ML20204F392

Proposed Tech Specs Re Reactivity Limits,Experiments,Control Rods,Radiation Monitoring Equipment,Reactor Bldg & Solid Waste

ML20204F392
Person / Time
Site:	05000113
Issue date:	10/17/1988
From:	ARIZONA, UNIV. OF, TUCSON, AZ
To:
Shared Package
ML20204F348	List: ML20204F346 ML20204F367 ML20204F392 ML20204F434 ML20204F531
References
NUDOCS 8810210569
Download: ML20204F392 (36)

Text

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' s TECHNICAL SPECIFICATIONS l

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8810210569 G31017 PDR ADOCK 05000113 PDC P

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• o 1.0 DEFINITIONS Channel - A channel is a combination of sensors, electronic circuits, and output devices connected by the appropriate communications network in order to measure and display the value of a parameter.

Channel Calibration - A channel calibration is an adjustment of a channel such that its output corresponds with acceptable accuracy to knosvn values of the parameter which the channel measures.

Calibration shall encompass the entire channel, including equipment, actuation, alarm, or trip and shall include a Channel Test.

Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. The verification shall include comparison of the channel output with previous readings or performance or with other independent channels or systems measuring the same variable, whenever possible.

Channel Test - A channel test is the introduction of a signal into the channel for verification that it is operable.

l Cold Critical - The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures the same (-20*C).

Experiment - An experiment is any device or material, not normally part of the reactor, which is l introduced into the reactor for the purpose of exposure to radiation, or any operation which is designed to investigate non-routine reactor characteristics.

Experimental Facilities - Experimental facilities are the thermal column, pneumatic transfer systems, central thimble, rotary specimen rack, beam tube, and the in-core facilities.

l Limiting Conditions for Operation - Limiting Conditions for Operation (LCO) are administratively established constraints on equipment and operational characteristics which shall be adhered to during operation of the reactor.

Limiting Safety System Setting (LSSS) - The LSSS is the actuating level for automatic protective devices related to those variables having significant safety functions.

Manual Mode - The reactor is in the manual mode when the reactor mode selection switch is in the manual or automatic position. In this mode, reactor power is held constant or is changed on periods of approximately one second or longer.

Measured Value - The Measured Value is the value of a parameter as it appears on the output of a channel.

Movable Experiment - An experiment is movable when it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.

Operable - Operable means a component or system is capable of performing its intended function.

Operating - Nrating means a component or system is performing its intended function.

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Pulse Mode - The reactor is in the pulse mode when the reactor mode selection switch is in the pulse position. In this mode, reactor power may be increased on periods less than one second by motion of the transient control rod.

Reactivity Worth of an Experiment - The reactivity worth of an experiment is the maximum value of the reac*.ivity change that would occur as a result of planned changes or credible malfunctions that alter experiment position or configuration.

Reactor Committee - The group of persons at the University who are assigned responsibility for review and audit of facility operation and review of changes and experiments in accordance with 10 CFR 50.59.

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

Reactor Safety Systems - Reactor Safety Systems are those systems, including associated input channels, which an designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

Reactor Secured - The reactor is secured when:

a. It contains insufficient fissile material or moderator present in the reactor, adjacent experiments or control rods, to attain criticality under optimum available conditions of moderation and reflection, or

b. 1. The minimum number of neutron absorbing control rods are fully inserted or other safety devices are in shutdown position, as required by technical specifications, and

2. The console key switch is in the off position and the key is removed from the lock, and

3. 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, and

4. No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth of one dollar or more.

Reactor Shutdown - The reactor is in a shutdown condition when sufficient control rods are inserted to assure that it is suberitical by at least 51.00 of reactivity.

Reportable Occurrence - A Reportable Occurrence is any of the following which occurs during reactor operstbre a) Operation with actual safety-system settings for required systems less conservative than the limiting safety-system settings specified in Technical Specification 2.2.

3 b) Operation in violation of limiting conditions for operation established in the Technical

Specifications.

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f 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 mi* unction or condition is discovered during maintenance tests or periods of reactor shuidown.

d) Any unanticipated or uncontrolled change in reactivity greater than one dollar, e) Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary which could result in exceeding of prescribed radiation exposure or release limits, f) An observed inadequacy in the implementation of either administrative or procedural controls which could result in operation of the reactor outside the limiting conditions for operation.

g) Release of radioactivity from the site above limits specified in 10GFR20.

Control Rod - A control rod is a device fabricated from neutron absorbing material or fuel which is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod may be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

Transient Rod - The trans:ent rod is a control rod with scram capabilities that is capable of providing rapid reactivity insertion to produce a pulse.

Safety Limit - A Safety Limit is a limit on an important process variable which is found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity. The principal physical barrier is the fuel element cladding.

Secured Experiment - A Secured Experiment is any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means.

The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, bt:oyant, or other forces which are normal to the operating environment of the expriment, or by forces which can arise as the resitlt of credible malfunctions.

Shall, Should, and hiay - The word "shall' is used to denote a requirement, the word "should*

denotes a recommendation, and the word 'may" denotes permission, neither a requirement nor a <

recommendation.

Shutdown hfargin - Shutdown hfargin is the reactivity existing when the most reactive control rod is Mily withdrawn from the core and the other control rods are fully inserted into the core.

Time Interval - Specification of the time between successive events as follows:

a) Diennially - at two year intervals (interval not to exceed 30 months) b) Annually - at one-year intervals (interval not to exceed 15 months) c) Semiannually - at 6-month intervals (interval not to exceed seven and one-half months) d) Quarterly - at 3-month intervals (interval not to exceed four months) e) hionthly - at one-month intervals (interval not to exceed six weeks)

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f) Weekly - at seven-day intervals (interval not to exceed ten days)

, 3) Daily - (must be done during the calendar day)

Untried Experiment - An untried experiment is any experiment not previously performed in this reactor.

' s 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit - Reactor Power Level Applicability.

This specification app!!es to the steady-state reactor power level.

Objective The objective is to define a power level below which it can be predicted with confidence that no damage to the fuel elements will occur.

Specification The power level of the reactor in the steady-state mode shall not exceed 1000 kW under any conditions of operation.

Basis It has been shown by expensive experience that operation of TRIGA reactors at a power level of 1000 kW will not result in damage to the fuel. Reactors of this type have operated successfully for decades at power levels in the range from 1000 to 1500 kW and above. The criterion for assuring the integrity of a TRIGA fuel element is that the fuel temperature be maintained below 1000*C. It has been shown by analysis and by measumments on other TRIO A reactors that a power level of 1000 kW corresponds to a peak fuel temperature of approximately 400*C. Thus, a Safety Limit on power level of 1000 kW provides an ample margin of safety for operation.

'2.2 Limiting Safety System Setting - Steady State Reactor Power Level Applicability This specification applies to the reactor power level safety system setting for steady state operation.

Objective The objective is to assure that the Safety Limit is not exceeded.

Specification The setting for the power leyw scram in steady state operation shall be no greater than 110 kW.

Basis Calculations and measurements show that at 110 kW, the peak fuel temperature in the core will be less than approximately 150*C which is well below the safety criterion of 1000*C and provides an ample safety margin to accommodate errors in measurement and anticipated operational transients.

2.3 Limiting Safety System Setting - Pulse Mode Reactor Power Level Applicability This specification applies to the reactor power level safety system setting for pulse mode operation.

Objective The objective is to assure that the fuel temperature specified by the Safety Limit is not exceeded in pulse mode operation.

Specification The setting for the peak power level scram in pulse mode operation shall be no greater than 1100 MW.

Basis Calculations and measurements show that at a peak power of 1100 MW in pulse mode operation, the peak fuel temperature in the core will be less than approximately 400*C. This provides an ample safety margin to accommodate errors la measurements and anticipated operational transients.

a 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits Applicability These specifications apply to the reactivity condition of the reactor, and the reactivity worths of control rods and experiments, and apply for all modes of reactor operation.

Objective The objective is to assure that the reactor can be shut down at all times and to assure that the safety limit will not be exceeded.

Specifications The reactor shall not be oper.ted unless the following conditions exist

a. The shutdown margin referred to the cold xenon-free condition is greater than 50.50 with the highest worth rod fully withdrawn and with the highest worth non-secured experiment in its most positive reactivity state.

b. Any experiment with a reactivity worth graater than $1.00 iJ secured so as to prevent unplanned reactivity removal from or insertion into the reactor;

c. The reactivity available to be inserted by the pulse rod is determined and is limited by a mechanical block to a maximum of $2.50,

d. The reactivity worth of an individual experiment is not more than 53.00;

e. The total of the absolute values of the reactivity worth of all experiments in the reactor is less than 55.00; .

f. A ramp or oscillating rod placed in the reactor cannot add more than $1.50 of reactivity;

g. The drop time of each standard control rod from the fully withdrawn position to 90 percent of full reactivity insertion is less than one second; and

h. The neutron count rate on the startup channel is greater than one count per second.

1. The maximum reactivity insertion rate by control rods for non-pulsed operation is less than 50.20/second.

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Bases The shut down rr.argin required by specification 3.la is necessary so that the reactor can be shutdown from any operating condition and remains shutdown after cooldown and xenon decay even if one control rod should stick in the fully withdrawn position.

Specification 3.lb is based on pulse measurements and analysis at the University of Arizona which indicate that as much as 53.00 reactivity could be inserted without increasing fuel temperature by more than 416'C. By restricting each non-secured experiment to a reactivity worth of one dohtr, r.n smple .nargin is provided.

Specifications 3.lc through 3.lf are intended to provide additional margins between those .

values of reactivity changes encocntered during the course of operations involving experiments ,

and those v. dues of enetivity which, if exceeded, might cause a safety limit to be exceeded, l Specific.d- ' " g it intuded to sssure prompt shutdown of the reactor in the event a : cram sigr# .

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n 'xnded t< mure that sufficient neutrons are available in the coro to prc'.ad? a s.i si : t e output of the startup channel during approaches to criticality.

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9 3.2 High Power Operation Applicability This specification applies to operation of the reactor at high steady-state power.

Objective The objective is to prevent inadvertent pulse operation of the reactor while it is at a high power level.

Specification The reactor shali not be operated in the steady-state mode at power levels above 10 kW unless, in addition to the conditions of Section 3.1 the transient rod is fully withdrawn.

Basis This specification is intended to prevent inadvertent pulse operation when the fuel temperature is above PC (corresponding to a power level of 10 kW) as measured in the B-ring. See Specification 3.3b.

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3.3 Pulse Operation Applicability The specifications apply to operation of the reactor in the pulse mode.

Objective The objective is to prevent the fuel temperature safety limit from being exceeded during pulse mode operation.

Specifications The reactor shall not be operated in the pulse mode unless, in addition to the requireinents of Section 3.1, the fo!!owing conditions exist t.. The transient rod is set such that the reactivity worth upon withdrawal is not greater than 52.50; and

b. The temperature of the fuel immediately prior to the pulse is essentially in equilibrium with the bulk water temperature. This is controlled by limiting the reactor power orior to pulsing.

Basis The specification 3.3a will maintain the maximum temperature of the fuel after a 52.50 pulse below 400cc above the bulk pool tempers: ore, and thus well below the 1000*C fuel safety criterion.

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3.4 - Reactor lastrumentation

- Applicability This specif*. cation applies to the information which must be available to the reactor operator during reactor operation.

Objective I The objective is to require that sufficient information is available to the operater to assure safe operation of the reactor. t Specification The reactor shall not be operated unless the measuring channels described in the following '

table are operable and the information is available in the control room:

Minimum Operating Mode Number in which Measurins Channel Operable Required Reactor Power Level (Linear) 1 Steady State i

Wide-range Los Power i Steady State Level (Startup count rate)

Reactor Power Level (high range) 1 Pulse Mode l Reac'or Tank Water Temperature  ! All Modes Area Radiation Monitors 2 All Modes Particulate Air Radiation Monitor 1 All Modes Reactor Water Activity Monitor i All Modes pa,jts, The neutron detectors assure that mearurements of the reactor power level are ade';aately covered at both low and high ranges.

The radiation monitors provide information to operating personnel of radiation above a preset level so that there will be sufficient time to evacuate the facility or take action to prevent the release of radioectivity to the surroundings.

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3.5 Reactor Safaty System Applicability This specification applies to the reactor safety system channels and interlocks.

Objective The objective is to require the minimum number of reactor safety system channels and interlocks that must be operable in order to assure that the safety limits are not exceeded, t Specification "

The reactor shall not be operated unless the safety system channels and Interlocks described in the following tables are operable.

.' Minimum Operating Mode Number i 6hich ,

Measurins Channel Operable Function

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Reactor Power Level 2 Scram Steady State Mode Peak Reactor Power i Sctsm Pulse Mode Manual Scram 1 Scram All Modes i

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hiinimum Operating htode Number in 'which Interlov: Operable Function Required Startup Count Rate  ! Prevent control Reactor Startup Interlock rod withdrawal when neutre- count rate is less than I per second Tansient Rod Interlock i Prevent withdraws! Steady State of a transient rod blode when its shock ab-sorber anvil is uot fully inserted Simultaneous Control Rod 1 Prevent simultaneous All blodes Wittdrawal Prohibit manual withdrawal of Interlock two control rods Reactor Power Level 1 Prevent transient Pulse blode Interlock rod withdraw 21 .

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Bases The power level scrams are provided in all modes of operation as protection against sonormally high fuel temperatures and to assure that the reactor operation stays within the licensed limits. The manual scram allows the operator to shut down the system if an unsafe i or abnormal condition occurs, l l l l

The interlocks which prevent the withdrawal of the transient rod in the steady state mode and l when the power level is greater than 10 kW prevent inadvertent pulses. The interlock to l prevent startup of the reactor with less than one peutron per second indicated on the startup channel assures that sufficient neutrons are available to provide indicators on the measuring channels and to provide negative reactivity feedback through the fuel temperature coefficient of reactivity.

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• 3.6 Ventilation System Applicability This specification applies to the operation of the reactor facility ventilation system.

Objective  !

t The objective is to assure that the ventilation system is in operation to mitigate the consequences of the possible release of radioactive materials resulting from reactor operation. i l

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The reactor shall not be operated unless the facility ventilation system is in operation except ,

for periods of time not to exceed two days to permit repairs to the system. During such [

periods of repair.

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a. The reactor shall not be operated at power levels above 10 kW and;

b. The reactor shall not be operated with experiments in place whose failure could result in the release of radioactive gases or aerosols.  !

Basis it is shown in The Safety Analysis Report that operation of the ventilation system reduces doses due to argon-41 during operation of the pneumatic transfer system, and also in the event of a TRIGA fuel element failure. The specifiestions governing operation of the reactor while the ventilation system is undergoing repairs limit the generation of argon-41 and also reduce the probability of fuel element failure during such times.

3.7 Experiments Applicacility ,

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

O)]ective The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.

Specifications The reactor shall not be operated unless the following conditions exist:

a. fueled experiments shall be limited such that the total inventory of lodine isotopes 131 through 135 in the experiment is not greater than 1.5 curies and the strontium-90 inventory is not greater than 5 millicuries;

b. experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials, or liquid fissionable materials shall be doubly encapsulated.

Basis The limits of Specification 3.7a prevent the dose in unrestricted areas resulting from experiment failure from exceeding 10 CFR Part 20 limits. Specification 3.7b is intended to reduce the prc5 ability of damage to reactor components resulting from experiment failure.

4.0 SURVE!LLANCE REQUIREMENTS 4.1 Fuel Applicability This specification applies to the surveillance requirements for the fuel elements.

Objective The objective is to assure that the dimensions of the fuel elements remain within acceptable limits.

Specifications

,s. The standard fuel elements shall be measured for length and bend at intervals separated by not more than 500 pulses of magnitude greater than 52.00 of reactivity,

b. A fuel element indicating an elongation greater than 1/4 of an inch over its original length or a lateral bending greater than 1/16 of an inch over its original bending shall be considered to be damaged and shall not be used in the core for further operation,

c. Fuel elements in the B- and C- rings shall be measured for possible distortion in the event that there is indication that the Limiting Safety System Settings may have been exceeded.

Basis The most severe stresses induced in the fuel elements result from pulse operation with high reactivity input, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above limits on the a!!owable distortion of a fuel element correspond to strains that are considerably lower than-the strain expected to cause rupture of a fuel element. ,

4.2 Control Rods Applicability This specification applies to the surveillance requirements for the control rods.

Objective l The objective is to assure the operability of the control rods.

Specification

a. The reactivity worth of each control rod sha!! be determined annuall",

b. Control rod drop times shall be determined annually and after disassembly of control rod drives or removal of cont.o! elemenu.

c. The control rods shall be visually inspected for deterioration biennially.

d. On each day that pulse mode operation of the reactor is planned, a functional performance check of the transient (pulse) rod system shall be performed prior to pulse mode operation, t

e. Semlannually, the transient (pulse) rod drive cylinder and the associated air supply system shall be inspected, cleaned and lubricated as necessary.

f. The maximum control rod reactivity insertion rates shall be determined annually.

Basis The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide a means for determining the reactivity worths of experiments insertM in the core. The visual inspection of the control rods and measurement of their drop times are made to determine whether the control rods are espable of performing their functions properly.

• 4.3 Reactor Safety System Applicability The specification applies to the surveillance requirements for the measuring channels of the reactor safety system.

Objective The objective is to assure that the safety system will remain operable and will prevent the fuel temperature safety limit from being exceeded, f,pecification

a. A channel test of each of the reactor safety system channels required in the operating mode to be followed shall be performed prior to each day's operation or prior to each operation extending more than one day,

b. A channel check of the power level measuring channels required in the operatina mode to be followed shall be performed daily whenever the reactor is in operation.

c. A channel calibration by the calorimetric method shall be performed for the reactor power level measuring channels annually.

Basis The daily tests and channel checks will assure that the safety channels are operable. The annual calibration and verifications will permit any long-term drift of the channels to be corrected.

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l 4.4 Radiation Monitoring Equipmg _

Applicability j This specification applies to the radiation monitoring equipment required by Section 3.4 of these specifkations.

4 Objective l i

The objective is to assure that the radia' ion monitoring equbment is operating and to verify .

the appropriate alarm settings.

Specification t i .

a. The alarm set points for the radiation monitoring instrumentation shall be verified prior l to each day's run. j

b. The radiation monitoring equipment shall be calibrated annually, Basis t Because of the redundancy of radiation monitoring instrumentation provided, periodic surveillance of the equipment will be adequate to assure that sufficient protection against radiation is available. I I

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4.5 Maintenance Appli ability This specification applies to the surveillance requirements following maintenance of a control or safety system, Object:ye The objective is to assure that a system is operable Sefore being used aftee maintenance has been performed.

Specificaticn Following maintenance cc. modification of a control or safety system or component, it shall be verified that the system is operable prior to its return to service.

Basis The intent of the specification is to assure that work on the system or component has been properly carried out and that the system or component has been properly reinstalled or reconnected.

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5.0 DESIGN FEATURES 5.1 Reactor Fuel Applicability This specification applies to the fuel elements used in the reactor core.

Objective The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degre* of reliability with respect to their mechanical integrity.

Specifications *

a. Standard Fuel Element ~

The standard fuel element shall be of the TRIGA type and shall contain uranium zirconium hydride, clad in 0.000 inch of 304 stainless steel. It shall contain a maximum of 9.0 weight percent uranium which has a maximum enrichment of 20 percent. There shall be 1.55 to 1.80 hydrogen atoms to 1.0 zirconium atom,

b. Loading: The elements shall be placed in a closely packed array except for experimental facilities or for positions occupied by control rods, elemeats fully loaded with graphite, a neutron startup source, or single positions within the array filled with water.

Basis This type of fuel element has a long history of successful use in TRIGA reactors.

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e 5.2 Reactor Building Applicability This specification applies to the facility which houses the reactor.

Objective The objective is to assure that provisions are made to restrict the radioactivity released into the environment.

Specifications

a. The reactor shall be housed in a closed room of a facility designed to restrict leakage,

b. The free volume of the reactor room shall be at least 6,000 cubic feet.

c. All air or other gases exhausted from the reactor room during reactor operation shall be released at a minimum of 12 feet above ground level,

d. The reactor facility shall be equipped with a ventilation system capable of exhausting air or other gases from the reactor room from a stack at a minimum of 50 feet above ground level under emergency conditions.

Basis In order that the movement of air can be controlled, the facility contains no windows that can be opened. Under emergency conditions the room air is exhausted thcough an absolute filter and discharged through a stack at a minimum of 50 feet above ground to provide dilution.

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5.3 Fuel 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 fuct which is being stored will not become supercritical and will not reach unsafe temperatures.-

Specifications

a. All fuel elements sha!! be stored in a geometrical array where the value of k rt is less than 0.9 for all conditions of moderation and reflection using light water.

b. Irradiated fuel elemens and fueled devices shall be stored in an array which will permit sufficient natural convwtion cooling by water or air such that the fuel element or fueled device temperstn e will not exceed 800*C.

Basis New fuel is stored in shipping t ctsiners or in a 30-position rack or in a 13-position holster in the reactor pool. Used fuel is stored in the same 30-position rack or 13-position holster in the reactor pool. These provisions have been shown by experience to be sufficient to prev 9nt inadvertent criticality or overheating of the fuci.

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6.0 ADMINISTRATIVE CONTROLS 6.1 Organization a.

The reactor facility shall be an integral part of the Nuclear and Energy E;gineering Department of the Co!!ege of Engineering and Mines at the University of Arizona as shown in the diagram below, b.

The reactor facility shall be under the supervision of a licensed senior operator for the reactor. He shall be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license and the provisions of the Reactor Committee.

c.

There shall be a Health Physicist responsible for assuring the safety of reactor operations from the standpoint of radiation protection. +

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An NRC licensed operator must be present when the key switch is on. An operator and one other person authorized by the Reactor Supervisor must be present whenever the reactor is not shut down.

COLLEGE of ENGINEERING andMINES RADIATION CONTROL DEAN DIRECTOR HEAD of NUCLEAR and ENERGY ENGINEERING RADIATK N CEPARTMENT CONTROL 0}TICE REACTOR REACTOR REACTOR Y HEALTH SUPERYlSOR COMMITTEE PHYSICIST REACTOR OPERATIONS

6.2 Review

a. There shall be a Reactor Committee which shall review reactor operations to assure that the facility is operated in a manner consistent with public safety and within the terms of the facility license,

b. The responsibility of the Committee includes, but is not limited to, the followirg:

1. Review and approval of experiments utilidng the reactor facilities; i

2. Review and approval of all proposed changes to the facility, procedures, and Technical Specifications;

3. Determination of whether a proposed change, test, or experiment would constitute an unrevicwed safety question or a change in the Technical Specifications as required by 10CFR 50.59, and review and approval of required safety analyses;

4. Review of the operation and operational records of the facility;

5. Review of abnormal performance of plant equipment and operating anomalles; and

6. Review of unusual or abnormal occurrences and incidents which are reportable ,

under 10 CFR 20 and 10 CFR 50,

7. Review and audit of the retraining and requalification program for the operating staff.

8. Biennial audit of the Emergency Plan,

c. The Committee shall be com;xned of at least five members, one of whom shall be a liealth Physicist. The membership of the Committee shall be such as to maintain a degree of technical proficiency similar to that of the present Reactor Committee for the University of Arizona TRIGA blark I reactor.

d. The Committee shall establish a written charter defining such matters as tue authority  ;

of the Committee, review and audit functions, and other such administrative provisions as are required for effective functioning of the Committee, hiinutes of all meetings of the Committee shall be kept and submitted to committee members and to the liend of

the Depatment of Nuclear and Energy Engineering in a timely manner.

e. A quorum of the Committee shall consist of not less than three members of the '

Committee and shall include the chairman or his designee.

f. The Committee shall meet at least quarterly, t

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6.3.1 Operating Procedures Written procedures, reviewed and approved by the Reactor Committee, shall be in effect and followed for the following items. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use of independent judgment and action should the situation require such.

a. Startup, operation, and shutdown of the reactor,

b. Installation or removal of fuel elements, control rods, experiments, and experimental facilities.

c. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary coolant system leaks, and abnormal reactivity changes.

d. Emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, re-entry, recovery, and medical supnort.

e. Maintenance procedures which could have an effect on reactor safety,

f. Periodic surveillance of reactor instrumentation and safety system, area monitors and continuous air monitors.

Substantive changes to the above procedures shall be made only with the approval of the Reactor Committee. Temporary changes to the procedures that do not change their original intent may be made with the approval of the Reactor Laboratory Director. All such temporary changes to procedures shall be documented and subsequently reviewed by the Reactor Committee.

6.3.2 ALARA Program A program shall be established to assure that radiation exposures and releases are kept as low as reasonably achievable.

6.4 Action to be Taken in the Event a Safety Limit is Exceeded in the event a safety limit is exceeded, or thought to have been exceeded:

a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

b. An immediate report of the occurrence shall be made to the Chairman of the Reactor Committee and reports shall be made to the NRC in accordance with Section 6.7 of these specifications,

c. A report shall be made which shall include an analysis of the causes and extent of possible resultant damage, efficiency of corrective action, and recommendations for measures to prevent or reduce the probability of recurretace. This report shall be submitted to the Reactor Committee for review, and a similar report submitted to the NRC when authorization to resume operation of the reactor is requested. i.

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6.5 Action to be Taken in the Event of a Reportable Occurrence in the event of a Reportable Occurrence, the following action shall be takere

a. The Reactor Laboratory Director shall be notified and corrective action taken prior to resumption ot' the operation involved,

b. A report shall be made which shall include an analysis of the cause of the occurrence, efficiency of corrective action and recommendations for measures to prevent or reduce the probability of reoccurrence. This report shall be submitted to the Reactor Committee for review,

c. A report shall be submitted to the NRC in accordance with Section 6.7 of these specifications.

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6.6 Plant Operating Records In addition to the requirements of applicsble regulations, and in no way substituting therefor, records and logs of the following items sha!! be prepared and retained for a period of at least 5 years (except as otherwise spe:ified in the Commission's regulations);

2. Normal plant operation (but not including supporting documents such as checklists, and recorder charts, which shall be maintained for a period of at least one year);

b. Principal maintenance activities;

c. Reportable Occurrences;

d. Equipment and component surveillance activities required by the Technical Specification;

e. Experiments performed with the reactor; Logs and records of the following items shall be prepared and retained for the life of the facility,

f. Gaseous and liquid radioactive effluents released to the environs; 3 Off-site ertvironmental monitoring sune>1;

h. Fuel inventories and transfers;

1. Facility radiation and contamination surve>1; J. Radiation exposures for all personnel; and

k. Updated, corrected, and as built drawings of the facility, i

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6.7 Reporting Requirements in addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the NRC as follow 1:

a. A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph to the appropriate regional USNRC Office of Inspection and Enforcement of;

1. Any accidental offsite release of radioactivity above permissible limits, whether or not the release resulted in property damage, personal injury, or exposure;

2. Any violation of a Safety Limit; and

3. Any reportable occurrences as defined in Section 1.13 of these specifications in writing.

b. A report within ten dan (in writing to the appropriate regional USNRC Office of Inspection and Enforcement, with copies to the Director, Office of Inspection and Enforcement, USNRC, Washington, D. C. 20555 and to the Director, Office of Management and Program Analysis, USNRC, Washington, D. C. 20555) of:

1. Any significant variation of measured values from a corresponding predicted value of previously measured value of safety-connected operating characteristics occurring during operation of the reactor,

2. Incidents or conditions relating to operation of the facility which prevented or could have prevented the performance of engineered safety features as described in these specifications;

3. Any reportable occurrences as defined in Section 1.13 of these specifications; and

4. Any violation of a Safety Limit,

c. a report within 30 days (in writing to the appropriate regional USNRC Office of Inspection and Enforcement, with copies to the Director, Office of Inspection and Enforcement, USNRC, Washington, D. C. 20555 and to the Director, Office of Management and Program Analysis, USNRC, Washington, D. C. 20555) of:

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1. Any substantial variance from performance specifications contained in these specifications or in the Safety Analysis Rer> ort;

2. Any significant change in the transient or accident analysis as described in the Safety Analysis Report;

3. Any changes in facility organization; and

.f . Any obetrved inadequacies in the implementation of administrative or procedural controls.

d. A report within 60 days after completion of startup testing of the reactor (in writing to the appropriate regional USNRC Office of Inspection and Enforcement, with copies to the Director, Office of Inspection and Enforcement, USNRC, Washington, D. C. 20555) upon receipt of, or an amendment to the license authorizing an increase in reactor power level, describing the measured values of the operating conditions or characteristics of the reactor under the new conditions including:

1. An evaluation of facility performance to date in comparison with design predictions and specifications; and

2. A reassessment of the safety analysis submitted with the license appiiestion in light of measured operating chancteristics when such measurements indicate that there may be substantial variance from prior anal >1is,

e. An annual report within 60 days fo!!owing the 30th of June each year (in writing to the appropriate regional USNRC Office of Inspection and Eaforcement, with copies to the Director, Office of Inspection and Enforcement, USNRC, Washington, D. C. 20555) providing the following information:

1. A brief narntive summary of (1) operating experience (including experiments performed), (2) changes in facility design, performance characteristles, and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections;

2. Tabulation of the energy output (in megawatt da)1) of the reactor, amount of pulse operation, hours reactor was crities), and the cumulative total energy output since initial criticality;

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3. The number of emergency shutdowns and inadvertent scrams, including reasons therefore;

4. Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor, and the reasons for any corrective maintenance required;

5. A brief description including a summary of the safety evaluations of changes in the facility or in procedures and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 50;

6. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effeulve control of the licensee as measured at or prior to the point of such release or discharge; Liquid Waste (summarized on a monthly basis)

a. Radioactivity discharged duritis the reporting period.

(1) Total radioactivity released (in curies).

(2) The MPC used and the isotopic composition if greater than I x 10-7 microcuries/cc for fission and activation producu.

(3) Total radioactivity (in curies), released by nuclide, during the reporting period, based on representative isotopic analysis.

(4) Average concentrntbn at point of release (in microcuries/cc) during the reporting period.

b. Total volume (in ga!!ons) of ef0uent water (including diluent) during periods of release.

Gaseous Waste (summarized on a monthly basis)

a. Radioactivity dischar ed during the reporting period (in curies) for:

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(1) Gases.

(2) Particulates with half lives greater than eight days.

b. The MPC used and the estimated activity (in curies discharged during the reporting period, by nuclide, for all gases and particulates based on representative isotopic analysis.

Solid Waste

a. The total amount of solid waste packaged (in cubic feet).

b. The total activity involved (in curies).

c. The dates of transfer or shipment and disposition

7. A summary of radiation exposures received by facility personel and visitors, including dates and time of significant exposures, and a summary of the results of radiation and contamination surveys performtd within the facility; and

8. A description of any enviro . mental surveys performed outside the facility.

6.8 Review of Experiments

a. All proposed new experiments utilizing the reactor shall be evaluated by the experimenter and the Ructor Committee. The evaluation sha!! be reviewed by a licensed Senior Operator of the facility (and the Health Physicist when appropriate) to assure compliance with the provisions of the utilization license, the Technical Specifications,10 CFR 20, and the requirements of 10 CFR 50.59. If, in his judgment, the experiment meets with the above provisions and does not constitute a threat to the integrity of the reactor, he shall submit it to the Reactor Supervisor for scheduling or to i the Reseter Committee for final approval as indicated in Section 6.2 above. When pertinent, the evaluation shall include:

1. The reactivity worth of the experiment;

2. The integrity of the experiment, including the effects of chstges in temperature, pressure, or chemical composition;

3. Any physical or chemical interaction that could occur with the reactor components; and

4. Any radiation hazard that may result from the activation of materials or from external beams,

b. Prior to performing an experiment not previously performed in the reactor, it shall be reviewed and approsed in writing by the Reactor Committee. This review shall consider the following informatiore

1. The purpose of the experiment;

2. A procedure for the performance of the experiment; and

3. The evaluation approved by a licensed Senior Operator,

c. For the irradiation of materials, the applicant shall submit an "Irradiation Request

• to the Reactor Supervisor. This request shall contain information on the target material including the amount, chemical form, and packaging. For routine irradiations (which do not contain known explosive materials and which do not constitute a signifiesnt threat to the integrity of the reactor or to the safety of individuals) the approval for the Reactor Committee may be made by the Reactor Supervisor.

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d. In evaluating experiments, the following assumptions should be used

• r the purpose of determining whether failure of the experiment would cause the ap, 'oriste limits of 10 CFR 20 to be exceeded:

1. If the possibility exists that altborne concentrations of radioactive gases or aerosols may be released within the facility,100 percent of the gases or aerosols will e: cape;

2. If the effluent exhausts through a filter installation designed for greater than 99 percent efficiency for 0.3 micron particles,10% of the particulates will escape; and

3. For a material whose boiling point is above 130*F and where vapors formed by boiling this material could escape only through a column of water above the core, 10% of these vapors will esespe.

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