ML20058F391
| ML20058F391 | |
| Person / Time | |
|---|---|
| Site: | U.S. Geological Survey |
| Issue date: | 10/30/1990 |
| From: | Danni Smith INTERIOR, DEPT. OF, GEOLOGICAL SURVEY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9011080250 | |
| Download: ML20058F391 (12) | |
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^3 United States Department of the Interior E
Gl:OI.OGICAl. $1'It VEY llON 25010.
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October 30, 1990 Document Conte..!%sk U.S. Nuclear Regalatory Commission Washington, D.C.
20555 Re USG8 TRIGA Reactor Facility Docket No. 50-274, License R-113 Gentlemen:
This letter is to request an amendment to the license and technical specifications for the US Geological Survey TRIGA research reactor facility (G8TR),
Specifically, this request is to allow the installation and use of a computerized control system at the reactor facility.
I request that this amendment become effective only upon installation of the new control system. This will allow continued use of the current control system until the new system is installed at some future, unspecified date. This request is made in accordance with the provisions of 10 CFR 50.90.
If you have questions concerning this request, please contact Tim DeBey, the facility supervisor, at. (303) 236-4726 or FTS 776-4726.
Sincerely, l
JA
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David Smith l:
GSTR Administrator I
1-l Copy to USNRC Region IV Ow 9011080250 901030 PDR ADOCK 05000274 P-PDC
INTRODUCTION The US Geological Survey has been operating its TRIGA reactor facility since 1969 with the original control system.
Although this system has performed well over the years, it is desirable to take advantage of new computerized technology for enhanced operation.
A new computerized control system is available and should provide improved reliability, easier repair, and increased information for the operator.
This change of control systems has been reviewed and approved by the GSTR Reactor Operations Committee for submission to the NRC for review.
At least three other TRIGA reactor facilities have been U
authorized use of essentially identical computerized control systems: the McClellan AFB TRIGA, the General Atomics TRIGA Hark I,
and the ATRRI TRIGA Hark F.
The new control system would provide a complete replacement of the existing control console and the existing regulating rod drive..This includes the replacement of all safety circuits and power indicating channels, although many of the same sensors will be used with the new system.
The Technical Specifications
~
need minor revisions to carry out the control system replacement.. Details of the control system change and the requested Technical Specification changes are in the following pages.
It is important to note that no date is scheduled for the actual control system replacement and this request should only be approved to become effective upon installation of the new control system.
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Q-t NEW SYSTEM OVERVIEW The computerized control system has the following key elements: a control console (CSC), a data acquisition unit (DAC), and three reactor power monitors.
(See rigure 1.)
The CSC is a desk-type unit that is similar in style to the old console.
Reactor operators will give commands to the system through switches on the console and will receive feedback information through two video monitors, bar displays, lights, and audible signals.
One video monitor provides symbolic displays of important reactor parameters while the other monitor provides a listing of the values of all available reactor parameters.
The second display has three windows: one for reactor. status, one for warning messages, and one for scram messages.
The CSC also cantains a near letter-quality printer that allows the production of hard copies of the video screens and other reactor data.
Historical records also can be kept on removable computer disks for permanent storage.
This provides the capability of maintaining records for replay, analysis, and training.
-A strip chart recorder is also provided to give a hard copy of reactor power whenever desired by the operator.
.The new. system operates in the same modes as the original controlsC Hanual, Automatic, Pelse, 'and square Wave.
A new rod drive mechanism for the regulating rod is included with'the'new system because of the digital method that ic used to. control the rod position during automatic operation.
This new mechanism.uses a stepping motor to provide rod motion.
The new regulating red drive is otherwise essentially the same as
!the original-drive.
The fuel-follower control rod and connecting rod will not be changed.
SAFETY CHANNELS Two high flux saf2ty channels report the reactor power level E
as-measured with two ion chambers placed near the core.
Each safety channel is a part of one NP(P)-1000 analog instrument.
For redundancy, two channels are used, operating identically during steady state operations.
The steady state-power level is displayed on two separate analog LED bargraph indicators and on the. reactor control CRT.
Only the CRT-displayed level involves digital processing.
'The NP-1000 instrument is shunted during pulse operation and the NPP-1000 instrument measures pulse power parameters.
Calibration checks of both instruments are performed automatically during the console prechecks.
Any failures detected by the prechecks will be automatically reported to the operator.
The NP(p)-1000 units form part of the scram logic circuitry.
When the. steady state power level, measured by either instrument, reaches the maximum power level determined by the trip setpoint, a bistable circuit is activated.
The bistable then opens a contact in the
-gnet power circuit to cause an immediate scram.of all control rods..
All the circuitry up to and including the magnet power circuit is analog and receives no digital processing.
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1 NH-1000 POWER CHANNEL The NH-1000 is a neutron flux monitoring instrument that performs digital processing on the power signal to provide
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linear.nd logarithmic power indications along with the reactor period.
This instrument does not to provide any of the required scrams however, it does provide a non-required period scram.
A fission chamber is the detector for the NH-1000, feeding either a count rate amplifier or a campbelling amplifier.
The count rate amplifier is in use at low powers while the campbelling amplifier is used at high powers.
The analog signal from these amplifiers is converted to a digital signal and processed.
The NH-1000 also provides interlock signals for the source interlock and the 1 kW pulsing interlock.
The NH-1000 is automatically checked for calibration errors and. component failures during the prestart checks of the console.
Any problems detected during the checks will be reported to the operator.
l TUEL TEMPERATURE CHANNELS Two fuel temperature measurement and indication channels exist on the now console.
These measure the temperatures of the
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fuel elements near the centerlines of the fuel elements.
There are.two identical channels to provide redundancy and reliability. The fuel temperature indications are displayed on separate analog LED bargraph indicators and the console CRTs.
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-Calibration checks are automatically performed during the console prechecks.- Any failures detected will be automatically reported to the operator.
The fuel temperature units do not provide any scram funct. ions for the USGS reactor.
Although bistable outputs are
.available;from the instruments, they-function as warnings only.
SCRAM SYSTEH The scram circuitry assures that.a minimum set of reactor conditions must be met before the reactor can be operated.
The scram circuit controls the application of electrical power to the control rod magnets.
This is done by placing a set of relay
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switches in series in the magnet power line.
These relays open on failure to cause a scram upon component failure or loss of power.
Magnet power also provides power to the transient rod l
air supply solenoid..
Interruption of magnet power by opening of any one relay switch will cause all control rods to drop to their-full down positions.
The scram system is comprised of inputs from analog instruments except the reactor period scram that is derived from the NH1000 digital power instrument.
The period scram is not required by'the technical specifications, but has been included in the GSTR console since initial operation.
The following list-is a complete itemization of all the.GSTR scrams for the new console'(see rig.2).
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Manual scram buttons this pushbutton on the reactor console allows the operator to scram the reactor manually.
External scram: this input allows an external signal to control two of the relay switches.
NP(P)-1000 sc Oms: these scrams occur when the measured core power exceeds the setpoints.
Loss of high voltages a loss of high voltage in eithtt NP(P)-1000 instrument will cause this scram.
Key switch: turning the console key switch off will scram the reactor.
Periodi power increasing on a period shorter than the setpoint will cause this scram. (Not a required scram.)
CSC watchdogs a watchdog timer in the CSC will scram the reactor if it is not frequently reset by a software routine executing in the computer.
This prevents continued i
operation if the CSC computer " hangs up."
DAC watchdogs a watchdog timer in the DAC will scram the reactor if it is not frequently reset by a software routine executing in the computer.
This prevents continued
. operation'if the DAC computer " hangs up".
SINGLE FAILURE ANALYSIS Single failure criteria analysis is to be performed in accordance with ANSI /ANS STD 15.15-1978 for all non-redundant reactor safety systems.
This analysis was performed on the General Atomics (GA)' console for the Armed Forces Radiobiology ResearchLInstitute (AFRRI)..The analysis showed that, except for the console key switch (which does not perform a safety function, but a security function) the-Hean Time Between Failure of-any single element of the new scram systen varies-from-23 to
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125 years.- This analysis is conservative for the GSTR facility i
sincefthe AFRRI f acility has a more complex i;afety system.
A failure analysis was-also performed by the University of Texas and GA.
This' analysis examined the probability of the reactor safety system failing to' perform its intended function.
This would occur only with two'or more simultaneous failures of system' components.
The analysis showed a probability of failure of 2x10-15 failures / hour and a mean time between failuree of 5x106 years.
OPERATIONAL DESCRIPTIONS.
HODES OF OPERATION There.are-four control system operating modes:
- manual, automatic, square wave, and pulse.
Manual and automatic modes are used for normal steady state operations.
Square wave and pulse modes produce power history shapes similar to their names.
Manual and automatic reactor operating modes are used for bteady-state operation from source 'evel to 100% power.
These two modes are used for manual reactor start up, changes in power level, and normal steady state adjustments.
Square wave operation allows reactor power to be raised quickly to the i
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desired power level and then stabilized, pulse mode operation generates very high power levels for very short periods of time.
Automatic operation can be entered by switching from the manual mode through depressing the Auto mode button on the console.. This action initiates control of the regulating rod position by the CSC computer.
The regulating rod is positioned in response to the power level and the reactor period.
The reactor power is compared with the demand level and the control rod is moved to bring the power level back to the demand level.
The reactor period input is used to prevent rapid power transients from the automatic controller.
Square wave operation is started by first bringing the reactor critical at a low power, leaving the transient rod at its full down position.
The transient rod is then fired up to a predetermined position,. placing the reactor on a very short positive period.
When reactor power reaches the demand power, the CSC computer controls the position of the regulating rod to maintain power at the demand level, pulsing operation consists of achieving low power criticality in the manual mode with the transient rod at its full down position.
The pulse mode button is then depressed and the transient rod is fired up to a predetermined position, placing the reactor in-a prompt critical condition, pulsing data are collected by the CSC computer and the reactor is automatically scrammed about 0.5 second after firing the transient rod.
The pulse data are stored on the CRC computer disk for future refe*.nce.
Normal rod. control is accomplished through a set of push buttons on the. control console.
The; top row of buttons is used to interrupt' magnet current to individual rod drives.
The middle row of buttons is used to give the "Up" command to the control. rod drives via the CSC and'DAC.,
Depressing a middle button will make a control rod move up unless some control rod interlock'is present to-prevent that notion.
The bottom row of buttons-is used to give the "DOWN" command to the control rod drives via the CSC and DAC.
Depressing a bottom button will make aLeontrol rod move down unless the rod is already fully down.
ROD DHIVES only.one rod drive will be changed when the new control-system in installed.
The existing regulating rod drive is
- incompatible with the new control system and will be replaced with a stepping-motor drive.
The new stepping motor drive is very similar to the other standard drives except that a stepping motor is used-instead of the standard DC motor.
The stepping motor drives a pinion gear and a 10-turn potentiometer via a
.i chain and pulley gear mechanism.
The pinion gear engages a rack attached to the magnet draw tube.
An electromagnet, attached to the end of the draw tube, engages an armature on the connecting rod.
Hovement of this rod results in movement of the control rod that is attached to its lower end, e
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Stepping motors operate on phase-switched DC power.
Each step of the motor results in a 1.8 degree rotation of the shaft.
Since current is continuously flowing in the motor windings, a high holding -torque is maintained even when the rotor is-not turning.
The torque vs speed characteristic of the stepping motor l
. depends greatly on the drive circuitry used.
A Translator Hodule-was selected to optimize the torque characteristic for the1 specific' motor frame size.
This produces the optimum torque at the operating speed of the control rod drive.
ROD INTERLOCKS
' The digital control system contains several. interlocks to
- prevent upward control rod motion during certain conditions.
..This prevents'further positive reactivity from being inserted into the core during. conditions-when that might be unsafe.,
The l
rod withdrawal-interlocks required in the control system are as follows:
i le No control rod withdrawal with less than two neutron-induced counts per secon on the startup channel 4
(steady state mode).
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'No simultaneous manual withdrawal of two control rods, j
. including the pulse rod (steady. state mode).
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No simultaneous manual withdrawal of two. control rods, excluding-the pulse rod (square ~ wave mode).
4.
No pulse initiation above 1 kW (pulse mode).
5.
No' application'of air pressure to the pulse rod. drive mechanism unless the cylinder is fully inserted (steady state mode).-
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'No-withdrawal of any centrol rod except the pulse rod 1
3, (pulse mode),
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1 COMPARISON OF OLD 5ND NEW CONTROL SYSTEMS
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-SAFETY SYSTEMS The-reactor technical specifications require certain minimum b
Ereactor: safety' systems.
'Both the new and old consoles meet these minimum requirements with the-following' systems -
' OLD NEW Linear power scram Safety Channel 1 scram; e
- Percent power scram Safety Channel 2 scram Hanual scram button Manual scram button
. Pulse' timer scram-Pulse tit:er scram
.Both systems;will also scraa the reactor if neutron detector
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high voltage is. lost, console power is lost, or the key-switch h
is turned off.
Individual rods can also be scrammed at both-O consoles.
In addition, the new system has hardwired scrams for timeout of the DAC and CSC watchdog circuits, and a digital scram in case of_an excessively short period.
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I The new digital control system has been designed to be better than the old, analog system.
This has been accomplished by hard-wiring all required scrams directly from the analog instruments in a redundant, fail-safe configuration.
These analog safety systems are completely independent of the system computers.
This means that even if both system computers were to fail, the' scrams would not be prevented from functioning.
i The~ watchdog circuite on the DAC and CSC computers monitor l
operation ointhe computers to initiate a scram upon failure of
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either computer.
The watchdog circuit contain timers that must be reset frequently or the subsequent timeout will trip a relay that in hardwired to cause a reactor scram.
The new control system contains safety system features that provide equal or better reactor safety than the present GSTR control system.
OPERATOR INFORMATION~
.The new control system provides significantly more
-information to the reactor operator than the old control system.
Examples of this ares coolant conductivity, area
' radiation levels, *2Ar release rate, continuous air monitor reading, cooling. water data, and current-pulse number.
In:
addition,~the CSC will print a hard copy of these data at any time.
The new system also archives operating history to' allow retrieval of reactor conditions for reconstruction and analysis of events.
' Pulsing operations are greatly enhanced. tar the detailed collection and-display of pulse data.
The'CSC acquires 5000 data points per pulse and calculates peak powe,-
peak temperature,-full width at half maximum, and pulse period.
All
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of.,thesefdata are displayed on-the console CRT!and a'hard copy U
.can tui easily produced.
Although most of the operator information is given on.the i
two CRTs, more critical information (power, period, fuel
- temperature, etc.)'is also provided on analog bar-graph displays 1
-on-the CSC. -These. analog displays do not depend on operation of either the CSC' computer or the DAC computer, thus providing
.1 backup indications.for these important parameters.
~ In conclusion, the new control system supplies'significantly more information=to'the reactor operator and provides redundant-
.i displays for the more critical parameters.
y CHANGES'TO TECHNICAL *PECIPICATIONS
- The Reacter Operations Committee for the GSTR has found that
.certain changes are required'to the Technical Specifications to allow installation and use of the new computerized control system.- The requested changes:are itemized on the following n
i pages'and. discussed thereafter, f
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L REQUESTED TECHNICAL SPECIFICATION CHANGES T
Page 8, Item E.12.e.
The linear power channel indicates the actual power level as determined by a
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thermal power measurement."
Change to read The safety channels indicate the actual L
power level as determined by a thermal y
power measurement."
This. change makes the. nomenclature for the-power measuring instruments condistent with the computerized control system.
L 7ABLE I L
HINIMUH REACTOR SAFETY SYSTEMS Current version:
Originating.
Hode in which effective Channel SetPoint SS pulse SW 1.
Linear =
110% of full power X
X 2.
Percent power 110% of. full power X
X 3 ', Scram button Manual push X
X X
4.
Preset time Less:than.or equal X*
.to 15 seconds
- !Does not scram reactor?- only drops pulse rod following a pulse.
CHANGE ~ Table:I to-reads-Originating Hode'in'which effective Channel' GetPoint SS Pulse SW 4
1.
Safety Channel 1 110% of full power X
X-
- 2. Safety' Channel 2 110% of full = power X
X 3.
Scram button-Hanual push X
X X
4.
Preset' timer:
Lessithan-or equal ~
.X
-to 15 seconds
- 5. CSC. watchdog. timer Loss of~ refresh signal.
X
-X X-6.-DAC watchdog timer. Loss of refresh' signal
.X X
X Thesetchanges make the nomenclature for the power: measuring channels consistent with the computerized-control system and add E
.the. watchdog timer scrams.
The watchdog timer scrams ensure that the reactor will scram if either of the control system. computers
" lock up".
This'is accomplished by requiring each computer to
- supply periodic refresh signals to its watchdog circuit.
If the refresh signals are interrupted for any reason, the watchdog timer will: time out and ' cause a reactor scram.
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~jj I.6 E l ft l l HINIMUM INTERLOCKS Current item 1.
Mode in which effective Action Prevented SS Pulne SW 1.-
Control rod withdrawal with less than X
.two. neutron induced counts per second on the startup channel.
4 CHANGE item 1 to readi Hode in which effective Action Prevented SS Pulse SW 1.-
Contro1Hrod withdrawal.with neutron X
level less than 10-?% power on the digital power channel.
,y This-interlock needs to be changed to have the neutron-level indicated in percent power instead of counts per second since the
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' newistartup channel does not provide a counts pertsecond indication
. for;the' operator.
This interlock. prevents a sourceless startup'of n
J, the; reactor.
When the reactor is shutdown with.the neutron source
- inLthe core, the neutron level indicated on the~ digital power channelHis typically aroundL5 x 10-7%.-
This level normally drops more.than an order of magnitude when the neutron source is removed l
.i-from the! core.
The interlock would thus prevent control rod, Jn
. withdrawal if the neutron source was not infplace in the core.-
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