ML20011F034

Forwards Revised Tech Specs to Replace All Previous Tech Specs Submittals

ML20011F034
Person / Time
Site:	University of Texas at Austin
Issue date:	02/12/1990
From:	Bauer T TEXAS, UNIV. OF, AUSTIN, TX
To:	Alexander Adams Office of Nuclear Reactor Regulation
Shared Package
ML20011F035	List: ML20011F034
References
NUDOCS 9002230501
Download: ML20011F034 (2)

Text

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AI DEPARTMENT OF MECHANICAL ENGINEERING e

{f 2.km p THE UNIVERSITY OF TEXAS AT AUSTIN

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.wy February 12, 1990 U.

S. Nuclear Regulatory Commission Director of Reactor Regulation Washington, D.

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20555 Attn:

Al Adams Project Manager, Docket 50-602

Subject:

Revision of Technical Specifications (January 1990)

Dear Sir:

Pursuant to section 50.54b(2):

10 copies of the Technical Specifications (January 1990) for docket 50-602 are submitted.

Please replace all previous Technical Specifications submittals with the present submittal dated January 1990.

Sincerely, d'

2. b Thomas L.

Bauer Assistant Director / Reactor Supervisor of NETL APPROVED:

Bew W/. w/cl&

Bernard W.

Wehring, 0:irector Nuclear Engineering 'Ieaching Laboratory TLB:jgr Enclosures cc:

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Lott H. Walls J. R. Howell H. Marcus L. J.

Callan (Region IV NRC) 9002230501 900212 Q

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dEh DEPARTMLNT OF MICHANICAL ENGINEERING g5 a% y lL]w THE UNIVERSITY OF TEXAS AT AUSTIN n

4,I Q:y? Nudur Enginanng %hing Ltleratory ' 10100 Bunwt Rud* Austin, has 78738 (512) 471 5787

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of S

S The University of Texas 5

at Austin 5

Docket No. 50-602 S

Balcones Roscarch Center S

Nuclear Engineering Teaching S

Laboratory (NETL) 5 AFFIDAVIT Gerhard J.

Fonken being duly sworn, hereby deposes and says that he is Executive Vice President and Provost, The University of Texas at Austin; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the enclosed-Technical Specifications (January 1990) as part of the docket 50-602 submittal; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge and belief.

H.

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Gerhatd J.

Fonken Executive Vice President and Provost STATE OF TEXAS S

Subscribed and sworn to befor e,

a Notary Public in and for the State of Texas, this /.ge.a Jday of Jaluenbj, 1970

(]itt L2 h Y nio-jotary Public in ahd for the State of Texas

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I Technical Specifications Docket 50-602 r

4 The University of Ter.er. at Austin j

TRICA Reactor i

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l-January 1990 i

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Technical Specifications 1/80 0

Table of Conterts

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1.0 DEFINITIONS 1.1 Certified Operators 5

1.1.1 Clast A Peactor Operator 5

1.1.3 Class B Reactor Operator 5

1.2 Channel 5

1.2.1 Channel Test 5

1.2.2 Channel check 5

1.2.3 Channel Calibration 5

1.3 Confinement 5

1.4 Experiment 5

1.4.1 Experiment, Moveable 6

1.4.2 Experiment, Secured 6

1.4.3 Experimental Facilities 6

1.5 Fuel Element, Standard 6

1.6 Fuel Element. Instrumented 6

1.7 Mode; Manual Auto, Square Wave, Pulse 6

1.8 Steady State 6

1.9 Opersble 6

1.10 Operating 7

1.11 Protective Action 7

1.11.1 Instrument Channel Level 7

1.11.2 Instrument Subsystem Level 7

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1.11.3 O'

Instrument System Level 7

1.11.4 1.12 Reactivity, ExcessReactor Safety System Level 7

7 1.13 Reactivity Limit 7

1.14 Reactor Core, Standard 7

1.15 Reactor Core, Operational 8

1.16 Reactor Operating 8

1.17 Reactor Safety System 8

1.18 Reactor Secured 8

1.19 Reactor Shutdown 8

1.20 Reference Core Condition 9

1.21 Research Reactor 9

1.22 Rod, Control 9

1.22.1 Shim Rod 9

1.22.2 Regulating Rod 9

1.22.3 Standard Rod 9

1.22.4 Transient Rod 9

1.23 Safety Limit 9

1.24 Scram Time 10 1.25 Shall, Should and May 10 1.26 Shutdown Margin 10

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1.27 Shutdown, Unscheduled 10 1.28 Surveillance Activities 10 1.29 Surveillance Intervals 10 1.30 Value, Measured 11 1.31 Value. True 11 O

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Toshnical Sp cifiesti:ns 1/8)

,,'s l Table of Conterts 1.0 DEFINITIONS 5

1.1 Certified Operators 5

1.1.1 Class A Peactor Operator 5

1.1.3 Class B Reactor Operator 5

1.2 Channel 5

1.2.1 Channel Test 5

L 1.2.2 Channel check 5

1.2.3 Channel Calibration 5

1.3 confinement 5

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1.4 Experiment 6

1.4.1 Experiment Moveable 6

1.4.2 Experiment. Secured 6

1.4.3 Experimental Facilities 6

1.5 Fuel Element, Standard 6

1.6 Fuel Element. Instrumented 6

1.7 Node; Manual. Auto, Square Wave Pulse 6

1.8 Steady State 6

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

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1.11.4 Reactor Safety System Level 7

1.12 Reactivity, Excess 7

1.13 Reactivity Limit 7

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 8

1.19 Reactor Shutdown 9

1.20 Reference Core Condition 9

1.21 Research Reactor 9

1.22 Rod, Control 9

1.22.1 Shim Rod 9

1.22.2 Rt.gulating Rod 9

1.22.3 Standard Rod 9

1.22.4 Transient Rod 9

1.23 Safety Limit 10 1.24 Scram Time 10 1.25 Shall, should and May 10 1.26 Shutdown Margin 10 1.27 Shutdown, Unscheduled 10 1.28 Surveillance Activities 10 1.29 Surveillance Intervals 11 1.30 Value, Measured 11 1.31 Value. True 11 O

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T;chnisc1 Specific:ticns 1/89 i

1 75 kJ 2.0 SAFETY L7.MITS AND LIMITING SAFETY SYSTEM SETTINGS 12 i

i 2.1 SAFETY LIMIT 12 2.2 LIMITING SAFETY SYSTEM SETTINGS 12 2.2.1 Fuel Temperature 12 2.2.2 Power Level (Non Pulse) 12 2.2.3 Reactivity Insertion (Pulse) 12 l

t 3.0 LIMITING CONDITIONS FOR OPERATION 13 i

3.1 REACTOR CORE PARAMETERS 13 3.1.1 Excess Reactivity 13 3.1.2 Shutdown Margin 13 3.1.3 Transient Insortions 13 3.1.4 Fuel Elements 13 l

3.2 REACTOR CONTROL AND SAFETY SYSTEM 14 i

3.2.1 Control Assemblies 14 r

3.2.2 Reactor Control System 14 3.2.3 Reactor Safety System 15 3.2.4 Reactor Instrument System 15 3.3 OPERATIONAL SUPPORT SYSTEMS 16 3.3.1 Water Coolant Systems 16 3.3.2 Air Confinement Systems 17 3.3.3 Radiation Monitoring Systems 17 3.4 LIMITATIONS ON EXPERIMENTS 17 O

3.4.1 Reactivity 17 3.4 1 Irradiations 17 3.4.3 Materials 17 4.0 SURVEILLANCE REQUIREMENTS 19 4.1 REACTOR CORE PARAMETERS 19 4.1.1 Excess Reactivity 19 4.1.2 Shutdown Margin 19 4.1.3 Transient Insertion 19 4.1.4 Fuel Elements 19 4.2 REACTOR CONTROL AND SAFETY SYSTEM 19 4.2.1 Control Assemblies 19 4.2.2 Reactor Control System 20 4.2.3 Reactor Safety System 20 4.2.4 Reactor Instrument System 20 4.3 OPERATIONAL SUPPORT SYSTEMS 20 4.3.1 Water Coolant Systems 20 4.3.2 Air Confinement Systems 21 4.3.3 Radiation Monitoring Systems 21 4.4 LIMITATIONS ON EXPERIMENTS 21 4.4.1 Reactivity 21 4.4.2 Irradiations 21 4.4.3 Materials 21

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Tcchnital Sp:cific tions 1/89 5.0 DESIGN FEATURES 22 5.1 SITE AND FACILITY DESCRIPTION 22 5.1.1 Location 22 5.1.2 Confinement 22 5.1.3 Safety Related Systems 22 l

5.2 REACTOR COOLANT SYSTEM 23 5.3 REACTOR CORE AND FUEL 23 5.3.1 Fuel Elements 23 5.3.2 Control Rods 23 5.3.3 Configuration 23 5.4 REACTOR FUEL ELEMENT STORAGE 24 6.0 ADMINISTRATIVE 25 6.1 ORGANIZATION 25 6.1.1 Structure 25 6.1.2 Responsibility 25 l

6.1.3

' Staffing 26 i

6.1.4 Selection and Training of Personnel 26 6.2 REVIEW AND AUDIT 26 6.2.1 Composition and Qualifications 26 6.2.2 Charter and Rules 27 6.2.3 Review Function 27 6.2.4 Audit Function 27 6.3 OPERATING PROCEDURES 28 6.4 EXPERIMENT REVIEW AND APPROVAL 29

• O 6.5 REQUIRED ACTIONS 29 6.5.1 Action to be Taken in Case of 29 Safety Limit Violation 6.5.2 Action to be Taken in the 29 Event of a Reportable Occurrence 6.6 REPORTS 30 6.6.1

' Operating Reports 30 6.6.2 Special Reports 31 6.7 RECORDS 32 6.7.1 Records to be Retained for 32 the Lifetime of the Facility 6.7.2 Records to be Retained for 33 a Period of at Least Five Years or the Life of the Component 6.7.3 Records to be Retained for 33 at Least One Training Cycle APPENDIX A.1 Startup Program 34 A.2 Objectives & Bases for Safety Limits 37 A.3 Objectives & Bases for Limiting Conditions for Operations 39 A.4 Objectives & Bases for Surveillance Requirements 46 A.5 Objectives & Bases for Design Features 51 Page 4

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Tcch:te:1 Sp;cific:ti::s 1/89 i

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'V 1.0 DEFINITIONS 1.1 Certified Operators An individual authorised 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 B 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 referred to as a Reactor Operator.

l 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 operable.

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.

Y 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.

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4 Tcchnico: Sp;cificetts:s 1/89 l

i l.4 Experiment Any operation, component, or target (excluding devicer ruch as detectors, foils, etc.), which is designed to investigate 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 chield structure so as to be part of their derign.

1.4.1 Experiment. Hoveable A moveable experiment is one where it is intended that the entire experiment way be roved 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 substantially 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 forcen which can arise as a result of credible conditions.

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1.4.3 Fxperimental Facilities i

Experimenta) facilities shall mean rotary specimen rack, pneuratic transfer tube, central thimble, beamtubes and irradiation facilities in the core or in the pool and will include the cobalt-60 facility.

l.5. Fuel Element, Standard

-A fuel element is a single TRIGA element of standard type.

Fuel is U-ZrH clad in stain 1 con steel clad. Hydrogen to zirconium ratio is nominal 1.6.

1.6 Fuel Element, Instrumented An instrumented fuel element in a special fuel element fabricated for temperature measurement. The element shall have at leest one thermocouple erhedded in the fuel near the axial and radial midpoints.

1.7 Mode; Manual, Auto. Pulre. Square Wave Each mode operation shall mean operation of the reactor with the mode selection switches in the manual, auto, pulse or square wave position.

1.8 Steady-state

-/s Steady-state mode operation tha33 mean any operation of the reactor k--)

with the mode selection switcher in the manual, auto or square wave position. The pulse mode switch will define pulse operation.

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Technical Specific 0 tic s 1/89 e

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1.9 Operable Operable means a component or syster is capable of performing its intended function.

I 1.10 Operating j

Operating means a component or system is performing its intended i

function.

4 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 I

At the protective instrument channel level, protective action is the

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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 level, 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 i

At the protective instrument system level, protective action is the generation and transmission of the command signal for the safety i

shutdown equipment to operate.

1.11.4 Reactor Safety System Level At the reactor safety system level, protective action is the operation of sufficient equipment to immediately shut down the reactor.

1.12 Reactivity, Excess Excess reactivity is that amount of reactivity that would exist if all t

the control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical.

1.13 Reactivity Limits The reactivity limits are those limits imposed on the reactor core excess reactivity. Quantities are referenced to a reference core condition.

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• w 1.14 Reactor Core, Standard A standard core is an arrangement of standard TRIGA fuel in the reactor grid plate and may include installed experiments.

1.15 Reactor Core, Operational An operational core is a standard core for which the core parameters of excess reactivity, shutdown margin, fuel temperature, power calibration, and reactivity worths of control rods and axperiments have been detetained to satisfy the requirements set forth in the Technical Specifications.

1.16 Reactor Operating The reactor is operating whenever it is not secured or shutdown.

1.17 Reactor Safety Systems Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor

. protection or to provide information for initiation of manual protective action.

1.18 Reactor Secure The reactor is secure when:

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 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 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.

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Technical Sp;cific;ti:ns 1/89

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1.19 Reactor chutdown l

The reactor le shutdown if it is suberitical by at least ore dollar in the reference core condition and the reactivity of all experiments is accounted for, l

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.

1.22 kod, Control A control tod in n device fabricated from neutron abcorbing material ot fuel which is used to establish neutron ilux changes and to compensate for routine reactivity loses. A control rod may be coupled to its drive unit allowing it to perf orm a safety function when the f-,

coupling is disengaged.

1.22.1 Shim Rod A ehim rod is a control rod having an electric motor drive and scram capabilitico.

1.22.2 Regulating Red A regulating rod is a control rod used to maintain an intended powe-level and may be varied manually or by a servo-controller. The regulating rod shall have scram capability.

1.22.3 Stardard Rod The regulating and chim rods are standard control rods.

1.22.4 Transfert Pod 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 have scram capability.

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i Tcchnical Sp cifications 1/89 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 unc mtrolled 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.

1.25 Shall, should and May The word shall is used to denote a requirement. The word should is used to denote a reconenendation. The word may is used to denote permissica, neither a requirement nor a recommendation.

i' 1.26 Shutdown Margin r

uhutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical 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 unscheduled 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.Z7 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 annuelly to biennially.

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'l.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

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following:

1.29.1 5 years (interval not to exceed 6 years).

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1.29.2 2 years (interval not to exceed 2-1/2 years).

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1.29.3 Annual (interval not to exceed 15 months).

i 1.29.4 Semiannual 2

J (interval not to exceed 7-1/2 months).

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1.29.5 Quarterly l-(interval not to exceed 4 months).

l 1.29.6 Monthly (interval not to exceed 6 weeks).

1 1.29.7 Weekly l

(interval not to exceed 10 days).

l 1.29.8 Daily (must be done during the calendar day).

1.30 Value. Measured The measured value is the value of a parameter es 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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Tcchnical Sp:cificctirne 1/89 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit Specification (s)

The maximum temperature in a standard TRIGA fuel element shall not exceed 1150*C for fuel element clad temperatures less than 500'C and shall not exceed 950'C for fuel element clad temperatures greater than 500'C.

l Temperatures apply to any condition of operation.

2.2 Limitina safety System Settings 2.2.1 Fuel Temperature Specification (s)

The limiting safety system setting shall be 550'c es measured in an

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instrumented fuel element. One instrumented element shall be located in 4

the B or C ring of the reactor core configuration.

2.2.2 Power Level (Manual, Auto. Square Wave)

Specification (s)

O' The maximum operating power level for the operation of the reactor shall be 1100 kilowatts in the manual, auto and square wave modes.

2.2.3 Reactivity Insertion (Pulse)

Specification (s)

The maximum transient reactivity insertion for the pulse operation of the reactor shall be 2.2% Ak/k in the pulse mode.

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Tcchnic 1 Sp;cific:ti:ns 1/69 7s

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3.0 LIMITING CONDITIONS FOR OPERATION

-3.1 Reactor Core Parameters 3.1.1 Excess Reactivity Specification (s)

Maximum excess reactivity shall be 4.9% Ak/k.

3.1.2 Shutdown Margin Specification (s)

The reactor shall not be operated unless the shutdown margin provided by control rods is greater than 0.2% ok/k with:

a.

The reactor in the reference core condition.

b.

The most reactive control rod fully withdrawn.

c.

The highest worth moveable experiment in its most reactive state.

3.1.3 Transient insertions

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Specification (s)

Total worth of the transient rod shall be limited to 2.8% ak/k, and a.

The pulse reactivity insertion shall be limited by a mechanical block on the pulse rod, b.

The reactivity for pulse operation shall be initiated from power levels less than 1 kilowatt.

3.1.4 Fuel Elements 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 ift a.

In measuring the elongation, the length exceeds 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 em (1/16 inch).

c.

A clad defect exists as indicated by release of fission products.

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V 3.2 _ Reactor Control and Safety System 3.2.1 Control Assemblies Specification (s)

The renetor chall not be operated unleon the control rods are operable, and a.

Control rods shall not be operable if damage is apparent to the rod or drive acremblies.

b.

The scram time measured from the instant a simulated signel reaches the value of a limiting a:afety system setting to the instant that the slowest scrammable control rod reaches its fully inserted position shall not exceed 1 second.

c.

Maximum reactivity insertion rate of a standard control rod shall be less than 0.2% Ak/k per second.

3.2.2 Reactor Control System Spn1fication(s)

The reactor shall not he operable unless the minimum safety interlocks are operable. The following control system safety interlocks shall be

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operable:

Interlocks Number Effective Mode Rod Drive Control Operable Function Steedy-state Pulse

a. Startup Withdrawal 4

prevent rod X

All Control P.odr withdrawal for less than 2 counts per see

b. Simultaneouc Withdrawal 4 prevent rod X

All Control Rods withdrawal for j

two or more rods

c. Pulse Reactivity I

set position X

X Insertion for air impulse

d. Standard Rod 3

prevent standard X

Withdrawal rod withdrawal

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e. Transient Rod 1

prevent rod X

Withdrawal withdrawal for more 1

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Tcchniec1 Sp:cific tions 1/89 i

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3.2.3 Reactor Safety System l

i Specifiention(e)

The reactor shall not he operable unless the minimum safety chancelp are opereble. The following control rod scram safety channels shall be i

operable.

Number Effective Mode Safety Channel Operable Function Steady-state Pulse a.

Fuel Temperature 2

Scram at 550*C X X

b.

Percent Power Level 2

Scram at 110%

X c.

liigh Voltage 2

Scram on loss X

X d.

Pulse ny, nyt 1

Scram at 20 Mw - secs.

X 2000 Mws.

X e.

Magnet Current 1

Scram on loss X

X f.

Manual Scram 1

Scram on demand X X

Console Button 3.2.4 Reactor Instrument System Specification (s) i A minimum configuration of measuring channels shall be operable. The following minimum reactor parameter measuring channelt shall be operable Number Effective Mode'~

Measuring Channel Operable Steady-state Pulse a.

Fuel Temperature 2

X X

b.

Percent Power Level 2

X c.

Pulse Power 1

X d.

Pulse Energy 1

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T;chnical Sp:cific:tions 1/89 i

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3.3,0perational, Support Systems 3.3.1 Water Coolant Systems Specification (s)

Corrective action shell be taken or the reactor shut down if the following reactor coolant vnter conditions are observed a.

The bulk pool water temperature exceeds 48'C.

h.

The water depth is less than 6.5 meters measured from the pool bottom to the pool water surface, c.

The water conductivity ic greater than 5.0 vaho/cm averaged for measurement periods of one month, d.

The pressure difference during heat exchanger operation is less than 7 kPa (1 psip) measured between the chilled water outlet pressure and the pool water inlet pressure to the heat exchanger.

3.3.2 Air Confinement Systems Specification (s)

The reactor shall not be operated unless minimum conditions for air confinement are funettonal. The following minimum conditions shall exists a.

Equipment shall be operable to isolate the reector area by closure of ventilation supply and exhaust dampers.

h.

A radioactivity level signal shall initiate confinement of the reactor area ventilatien nyntem.

c.

An auxiliary air purge system shall be opereM e to control air activation in experiment systems.

1 argon 4 from the facility shall be d.

The annual averer'f re ease 3 6

limited to 2 x 10 pCi/cm 3.3.3 Kadiation Monitoring Systems Specification (s)

The reactor chall not be operated unless minimum conditions for radiation ocasurement are functional. The following minimum conditions shall exicts e.

A continuous air monitor (particulate) shall be operable with readout and audible alarn. 1.ocation of the monitor shall be at the pool accesc icvel and within 5 meters of the pool deck.

b.

Three area radiation monitorr (gamma) shall be operable with readout and audible alarm, one of which shall be located in the vicinity of the top of the reactor pool, page 10

To:hnical Specificctims 1/89

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c.

A portable radiation monitoring device shall be available. The portable radiation monitor may be substituted for a monitor that

-is not operable for periods not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.'

3.4 Limitations on Experiments 3.4.1 Reactivity Specification (s)

The reactor shall not be operated unless the following conditions governing experiment reactivity exists a.

A moveable experiment shall have a reactivity worth less than 1.00 dollar.

b.

The reactivity worth of any single experiment shall be less than 2.50 dollars.

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.

3.4.2 Irradiations Specification (s)

A device or material shall not be irradiated in an irradiation facility under the classification of an irradiation unless the following conditions exists a.

The irradiation meets all the specifications of Section 3.4 for an experiment, b.

The device or material is encapsulated in a suitable container.

c.

The reactivity worth of the device or material is 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 I rem /hr at one meter after 10 min., otherwise it shall be classed as an i

experiment.

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e.

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.

3.4.3 Materials Specification (s)

The reactor shall not be operated unless the following conditions governing experiment materials exist:

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Technical Sp3cificatiens 1/89 j

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a.

Experiments containing materials corrosive to reactor components.

l1 compounds highly reactive with water, potentially explosive

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materials, and liquid fissionable materials shall be doubly encapsulated.

i b.

If a capsule fails and releases material which could damage the

)

reactor fuel or structure by corrosion or other means, removal l

s and physical inspection shall be performed to determine the consequences and need for corrective action. The results of the inspection 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.

l c.

Each fueled experiment shall be controlled such that the total

[

inventory of iodine isotopes 131 through 135 in the experiment is 1

l no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 mil 11 curies.

d.

Explosive materials, such as gunpowder, nitroglycerin, trinitrotoluene, or pentaerythritol tetranitrate in quantities i

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 l

produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.

e.

Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible l

accident conditions in the reactor, (3) possible 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 e. 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.

O Page 18

I Tcchnical Sp:cific tiens 1/89 l

l

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r 4.0 SURVEILLANCE REQUIREMENTS w

4.1 Reactor Core Parameters

{

4.1.1 Excess Reactivity Specification (s)

Excess reactivity shall be determined annually or after significant reactor core or control rod changes.

4.1.2 Shutdown Margin Specification (s)

Excess reactivity and shutdown margin shall be determined annually or after significant core or control rod changes.

l 4.1.3

. Transient Insertion Specification (s)

The transient rod drive cylinder and associated air supply shall be inspected, cleaned, and lubricated annually, 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.

4.1.4 Fuel Elements 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.

4.2 Reactor Control and Safety System 4.2.1 Control Assemblies Specification (s)

Control rod worths shall be determined annually of after significant core or control rod changes, and O

Page 19

y T chnical Sp;cific tir s 1/89 I

b,

{

a.

Each control rod shall be inspected at biennial intervals by f

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.

i 4.2.2 Reactor Control System Specification (s)

The minimum safety interlocks shall be tested at semiannual intervals.

1 i

4.2.3 Reactor Safety System Specification (s) i The minimum safety channels shall be calibrated annually and tested prior j

to each days operation or prior to each extended period of operation.

j 4.2.4 Reactor Instrument System j

Specification (s)

O 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, j

l 4.3 Operational Support Systems 4.3.1 Water Coolant Systems 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.

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.

O I

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L Technical Specific tions 1/89 r

4.3.2 Air Confinement Systems Specification (s)

The following actions shall demonstrate the air confinement conditions:

a.

Annual examination of door seals and isolation dampers, b.

Monthly functional tests of air confinement isolation.

i c.

Calibration of argon-41 measurements shall be made annually and measurements or calculations performed quarterly.

4.3.3 Radiation Monitoring Systems l

Specification (s)

The minimum radiation monitors specified to be operable during reactor operation shall be 4

a.

Calibrated at semiannual intervals.

b.

Checked at monthly intervals for fixed monitor.

l c.

Checked daily for portable monitor.

O L/ -

4.4 Limitations on Experiments 4.4.1 Reactivity Specification (s)

The reactivity of an experiment shall be measured before an experiment is considered functional.

4.4.2 Irradiations Specification (s)

Experiments classified as irradiations shall be identified and a log or cther record maintained while the sample is in the reactor.

4.4.3 Materials Specification (s)

Any surveillance conditions or special requirements shall be specified as a part of the experiment approval.

page 21

4 t;.

'l Technic 1 Sp;cific tiens 1/89 5.0 DESIGN FEATURES 5.1= Site'and Facility Description 5.1.1 Location Specification (s) a.

The site location is in the northeast corner of The University of Texas at Austin Balcones Research Center.

C b.

The.TRIGA reactor is installed in a designated room of a building constructed as a Nuclear Engineering Teaching Laboratory.

c.

The reactor core is assembled in an above ground shield and pool structure with horizontal and vertical access to the core, d.

Restricted access area of the facility shall consist of the room enclosing the reactor shield and pool structure, and the adjacent area for reactor control.

5.1.2 Confinement Specification (s)

[d^T.

a.

The reactor room shall be designed to restrict leakage and will have a minimum enclosed' air volume of 4120 cubic maters.

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 level.

c.

An air purge system should exhaust experiment air cavities and shall be filtered by high efficiency particulate absorption filters.

J.

All exhaust air from the reactor area enclosure shall be ejected vertically upward at a point above the facility roof level.

5.1.3 Safety Related Systems 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, i

Page 22

Techniccl Sp:cific:tiens 1/89

_7}

5.2. Reactor Coolant System' Specification (s)

a..

The reactor core shall be cooled by natural convection flow of

~

water.

b.

Pool water level shall be protected by holes for siphon breaks in pool water system pipe lines, 5.3 Reactor Core and Fuel 5.3.1 Fuel Elements Specification (s)

The standard TRIGA fuel element at fabrication shall have the following characteristics:

.l

\\

a.

Uranium content: 8.5 We% uranium enriched to a nominal 19.7%

l Uranium-235.

l b.

Zirconium hydride aton ratio: nominal 1.6 hydrogen to zirconium, ZrH.

i c.

Cladding: '304 stainless steel, nominal.020 inches thick.

5.3.2 Control Rods i

-Specification (s) j

-l The shim, regulating, and transient control rods shall have scram capability -and I

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.

l b.

Contain borated graphite, B C powder, or boron and its compounds in solid form as a poison. 4 c.

The transient rod shall have an adjustable limit to allow a variation of reactivity insertions.

Page 23

t 1

Technical Specificaticas 1/89 5.3.3 configuration Specification (s)

'fhe reactor shall be an arrangement of core single grid positions occupied by fuel elements, control rods, and graphite elements. Single element positions may.be occupied by voids, water or experiment facilities.

l Special multielet ert positions or single element positions may be. occupied by approved experiments.

5.4 Reactor Fuel Element Storage Specification (s)

.a.

All fuel element nall be stored in a geometrical array where the effective multiplication is less than 0.8 for all conditions of moderation.

b.

Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by.

water or air such that the fuel element or fueled device j

temperature will not exceed design values.

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1 4

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Tschnical Sptcificcticus 1/89 I

L 6.0 ADMINISTRATIVE CONTROLS 4

6.1 Organization.

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 l

Dean College of Engineering l

Radiation Chairman Nuclear Safety Department of Reactor O

Committee Mechanical Engineering Committee l

l 8

l Director I

i Nuclear Engineering i

l._. _ _ _ _

Teaching Laboratory

..__..i i

I I

l_,

Reactor Supervisor, Health Physicist l

Reactor Operators. Technicians, Others r

Responsibility L

Communication l'

(

.2 Responsibility The Director shall be responsible to the Dean of the College of Engineering and the Chairman of the Department of Mechanical Engineering for safe operation and maintenance of the reactor and its L

l associated equipment. The Director or his appointee shall review and l-approve all experiments and experimental procedures prior to their use in the reactor.

Individuals of the management organization shall l

(~N be responsible for the policies and operation of the facility, and L

shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to the operating license and technical specifications.

L Page 25 l

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Technical Spscificatiens 1/89

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• =

,x 6.1.3 Staffing The minimum staffing when the reactor is not secured shall be

-a.

A certified operator in the control room.

b.

A second person in the facility area that can perform prescribed written instructions. Unexpected absence for two hours shall require inuediate action to obtain an

, alternate person.

c.

A designated Class A operator readily available. The available operator should be within thirty minutes of the facility and reachable by telephone.

Events requiring the direction of a Class A operator shall bet a.

All fuel element or control rod relocations within the 4

t reactor core region.

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 includes a.

Management personnel, b.

Radiation safety personnel, c.

Other operations personnel.

6.1.4 Selection and Training of Personnel The selection, training and regulification of operators shall meet or exceed the requirements of American National Standard for Selection and Training of Personnel for Research Reactors ANSI /ANS - 15.4.

i Qualification and requalification of certified operators shall be l

subject to an approved NRC (Nuclear Regulatory Con: mission) program.

1

[;

6.2 Review and Audit i

6.2.1 Cotoposition and Qualifications A Nuclear Reactor Committee shall consist of at least three (3) members appointed by the Dean of the College of Engineering that are l

knowledgeable in fields which relate to nuclear safety. The University Radiological Safety Officer shall be an ex-officio member of the Nuclear Reactor Committee. The committee will perform the functions of review and audit or designate a knowledgeable person for audit functions.

Page 26

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Technical Sp:cificstions 1/89 t

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s

's/ :.

6.2.2

' Charter and Rules The operations of the' Nuclear Reactor Committee shall be in accordance'with an established charter, including provisions for p

a.

Meeting frequency (at leant once each six months).

b..

Ouorunn-(not less than one-half the memberchip).

c.

Dissemination, review, and approval of minutes.

d.

Use of subgroupn.

6.2.3 Review Function The review function shall include facility operations related to reactor and radiological safety. -The following items shall be reviewed:

a.

Determinations that proposed changes in equipment, systems, testo, experiments, or procedures do not involve an unreviewed safety question.

b.

All new procedures and major revisions thereto, and proposed changco in reactor facility equipment or systems having safety significance.

c.

All new experiments or classen of experiments that could affect reactivity or result in the release of radioactivity.

L d.

Changes in technical specifications or license, e.

Violations of technical specifications or license.-

f.

Operating abnormalities or violations of procedures having safety significance, g.

Other reportable occurrences.

h.

Audit reports.

6.2.4 Audit Function b

The audit function shall be a selected examination of operating records, logs, or other documents.

An audit will be by a person not directly responcibic for the records and may include discussions with cognizant personnel or observation of operctierc. The following items shall he audited red a report made to the Reactor Supervisor

,o and Nuclear Practor Operation Committees b

Page 27

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Tschniccl Sp cificatitna 1/89

)

g V

a.

Conformance of f acility operations with license and technical specifications at least once each calendar year.

j b.

Results of actions to correct deficiencies that may occur in reactor facility equipment, structures, systems, or methods of operation that affect safety at least once per calendar year.

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.

6.3 Operating Procedures

'4ritten operating procedures shall be prepared reviewed and approved by the Director or a designated alternate and the Nuclear Reactor Committee prior to initiation of the following activities a

Startup, operation, and shutdown of the reactor.

- /~%

b.

Fuel loading, unloading and movement in the reactor.

U c.

Routine maintenance of major components of systems 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 that could have an effect on reactor safety.

f.

Personnel radiation protection consistent with applicable l

regulations.

g.

Implementation of required plans such as the emergency plan j

or physical security plan.

L Substantive changes to the above procedure shall be made effective after approval of the Director a designated alternate and the Nuclear Reactor Committee. Minor modifications to the original procedures which do nut change the original int'ent may be made by a Class A operator but 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 circumstances or conditions. Such deviations shall be documented and reported to the Director or the designated alternate.

l Page 28

i Technical Sp cifiesticns 1/89 fi k,{

6.4 Experiment Review and Approval All new experiments or classes of experiments shall be approved by the-Director or designated alternate and the Nuclear Reactor Committee.

t a.

Approved experiments shall be carried out in accordance l

with establishe6 and approved procedures.

b.

Substantive changes to previously approved experiments shall require the same review as a new experiment.

c.

Minor changes to an experiment that do not significantly alter the experiment may be made by a Class A operator.

6.5 Require'd Actions 6.5.1 Action to be Taken in Case of a Safety Limit Violation-In the event of a safety limit violation, the following action shall be takent 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 Nuclear Regulatory Commission (NRC).

b.

The safety limit violation shall be-promptly reported to the Director of the facility or a designated alternate, c.

The safety limit violation shall be subsequently reported o

to the NRC.

d.

A safety limit violation report shall be prepared and submitted to the Nuclear Reactor Committee. The report shall describe: (1) Applicable circumstances leading to the violation including, 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 actions taken to prevent recurrence.

6.5.2 Action to be Taken in the Event of an Occisrrence that is Reportable.

In the event of a reportable occurrence, the following action shall be taken:

a.

Reactor conditions shall be returned to normal or 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.

Page 29

Tcchnic:1 Spscific:tions 1/89

,A J'

b.

Occurrence shall be reported to the Director or his t

designated alternate and to licensing authorities as required.

c.

Occurrence shall be reviewed by the Nuclear Reactor Committee at the next regularly scheduled meeting.

6. 6' Reports All written reports shall be sent within the prescribed interval to l

the NRC, Washington D.C. 20555, Atten: Document control Desk, with a copy to the Regional Administrator, Region IV.

6.6.1 Operating Reports Routine annual reports covering the activities of the reactor facility during the previous calendar year shall be submitted within three months following the end of each prescribed year. Each annual operating report shall include the following information:

a.

A narrative summary of reactor operating experience including the energy produced by the reactor or the hours the reactor was critical, or both.

'q b.

The unscheduled shutdowns including, where applicable,

(/

corrective action taken to preclude recurrence.

c.

Tabulation of major preventive and corrective maintenance operations having safety significance.

d.

Tabulation of major changes in the reactor facility and procedures, and tabulation of new tests or experiments, or both, that are significantly different from those performed previously, including conclusions that no unreviewed safety.

questions were involved.

e.

A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge. The summary shall include 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 is sufficient.

f.

A summarized result of environmental surveys performed outside the facility.

g.

A summary of exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed or recommended.

Page 30

o at Technic:1 Spscifications 1/89

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.l 1

ff~')L 6.6.2-Special Reports A written report within 30 days to the NRC oft a.

Permanent changes in the facility organization involving i

Director or Superviser.

b.

Significant changes in transient or accident analysis as

,)

described in the Safety Analysis Report.

A report to Region IV by telephone not later than the following working day and confirmed in writing by telegraph or similar-conveyance to be followed by a written report within 14 days that describes the circumstances of the event of any of the following a.

Violation of fuel element temperature safety limit, b.

Release of radioactivity above allowable limits.

c.

Other reportable occurrences.

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 in addition to those required by the

()

technical-specifications, 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.)

a.

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 safety function unless the malfunction or' condition is discovered during maintenance tests or periods of reactor shutdowns.

d.

Abnormal and significant degradation in reactor fuel, or cladding, or both, coolant boundary, or confinement boundary (excluding minor leaks) where applicable which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both.

('T e.

An observed inadequacy in the implementation of

\\s /

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.

Page 31 3.

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s Tcchnical Spseific:ticas 1/89 f^y

^

A written report within 90 days or 9 months, which ever 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 the rod of highest reactivity worth, b.

Minimum shutdown margin of the' reactor both at ambient and operating temperatures.

c.

An evaluation of facility performance to date in comparison with design conditions and measured operating characteristics, and a reassessment of the safety analysis when measurements indicate that there may be substantial variance from prior analysis submitted with the license application.

6.7 Records The records may be in the form of logs, data sheets, or other

' ~

suitabit forms. The required information may be contained in single.

- l' or multCple records, or a combination thereof.

6.7.1

. Records to be. Retained-for the Lifetime of the Reactor Facility:-

(Note:. Applicable annual reports, if they contain all of the required information, may be used as records in this section.)

a.

Gaseous and liquid radioactive effluents released to the environs.

b.

Offsite environmental monitoring surveys required by technical specifications, c.

Radiation exposure for all personnel monitored, d.

Updated drawings of the reactor facility.

Page 32

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Technicc1 Spscificctions 1/89 t L

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6.7.2 Records to be Retained for a Period of at Least Five Years or' for the Life of the Component Involved Whichever is-Shorter a.

. Normal reactor facility' operation.(supporting documents such as checklists, log sheets, etc. shall be maintained-for_a period of.at least one year).

s b.

Principal maintenance: operations.

c.

Reportable occurrences.

d.

Surveillance activities required by technical specifications.

e '.

Reactor facility radiation and contamination surveys there required by applicable regulations.

f._

Experiments performed with the reactor.

g.

Fuel inventories, receipts, and shipments, h.

Approved changes in operating procedures, i.

Records of meeting and audic reports of the review and it audit group.

6.7.3 Recorda~to be Retained for at Least One Training Cycle Retraining and requalifications of certified operations personnel.

Records of the most'recent complete cycle shall be maintained at all times the individual is employed.

9 Page 33

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• 1 Tschnic:1 Specifications 1/89 1

g APPENDIX Information in this appendix supplements the Technical Specifications. Section A.1 defines the Startup Program by which the licensee w!11 initial operation of the reactor.

Sections A.2 through A.5 document the intent of each specification by defining the applicability, objective and bases that relate to each specification.

A.1.0 STAkTUP PROGRAM The following is a brief outline of the procedures to be followed in performing the startup and calibration of the TRIGA reactor.

By following these procedures, all necessary instrument checks and nuclear calibrations will be performed. At each step comparison of measurements with calculated or expected values will be made and explanation of discrepancies provided before proceeding to the next test.

A.1.1.

Pre-Critical Tests a.

Functional Tests of Mechanical Equipment This includes verification of proper installation and performing operational tests on the isotope production facility, pneumatic system, beam ports, fuel element

_!.( )-

handling tool, water system and other related equipment.

Also included are area radiation monitors, continuous air monitors, and ventilation system oporation, control and confinement.

b.

Checkout of Instrumentation and Control System i

This includes a complete checkout of the control console by el.etronics personnel and ensures that all circuits are calibrated and operating satisfactorily.. Additionally, the pneumatic and standard rod drives are checked for proper operation during rod withdrawal and scram situations. The fission and ion chambers are shown to be sensitive and

-responsive te neutrons.

A picoammeter and a scaler are used as a temporary addition to the console instrumentation to provide more information during initial startup.

I O

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Tcchnical Sp::cifications 1/89

,~

A.I.2.1 Initial Loading of Fuel Elements to Criticality Using Inverse Multiplication Data Loading proceeds according to standard procedures used on the many TRIGA reactors in operation.

At the end of each loading _

step data are recorded (the number of elements loaded.in each step is determined by 1/2 the number projected for criticality, or one, whichever is greater). The first four loading steps-proceed with all control rods withdrawn. The remainder of the loading. steps proceed with the shim rod inserted and the other rods withdrawn. A plot of the number of fuel elements loaded versus the inverse multiplication is maintained and extrapolated to predict criticality. The weight of uranium in each fuel element-varies slightly, so the elements are positioned selectively to group the heavier elements at the core's center, All inner ring grid plate holes will be filled with fuel elements, control rods or other core components. The source is removed temporarily upon reaching criticality to verify criticality.

A.I.3.

Tests to be Performed Upon Reaching Criticality a.

Approximate Calibration of Control Rods by Rod Dron Method i

This is only a rough approximation used to give some

~-

indication of the rods' values.

b._

Adjustment of Excess Reactivity by Addition of Fuel Elements and Preliminary Control Rod Calibration Additional fuel is added to the core..The outer ring of the core will initially contain the pneumatic 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 made during the loading of the excess reactivity.

c.

Final Calibration of Control Rods by the Period Method and Determination of Final Excess Reactivity and Core Shutdown Reactivity 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 Meter Using the rod calibration curves, a predetermined amount of excess reactivity is inserted. Using the in-hour curve.

Q the period corresponding to that insertion can be V

determined and the period meter adjusted accordingly.

Page 35

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Technical Sp*cificctiens 1/89 j

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J-

~ Tests to be Performed at Higher Powers A.1.4.

a.

Power Calibration by the Method of Rate of' Rise of Water Temperature j

Reactor heat added to the water at a constant rate raises the water temperature X degrees per hour. This rate of i

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 analysis of other standard TRIGA reactor tanks.

b.

Check of Instrument Linearity by Going to Full Power in Decade Steps Console instrumentation is compared with the reference picoammeter readings.

A.1.5.

Demonstration Tests at Rated Power a.

Demonstration of performance and reliability of reactor system by operation at rated power, b.

Radiation Survey to Demonstrate Adequacy of Reactor q

Shielding V

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 these measurements is to determine the pulse characteristics and to check the nv circuits calibrated-extrapolation from full steady state power. A high-speed recorder, calibrated potentiometer and a picoammeter are used as auxiliary instrumentation. The high-speed recorder is used to give a graphic picture of the pulse, the shape of which can be compared with pulses from other TRIGA reactors. The peak power shown on the high-speed recorder is compared with the peak power information displayed on the console recorder to assure accurate calibration.

Temperature information from thermocouples embedded in the fuel elements can be read out on both recorders and compared. Pulse data are recorded.

Page 36 o

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Tschnical Spscifications 1/89 iG V~

A. 2' 0 SAFETY LIMITS 6 LIMITING SAFETY SYSTEM SETTINGS

~ OBJECTIVES AND BASES A.2.1 Safety Limit Applicability i

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.

Bases The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well suited as a single 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

-f-)s 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 ll50*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.

A.2.2.1 Fuel Temperature Applicability This specification applies to the protective action for the reactor fuel element temperature, h

Objective L

The objective is to prevent the fuel element temperature safety limit from being reached.

L.

lL l

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-Tcchnical Spacificoticns 1/89 9

D..

D-Bases 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 l

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 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 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 wi1.1 apply.

However, 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.

A.2.2.2 Power Level (Manual, Auto, Square Wave)

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.

Bases 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.

A.2.2.3 Reactivity Insertion (Pulse) 1 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.

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-Bases

(

Calculations-indicate that standard TRIGA fuel elements may be safely operated at transient conditions in excess of 2.2% ok/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% ok/k.

A.3.0 LIMITING CONDITIONS FOR OPERATION OBJECTIVES & BASES A.3.1 Reactor Core Parameters A.3.1.1 Excess Reactivity Applicability This specification applies to the reactivity condition of the reactor core in terms of the available excess above the enld xenon free, critical condition.

Objective The cbjective is to prevent the fuel element temperature safety limit fron. being reached-by limiting _the potential reactivity available in

_(

the reactor for any condition of operation.

Bases Maximum excess core reactivity is sufficient to provide the core rated power, xenon compensation and reactivity for 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.

y l

A.3.1.2 Shutdown Margin L

Applicability H

This specification applies to the reactivity margin by which 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.

Bases i

The value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the highest worth control Page 39 0

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• O rod should remain in the fully withdrawn position and an unsecured experiment is in a high reactivity state.

A.3.1.3 Transient Insertions l

Applicability This specification applies to the total potential worth of the transient rod and the allowable reactivity insertion for reactor pulse operation.

Objective The objective is to limit the reactivity available for pulse insertion to a-value that will not csuse the fuel temperature safety limit to be exceeded.

Bases Calculations demonstrate that the total insertion of all the transient rod worth will not exceed the fuel temperature safety limit. For a 2.8% ok/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 7s operation of TRIGA reactors by the manufacturer indicate that

(

insertions up to 3.5% ok/k have not exceeded the fuel temperature safety limit.

'A.3.1.4 Fuel Elements Applicability This specification applies to the measurement parameters for the fuel elements..

Objective The' objective is to verify the physical condition of the fuel element cladding.

Bases 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 TRICA reactors has shown that fuel element bowing f'

that could result in touching has occurred without deleterious effects.

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A.3.2 Reactor Control and Safety System A.3.2.1 Control Assemblies 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 scrammable control rods and reactivity-insertion rates for standard control rods.

Bases The apparent condition of the control rod assemblies will provide assurance *.nat 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 apecification for rod reactivity insertion rates assures that the reactor will start up controllahly 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.

A.3.2.2 Reactor Control System Applicability These specifications apply to logic of the reactor control system.

Objective The objective is to determine the minimum control system interlocks

. operable for operation of the reactor.

Bases Interlocks are specified to prevent function of the control rod drives unless certain specific conditions exist. The interlock to prevent startup of the reactor at power levels less than 2 neutron eps, which corresponds to approximately 4 milliwatts, assures that sufficient neutrons are available 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 steady state 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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~A.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.

Bases t

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 powcr 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 magnet current.

A.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.

-Bases 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. Measurements of the same or different parameters provide redundancy.

A.3.3 Operational Support System A.3.3.1 Water Coolant Systems Applicability This specification &pplies to the operating conditions for the reactor pool and coolant water systems.

Objective

-p/

The objective is to assure that adequate conditions are maintained to Page 42

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provide shielding of the reactor radiation, protection against corrosion of the reactor components, cooling of the reactor fuel, and prevent leakage from the primary coolant.

Bases The specifiestions for conditions of thn 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 exist.s 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, c.

Average measurements of pool coolant water conductivity of 5.0 paho/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.

A.3.3.2 Air Confinement Systems Applicability 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 l

experimental facilities, i

Bases The specifications for exhaust ventilation and isolation of the I

reactor bay provide control for radioactive releases by both routine and non routine operating conditions.

A.3.3.3 Radiation Monitoring Systems O(_/

Applicability This specification applies to the radiation monitoring conditions in Page 43

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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.

Bases The radiation monitors provide information to operating 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 personnel and relesse of radioactivity.

A.3.4 Limitations on Experiments A.3.4.1 Reactivity Applicability This specification applies to the reactivity associated with experiments located in the reactor core.

q Objective V

The objective is control the amount of reactivity associated with experiments to values that will not endanger the reactor safety-limit.

Bases 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.

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

-l 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.

o'

c. The maximum worth of all experiments is lin'ited so that removal of the total worth of all experiments will not exceed the fuel element temperature safety limit.

O A 3.4.2 Irradiations Q

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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.

Basen This specification is intended to provide assurance that the special class of experiments called irradiations will be performed in a safe manner.

A.3.4.3 Materials Applicability i

These specifications apply to experiments installed in the reactor and its experimental facilities.

,s V

Objective The objective'is to prevent the release of radioactive material in the event of an experiment failure..cither by failure of the experiment or subsequent damage to the reactor components.

Bases

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-curie 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 exposure dose at the exclusion area boundary is limited.

~

d. This epecification 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 airborno activities in excess of the maximum allowable limits will be s

(,)

released to the atmosphere outside the facility boundary,

f. This specification provides guidance for the calculation of conditions in part (e).

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'l A.4.0" SURVEILLANCE REQUIREMENTS OBJECTIVES & BASES A.4.1 Reactor Core Parameters A.4.1.1 Excess Reactivity Applicability This specification applies to the measurement of reactor excess reactivity.

s Objective The objective is to periodically determine the changes in core excess reactivity available for power generation.

Bases 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'.

A.4.1.2 Shutdown Margin j

, ~i.

Applicability

)

x This specification applies to the measurement of reactor shutdown margin.

Objective The objective is to periodically determine the core shutdown reactivity availabic for reactor shutdown.

l Bases 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.

A.4.1.3 Transient Insertion Applicability This specification applies to surveillance of the transient rod mechanism and to observation of the reactor transient response.

Objective

.f S The objective is to assure the function of the transient rod drive

(_j.

and to compare the reactor pulse insertion parameters.

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i Bases Ny Annual inspections lof 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.

i' A.4.1.4 Fuel Elements Applicability This specification applies to the inspection requirements for the fuel elements.

Objective r

The objective is to inspect the physical condition of the fuel-element cladding.

H Bases

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 characteristics are well known.

A.4.2 Reactor Control and Safety System A.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 operable condition of the rod by

-periodic measurement of the scram times and insertion rates.

i Bases Annual determination of control rod worths or measurements after significant core changes provide information about changen 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 assemblies. 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.

l 1

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~-V A.4.2.2

. Reactor Control System Applicability This specification applies to the tests of the logic of the reactor.

control system.

4

Objective The. objective-is to specify intervals for test, check or calibration of the minimum control system interlocks, a

Bases t

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.

A.4.7.3 Reactor Safety System i

Applicability Thi$ specification applies to tests of the function of the reactor s:

safety system.

fs Objective i

The objective is to specify intervals for test, check or calibration of-the minimum safety system scrams.

4 Bases

'The periodic calibration at annual intervals provides adequate 4

information that the setpoints of the safety system screms are j

functional. Tests of the safety system prior to each planned operation assure that each intended scram function is' operable.

A.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 L

of the minimum instrument channels.

l-Bases L

O Annual calibration of instrument channels are scheduled to allow l

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.l adjustments for changes in reactor. and instrumentation parameters.

Checks and tests'are applied prior to system operation to verify function of'the system.

A.'4.3

-Operational' Support Systems y

A.4.3.1 Water Coolant Systems Applicability i

-This specification applies to surveillance conditions for the reactor pot ? and coolant water systems.

Objective The objective is to maintain the reactor coolant conditions within acceptable specifications.

Bases Conditions for the reactor coolant are monitored by visual observation of measurements or automatic action of sensors. Periodic-L' checks and tests of measurement devices for the reactor coolant system parameters assure that the coolant system will perform its intended function.

.t' j'

A. 4. 3 J Air Confinement Systems Applicability This specification applies to surveillance conditions for the air ventilation in the reactor area.

Objective 5

The objective is to demonstrate the function of confinement and release of air from the reactor bay.

- Bt.ses Periodic' evaluations of air confinement criteria are determined by examination, test, and calibration of the appropriate ventilation functions. the reactor bay provide control for radioactive releases by both routine and non routine operating conditions.

A.4.3.3 Radiation Monitoring Systems Applicability This specification applies to the surveillance conditions of the g'~

radiation monitoring channels.

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Objective The objective is to assure the radiation monitors are functional.

Bases Periodic calibrations and frequent checks are specified to maintain reliable performance of the radiction monitoring instruments.

A.4.4 Limitations on Experiments A.4.4.1 Reactivity 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.

Bases G.-

The measured reactivity or determination that the reactivity is not significant will provide data that configuration of the experiment or experiments is allowable.

A.4.4.2 Irradiations Applicability This specification applies to the surveillance requirements for reactor irradiations.

Objective The objective is to provide a record of experiments inserted in the reactor as irradiations.

Bases Experiments performed as irradiations are monitored by data on the sample location, identification and other pertinent information.

A.4.4.3 Materials Applicability O

This specification applies to the surveillance requirements for materials inserted into the reactor.

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Objective The objective is to prevent the introduction of materials that could damage the reactor or its components.

Bases A careful evaluation of all experituents is performed to classify the experiment as an approved experiment.

A.5.0 DESIGN FEATURES OBJFCTIVES & BASES A.5.1 Site and Facility Descriptions A.5.1.1 Location Applicability This specification applies to the TRIGA reactor site location and specific facility design features.

Obj ective The objective is to specify those features related to the Safety Analysis evaluation.

Bases The TRICA facility site is located in an area controlled by The a.

University of Texas at Austin.

b.

The room enclosing the reactor has been designed with characteristics related to the safe operation of the facility.

The shield and pool structure have been designed for radiation c.

levels of less than 1 mrem /hr at locations that are not access ports to the reactor structure, 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.

A.S.I.2 Confinement Applicability This specification applies to he boundary for control of air in the area of the reactor.

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Obj ective t

i The objective is to assure that provisions are made to control or restrict the amount of release of radioactivity into the environment.

Bases Calculations of the concentrations of released radionuclides a.

within the reactor area depend on the available enclosed air volume to limit the concentrations to acceptable levels.

3 Control of the reactor area air exchange is by fan motors and b.

isolation dampers for the supply and exhaust air which are controlled by a logic signal from a radiation sensor to provide automatic air confinement.

Energency air ventilation is filtered to control the release of i

c.

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.

A.S.I.3 Safety Related Systems 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.

Bases 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 specifications are assumed to meet the presently accepted operating criteria.

Questions that may include an unreviewed safety question are referred to the reactor operation committee.

A.S.2 Reactor Coolant System Applicability This specification applies to the reactor coolant system composed of

(

-deionized water.

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Objective The objective is to assure that adequate water is available for cooling and shielding during reactor operation.

Bases a.

This specification is based on thermal and hydraulic calculations which show that a standard 85 element 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, b.

Siphon breaks set the subsequent pool water level for loss of coolant without an associated water return caused by inadvertant pumping or accidental siphon of water from the pool.

A.5.3 Reactor Core and Fuel A.S.3.1 Fuel Elements Applicability This specification applies to the fuel elements used in the reactor core.

U Objective The objective is to assure that the fuel elements are of such a design and fabricated in auch a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Bases The Design basis of the standard TRIGA 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. The fuel temperatures are not expected to exceed 550*C during any condition of normal operation.

A.S.3.2 Control Rods Applicability This specification applies to the control rods used in the reactor core.

Obj ective

/]

The objective is to assure that the control rods are of such a design

'Q as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

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Bases The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, B C powder, or boron and its 4

compounds. These materials must be contained in a suitable clad material, such as aluminum or stainless steel, to insuro mechanical stability during movement and to isolate the poison from the pool t

water c vironment.

Scram capabilities are provided for rapid

inse, of the control rods which is the primary safety feature of the reust;

'he transient control rod is designed for a reactor pulse.

A.S.3.3 configuration Applicability This specification applies to the configuration of fuel elements, control rods, experiments and other reactor grid plate components.

Objective The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.

fN Bases

\\J Standard TRIGA cores have been in use for years and their i

characteristics are well documented.

A.5.4 Reactor Fusi Element Storage Applicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Objective The objective is to assure that fuel storage will not become critical and will not exceed design temperatures.

Bases The limits imposed by these specifications are considered sufficient

)

to provide conservative fuel storage and assure safe storage.

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