ML20148J706

Insp Repts 50-321/87-32 & 50-366/87-32 on 871207-10.No Violations or Deviations Noted.Major Areas Inspected:Review of Selected Procedures & Representative Records & Emergency Response Facilities & Related Equipment

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: ML20148J684 ML20148J706
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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$ RUCLEAR REGULATORY COMMISSION

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C' AT LANT A. G EoRGI A 3032.'

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\,+.**+/ MAR 0 81988

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!/

A. Gooden

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D' ate Signed

Accompanying Personnel: G. W. Bethke

W. G. Gloersen .

K. C. McBride 1

J. V. Ramsdell l

Approved by: /f OO I

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.

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8803300354 880309

PDR 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 Technical Support Center

3.1 Regulatory Guide 1.97 Variable Availability

3.2 Functional Capabilities

3.3 Habitability

3.4 Data Collection, Storage, Analysis and Display

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4.0 Emergency Operations Facility

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  ;

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7.0 Licensee Actions on Previously Identified Findings

8.0 Glossary of Abbreviations l

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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 i

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 ~ .

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Calibration factors were updated by procedure on a i

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 I

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 i

Specification maximum allowable leak rate) or by assuming l

the drywell to be vented at a given flow rate.

Containment (drywell) air sample analyses obtained using l

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

Regulatory Guide 1.145. 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 j

Delta T. The licensee agreed to consider sigma theta for  ;

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estimating horizontal diffusion coefficients in the dose

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 i

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basis for estimating vertical diffusion coefficients.

Meteorological and source term data were manually entered  ;

into the dose assessment procedure. Meteorological data l

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). -

During the exercise, the inspector noted _that dose

assessment personnel in the EOF. used the manual,

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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 j

had difficulty in reading the graphs. The licer.see agreed  ;

to consider providing enlarged versions of these graphs in

the Control Room and TSC. ,

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Based on the above review, the licensee agreed to evaluate I

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the following: l

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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).

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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 - '

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

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meteorological systems. The primary system includes' the following

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

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difference between the 10 and 45 m levels. No problems related to

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

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located near the bases of the towers, wnere they would be conditioned

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

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 ,

i checks, and semi-annual calibrations of the meteorological instrument

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 '

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

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data were also available directly from the primary and backup towers.

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.

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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

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AutoData 10 to average wind directions was incorrect, in that the

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algorithm produced an arithmetic average that could be ir error by as ,

much as 180 degrees under rather common wimi direction comvnations.

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Based on the above review, and as discussed in Paragraph 1, above

(although the licensee's meteorological system complies with i

RG 1.97), relevant recording systems and procedures for obtaining '

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 Technical Support Center

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 1

Power Company received a final, satisfactory SER from NRC l

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 l

daily surveillance testing to insure reliability of data i

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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

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emergency plan, three dedicated data-oriented telephone

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: TSC 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. ,

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Based on the above review, TSC habitability appeared to be i

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

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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.

SPDS 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 I

other at 5 megabits /second. This data was written to I

shared memory in each computer system so that immediate l

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

displayed to the ERF CRTs..

In the TSC, there were two graphic display CRTs (cathode

ray tubes) controlled by Ramtek-9400M/I graphics generators i

(located in the Control Room), Ramtek GM-850 display CRTs '

were driven by RGB (red green-blue) video signals l

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,

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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 l

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 l

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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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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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,

TSC, and the EOF.

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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4.0 Emergency Operations Facility

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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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.

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Locations of the primary and alternate EOFs are defined in

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Section 4.1,

'

above. Both facilities were ' on a major

i north-south 115 KV distribution line; however, the line was

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. l

but the facility had several wall mounted, battery powered  !

emergency lighting units. l

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, Based on the above review, the licensee agreed to evaluate

,

and take appropriate action on the following: '

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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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as transportation, equipmer.t, etc. would be provided  !

(50-321,366/87-32-04). l

4.3 Regulatory Guide 1.97 Variable Availability l

The data availability in the E0F was essentially the same j

as that in the TSC, with both the ERDS computer system and 1

Communicator / Recorder personnel serving as the primary and l

backup sources of data. With minor exceptions, information I

in the paragraphs that follow were identical to the  !

corresponding TSC section of this report. l

The Georgia Power Company (GPC) RG 1.97 SAR submission to l

NRC was reviewed. This document contained an excellent i

summary of the RG 1.97 variables available in the TSC and  ;

EOF on the ERDS. ERDS consists of two Milspec subsystems, I

SPDS and the ERFDS. GPC received a final, satisfactory SER l

from NRC on implementation of RG 1.97 on July 30, 1985. 1

The only "missing" radiological parameter of interest in

the ERFs from the ERDS was the Unit 1 Recombiner Building i

Vent Radiation Monitor. The Recombiner Building vent

'

isolates on any accident of concern and the value would be

available by telephone.

l

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

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Based on the above review, EOF variable availability was i

determined to be adequate.  ;

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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 I

• C, Coggin, Manager, Training and Emergency Preparedness  !

• G. Creighton, Regulatory Specialist, Nuclear Safety and Compliance  !

• R. Dedrickson, Assistant to Vice-President i

1. S. Ewald, Manager, Radiological Safety l

• P. Fornel, Manager, Maintenance l

• 0. Fraser, Manager, Site ~ Quality Assurance j

• R. Hayes, Deputy Manager, Operations

• J. Heidt, Manager, Nuclear Licensing

"L. Hill, Manager, Nuclear Emergency Preparedness

1. 1. 1. • 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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Other Organizations

• B. Edmark, Senior Startup Engineer, Bechtel Corporation

Nuclear Regulatory Commission

• L. Crocker, Project Manager, NRR

1. A. Cunningham, Senior Radiation Specialist, RII

1. 1. 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

1. Participated in conference call on January 15, 1988

1. 1. 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 l

report. j

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Item Number Status Description / Reference Paragraph j

50-321, 366/87-32-01 Open IFI - Evaluation of the adequacy l

of the EOF dose assessment model l

(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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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 l

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 )

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the licensee and offsite officials

(Paragraph 7.c).

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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

Emergency Plan.

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

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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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8,0 Glossary of Acronyms and Initialisms

CPU Central Processing _ Unit

CR ' Control Room

CRT Cathode Ray Tube

DAS Data Acquisition System ,

'DP Differential Pressure  ;

EAL Emergency Action-Level i

E0F Emergency Operations Facility

ERDS Emergency Response Data System

ERFs Emergency Response Facilities )

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ERFDS Emergency Response Facility Display System l

GM Geiger Muller

GPC Georgia Power Company

HEPA High Efficiency Particulate Air (Filter)

HPCI High Pressure Coolant. Injection  :

HVAC Heating, Ventliation, Air. Conditioning )

IRDAM Interactive Rapid Dose Assessment Model '

MDCS Meteorological Data Collection System ,

NSSS Nuclear Steam Supply System j

PASS Post Accident Sampling System. '

RCIC Reactor Core Isolation Cooling

RG Regulatory Guide

RGB Red Green Blue l

RHR Residual Heat Removal System-

SER Safety Evaluation Report

SPDS Safety Parameter Display System l

TSC Technical Support Center J

UPS Uninterruptable Power Supply

V&V Validation and Verification l

4' wg Water Gauge  ;

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