ML20148J706
| ML20148J706 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/19/1988 |
| From: | Decker T, Gooden A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20148J689 | List: |
| References | |
| 50-321-87-32, 50-366-87-32, TAC-45950, TAC-45951, TAC-46021, TAC-46022, TAC-46093, TAC-46094, TAC-461310, TAC-46309, TAC-46310, NUDOCS 8803300354 | |
| Download: ML20148J706 (25) | |
See also: IR 05000321/1987032
Text
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UNITED STATES
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RUCLEAR REGULATORY COMMISSION
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REGloN 11
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101 MARIETTA STREET,N.W.
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AT LANT A. G EoRGI A 3032.'
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MAR 0 81988
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Report Nos.: 50-321/87-32, 50-366/87-32
Licensee: Georgia Power Company.
P. O. Box 4545
Atlanta, GA 30302'
Docket Nos.: 50-321, 50-366
License Nos.: .0PR-57, NPF-5
Facility Name:
Hatch
Inspection Conducted: DecmberJ-10,1987
-;
Inspector:
k!/
^
v
fm
/f/88
A. Gooden
.
D' ate Signed
Accompanying Personnel:
G. W. Bethke
W. G. Gloersen
.
K. C. McBride
1
J. V. Ramsdell
Approved by:
/f OO
T. R. Decker, Section Chief
Date Signed
Division of Radiation Safety and Safeguards
SUMMARY
Scope:
This special, announced inspection was an Emergency Response Facility
Appraisal. Areas examined during the appraisal included a review of selected
procedures and representative records, the ERFs and related equipment, and
interviews with licensee personnel. Select activities were observed during the
1987 annual exercise to ascertain the adequacy of the ERFs and related
equipment.
J
Results:
No violations or deviations were identified.
,
8803300354 880309
ADOCK 05000321
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TABLE OF CONTENTS
Details
1.0
Assessment of Radioactive Releases
1.1
Source Term
1.2
Dose Assessment
2.0
Meteorological Information
3.0
3.1
Regulatory Guide 1.97 Variable Availability
3.2
Functional Capabilities
3.3
Habitability
3.4
Data Collection, Storage, Analysis and Display
4.0
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4.1
Location and Habitability
4.2
Functional Capabilities
j
4.3
Regulatory Guide 1.97 Variable Availability
4.4
Data Collection, Storage, Analysis and Display
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5.0
Persons Contacted
6.0
Exit Interview
7.0
Licensee Actions on Previously Identified Findings
8.0
Glossary of Abbreviations
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1.0
Assessment of Radioactive Releases
1.1
Sou'rce Term
There were four normal release pathways at Hatch: the main
stack, Unit 1 Reactor Buildina vent, Unit 2 Reactor
Building vent, and the Recombb.or Building vent. The main
stack and the Reactor Building vents for Units 1 and 2
respectively, had both normal range and accident. range
(KAMAN) monitors. The Recombiner Building vent only had a
normal range monitor since this pathway isolated on any
accident of concern.
The normal range monitors were
scintillation detectors with two channels and a range of
101 to 105 cpm. Accident range monitors were GM detectors
with one channel and a range of 5.0 E-2 to 1.0 E 5 pCi/cc.
Calibration factors had been generated for normal range
monitors which would convert from a counts per second (cps)
'
or counts per minute (cpm) monitor readout to a release
rate in pCi/s using an appropriate vent or stack flow rate ~
.
Calibration factors were updated by procedure on a
quarterly basis with
updated
graphs
showing _ the
relationship between monitor reading and release rate found
in the Control Room. Accident range monitors readout in
pC1/cc and could be converted to a release rate using an
appropriate vent or stack flow rate.
The drywell wide range radiation monitor could be used to
generate a source term.
The monitor, located inside the
drywell, was a sensitive ionization detector with two
channels and a range of 1 to 10' R/h.
Using the
methodology from General Electric's core damage guidance
(NEDO-22215), the drywell monitor reading could be related
to percent core damage. The percent core damage could then
be related to pCi/cc concentration of noble gases and
radiciodines in the drywell based on the time after reactor
shutdown.
A source term (pCi/s) could then be generated
assuming a leak rate of 1.2% per day (the Technical
Specification maximum allowable leak rate) or by assuming
the drywell to be vented at a given flow rate.
Containment (drywell) air sample analyses obtained using
the post-accident sampling system (PASS) could be used to
determine a release rate.
The isotopic sample results
could be converted to Xe-133 dose equivalent for noble
gases and I-131 dose equivalent for radioiodines.
Then a
release rate could be calculated assuming a 1.2% leak rate
or by assuming the drywell to vent at a given flow rate.
If all monitoring instrumentation used to generate a
release source term were inoperable or offscale, a source
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tera could be generated by obtaining grab samples from or
near the release points.
1.2
Dose Assessment
The licensee's primary dose assessment model, entitled
"DOSE," was run on an IBM-AT, and was available for use in
-the EOF and the corporate office. The capabilities of DOSE
and
the
operating
instructions
were
found
in
Procedure 73EP-EIP-015-OS
entitled,
"Offsite
Dose
Assessment." DOSE calculates whole-body and child thyroid
dose rates for both. elevated and ground-level releases.
The main stack releases were considered elevated and the
vent releases (Reactor Building and Recombiner Building)
were considered ground-level. Multiple source terms could
be handled by this model (e.g. release from the main stack
and the Reactor Building vent).
Integrated ' doses were
calculated for distances of
1,
2,
5,
and 10 miles.
Calculations
could
be
updated
every
15 minutes.
Additionally, DOSE had the capability for calculating off
center-line dose rates for specific offsite distances and
center-line dose rstes at distances other than 1, 2, 5, and
10 miles.
Source terms
from the drywell wide range radiation
monitors, chemistry samples (drywell air samples or grab
samples from the release vents' or the main stack), and
effluent monitors could be entered into the model .
A
direct entry of C1/s could also be handled by the model.
The computer code used for dose assessment in the E0F used
a straight-line Gaussian model for estimating the transport
and diffusion of radioactive material released to the
atmosphere.
This
model
satisfied
the
regulatory
requirements and was appropriate for an initial assessment
performed in the Control Room, but had limitations as a
primary model for use in the EOF regarding tracking of
radioactive material releases if the wind direction changes
following the release.
In reviewing DOSE (Version IBM-2.02), it was also noted
tnat the magnitude of the horizontal diffusion coefficient
(sigma y) is limited to 1,000 m.
The license did not
provide the basis for this limitation. In addition, it was
noted that the primary diffusion model in the code includes
a meander correction term in calculations for ground-level
releases. The manner in which the term was applied follows
from Regulatory Guide 1.145. However, the Regulatory Guide
assumes that the calculations are based on hourly average
meteorological data.
It is therefore inappropriate to
apply the meander correction when diffusion computations
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are based on short-term averages of data and Jare made' at.
frequent intervals.
Thus, the seander correction should'
not be applied. to the 15-minute dose estimates. However,
it is appropriate to. ccatinue to apply. the correction to
the 2-hour dose pr
-tions.
The licensee _ agreed to
evaluate the limita
on the sigma y and the meander
correction factor us..
DOSE estimates the distance to the location of the maximum-
dose for elevated releases using a relationship attributed
to Turner's "Workbook of Atmospheric Dispersion Estimates."
This relationship is based on a specific set of curves that
describe vertical diffusion.
The relationships used to'
evaluate vertical diffusion in DOSE are based on a
_
different set of curves.
As a result,- the distances-
reported in the computer code output are not consistent
with the actual model used to estimate doses.
An
additional computer code used for dose asserment in the
E0F includes an optional diffusion model for . elevated
releases called a fumigation model that is taken from NRC
The inspector discussed with
licensee
representatives
the
guidance
provided
in
73EP-EIP-015-0S (Page 9) regarding the use of the cooling
tower plume to determine when to use the fumigation model.
Since the physical process that causes fumigation is
entirely different from the processes that cause downwash
of the cooling tower plume, the licensee agreed to evaluate
this use in the procedure.
Both the primary and backup meteorological systems included
modules that compute the standard deviation of the wind
direction (Sigma Theta). This information was recorded on
strip charts in the Control Room and E0F.
The inspector
discussed with licensee representatives the fact that sigma
theta was a better predictor of horizontal diffusion than
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Delta T.
The licensee agreed to consider sigma theta for
estimating horizontal diffusion coefficients in the dose
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assessment model .
lhe inspector noted that Regulatory
Guide 1.23 provides a means of converting sigma theta to
,
stability class, and that delta T could continue to be the
basis for estimating vertical diffusion coefficients.
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Meteorological and source term data were manually entered
into the dose assessment procedure.
Meteorological data
could be obtained from strip charts or the SPDS terminals
in the E0F,
or from the Control
Room via phone
communications.
Similarly, affluent monitor readings and
flow rate readings could
t, .; obtained from the SPOS
terminals in the EOF or from the Control Room via phone
communications. Current calibration factors for converting
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normal range effluent monitor readings from cps to pCi/cc
were obtained from the Control Room.
The inspector observed that the procedure used to obtain
meteorological data for use in dose assessmen't was subject
to error and bias.
Meteorological data used for dose
assessment in the EOF was obtained from the Control Room.
In the Control Loom, the meteorological data were manually
estimated from strip charts containing instantaneous
measurements. Meteorological data was also available from
the SPDS.
However, the meteorological data displayed on
the SPDS was not averaged; the SPDS data reflected
instantaneous variations
as
they
occurred
in
the
atmosphere. Consequently, the SPDS data was inappropriate
for use in dose assessment.
On January 20, 1988, members
of the Region II Emergency Preparedness staff discussed
with the licensee's Site Emergency Preparedness Coordinator
the current method for obtaining meteorological data used
in dose assessment.
The licensee was informed that the
current method of obtaining meteorological data for dose
assessment presented several probable sources of error.
The licensee agreed to conduct an evaluation to determine
if a more reliable and less subjective prc :edure was
necessary to ensure that meteorological data was being
compiled and computed in accordance vith RG 1.23.
DOSE must consider a certain mix of radionuclides when
entering release rate information in C1/s determined from
effluent monitoring readings.
The whole-body dose rates
were calculated using a fixed mixture of noble gas
radionuclides that represent an average mixture for two
hours after reactor shutdown.
No consMeration was given
to the change of this mixture beyona two hcurs after
shutdown. The inspector noted that by not considering this
mixture of noble gases to change, the offsite whole-body
dose rates could be overestimated.
Calculations using
IRDAM indicated that whole-body dose rates could be
approximately a factor of 10 greater for a unit release of
noble gases immediately following shutdown compared to a
unit release of noble gases 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown.
Child thyroid dose calculations with DOSE used the I-131
dose conversion factor from Regulatory Guide 1.109, thereby
assuming all the radioiodine activity to be I-131.
The
inspector noted that similar to the discussion above
regarding noble gas nuclide mixtures, the mixture of
radioiodine nuclides would also vary with time after
reactor shutdown.
Considering all the radioiodines to be
I-131 would overestimate tha child thyrrid dose during the
first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown.
Calculations using
IRDAM showed that the ch11d thyroid dose could be
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approximately a factor of 5 greater for a unit release. of
I-131 compared to a unit release of a "0-hour after
shutdown mix" of radioiodines.
DOSE had an option entitled "chemistry samples" which
allows the input of individual radionuclide concentrations.
These concentrations were converted into dose equivalent
Xe-133 and dose equivalent I-131 values by - the model for
calculation-purposes. - DOSE only allowed the input of noble
gas and radiciodine nuclides.
The licensee agreed to
consider improving the model to allow evaluation .of
additional fission products such as Cs-134 and'Cs-137.
'
Backup dose calculations would -be performed in the
Corporate Office using DOSE. If the E0F must be evacuated,
the dose assessment responsibility would be transferred to
the Corporate Office until the Alternate EOF was activated.
Manual calculations could be performed in the EOF as a
backup
method
using
the
verification
tables. in
Procedure 73EP-EIP-015-05
-(Of f site
Dose
Assessment).
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During the exercise, the inspector noted _that dose
assessment personnel
in the EOF. used the manual,
labur-intensive method to verify DOSE results.
The
.
inspector discussed with the licensee the use of the DOSE
results in the EOF to compare with those being done
simultaneously in the Corporate Office, since the manual
calculations using the verification tables -could be more
subject to error.
Procedure 63EP-EIP-053-0S (Prompt Offsite Dose Assessment)
was used to perform the ir.itial dose assessment required
for decisionmaking in the Control Room and the TSC.
The
'
procedure provides for a manual method that calculated site
boundary whole-body dose rates, assuming worst case
meteorological conditions.
The method could be used for
making initial assessment from any of the four release
pathways (main stack, Unit 1 Reactor Building vent, Unit 2
Reactor Building vent, and Recombiner Building).
The
procedure contained nomograms relating effluent monitor
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readings to site boundary dose rates, During the exercise
observation, the inspector noted that a Control Room player
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had difficulty in reading the graphs. The licer.see agreed
to consider providing enlarged versions of these graphs in
the Control Room and TSC.
,
Based on the above review, the licensee agreed to evaluate
the following:
'
Review the primary dose assessment code (DOSE) to
assure that its use of atmospheric transport and
diffusion models is appropriate, that corrections are
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made for radioactive decay and radionuclide mix after-
reactor shutdown, and that it can adequately track the
release during changes in wind direction '(50-321,
366/87-32-01). .
Review the meteorological equipment and procedures
used for the collection of meteorological data used in
dose assessment to' assure it is in accordance with
RG 1.23 (50-321, 366/87-32-02).
The following items - should be_ considered .by the licensee
<
for improvement:
Use of Sigma Theta to estimate stability class rather-
than Delta T.
,
Inclusion of additional fission products to: the
"chemistry samples" option of "DOSE" which presently -
allow entry of only noble gas and radioiodine
nuclides.
Elimination of the use of hand calculations in the EOF -
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to verify "DOSE" results. Comparison of results with
Corporate Office
personnel
conducting
parallel
calculations using "00SE."
Providing the Control Room with enlarged versions of
prompt dose assessment normograms relating effluent
monitor readings to site boundary dose rates.
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2.0
Meteorological Information
Onsite meteorological data is provided by primary and backup
meteorological systems. The primary system includes' the following
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measurements:
wind at 10, 60 and 100 m; temperature differences
between 10 and 100 m, and 10 and 60 m; temperature at 10 m; and dew
point at 10 m.
In addition, precipitation is measured near the
primary meteorological tower.
Meteorological measurements made by
the backup system include wind at the 23 m level and temoerature
difference between the 10 and 45 m levels. No problems related to
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instrument specifications, installation or exposure were no ted. The
inspector discussed with the licensee the fact that there could be
problems in the future if trees surrounding the towers are allowed _to
grow without restriction.
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Signals from the meteorological instruments go to instrument sheds
located near the bases of the towers, wnere they would be conditioned
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and displayed. The signals would then be transmitted from the shed
to the Control Room and EOF.
Although the sheds are not within
tightly controlled areas, they were within the plant fence.
The
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instruments and towers were protected from lightning, and the sheds-
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appeared to have adequate environmental control to permit the
instrumentation to operate reliably. Instrument electrical power was
obtained from an uninterruptable power supply.
Plant procedures provided for daily inspections, quarterly functional
,
checks, and semi-annual calibrations of the meteorological instrument
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systems.
Records indicated that there have been some reliability
problems with. individual instruments.
However, as a result of the
redundancies
in .the
systems,
the
availability
of
onsite
meteorological data had been maintained at an acceptable level.
Meteorological data were available in the Control Room from strip
chart recorders and via the .SPDS.
To obtain meteorological data
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appropriate for use in dose assessment, average wind directions,
speeds and stabilities had to be estimated -from the strip chart
recorders.
As a result, biases and errors in the estimation of
average meteorological conditions were possible as discussed above in
Paragraph 1.
It was noted that several of the recorders for wind
data were adjusted so that the wind speed zero was off-scale on the
low side.
This data could subsequently result in a source of
potential error in the data used in dose assessment.
The
meteorological data presented by the SPDS was not a 15 minute
average.
The data reflected the second-to-second variations that
were common in the atmosphere. As a result, the SPDS meteorological
data were not appropriate for use in dose assessment.
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Meteorological data for use in dose assessment were passed to the EOF
from the Control Room and posted on a status board. Meteorological
data were also available directly from the primary and backup towers.
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The data were recorded on strip charts. The inspector discussed with
the licensee the fact that the strip chart in the EOF containing the
temperature differences was difficult to read due to the location.
The area was dimly lit.
In addition, the range of the temperature
,
difference recorder was too wide to permit easy determination of the
differences required to resolve stability Classes B and C.
a
Fifteen minute averaged meteorological deta were available in the E0F
from the AutoData 10 data logger. Howeve.r, 73EP-EIP-015-OS did not
list the AutoData 10 as a source of meteorological data.
The only
sources listed were the status board containing data from the Control
Room and the General Office Operations Center in Atlanta (which did
not provide averaged, onsite meteorological data)
The inspector
,
discussed with the licensee that the algorithm esed by the
AutoData 10 to average wind directions was incorrect, in that the
,
algorithm produced an arithmetic average that could be ir error by as
<
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much as 180 degrees under rather common wimi direction comvnations.
Based on the above review, and as discussed in Paragraph 1, above
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(although the licensee's meteorological
system complies with
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RG 1.97), relevant recording systems and procedures for obtaining
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meteorological data used in dose assessment might not provide
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- reliable 15 minute averages of meteorological data. The inspector
discussed the above issues with licensee representatives who agreed
to consider the following:
Evaluation of the algorithm used by the AutoData 10 to compute
average wind directions for tecuracy.
Relocation, or improvement cf the lighting for the E0F
meteorological strip charts u;ed to record Delta T and Sigma
Theta.
Verification that the zero adjustment setting for the wind speed
recorders are on scale.
3.0
The Technical Support Center (TSC) was located in the Service
Building Annex, approximately two minutes walk from the main Control
Room.
The total size of the TSC was over 1,800 square feet of space.
3.1
_ Regulatory Guide 1.97 Variables Availability
P.egulatory Guide 1.97 variables were provided in the TSC
via the Emergency Response Data System (ERDS).
Two
i
components of ERDS, the Safety Parameter Display System
(SPDS) and Emergency Response Facility Display System
(ERFDS),
provided the TSC Managers with information
required for performance of their emergency response
functions.
Units 1 and 2 shared a common ERDS console in
the TSC.
The inspector reviewed a Safety Evaluation Report (SER)
provided by the licensee to demonstrate conformance with
,
RG 1.97 requirements.
According to documentation, Georgia
Power Company received a final, satisfactory SER from NRC
on the implementation of RG 1.97 on July 30, 1985.
The
only missing thermal-hydraulic or radiological parameter of
interest in the ERFs from the ERDS was the Unit 1
Recombiner Building vent radiation monitor; however, this
parameter alarms in the TSC on the analog annunciator
panel.
The Recombiner Building vent would isolate on any
accident of concern and the value would be available by
telephone.
Since 1985, several additional parameters were
added to the ERDS system,
including meteorological
parameters and drywell
sump level.
No problems were
observed regarding the availability of parameters on the:
ERDS computer system.
The ERDS system was scheduled for
daily surveillance testing to insure reliability of data
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The primary means of obtaining RG 1.97 variables was via
the ERDS system.
In addition to the parameters available
electronically on the EROS system, during activation of the
emergency plan, three dedicated data-oriented telephone
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Communicator / Recorder personnel were stationed in the TSC
with status boards. These personnel were in communication
with personnel in the Control Room and other locations, and
continuously
recorded
plant
status,
radiological
parameters, and key thermal-hydraulic parameters on large
status boards. The status boards provided a rough. trending
capability, using 15 minute data updates, to supplement the
trending capabilities of the ERDS.
In addition to the ERDS and status boards discussed above,
the TSC had an analog status wall consisting of an
annunciator panel and a Nuclear Steam Supply System (NSSS)
mimic for each reactor unit.
The annunciator panels
contained repeater annunciators for selected annunciators
located in the Control Room.
The NSSS mimics (data
acquisition panels) contained pump and valve indicators for
the primary emergency and normal core cooling / recirculation
systems (e.g. HPCI, RCIC, RHR), safety relief valves, main
steam isolation valves, etc.
Based on the above review and written procedures, the
RG 1.97 variable availability was determined to be
adequate.
3.2
TSC Functional Capabilities
The following areas of power continuity were considered in
evaluating the ability of the TSC to function during a
station
blackout
without
interruption:
data
acquisition systems, communications equipment, lighting,
and the ventilation system (HVAC).
Non-instrument loads (e.g. ventilation) in the TSC were
powered from the following diverse sources:
4 KV Bus IF and 2F (Class 1E)
Unit 1 and 2 Offsite 115 KV Power
Essential Diesel B
Lighting and instrument loads (e.g. lights, EROS Computer
System) were powered from the following diverse sources:
Security diescl
Normal plant 600 V station service
Battery backed uninterruptable power supply (UPS)
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Therefore, all TSC equipment, lighting, and ventilation
systems were supplied by reliable, redundant power
The
inspector noted that some distribution panel switches in
the TSC mechanical equipment room were poorly labeled .or
mis-labeled.
The licensee,'when informed of this matter,
provided documentation to' indicate that a plant-wide
distribution panel labeling program was reviewed - by the
Plant Safety Review Board, and would be implemented upon
approval by the appropriate authorities.
This program
should correct the problems noted in the TSC.
Several systems were available to support th? TSC in the
performance of emergency functions.
Plant system status
was available on the ERDS-(SPDS, ERFDS), the annunciator
panels, NSSS mimics, and status boards.
The SPDS portion
of ERDS had considerable trending capability for parameters
associated with the primary displays. The ERDS stored two
hours of historical data on disk for rapid access and
trending back in time. The standard ERDS trend plots were
selectable to either six or sixty. minutes of historical
data.
Trends were also maintained by listing the most
recent 4 sets of 15 minute data updates on the TSC manual
,
status boards.
Emergency Plan Implementing Procedure 63EP-RCL-005-OS
(Determination of the Extent of Core Damage Under-Accident-
conditions) described a procedure for determining the
extent of core damage as a function of the drywell
radiation monitor under accident conditions, based on the
General Electric NED0s.
Pre-calculated relationships for
the following were contained:
Coolant radionuclide concentration to core damage
Containment radiation levels to core damage
Drywell hydrogen concentration versus core damage and
percent metal-water reaction
Containment airborne radioactivity concentration to
core damage
The
inspector
further
noted
that
the
SPDS/ERFDS
thermal-hydraulic alarm setpoints were closely related to
both Technical Specification limits and Emergency Action
Level (EAL Procedure 73EP-EIP-001-05) trigger points.
While SPDS/ERFDS radiological trigger points were not
generally programmed with setpoints, the TSC annunciator
pane! contained alarms which were anticipatory to, or the
same as the EAL radiological trigger points.
Radiological
parameters alarmed on the TSC annunciator panels included
Reactor Building vents, main stack, containment high range
dome monitor, Unit 1 Recombiner Building vent, main steam
line monitors, and offgas pre-treatment.
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Based on the above review,. the area of the-TSC functional
capability.was determined to be' adequate.
3.3
TSC Habitability
The TSC Ventilation System operated satisfactorily to
. pressurize the TSC via a filter train consisting of series
components in the following order:
prefilter - heater -
HEPA - charcoal - charcoal - HEPA.
This arrangement for
the emergency ventilation system was consistent with the
recommendations in - RG 1.52.
The emergency ventilation
train was aligned upstream of and in series with the normal
ventilation train, thereby minimizing the chance of. bypass
flow during emergency operation. During system operation,
the differential
pressure
across
all- filters was
satisfactory.
A radiation monitor was installed at the-
comoined normal / emergency - suction louver to the outside
atmosphere which automatically initiated ' the system upon
sensing increased radiation . levels.
The system was
!
observed
to
pressurize
the TSC
to approximately
.09 inches wg during the emergency exercise.
The low
differential
pressure (DP) alarm was- set at about
0.1 inches wg.
This resulted in frequent low DP alarms.
The inspector discussed with licensee representatives that
it appeared that the 0.09 inches wg of DP was satisfactory
to pressurize the facility and they needed to either lower
the setpoint or take actions to improve the sealing of the
l
TSC to eliminate the spurious low DP alarms. The inspector
also noted that the charcoal adsorber sample laboratory
'
test
conditions
were
not
specified
in
Procedure 425V-X75-001-1 (Testing of TSC Filter Train by
Vendor, Revision 1,
October 22, 1985).
The licensee
representative agreed to evaluate these issues.
,
i
Based on the above review, TSC habitability appeared to be
'
adequate.
Based on the above, licensee representatives agreed to
evaluate the following:
Elimina' ion of spurious TSC HVAC low DP alarms by
lowering alarm setpoint or improving the facility
sealing
Inclusion
of
the
laboratory
test
condition
requirements
to
the TSC charcoal
vendor test
procedure,
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-3.4
TSC Data Collection, Storage, Analysis and Display
Real-time data acquisition, storage, and display were
performed by two separate computer systems for Hatch
Units 1 and 2, namely:
(1) a Safety Parameter Display
System; and (2) an Emergency Response Facility Display
_
System.
The computers, signal generators, and graphics
display devices were military standard units.
The SPDS
consisted of 1 R0W MSE/14 computer, _1 OTI graphic display
i
system, and a fiber optic communications link.
The
configuration of the ERFDS included 1 R0 W MSE/14 computer
system, 1 Ramtek graphics display generator,_ 2 Ramtek
graphics display CRTs _(GM-850's) in the TSC and 2 GM-850's
in the EOF, a magnetic tape system, and a fiber optic
communications link.
The following table shows the sensor distributions for
Hatch Units 1 and 2.
ERFDS
ERFDS
Analog
Analog
Digital
Total
Unit #
Sensors
Sensors
Sensors
Sensors
>
1
81
110
698
889
2
87
113
704
'904-
,
Sensor information was collected from Foxboro analog,
Validlyne analog, and Cutler Hammer Digital intelligent
front ends.
With the use of - a fiber optic link and
R0 W 3552 transducers, sensor information gathered by
,
either the SPDS or the ERFDS was routinely passed to the
other at 5 megabits /second.
This data was written to
shared memory in each computer system so that immediate
access to all sensors monitored by both systems was
available for processing.
Failure of either the SPDS or
the ERFDS would cause only part of the safety parameter
data to be available at the ERF display CRTs.
SPDS and ERFDS systems were configured with R0W 4050/5042
8" hard di sks.
Every second, complete sensor sample sets
were collected,
analyzed, stored to hard disk, and
In the TSC, there were two graphic display CRTs (cathode
ray tubes) controlled by Ramtek-9400M/I graphics generators
(located in the Control Room), Ramtek GM-850 display CRTs
'
were driven
by RGB (red green-blue) video signals
transmitted from the Ramtek graphics generator at 10 MHZ
(megaherz).
Users could select the Hatch Unit desired and
display safety parameters or parameter sets of interest,
a
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On request an RGB encoder translated the video signal into
.a faster format to produce a black and white hard copy.
Graphics displays were updated by the SPDS/ERFDS computer _
systems every 1 to 10 seconds depending on the priority of
the graphics display task in use.
Users could select
'
critical parameter displays using function keys and a
screen menu. Displays that the user could select included
i
the following:
Primary
Trend
Diagnostic
Emergency
Log in/ Log out
Typical displays included a graphical representation of a
I
plant system with real-time updated parameters. - In.several
cases, trend graphs of safety parameters could be shown on
the display to provide additional information about how
these critical parameters were changing.
A user's manual defining operation and use of the display
system was reviewed.
The manual
provided adequate
instructions and examples of typical safety parameter
displays.
The function key / menu system was user-friendly
and systems' functions could be easily learned.-
i
Unit 1 SPDS and ERFDS computers read, analyzed, and stored
to hard disk 889 analog and digital sensors every second.
Unit 2 SPDS/ERFDS computers read, analyzed, and stored to
i
hard disk 904 analog and digital sensors overy second. The
SPDSs were dedicated systems and were not used to support
other data processing tasks.
The ERFDSs were also
dedicated systems; however, they did read meteorological
'
data from the MDCS and produced a meteorological display on
'
request.
Licensee representatives reported that load
testing, such as requesting displays concurrently on all
,
aveilable display terminals, produced no observed system
'
degradation.
,
Class 1E optical isolators had been used to isolate the
SPDS from the signal source. Also, the ERFDS was reported
to be isolated from Class 1E plant systems through suitable
means.
Validation
and
verification
(V&V) performed
independently by Bechtel, Gaithersburg indicated that no
evidence of signal degradation, interference, or damage was
l
observed (documentation entitled "Emergency Response Data
!
Systems Units 1 and 2 Validation Report," by H. S. Kassel,
Jr. Project Engineer and L. D. Wechsler, both of Bechtel
Eastern Power Corporation, July 1986).
,
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14
The data communications components of. the SPU5 and ' the
ERFDS were adequate .to support data acquisition, analysis,
display, and storage for TSC requirements. As mentioned
earlier, the SPDS and the ERFDS shared data via a
5 megabit /second fiber optic CPU to CPU link.
Error
checking devices were installed to insure correct data
transmissions.
CRT displays in the TSC were updated
through the use of RGB balanced line cabling at a rate of
45 megahertz.
Time resolution for data communications was
adequate.
The SPDS and the ERFDS were dedicated systems designed to
support plant safety monitoring and reporting needs.
Utility personnel interviewed reported that running all
display monitors concurrently caused no observable system
degradation.
The ROLM computers were multi-tasking units
that were programmed to acquire, analyze, and store data on
a first priority approach.
The SPDS primary display
showing real-time critical reactor parameters also executed
at highest priority.
Other displays and functions -were
handled at lower priorities.
The way that software tasks
were scheduled allowed the SPDS and ERFDS systems to
complete critical tasks without degrading the systems.
Data
storage was functionally implemented to meet
NUREG-0696 requirements.
Interviewees reported that at any
time, three hours of historical data were available .to
provide trending information on critical plant parameters.
On demand, or on a reactor "SCRAM," plant parameter data
would be continuously stored on a magnetic tape. This data
would continue to be stored until the tape storage process
was manually halted or until another tape was mounted.
Tapes would fill to capacity every 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> and require
changing.
Data
storage
capability
could
continue
indefinitely as long as it was needed.
Currently if a
computer failed and needed restarting ("rebooting"),
historical data filed on the computer disks would be lost.
If this occurs, data storage requirements would not be met.
The utility was in the process of correcting this sof tware
problem (Software Modification Request (SMR) No. 45).
The V&V report by Bechtel Eastern Corporation was discussed
above.
In addition, the inspector reviewed the model
descriptions in the Utility Functional Specification.
On
reviewing model descriptions in the Utility Functional
Specification, an error was discovered in the "rate"
algorithm documentation.
The error was reported to the
licensee,
and later during the ERF appraisal,
the
Functional Specification document was updated with a new
hand written "rate" algorithm (Page 3.1 of the Functional
Specification).
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15
Utility contacts reported that SPDS analog sensors were
redundantly sampled.
The following describes how two
independent values for each safety parameter were handled:
If both sensor values were within pre-established
limits, their average was displayed using a green
background
If only one value was outside specified limits, the
other was displayed using a yellow background
If both sensor values were outside limits, their
average was displayed using a red background
Manual
logging of computer system unavailability was
performed by the licensee. The unavailability was reported
to be less than .05 which is acceptable.
However, the
licensee contact stated that no manual data entry processes
were employed.
The computer systems used to support ERF
functions were military standard units and did not require
special
environmental
control.
Air conditioning was
reported b, licensee personnel to be functional in the CR,
The licensee received a final SER on RG 1.97 variable
implementation in 1985.
Additional parameters have been
2
added to the ERDS system since the 1985 SER. All of the RG 1.97 parameters approved by NRC were included as inputs to
the ERDS. Plant Hatch was one of the six plants evaluated
in the 1985 SPDS Pilot Evaluation Program.
Although no
formal SER was received by GPC on the SPDS, the Pilot
Program report stated that the system met or exceeded the
NUREG-0737, Supplement I requirements.
According to the
documentation, the SPDS was shown as complete on the Hatch
integrated schedule.
As a result of the aforementioned
reports and additional reviews of the ERDS system during
the ERF Appraisal, the TSC data base met the ERF
requirements of NUREG-0737.
Based on the above review the licensee agreed to evaluate
and take appropriate action on the following:
Verify that the completionof the data acquisition
system
prog ram
upgrade
(SMRN45)
includes
the
capability
to
produce
full
data
(50-321,
366/87-32-03).
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16
4.0
4.1
EOF Location and Habitability
The EOF was located.in the East Wing of the Training Center-
approximately 0.2 miles south of the plant. An Alternate
E0F was located in the Georgia Power Company District
Office in Baxley, GA beyond the 10 mile radius (i' e. ,
approximately 10.1 miles from _ the power block).
This was
consistent with Option 1
in Table 1 of NUREG-0737,
Supplement 1.
The EOF ventilation system emergency mode isolated the
facility from the outside atmosphere and -placed a HEPA
filter in the recirculation flow path.
The system was
designed to meet the minimum requirements of NUREG-0737 for
near-site EOFs. A periodic testing program (including 00P
testing) was in place.
The inspector discussed- with
licensee representatives test procedure, 42SP-111187-RY-1N,
"Test of EOF Filter Train by Vendor," regarding its
revision and status.
The licensee agreed to review the
procedure to determine the need for additions to a final
procedure to include provisions to perform bypass testing
to insure closure of the outside air damper and the two
dampers which isolate the normal, parallel dust filter flow
path.
The licensee agreed to investigate and correct the apparent
HEPA filter bypass flow problem which was identified during
the Appraisal.
In addition, the licensee agreed, through
the use of procedures and warning signs, to keep the doors
to the three access points to the EOF closed during
potential airborne radiological problems (and exercises).
Based on the above review, EOF location and habitability
appeared to be adequate.
i
Licensee representatives agreed to evaluate and take
i
appropriate action on the following items:
'
Revision of and establishment as permanent the EOF
HVAC testing procedure, and inclusion of steps to test
the isolation dampers in the system.
Revision of Section H-3 of the Hatch Emergency Plan to
reflect
the
E0F HVAC
system as
an
isolated
recirculating system versus a pressurized system.
Investigation and correction of the apparent HEPA
filter bypass flow problem which was identified during
the appraisal.
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r
Use of procedures and warning signs to keep the doors
1
to the -three access points to the EOF closed .during
potential airborne radiological problems _ and annual
emergency preparedness exercises.
4.2
E0F Functional Capabilities
With the exception of the TSC analog annunciator and mimic
panels, the EOF data acquisition system and procedures were
almost identical to those in the TSC- (see Paragraph 3).
Status boards in the EOF were more oriented toward dose
assessment than those in the TSC, since the EOF would be
the primary facility for this function.
In the Alternate EOF, four work areas were available. Two
of the areas provided room for six desks, and the remaining
two areas had room for about two desks.
There ' were
approximately eight phones available in the building. A
storeroom was packed with extra tables and approximately 50
chairs.
Backup power was provided- by a gas powered
generator which was tested weekly.
All ' other equipment
would be brought from the primary E0F.
The Alternate EOF-
_'
was not capable of handling all personnel staffing the
primary EOF. The licensee agreed to identify the key staff
i
to be required to report to the Alternate EOF and define
their placement in the facility.
~
j
Locations of the primary and alternate EOFs are defined in
'
Section 4.1,
above.
Both facilities were ' on a major
north-south 115 KV distribution line; however, the line was
i
divided between the facilities by the Baxley Substation,
j
The two sections of the 115 KV line had independent feeds.
The likelihood of common mode failure was very small since
both sides of the Substation would have to be damaged to
disrupt power to both EOF facilities.
On this-basis, the
near-site E0F power supplies were not reviewed in detail or
tested during the Appraisal. All instruments (e.g. ERDS
-
and Do e Assessment computers) in the near-site EOF were
supplied by reliable, UPS backed power'.
Non-instrument
loads (e.g. lights) were on normal offsite -115 KV power.
but the facility had several wall mounted, battery powered
emergency lighting units.
=
'
Based on the above review, the licensee agreed to evaluate
,
and take appropriate action on the following:
'
,
.
4
'
Verification that plans and/or procedures for the
backup EOF to identify key staff who would report to
the Alternate EOF, where they would be positioned in
the facility, and other logistical requirements such
_
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18
as transportation, equipmer.t, etc. would be provided
(50-321,366/87-32-04).
4.3
Regulatory Guide 1.97 Variable Availability
The data availability in the E0F was essentially the same
j
as that in the TSC, with both the ERDS computer system and
Communicator / Recorder personnel serving as the primary and
backup sources of data. With minor exceptions, information
in the paragraphs that follow were identical to the
corresponding TSC section of this report.
The Georgia Power Company (GPC) RG 1.97 SAR submission to
NRC was reviewed.
This document contained an excellent
summary of the RG 1.97 variables available in the TSC and
EOF on the ERDS. ERDS consists of two Milspec subsystems,
SPDS and the ERFDS. GPC received a final, satisfactory SER
from NRC on implementation of RG 1.97 on July 30, 1985.
The only "missing" radiological parameter of interest in
the ERFs from the ERDS was the Unit 1 Recombiner Building
'
Vent Radiation Monitor.
The Recombiner Building vent
isolates on any accident of concern and the value would be
available by telephone.
As discussed in the above paragraph, parameter availability
was satisfactory in 1985. Subsequently, several additional
parameters were added to the ERDS
system including
meteorological parameters and drywell sump level.
No
problems were noted with availability of containment
condition and radiological parameters on the ERDS computer
system.
In
addition
to
the
parameters
available
electronically on the ERDS system, GPC stationed dedicated
telephone Communicator / Recorder personnel in the E0F,
These personnel were in communication with personnel in the
Control Room, the TSC, and other locations.
Additionally
plant status records, radiological parameters, and key
thermal-hydraulic parameters were periodically posted on
large status boards.
The status boards provided a rough
trending capability, using 15 minute data updates, to
supplement the trending capabilities of th ERDS.
The combination of data available on the ERDS system,
facility status boards, and communicators appeared to be
)
satisfactory.
j
l
Based on the above review, EOF variable availability was
i
determined to be adequate.
1
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19
4.4
E0F Data Collection, Storage, Analysis and Display
The same computers supporting TSC activities also supported
the EOF. These systems and details of their functions have
have been described in Paragraph 3 above. EOF display CRTs
were the same as those located in the TSC, and allowed
users to view Units 1- or 2 parameters on request. - The EOF
'
was reported to be approximately 0.2 ' miles from the'~ ERF
!
computers in the service building, and as such, used the
same cabling-scheme as the TSC for the display CRTs.
'
GPC received a final SER on RG 1.97 variable implementation
in 1985. Additional parameters have been added to the ERDS
,
system since the 1985 SER. All of the RG 1.97 parameters
approved by NRC were included as inputs to the ERDS. Hatch
,
was one of the six plants evaluated in the 1985 SPDS Pilot
Evaluation Program. Although no formal SER was received by
GPC on the SPDS, the Pilot Program report stated that the
.
system met or exceeded the NUREG-0737, Supplement 1
l
requirements.
Based on these two reports and additional
reviews of the ERDS system during the ERF Appraisal, the
E0F data base met the ERF requirements of the referenced
supplement.
Based on the above review, EOF data collection, storage,
I
analysis, and display were determined to be adequate.
5.0
Persons Contacted
- J. Beckham, Vice-President
- S. Bethay, Supervisor, Nuclear Safety and Compliance
- C, Coggin, Manager, Training and Emergency Preparedness
- G. Creighton, Regulatory Specialist, Nuclear Safety and Compliance
- R.
Dedrickson, Assistant to Vice-President
i
- S. Ewald, Manager, Radiological Safety
- P. Fornel, Manager, Maintenance
- 0. Fraser, Manager, Site ~ Quality Assurance
j
- R. Hayes, Deputy Manager, Operations
- J. Heidt, Manager, Nuclear Licensing
"L. Hill, Manager, Nuclear Emergency Preparedness
- R. Mothena, Supervisor, Nuclear Emergency Preparedness
- H. Nix, Plant Manager
- T. Powers, Manager, Engineering Support
- D. Read, Manager, Plant Support
- D. Smith, Superintendent, Health Physics
P. Underwood, Shift Technical Advisor
- E. Wahab, Superintendent, Balance of Plant Engineering
- R. Zavadoski, Manager, Health Physics / Chemistry
Other licensee employees contacted included engineers, technicians,
operators, and security force members.
'
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.
20
Other Organizations
- B. Edmark, Senior Startup Engineer, Bechtel Corporation
Nuclear Regulatory Commission
- L. Crocker, Project Manager, NRR
- A. Cunningham, Senior Radiation Specialist, RII
- T. Decker, Chief, Emergency Preparedness Section, RII
- G. Lapinsky, Engineering Psychologist, NRR
..
- J. Menning, Resident Inspector
- M.
Sinkule, iaction Chief, Reactor Projects, RII
- Attended exit interview on December 10, 1987
- Participated in conference call on January 15, 1988
- Participated in telephone exit interview on January 20, 1988
6.0
Exit Interview
The inspection scope and findings were summarized on December 10,
1987, with those persons indicated in Paragraph 5 above.
The
inspector described the areas inspected and discussed in detail the
inspection findings listed below.
Additionally, the inspector
discussed actions initiated by the licensee during the inspection to
address some of the inspection findings (as noted in Paragraphs 3.2
and 3.4).
Members of the Region II Emergency Preparedness staff telephoned
licensee representatives on January 15, 1988, and January 20, 1988,
to inform the licensee that a review of the report details presented
in Paragraph 1.2 above resulted in an additional open item.
A
detailed review of the licensee's dose assessment procedure resulted
in an Inspector Followup Item (review and assure that the
meteorological data used in dose assessment is 15-minute averaged
data).
No dissenting comments were received from the licensee.
Although proprietary material was reviewed during the i nsper.ti on ,
i
such material was neither removed from the site nor entered into this
report.
j
Item Number
Status
Description / Reference Paragraph
j
50-321, 366/87-32-01
Open
IFI - Evaluation of the adequacy
of the EOF dose assessment model
(assumptions, correction factors,
'
computer
codes,
etc.)
for
continuously
assessing
the
consequences
of
a
radioactive
release
to
the
environment
(Paragraph 1.2).
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4
50-321, 366/87-32-02
Open
Review'of the meteorological
instruments and/or equipment and
procedures, and assure that the
meteorological data used in dose
assessment is in accordance with
RG 1.23 (Paragraph 1.2).
50-321, 366/87-32-03
Open
IFI - Complete Software
Modification Request No. 45 to
ensure that the data acquisition
system program upgrade includes
the capability to produce full
<
data
file
recovery- following
computer failures (Paragraph 3.4).
,
50-321, 366/87-32-04
Open
IFI - Verification that the plans
and/or procedures for the backup
E0F identify key staff who would
report to the alternate EOF, where
they would be positioned in the
facility.
and other logistical
requirements
such
as
transportation,
equipment
etc.
l
(Paragraph 4.2).
50-321, 366/87-18-04
Closed
IFI - Verification of timely shift-
staff augmentation times using
periodic announced and unannounced
communications
drills
(Paragraph 7.a).
50-321, 366/87-18-06
Closed
IFI - Compare dose assessment
results
from the prompt dose
assessment,
computerized
dose
assessment, state dose assessment,
,
and NRC dose assessment (IRDAM)
methods using standard benchmark
problems (Paragraph 7.6).
50-321, 366/85-25-03
Closed
Exercise Weakness - Improvements
to Public Information program in
coordination of press releases and
,
emergency news information between
the licensee and offsite officials
'
(Paragraph 7.c).
.
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22
7.0
Licensee Actions on Previously Identified Inspection Findings (92701)
a.
(Closed) Inspector Followup Item 321, 366/87-18-04. . Verify
timely shif t staff augmentation times using periodic announced
and unannounced communication drills.
The inspector determined that the availability of augmentation
personnel for the onsite emergency organization, as specified in
the Emergency Plan, was uncertain because such availability had
not been tested by announced or unannounced drills.
The inspector reviewed documentation dated September 14, 1987,
and October 18, 1987, entitled, "Announced and Unannounced
Callout."
The aforementioned documentation verified that
personnel were contacted for estimated time of arrival at the
plant to ensure Table B-1 augmentation times as specified in the
b.
(Closed) Inspector Followup Item 321, 366/87-18-06.
Compare
dose assessment results from the prompt dose assessment,
computerized dose assessment, State dose assessment, and NRC
dose assessment (IRDAM) methods using standard benchmark
problems.
A previous inspection resulted in the absence of documentation
l
by the licensee to show that a comparison test between the
'
licensee, NRC, and State dose assessment model had been
conducted.
The inspector was provided documentation which verified a
comparison of the licensee computerized method ("DOSE") and
prompt dose assessment method to NRC's IRDAM (Rev. 5) and the
State of Georgia's had been performed for 10 test cases.
>
Differences between the models were identified.
c.
(Closed) Exercise Weakness 321, 366/85-25-03.
Improvements to
Public Information Program in the coordination of press releases
,
and emergency news information between the licensee and offsite
officials.
.
During a FEMA /NRC meeting following the 1985 annual exercise,
problems were noted in the coordination of news releases and
emergency information between the licensee
and offsite
officials.
The inspector noted during the 1987 annual exercise that press
releases were well coordinated between the licensee and offsite
officials, and releases were maae in a timely manner.
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23'
1
8,0
Glossary of Acronyms and Initialisms
CPU
Central Processing _ Unit
CR
' Control Room
Cathode Ray Tube
Data Acquisition System
,
'DP
Differential Pressure
Emergency Action-Level
i
E0F
Emergency Response Data System
)
Emergency Response Facilities
-i
.
ERFDS
Emergency Response Facility Display System
GM
Geiger Muller
GPC
Georgia Power Company
High Efficiency Particulate Air (Filter)
High Pressure Coolant. Injection
Heating, Ventliation, Air. Conditioning
)
IRDAM
Interactive Rapid Dose Assessment Model
'
MDCS
Meteorological Data Collection System
,
Nuclear Steam Supply System
j
Post Accident Sampling System.
'
Reactor Core Isolation Cooling
Regulatory Guide
RGB
Red Green Blue
Residual Heat Removal System-
Safety Evaluation Report
Safety Parameter Display System
J
Uninterruptable Power Supply
Validation and Verification
4'
wg
Water Gauge
's
i
La
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