ML20133G854
| ML20133G854 | |
| Person / Time | |
|---|---|
| Site: | University of Texas at Austin |
| Issue date: | 09/30/1985 |
| From: | TEXAS, UNIV. OF, AUSTIN, TX |
| To: | |
| Shared Package | |
| ML20133G852 | List: |
| References | |
| NUDOCS 8510160158 | |
| Download: ML20133G854 (55) | |
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s Appendix A Technical Specifications The University o f T e :: a s at Austin TRIGA Reactor September 1985 8510160150 851009 ADOCK 050 g 2 PDR
Technical Specifications 9/85 Table of Contents 1'O DEFINITIONS 5
1.1 Certified Operators 5
1.1.1 Class A Reactor 0.perator 5
1.1.3 Claas D Reactor Oaerator 5
1.2 Channel 5
1.2.1 Charnel Test 5
- 1. I'. 2 Channel Check 5
1.2.3 C h a n-s e l Calibratios 5
1.3 Confinement 6
1.4 Experiment 6
1.4.1 Experiment, llo v e a b l e 6
1.4.2 Experiment, Secured 6
1.4.3 Experimental Facilit ie s 6
1.5 Fuel Element, Standard 6
1.6 Fuel Element, Instrumented 6
1.7
!!o d e, Steady State 7
1.8
!!o d e, Pulse 7
1.9 Operable 7
1.10 Operating 7
1.11 Protective Action 7
1.11.1 Instrument Channel Level 7
1.11.2 Instrument Subsystem Level 7
1.11.3 Instrument System Level 7
1.11.4 Reactor Safety System Level 8
1.12 Reactivity,' Excess 8
1.13 Reactivity Limit 8
1.14 Reactor Core, Standard 8
1.15 Reactor Core, Operational 8
1.16 Reactor Operating 8
1.17 Reactor Safety System 8
1.18 Reactor Secured 9
1.19 Reactor Shutdown 9
1.20 Reference Core Condition 9
1.21 Research Reactor 9
1.22 Rod, Control 10 1.22.1 Regulating Rod 10 1.22.2 Safety Rod 10 1.'22.3 Shim Rod 10 1.22.4 Standard Rod 10 1.22.5 Transient Rod 10 1.23 Safety Limit 10 1.24 Scram Time 10 1.25 Shall, Should and !!ay 11 1.26 Shutdows Margin 11 1.27 Shutdown, Unscheduled 11 1.28 Surveillance Activities 11 1.29 Surveillance Intervals 12 1.30 Value, !!e a s u r e d 12 1.31 Value, True 12 Page 2
Technical Specifications 9/85 2.0 SAFETY LI!!ITS AND LIf!ITING SAFETY SYSTEli SETTINGS 13 2.1 SAFETY LI!!IT 13 2.2 LIltITING SAFETY SYSTEM SETTINGS 14 2.2.1 Fuel Temperature 14 2.2.2 Power Level (Non Pulse) 15 2.2.3 Reactivity Insertion (Pulse) 15 3.0 LI?!ITING CONDITIONS FOR OPERATION 16 3.1 REACTOR CORE PARA}!ETERS 16 3.1.1 Excess Reactivity 16 3.1.2 Shutdown !!a r g i n 16 3.1.3 Transient Insertions 17 3.1.4 Fuel Elements 18 3.2 REACTOR CONTROL AND SAFETY SYSTE!!
19 3.2.1 Control Assemblies 19 3.2.2 Reactor Control System 19 3.2.3 Reactor Safety System 21 3.2.4 Reactor Instrument System 22 3.3 OPERATIONAL SUPPORT SYSTE!!S 23 3.3.1 Vater Coolant Systems 23 3.3.2 Air Confinement Systems 24 3.3.3 Radiation !!onitoring Systems 25 3.4 LI!!ITATIONS ON EXPERIl!ENTS 26 3.4.1 Reactivity 26 3.4.2 Irradiations 27 3.4.3
!!a t e r i a l s 28 4.0 SURVEILLANCE REQUIRE!.!ENTS 30 4.1 REACTOR CORE PARA!!ETERS 30 4.1.1 Excess Reactivity 30 4.1.2 Shutdown !!a r g i n 30 4.1.3 Transient Insertion 31 4.1.4 Fuel Elements 31 4.2 REACTOR CONTROL AND SAFETY SY STE!!
32 4.2.1 Control Assemblies 32 4.2.2 Reactor Control System 33 4.2.3 Reactor Safety System 33 4.2.4 Reactor Instrument System 34 4.3 OPERATIONAL SUPPORT SYSTEllS 34 4.3.1 Water Coolant Systems 34 4.3.2 Air Confinement Systems 35 4.3.3 Radiation I:onitoring Systems 36 4.4 LIFIITATIONS ON EXPERIMENTS 36 4.4.1 Reactivity 36 4.4.2 Irradiations 37 4.4.3 Materials 37 Page 3
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Technical Specifications 9/85 i
5.0 DESIGN FEATURES 38 i
5.1 SITE AND FACILITY DESCRIPTION 38 5.1.1 Location 38 5.1.2 Confinement 39 5.1.3 Safety Related Systems 40 5.2 REAC't0E COOLANT SYSTEM 40 5.3 REACTOR CORE AND FUEL 41 5.2.1 Fuel Elements 41 l
5.2.2 Control Rods 42 5.2.3 Configuration 42
{
5.4 REACTOR FUEL ELEMENT STORAGE 43 I
6.0 ADMINISTRATIVE 44 l
6.1 ORGANIZATION 44 6.1.1 Structure 44 i
6.1.2 Responsibility 44 6.1.3 Staffing 45 6.1.4 Selection and Training of Personnel 45 6.2 REVIEW AND AUDIT 46 6.2.1 Composition and Qualifications 46 6.2.2 Charter and Rules 46 3
6.2.3 Review Function 46 6.2.4 Audit Function 47 1
6.3 OPERATING PROCEDURES 47 I
6.4 EXPERIMENT REVIEW AND APPROVAL 48 i
6.5 REQUIRED ACTIONS 48 l
6.5.1 Action to be Taken in Case of i
Safety Limit Violation 48 6.5.2 Action to be Taken in the i
Event of a Reportabic I
Occurrence 49 L
6.6 REPORTS 49 I
6.6.1 Operating Reports 49 f
j 6.6.2 Special Reports 50 6.7 RECORDS 52 6.7.1 Records to be Retained for the Lifetime of the Facility 52 6.7.2 Pecords to be Retained for a Period of at Least Five i
Years or the Life of the Component 52 i
6.7.3 Records to be Retained for at Least One Training Cycle 52 l
APPENDIX Startup Program 53
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Technical Specifications 9/85 1.0 DEFINITIONS 1.1 Certified Operators An individual authorized by the chartering or licensing organization to carry out the responsibilities associated with the position requiring the certification.
1.1.1 Class A Reactor Operator An individual who is certified to direct the activities of Class D reactor operators.
Such an individual is also a reactor operator and is commonly referred to as a Senior Reactor Operator.
1.1.2 Class B Reactor Operator An individual who is certified to manipulate the controls of a reactor.
Such an individual is commonly l
referred to as a Reactor Operator.
1.2 Instrumentation Channel A channel is the combination of sensor, line, amplifier, and output device which are connected for the purpose of measuring the value of a parameter.
1.2.1 Channel Test Channel test is the introduction of a signal into the channel for verification that it is operabic.
1.2.2 Channel Check Channel check is a qualitative verification of acceptable performance by observation of channel behavior.
This verification where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable.
1.2.3 Channel Calibration Channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures.
Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a channel test.
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Technical Specifications 9/85 1.3 Confinement Confinement means an enclosure on the overall facility which controls the movement of air into it and out through a controlled path.
1.4 Experiment Any operation, component, or target (excluding devices such as detectors, foils, etc.), which is designed to investigato non-routine reactor characteristics or which is intended for irradiation within the pool, on or in a beamport or irradiation facility and which is not rigidly secured to a core or shield structure so as to be part of their design.
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1.4.1 Experiment, !!o v e a b l e i
A moveable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
1.4.2 Experiment, Secured A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining force must be s'ubstantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credibic conditions.
1.4.3 Experimental Facilities Experimental facilities shall mean rotary specimen rack, pneumatic transfer tube, central thimble, beamtubes and irradiation facilities in the core or in the pool and will include the cobalt-60 facility.
1.5 Fuel Element, Standard A fuel element is a single TRIGA element of standard type.
Fuel is U-ZrH clad in stainless steel clad.
Zirconium to hydrogen ratio is nominal 1.6.
i 1.6 Fuel Element, Instrumented An instrumented fuel element is a special fuel element fabricated for temperature measurement.
The element shall have at least one thermocouple embedded in the fuel near the axial and radial midpoints.
Fage 6
Technical Specifications 9/85 1.7 Mode, Steady State Steady state mode operation shall mean operation of the reactor with the mode selector switch in the steady-state position.
1.8 Mode, Pulse Pulse mode operation shall mean any operation of the reactor with the mode selector switch in the pulse position.
1.9 Operable Operable means a component or system is capable of performing its intended function.
1.10 Operating Operating means a component or system is performing its intended function.
1.11 Protective Action Protective action is the initiation of a signal or the operation of equipment within the reactor safety system in response to a variable or condition of the reactor facility having reached a specified limit.
1.11.1 Instrument Channel Level At the protective instrument channel level, protective action is the generation and transmission of a trip signal indicating that a reactor variable has reached the specified limit.
1.11.2 Instrument Subsystem Level At the protective instrument subsystem Icvel, protective action is the generation and transmission of a trip signal indicating that a specified limit has been reached.
1.11.3 Instrument System Level At the protective instrument system level, protective action is the generation and transmission of the command signal for the safety shutdown equipment to operate.
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Technical Specifications 9/85 l
1.11.4 Reactor Safety System Level At the reactor safety system level, protective action 4
is the operation of sufficient equipment to immediately 4
shut down the reactor.
i 1.12 Reactivity, Excess Excess reactivity is that amount of reactivity that i
would exist if all the control rods were moved to the maximum reactive condition from the point where the i
reactor is exactly critical.
4 1
1 1.13 Reactivity Limits l
i The reactivity limits are those limits imposed on the l
reactor core excess reactivity.
Quantities are referenced to a reference core condition.
1.14 Reactor Core, Standard i
A standard core is an arrangenent of standard TRIGA fuel in the reactor grid plate and may include installed experiments.
1.15 Rtactor Core, Operational An operational core is a standard core for which the core parameters of excess reactivity, shutdown margin, fuel temperature, power calibration, and reattivity i
worths of control rods and experiments have been j
determined to satisfy the requirements set forth in the j
Technical Specifications.
1.16 Reactor Operating l
The reactor is operating whenever it is not secured or
}
shutdown.
]
1.17 Reactor Safety Systems r
i Reactor safety systems are those systems, including their associated input channels, which are designed to i
initiate automatic reactor protection or to provide j
information for initiation of manual protective action.
1.18 Reactor Secured i
The reactor is secured when:
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Technical Specifications 9/85 1.18.1 It contains insufficient fissile material or moderator present in the reactor, control rods or adjacent experiments, to attain criticality under optimum available conditions of moderation and reflection, or 1.18.2 a.
The minimum number of neutron absorbing control rods are fully inserted or other safety devices are in shutdown position, as required by technical specifications.
b.
The console key switch is in the off position and l
the key is removed from the lock.
c.
No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from I
the control rods, d.
No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum allowed for a single experiment or one dollar which ever is smaller.
1.19 Reactor Shutdown The reactor is shutdown if it is subcritical by at least one dollar in the reference core condition and the reactivity of all experiments is accounted for.
1.20 Reference Core Condition The condition of the core when it is at ambient temperature (cold) and the reactiviity worth of xenon is negligible
(<.30 dollars).
1.21 Research Reactor A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research, development, educational, training, or experimental purposes, and which may have provisions for the production of radioisotopes.
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Technical Specifications 9/85 1.22 Rod, Control A control rod is a device fabricated
- rom neutron absorbing material or fuel which is used to establish neutron flux changes and to compensate for routine reactivity loses.
A control rod may be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.
1.22.1 Regulating Rod A regulating rod is a control rod used to maintain an intended power level and may be varied manually or by a servo-controller.
The regulating rod shall have scram capability.
1.22.2 Safety Rod A safety rod is a control rod having an electric motor drive and scram capabilities.
1.22.3 Shim Rod A shim rod is a control rod having an electric motor drive and scram capabilities.
1.22.4 Standard Control Rod The regulating, sa fe ty and shim rods are standard control rods.
1.22.5 Transient Rod A transient rod is a control rod used to initiate a power pulse that is operated by a motor drive and/or air pressure.
The transient rod shall k,a v e scram capability.
1.23 Safety Limits Safety limits are limits on important process variables which are found to be necessary to protect reasonably the integrity of the principal barriers which guard against the uncontrolled release of radioactivity.
The principal barrier is the fuel element cladding.
1.24 Scram Time Scram time is the elapsed time between reaching a limiting safety system set point and a specified control rod movement.
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Technical Specifications 9/85 1.25 Shall, Should and May The word shall is used to denote a requirement.
The word should is used to denote a recommendation.
The word may is used to denote permission, neither a requirement nor a recommendation.
1.26 Shutdown Margin Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems starting from any permissible operating condition although the most reactive rod is in its most reactive position, and that the reactor will remain subcritical without further operator action.
1.27 Shutdown, Unscheduled An unschedule' shutdown is defined as any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operation, not including shutdowns which occur during testing or check-out operations.
1.27 Surveillance Activities Surveillance activities, (except those specifically required for safety when the reactor is shutdown) may be deferred during reactor shutdown, however they must be completed prior to reactor startup.
Surveillance activities scheduled to occur during an operating cycle which cannot be performed with the reactor operating may be deferred to the end of the cycle.
In general, two types of surveillance activities are specified, operability checks and calibrations.
Operability checks are generally specified as monthly to quarterly.
Calibrations are generally specified as annually to biennially.
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Technical Specifications 9/85 1.29 Surveillance Intervals Maximum intervals are to provide operational flexibility and not to reduce frequency.
Established frequencies shall be maintained over the long term.
Allowable surveillance intervals shall not exceed the following:
1.29.1 5 years l
(interval not to exceed 6 years).
1.29.2 2 years (interval not to exceed 2-1/2 years).
1.29.3 Annual l
(interval not to exceed 15 months).
1.29.4 Semiannual I
(interval not to exceed 7-1/2 months).
1.29.5 Quarterly g
1 (interval not to exceed 4 months).
1.29.6 Monthly j
(interval not to exceed (: weeks).
1.29.7 Weekly l
(interval not to exceed 10 days).
1.29.8 Daily (must be done during the calendar day).
1.30 Value, Measured The measured value is the value of a parameter as it appears on the output of a channel.
1.31 Value, True The true value is the actual value of a parameter.
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Technical Specifications 9/85 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit Applicability This specification applies to the temperature of the reactor fuel in a standard TRIGA fuel element.
Objective The objective is to define the maximum temperature that can be permitted with confidence that no damage to the fuel element cladding will result.
Specification (s)
The maximum temperature in a standard TRIGA fuel element shall not exceed 11500C for fuel element clad temperatures less than SOC *C and shall not exceed 950*C for fuel element clad temperatures greater than 500*C.
Temperatures apply to any condition of operation.
Basis The important parameter for a TRIGA reactor is the fuci element temperature.
This parameter is well suited as a singic specification since it can be measured directly.
A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit.
The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the fuel-moderator.
Hydrogen pressure is the most significant component.
The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium in the alloy.
The safety limit for the standard TRIGA fuel is based on calculations and experimental evidence.
The results indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1150*C and the fuel cladding does not exceed 500*C.
For conditions that might cause the clad temperatures to exceed 500*C the safety limit of the fuel should be set at 950*C.
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Technical Specifications 9/85 2.2 Limitinc Safety Systen Settines 2.2.1 Fuel Temperature Applicability This specification applies to the protective action for the reactor fuel element temperature.
Objective The objective is to prevent the fuel element temperature safety limit from being reached.
Specification (s)
The limiting safety system setting shall be 550*C as measured in an instrumented fuel element.
The instrumented element shall be located in the D ring of the reactor core configuration.
Dasis For non pulse operation of the reactor, the limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded.
A setting of 550*C provides a safety margin at the point of measurement of at least 400*C for standard TRIGA fuel elements in any condition of operation.
A part of the safety margin is used to account 4
for the difference between the true and measured temperatures resulting from the actual location of the thermocouple.
If the thermocouple element is located in the hottest position in the core, the difference between the r
true and measured temperatures will be only a few degrees since the thermocouple junction is near the center and the mid plane of the fuel element.
For pulse operation of the reactor, the same limiting safety system setting will apply,
!!o w e v e r, the temperature channel will have no effect on limiting the peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds).
In this mode, however, the temperature trip will act to limit the energy release after the pulse if the transient rod should not reinsert and the fuel temperature continues to increase.
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Technical Specifications 9/85 2.2.2 Power Level (Non Pulse)
Applicability This specification applies to the protective action for the reactor during non pulse operation.
Objective The objective is to prevent the fuel element temperature safety limit from being reached.
Specification (s)
The maximum operating power level for the non pulse operation of the reactor shall be 1100 kilowatts.
Basis Thermal and hydraulic calculations indicate that standard TRIGA fuel elements may be safely operated at power levels in excess of 1500 kilowatts with natural convection cooling.
Conservative estimates indicate that a departure from nucleate boiling ratio of approximately two will occur at about 1900 kilowatts.
2.2.3 Reactivity Insertion (Pulse)
Applicability This specification applies to the reactivity insertion for the reactor during pulse operation.
Objective The objective is to prevent the fuel element temperature safety limit from being reached.
Specification (s)
The maximum transient reactivity insertion for the pulse operation of the reactor shall be 2.2% Ak/k.
Basis Calculations indicate that standard TRIGA fuel elements may be safely operated at transient conditions in excess of 2.2%
Ak/k with ambient cooling conditions.
Conservative estimates indicate that a substantial safety margin exists for the rise of peak fuel temperature with reactivity insertions as large as 2.8% Ak/k.
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4 Technical Specifications 9/85 3.0 LIMITING CONDITIONS FOR OPERATION d
3.1 Reactor Core Parameters 3.1.1 Excess Reactivity Applicability t
j This specification applies to the reactivity condition of the reactor core in terms of the availabic excess above the cold renon free, critical condition.
Objective The objective is to prevent the fuel element temperature safety limit from being reached by limiting the potential reactivity availabic in the reactor for any condition of operation.
Specification (s) l Maximum excess reactivity shall be 4.9% Ak/k.
f Basis Maximum excess core reactivity is sufficient to provide the core rated power, renon compensation and reactivity for j
shutdown.
Analysis of the reactor core demonstrates that no single component represents sufficient potential reactivity to reach the fuel element temperature safety limit during any condition of operation.
l 3.1.2 Shutdown Margin i
Applicability i
This specification applies to the reactivity margin by which j
the reactor core will be considered shutdown when the reactor is not operating.
Objective The objective is to assure that the reactor can be shut down safely by a margin that is sufficient to compensate for the failure of a control rod or the movement of an experiment.
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Technical Specifications 9/85 Specification (s)
The reactor shall not be operated unless the shutdown margin provided by control rods is greater than 0.2% Ak/k with:
l a.
The reactor in the reference core condition.
l b.
The most reactive control rod fully withdrawn.
l c.
The highest worth roveable experiment in its most reactive state.
Basis The value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the highest worth control rod should remain in the fully withdrawn position and an unsecured experiment is in a high reactivity state.
3.1.3 Transient Insertions Applicability This specification applies to the total potential worth of the transient rod and the allowable reactivity insertion for reactor pulse operation.
4 Objective The objective is to limit the reactivity available for pulse insertion to a value that will not cause the fuel temperature safety limit to be exceeded.
Specification (s)
Total worth of the transient rod shall be limited to 2.8%
Ak/k, and J
g a.
The pulse reactivity insertion shall be limited by I
a mechanical block on the pulse rod.
b.
The reactivity for pulse operation shall be initiated from power levels less than 1 kilowatt.
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Technical Specifications 9/85 Basis Calculations demonstrate that the total insertion of all the transient rod worth will net exceed the fuel temperature safety limit.
For a 2.8% Ak/k pulse a safety margin would exist between the fuel element safety limit and the rise of peak fuel temperature above an assumed ambient pool temperature of 50*C.
Experiments with pulsed operation of TRIGA reactors by the manufacturer indicate that insertions up to 3.5% Ak/k have not exceeded the fuel temperature safety limit.
3.1.4 Fuel Elements Applicability
{
This specification applies to the measurement parameters for the fuel elements.
1 Objective The objective is to verify the physical condition of the fuel element cladding.
Specification (s)
The reactor shall not be operated with damaged fuel.
A fuel element shall be considered damaged and must be removed from the core if:
a.
In measuring the elongation, the length exceeds I
the original length by 2.54 mm (1/10 inch).
b.
In measuring the transverse bend, the bend exceeds the original bend by 1.5875 mm (1/16 inch).
c.
A clad defect exists as indicated by release of fission products.
Basis The elongation limit has been specified to assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow.
The limit of transverse bend has been shown to result in no difficulty in disassembling the reactor core.
Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching.
Experience with TRIGA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects.
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Technical Specifications 9/85 3.2 Reactor control and Safety System 3.2.1 Control Assemblics Applicability This specification applies to the function of the control rods.
Objective The objective is to determine that the control rods are operable by specification of apparent physical conditions, the scram times for scrammabic control rods and reactivity insertion rates for standard control rods.
Specification (s)
The reactor shall not be operated unless the control rods are operable, and a.
Control rods shall not be operable if damage is apparent to the rod or drive assemblies, b.
The scram time measured from the instant a simulated signal reaches the value of a limiting safety system setting to the instant that the slowest scrammable control rod reaches its fully inserted position shall not exceed 1 second.
c.
}!aximum reactivity insertion rate of a standard control rod shall be less than 0.2% Ak/k per second.
Basis The apparent condition of the control rod assemblies will provide assurance that the rods will continue to perform reliably and as designed.
The specification for rod scram time assures that the reactor will shut down promptly when a scram signal is initiated.
The specification for red reactivity insertion rates assures that the reactor will start up contro11 ably when rods are withdrawn.
Analysis has indicated that for the range of transients anticipated for a TRIGA reactor the specified scram time and insertion rate is adequate to assure the safety of the reactor.
3.2.2 Reactor Control System Applicability These specifications apply to logic of the reactor control system.
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Technical Specifications 9/85 Objective The objective is to determine the minimum control system interlocks operable for operation of the reactor.
Specification (s)
The reactor shall not be operable unless the minimum safety interlocks are operable.
The following control system 1
safety interlocks shall be operable:
Interlocks Number Effective Mode Rod Drive Control Ooerable Function Non-Pulse Pulse a.
All Control Rods 1
prevent for I
Startup Withdrawal less than 2 counts per sec b.
All Control Rods 1
prevent for I
Simultaneous Withdrawal two or more rods c.
Pulse Reactivity I
set rate for I
X Insertion transient rod d.
Standard Control Rod 1
prevent for X
Withdrawal each control rod e.
Transient Rod 1
prevent for X
Withdrawal more than a period of 15 secs Basis Interlocks are specified to prevent function of the control 1
rod drives unless certain specific conditions exist.
The interlock to prevent startup of the reactor at power levels less than 2 neutron cps, which corresponds to approximately i
4 milliwatts, assures that sufficient neutrons are availabic for controlled reactor startup.
Simultaneous withdrawal of more than one control rod is prevented by an interlock to limit the maximum positive reactivity insertion rate available for s tek iy s ta te operation.
Several interlocks applied to the transient rod determine the proper rod operation during pulse operation and protect against inadvertent pulse operation.
The interlock to prevent withdrawal of the motor driven rods in the pulse mode is designed to prevent changing the critical state of the reactor prior to the pulse.
A preset timer insures that the transient rod will not remain in the pulse position for an extended time after the pulse.
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Technical Specifications 9/85 3.2.3 Reactor Safety System Applicability These specifications apply to operation of the reactor safety system.
Objective The objective is to determine the minimum safety system scrams operable for the operation of the reactor.
Specification (s)
The reactor shall not be operable unless the minimum safety channels are operable.
The following control rod scram safety channels shall be operable.
Number Effective Mode Safety Channel Overable Function Non-Pulse Pulse a.
Fuel Temperature 1
Scram at 550*C I
X b.
Linear Power Level 1
Scram at 110%
X c.
Percent Power Level 1
Scram at 110%
X d.
Peak Pulse Power 1
Scram at 110%
X 4
e.
High Voltage 1
Scram on loss X
X f.
Magnet Current 1
Scram on loss X
X g.
Hanual Scram 1
Scram on demand I X
Console Button Basis Manual operation of the reactor safety system is considered part of the protective action of the reactor safety system.
Signals for control rod insertion and reactor shutdown provide scrams on excessive fuel temperature and power level that is short of the fuel element temperature safety limit.
Operation without adequate control and safety system power supplies is prevented by scrams on neutron detector high voltage and control rod maEnet current.
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Technical Specifications 9/85 3.2.4 Reactor Instrument System Applicability These specifications apply to measurements of reactor operating parameters.
Objective The objective is to determine the minimum instrument system channels to be operable for continued operation of the reactor.
Specification (s)
A minimum configuration of measuring channels shall be operable.
The following minimum reactor parameter measuring channels shall be operable:
Number Effective !!o d e
!!easurine Channel Overable Non-Pulse Pulse a.
Fuel Temperature 1
X X
b.
Linear Power Level 1
X c.
Percent Power Level 1
X d.
Peak Pulse Power 1
X Basis The minimum measuring channels are sufficient to provide signals for automatic safety system operation.
Signals from the measuring system provide information to the control and safety system for a protective action.
)!casurements of the same or different parameters provide redundancy.
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Technical Specifications 9/85 3.3 operational Sunnort Systems 3.3.1 Water Coolant Systems Applicability This specification applies to the operating conditions for the reactor pool and coolant water systems.
Objective The objective is to assure that adequate conditions are maintained to provido shielding of the reactor radiation, protection against corrosion of the reactor components, cooling of the reactor fuel, and prevent leakage from the primary coolant.
Specification (s)
Corrective action shall be taken or the reactor shut down if the following reactor coolant water conditions are observed:
l a.
The bulk pool water temperature exceeds 488C.
l b.
The water depth is less than 6.5 meters measured from the pool bottom to the pool water surface.
l c.
The water conductivity is greater than 5.0 pmho/cm averaged for measurement periods of one month, d.
The pressure difference during heat exchanger operation is less than 7 kPa (1 psig) measured between the chilled water outlet pressure and the pool water inlet pressure to the heat exchanger.
Basis The specifications for conditions of the pool water coolant system provide controls that are to control the radiation exposures and radioactive releases associated with the reactor fission product inventory.
a.
The bulk water temperature constraint assures that sufficient core cooling exists under all anticipated operating conditions and protects the resin of the water purification system from deterioration.
b.
A pool water depth of 6.5 meters is sufficient to provide more than 5.25 meters of water above the reactor core so that radiation levels above the reactor pool are at reasonable levels.
Page 23 2
Technical Specifications 9/85 c.
Average measurements of pool coolant water conductivity of 5.0 pmho/cm assure that water purity is maintained to control the effects of corrosion and activation of coolant water impurities.
d.
A pressure difference at the heat exchanger chilled water outlet and the pool water inlet of 7 kpa will be sufficient to prevent loss of pool water from the primary reactor coolant system to the secondary chilled water system in the event of a leak in the heat exchanger.
3.3.2 Air Confinement Systems Applicability i
This specification applies to the air ventilation conditions in the reactor area during reactor operation.
Objective The objective is to control the release of air in the reactor area or experimental facilities.
Specification (s)
The reactor shall not be operated unless minimum conditions for air confinement are functional.
The following minimum conditions shall exist:
a.
Equipment shall be operable to isolate the reactor area by closure of all access doors and closure of ventilation supply and exhaust dampers.
b.
A radioactivity level signal shall initiate confinement of the reactor area ventilation and an emergency air purge shall be operable with high efficiency particulate adsorption filters.
c.
The annual average release of argon-g1 from ghe facility shall be limited to 2x 10 pCi/cm Basis The specifications for exhaust ventilation and isolation of the reactor bay provide control for radioactive releases by both routine and non routine operating conditions.
l Page 24
Technical Specifications 9/85 3.3.3 Radiation Monitoring Systems l
Applicability j
This specification applies to the radiation monitoring l
conditions in the reactor area during reactor operation.
Objective The objective is to monitor the radiation and radioactivity conditions in the area of the reactor for indication of a radioactive release.
Specification (s)
The reactor shall not be operated unless minimum conditions for radiation measurement are functional.
The following minimum conditions shall exist:
a.
A continuous air monitor (particulate) shall be operabic with readout and audible alarm.
b.
An area radiation monitor (gamma) shall be operabic with readout and audible alarm.
c.
A portable radiation monitoring device shall be available.
The portable radiation monitor may be i
substituted for a monitor that is not operable for i
periods not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Basis The radiation monitors provide information to operating i
personnel of impending or existing hazards from radiation so that there will be sufficient time to evacuate the facility or take the necessary steps to control the exposure of 1
)
personnel and release of radioactivity.
Page 25
~
Technical Specifications 9/85 3.4 Limitations on Experiments 3.4.1 Reactivity Applicability This specification applies to the reactivity associated with experiments located in the reactor core.
Objective I
The objective is control the amount of reactivity associated with experiments to values that will not endanger the reactor safety limit.
Specification (s)
The reactor shall not be operated unless the following conditions governing experiment reactivity exist:
i a.
A moveable experiment shall have a reactivity worth less than 1.00 dollar.
l b.
The reactivity worth of any single experiment shall be less than 2.50 dollars.
I c.
The total reactivity worth of in-core experiments shall not exceed 3.00 dollars. including the potential reactivity which might result from malfunction. flooding, voiding, or removal and insertion of the experiments.
Basis
- a. The worth of single moveable experiment is limited so that sudden removal movement of the experiment will not cause prompt criticality.
Worth of a single unsecured experiment will not cause a reactivity insertion that would exceed the core temperature safety limit.
1
- b. The maximum worth of a single experiment is limited so that the fuel element temperature safety limit will not be exceeded by removal of the experiments.
Since experiments of such worth must be secured in place, removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent excessive power levels from being attained.
I i
- c. The maximum worth of all experiments is limited so that removal of the total worth of all experiments will not exceed the fuel element temperature safety limit.
)
l page 26
Technical Specifications 9/85 3.4.2 Irradiations Applicability This specification applies to irradiations performed in the installed experimental facilities contained in the reactor pool. Irradiations are a subclass of experiments that fall within the specifications hereinafter stated in this section.
Objective The objective is to prevent damage to the reactor, excessive release of radioactive materials, or excessive personnel radiation exposure during the performance of an irradiation.
Specification (s)
A device or material shall not be irradiated in an irradiation facility under the classification of an irradiation unless the following conditions exist:
a.
The irradiation meets all the specifications of Section 3.4 for an experiment.
l b.
The device or material is encapsulated in a suitable container.
c.
The reactivity worth of the device or material is l
0.25 dollars or less, otherwise it shall be classed as an experiment.
d.
The expected radiation field produced by the device or sample upon removal from the reactor is not more than 1 rem /hr at one meter after 10 min.,
I 1
otherwise it shall be classed as an experiment.
i c.
The device or material does not remain in the reactor for a period of over 15 days, otherwise it shall be classed as an experiment.
Basis This specification is intended to provide assurance that the special class of experiments called irradiations will be performed in a safe manner.
Page 27 I
Technical Specifications 9/S5 l
r I
3.4.3 Materials Applicability These specifications apply to experiments installed in the reactor and its experimental facilities.
Objective The objective is to prevent the release of radioactive material in the event of an experiment failure, either by failure of the experiment or subsequent damage to the reactor components.
Specification (s)
The reactor shall not be operated unless the following conditions governing experiment materials exist:
a.
Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials, and liquid fissionable materials shall be doubly encapsulated.
b.
If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, removal and physical inspection shall be performed to determine the consequences and need fer corrective action. The results of the inspectica and any corrective action taken shall be reviewed by the Director, or his designated alternate, and determined to be satisfactory before operation of the reactor is resumed.
i c.
Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 millicuries.
d.
Explosive materials, such as gunpowder, nitroglycerin, trinitrotoluene, or pentaerythritol tetranitrate in quantities greater than 25 milligrams shall not be irradiated in the reactor or experimental facilities. Explosive materials in quantities less than 25 milligrams may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.
Page 28
Technical Specifications 9/85 e.
Experiment materials, except fuel materials, which could off gas, sublime, volatilize, or produce acrosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor, (3) possibic accident conditions in the experiment shall be limited in activity such that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the occupational limits for maximum permissible concentration, f.
In calculations pursuant to c.
above, the following assumptions shall be used: (1) If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.
(2) If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.25 micron particles, at least 10% of these vapors can escape. (3) For materials whose boiling point is above 55'C and where vapors formed by boiling this material can escape only through an undisturbed column of water above the core, at least 10% of these vapors can escape.
Basis a.
Double encapsulation is required to lessen the experimental hazards of some types of materials.
b.
Operation of the reactor with the reactor fuel or structure damaged is prohibited to avoid release of fission products.
- c. The 1.5-curic limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, the cxposure dose at the exclusion area boundary is limited,
- d. This specification is intended to prevent damage to reactor components resulting from failure of an experiment involving explosive materials, e.
This specification is intended to reduce the likelihood that airborne activities in excess of the maximum allowable limits will be released to the atmosphere outside the facility boundary,
- f. This specification provides guidance for the calculation of conditions in part (e).
Page 29
Technical Specifications 9/85 4.0 SURVEILLANCE REQUIREMENTS 4.1 Reactor Core Parameters 4.1.1 Excess Reactivity Applicability This specification applies to the measurement of reactor excess reactivity.
Objective The objective is to periodically deter ine the changes in core excess reactivity available for power generation.
Specification (s)
Excess reactivity shall be determined annually or after significant reactor core or control rod changes.
Basis Annual determination of excess reactivity and measurements after reactor core or control rod changes are sufficient to monitor significant changes in the core excess reactivity.
4 1.2 Shutdown Margin Applicability This specification applies to the measurement of reactor shutdown margin.
Objective The objective is to periodically deterrine the core shutdown reactivity available for reactor shutdown.
Specification (s)
Excess reactivity and shutdown margin shall be determined annually or after significant core or control rod changes.
Basis Annual determination of shutdown margin and measurements after reactor core or control rod changes are sufficient to monitor significant changes in the core shutdown margin.
Page 30
Technical Specifications 9/85 s
4.1.3 Transient Insertion Applicability This specification applies to survelliance of the transient rod mechanism and to observation of the reactor transient response.
Objective The objective is to assure the function of the transient rod drive and to compare the reactor pulse insertion parameters.
Specification (s)
The transient rod drive cylinder and associated air supply shall be inspected, cleaned, and lubricated snnually, and a.
A reactor pulse shall be performed annually to compare fuel temperature, peak power and energy measurements with those of previous pulses.
b.
The reactor shall not be pulsed routinely until such comparative measurements have been made.
Basis Annual inspections of the pulse rod drive system should be sufficient to detect and correct changes in the system that could impair operability.
The annual measurement of pulse parameters provides data to monitor changes in the reactor core transient characteristics.
4.1.4 Fuel Elements Applicability This specification applies to the inspection requirements for the fuel elements.
Objective The objective is to inspect the physical condition of the fuel elemeat cladding.
Specification (s)
The reactor shall not be operated with damaged fuel and a visual inspection of the fuel elements shall be made at biennial intervals.
Page 31
Technical Specifications 9/85 Basis The frequency of inspection and measurement schedule is based on the parameters most likely to affect the fuel cladding of a pulsing reactor operated at moderate pulsing levels and utilizing fuel elements whose charac te ris t ic s are well known.
4.2 Reactor Control and Safety System 4.2.1 Control Assemblies Applicability This specification applies to the surveillance of the control rods.
Objective The objective is to inspect the physical condition of the reactor control rods and establish the operabic condition of the rod by periodic measurement of the scram times and insertion rates.
t Specification (s)
Control rod worths shall be determined annually of after significant core or control rod changes, and a.
Each control rod shall be inspected at biennial intervals by visual observation, b.
The scram time of a scrammable control rod shall be measured annually.
c.
The reactivity insertion rate of a standard control rod shall be measured annually.
Basis Annual determination of control rod worths or measurements after significant core changes provide information about chanses in reactor total reactivity and individual rod worths.
The frequency of inspection for the control rods will provide periodic verification of the condition of the control rod assemblics.
Verification will be by measurement of fueled sections and visual observation of absorber sections plus examination of linkages and drives.
The specification intervals for scram time and insertion rate assure operable performance of the rods.
Deviations that are significant from acceptable standards will be promptly corrected.
Page 32
Technical Specifications 9/85 4.2.2 Reactor Control System Applicability This specification applies to the tests of the logic of the reactor control system.
)
Objective The objective is to specify intervals for test, check or calibration of the minimum control system interlocks.
t Specification (s)
The minimum safety interlocks shall be tested at semiannual intervals.
I Dasis The periodic test of the interlock logic at semiannual intervals provides adequate information that the function of the control system interlocks are functional.
Checks or calibrations of the control system logic are not considered applicable functions.
4.2.3 Reactor Safety System Applicability This specification applies to tests of the function of the reactor safety system.
Objective The objective is to specify intervals for test, check or calibration of the minimum safety system scrams.
Specification (s)
The minimum safety channels shall be calibrated annually and tested prior to each days operation or prior to each extended period of operation.
Basis The periodic calibration at annual intervals provides adequate information that the setpoints of the safety system scrams are functional.
Tests of the safety system prior to each planned operation assure that each intended scram i
function is operable.
Page 33
Technical Specifications 9/85 4.2.4 Reactor Instrument System Applicability These specifications apply to calibrations, checks, and tests of reactor measurement channels.
Objective The objective is to specify intervals for test, check or calibration of the minimum instrument channels.
Specification (s)
The minimum instrument channels shall be calibrated annually.
Calibration of the power measuring channels shall be by the calorimetric method.
A check and test of each channel shall be made prior to each days operation or prior to each extended period of operation.
Basis Annual ca:ibration of instrument channels are scheduled to allow adjustments for changes in reactor and instrumentation parameters.
Checks and tests are applied prior to system operation to verify function of the system.
4.3 Operational Sunnort Systems 4.3.1 Water Coolant Systems Applicability This specification applies to surveillance conditions for the reactor pool and coolant water systems.
Objective The objective is to maintain the reactor coolant conditions within acceptable specifications.
Specification (s)
The following measurements shall monitor the reactor coolant conditions:
a.
The water temperature sensor shall be checked annually, tested monthly and monitored continuously during reactor operation.
b.
The water depth sensor shall be checked annually, tested monthly and monitored continuously during operation of the reactor.
Page 34
Technical Specifications 9/85 l
1 c.
The water conductivity sensor shall be checked annually and pool water conductivity measured monthly.
d.
The pressure difference sensor shall be tested prior to each days operation or prior to each extended period of operation of the heat exchanger and monitored continuously during operation.
Basis Conditions for the reactor coolant are monitored by visual observation of measurements or automatic action of sensors.
Periodic checks and tests of measurement devices for the reactor coolant systen parameters assure that the coolant system will perform its intended function.
4.3.2 Air Confinement Systems Applicability This specification applies to surveillance conditions for the air ventilation in the reactor area.
Objective l
The objective is to demonstrate the function of confinement and release of air from the reactor bay.
Specification (s)
The following actions shall demonstrate the air confinement conditions:
a.
Annual examination of door seals and isolation
~
daupers.
b.
!!o n t h l y functional tests of air confinement 4
isolation, c.
Calibration of argon-41 measurements shall be made annually and measurements or calculations performed quarterly.
I Dasis Periodic evaluations of air confinement criteria are determined by exanination, test, and calibration of the appropriate ventilation functions, the reactor bay provide control for radioactive releasea by both routine and non routine operating conditions.
Page 35
Technical Specifications 9/85 4.3.3 Radiation Monitoring Systems Applicability This specification applies to the surveillance conditions of the radiation monitoring channels.
Objective The objective is to assure the radiation monitors are functional.
Specification (s)
The minimum radiation monitors specified to be operable during reactor operation shall be a.
Calibrated at semiannual intervals.
b.
Checked at monthly intervals for fixed monitor.
c.
Checked daily for portable monitor.
Basis Periodic calibrations and frequent checks are specified to maintain reliable performance of the radiation monitoring instruments.
4 4.4 Limitations on Experiments 4.4.1 Reactivity i
Applicability This specification applies to surveillance of the reactivity of experiments.
Objective The objective is assure the reactivity of an experiment does not exceed the allowable specification.
Specification (s)
I The reactivity of an experiment shall be measured before an experiment is considered functional.
Basis i
The measured reactivity or determination that the reactivity is not significant will provide data that configuration of the experiment or experiments is allowable.
l 4
I Page 36
Technical Specifications 9/85 4.4.2 Irradiations Applicability This specification applies to the surveillance requirements for reactor irradiations.
i Objective The objective is to provide a record of experiments inserted in the reactor as irradiations.
Specification (s)
Experiments classified as irradiations shall be identified i
and a log or other record maintained while the sample is in the reactor.
Basis Experiments performed as irradiations are monitored by data on the sample location, identification and other pertinent information.
4.4.3
!!aterials Applicability This specification applies to the surveillance requirements for materials inserted into the reactor.
Objective The objective is to prevent the introduction of materials l
that could damage the reactor or its components.
Specific tion (s)
Any surveillance conditions or special requirements shall be specified as a part of the experiment approval.
Basis A careful evaluation of all experiments is performed to classify the experiment as an approved experiment.
Page 37
Technical Specifications 9/85 5.0 DESIGN FEATURES 5.1 Site and Facility Description 5.1.1 Location Applicability This specification applies tc the TRIGA reactor site location and specific facility design features.
Objective 4
The objective is to specify those features related to the Safety Analysis evaluation.
Specification (s) a.
The site location is in the northeast corner of The University of Texas at Austin Balcones Research Center.
I b.
The TRIGA reactor is installed in a designated room of a building constructed as a Nuclear Engineering Teaching Laboratory.
4 c.
The reactor core is assembled in an above ground l
shield and pool structure with horizontal and vertical access to the core, j
d.
Restricted access area of the facility shall consist of the room enclosing the reactor shield j
and pool structure, and the adjacent area for reactor control.
Basis a.
The TRIGA facility site is located in an aren controlled by The University of Texas at Austin.
b.
The room enclosing the reactor has been designed with characteristics related to the safe operation of the facility.
c.
The shield and pool structure have been designed for radiation levels of less than 1 mrem /hr at locations that are not access ports to the reactor structure.
4 l
d.
The resticted access to specific facility areas assure that proper controls are established for the safety of the public and for the security of special nuclear materials.
Page 38
. - ~
Technical Specifications 9/85 S.1.2 Confinement Applicability This specification applies to the boundary for control of air in the area of the reactor.
Objective The objective is to assure that provisions are made to control or restrict the amount of release of radioactivity into the environment.
Specification (s) a.
The reactor room shall be designed to restrict leakage and will have a minimum enclosed air volume of 4120 cubic meters.
b.
Ventilation system should provide two air changes per hour and shall isolate air in the reactor area upon detection of a limit signal related to the radiation IcVel.
c.
Emergency ventilation system shall be filtered by high efficiency particulate adsorption filters and should maintain a negative pressure compared to ambient air conditions, d.
All exhaust air from the reactor area enclosure shall be ejected vertically upward at a point above the facility roof level.
Basis a.
Calculations of the concentrations of released radionuclides within the reactor area depend on the available enclosed air volume to limit the concentrations to acceptable levels, b.
Control of the reactor area air exchange is by fan motors and isolation dampers for the supply and exhaust air which are controlled by a logic signal from a radiation sensor to provide automatic air confinement.
c.
Emergency air ventilation is filtered to control the release of particulates and a pressure difference relative to the external ambient pressure is intended to prevent leakage of air without filtration.
d.
Exhaust air during reactor operation is released at an elevated level for dispersion and is designed to provide a relative pressure difference to the external ambient pressure.
Page 39
Technical Specifications 9/85 5.1.3 Safety Related Systemn Applicability This specification applies to the requirements of any system related to reactor safety.
Objective The objective is to assure the proper function of any system related to reactor safety.
Specifications Any modifications or maintenance to the air confinement or ventilation system, the reactor shield, the pool or its penetrations, the pool coolant system, the core and its associated support structure, the rod drive mechanisms or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated.
Alternate specifications may be approved by the reactor operation committee.
A system shall not be considered operable until after it is tested successfully.
Basis This specification relates to changes in reactor systems which could affect the safety of the reactor operation.
Changes or substitutions to these systems that meet or exceed the original design srecifications are assumed to meet the presently accepted operating criteria.
Questions that may include an unreviewed safety question are referred to the reactor operation committee.
5.2 Reactor Coolant Systen Applicability This specification applies to the reactor coolant system composed of deionized water.
Objective The objective is to assure that adequate water is available for cooling and shielding during reactor operation.
Specification (s) a.
The reactor core shall be cooled by natural convection flow of water.
Page 40
Technical Specifications 9/85 4
b.
Pool water level shall be protected by holes for siphon breaks in pool water system pipe lines.
Basis a.
This specification is based on thermal and hydraulic calculations which show that a standard 85 element l
TRIGA core can operate in a safe manner at power levels up to 1.900 kW with natural convection flow of the coolant water and a departure from nucleate boiling ratio of 2.0.
I b.
Siphon breaks set the subsequent pool water level for loss of coolant without an associated water return i
caused by inadvertant pumping or accidental siphon of water from the pool.
5.3 Reactor Core and Fuel l
5.3.1 Fuel Elements Applicability This specification applies to the fuel elements used in the reactor core.
Objective The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Specification (s)
The standard TRIGA fuel element at fabrication shall have the following characteristics:
a.
Uranium content: 8.5 Wt% uranium enriched to a I
n o m i r. a 1 19.7% Uranium-235.
i l
b.
Zirconium hydride atom ratio: nominal 1.6 hydrogen to zirconium. Z rli,.
l c.
Cladding: 304 stainless steel, nominal.020 inches J
thick.
1 Basis i
The Design basis of the standard TitIGA core demonstrates that 1.5 megawatt steady or 36 megawatt-sec pulse operation presents a conservative limitation with respect to safety limits for the maximum temperature generated in the fuel.
Page 41 i
i I
i Technical Specifications 9/85 I
l l
The fuel temperatures are not expected to exceed $50*C during any condition of normal operation.
l 5.3.2 Control Rods Applicability This specification applies to the control rods used in the reactor core.
Objective l
The objective is to assure that the control rods are of such a design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Specification (s)
The safety, shim, regulating, and transient control rods shall have scram capability, and a.
Include stainless steel or aluminum clad and may be followed by air or aluminum, or for a standard rod may be followed by fuel with stainless steel
- clad, b.
Contain borated graphite, 1 C powder, or boron and 4
its compounds in solid form as a poison.
c.
The transient rod shall have an adjustable limit to allow a variation of reactivity insertions.
Easis The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite. E C powder, or 4
boron and its compounds.
These materials must be contained in a suitable clad material, such as aluminum or stainless steel, to insure mechanical stability during movement and to isolate the poison from the pool water environment.
Scram capabilitiea are provided for rapid insertion of the control l
rods which is the primary safety feature of the reactor.
l The transient control rod is designed for a reactor pulse.
l 5.2.1 Configuretion Applicability This specification applies to the configuration of fuel elements, control rods, experiments and other reactor grid plate components.
Page 42 L
i l
j Technical Specifications 9/85 i
j Objective j
The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so t
i as to provide assurance that excessive power densities will j
not be produced.
Specification (s) 1 The reactor shall be an arrangement of core single grid I
positions occupied by fuel elements, control rods, and graphite elements.
Single element positions may be occupied j
by voids. water or experiment facilities.
Special multielement positions or single element positions may be l
occupied by approved experiments.
1l Basis l
Standard TRIGA cores have been in use for years and their l
characteristics are well documented.
d 5.4 Peactor Fuel I?lement Storstg 1
Applicability This specification applies to the storage of reactor fuel at l
times when it is not in the reactor core.
l Objective 1
The objective is to assure that fuel storage will not become j
critical and will not exceed design temperatures.
I Specification (s) a.
All fuel elements shall be stored in a geometrical array where the effective multiplication is less l
than 0.8 for all conditions of moderation.
l b.
Irradiated fuel elements and fueled devices shall
(
j l
be stored in an array which will permit sufficient a
natural convection cooling by water or air such that the fuel element os fueled device temperature will not exceed design values.
l Dasis l
)
i The limits imposed by these specifications are considered sufficient to provide conservative fuel storage and assure i
safe storage.
}
I 1
i i
i Page 43 i
(
Technical Specifications 9/85 6.0 ADMINISTRATIVE CONTROLS 6.1 Oreanization 6.1.1 Structure The facility shall be under the direct control of the Director or a licensed senior operator designated to be in direct control.
The management for operation of the facility shall consist of the organizational structure established as follows:
Office, President University of Texas at Austin i
Vice President for Academic Affairs and Research I
Dean College of Engineering I
Radiation Chairman Reactor Safety Department of Operation Committee Mechanical Engineering Committeo I
8 l
I Director g
luclear Engineering
._, a reaching Latoratory I
Supervisor Responsibility Reactor Communication Operations 6.1.2 Responsibility The Director shall be responsible to the Dean of the College of Engineering and the Chairman of the Department of Hechanical Engineering for safe operation and maintenance of the reactor and its associated equipment.
The Director or Lis appointee shall review and approve all asperiments and experimental procedurea prior to their use in the reactor.
Individuals of the management organization shall be responsible for the policioa and operation of the facility, and shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to the operating license and technical specifications.
Page 44
~
E{'
Technical Specificationa 9/85 t
6.1.3 Staffing The minimum staffing when the reactor is i. o. secured shall be:
l a.
A certified operator in the control reom.
i b.
A second person in the facility area that can perform prescribed written instructions.
Unexpected absence for two hours shall requito l
immediate action to obtain an alternate person.
c.
A designated Class A operator readily available.
The available operator should be within thirty 3
l minutes of the facility and reachable by telephone.
N i
j Events requiring the direction of a Class A operator shall be a.
All fuel element or control rod relocations within the reactor core region.
i b.
Relocation of any experiment with a reactivity worth of greater than one dollar, c.
Recovery from an unscheduled shutdown will require documented verbal concurrence if the cause is unknown.
A list of reactor facility personnel by name and telephone number shall be available to the operator in the control room.
The list shall include:
a.
Management personnel.
b.
Radiation safety personnel.
c.
Other operations personnel.
6.1.4 Selection and Training of Personnel The selection, training and regullfication of operators I
shall meet or exceed the requirements of American National Standard for Selection and Training of Personnel for i
Research Reactors ANSI /ANS - 15.4.
Qualification sta I
requalification of certified operators shall be subject to an approved NRC (Nuclear Reguistory Commission) program.
I i
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i Technical Specifications 9/85 l
l 6.2 Review and Audit 6.2.1 Composition and Qualifications i
A Reactor Operation Committee shall consist of at least three (3) members appointed by the Dean of the College of Engineering that are knowledgeable in fields which relate to nuclear safety.
The University Radiological Safety Officer shall be an ex-officio member of the Reactor Committee.
The committee will perform the functions of review and audit or designate a knowledgeable person for audit functions.
6.2.2 Charter and Rules I
The operations of the Reactor Committee shall be in accordance with an established charter, including provisions for:
a.
Meeting frequency (at least once each semester).
b.
Quorums (not less than one-half the membership).
l c.
Dissemination, review, and approval of minutes.
d.
Use of subgroups.
6.2.3 Review Function i
The review function shall include facility operations i
related to reactor and radiological safety.
The following j
-items shall be reviewed:
s a.
Determinations that proposed changes in equipment, systems, tests, experiments, or procedures do not involve an unreviewed safety question, b.
'611 new procedures and major revisions thereto, aid proposed changes in reactor facility equipment or tystems having safety significance.
1 c.
All nei experiments or classes of experiments that could a f'ipc t reactivity or result in the release o f rad ioastQvity.
4 d.
Changes in to'ts&nical specifications or license.
'h teY nical specifications or license.
e.
Violations of h
f.
Operating abnormal bos or violations of procedures having s a f'2Ky significance.
V4 g.
Other reportable occurredbys.
h
,1 ss l
h.
Audit reports.
N'y
<s
'4 Page 46
Technical Specifications 9/85 6.2.4 Audit Function The audit function shall be a selected examination of operating records, logs, or other documents.
An audit will be by a person not directly responsible for the records and may include discussions with cognizant personnel or observation of operations.
The following items shall be audited and a report made to the Reactor Supervisor and Reactor Operation Committee:
a.
Conformance of facility operations with license and technical specifications at least once each calendar year.
b.
Results of actions to correct deficiencies that i
may occur in reactor facility equipment, structures, systems, or methods of operation that affect safety at least once per calendar year.
l c.
Function of the retraining and requalification program for certified operators at least once every other calendar year.
d.
The reactor facility emergency plan and physical security plan, and implementing procedures at least once every other year.
l 6.3 Operatine Procedures Written operating procedures shall be prepared reviewed and approved by the Director or a designated alternate and the Reactor Operations Committee prior to initiation of the following activities:
l a
Startup, operation, and shutdown of the reactor.
b.
Fuel loading, unloading and movement in the l
reactor.
t c.
Routine maintenance of major components of systems l
that could have an effect on reactor safety.
d.
Surveillance calibrations and tests required by the technical specifications or those that could have an effect on reactor safety.
e.
Administrative controls for operation maintenance, and the conduct of experiments or irradiations l
that could have an effect on reactor sa fe ty.
f.
Personnel radiation protection consistent with I
applicable regulations.
i Page 47
Technical Specifications 9/85 F
g.
Implementation of required plans such as the emergency plan or physical security plan.
Substantive changes to the above procedure shall be made effective after approval of the Director a designated alternate and the Reactor Operations Committee.
Minor modifications to the original procedures which do not change the original intent may be made by a Class A operator but l
the modifications must be approved by the Director or a designated alternate.
Temporary deviations from the procedures may be made by a Class A operator in order to deal with special or unusual circumstancea or conditions.
Such deviations shall be documented and reported to the Director or the designated alternate.
1 6.4 Experimen,t Review and Anoroval All new experiments or classes of experiments shall be approved by the Director a designated alternate and the Reactor Operations Committee.
1 a.
Approved experiments shall be carried out in accordance with established and approved procedures.
b.
Substantive changes to previously approved experiments shall require the same review as a new l
experiment.
c.
Minor changes to an experiment that do not significantly alter the experiment may be made by 1
a Class A operator.
i 6.5 Recuired Actions 6.$.1 Action to be Taken in Case of a Safety Limit Violation In the event of a safety limit violation, the following action shall be taken:
r i
a.
The reactor shall be shut down and reactor operation shall not be resumed until a report of the violation is prepared and authorization of the Nucle a r Ite gula t ory Commi s s ion (NRC).
b.
The safety limit violation shall be promptly reported to the Director of the facility or a designated alternate.
I c.
The safety limit violation shall be subsequently reported to the NRC.
j Page 48
!l-Technical Specifications 9/85 i
d.
A safety limit violation report shall be prepared and submitted to the Reactor Operations Committee.
l The report shall describe: (1) Applicable l
circumstances leading to the violation including, t
when known the cause and contributing factors, (2)
Effect of the violation on reactor facility components, systems. or structures and on the health and safety of the public, (3) Corrective 4
j actions taken to prevent recurrence.
j 6.5.2 Action to be Taken in the Event of an l
Occurrence that is Reportable.
1 In the event of a reportable occurrence, the following action shall be taken:
)
a.
Reactor conditions shall be returned to normal or j
the reactor shutdown.
If it is necessary to shut down the reactor to correct the occurrence,
]
operations shall not be resumed unless authorized by the Director or his designated alternate.
i b.
Occurrence shall be reported to the Director or j
his designated alternate and to licensing i
authorities as required.
i c.
Occurrence shall be reviewed by the Reactor Operacion Committee at the noxt regularly scheduled meeting.
i 6.6 Recorts All written reports shall be sent within the prescribed j
interval to the NRC, Washington D.C.
20555, Atten: Document l
Control Desk, with a copy to the Regional Administrator, l
l Region IV.
l l
j 6.6.1 Operating Reports 4
)
Routine ansual reports covering the activities of the reactor facility during the previous calendar year shall be i
j submitted within three months following the end of eas h j
prescribed year.
Each annual operating report shall include f
the following information:
1 i
j a.
A narrative summary of reactor operating expirlwnce I
including the energy produced by the reactor or the hours the reactor was critical, or both.
i j
b.
The unscheduled shutdowns including, where j
spplicable, corrective action taken to preclude j
recurrence.
,i l
c.
Tabulation of major preventive and corrective 1
maintenance operations having safety significance, i
i 4
Page 49
_ =_
i Technical Specifications 9/85
+
4 d.
Tabulation of major changes in the reactor j
facility and procedures, and tabulation of new tests or experiments, or both, that are significantly different from those performed previously, including conclusions that no unroviewed safety questions were involved.
1 e.
A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the offective control of the owner-operator i
as determined at or before the point of such I
4 release or discharge.
The summary shall include l
to the extent practicable an estimate of individual radionuclides present in the effluent.
If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended, a statement to this effect I
l is sufficient.
i f.
A summarized result of environmental surveys I
performed outside the facility.
[
g.
A summary of exposures received by facility l
personnel and visitors where such exposures are l
4 J
greater than 25% of that allowed or recommended.
j j
6.6.2 Special Reports
[
l A written report within 30 days to the NRC of.
l i
j a.
Permanent changes in the facility organization I
involving Director or Supervisor.
i.
{
}
b.
Significant changes in transient or accident l
l analysis as described in the Safety Analysis Report.
]
A report to Region IV by telephone not later than the j
following working day and confirmed in writing by telegraph j
or similar conveyance to be followed by a written report 1
within 14 days that describes the circumstances of the event of any of the follosing:
a.
Violation of fuel element temperature safety limit.
l b.
Release of radioactivity above allowable limits.
i c.
Other reportable occurrences.
I Page 50 4
Tochnical Specificotions 9/89 Other events that will be considered reportable events are listed in this section.
A return to normal operation or curtailed operation until authorized by management will occur.
(Note:
Where components or systems are provided ir addition to those required by the technical specifications, i
the failure of components or systems is not considered reportable provided that the minimum number of components or systems specified or required perform their intended reactor safety function.)
s.
Operation with actual safety-system settings for required systems less conservative than the limiting safety system settings specified in the technical specifications.
b.
Operation in violation of limiting conditions for operation established in technical specifications unless prompt remedial action is taken.
c.
A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended #4fety function unless the malfunction or conditica is discovered during maintenance tests or periods of reactor shutdowns.
d.
Abnormal and significant d o B r a d a t i r.
in reactor fuel, or cladding, or both, c o o j a r. t bcondary, or confinement boundary (excluding.sinne 1. ass) where applicable which could result in exceeding prescribed radiation exposure 1smits of personnel or environment, or both.
c.
An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.
A written report *eithin 90 days or 9 months, which over is earlier, after the initial criticality of the startup test program (see attachment), to the NRC of Characteristics upon receipt of a new facility license, of the reactor under the new conditions, describing the measured values of the operating conditions including a.
Total control reactivity worth and reactivity of g
the rod of higheet reactivity worth, b.
Minimum shutdown margin of the reactor both at l
ambient and operating temperatures.
l'a g e $1
l, Technical Specifications 9/85 i
1 c.
An evaluation of facility performance to date in i
comparison with design conditions and measured operating characteristics, and a reassessment of the safety analysis when measurements indicate i
that there may be substantial variance from prior analysis submitted with the license appilcation.
]
6.7 Records J
The records may be in the form of logs, data sheets, or r
i other suitable forms.
The required information may be contained in single or multiple records. or a combination thereof.
f 6.7.1 Records to be Retained for the Lifetime of I
the Reactor Facility:
(Note:
Applicable annual reports, if they contain all of the required information, may be used as records in this section.)
1 j
a.
Gaseous and 11guld radioactive effluents released l
l to the environs.
b.
Offsite environmental monitoring surveys required by technical specifications.
[
c.
Radiation esposure for all personnel monitored.
)
d.
Updated drawings of the reactor facility.
j 6.7.2 Records to be Retained for a Period of at f
Least Five Years or for the Life of the i
Component Involved Whichever is Shortert I
s.
Normal reactor facility operation (supporting documents such as checklists. log sheets, etc.
f shall be maintained for a period of at least one
.i year).
b.
Principal maintenance operations.
]
c.
Reportable occurrences.
d.
Surycillance activities required by technical specifications.
l l
e.
Reactor facility radiation and contamination 1
]
surveys where toquired by applicable regulations.
l f.
Experiments performed with the reactor.
4 g.
Fue1 inventories, receipts, and shipments.
l h.
Approved changes in operating procedures.
l Records of meeting and audit reports of the review and audit group.
l j
6.7.3 Records to be Retained for at Least one l
l Training Cycles j
Retraining and requalifications of certified operations personnel.
Records of the most recent complete cycle shal!
be unintained at all times the individual is employed.
l t
i Page $2
r Technical Specifications 9/85 4
ATT ACll!!EllT TO TECil!!IC AL SPECIFIC ATI0!!S Outline of THIGA Reactor Startup and Calibration Procedures The following is a brief outline of the procedurco to be followed in performing the startup and calibration of the TRIGA reactor.
By following theoc procedurco, all necessary instrument checks and nuclear calibrations will be performed.
1.
Pre-Critical Tests a.
Functional Testo of !!ochanical Equipment This includes verification of proper installation and performing operational tests on the isotopo production facility, pneunatic system, beam ports, water system, fuel olomont handling tool and other related equipment.
b.
Checkout of Instrumentation and Control System This includes a complete checkout of the control consolo by electronics personnel and ensures that all circuite are calibrated and operating satinfactorily.
Additionally, the pnoinatic and standard rod driven are checked for proper operation during rod withdrawal and acram situations.
The fission and ion chambora are shown to be noncitive and responsive to neutronn.
A piconemeter and a scalor are used as a temporary addition to the concolo instrumentation to provido more information during initial startup.
2.
Initial Loading of Fuol Elements to Criticality Using Invorco !!ultiplication Data Loading proccoda according to standard proceduron used on the many TRICA renotors in operation.
At the end of each loading atop data are recorded (the number of clomonta loaded in each atop in determined by 1/2 the number projected for criticality, or one, whichevor in greator).
The first four loading stops procond with all control rods withdrawn.
The romainder of the loading stops proccod with the shim rod incerted and the other rods withdrawn.
A plot of the number of fuel olomonta loaded versus the inverno multiplication in maintained and extrapolated to predict criticality.
The weight of uranium in each fuel clomont varion olightly, no the olomonta are positioned colectively to group the heavior clomonto at the core's contor.
All Pago 53
~
i Technical Specificctions 9/85 inner ring grid plate holes will be filled with fuel elements, control rods or other core components.
The source is re=oved temporarily upon reaching criticality to verify criticality.
3 Tests to be Perforced Upon Roaching Criticality a.
Approximate Calibration of Control Rods by Rod Drop !!ethod This is only a rough approximation used to give some indication of the rods' values.
b.
Adjustment of Excess Deactivity by Addition of Fuel Elements and Preliminary Control Rod Calibration Additional fuel is added to the core.
The outor ring of the core will initially contain the pneuuntic system terminus and graphite dummy elements.
It may also contain enough fuel elements to provide the specified excess reactivity.
A preliminary control rod calibration is mado during the loading of the excess reactivity.
c.
Final Calibration of Control Hods by the Period
!!cthod and Determination of Final Excess Henctivity These accurate calibration curves are calculated using measured periods and the in-hour curve and plotted after the final core configuration has been established.
They are used throughout the rest of the startup tests, d.
Calibration of Period !!otor Using the rod calibration curvos, a prodotermined amount of excess roactivity is inserted.
Using the in-hour curvo, the period corresponding to that insortion can be determined and the period motor adjusted accordingly.
4 Tests to be Performed at !!1ghor Powers a.
Chock of Instrument Linearity by Going to Full Power in Docado Stops Consolo instrumontation is compared with the reference piconomotor readings.
Pago 54
r Technical Specifications 9/85 b.
Power Calibration by the Method of Rato of Rise of Water Temperature Reactor heat added to the water at a constant rate raises the water temperature X degrees per hour.
This rate of rise compared with the rate of rise for a known kilowatt heat input establishes the reactor power level.
The parameter X has been determined by electric heaters in other standard TRIGA reactor tanks.
5.
Demonstration Tests at Rated Power a.
Demonstration of performanco and reliability of reactor system by operation at rated power, b.
Radiation Survey to Demonstrate Adequacy of Reactor Shielding Readings will be taken at different positions at the top of the reactor tank over the water, on the bridge, at the beam tubes, step in the shield, and various other points on the shield.
c.
Demonstration of Pulsing Performance of Reactor System The purpose of those measurements is to determine the pulso characteristics and to check the nv circuits calibrated extrapolation from full steady state powor.
A high-speed recorder, calibrated potentiometer and a picoammoter are used as auxiliary instrumentation.
The high-spcod recorder is used to give a graphic picture of the pulse, the shapo of which can be compared with pulsos from other TRIGA reactors.
The peak power shown on the high-spood recorder is compared with the peak power information displayed on the consolo recorder to assure accurate calibration.
Temperature information from thercocouples ombedded in the fuel elements can be read out on both recorders and comparod.
Pulse data are recorded.
Pago 55 i