitt, Radiological Controls Division Manager
R. Thompson, Health Physics Foreman
J. Tirone, Radiochemistry Technician
B. Whitmer, Health Physics Foreman
- J. L. Smith, Manager, Nuclear Operations Support
- R. A. Kubinak, Director, Quality Assurance
- R. A. Wieman, Engineer, I&C
- A. R. Muller, QC Division Manager
- D. Terry, Manager, Maintenance Division
- E. P. Stergakes, Manager, Radiation Protection Division
- N. Steiger, Plant Manager
- K. K. Taylor, Section Head, Radiological Assessment
- K. McLaughlin, Clerk
- L. F. Britt, Manager, Licensing and Regulatory Affairs 4
B. Contractor Personnel
- R. J. Rossin, Engineer, Stone and Webster
- W. Burnett, Compliance Engineer, Impe11
- T. S. Bulischeck, Acting Rad. Chem. Lab. Supervisor, NUS
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--
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Attachment-.1 2
.C. ~NRC'
E
- P. W. Eselgroth, Senior Resident Inspector
-
Other members of the licensee's staff were also contacted and/or
participated in an exercise of post-accident and effluent monitoring
-systems during the inspection.
'* Denotes attendance at exit interview on' February 15, 1985.
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ATTACHMENT 2
TO INSPECTION REPORT 50-322/85-04
DOCUMENTATION FOR NUREG-0737, II.B.3
Shoreham Nuclear Power Station Emergency Procedures
--
EPIP 2-9, " Post-Accident Primary Coolant Sampling," Revision 2,
November 19, 1984.
--
EPIP 2-10, " Post-Accident Primary Coolant Sample Analysis," Revision 2,
November 19, 1984.
--
EPIP 2-11, " Post-Accident Containment Air Sampling," Revision 2,
November 19, 1984.
--
EPIP 2-12, " Post-Accident Containment Air Sample Analysis," Revision 2,
November 19, 1984.
--
EPIP 2-25, " Determination of PASS Sample Location," Revision 1,
November 15, 1983.
--
EPIP 2-26, " National Lab Shielded Sample Flask Handling," Revision 2,
April 1, 1984.
Other Licensee Procedures
--
SP 72.720.01, " Alternate Method for Analysis of Post-Accident Samples,"
Revision 0, dated September 27, 1983.
--
SP 73.631.23, "PASF Airborne Radiation Monitor Operation," Revision 0,
dated December 9, 1983.
--
SP 73.720.10, " Boron /pH Analyzer 1610 Operation and Calibration,"
Revision 0, dated September 29, 1983.
--
SP 73.720.12, "Orbisphere Dissolved Oxygen Analyzer Operation and
Calibration," Revision 0, dated October 6, 1983.
--
SP 73.720.14, "Dionex Ion Chromatograph - Operation and Calibration,"
Revision 0, dated September 24, 1983.
--
SP 76.033.12, " PASS Gamma Spectrometer System Calibration and Calibration
Check," Revision 0, dated October 19, 1983.
--
SP 76.631.23, " Post-Accident Sampling Facility Airborne Radiation Monitor
Calibration and Functional Test," Revision 2, dated September 20, 1984.
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Attachment 2 2
I --
SP 76.720.01, "Victoreen 3-Channel Gross Gamma Monitoring System
Calibration and Functional Test," Revision 0, dated September 28, 1983.
--
SP 77.631.23, "PASF Airborne Radiation Monitor Functional Test,"
Revision A, undated.
--
SP 78.011.21, " Boron Analysis, Fluoroborate Selective Ion Electrode,"
Revision 0, dated September 21, 1982.
,
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SP 78.030.32, "pH Determination, Flat-Surface Electrode," Revision 0,
dated August 23, 1982.
Licensee Reports
--
" Orion Model 1610 Boron /pH Monitor Operations Manual," Revision F,
August 31, 1983.
--
"NUPAC PAS-1 Post-Accident Sample Cask Operation and Maintenance Manual,"
! Revision 1, June 29,1984.
--
" System Description," Revision 0, dated January 5,1983. '
Licensee Correspondence
--
J. P. Novarro, Proj. Mgr. , SNPS, to H. R. Denton, Dir. , NRR, dated
May 15, 1981.
--
8. R. McCaf frey, Mgr. , Proj . Engr. , SNPS, to H. R. Denton, Dir. , NRR,
dated July 23, 1981.
--
B. R. McCaffrey, Mgr., Prof. Engr., SNPS, to H. R. Denton, Dir., NRR,
dated July 31, 1981.
--
J. L. Smith, Mgr. , Spec. Proj. , SNPS, to H. R. Denton, Dir. , NRR, dated
November 23, 1981.
--
J. L. Smith, Mgr. , Spec. Proj. , SNPS, to H. R. Denton, Dir. , NRR, dated
December 11, 1981.
--
J. L. Smith, Mgr. , Spec. Proj. , SNPS, to H. R. Denton, Dir. , NRR, dated
January 7,1982.
!
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J. L. Smith, Mgr. , Spec. Proj. , SNPS, to H. R. Denton, Dir. , NRR, dated
- January 11, 1982.
,
--
J. L. Smith, Mgr. , Spec. Proj. , SNPS, to H. R. Denton, Dir. , NRR, dated
,
January 13, 1982.
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Attachment 2 3
--
~ J. L. Smith, Mgr. , Spec. Proj. , SNPS, to H. R. Denton, Dir. , NRR, dated
January 12, 1983.
--
J. L. Smith, Mgr. , Spec. Proj. , SNPS, to H. R. Denton, Dir. , NRR, dated
April 8, 1983.
Shoreham Nuclear Power Station Drawings
--
- M-13449-6, " Flow Diagram, Ventilation - Misc. Buildings," Sheet 3,
Revision 6, dated May 31, 1983.
_
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M-13432-7, " Flow Diagram, Post-Accident Sample System IZ96," Sheet 2,
Revision 7, dated August 30, 1983.
---
M-13389-7,. " Flow Diagram, Post-Accident Sample System 1Z96," Sheet 1,
Revision 7, dated August 30, 1983.
--
. M-12405-14, " Flow Diagram, Primary Containment, Atmospheric Control
System 1T48," Revision 14, date not legible.
-- . M-10155-B, " Flow Diagram, Reactor Vessel - Instruments," Sheet 2,
Revision 8, dated September 23, 1982.
--
M-10111-19, " Flow Diagram, Residual HT. Removal Ssy. No. IE11," Sheet 1,
Revision 19, April 13, 1984.
--
M-10112-20, " Flow Diagram, Residual Heat Removal System, No.1E11,"
Sheet 2, Revision 20, date not legible.
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ATTACHMENT 3
.
TO INSPECTION REPORT 50-322/85-04
COMPARISON OF ANALYTICAL RESULTS
A .~ Chemical Analysis
.-
-Boron
The on-line analysis capability was not tested due to equipment
failure. The following data are the results of the analysis of the
Analysis NUREG-0737 Licensee
. Standard- -Results Error Requirements Commitment
99.8 ppm 100 ppm 0.2 ppm i 50 ppm i 3%
,- u489.5 ppm 500 ppm 10.5 ppm i 50 ppm i 5%
998 ppm 1000 ppm 2 ppm i 50 ppm ' t 5% -
-
l The on-line analysis capability was not tested due to on going
equipment modifications.
-
pH
The on-line analysis capability could not be tested due to equipment
failure.
B. Isotopic Analysis
Isotopic Analysis could not be conducted due to low activity level of
coolant.
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f ATTACHMENT 4
TO IN5PECTION REPORT 50-322/85-04
L
DOCUMENTATION FOR NUREG-0737, II.F.1-1&2
'
-Long Island Lighting Company - System Description
Radiation Monitoring System 1020.631 (S&W 1011 and 1D21 Shoreham Nuclear Power
Station - Unit 1
Stone and Webster Engineering Corporation Drawings
'
'
Logic Diagram Reac' tor Building Standby Ventilation System, Shoreham
Nuclear Power Station - Unit 1, Long Island Lighting Company, 11600.02 -
LSK-278A,B,C,E,G.J.
Stone and Webster Calculations
Calculation SNPS-1-URB-23-Q, " Dose Rates in the Vicinity of 1011*PNL-126 Post
LOCA", Preparation date June 6, 1982.
KaManSciencesCorporationDrawings
P and I Diagram Auxiliary Pumping Skid, 400569, LILC0 #11600.02-7.23-84E.
P and I Diagram KMG-HRH, 400454-TAB, LILCO #11600.02-7.23-82G.
Enhancement Shield Top, 913201-001, LILC0 #11600.02-7.23-119A.
Enhancement Shield Bottom, 913200-001, LILC0 #11600.02-7.23-118A.
Long Island Lighting Company Emergency Preparedness Implementing Procedures
EPIP 2-7 Post-Accident Gaseous Effluent Sampling, Revision 2, November 19, 1984.
EPIP 2-8 Post-Accident Gaseous Effluent Sampling Analysis, Revision 2,
February 22, 1985.
Long Island Lighting Company Procedures
Post-Accident Station Vent and RBSVS Radiation Monitor Functional Test,
Revision 0, February 9, 1984.
Post-Accident Vent and RBSVS Radiation Monitor Calibration and Functional
Test, Revision 2, May 28, 1984.
Process Radiation Monitor (Post-Accident) CS.630.006.
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Attachment 4 2
.
NRC Memoranda
>
W. E. Kreger, Asst. Dir. , Rad. Prot. , to R. L. Tedesco, Asst. Dir. , DOL, dated
January-12, 1981.
W. E. Kreger, Asst. Dir. , Rad. Prot. , to R. L. Tedesco, Asst. Dir. , DOL, dated -l
January 22, 1981.
L. Rubenstein, Asst. Dir. for Core Containment Sys. to R. L. Tedesco, Asst.
Dir. , DOL, dated June 19, 1981.
W. E. Kreger, Asst. Dir. , Rad. Prot. , to R. L. Tedesco, Asst. Dir. , DOL, dated
June 19, 1981.
W. E. Kreger, Asst. Dir. , Rad. Prot. , to R. L. Tedesco, Asst. Dir. , D0L, dated
L June 22, 1981.
!
L. S. Rubenstein, Asst. Dir. for Core Containment Sys, to R. L. Tedesco, Asst.
Dir., DOL, dated August 19, 1981.
L '
R. W. Houston, Asst. Dir. , Rad Prot. , to R. L. Tedesco, Asst. Dir. , DOL, dated
February 12, 1982.
( R. W. Houston, Asst. Dir. , Rad. Prot. , to R. L. Tedesco, Asst. Dir. , DOL, dated
April 5, 1982.
Licensee Correspondence -
B. R. McCaf frey, Mgr. , Proj. Eng. , SNPS, to H. R. Denton, Dir. , NRR, dated
July 22, 1981.
B. R. McCaffrey, Mgr., Proj. Eng., SNPS, to H. R. Denton, Dir., NRR, dated
August 7, 1981.
B. R. McCaffrey, Mgr. , Proj. Eng. , SNPS, to H. R. Denton, Dir. , HRR, dated
October 13, 1981.
J. L. Smith, Mgr. , Spec. Proj. , SNPS, to H. R. Denton, dated February 7,1982.
Vendor Manual and Documents
Shoreham Nuclear Power Station EMSP Software, Rev. B.0 - Patch 11, Documenta-
tion of revisions to incorporate the high-range monitors and to upgrade the
dose-assessment capabilities to include up-to-date dispersion models and
finite-cloud techniques, Entech Engineering, Inc., dated February 1983.
Shoreham Nuclear Power Station EMSP Software, Rev. B.1, Documentation of
revisions to accommodate the PM21/PM22 sample flow discharge, the incorporation
of two plume capability and other modifications, Entech fingineering, Inc., dated
June 1983.
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Attachment 4 3
Shoreham Nuclear Power Station, Tables of Atmospheric dispersion and depo-
sition data for use during normal operation and accident conditions, Entech
Engineering, Inc., dated July 1983.
Shoreham Nuclear Power Station EMSP Software, Rev. B.1, Documentation of the
release and dose models for assessing the radiological impact of the liquid and
gaseous routine effluents, Entech Engineering, Inc., dated July 1983.
l Shoreham Nuclear Power Station, ACCDOS, An HP-858 computer software package for
backup off-site dose assessment capability under accident conditions and pro-
tective action requirements, Entech Engineering, Inc., dated November 1983.
" Primary Isotopic Calibration," Kaman Instrument Corp., dated June 16, 1982.
" Transfer Calibration," Kaman Instrument Corp., dated November 8, 1982.
" Summary Report of Radiation Monitoring System Detector Calibration for Long
I
Island Lighting Company Shoreham", K-83-14-6-(R), Kaman Instrument Corp.
" Report of Calibrations Noble Gas Radiation Monitor Model KMG-HRH High Range
Gas Channel (Enhanced Design)," K-82-73-6-(R), Kaman Instrument Corp.
"High Range Gas Effluent Radiation Monitor Model KMG-HR," Kaman Instrument
Corp.
" Microcomputer Model KEM-P," Kaman Instrument Corp.
" Universal Controller," Kaman Instrument Corp.
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ATTACHMENT 5
TO INSPECTION REPORT 50-322/85-04
DOCUMENTATION FOR NUREG-0737, II.F.1-3
-Shoreham Nuclear Power Station Final Safety Analysis Report,Section II.F.1,
" Additional Accident Monitoring Instrumentation."
Shoreham Technical Specifications - Table 3.3.7.S-1, " Accident Monitoring
Instrumentation."
!
'Shoreham Startup Form 8.7, " Calibration of Containment High-Range Monitors !
D21*RE-085A and D21*RE-085B", approved December 19, 1983. j
l
Stone and Webster Interoffice Correspondence No. 11605.02, dated 1
September 26, 1980.
Franklin Research Center Test Report No. 455-01, dated September 2, 1980.
D. G. O'Brien Connectors Test Report No. 159-13.
- Litton Vean Connectors Test Report No. 475-03, dated July 1,1981.
Kaman Report No. 16603-1, dated February 16, 1983.
Kaman Construmentation Specifications for Containment Area Radiation Monitor
Model KMA-1 1000.
Kaman Report K-82-70-U-(R), approved by Stone and Webster, February 1, 1983.
Kaman KDA-HR Factory Calibration, dated February 17, 1982.
Equipment Justification, Mark No.1D21RE085A,B, Revision dated October 1982.
' Letters
Shoreham-
SNRC Letter No. 836, dated April 14, 1983
G. K. Price to M. H. Milligan, dated August 19, 1983
A. E. ' Parker to J. F. Etzweiler, dated January 31, 1985, NSD85-210
_.
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Attachment 5 2
,
N.RC
R. Scarano, NRC Region V, to D. Eisenhut, Director, Division of Licensing, NRR,
dated December 20, 1983
J. Wiggenton, DEQA, IE,.NRC, to J. Joyner, NRC Region I, Part 21 Report, dated
October. 13, 1982 i
Procedures
EPIP'l-11, " Operational Assessment," dated November 18, 1983
EPIP 1-1 thru EPIP 1-3 and Emergency Action Level No.13, "Drywell High
Radiation Monitor"
'SP 73.650.15, " Containment High Range Radiation Monitor Operation," dated '
November. 23, 1983
SP 74.640.15, " Containment High Range Radiation Monitor Functional Test,"
dated November, 1983
SP:74.630.16, " Containment High Range Radiation Monitor Calibration and
Functional Test, dated-April, 1984
i Training
Lesson Plan-EP-14, " Implementing Actions," dated October 30, 1984
EP-14, Examinations 1983, 1984, and 1985
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ATTACHMENT 6
TO INSPECTION REPORT 50-322/85-04 l
DOCUMENTATION FOR-ITEM III.D.3.3
Procedures'
SP No. 63.030.03, Revision 0, " Operation of the Eberline AMS-3 (CAMS)"
- SP No. 66.030.03, Revision 1, " Calibration of the Eberline AMS-3 (CAMS)"
- 'SP No. 63.030.01, Revision 0, " Operation of Continuous Air Monitors
.(CAMS)
- SP No. 66.030.01, Revision 1 and Revision C, " Calibration of the Eberline
PING - (CAMS)"
- SP No. 66.032.01, Revision 6, " Calibration of Portable Air Samplers"
- SP.No. 61.080.01, Revision'3, " Control of Health Physics Instrumentation"
- -SP No. 63.032.02, Revision 0, " Operation and Use of the Gilian.Model HFS
.113T Lapel Air Sampler"
'
- SP No.~66.032.02, Revision 0, " Calibration of the Gilian Model HFS 113T
Lapel Air Samplers"
- SP No. 63.032.01, Revision 2, " Operation of Portable Air Samplers"
- EPIP No. 2-4, Revision 1, "Inplant Surveys" .l
-SP No. 62.030.01, Revision 2, " Airborne Survey Techniques and
.
,
Determinations"
Letters
- - -SNRC-563, dated'May 15, 1981'(Novarro to Denton)
- SNRC-602, dated July 22, 1981 (McCaffrey to Denton)
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ATTACHMENT 7
TO INSPECTION REPORT 50-322/85-04
QUALITY ASSURANCE AND DESIGN REVIEW
A. Stone and Webster Documents
1) Flow Diagrams M-13389-7 and M-13432-7, " Post-Accident Sampling System
-1Z96," Sheets 1 and 2, dated August 30, 1983.
2) Flow Diagram M-12405-14, " Primary Containment Atmospheric Control
System IT48."
3) Flow Diagram M-10155-8, " Reactor Vessel Instruments," Sheet 2, dated
September 23, 1982, portion of the jet pump flow instrument lines
connected to PASS sample input.
4) -Flow Diagrams M-10111-19 and M-10112-20, " Residual Heat Removal
'
System 1E11, Sheets I and 2, portion of the 'RHR pump discharge
-
connected to the PASS semple input.
5) Flow Diagram M-10157-15, " Demineralized and Makeup Water System
IP11," Sheet 1, portion of the Condensate Transfer Pump Discharge.
supplying cooling water to PASS sample cooler.
6) Instrument Piping Diagrams M-13464-3, M-13465-3, M-13466-5,
M-13467-5, M-13468-4, " Post-Accident Sample System, Reactor
Building," Sheet I through Sheet 5.
-7) Drawing Nos. M-13451-4 and M-13542-6, " Post-Accident Sampling System
Skid 1Z96-PNL-002," Sheets 1 and 2.
-
8) Installation Specification of Instrument Tubing, Specification No.
SH1-343', Revision 4, dated January 24, 1983, portion that applies to
sample tubing installation.
B. Long Island Lighting Company Documents
1) Purchase Order No. 310677-29 to Reliance Electric Company for
fabrication of the Post-Accident Sampling Panel, dated March 23,
1981.
2) Purchase Order No. 310979 to Valcor Engineering Corporation for
PASS Containment Isolation Valves, dated January 6, 1981, including
. Attachment No. 1, Valve Requirements,-dated October 2, 1980.
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