ML17254A569
| ML17254A569 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/25/1985 |
| From: | Baum J, Doagoun T, Dragoun T, Knox W, Paulino R, Shanbaky M, Weadock A, Jason White NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17254A568 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 50-244-85-08, 50-244-85-8, NUDOCS 8510080032 | |
| Download: ML17254A569 (44) | |
See also: IR 05000244/1985008
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report
No.
50-244/85-08
Docket No.
50-244
License
No.
Pri ority
Category
C
Licensee:
Rochester
Gas
and Electric Cor oration
Facility Name:
Ginna Nuclear Power Plant
Inspection At:
Ontario
Inspection
Conducte
.
85
Inspectors:
J.
. Wh',
a 'ation Specialist,
NRC
R.
P
in
Rea
or
n
r,
NRC
dat
p ~s
P'~
T.
D.
ag
n,
Ra
ecialist,
NRC
da
2.> /W
.
Baum
Health Physic'st
Brookhav
Nat
al.
o
tory (BNL)
W.
H.
nox, Healt
Ph sicist,
date
Z~- QJ-
date
A. A. Weadock,
Ra iation Specialist,
NRC
Approved by:
M.
M. Shanba
, Chief,
R Radiation Safety
Section
date
ZD
date
Ins ection Summar:
Ins ection
on June
10-14
1985
Re ort No. 50-244/85-08
Areas
Ins ected:
Special,
announced
safety inspection of the licensee's
implemen-
tation and status of the following task actions iden'tified in NUREG-0737: Post-
accident
sampling of reactor coolant
and containment
atmosphere;
increased
range
of noble gas radiation monitors; post-accident
effluent monitoring; containment
high range radiation monitoring;
and in-plant radioiodine measurements.
The in-
spection
involved 170 hours0.00197 days <br />0.0472 hours <br />2.810847e-4 weeks <br />6.4685e-5 months <br /> by four region-based
inspectors
and two contractors
from Brookhaven National Laboratory.
Results:
No violations were identified in the areas
inspected.
areas
requiring improvements
were identified.
OOSOo~'SOOO~~~g
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However,
several
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DETAILS
1.0
Persons
Contacted
1. 1
During the course of the inspection,
the following licensee
personnel
were contacted
or interviewed:
- R. W. Kober, Vice President
~B. A. Snow, Superintendent-Nuclear
Production
~R.
C. McCredy, Manager,
Nuclear Engineering
G. Daniels,
Manager, Electrical Engineering
"D. Fi lkins, Manager,
Health Physics
and Chemistry
T. A. Meyer, Technical
Manager
- R. Baker, Electrical
Engineering
C. Boucher,
Chemistry Technician
G. Caine,
Instrumentation
and Controls
- D. Filion, Radiochemist
- W. Goodman,
HP Foreman
~B.
R. guinn, Corporate
Health Physicist
- C. Mambretti,
Systems
Engineer
"F. J. Mis, Health Physcist
- J. T. St. Martin, Station
Engineer
- S. B. Warren, Health Physicist
- Denotes attendance
at the Exit Interview conducted
June
14,
1985.
Other members of the licensee's
staff were also contacted
and/or
participated
in exercises
of post accident
and effluent monitoring
systems
during the inspection.
2.0
~Pur use
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
Re uirements:
Task No.
Title
II.B.3.
II.F.I-1
II.F.1-2
II.F.1-3
III.D.3.3
Post Accident Sampling
Capabi
1 ity
Noble Gas Effluent Monitors
Sampling
and Analysis of Plant Effluents
Containment
High-Range Radiation Monitor
Improved Inplant Iodine Instrumentation
under
Accident Conditions
3
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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
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t,
following documents:
TMI-2 Lessons
Learned
Task Force Status
Re ort and Short-
Term Recommendations
dated July 1979.
Letter from Darrel
G. Eisenhut,
Acting Director, Division of
Operating
Reactors,
to all Operating
Power Plants,
dated October 30,
1979.
NUREG-0737, Clarification of TMI Action Plan
Re uirements
dated
November,
1980.
Generic Letter 82-05, letter
from D.
G. Eisenhut,
Director, Division
of Licensing, to all Licensees
of Operating
Power Reactors,
dated
March 14,
1982.
Letter from D.
G. Eisenhut,
Director, Division of Licensing, 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.
"Order Confirming Licensee
Commitments
on Post-TMI Related Issues",
dated
March 14,
1983.
Regulatory
Guide 1.3 "Assumptions
Used for Evaluating Radiological
Consequences
of a Loss of Coolant Accident for Boiling Water
Reactors".
Regulatory
Guide 1.4,
"Assumptions
Used for Evaluating Radiological
Consequences
of a Loss of Coolant Accident for Pressurized
Water
Reactors".
Regulatory
Guide 1.97,
Rev.
2, "Instrumentation for Light-Mate<-
Cooled Nuclear
Power Plants to Assess
Plant
and Environs Conditions
During and Following an Accident".
Regulatory
Guide 8.8,
Rev.
3, "Information Relevant to Ensuring that
Occupational
Radiation
Exposure at Nuclear
Power Station will be As
Low As Reasonably
Achievable".
R.
E. Ginna Nuclear
Power Plant Updated Safety Analysis Report final
draft dated
November
1984.
following documents:
TMI-2 Lessons
Learned
Task Force Status
Re ort and Short-
Term Recommendations
dated July 1979.
Letter from Darrel
G. Eisenhut,
Acting Director, Division of
Operating
Reactors,
to all Operating
Power Plants,
dated October 30,
1979.
NUREG-0737, Clarification of TMI Action Plan
Re uirements
dated
November,
1980.
Generic Letter 82-05, letter from D.
G. Eisenhut,
Director, Division
of Licensing, to all Licensees
of Operating
Power Reactors,
dated
March 14,
1982.
Letter from D.
G. Eisenhut,
Director, Division of Licensing, 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.
"Order Confirming Licensee
Commitments
on Post-TMI Related Issues",
dated
March 14, 1983..
Regulatory
Guide 1.3 "Assumptions
Used for Evaluating Radiological
Consequences
of a Loss of Coolant Accident for Boiling Water
Reactors".
Regulatory
Guide 1.4,
"Assumptions
Used for Evaluating Radiological
Consequences
of a Loss of Coolant Accident for Pressurized
Water
Reactors".
Regulatory
Guide 1.97,
Rev.
2, "Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess
Plant
and Environs Conditions
During and Following an Accident".
Regulatory
Guide 8.8,
Rev.
3, "Information Relevant to Ensuring that
Occupational
Radiation
Exposure at Nuclear
Power Station will be As
Low As Reasonably
Achievable".
R.
E. Ginna Nuclear
Power Plant Updated Safety Analysis Report final
draft dated
November
1984.
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4.0
Post Accident
Sam lin
S stem
Item II.B.3.
Position
Item II.B.3, specifies that licensee
shall
have the
capability to promptly collect,
handle
and analyze
post accident
samples
which are representative
of conditions existing in the
and containment
atmosphere.
Specific criteria are
denoted
in commitmgnts to the
NRC relative to the specifications
contained
in NUREG-0737.
Documents
Reviewed
The'mplementation,
adequacy
and status of the licensee's
post-
accident
sampling
and monitoring systems
were reviewed against
the
criteria identified in Section
3 '
of this report and in regard to
licensee letters,
memoranda,
drawings
and station
procedures
as
listed in Attachment I.A.
4.2
S stem Oescri tion
The Ginna Post Accident Sampling
System
(PASS)
was designed
and built
by Sentry.
The system consists of three panels
and associated
lines.
A shielded
Liquid and
Gas
Sample
Panel
( LGSP), located at elevation
253'n the south section of the Intermediate
Building provides
containment
and contai'nment
atmosphere
sampling
capability.
Control
and monitoring is accomplished
by a Control
Panel
(CP)
and Instrument
Panel
(IP) which is located in the Hot
Shop at elevation 253'.
The system is designed
to provide on-line analysis of boron,
pH,
dissolved
and dissolved
Diluted grab
samples
can
be obtained
and transported
to the labora-
tory for isotopic and back-up analyses
of non-radiological
para-
meters.
4.3
PASS Performance
Test
A test of the
PASS was conducted
on June
12,
1985 in order to esta-
blishh
the licensees
capability for sample
acqui sition and analysis
in accordance
with the specifications of NUREG-0737
and the licensee's
commitment.
Although samples
were obtained
and analyzed,
based
on existing pro-
cedures
the licensee
was unable to demonstrate
that all activities
could be conducted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
This delay was largely due to
extensive
preparation,
system flush times,
and compartmentalized
analysis
processes.
I
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k
The system
had not been fully tested to demonstrate
that represen-
tative samples
could be obtained.
The following findings were
made
in the course of this inspection relative to testing:
1.
Although samples
had
been collected
from the "B" loop,
sample
acquisition
and representativeness
have not been validated in
accordance
with the "B" Loop Verification Test Procedure.
2.
There
have
been
no tests to determine that
a representative
sample
can
be obtained at low pressure.
Since the system
does
not include
a pump, but relies
on
RCS pressure
as
a
motive force, the ability to obtain
a low-pressure
sample is
questionable.
3.
No test data
were available to demonstrate
the performance
of
the analytical
instrumentation
or techniques
in the presence
of elements
in the Standard
Test Matrix.
4.3. 1 Recommendation
for Im rovements
based
on the above findin s
1
~
Demonstrate
that
a representative
sample
can
be collected
at low RCS pressure
(85-08-01).
2.
Complete the remaining
system tests
and provide data to
demonstrate
that the stated
accuracies,
ranges
and sensi-
tivities of the analytical
instrumentation
and techniques
can
be achieved with the Standard
Test Matrix solution
(85-08-02).
These matters will be reviewed in a future inspection.
4.4
Findin
s
Personnel
and Procedures
The licensee
was able to demonstrate
a satisfactory analytical
capa-
bility.
However,
an insufficient number of personnel
have
been
trained to operate
the system.
The training program had'not
been
formalized.
The following technical deficiencies
were noted in the
procedures
used to detail
PASS operation:
1.
The system operating
procedure,
PC-25.7
"Operation of the
Under Accident Conditions", requires extensive
preparation
and
flushing times.
As
a result,
the collection and analysis
process
could not be completed within the
3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time limit specified in
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2.
An inappropriate
pressure
correction factor was used in PC-25.4,
equation 6.6.5.1.
3.
Several
parameters
were omitted from equation
6. 12 '
in the core
damage
assessment
procedure.
4.
There were
no provisions in the procedures
for providing back-up
capability for on-line
pH analysis.
However, it was noted that
the on-line byron meter is also
was capable of determining
pH.
5.
The isotopic analysis
procedure
does not indicate the specific
method to be used for processing
post-accident
samples.
For
example,
the procedure
indicates that samples
would be boiled to
remove noble gases.
However,
from discussions
with personnel,
it is unlikely that post-accident
samples
would not be boiled
due to radiological considerations.
6.
Several libraries are required
by the multi-channel
analyzer
(MCA)
to completely determine
the concentration
of all isotopes
used
in the core
damage
assessment
procedure.
This requires multiple
analysis of the
same
sample.
7.
In the process of obtaining grab samples,
personnel
were re-
quired to stand close to an unshielded
holding tank which could
contain
up to
18 gallons of coolant.
The holding tank can
be
manually flushed to minimize personnel
exposures.
A procedure
change
request
was issued during the inspection period to
require this flush prior to obtaining grap samples.
4.4. 1 Recommendations
for im rovements
based
on the
above findin s:
1.
Streamline
or consolidate
processes
in the system operating
procedure
to assure
that samples
can
be collected
and
analyzed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
2.
Make appropriate
corrections to equations
in the core
damage
assessment
procedure.
3.
Consider the use of the boron
pH probe
as
a back-up to the
primary in-line pH analysis
instrumentation.
4.
Indicate the actual
method
used to isotopically analyze
sample in procedures
and develop
a consolidated library
for post accident analysis.
These matters will be reviewed in a future inspection.
(85-08-03)
4.5
Findin
s
Hardware
Based
on an evaluation of the
PASS System equipment,
the following
items were noted:
l.
A limit switch .was
used to indicate
sample air flow from the
containment.
However, there were
no data available to indicate
that
a representative
sample could be achieved at the limit
switch activation threshold.
2.
The spare parts
needed
to maintain the system
have not been
included in the inventory system.
3.
During the performance test it was noted that:
a.
Valve V33 in position
A was inoperative.
An alternate
flow path
was available.
b.
The position indicator light for Valve 955 (Hot leg isola-
tion valve) indicated that the valve was in an
open
position,
regardless
of actual
valve position.
c.
Flow indicators
were labeled
FS1
and
FS4
on the control
panel.
The procedure
designations
were Fl and
F4,
respectively.
4.5.1
Recommendations
for im rovement
Based
on the above findings:
1.
Provide assurances
that representative
containment
air samples
can
be obtained at the flow indicator's
activation threshold.
2.
Repair the system valves
and correct the flow
indicator designations.
3.
Include the system
spare
parts in the inventory
system.
These matter s will be reviewed in a future inspection.
(85-08-04)
4.6
ualit
Assurance
and Desi
n Review
As part of this inspection effort a review was performed to verify
and validate the adequacy of the level of design
and quality assurance
program for the installation.
Documents
Reviewed
The procurement,
installation, construction
and inspection of the
licensee's
Post-Accident
Sampling
System were reviewed against
the
criteria identified in the following documents:
RG&E Procedure
No. QCIP-8; Final Inspection of Station Modifi-
cations
RG&E Procedure
No. QCIP-12;
Pipe
and Tubing Installation Proce-
dure
As-Built Drawing Nos.
1998-E-090
Rev.
1, 1998-E-290,
Rev.
4,
1998-M-001
Rev.
0,
1998-E-100
Rev.
2 and
1998-M-212 Rev.
2
PASS General
Arrangement drawing No. 33013-1143
Rev.
1
Post-Accident
Sampling
System Piping and Instrumentation
Diagram
No. 33013-1142
Rev.
1
Post-Accident
Sampling
System
P&ID Drawing No. 33013-1279
Rev.
1
Purchase
Order No.
17896 dated August ll, 1981
Engineering
Work Requests
(EWR)
2606C
Station Modification (SM) No.
2606
SM No. 2606. 1 - Mechanical Installation of the
PASS System
SM No. 2606.2
PASS System Electrical Installation
SM No. 2606. 1A - S/G1A and
1B Blowdown Sampling Line Installation
SM No. 2606.3 -
System
Equipment Mounting and Masonry Work
SM No. 2606.4A
Hydrostatic Test of PASS Inside Containment
SM 2606 'C
Hydrostatic/Pneumatic
Test of Post-Accident
Sampling
System
Balance of Piping
SM 2606.5A
Panel
Energization
Test
SM 2606.5F - Heat Trace
System Test,
SM 2606.5J
PASS System Instrumentation
Calibration
and
Verification Test
Procedure
Nos.
A-1102 and A-1002 Qualification of Test Personnel
Training and Attendance
Records
dated
November
1,
1982 and
January
4,
1983
PASS Implementation
and Safety Analysis
Rev.
1 dated
May 25,
1982
PASS Implementation
Design Criteria Rev.
1 dated July 19,
1982
~Findin
s
In addition to reviewing the above documents,
the inspector verified
the
PASS as-built configuration
as well as design
changes
determining
that the
PASS is classified non-seismic
except for the component
cooling water and volume control tank purge line tie-ins which are
seismic category l.
All equipment
supports
have
been
analyzed for
seismic
loadings to preclude
gross failure which could damage
nearby
safety-related
components.
Generally, all gA and design control requirements
and procedures
were
adequately
performed.
Design changes,
nonconformances
and rework were
wel'1 documented.
Records
are easily retrievable,
current
and signed
by authorized cognizant personnel.
The design
and quality assurance
program for the
system instal-
lation was adequate.
5.0
Noble Gas Effluent Monitor
Item 11.F. 1-1.
and
Sam lin
and Anal sis of
Plant Effluents
Item II'. 1-2
5.1
Position
Item II.F.1-1 requires
the installation of noble
gas
monitors with an extended
range designed
to function during normal
operation
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.l-l of NUREG-0737.
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 I.B.
The licensee's
performance relative to these criteria was determined
by interviews with the principal persons
associated
with the design,
testing, installation
and surveillance of the high range
noble gas
monitoring systems,
a review of the associated
procedures
and docu-
mentation,
an examination of personnel
qualifications,
and direct
observations
of the systems.
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5.2
~Findin
s
Within the
scope of this review the following were identified:
5.2. 1 Descri tion and
Ca abi lit
To meet
NUREG-0737 Requirements
for the extended
range monitor-
ing of noble gases
in plant effluents,
the Licensee
has instal-
led three Eberline
SPING-4 radiation monitors
and two Eberline
GM tube detector
steam line monitors.
Two of its SPING-4 units
are identical
and are
employed for containment
vent and plant
vent effluent monitoring.
Each
has
a capability for monitoring
noble gases
in concentrations
from 10
to
10
uCi/cm
of Xe
or its equivalent
using
a low-range beta scintillation detector
(RDA-3); a mid-range
channel
which, employs
an energy
com-
pensated
GM tube (SA-12) with a sensitivity from 2.5
x 10
to
1 x 10
uCi/cm
and
a high-range
energy
compensated
GM tube
3
3.
(SA-9) with a range
from 10
to
10
uCi/cm
.
The SPING-4 is basically identical to the Eberline SPING-3, with
the addition of a high range
noble
gas detector
(SA-9) so as to
extend its upper
range
from 10
to
10
uCi/cm
(Xe
) equiva-
2
5
.
3
"lent). It was developed
by Eberline in 1980 to meet the require-
ments of NUREG-0578.
However,
from contacts with the vendor it
is apparent that Eberline
does not claim that it meets
the
requirements
of NUREG-0737, II.F. 1-1 due to the vulnerability of
its microcomputer.
Also, since there is no feature in the
SPING-4 whereby its low and intermediate
range detectors
would
be by-passed
or deactivated
when subject to high concentrations
of radiogases,
their recovery following such exposure
is
questionable.
These
two SPING-4 units each include
a particulate filter and
a
standard
2" x 3/4" cartridge containing silver zeolite,
which are
viewed by low range detectors.
These filters are the sole
components of the
SPING-4 for the sampling of particulates
and
radioiodines
as required
by NUREG-0737, II.F. 1-2.
The system
also includes
a fourth shielded
GM-tube detector
which monitors
gamma background
in the area of the
SPING unit.
The output from
this channel
is used for gamma
compensation
on the low-range
and
mid-range noble-gas detectors.
Single-nozzle isokinetic sampling probes
are installed in the
containment
vent and plant vent.
For the containment
vent
sample line,
an isokinetic nozzle of 5/8" O.D. stainless
steel
tube with an
ID of .481" is used.
The plant vent isokinetic
nozzle
employs
a 3/8" O.D. stainless
steel
tube with .277" ID.
ANSI standard
N13. 1 was
used for design criteria.
~ ~
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11
The plant vent and containment
vent are
sampled at
a flow rate
of 60 liters per minute.
The sampling lines are about
25 feet
and
50 feet long, respectively.
They are not heat traced.
It
was noted that the lower and upper alarm settings
on the flow-
rate meters for the plant vent were set at 30 and
71 liters per
minute.
Also, the pressure
in the
sample
chambers
was approxi-
mately
5 inches of mercury below atmospheric.
Effluent monitor calibrations
are
done
by I & C during annual
outages.
Examination of their procedures
revealed
no specifi-
cation of the proper alarm setpoints for the sampling flow rate
meters.
Since the containment
vent fans were not utilized
during the inspection,
the containment
vent monitoring system
was not operating,
which is according to procedures.
However,
the flow meter incorrectly indicated
a flow rate of 40 liters
per minute.
This problem was corrected
before completion of the
inspection.
A third SPING-4 unit is used to monitor the gaseous
effluent
from the air ejector vent system.
This system
has low-,
medium-,
and high-range
noble-gas
channels
identical to those
described
above for the containment vent and plant vent systems.
Samples
are
drawn isokinetically,
and passed
through
a Johnson
Controls cooler to remove moisture before entering the detector
chambers.
Each
SPING-4 unit includes valves
and plumbing which
make it possible
to purge the unit and to obtain
an unshielded
Counting facilities with extended
sample
shelves
are available,
with the stated capability of handling
a sample containing
several
curies.
Procedures
include conversions
of noble gas
and iodine activity concentrations
to dose rates,
for times
up to
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after shutdown.
Time and motion, shielding design,
and
counting rate studies
were not available
to indicate (a) if grab
samples
and filter samples
can
be obtained without exceeding
5
rem whole body or 75 rem extremity doses;
and (b) to ensure that
design basis
samples
could be analyzed.
A design
review has not been
performed to evaluate
the possible
impact
on detectors
and filters of moisture
laden effluents at
high temperatures,
such
as those which may occur in the contain-
ment effluent vent pathway under accident conditions.
The
operating
temperature limit for the SPING-4 units is specified
as
32
F to 122
F.
Since
gases
from the containment building may
be moisture saturated
and in excess
of 122 F,
an analysis is
needed
to ensure that the sampling lines, filters, scintillation
detectors,
and
GM tubes
would continue to function properly
under these conditions.
I
I
]I
I4
lt
't
I
A
'0
~
~
12
The detector
response
to the variable mix of photon energies
for various periods of time up to 30 days after
a loss of coolant
accident
has
been calculated
by the Technology for Energy Corpora-
tion (TEC).
Conversion
curves
were developed
to convert detector
133
output in mR/hr to uCi/cm
of Xe
or equivalent.
However,
no
empirical confirmation of the calculated initial response
has
been demonstrated.
Also, the instrument calibration is rou-
tinely checked for reproducibility at
a single dose rate which
is about
a factor of 300 to 600 below the upper design counting
rate.
Details
on geometry,
attenuation,
assumptions,
and
methods
employed in sensitivity calculations
were lacking in
the
TEC report.
Noble gas release
rates
can
be quantified by taking the product
of steam flow rate, radioactivity concentration,
and the time
duration that the valves
open.
Read-out
can
be
made locally, in
the main control
room,
and at the technical
operations facility.
Computer controlled readout with alarmed setpoints for all of
these effluent monitors is provided through Eberline
CT-1 con-
trol terminals in the control
room and the
TOF.
For the con-
tainment vent and plant vent effluent monitors, trend alarm,
alert alarm,
and high alarm settings
are available for the
several
detectors.
In order to meet the
NUREG-0737 requirements
for monitoring
activity release
rates
from the main
steam safety valves
and
power operated relief valves,
shielded
Eberline
SA-11 detector
assemblies
are located adjacent
to steam lines
A and
B on the
296 foot level
on the clean
side of the intermediate building.
These monitors
have energy
compensated
GM tubes which are
shielded
by approximately 3" of lead, with a ~" diameter cylin-
drical
open face to the
steam line.
Detector
response
is
designed
to cover the range
from
1 to
10~ uCi/cm
Xe
or its
equivalent.
The detector
on
steam line ¹1 alarmed (for no apparent,
reason)
while the inspectors
were in the area for a few minutes.
This monitor is in an uncomfortably warm area
and subject to
considerable
vibration.
The Eberline
CT-1 control terminal
has
6 lights on its display
panel
which indicate normal, maintenance,
fail, trend alarm,
alert alarm, or high alarm.
The alarms
can
be triggered
by any
one of the above mentioned detector
systems.
During normal
operation,
hourly averages
of monitor readouts
are printed out.
There is
a capability for an instantaneous
readout
on demand.
Once
an alarm has
been tripped,
the monitor actuates
10 minute
average
readouts for all detectors
which have not exceeded their
upper limits.
Checks of alarm setpoints
indicated they were
generally set equal
to those indicated in the Ginna/UFSAR or in
some
cases
at
40% of the
FSAR value.
Cl
13
5.3
Acce tabilit
of Noble Gas
and Steam
Line Monitorin
S stems
The systems
as reviewed,
do not fully appear to meet the requirements
of NUREG-0737, Attachment II.F. 1-1, in that their ability to function
during and following an accident which produces
design basis
release
rates
over an extended
time is questionable.
However,
the Steam
Line
Monitor appears
acceptable
under these
same criteria.
5.4
Recommendations
for Im rovement
Noble
Gas
and Steam
Line Monitors
~
Calibration corrections
are
needed for the SPING-4 monitoring
systems
to correct for the reduced
pressure
in the noble
gas
detector
volumes.
Due to pressure
drops in the sampling lines
the pressure
in the noble
gas detector
chambers
was
seen
to be
about
5 inches of mercury below atmospheric.
Since the detec-
tors were originally calibrated at atmospheric
pressure,
this
entails
an approximately
17% correction.
Alternatively a retro-
fit kit available
from Eberline,
which provides for a mass flow
measurement,
could be installed.
(85-08-05)
~
Empirical checks
should
be
made to verify the calculated
sensi-
tivities of the shielded detectors
used for monitoring steam
lines
A and B.
(85-08-06)
~
An analysis
should
be
made of the potential radiation
damage
to
the microcomputer in the SPING-4's
and of the ability of its low-
and mid-range detectors
to withstand high concentrations
of
noble gases.
(85-08-07)
~
Upper and lower set points for triggering the
sample flow rate
alarms
on the effluent monitors should
be set at points reason-
ably close to normal
sampling rates.
I&C calibration procedures
should include the restoration of flow limits to appropriate
values after testing.
(85-08-08).
Routine calibrations of the
steam line, containment vent, plant
vent,
and air ejector vent monitors should include tests of GM
plateaus
and detector
responses
at maximum intended counting
rates.
These tests
are
needed
to ensure that the detectors
are
not nearing
the
end of their useful periods.
(85-08-09)
These matters will be reviewed in a future inspection.
6.0
Sam lin
and Anal sis of Plant Effluents
Item II.F. 1-2
6. 1
Position
Item II'. 1-2, requires
the provision of a capability for
the collection, transport,
and measurement
of representative
samples
of radioactive
and particulates
that
may accompany
gaseous
~
~
lt W
0
14
effluents following an accident.
It must
be performable within
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, Attachment 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,
memo-
randa,
drawings
and station
procedures
as listed in Appendix I.B.
The licensee's
performance relative to these criteria was determined
by interviews with the principal persons
associated
with the design,
testing, installation,
and surveillance of the
systems for sampling
and analysis of high activity radioiodine
and particulate effluents,
by
a review of the associated
procedures
and documentation,
by an
examination of personnel
qualifications,
and by direct observation of
the systems.
6.2
~Findin
s
This system
does not meet
NRC guidance
given
in,NUREG-0737 II.F. 1-2
for the following reasons:
A shield design
study,
time study,
and related
procedures
are
needed
to ensure that particulate
and iodine filters can
be
changed,
transported,
and analyzed
and that grab
samples
can
be
obtained,
transported
and analyzed
in terms of the design cri-
teria of 30 minutes of sampling at
100 uCi/cm~.
This would
yield about
180 curies of activity on the filter at the sampling
rates
being employed.
(85-08-10)
~
Empirical data are
needed to demonstrate
that representative
particulate
and iodine samples
can
be obtained
under
accident
conditions
from the plant vent and containment
vent sampling
lines.
Although the
sample lines
have isokinetic nozzles, their
long thin dimensions
and lack of heat tracing
may lead to losses
due to plateout
and vapor condensation.
(85-08-11)
~
Data are
needed
to demonstrate
that the particulate
and iodine
filters for the containment
vent monitors will not degrade
or
saturate
with moisture
under accident
type conditions.
Similar
data
are
needed to assure
that these detectors will not be sub-
jected to excessively
high temperatures
due to sampling of hot
gases
which may be produced during
some accidents.
(85-08-12)
These matters
are unresolved
and will be reviewed in a future
inspection.
~
~
15
7.0
Containment
Hi
h Ran
e Radiation Monitor
Item II.F. 1-3
Position
NUREG-0737 item II.F. 1-3 requires
the installation of two high range radia-
tion monitors capable of detecting
and measuring radiation levels within
the reactor
containment during and following an accident.
Specific
requirements
are set forth in Table II.F. 1, Attachment 3.
Observations
The licensee
has installed
two Yictoreen
Model
875 High Range
Containment
Area Monitoring Systems with readouts
mounted in the control
room.
Test
data provided by the manufacturer certifies that the range
and response
of
the monitoring systems
meet Table II.F. 1-3 technical
specifications.
In-
situ tests of detector
response
to radiation were conducted
in May 1982
and
May 1984 using
an Iridium-192 radiography
source.
The licensee
recently purchased
and
has on-site
a Victoreen High Range Field Calibrator
Model
No. 878-10-5 containing
250
mCi of Cesium-137 to be used for res-
ponse
checks
and calibration of the monitoring system in the future.
The positioning of the detectors
inside containment
could not be verified
by direct observation.
A review of installation drawings
was inadequate
to establish
the field of view of the detectors.
The licensee
was
requested
to verify that the detectors
view a large fraction of the con-
tainment volume.
This item will be reviewed in a future inspection.
(85-08-13)
The system
component parts
were verified to be environmentally qualified
to withstand design basis accident conditions with the exception of the
containment detector/connector/cable
assembly.
Victoreen Environmental
Qualification (EQ) test report
No.
950 '01 cites eight test failures in
attempting to qualify this assembly.
The ninth and only successful
test
consists of complete isolation of the cable
and termination
from the
environment
using metal conduit.
The licensee's
installation
does not conform to the qualified Yictoreen
configuration.
The licensee
is using
Raychem shrink sleeve to achieve
a
sealed
cable termination
based
on engineering
assumptions
for qualified
shrink sleeving
and test results for multi-layered
Raychem shrink sleeve
configuration over the cables.
The licensee
determined that stress
cracking of shrink sleeves
during testing
by Victoreen
was due to mis-
applications
and that properly applied shrink sleeves
would not experience
stress
cracks.
However,
no environmental
testing of properly applied
shrink sleeves
on coaxial cable were performed.
In reviewing the licensee's
evaluation of the Victoreen test report the
inspector
noted that the licensee
does not address all problem areas
~ z
A
II,
4k
16
identified.
Critical items omitted in the licensee's
review include
hardening of the cable,
and red sealant material;
and powdering
of the cable electrical insulation.
This item is unresolved
pending the licensee's
re-evaluation
of the
Victoreen
Eg test results
and supporting data establishing
the
environmental qualification of the installed cable
assembly.
(85-08-14)
8.0
Im roved In lant Iodine Instrumentation
Under Accident Conditions
Position
NUREG-0737 Item III.D.3.3 requires
the licensee
to provide equipment,
associated
training,
and procedures
for accurately determining the air-
borne iodine concentration
in areas within the facility where plant
personnel
may be present
during
an accident.
Observations
and Findin
s
The licensee will use silver zeolite sampling cartridges
to selectively
monitor airborne iodine concentrations.
Cartridges will be counted in the
field using survey meters with geiger probes.
Appropriate conversion
formulae to obtain iodine concentrations
are found in procedure
SC-421.
Sample cartridges
may also
be analyzed
in the radiochemistry laboratories
multichannel
analyzer or in the environmental
laboratory.
The licensee
stated that
an analysis of post accident conditions indicated that background
radiation levels in the laboratories will be low enough to allow use of
the multichannel
analyzers.
A supply of silver zeolite cartridges
and various types of air sampling
pumps are
stocked
in emergency
lockers located in the Operations
Support
Center/Technical
Support Center
and the
Emergency Operations Facility.
These
are the only site areas
(including the Control
Room) to be occupied
after
an accident.
These
supplies
are inventoried monthly per procedure
SC-410.
The availability of procedures
and the associated
training of personnel
appeared
to be adequate.
Recommendation
for
Im rovement
The auxiliary multichannel
analyzer (Tracor Northern TN-ll) located
in the environmental
counting laboratory
was not fully functional.
The operating
procedure
PC-1.4 was incomplete
and difficult to
follow.
Counting efficiency factors for the silver zeolite cartridge
configuration
was not available for the full range of anticipated
iodine concentrations.
Only a small
number of technicians
were
trained in the
use of the equipment.
There
was
no requirement to
purge
a sample cartridge with clean air to remove noble
gas prior
to analysis.
These matters will be reviewed in a future inspection.
(85-08-15)
Attachment I.A
Documentation for NUREG-0737
II.B.3
Corres
ondence
D.M. Crutchfield, Chief
ORB ¹5,
DOL, to J.E.
Maier, V.P.
RGE,
dated
March 2,
1982.
J.E.
Maier, Vice President,
RGE, to D.M. Crutchfield, Chief
ORB ¹5,
DOL, dated April 23,
1982.
J.E. Maier, Vice President,
RGE, to D.M. Crutchfield, Chief
ORB ¹5,
DOL dated
June
17,
1982.
D.M. Crutchfield, Chief
ORB ¹5,
DOL to J.E.
Maier, Vice President,
RGE, dated
September
2,
1982.
D.M. Crutchfield, Chief
ORB ¹5,
DOL, to J.E.
Maier, V.P.,
RGE,
dated October 8,
1982.
J.E. Maier, Vice President,
RGE, to D.M. Crutchfield, Chief
ORB ¹5,
DOL, dated
December 2,
1982.
R.W. Kober,
RGE to D.M. Crutchfield, Chief
ORB ¹5,
DOL, dated
February 6,
1984.
D.M. Crutchfield, Chief ORB ¹5,
DOL to R.W. Kober,
RGE dated
April 24,
1984.
Procedures
PC-4,
"Sampling in Nuclear
Sample
Room to Determine
Hydrogen Concen-
"tration and Radiogas Activity in Primary Coolant",
Rev. 8, dated
February
22,
1985.
PC-5,
"Gamma Isotopic Analysis of Degassed
Primary Coolant",
Rev. 9,
dated
August 9,
1983.
PC-23. 1, "Alternate Emergency
Sampling of Primary Coolant",
Rev.
10,
dated April 28,
1985.
PC-23.2,
"Containment Atmosphere
Sampling
and Analysis During Contain-
ment Isolation",
Rev. 8, October
18,
1983.
PC-25.7,
"Operation of Post Accident Sample
System
(PASS)
Under
Accident Conditions
Master Procedure",
Rev. 2, dated
June
7,
1985.
~
~ ~ =
~
'1
0
Attachment I.A
PC-25.7. 1, "Conductivity, pH, Dissolved
Oz at PASS",
Rev.
2, dated
June
7,
1985.
PC-25.7.2,
"Boron Analysis at Pass",
Rev.
1, dated
June
7,
1985.
PC-25.7.3,
"Liquid Sample Dilution at PASS",
Rev.
2, dated June
7,
1985.
PC-25.7.4,
"Liquid Sample
Degassing
at PASS",
Rev.
4, dated
June
10,
1985
'C-25.7.5,
"Gas Chromatograph
Analysis at PSS",
Rev.
2, dated
June
7,
1985.
PC-25.7.6,
"Gas Dilution at PASS",
Rev.
1, dated
June
7,
1985
'C-25.7.7,
"Leak Testing of the
Gas Sampling Lines with Helium
Detection",
Rev.
1, dated
February
12,
1985.
PC-25.7.9,
"Dissolved
Oz Sensitivity Check
and Calibration Check",
Rev.
1, dated
May 24,
1985.
PC-25.7.8,
"PASS
Gas
Chromatograph
Calibration
& Calibration Check",
Rev.
3, dated
June
7,
1985.
WC-16. 1, "Post Accident Analysis of Chloride Using Ion Chromatography",
Rev.
2, dated April 2,
1985.
SM-2606.20A,
"pH Monitor Acceptance Test",
Rev.
1, dated
February
28,
1985.
SM-2606.20B,
"Determination of Liquid Dilution Ratio",
Rev.
0, dated
February
20,
1985.
SM-2606.20C,
"Gas Dilution Verification", Rev.
1, dated
May 2,
1985.
~ .
SM-2606.20D, "Liquid Degassing",
Rev.
0, dated
February 8,
1985.
SM-2606.20E,
"Boron Analyzer and Calibration Check",
Rev.
0, dated
February
13,
1985.
SM-2606.20F,
"PASS Degassing Calibration",
Rev. 0, dated
February 7,
1985.
5U,
W
~
~
Attachment I.B
Documentation for NUREG-0737
II.F. I-1&2
Rochester
Gas
and Electric Cor oration
Ginna Station
Licensee
Corres
ondence
John
E. Maier,
VP to Dennis
M. Crutchfield, Chief,
ORB ¹5,
DOL,
dated January
30,
1981.
John
E. Maier,
VP to Dennis
M. Crutchfield, Chief,
ORB ¹5,
DOL,
dated
November
19,
1981.
John Arthur,
VP and Chief Eng., to Dennis
M. Crutchfield, Chief,
¹5,
DOL, dated
November 25,
1981.
John
E. Maier,
VP to Dennis
M. Crutchfield, Chief,
ORB ¹5,
DOL, dated
January
19,
1982.
Richard J. Watts to James
Lockridge, Tech. for Energy Corp.,
dated
March 29,
1982.
NRC Memoranda
& Re orts
R.
W. Houston, Asst. Dir. for Rad. Prot. to Thomas
M. Novak, Asst.
Dir
. for Op. Reactor.
"Radiation Protection Input to Ginna Restart
SER", dated
May 12,
1982.
Inspection
Report
No. 50-2'44/82-18 for Rochester
Gas
and Electric Co.
from US NRC, Region I, dated October
13,
1982.
Daniel
R. Muller, Asst. Dir. for Rad. Prot. to Frank J. Miraglia,
Asst. Dir. for Safety Assess.,
DL, "R.E. Ginna Nuclear
Power Plant
Effluent Radioactivity Monitoring System Alarm Setpoints
and
Surveillance
Requirements
dated
May 3,
1983.
Licensee Inter-Office Memoranda
Richard A. Baker,
Responsible
Eng.,
EWR 2608A, to Duane
L. Fi lkins,
"EWR 2608A SPING-4 Effluent Monitors", dated
May 5,
1981.
David P. Hamelink, Proj. Liaison Eng., to J.C.
Bodine,
QC Eng,
"SM
Procedure
SM-2921.3 - Installation of Steamline
Radiation Monitors",
dated
December
22,
1981.
Steven
T. Adams,
Liaison Eng., to R. Baker,
Responsible
Eng.,
"EWR-
2608A High Range Effluent Monitor", dated October 26,
1983.
~
~
Attachment I.B
NRC Contractor Corres
ondence
Andrew P. Hull, NUREG-0737 Project
Leader,
BNL to Matt Haapola,
Sys.
Eng., Eberline Div. Thermo-Electric Corp., dated
June
27,
1985.
Matt Haapola,
Sys.
Eng., Eberline Div. Thermo-Electric
Corp. to
Andrew P. Hull, NUREG-0737 Project
Leader,
BNL dated July 8,
1985.
Licensee
Desi
n and 0 eratin
Documents
"Design Analysis Ginna Station High Range Effluent Monitor Isokinetic
Nozzle Design",
EWR 2608A,
Rev.
0, dated August 15,
1980.
"Calculations of Dose to Activity Conversion
Curves for Determination
of Activity Concentrations
within Lines A & B of Ginna Nuclear
Station" Technology for Energy Corporation
Report
No. R-81-020,
dated
August 1981.
Rochester
Gas
& Electric Corporation,
"Nuclear Emergency
Response
Plan",
Rev.
0, dated July 1984.
Licensee
Procedures
"Installation of Steamline
Radiation Monitor s",
Rev.
0, dated
September
29,
1981
'Estimation
of Noble Gas Release
Rate from the Plant Vent During
Accident Conditions",
Rev. 9, dated
September
27,
1983.
"Calibration and/or Maintenance of DAM-3 Steam
Line Radiation
Monitors R-31 or R-32",
Rev.
1, dated
May 8,
1984.
"Determination of Iodine or Particulate Activity", Rev.
4, dated
July 12,
1984.
"Determination of Radioactivity Release
Rates
using the High-Range
Effluent (SPING-4) Monitors", Rev.
4, dated August 16,
1984.
"Protective Action Recommendation",
Rev.
3, dated August 16,
1984.
"Estimating Off-site Doses",
Rev.
6, dated January
23,
1985.
"Calibration and/or
Maintenance
or SPING-4 Radiation Monitors for
R-12A,
R-14A and R-15A Radiation Monitoring Channels",
Rev.
6, dated
March 13,
1985.
~
~ ~
~
~
Attachment I.B
Vendor Literature
and Manuals
"
"SPING-4 Monitor Specifications",
1980
"ARM Monitor Specifications",
1981.
"Calibration of Airborne Effluent Monitors", 4506,
dated
March 23,
1984.
"Primary Calibration SA-12,
SA-13 Mid-Range
Gas Detector", dated
April 10,
1980.
"Primary Calibration SA-9, High Range
Gas Detector",
dated April 15,
1980.
Attachment I.C
Documentation for NUREG-0737
II.F ~ 1-3
Corres
ondence
and
Re orts
B.
R.
Quinn to D. L. Fi lkins, "Test of High Range
CV Radiation
Monitors" dated
May 16,
1984
B.
R. Quinn to R. J. Watts et al, "Test of High Range
CY Radiation
Monitors" dated
May 24,
1982.
R.
P.
Ulman (Victoreen) to
R. Baker (RGE), "Data Package
for High
Range
Containment Monitor" dated
December
10,
1980.
Electrical Installation Inspection
No.
Final Inspection
Report
No. QCIP-8
GINNA Design Criteria,
Rev.
1, dated April 7,
1980
GINNA Safety Analysis,
Rev.
1, dated April 7,
1980
Surveillance
Report Nos.
111804,
111883,
111904,
111921
and
112668
Drawing Nos.
GEL 876-1,
GEL877-1,
03021-233,
03021-360,
03021-364,
03021-409,
E-40830
and
E-40831
Victoreen Environmental Qualification Report
No. 950.301
Wyle Laboratory Report
No. 45050-1
Procedures
Engineering
Work Request - 2608B
Work Start Authorization Nos. 80-23,80-160,
80-176,81-219,
and 82-14
Work Procedure
EWR-2608-1
and
EWR-2608-3
Procedure
Change
Request
81-T-188,
SM-80-2608-B. 1 and 81-1607
Procedure
CP-224, "Calibration and/or Maintenance of Containment
High Range
Area Monitors" Rev.
2
, Attachment I.D
Documentation for NUREG-0737 item III.D.3.3
Procedures
HP-11. 1, "Iodine in Plant Air, Drying Tube Method" Rev.
4
HP-11.2, "Iodine in Air - Charcoal
and Silver Zeolite Cartridge
Method" Rev.
20
PC-1.4 "Operation of TN-ll Gamma Analyzer" Rev.
4
PC-1.5 "Operation of TN-4000
Gamma Analyzer" Rev.
4
PC-23.5 "Determination of Radioactivity Release
Rates
Using the High Range
Effluent (SP ING-4) Monitors" Rev.
4
RD-1.1 "Sampling
and Analysis of Containment
Iodine and Particulates"
Rev.
10
RD-3 "Sampl,ing
and Analysis of Plant Vent Iodine and Particulate
Releases"
Rev.
13
SC-410 "Inspection of Emergency
Equipment"
Rev.
16
SC-421 "Determination of Iodine or Particulate Activity" Rev.
4
SC-430 "Administration of Potassium
Iodine" Rev.
1
~
e
~
A.
Chemical
Analyses
1.
Attachment II
Com arison of Anal tical Results
PASS (Error)
653
ppm (3.4%)
2.
pH
6.8 pH(0.1pH)
3.
Dissolved
H dro
en
Normal
700
ppm
6.9
pH
Grab Sample (Error)
698
ppm (0.14%)
6.9
pH (OpH)
Acceptance
Tolerance
+/- 5/
+/- 0.3pH units
25.9 cc/kg(1.55)
26.7 cc/kg
0 (nondetectable)
B.
Isotopic Analyses
1.
Containment
Atmos here
+/-
10%
Isotope
Xe-135
~Ex ected
Sink
Grab
Sam
1 e
uCi/cc
3.3E-S
1.79E-7
Observed
Grab
Sam le
uCi/cc
2.95E-5
8.08E-8
Factor
.94
.72
~Acne tance
Factor
0.5 to 2.0
O.S to 2.0
2.
~Ex ected
Sink
~uCi~/cc
Observed
~Ieoto
e
Cs138
I131
I132
I133
I134
I135
3.29E-2
2.34E-3
5.38E-2
2.83E-2
1. 04E-1
4. 91E-2
~uCi/cc
2 '3E-2
2.72E-3
5.29E-2
2.98E-2
8.36E-2
4.97E-2
Factor
.91
1.08
.99
1.02
.90
1.00
~Acce table
O.S to 2.0
0.5 to 2.0
0.5 to 2.0
0.5 to 2.0
0.5 to 2.0
0.5 to 2.0
'U
Attachment II
3.
Dissolved
Gas
~Ex ected
Sink
~uCi/cc
Observed
~iaoto
e
'I131
I132
I133
I134
I135
5.86E-3
1. 73E-1
8.79E-2
3.09E-1
1. 48E-1
~uCi/cc
2. 72E-4
5.29E-3
2.98E-3
8.36E-3
4.97E-3
Facto
21.5
3.3
2.9
3.7
3.0
A~cue taoce
0.5 to 2.0
0.5 to 2.0
0.5 to 2.0
0.5 to 2.0
0.5 to 2.0
~
~ ~
J
S
4
C