ML20078S688

Proposed Tech Specs Re Changes to Reflect Upgrades in Electronics & Instrumentation & Reorganization of University Administration

ML20078S688
Person / Time
Site:	University of New Mexico
Issue date:	12/20/1994
From:	NEW MEXICO, UNIV. OF, ALBUQUERQUE, NM
To:
Shared Package
ML20078S674	List: ML20078S673 ML20078S688
References
NUDOCS 9412290131
Download: ML20078S688 (35)

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APPENDIX A ,

j LICENSE NUMBER R-102

'ECHNICAL SPECIFICATIONS .-

FOR I

UNIVERSITY OF NEW MEXICO AGN-20lM REACTOR 1

, SERIAL NUMBER 112 i

DOCKET NUMBER 50-252 i

. REVISED DECEMBER 1994 1 I

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1.0 INTRODUCTION

. . . . . I 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS . . .5 2.1 Safety Limits . . . .. . .. .5 2.2 Limiting Safety System Settings . . .. 6 3.0 LIMITING CONDITIONS FOR OPERATION . . . .7 3.1 Reactor Core Parameters . . .7 3.2 Reactor Control and Safety Systems .8 3.3 Limitations on Experiments . . . 11 3.4 Radiation Monitoring, Control And Shielding .. .. 12 4.0 SURVEILLANCE REQUIREMENTS . . . 14 4.1 Reactivity Limits . . . 14 4.2 Control and Safety Systems 15 4.3 Reactor Structure 16 4.4 Radiation Monitoring and Control . 17 5.0 DESIGN FEATURES . . . 18 5.1 Reactor . . .. 18 5.2 Fuel Storage . . . . 18 5.3 Reactor Room and Reactor Control Room .. .. . 19 6.0 ADMINISTRATIVE CONTROLS . . . . .20 6.1 Organization . . .20 6.2 StafTQualifications . . 25 6.3 Training . . . 25 l 6.4 Reactor Safety Advisory Committee . . . .25 l 6.5 Approvals . . . . .27 6.6 Procedures . 27 6.7 Experiments . . .28 6.8 Safety Limit Violations . . . 28 l 6.9 Reporting Requirements . . 28 6.10 Record Retention . .. .32 Resised December 1994

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1.0 INTRODUCTION

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] The terms Safety Limit (SL), Limiting Safety System Setting (LSSS), and Limiting Conditions for

! Operation (LCO) are as defined in 10 CFR Part 5036.

I j l.1 Definitions i

j 1.1.1 Certified Operator - A certified operator is an individual authorized by the Nuclear

] Regulatory Commission (NRC) to carry out the duties and responsibilities associated with j operation of the reactor.

i l 1.1.2 Channel Calibration - A channel calibration is an adjustment of the channel such that its j output responds, within acceptable range and accuracy, to known values of the parameter - 1

which the channel measures. Calibration shall encompass the entire. channel, including i equipment, actuation, alarm, or trip.

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j 1.1.3 Channel Check - A channel check is a qua'litative verification of acceptable performance l by observation of channel behavior. This verification may include comparison of the channel with other independent channels or methods measuring the same variable.

l 1.1.4 Channel Test - A channel test is the introduction of a signal into the channel to verify that l it is operable.  :

l l 1.1.5 Control Rod - Any of the four moveable rods loaded with fuel which are manipulated by j the operator to change the reactivity of the reactor.

! 1.1.6 Coarse Control Rod - The control rod with a scram function which can be mechanically j withdrawn / inserted at two possible speeds (minimum 40 seconds full insertion time for .

j slow speed or minimum 20 seconds full insertion time for fast speed). l l 1.1.7 Excess Reactivity - The amount of reactivity above K,g = 1. This is the amount of

! reactivity that would exist if all control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical (K.a = 1) l 1.1.8 Experiment - An experiment is any of the following:

i l a. An activity utilizing the reactor system or its components or the neutrons or j radiation generated therein; i b. An evaluation or test of a reactor system operational, surveillance, or maintenance j technique;

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I 1.1.9 Experimental Facilities - Experimental facilities are those portions of the reactor assembly that are used for the introduction of experiments into or adjacent to the reactor core region or allow beams of radiation to exist outside the reactor shielding. Experimental facilities shall include the thermal column, glory hole, and access ports.

1.1.10 Explosive Material - Explosive material is any solid or liquid which is categorized as a Severe, Dangerous or Very Dangerous Explosion Hazard in " Dangerous Properties of l Industrial Materials" by N.I. Sax, third Ed. (1968), or is given an Identification of Reactivity (Stability) index of 2, 3, o; 4 by the National Fire Protection Association in its publication 704-M,1966. " Identification System for Fire Hazards of Materials," also enumerated in the " Handbook for Labordory Safety," 2nd Ed. (1971), published by The

! Chemical Rubber Co .

j 1.1.11 Fine Control Rod A low worth control rod used primarily to maintain an intended power level. Its position is normally varied manually.

j 1.1.12 Major Change - Any change in reactor configuration which affects the probability or

consequences of an event.

I 1.1.13 Measured Value - The measured value is the value of a parameter as it appears on the output of a measuring channel.

1.1.14 Measuring Channel - A measuring channel is the combination of sensor, lines, amplifiers, 2

and output devices which are connected for the purpose of measuring or responding to the j value of a process variabic.

1.1.15 Movable Experiment - A movable experiment is one which may be inserted, removed, or manipulated while the reactor is critical.

1.1.16 Operable - Operable means a component or system is capable of performing its intended function in its normal manner.

1.1.17 Operating - Operating means a component or system is performing its intended function in its normal manner.

1.1.18 Potential Reactivity Worth - The potential reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result ofintended or anticipated changes or credible malfunctions that alter experiment position or configuration.

1.1.19 Reactor Component - A reactor component is any apparatus, device, or material that is a normal part of the reactor assembly.

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1.1.20 Reactor Operation - Reactor operation is any condition wherein the reactor is not shutdown.

1.1.21 Reactor Safety System - The reactor safety system is that combination of safety channels and associated circuitry which forms an automatic protective system for the reactor or provides information which requires manual protective action be initiated.

1.1.22 Reactor Secured - The reactor shall be considered secured whenever:

a. either: 1. All safety and control rods are fully with-drawn from the core; or

2. The core fuse melts resulting in separation of the core.

and:

b. The reactor console key switch is in the "off' position and the key is removed from the console and under the control of a certified operator.

1.1.23 Removable Experiment - A removable experiment is any experiment, experimental facility, or component of an experiment, other than a permanently attached appurtenance to the reactor system, which can reasonably be anticipated to be moved one or more timc; during the life of the reactor, 1.1.24 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 producing radioisotopes.

1.1.25 Safety Channel- A safety channel is a measuring channel in the reactor safety system.

1.1.26 Safety Control Rod - One of two scrammable control rods which can be mechanically withdrawn / inserted at only one speed (minimum 35 seconds full insertion time).

1 1.1.27 Scram Time - The time for the control rods acting under gravity to change the reactor from a critical to a suberitical condition. b most cases, this is less than or equal to the time 4

it takes for the rod to fall from full-in to full-out position.

1.128 Secured Experiment - any experiment, or component of an experiment is deemed to be secured, or in a secured position, ifit is held in a stationary position relative to the reactor by mechanical means. The restraint shall exert sufficient force on the experiment to overcome the expected effects of hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment or which might arise as a result of credible malfunctions .

1.1.29 Shall, Should and May - The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" to denote permission--neither a requirement nor a recommendation.

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! e 1.1.30 Shutdown Margin - Shutdown margin shall mean the minimum shutdown reactivity l necessary to provide confidence that the reactor can be made subcritical by means of the i

control and safety systems starting from any permissible operating condition although the most reactive rod is in its most reactive condition, and that the reactor will remain ,

subcritical without further operator action.

l.1.31 Static Reactivity Worth - The static reactivity worth of an experiment is the value of the reactivity change measurable by calibrated control or regulating rod comparison methods between two defined terminal positions or configurations of the experiment. For removable experiments, the terminal positions are fully removed from the reactor and fully inserted

, or installed in the normal functioning or intended position. ,

,- 1.1.32 Surveillance Time - A surveillance time indicates the frequency of tests to demonstrate performance. Allowable surveillance intervals shall not exceed the following:

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a. Two-year (interval not to exceed 30 months)

b. Annual (interval not to exceed 15 months)

c. Semi-annual (interval not to exceed seven and one-half months) i d. Quarterly (interval not to exceed four months) j e. Monthly (interval not to exceed six weeks).

1 l 1.1.33 True Value - The true value is the actual value of a parameter J

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2.0 SAFETY LIMITS 'AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits Applicability This specification applies to the maximum core temperature during operation.

Objective ,

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, To assure that the integrity of the fuel material is maintained and that all fission products are  !

i retained in the core matrix. >

I j Specification i a. The maximum core temperature shall not exceed 200*C during operation.

I j Basis 1

l The polyethylene core material does not melt below 200*C and is expected to maintain its integrity j and retain essentially all of the fission products at temperatures below 200*C. The Hazards 1 Summary Report dated February 1962 submitted on Docket F-15 by Aerojet-General Nucleonics j (AGN) calculated a core average temperature rise of 0.44*C/ watt. The corresponding maximum

core temperature would be below 200 C thus assuring integrity of the core and retention of fission

products.

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! 2.2 Limiting Safety System Settings l

Applicability This specification applies to the parts of the reactor safety system which will limit maximum power and core temperature.

Objective To assure that automatic protective action is initiated to prevent a safety limit from being exceeded.

Specification

a. The safety channels shall initiate a reactor scram at the following limiting safety system settings:

Channel Condition LSSS Nuclear Safety #2 High Power 6 Watts Nuclear Safety #3 High Power 6 Watts

b. The polystyrene core thermal fuse melts when heated to a temperature of about 120*C resulting in core separation and a reactivity loss greater than 5% Ak/k.

Basis l Based on instrumentation response times and scram tests, the AGN Hazards Report concluded l that reactor periods in excess of 30-50 milliseconds would be adequately arrested by the scram i l system. Since the maximum available excess reactivity in the reactor is less than one dollar, the reactor cannot become prompt critical, and the corresponding shortest possible period is greater than 200 milliseconds. The high power LSSS of 6 watts in conjunction with automatic safety j systems, and the average temperature rise of 0.44*C/ watt, and/or manual scram capabilities will j assure that the safety limits will not be exceeded during normal operation or as a result of the most  :

I severe credible transient.

l In the event of failure of the reactor to scram, the selflimiting characteristics due to the high l negative temperature coefficient, and the melting of the thermal fuse at a temperature below 120*C will assure safe shutdown without exceeding a core temperature of 200 C (the Safety Limit).

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l 3.0 LIMITING CONDITIONS FOR OPER.ATION 3.1 Reactor Core Parameters Applicability This specification applies to the reactivity condition of the reactor and the reactivity worths of control rods and experiments.

Objective To assure that the reactor can be shut down at all times and that the safety limits will not be exceeded.

Specification

a. The available excess reactivity with all control and safety rods fully inserted and including the potential reactivity worth of all experiments shall not exceed 0.65% Ak/k.

b. The shutdown margin with the most reactive safety or control rod fully inserted shall be at least 1% Ak/k.

c. The reactivity worth of the control rods shall ensure subcriticality on the withdrawal of the coarse control rod or any one safety rod.

d. The excess reactivity with no experiments in the reactor and the control and safety rods fully inserted shall not exceed 0.25% Ak/k.  !

l Basis

! The limitations on total core excess reactivity assure reactor periods of sufficient length so that the reactor protection system and/or operator action will be able to shut the reactor down without exceeding any safety limits. The shutdown margin and control and safety rod reactivity limitations assure that the reactor can be brought and maintained subcritical if the highest reactivity rod fails to scram and remains in its most reactive position 7 Revised December 1994 l

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3.2 Reactor Control and Safety Systems Applicability These specifications apply to the reactor control and safety systems.

Objective To specify lowest acceptable level of performance, instrument set points, and the minimum number of operable components for the reactor control and safety systems.

Specification

a. The fine control rod, coarse control rod, and the two safety rods shall be operable and the carriage position of the fine and coarse control rods shall be displayed at the console whenever any rod is above its lower limit.

b. The total scram withdrawal time of the safety rods and coarse control rod shall be less than Isecond.

c. The average reactivity addition rate for each control or safety rod shall not exceed 0.065%

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d. The safety rods and coarse control rod shall be interlocked such that:

1. Reactor startup cannot commence unless both safety rods and the coarse control l rod are fully withdrawn from the core.

2. Only one safety rod can be inserted at a time.

3. The coarse control rod cannot be inserted unless both safety rods are fully inserted.

4. At any operating power above 50 x 104 Watts, none of the rods can be moved to a more reactive position.

e. Nuclear safety channel instrumentation shall be operable in accordance with Table 3.1 whenever the reactor control or safety rods are not in their fully withdrawn position.

f. A manual scram shall be provided on the reactor console, and the safety circuitry shall be designed so that no single failure can negate both the automatic and manual scram capability.

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! g. The shield water level interlock shall be set to prevent reactor startup and scram the reactor if the shield water level falls more than 18 cm below the highest point on the j reactor shield tank manhole opening.

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h. The shield water temperature interlock shall be set to prevent reactor startup and scram

! the reactor if the shield water temperature falls below 18*C.

l l i. The seismic displacement interlock shall be installed in such a manner to prevent reactor startup and scram the reactor during a seismic displacement.

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' A loss of electric power shall cause the reactor to scram.

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The specification on operability of the rods assure console control over reactivity conditions within j the reactor. Display of the positions of the fine and coarse control rods assures that the positions i of these rods be available to the operator to evaluate the configuration of the reactor.

l The specifications on scram withdrawal time in conjunction with the safety system instrumentation i and set points assure safe reactor shutdown during the most severe foreseeable transients.

Interlocks on control rods assure an orderly approach to criticality and an adequate shutdown capability. The limitations on reactivity addition rates allow only relatively slow increases of

, reactivity so that ample time will be available for manual or automatic scram during any operating j conditions.

i The neutron detector channels (Nuclear Safety Channels #1 through #3) assure that reactor power i levels are adequately monitored during reactor startup and operation. The power level scrams

! initiate redundant automatic protective action at power level scrams low enough to assure safe

shutdown without exceeding any safety limits. The manual scram assures a method of shutdown without reliance on safety channels and circuitry.

j The AGN-201's negative temperature coefficient of reactivity causes a reactivity increase with l decreasing core temperature. The shield water temperature interlock will prevent reactor operation j at temperatures below 18*C thereby limiting potential reactivity additions associated with j temperature decreases.

I j Water in the shield tank is an important component of the reactor shield and operation without the j water may produce excessive radiation levels. The shield tank water levelinterlock will prevent i reactor operation without adequate water levels in the shield tank.

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! The reactor is designed to withstand 0.6g accelerations and 6 cm displacements. A seismic l instrument causes a reactor scram whenever the instrument receives a horizontal acceleration that

causes a horizontal displacement of 0.16 cm or greater. The seismic displacement interlock assures

! that the reactor will be scrammed and brought to a subcritical configuration during any seismic l disturbance that may cause damage to the reactor or its components.

l l The manual scram allows the operator to manually shutdown the reactor if an unsafe or otherwise

! abnormal condition occurs that does not scram the reactor. A loss of electrical power de-energizes l the safety and coarse control rod holding magnets causing a reactor scram thus assuring safe and j immediate shutdown in case of a power outage.

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l Table 3.1 Nuclear Instrumentation j Channel No. Function Operating Limits i

1 Monitor None 2 High Power Scram 120% (6 Watts) oflicensed

power l 3 High Power Scram 120% (6 Watts) oflicensed i power I

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1 3.3 Limitations on Experiments j; )

l Applicability l

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This specification applies to experiments installed in the reactor and its experimental facilities. j r

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To prevent damage to the reactor or excessive release of radioactive materials in the event of an l experimental failure. )

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Specification -

i a. Expenments (within the reactivity limits defined in Specification 3.1) containing materials )

] corrosive to reactor components or which contain gaseous or liquid, fissionable materials  !

! shall be doubly encapsulated. l L

l b. Explosive materials or materials which might combine violently shall not be inserted into j experimental facilities of the reactor or irradiated in the reactor.

i c. The radioactive material content, including fission products, of any doubly encapsulated l l experiment should be limited so that the complete release of all gaseous, particulate, or 1' i volatile components from the encapsulation could not result in:

i (1) a dose in excess of the current 10CFR20 limits to any person occupying an i unrestricted area or I l (2) a dose in excess of the current 10CFR20 limits to any person occupying a

) restricted area.

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Basis I

l These specifications are intended to reduce the likelihood of damage to reactor components and 4

or radioactivity releases resulting from an experimental failure and to protect operating personnel i and the public from excessive rediation doses in the event of an experimental failure. Specification l 3.3c conforms to the regulatory position put forth in Regulatory Guide 2.2, issued November l 1973.

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l 3.4 Radiation Monitorina_. Control And Shieldmu .

1 Applicability This specification applies to radiation monitoring, control, and reactor shielding required during l reactor operation. )

Objective I

The objective is to protect facility personnel and the public from radiation exposure.

Specification

a. An operable portable radiation survey instrument capable of detecting gamma radiation shall be immediately available to reactor operating personnel whenever the reactor is not secured.

b. The reactor room shall be considered a restricted area whenever the reactor is not secured.

c. The reactor room shall be considered a radiation area whenever the reactor is operated.

d. Whenever the reactor is operated, the top of the reactor shall be considered a high radiation area. The access stairs to the top of the reactor shall be locked with the keys to the lock under the control of a Reactor Supervisor,

e. The following shielding requirement shall be fulfilled during reactor operation:

The thermal column shall be filled with water or graphite except during a critical experiment (core loading) or during other approved experiments which require the thermal column to be empty. ,

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f. The core tank shall be sealed during reactor operation.

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Basis Radiation surveys performed under the supervision of a qualified health physicist have shown that the total gamma, thermal neutron, and fast neutron radiation dose rate in the reactor room, at the closest approach to the reactor (but without access to reactor top), is less than 50 mrem /hr at reactor power levels of 5.0 watts .

The facility shielding in conjunction with radiation monitoring, control, and restricted areas is designed to limit radiation doses to facility personnel and to the public to a level below 10 CFR 20 limits under all conditions. The control of access to the top of the reactor is in compliance with 10CFR20.1601.a3.

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{ 4.0 SURVEILLANCE REQUIREMENTS 1

i Actions specified in this section are not required to be performed if during the specified j surveillance period the reactor has not been brought critical or is maintained in a secured condition

! extending beyond the specified surveillance period. However, the surveillance requirements must

) be fulfilled prior to subsequent startup of the reactor, i

j 4.1 Reactivity Limits

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s Applicability 1

! This specification applies to the surveillance requirements for reactivity limits.

i Objective

] To assure that reactivity limits for Specification 3.1 are not exceeded.  ;

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! Specification l a. Control rod reactivity worths shall be measured annually.

i l b. Total excess reactivity and shutdown margin shall be determined annually.

j' c. The reactivity worth of an experiment shall be estimated or measured, as j appropriate, before or during the first startup subsequent to the experiment's first j insertion.

I j Basis The control and safety rod reactivity worths are measured annually to assure that no degradation i or unexpected changes have occurred which could adversely affect reactor shutdown margin or l total excess reactivity. The shutdown margin and total excess reactivity are determined to assure j that the reactor can always be safely shut down with one rod not functioning and that the

maximum possible reactivity insertion will not result in reactor periods sharter than those that can j be adequately termmated by either operator or automatic action. Based on experience with AGN reactors, significant changes in reactivity or rod worth are not expected within a 12 month period.

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r 4.2 Control and Safety Systems Applicability This specification applies to the surveillance requirements of the reactor control and safety systems.

Objective To assure that the reactor control and safety systems are operable as required by Specification 3.2.

Specification

a. Control rod scram times and average reactivity insertion rates shall be measured annually.

b. Control rods and drives shall be inspected for proper operation annually.

c. A channel test of Nuclear Safety Channels #2 and #3 shall be performed prior to the first reactor startup of the day or prior to each reactor operation extending more than one day.

d. A channel test of the seismic displacement interlock shall be performed annually.

e. A channel check of Nuclear Safety Channels #2 and #3 shall be performed daily whenever the reactor is in operation.

f. Prior to each day's reactor operation or prior to each reactor operation extending ,

more than one day, safety rod #1 shall be inserted and scrammed to verify operability of the manual scram system.

g. The power level measuring channels shall be calibrated and set points verified annually.

h. The shield water level interlock and shield water temperature interlock shall be caEbrated annually.

Basis The channel tests and checks required daily or before each startup will assure that the safety channels and scram functions are operable. Based on operating experience with reactors of this type, the annual scram measurements, channel calibrations, set point verifications, and inspections are of sufficient frequency to assure, with a high degree of confidence, that the safety system l settings will be within acceptable drift tolerance for operation.

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1 1 4.3 Reactor Structure i

Applicability 4

q This specification applies to surveillance requirements for reactor components other than control j rods.

i l Objective l The objective is to assure integrity of the reactor structures.

Specification l Visual inspection for water leakage from the shield tank shall be performed annually. Leakage sufficient to leave a puddle on the floor shall be corrected prior to subsequent reactor operation.

! Basis i

i Based on experience with reactors of this type, the frequency of inspection and leak test requirements of the shield tank will assure capability for radiation protection during reactor operation.

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4.4 Radiation Monitoring and Control Applicability ,

This specification applies to the surveillance requirements of the radiation monitoring and control systems.

Objective To assure that the radiation monitoring and control systems are operable and that all radiation and high radiation areas within the reactor facility are identified and controlled as required by Specification 3.4.

Specification

a. All portable radiation survey instruments assigned to the reactor facility shall be calibrated under the supervision of the Radiological Safety Office annually. The reactor area Remote Area Monitors shall be calibrated annually. These may be calibrated with the internal ' check source if one is present.

b. Prior to each day's reactor operation or prior to each reactor operation extending more than one day, the reactor access high radiation area alarm (Ref 3.4d) and reactor area Remote Area Monitors shall be verified to be operable.

c. A radiation survey of the reactor room shall be performed under the supervision of the Radiological Safety Ollicer annually to determine the location of radiation and high radiation areas corresponding to reactor operating power levels.

Basis The periodic calibration of radiation monitoring equipment and the surveillance of the reactor access high radiation area alarm (Ref 3.4d) will assure that the radiation monitoring and control systems are operable during reactor operation.

The periodic radiation surveys will verify the location ofradiation and high radiation areas and will assist reactor facility personnel in properly labeling and controlling each location in accordance with 10 CFR 20.

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l 5.0 DESIGN FEATURES 5.1 Rfactor

a. The reactor core, including control rods, contains approximately 667 grams of U-235 in the form ofnominal 19.5% enriched UO2 dispersed in approximately 1I kilograms of polyethylene. The lower section of the core is supported by an aluminum rod hanging from a fuse link. The fuse melts at a fuse temperature of about 120"C causing the lower core section to fall away from the upper section reducing reactivity by at least 5% Ak/k. Suflicient clearance between core and reflector is provided to ensure free fall of the bottom half of the core during the most severe transient.

b. The core is surrounded by a 20 cm thick high density (1.75 gm/cm') graphite reflector followed by a 10 cm thick lead gamma shield. The core and part of the graphite reflector are sealed in a fluid-tight aluminum core tank designed to contain any fission gases that might leak from the core.

c. The core, reflector and lead shielding are enclosed in and supported by a fluid-tight l steel reactor tank. An upper or " thermal column tank" may serve as a shield tank when filled with water or a thermal column when filled with graphite.

i d. The 198 cm diameter, fluid-tight shield tank is filled with water constituting a 55 cm thick fast neutron shield. The fast neutron shield is formed by filling the tank with approximately 3785 liters of water. The complete reactor shield shall limit doses to personnelin unrestricted areas to levels less than permitted by 10 CFR 20 under operating conditions.

e. Two safety rods and one control rod (identical in size) contain less than 15 grams of U-235 each in the same form as the core material. These rods are lifted into the core by electromagnets, driven by reversible DC motors through lead screw assemblies. De-energizing the magnets causes a spring-driven, gravity-assisted scram. The fourth rod or fine control rod (approximately one-half the diameter of the other rods) is driven directly by a lead screw. This rod may contain polyethylene with or without fuel.

5.2 Fuel Storage Fuel, including fueled experiments and fuel devices not in the reactor, shall be stored in Room

1. 077 (source storage room in the Nuclear Engineering Laboratory) which shall be locked at all times. The storage array shall be such that K,g is no greater than 0.9 for all conditions of moderation and reflection.

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. l 5.3 Reactor Room and Reagtor Controls i

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a. The reactor room houses the reactor assembly and accessories required for its operation and maintenance, and the reactor control console.

b. The reactor room is a separate room in the Nuclear Engineering Laboratory, constructed with adequate shielding and other radiation protective features to limit doses in restricted ,

and unrestricted areas to levels no greater than permitted by 10 CFR 20.  :

c. The access doors to the reactor room shall contain locks.

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6.0 ADMINISTRATIVE CONTROLS 6.1 Organization The administrative organization for control of the reactor facility and its operation shall be as set forth in Figure 1 attached hereto. The authorities and responsibilities set forth below are designed to comply with the intent and requirements for administrative ccntrols of the reactor facility as set forth by the Nuclear Regulatory Commission.

6.l.1 President The President is the chief administrative officer responsible for the University and in whose name the application for licensing is made.

6.1.2 Dean, College ofEngineering The Dean of Engineering is the administrative officer responsible for the operation of the College ofEngineering.

6.l.3 Reactorildministrator The Reactor Administrator provides fmal policy decisions on all phases of reactor operation and regulations for the facility. He/she is selected by the Chairman of the Chemical and Nuclear Engineering Department and shall hold a graduate degree in Engineering. He/she is advised on matters concerning personnel health and safety by the Radiological Safety Officer and/or the Radiation Control Committee. He/she is advised on matters concerning safe operation of the reactor by the Reactor Operations Committee and/or the Reactor Safety Advisory Committee.  ;

He/she designates Reactor Supervisors and names the Chief Reactor Supervisor. He/she approves 1 all regulations, instructions and procedures governing facility operation. He/she submits the annual <

report to NRC. He/she is responsible for control of and changes to the code on the cipher locks of the Nuclear Engineering Laboratory Building at the beginning of each semester.

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6.1.4 Radiation ControlCommittee l The Radiation Control Committee (RCC)is a permanent committee established to act on behalf of the President of the University for control of all University of New hiexico (UNN1) activities invoking sources ofionizing radiation. The Committee consists of at least five members from the

, UNh1 faculty / staff appointed for two year terms. hieetings are held at least quarterly (interval not to exceed four months). Responsibilities are (1) to establish policy and disseminate mies for radiological safety and control at UNht, (2) to serve as UNhi's liaison with the NRC in matters of registration, licensing, and radiological control, (3) to ensure periodic inspections and radiation surveys for the purpose of assuring the safety of radiological operations within any UNhi facility.

6.1.5 RadiologicalSafety Officer The Radiological Safety Oflicer is an ex-officio member of the Radiation Control Committee and normally represents the RCC in matters concerning the radiological safety aspects of reactor operation. He/she is available for advice and assistance on radiological safety problems. He/she is the Emergency Director if an emergency involves radiation safety.

6,1.6 Reactor Safety Advisory Committee Reviews, evaluates, and audits reactor operations and procedures to ensure that the reactor shall be operated in a safe and competent manner. There shall be at least four members on the RS AC with at least two members from organizations outside the University. The Committee is available for advice and assistance on reactor operation problems. They must approve any major change in the facility.

6.l.7 Reactor Operations Committee Consists of the Reactor Supervisors with the Chief Reactor Supervisor. Other qualified persons may also be members. They are directly responsible to the Reactor Administrator for the preparation and submission of detailed procedures, regulations, forms, and rules to ensure the maintenance, safe operation, competent use and security of the facility. The Committee ensures  ;

that all the activities, experiments, and maintenance involving the facility are properly logged and I are in accordance with established local and U.S. Nuclear Regulatory Commission regulations. l

~

They review all proposed changes in procedure or changes in the facility and must approve any minor change before the change is implemented.

21 Revised December 1994

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6.1.8 ChiefReactor Supervisor Holds a Senior Reactor Operator's license issued by the NRC. He/she is responsible for the distribution and enforcement of rules, regulations and procedures concerning operation of the l facility. He/she has the authority to authorize any experiments or procedures which have received appropriate prior approval by the Reactor Operations Committee, the Reactor Safety Advisory Committee and/or the Radiation Control Committee (or the Radiological Safety OfIicer) and have received prior authorization by the Reactor Administrator. He/she shall not authorize any proposed changes in the facility or in procedure until appropriate evaluation and approval has been made by the Reactor Operations Committee or the Reactor Safety Advisory Committee and authorization given by the Reactor Administrator. The Chief Reactor Supervisor is directly responsible for enforcing operating procedures and ensuring that the facility is operated in a safe, competent and authorized manner. He/she is directly responsible for all prescribed logs and records. He/she is the Emergency Director for emergencies not involving radiation.

6.1.9 Reactor Supervisors 1

Shall hold valid Senior Reactor Operator's licenses issued by the Nuclear Regulatory Commission.

A Reactor Supervisor shall be in charge of the facility at all times during reactor operation and must witness the startup and intentional shutdown procedures. The Reactor Supervisors are {

directly responsible to the Chief Reactor Supervisor. The Reactor Supervisors do not have to be l present other than when the reactor is going critical, being intentionally shut down, or when reactor experiments are loaded or unloaded. However, the location of the Reactor Supervisor must be known to the Reactor Operator at all times during operation so that it is possible to contact him/her if required.

6.l.10 Reactor Operators l Must hold a valid Reactor Operator's license issued by the NRC. They must conform to the rules, instructions and procedures for the startup, operation and shutdown of the reactor, including emergency procedures. Within the constraints of the administrative and supervisory controls outlined above, a reactor operator will be in direct charge of the control console at all times that l the reactor is operating. The reactor operator is required to maintain complete and accurate records of all reactor operations in the operational logs. ,

l 6.1.11 Authori:edOperators These are individuals who are authorized by the Reactor Supervisor to operate the reactor controls and who do so with the knowledge of the Supervisor and under the direct supervision of a Reactor Operator.

22 Revised December 1994

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6.1.12 Reactor Assistants i

j These are individuals who are present during reactor operation to provide assistance to the Operator as needed, with the exception that a Reactor Assistant does not operate the controls of the reactor. In an emergency they may push the Reactor Scram button.

6.1.13 OperatingStaff l I

a. The minimum operating staff during any time in which the reactor is not secured shall >

l consist of all of the following:

l One Reactor Operator in the reactor control room.

r One other person in the reactor room or Nuclear Reactor Laboratory qualified to activate ' i manual scram and initiate emergency procedures. l One health physicist who can be readily contacted by telephone and who can arrive at the reactor facility within 30 minutes. <

N One Senior Reactor Operator readily available on call. This requirement can be satisfied -

by having a licensed Senior Reactor Operator perform the duties stated in paragraph 1 or 2 above or by designating a licensed Senior Reactor Operator who can be readily contacted by telephone and who can arrive at the reactor facility within 30 minutes.

b. A Senior Reactor Operator shall supervise all reactor maintenance or modification which

• could affect the reactivity of the reactor.

c. A listing of reactor facility personnel by name and phone number shall be conspicuously l posted in the reactor control room.

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j 23 Revised December 1994 l

I

, PRESIDENT i OFTHE

, WIVERSITY 1

1 1

, VICE PRESIDBU PROVOST /VICE ASSOC. PROVCCT i FOR FRESIDENT FOR FOR SYUDENT AFFAIRS ACADD4IC AFFAIRS RESEARCH l

j l I L l DEAN, DIRECTOROF RAIMION FCLICE DEPARTMaiT COILEGE OF CONTROL DIGHTEERRIG COM)RTTEE l

i I j ._

CHAIRMAN, RADIGILGICAL CHEMICAL AND SAFEIY OFFICER NUCLEAR (for Radiation Emergency)

, B10lNEERR40 DEPT.

l l

l REACTOR NUCLEAR j ADMINISTRATOR ENGINEERDIO j LABORATORY j SUPERVISOR

., REACTOR

SAFEOUARDS I 1 ADVISORY

' l

! COMMITTEE CHIEF REACTOR RADIATION i SUPERVIOOR SUPERVISORS e

i l I REACTOR NUCLEAR SUPERVISORS ENGINEERDIO j STUDa4TS j

REACTOR OPERATORS i

)

Figure 1

Administration i

24 Revised December 1994

__ _ . _ . _ ._ ._= _ _~ .

l 6.2 StaffOualifications 1 i The Chief Reactor Supervisor, licensed Reactor Operators, and technicians performing reactor i

maintenance shall meet the minimum qualifications set forth in ANSI 15.4, " Standards for l Selection and Training of Personnel for Research Reactors". Reactor Safety Advisory Committee l members shall have a minimum of five (5) years experience in a technical profession or a '

baccalaureate degree and two (2) years of professional experience. The Radiological Safety Officer shall have a baccalaureate degree in biological or physical science and have at least two (2) years l experience in health physics.

l l 6.3 Training I

l The Reactor Administrator shall be responsible for directing training as set forth in-ANSI 15.4-1977, " Standards for Selection and Training of Personnel for Research Reactors". All licensed reactor operators shall participate in requalification training as set forth in 10 CFR 55.

6.4 Reactor Safety Advisory Committee 6.4.1 Meetings and Quorum The Reactor Safety Advisory Committee shall meet as often as deemed necessary by the Reactor Safety Advisory Committee Chainnan but shall meet at least semiannually (interval not to exceed 7 months). A quorum for the conduct of official business shall be three members.

6.4.2 Reviews The Reactor Safety Advisory Committee shall review:

a. Safety evaluations for changes to procedures, equipment or systems, and tests or experiments, conducted without Nuclear Regulatory Commission approval under the provision of 10 CFR 50 paragraph 50.59 to verify that such actions do not constitute an unreviewed safety question.

b. Proposed changes to procedures, equipment or systems that change the original intent or ,

use, and are non-conservative, or those that involve an unreviewed safety question as '

defined in 10 CFR 50 paragraph 50.59. j

c. Proposed tests or experiments which are significantly different from previous approved l tests or experiments, or those that involve an unreviewed safety question as defined in 10 CFR 50 paragraph 50.59.

1 25 Revised December 1994 l -

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d. Proposed changes in Technical Specifications or licenscs.

l

e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

f. Significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect nuclear safety.

g. Reportable occurrences.

h. Audit reports.

6.4.3 Audits Audits of facility activities shall be performed at least annually (interval not to exceed 15 months)

under the cogmzance of the Reactor Safety Advisory Committee but in no case by the personnel i responsible for the item audited. These audits shall examine the operating records and encompass, l but shall not be limited to, the following

1

a. The conformance of the facility operation to the Technical Specifications and applicable license conditions, at least annually (interval not to exceed 15 months).

b. The Facility Emergency Plan and implementing procedures, at least every two years (interval not to exceed 30 months).

c. The Facility Security Plan and implementing procedurer, at least every two yeau (interval not to exceed 30 months).

d. Operator requalification program and records, at least every two years (interval not to exceed 30 months).

e. Results of actions taken to correct deficiencies, at least annually (interval not to exceed 15 months).

6.4.4 Authority l l

The Reactor Safety Advisory Committee shall report to the Reactor Administrator and shall advise i the Chief Reactor Supervisor on those areas of responsibility outlined in Section 6.1.5 of these Technical Specifications.

26 Revised December 1994

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6.4.5 Minutes of the Reactor Safety Advisory Committee i'

One member of the Reactor Safety Advisory Committee shall be designated to direct the preparation, maintenance, and distribution of minutes ofits activities. These minutes shall include a summary of all meetings, actions taken, audits, and reviews. Minutes shall be distributed to all

. RSAC members, all administrative levels, and the Radiation Safety Officer within 6 months (interval not to exceed 7 months) after each meeting. i 6.5 Approvals l I l The procedure for obtaining approval for any change, modification, or procedure which requires I approval of the Reactor Safety Advisory Committee shall be as follows:

i ,

i a. The Chief Reactor Supervisor shall prepare the proposal for review and approval by the Reactor Administrator.

b. The Reactor Administrator shall submit the proposal to the Reactor Safety Advisory ,.

Committee for review, comment, and possible approval.

c. The Reactor Safety Advisory Committee can approve the proposal by majority vote.

l . d. The Reactor Administrator will provide final approval after receiving the approval of the Reactor Safety Advisory Committee.

6.6 Procedures There shall be written procedures that cover the following activities:

l

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

I

b. Fuel movement and changes to the core and experiments that could affect reactivity.

i

c. Conduct ofirradiations and experiments that could affect the operation or safety of the .

reactor.

d. Preventive or corrective maintenance which could affect the safety of the reactor. i

e. Routine reactor maintenance. -

f. Radiation Safety Protection far WI reactor related personnel.

g. Surveillance, testing and calibration ofi.rtruments, components, and systems as specified i in Section 4.0 of these Technical Srecifications.

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27 Revised December 1994

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h. Implementation of the Security Plan and Emergency Plan.

l The above listed procedures shall be approved by the Reactor Administrator and the Reactor l Safety Advisory Committee. Temporary procedures which do not change the intent of previously

! approved procedures and which do not involve safety question may be employed on approval by j the Chief Reactor Supervisor.

l 6.7 Experiments i a.. Prior to initiating any new reactor experiment, an experimental procedure shall be prepared i by the Chief Reactor Supervisor and reviewed and approved by the Reactor Safety j Advisory Committee.

i i b. Experiments shall only be performed under the cognizance of the Chief Reactor j Supervisor.

6.8 Safety Limit Violations

! The following actions shall be taken in the event a Safety Limit is violated:

i j a. The reactor will be shut down immediately and reactor operation will not be resumed

without authorization by the Nuclear Regulatory Commission (NRC). i 1

1

! b. The Safety Limit Violation shall be reported to the NRC Regional Administrator of Region i IV, the Director of NRR, the Reactor Safety Advisory Committee, and Reactor

! Administrator not later than the next work day. l j

j c. A Safety Limit Violation Report shall be prepared for review by the Reactor Safety i Advisory Committee. This report shall describe the applicable circumstances preceding the

{ violation, the effects of the violation upon facility components, systems, or structures, and

} corrective action to prevent recurrence.

1 i d. The Safety Limit Violation Report shall be submitted to the NRC and the Reactor Safety Advisory Committee within 14 days of the violation.

l I

i 6.9 Reoortina Reauirements f

In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the i following reports shall be submitted to the Director, Office of Nuclear Reactor Regulation, USNRC, Washington D.C.,20555, Attention: Document Control Desk; with a copy to Regional Administrator, Region IV.

i

\

i 28 Revised December 1994 1

i 6.9.1 Annual Operating Report Routine annual operating reports shall be submitted no later than ninety (90) days following June

30. The annual operating reports shall provide a comprehensive summary of the operating l

experience having safety significance that was gained during the year, even though some repetition of previously reported information may be involved. References in the annual operating report to previously submitted reports shall be clear.

Each annual operating report shall include: )

a. A brief narrative summary of

1. Changes in facility design, performance characteristics, and operating procedures related to reactor safety that occurred during the reporting period.

i

2. Results of major surveillance tests and inspections.

1 i b. A tabulation showing the hours the reactor was operated and the energy produced by the reactor in watt-hours.

, c. List of the unscheduled shutdowns, including the reasons therefore and corrective action taken, 1

ifany.

d. Discussion of the major safety related corrective maintenance performed during the period, including the effects, if any, on the safe operation of the reactor, and the reasons for the corrective maintenance required.

1

e. A brief description of:

1. Each change to the facility to the extent that it changes a description of the facility in the application for license and amendments thereto.

i

2. Changes to the procedures as described in Facility Technical Specifications.

! 3. Any new experiments or tests perfonned during the reporting period.

j f. A summary of the safety evaluation made for each change, test or experiment not submitted for NRC approval pursuant to 10 CFR 50, paragraph 50.59 which clearly shows the reason leading to the conclusion that no unreviewed safety question existed and that no Technical Specifications l change was required. I 1

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29 Revised December 1994 i! ,

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g. A summary of the nature and amount of radioactive efIluent released or discharged to the environs beyond the effective control of the licensee as determined at or prior to the point of such release or discharge.

1. Liquid Waste (summarized for each release)

a. Total estimated quantity of radioactivity released (in Curies) and total volume (in liters) of effluent water (including diluent) released.

2. Solid Waste (summarized for each release)

a. Total volume of solid waste packaged (in cubic meters)

b. Total activity in solid waste (in Curies) l c. The dates of shipment and disposition (if shipped off site).

h. A description of the results of any environmental radiological surveys performed outside the facility.

i. Radiation Exposure - A summary of personnel exposures received during the reporting period by facility personnel and visitors similar to the report described in the applicable section of 10 CFR 20.

6.9.2 Reportable Occurrences Reportable occurrences, including causes, probable consequences, corrective actions and measures to prevent recurrence, shall be reported to the NRC as described in Section 6.9. Supplemental reports may be required to fully describe final resolution of the occurrence. In case of corrected or supplemental reports, an amended licensee event report shall be completed and reference shall be made to the original report date.

a. Prompt Notification with Written Follow-up The types of events listed below are considered reportable occurrences and shall be

, reported as expeditiously as possible by telephone and confirmed by telegraph, mailgram,

or facsimile transmission to the Administrator of the NRC Region IV Office, or his designated representative no later than the first work day following the event, with a written follow-up report within two weeks as described in Section 6.9. Information provided shall contain narrative material to provide complete explanation of the circumstances surrounding the event.

30 Revised December 1994

r l 1. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored

parameter reached the setpoint specified as the limiting safety system setting in the l Technical Specifications or failure to complete the required protective function.

2. Operation of the reactor or affected systems when any parameter or operation

( subject to a limiting condition is less conservative than the limiting condition for operation established in the Technical Specifications - without taking permitted remedial action.

3. Abnormal degradation discovered in a fission product barrier.

4. Reactivity balance anomalies involving:

a. Disagreement between expected and actual critical rod positions of approximately 0.3% Ak/k.

b. Exceeding excess reactivity limit.

c. Shutdown margin less conservative than specified in Technical i Specifications.

i

d. If sub-critical, an unplanned reactivity insertion of more than approximately t 0.5% Ak/k or any unplanned criticality.

5. Failure or malfunction of one (or more) component (s) which prevents or could l prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the Safety Analysis Report.

6. Personnel error or procedural inadequacy which prevents, could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the Safety Analysis Report.

7. Unscheduled conditions arising from natural or man-made events that, as a direct l result of the event, require reactor shutdown, operation of safety systems, or other l protective measures required by Technical Specifications.

8. Errors discovered in the analyses or in the methods used for such analyses as described in the Safety Analysis Report or in the bases for the Technical l Specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analysis.

l 31 Revised December 1994

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j e .. e l

9. Release of radiation or radioactive materials from site above allowed limits.

l

10. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed ~in the accident analysis in the Safety Analysis Report or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

6.9.3 Special Reports Special reports which may be required by the Nuclear Regulatory Commission shall be submitted to the Administrator of the NRC Region IV Office within the time period specified for each report.

This includes changes in Level 1,2 or 3 administration, as shown in Figure 1, which must be  !

reported within 30 days of such a change.

6.10 Record Retention 6.10.1 Records to be Retained for a Period of at Least Five Years ,

l

a. Operating logs or data which shall identify:

1. Completion ofpre-startup check-out, startup, power changes, and shutdown of the i reactor.

2. Installation or removal of fuel elements, control rods, or experiments that could affect core reactivity.

3. Installation or removal ofjumpers, special tags or notices, or other temporary changes to reactor safety circuitry.  !

4.' Rod worth measurements and other reactivity measurements.

b. Principal maintenance operations.

c. Reportable occurrences.

d. Surveillance activities required by Technical Specifications.

e. Facility radiation and contamination surveys.

32 Revised December 1994

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, f. Experiments performed with the reactor. This requirement may be satisfied by the normal l operations log book plus, l

1. Records of radioactive material transferred from the facility as required by license.

2. Records required by the Reactor Safety Advisory Committee for the performance  ;

ofnew or special experiments.

g. Records of training and qualification for members of the facility staff.

h. Changes to operating procedures.

6.10.2 Records to be Retained for the Life of the Facility -

a. Records ofliquid and solid radioactive effluent released to the environs.

b. Off-site environmental monitoring surveys.

c. Fuelinventories and fuel transfers.

d. Radiation exposures for all personnel.

e. Drawings of the facility.

f. Records of operational cycles for those components designed for a limited number of cycles.

g. Recod of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

l h. Records of meetings of the Reactor Safety Advisory Committee, and copies of RSAC

! audit reports. l l

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33 Revised December 1994

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