ML20211Q409

Proposed Tech Specs for Tx Engineering Experiment Station Tx A&M Univ Sys Nuclear Science Ctr Reactor Facility,Revised Through Amend 15

ML20211Q409
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Site:	05000128
Issue date:	09/07/1999
From:	TEXAS A&M UNIV., COLLEGE STATION, TX
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ML20211Q406	List: ML20211Q400 ML20211Q409
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NUDOCS 9909150059
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TECHNICAL SPECIFICATIONS

.FOR THE TEXAS ENGINEERING EXPERIMENT STATION TEXAS A&M UNIVERSITY SYSTEM NUCLEAR SCIENCE CENTER REACTOR FACILITY DOCKET NO. 50-128 LICENSE NO R-83 i MARCH 1983 REVISED THROUGH AMENDMENT NO.15 i i

9909150059 990907 PDR ADOCK 05000128 P PDR

TECHNICAL SPECIFICATIONS FOR THE NUCLEAR SCIENCE CENTER REACTOR FACILITY LICENSE NO. R-83 March 1983 Revised through Amendment No.15 TABLE OF CONTENTS IMEs 1.0 Definitions.. .. . . . . . .. .1 1.1 Abnormal Occurrence.. .. .. . .. . . .1 1.2 ALARA.. .. . . . . .. .1 1.3 Channel.. . . . . . . .1 1.3.1 Channel Test.. . .. . .I 1.3.2 Channel Calibration.. . . .. .. . .1 1.3.3 Channel Check . .I 1.4 Confinement. . .. . . . . . .1 1.5 Core Lattice Position . . . .2 1.6 Experiment.. . .. .. . . . . . . . .2 1.7 Experimental Facilities.. . . . . . . . . . . . . . . 2 1.8 Experiment Safety Systems . .. . . . .. .2 1.9 FLIP Core ... . . . . _ . . . . . . .2 1.10 Fuel Bundle.. . ... .. . . . . . . . . . . .2 1.11 Fuel Element.. . . . . . . .. . . .2 1.12 Instrumented Element.. . .. . . . .2 1.13 Limiting Safety System Setting. . . . . . . . . .. . . . . . . . . .2 1.14 Measuring Channel., . . . . . .. . .2 1.15 Measured Value.. .. . . .. ..3 1.16 Mixed Core.... . . . . . .. . .. 3 1.17 Movable Experiment .. . . . . . . . .3 1.18 Operable.. . . . . . . . . .3 1.19 Operating. . . . . .3 4 1.20 Steady State Operational Core.. , .. . . . . .. . . 'l l.21 Pulse Operational Core.. . . . . . . . . . . .3 1.22 Pulse Mode.. . . . . . . . . .. . ..3 1.23 Reactivity Worth of an Experiment.. . .3 1.24 Reactor Console Secured. . . . . . . .3 1.25 Reactor Operating.. .. . - , . . . . .3 1.26 Reactor Safety Systems . . . . . .4 1.27 Reactor Secured.. . . .. .4 1.28 Reactor Shutdown.. .. . . .4 1.29 Reportable Occurrence . .. .. ..4 1.30 Rod-Control .. . . . . . .5 1.31 Rod-Regulating.. . ... . .5 1.32 Rod-Shim Safety. .. . .. .5 1.33 Rod-Transient.. . . . . . . .. . . .5 1.34 Safety Channel.. . . . . .5 1.35 Safety Limit. . .5 1.36 Scram Time.. .. ... . . . . . . =5 1.37 Secured Experiment.... . . .5

1.38 Shall, Should and May... . . . . . . . . . . .. . .5 1.39 Shutdown Margin . . . .. .. .. . . . .6 1.40 Standard Core . .. . . . . . .. .6 1.41 Steady State Mode..... ... . . . . ..6 1.42 True Value.. . . ... . . . . . . . . . .6 1.43 Unscheduled Shutdown. . . . . ..6 2.0 Safety Limit and Limiting Safety System Setting.. .. .. . .. . .7 2.1 Safety Limit Fuel Element Temperature.. . .7 2.2 Limiting Safety System Setting.. . .. . . . . .7 3.0 Limiting Conditions for Operation.. . . . . . . . .9 3.1 Reactor Core Parameters . . . . .. .. .9 3.1.1 Steady State Operation .. .. .. .. .9 3.1.2 Pulse Mode Operation.. .. . . . . . . . ..9 3.1.3 Shutdown Margin.. . . . . . .10 3.1.4 Core Configuration Limitation.. . . . . .. . .11 3.1.5 Maximum Excess Reactivity... . . .. . . ..I1 3.2 Reactor Control and Safety Systems.. . . . . . . . . . .12 3.2.1 Reactor Control Systems.. . . . . .. . .12 3.2.2 Reactor Safety Systems.. . . . .. .12 3.2.3 Scram Time . . . . . . . . .. . .. . 14 3.3 Confinement . . . . . . . . . . . . . . . . 14 3.3.1 Operations that Require Confinement.. . . . . .. . 14 3.3.2 Equipment to Achieve Confinement. . . . . . .. . .15 3.4 Ventilation System... . . . . . 15 3.5 Radiation Monitoring Systems and Effluents.. . . . . . .. . . . .16 3.5.1 Radiation Monitoring . . . . . . . .16 3.5.2 Argon-41 Discharge Limit . . . . . . .. . . .17 3.5.3 Xenon and Iodine Monitoring.. .. . . . , . . . . . 17 3.6 Limitations on Experiments.. . . . .. . .. . . . 18 3.6.1 Reactivity Limits. . . . . .. . . . . .18 3.6.2 Material Limitations. . . . . . . . . . .18 i' 3.6.3 Failures and Malfunctions. .. . . . . . . . . 20 3.6.4 Xenon Irradiation for lodine Production... . . . .. . 21 3.7 As Low As Reasonably Achievable (ALARA) Radioactive Efiluents Released . . 21 3.8 Primary Coolant Conditions . . . . . . .. .. . . 22  ;

i 4.0 Sun'eillance Requirements.. . . .. . . . . . . . . . . . . . . . . .. . 24 1 4.1 General.... .. .. . . . . . . . . . . . . . 24 4.2 Reactor Core Parameters . .. .. . . 24 4.2>l Steady State Operation ... . . . . . . . . . , . .. .. .. . . 24 4.2.2 Pulse Mode Operation.. . ... . .. . . .. . . 25 4 4.2.3 Shutdown Margin.. . .. . . . 25 4.2.4 Reactor Fuel Elements.. . . . . . . . 25 4.3 Reactor Control And Safety Systems . .. .. . . 27 4.3.1 Reactor Control Systems. . . . . . 27 4.3.2 Reactor Safety Systems.. . .. . . . 27 4.3.3 Scram Time . . . .. . . 28 4.4 Equipment to Achieve Confinement: Ventilation System.. . . 28 4.5 Radiation Monitoring Systems and Efiluents.. . . . 29 4.6 Experiments.. . . . . . . . 29 5.0 Design Features.. . .. . . . .. . . . . .31 l 5.1 Reactor Fuel.. . . .. . . .. . . . . 31 l

5.2 Reacar Core .. . . . .. . . .. . . .32 5.3 Control Rods.. . . . . . .. . . . . . . . .. . . . 32

5.4 Radiation Monitt ,g System.. . . . 33 j j

5.5 Fuel Storage... . .. .. .. . _34 5.6 - Reactor Building and Ventilation System ... . .. ... ... .. . 34 5.7 Reactor Pool Water Systems-- . . 35 5.8 Physical Security - .. . ... . .36 6.0 Administrative Controls - .37 6.1 Organization . . .. . . . . .. . . 37 6.1.I' Structure . .. .. .. . . =37 6.1.2 ' Responsibility.. .. .. . . . . . _ . . . .. 37 6.1.3 Staffing -- -. 37 6.1.4 Selection and Training of Personnel _ .. . .. . . 38 6.2 Review and Audit.. - 39 6.2.1 Reactor Safety Board = -39 6.2.2 RSB Charter and Rules.. . - 39 6.2.3 . RSB Review Function.. . . 39 6.2.4 RSB Audit Function = .40 6.2.5 Audit of ALARA Program - . - 40 6.3 Operating Procedures-- =40 C.4 Experiments Review and Approval.. . .. =41-6.5 Required Actions .. ... . 41 6.5.1 Action to be Taken in the Event a Safety Limit is Exceeded .. . . .41 6.5.2 Action to be Taken in the Event of a Reportable Occurrence.. .._... . . 41 6.6 Reporting Requirements ; . .. . . 42 6.6.1 Annual Report ., . .. :42 6.6.2 Special Reports _ - . . . . . . . ... 43 6.7 Records.. . - 44 6.7.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved.; .. .. . . 44 6.7.2 Records to be Retained for at Least One Training Cycle -- 44 6.7.3 Records to be Retained for the Lifetime of the Reactor Facility.. . 44 1

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I NSC Technical Specifications, Version 09/01/1999 l

TECHNICAL SPECIFICATIONS FOR THE NUCLEAR SCIENCE CENTER REACTOR l FACILITY LICENSE NO. R-83 Included in this document are the Technical Specifications and the " Bases" for the Technical Specifications. These bases, which provide the technical support for the individual technical specifications, are included for informational i purposes only. They are not part of Technical Specifications and they do not constitute limitations or requirements  !

to which the licensee must adhere.

1.0 Definitions l l

1.1 Abnormal Occmrence An " Abnormal Occurrence" is defined, for the purposes for the reporting requirements of Section I 208 of the Energy Reorganization Act of 1974 (P.L.93-438) as an unscheduled incident or event  !

which the Nuclear Regulatory Commission determines is significant from the standpoint of public I health or safety. j 1.2 ALARA l The ALARA program (As Low As Reasonably Achievable)is a program for maintaining occupational exposures to radiation and release of radioactive effluents to the environs as low as reasonably achievable.

1.3 Channel l

A channel is the combination of sensors, lines, amplifiers, and output devices, which are ,

I connected for the purpose of measuring the value of a parameter.

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1.3.1 Channel Test A channel test is the introduction of a signal into the channel for verification that it is operable. i 1

1.3.2 Channel Calibration A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. l Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a channel test.

1.3.3 Channel Check A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable.

1.4 Confinement Confinement means a closure of the overall facility which controls the movement of air into it and l out through a controlled path. j 1

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NSC Technical Specifications, Version 09/01/1999 1.5 Core Lattice Position The core lattice position is that region in the core (approximately 3" x 3") over a grid plug hole. It may be occupied by a fuel bundle, an experiment, or a reflector element.

1.6 Experiment An operation, hardware, or target (excluding devices such as detectors, foils etc.) which is designed to investigate non-routine reactor characteristics or which its intended for irradiation within the pool, on or in a beam port or irradiation facility and which is not rigidly secured to a core or shield structure so as to be a part of their design.

1.7 Exprimental Facilities  :

Experimental facilities shall mean beam ports, including extension tubes with shields, thennal columns with shields, vertical tubes, through tubes, in-core irradiation baskets, irradiation cell, pneumatic transfer systems and in-poolirradiation facilities.

1.8 Experiment Safety Systems Experiment safety systems are those systems, including their associated input circuits, which are designed to initiate a scram for the primary pmpose of protecting an experiment or to provide information which requires manual protective action to be initiated.

1.9 FLIP Core A FLIP core is an arrangement of TRIGA-FLIP fuel in a reactor grid plate.

1.10 Fuel Bundle A fuel bundle is a cluster of two, three or four elements and/or non-fueled elements secured in a square array by a top handle and a bottom grid plate adapter. Non-fueled elements shall be fabricated from stainless steel, aluminum or graphite materials.

1.11 Fuel Element A fuel element is a single TRIGA fuel rod of either standard or FLIP type.

1.12 Instrumented Element An instrumented element is a special fuel element in which a sheathed chromal-alumel or equivalent thermocouple is embedded in the fuel near the horizontal center plane of the fuel element at a point approximately 0.3 inch from the center of the fuel body.

1.13 Limiting Safety System Setting The limiting safety system setting is the setting for automatic protective devices related to those variables having significant safety functions. l 1.14 Measuring Channel i A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device which are connected for the purpose of measuring the value of a variable.

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NSC Technical Specifications, Version 09/01/1999 I 1.15 Measured V.lue The measured value is the value of a parameter as it appears on the output of a channel.

1.16 Mixed Core A nUxed core is an arrangement of standard TRIGA fuel elements with at least 35 TRIGA-FLIP fuel elements located in a central contiguous region of the core.

l 1.17 Movable Experiment A movable experiment is one for which it is intended that the entire experiment may be moved in I or near the core or into and out of the reactor while the reactor is operating.

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

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

1.20 Steady State Operational Core A steady state operational core shall be a standard core, mixed core, or FLIP core for which the I core parameters of shutdown margin, fuel temperature and power calibration have been determined.

1.21 Pulse Operational Core A pulse operational core is a steady state operational core for which the maximum allowable pulse  !

reactivity insertion has been determined.

1.22 Pulse Mode Pulse mode operation shall mean any operation of the reactor with the mode selector switch in the pulse position.

1.23 Reactivity Worth of an Experiment The reactivity worth of an experiment is the maximum absolute value of the reactivity change that l would occur as a result ofintended or anticipated changes or credible malfunctions that alter the l experiment position or configuration. )

1.24 Reactor Console Secured I

The reactor console is secured whenever all scrammable rods have been fully inserted and verified down and the console key has been removed from the console.

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

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e NSC Technical Specifications, Version 09/01/1999 1.26 Reactor Safety Systems Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action. Manual protective action is considered part of the reactor safety system.

1.27 Reactor Secured A reactor is secured when:

a) It contains insufficient fissile material or moderator present in the reactor and adjacent experiments to attain criticality under optimum available conditions of moderation and reflection, or b) The reactor console is secured, and

1) 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

2) No experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum value of one dollar.

1.28 Reactor Shutdown The reactor is shut down when the reactor, at ambient temperature and xenon-free condition and  ;

including the reactivity worth of all experiments, is suberitical by at least one dollar.

1.29 Reportable Occurrence A reportable occurrence is any of the following which occurs during reactor operation:

a) Operation with actual safety system settings for required systems less conservative than the limiting safety-system settings specified in the Technical Specifications 2.2.

b) Operation in violation oflimiting conditions for operation established in the technical specifications.

c) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdowns. (Note: Where components or systems are provided in addition to those required by the technical specifications, the failure of the extra components or systems is not considered reportable provided that the minimum number of components or systems specified or required perform their intended reactor safety function.)

d) An unanticipated or uncontrolled change in reactivity greater than one dollar.

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

f) An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.

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NSC Technical Specifications, Version 09/01/1999 1.30 Rod-Control 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.

1.31 Rod-Regulating The regulating rod is a low worth control rod used primarily to maintain an intended power level that need not have scram capability and may have a fueled follower. Its position may be varied manually or by the servo-controller.

1.32 Rod-Shim Safety A shim-safety rod is a control rod having an electric motor drive and scram capabilities. It may have a fueled follower section.

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

1.34 Safety Channel i

A safety channel is a measuring channel in the reactor safety system.

1.35 Safety Limit Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioactivity.

1.36 Scram Time Scram time is the time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable control rods reaches its fully inserted position.

1.37 Secured Experiment A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially grester than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating c: vironment of the experiment, or by forces which can arise as a result of credible malfunctions.

1.38 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 denate permission, neither a requirement nor a recommendation. In order to conform to this standard, the user shall conform to its requirements but not necessarily to its recommendations.

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NSC Technical Specifications, Version 09/01/1999 1.39 Shutdown Margin Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by means of the control and safety systems starting from any permissible operating condition, if the most reactive rod is stuck in its most reactive position, and that the reactor will remain suberitical without further operator action.

1.40 Standard Core A standard core is an arrangement of standard TRIGA fuel in the reactor grid plate.

1.41 Steady State Mode Steady state mode operation shall mean operation af the reactor with the mode selector switch in the steady state position.

1.42 True Value The true value is the actual value of a parameter.

1.43 Unscheduled Shutdown An unscheduled shutdown is defined as any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operation, not to include shutdowns which occur during testing or check out operations.

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NSC Technical Specifications, Version 09/01/1999 2.0 Safety Limit and Limiting Safety System Setting 2.1 Safety Limit Fuel Eleinent Temperature Apolicability This specification applies to the temperature of the reactor fuel.

Obiective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding will result.

Snecifications a) The temperature in a TRIGA-FLIP fuel element shall not exceed 2100 F (1150 C) under any conditions of operation.

b) The temperature in a standard TRIGA fuel element shall not exceed 1830 F (1000 C) under any conditions of operation.

Bases The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well suited as a single specification especially since it can be measured. A loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of the hydrogen to zirconium in the alloy.

He safety limit for the TRIGA-FLIP fuel element is based on data which indicate that the stress i in the cladding due to the hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided the temperature of the fuel does not exceed 2100*F (1150 C) and the fuel cladding is water cooled.

The safety limit for the standard TRIGA fuel is based on data, including the large mass of experimental evidence obtained during high performance reactor tests on this fuel. These data j indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1830*F (1000*C) and the fuel cladding is water cooled.

2.2 Limiting Safety System Setting Annlicability This specification applies to the scram setting which prevents the safety limit from being reached.

Obiective The objective is to prevent the safety limits from being reached.

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NSC Technical Specifications, Version 09/01/1999 Snecification The limiting safety system setting shall be 975"F (525*C) as measured in an instrumented fuel  ;

element. The instrumented element shall be located adjacent to the central bundle with the exception of the corner positions.

D.alis He limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. A peak core temperature of 950*C in FLIP fuel and 800*C in standard fuel is the criteria established to provide a minimum safety margin of 200 C for all modes of operation. A part of this margin is used to acccunt for the difference between the maximum and measured temperatures resulting from the actual location of the thermocouple. If the thermocouple element were located in the hottest position in the core, the difference between the true and measured temperatures would be only a few degrees since the thermocouple junction is at the mid-plane of the element and close to the anticipated hot spot.

liowever, this position is normally not available due to the location of the transient rod. He location of the instrumented elements is therefore restricted to the positions closest to the central element. Calculations indicate that, for this case, the true temperature at the hottest location in the core will differ from the measured temperature by no more than 40%. Thus, for the steady state mode of operation when the temperature in the thermocouple element reached the trip setting of

$25*C, the true temperature at the hottest location in a standard core would be no greater than 632*C and 690 C in a mixed core, providing a safety margin of at least 368"C for standard fuel elements and 460*C for FLIP type elements. These margins are ample to account for the remaining uncertainty in the accuracy of the fuel temperature measurement channel and any )

overshoot in reactor power resulting from a reactor transient during steady state mode operation.

In the pulse mode of operation, the same limiting safety system setting will apply. Ilowever, the temperature channel will have no effect on limiting peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). In this mode, however, the temperature trip will act to reduce the amount of energy generated in the entire pulse transient by cutting the " tail" of the energy transient in the event the pulse rod remains I stuck in the fully withdrawn position. l 1

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l NSC Technical Specifications, Version 09/01/1999 3.0 Limiting Conditions for Operation r

3.1 Reactor Core Parameters 3.1.1 Steady State Operation Anolicability This specification applies to the energy generated in the reactor during steady state operation.

Obiective The objective is to assure that the fuel temperature safety limit will not be exceeded during steady state operation.

Snecifications The reactor power level shall not exceed 1.3 megawatts under any condition of operation.

The normal steady state operating power level of the reactor shall be 1.0 megawatts.

Ilowever, for purposes of testing and calibration, the reactor may be operated at higher power levels not to exceed 1.3 megawatts during the testing period.

D31it Thermal and hydraulic calculations indicate the TRIGA fuel may be safely operated up to

. power levels of at least 2.0 MW with natural convection cooling.

3.1.2 Pulse Mode Operation Annlicability This specification applies to the peak temperature generated in the fuel as the result of a pulse insertion of reactivity.

Obiective The objective is to assure that respective pulsing will not induce damage tot he reactor fuel.

Snecification a) The reactivity to be insened for pulse operation shall not exceed that amount which will produce a peak fuel temperature of 1526*F (830 C). In the pulse mode the pulse rod shall be limited by mechanical means or the rod extension physically shortened so that the reactivity insenion will not inadvertently exceed the maximum value.

b) Until the full FLIP fuel core has been calibrated, maximum pulse shall be limited to $2.00.  ;

Dalit TRIGA fuel is fabricated with a nominal hydrogen to zirconium ratio of 1.6 for FLIP fuel l and 1.65 for standard. This yields delta phase zirconium hydride which has a high creep i

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NSC Technical Specifications, Version 09/01/1999 strength and undergoes no phase changes at temperatures over 1000 C. Ilowever, after extensive steady state operation at 1 MW the hydrogen will redistribute due to migration from the central high temperature regions of the fuel to the cooler outer regions. When the fuel is pulsed, the instantaneous temperature distribution is such that the highest valuer occur at the surface of the element and the lowest values occur at the center. The higher temperatures in the outer regions occur in fuel whh a hydrogen to zirconium ratio that has now substantially increased above the nominal value. This produces hydrogen gas pressures considerably in excess of the expected for Zrlin. If the pulse insertion is such that the temperature of the fuel exceeds 874 C, then the pressure will be sufficient to cause expansion of microscopic holes in the fuel that grows with each pulse. The pulsing limit of 830*C is obtained by examining the equilibrium hydrogen pressure of zirconium hydride as a function of temperature. The decrease in temperature from 874*C to 830"C reduces hydrogen pressure by a factor of two, which is an acceptable safety factor. This phenomenon does not alter the safety limit since the total hydrogen in a fuel element does not change. Thus, the pressure exerted on the clad will not be significantly affected by the distribution of hydrogen within the element.

In practice, the pulsing limit of 830*C will be translated to a reactivity insertion limit for each specific core. The peaking factors from the thermocouple element to the hottest spot in the core must be calculated for each core configuration that is to be used.

Temperature would then be measured for small pulse insertions.

For new uncalibrated cores, the pulse insertions shall be increased by small increments to

'a maximum of $2.00 to allow an extrapolation of peak temperatures, thereby establishing the maximum allowed pulse insertion for a given core. Following approval by the NRC staff of the calibration of the new core, the $2.00 restriction shall be removed.

3.1.3 Shutdown Margin Aeolicability These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods and experiments. They apply for all modes of operation.

Obiective The objective is to assure that the reactor can be shutdown at all times and to assure that the fuel temperature safety limit will not be exceeded.

Specifications The reactor shall not be operated unless the shutdown margin provided by control rods is greater than 50.25 with:

a) The highest worth non-secured experiment in its most reactive state, J b) The highest worth control rod and the regulating rod (if not scrammable) fully withdrawn, and I c) The reactor in the cold condition without xenon.

D.alia The value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the highest worth control rod should remain in the fully withdrawn position, if the regulating rod is not scrammable, its worth is not used in detennining the shutdown reactivity.

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NSC Technical Specifications, Version 09/01/1999 3.1.4 Core Configuration Limitation Aeolicability This specification applies to mixed cores of FLIP and standard types of fuel and to full FLIP cores.

Obiective ne objective is to assure that the fuel temperature safety limit will not be exceeded due to power peaking effects in mixed cores and FLIP cores.

Soecifications a) The TRIGA core assembly may be standard, FLIP, or a combination thereof (mixed core) provided that any FLIP fuel core be comprised of at least thirty-five (35) fuel elements, located in a contiguous, central region.

b) The reactor shall not be taken critical with a core lattice position vacant except for positions on the periphery of the core assembly. Water holes in the inner fuel region shall be limited to single rod positions. Vacant core positions shall contain experiments or an experimental facility to prevent accidental fuel additions to the reactor core, c) The instmmented element shall be located adjacent to the central bundle with the exception of the corner positions (

Reference:

2.2 Limiting Safety System Setting).

Bases a) In mixed cores, it is necessary to specify the minimum number of FLIP elements and arrange them in a contiguous, central region of the core to control flux peaking and power generation values in individual elements.

b) Vacant core positions containing experiments or an experimental facility will l prevent accidental fuel additions to the reactor core. They will be permitted only on the periphery of the core or a single rod position to prevent power peaking in regions of high power density, c)

Reference:

2.2 Limiting Safety System Setting.

3.1.5 Maximum Excess Reactivity Aoolicability l

his specification applies to the maximum excess reactivity, above cold critical, which may be loaded into the reactor core at any time.

Obiective The objective is to e asure that the core analyzed in the safety analysis report approximates the operational core within reasonable limits.

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NSC Technical Specifications, Version 09/01/1999 Specifications The maximum reactivity in excess of cold, xenon-free critical shall not exceed 5.5% Ak/k

($7.85).

Euis Although maintaining a minimum shutdown margin at all times ensures that the reactor can be shut down, that specification does not address the total reactivity available within the core. This specification, although over-constraining the reactor system, helps ensure that the licensee's operational power densities, fuel temperatures, and temperature peaks are maintained within the evaluated safety limits. The specified excess reactivity allows for power coefficients of reactivity, xenon poisoning, most experiments, and operational flexibility.

3.2 Reactor Control and Safety Systems 3.2.1 Reactor Control Systems Aonlicability This specification applies to the information which must be available to the reactor operator during reactor operation.

Obiectis e The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor.

Soecifications j The reactor shall not be operated unless the measuring channels listed in the following table are operable.

EITective Afode Afeasuring Channel Afin. No. Operable S S. Pulse Fuel Element Temperature 1 X X Linear Power Level I X Log Power Level 1 X Integrated Pulse Power I X UaEl Fuel temperature displayed at the control console gives continuous information on this parameter, which has a specified safety limit. The power level monitors assure that the reactor power level is adequately monitored for both steady state and pulsing modes of operation. The specifications on reactor power level indication are included in this section, since the power level is related to the fuel temperature.

3.2.2 Reactor Safety Systems Aeolicability l

l This specification applies to the reactor safety system circuits.

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NSC Technical Specifications, Version 09/01/1999 Obiective The objective is to specify the minimum number of reactor safety system channels that must be operable for safe operation.

Snecifications The reactor shall not be operated unless the safety circuits described in the following table are operable.

Number Effective Mode Saferv Channel Operable Function S.S. Pulse 1 SCRAM @ LSSS X X Fuel Element Temperature 2 SCRAM @ 125% X liigh Power Level 1 SCRAM X X Console Scram Button liigh Power Level Detector 2 SCRAM on loss of supply voltage X j Power Supply  !

1 Transient rod scram 15 seconds or less X Preset Timer after pulse 1 Prevent withdrawal of shim safeties at X Log Power <4 x 10-5 W ,

1 Prevent Pulsing above 1 kW X l Log Power 1 Prevent application of air unless fully X Transient Rod position inserted Shim Safeties & Regulating i Prevent withdrawal X Rod Position Bases The fuel temperature and power level scrams provide protection to assure that the reactor can be shutdown before the safety limit on the fuel element temperature will be exceeded. The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs. In the event of failure of the power supply for a safety chamber, operation of the reactor without adequate instrumentation is prevented. The preset timer insures that the reactor power level will reduce to a low level after pulsing.

The interlock to prevent startup of the reactor at power levels less than 4 x 10-3 W which corresponds to approximately 2 cps assures that sufficient neutrons are available for proper startup.

The interlock to prevent the initiation of a pulse above I kW is to assure that the magnitude of the pulse will not cause the fuel element temperature safety limits to be exceeded. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted is to prevent pulsing of the reactor in steady state mode. The interlock to prevent the withdrawal of the shim safeties or regulating rod in the pulse mode is to prevent the reactor from being pulsed while on a positive period.

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NSC Technical Specifications, Version 09/01/1999 3.2.3 Scram Time Anolicability This specification applies to the time required for the scrammable control rods to be fully inserted from the instant that the fuel temperature safety channel variable reaches the Limiting Safety System Setting.

Obiective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.

Specification The scram time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable control rod reaches its fully inserted l position shall not exceed 1.2 seconds.

l Balis This specification assures that the reactor will be promptly shutdown when a scram ,

signal is initiated. Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.

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3.3 Confinement 3.3.1 Operations that Require Confinement Aonlicability This specification applies to confinement requirements during operation of the reactor I and the handling of radioactive materials.

Obiective To maintain normal or emergency air flow into and out of the reactor building during operations that produce or could potentially produce airborne radioactivity.

Specification Confinement of the reactor building will be required during the following operations.*

a) Reactor operating.

b) Handling of radioactive materials with the potential for airborne release.

i For periods of time for maintenance to the central exhaust fan, entry doors to the l reactor building will remain closed except for momentary opening for personnel  ;

entry or exit.

Bases a) This basis applies during the conduct of those activities defmed as reactor operations. Argon-41 is produced during operation of the reactor in experimental facilities and in the reactor pool; thus, air control within the 14

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building and the exhaust system in necessary to maintain proper airborne radiation levels in the reactor building and release levels in the exhaust stack.

Other radioactivity releases to the reactor building must be considered during reactor operation, such as fission product release from a leaking fuel element or a release from fixed experiments in or near the core.

b) The handling of radioactive materials can result in the accidental or controlled release of airborne radioactivity to the reactor building environment or direct release to the building exhaust system. In these cases, the control of air into a id out of the reactor building is necessary.

3.3.2 Equipment to Achieve Confinement Apolicability This specification applies to the equipment and controls needed to provide confinement of the reactor building.

Obiective The objective is to assure that a minimum of equipment is in operation to achieve confinement as specified in 3.3.1 and that the control panel for this equipment is available for normal and emergency situations.

Specifications a) The minimum equipment required to be in operation to achieve confinement of the reactor building shall be the central exhaust fan.*

b) Controls for establishing the operation of the ventilation system during normal and emergency conditions shall be located in the reception room.

For periods of time for maintenance to the central exhaust fan, entry doors to the reactor building will remain closed to assure closure except for the momentary opening for personnel entry or exit.

Bases a) Operation oithe central exhaust fan will achieve confinement of the reactor building during normal and emergency conditions when the controls for air input are set such that the central exhaust fan capacity remains greater than the amount of air being delivered to the reactor building. The exhaust fan has sufficient capacity to handle extra air intake to the building during momentary opening of doors, b) The control panel for the ventilation system provides for manual selection of air input to the reactor building and the automatic or manual selection of air removal. The air supply and exhaust systems work together to maintain a small negative pressure in the reactor building. These controls are located in the reception room for accessibility during emergency conditions.

3.4 Ventilation System Annlicability This specification applies to the operation of the facility ventilation system.

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NSC Technical Specifications, Version 09/01/1999 Obiective ne 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.

Soecification The reactor shall not be operated unless the facility ventilation system is operable, except for periods of time necessary to permit repair of the system. In the event of a substantial release of airbome radioactivity, the ventilation system will be secured automatically by signals from an exhaust air radiation monitor.

D.alis During normal operation of the ventilation system, the concentration of Argon-41 in unrestricted areas is below the Ef0uent Concentration (SAR,Section IX). In the event of a substantial release of airbome radioactivity, the ventilation system will be secured automatically. Therefore, operation of the reactor with the ventilation system shutdown for short periods of time to make repairs insures the same degree of control of release of radioactive materials. Moreover, radiation monitors within the building independent of those in the ventilation system will give warning of high levels of radiation that might occur during operation with the ventilation system secured.

3.5 Radiation Monitoring Systems and Effluents 3.5.1 Radiation Monitoring Aeolicability i

This specification applies to the radiation monitoring information which must be l available to the reactor operator during reactor operation.

Obiective i The objective is to assure that sufficient radiation monitoring information is available to i the operator to assure safe operation of the reactor.

I Specification 1 The reactor shall not be operated unless the radiation monitoring channels listed in the following table are operable.

Radiation Monitoring Channels

• Function Number Area Radiation Monitor Monitor radiation levels within I the reactor bay Continuous Air Radiation Monitor 1 Exhaust Gas Radiation Monitor Monitor radiation levels in the i exhaust air stack Exhaust Partictilate Radiation Monitor l For periods of time far maintenance to the radiation monitoring channels, the intent of this specification will be satisfied if they are replaced with portable gamma sensitive instruments having their own alarms or which shall be kept under visual observation. If two of the above monitors are not operating, the reactor shall be shutdown.

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NSC Technical Specifications, Version 09/01/1999 Bases

'Ihe radiation monitors provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.

3.5.2 Argon 41 Discharge Limit Anolicability This specification applies to the concentration of Argon-41 that may be discharged from the TRIGA reactor facility.

Obiective To insure that the health and safety of the public is not endangered by the discharge of Argon-41 from the TRIGA reactor facility.

Soecification The concentration of Argon.41 in the effluent gas from the facility as diluted by atmospheric air in the lee of the facility due to the turbulent wake effect shall not exceed 1.0 x 10-8 pCi/ml averaged over one year.

DAEis The maximum allowable concentration of Argon-41 in air in unrestricted areas as specified in Appendix B, Table II of 10CFR20 is 1.0 x 10 ' pCi/ml.Section IX of the SAR for the NSCR substantiates a 5.0 x 10-8 atmospheric dilution factor for a 2.0 mph wind speed. This dilution factor represents the conditions at the site building for a wind speed of 2.0 mph, which occurs less than 10% of the time on an annual basis.

3.5.3 Xenon and Iodine Monitoring Aeolicability This specification applies to the radiation monitoring' systems necessary to monitor and control the concentration of any effluent releases during the production of i2'I from the radioactive decay of i25xe, Obiective The objective is to assure that sufficient radiation monitoring information is available to the operator to insure that the health and safety of the general public is not endangered during the production of 25g, Specification

- No experiment that involves active handling of '25Xe and i2'l may be performed unless the following radiation monitoring systems are operable. No experiment may be perfonned, except decay of Xe, unless thet25Xe effluent monitoring channelis 25 operable.

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NSC Technical Specifications Version 09/01/1999 Radiation Monitorine Systems

• Number

'2'Xe Efiluent Monitoring Channel 1 i25 1 Air Monitor I For periods of maintenance to the '2'Xe effluent monitoring r5 mnel, the intent of this specification will be satisfied if it is replaced by building gas sampics.

Basis Normal facility effluent monitors are not calibrated for "'Xe or '251. The '2'Xe channel provides information to operators in the event there is a significant release of2 'Xe during the production of '253, 3.6 Limitations on Experiments 3.6.1 Reactivity Limits Anolicability This specification applies to the reactivity limits on experiments installed in the reactor and its experimental facilities.

Obiective The objective is to assure control of the reactor during the handling of experiments adjacent to or in the reactor core.

Snecifications The reactor shall not be operated unless the following conditions governing experiments exist.

a) Non-secured experiments shall have reactivity worths less than one dollar, b) The reactivity worth of any single experiment shall be less than two dollars.

Bases a) This specification is intended to provide assurance that the worth of a single unfastened experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were suddenly inserted.

b) The maximum worth of a single experiment is limited so that its removal from the cold critical reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

3.6.2 Material Limitations Annlicability This specification applies to experiments installed in the reactor and its experimental l

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Obiective The objective is to prevent damage to the reactor or excessive release of radioactivity by limiting materials quantity and radioactive material inventory of the experiment.

Specincations a) . Explosive materials in quantities greater than 5 pounds shall not be allowed within the reactor building. Irradiation of explosive materials shall be restricted as follows:

1) Explosive materials in quantities greater than 23 milligrams shall not be irradiated in the reactor pool. Explosive materials in quantities less than 25 milligrams may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.

2) Explosive materials in quantities greater than 25 milligrams shall be restricted from the reactor pool, the upper research level, the demineralizer room, cooling equipment room and the interior of the pool containment structure.

3) Explosive materials in quantities greater than 5 pounds shall not be irradiated in experimental facilities.

4) Cumulative exposures for explosive materials in quantities greater than 25 milligrams shall not exceed 10i2 n/cm 2for neutron or 25 Roentgen for gamma exposures.

b) Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 10 Ci.

BAK1 a) This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving explosive materials.

1) This specification is intended to prevent damage to the reactor core and safety related reactor components located within the reactor pool in the event of failure of an experiment involving the irradiation of explosive materials. Limited quantities ofless than 25 milligrams and proper containment of such experiment provide the required safety for in-pool irradiation.

2) This specification is intended to prevent damage to vital equipment by restricting the quantity and location of explosive materials within the reactor building. Explosives in quantities exceeding 25 milligrams are restricted from areas containing the reactor bridge, reactor console, pool water coolant and purification systems and reactor safety related equipment.

3) The failure of an experiment involving the irradiation of up to 5 pounds of explosive material in an experimental facility located external to the 19

NSC Technical Specifications, Version 09/01/1999 reactor pool structure will not result in damage to the reactor or the reactor pool containment structure.

4) This specification is intended to prevent any increase in the sensitivity of explosive materials due to radiation damage during exposures.

b) The 10 Cilimitation on Iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less than that allowed by 10 CFR 20 for an unrestricted area.

3.6.3 Failures and Malfunctions Aonlicability This specification applies to experiments installed in the reactor and its experimental facilities.

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

Snecifications a) Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the l experiment or reactor, (2) credible accident conditions in the reactor, or (3) possible accident conditions in the experiment shall be limited in activity such j I

that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the limit of Appendix B of 10CFR20.

b) In calculations pursuant to a) above, the following assumptions shall be used:

1) If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or acrosols produced will escape.

2) If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles at least 10% of these vapors can escape.

3) For materials whose boiling point is above 130*F and where wspors formed by boiling this material can escape only through an undisturbed colunm of water above the core, at least 10% of these vapors can escape.

c) If a capsule fails and releases material which could damage the reactor fuel or j structure by corrosion or other means, removal and physical inspection shall be performed to determine the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Director (NSC) or his designated attemate and determined to be satisfactory ,

before operation of the reactor is resumed.

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NSC Technical Specifications, Version 09/01/1999 Ilain a) This specification is intended to reduce the likelihood that airborne activities in excess of the limits of Appendix B of 10CFR20 will be released to the atmosphere outside the facility boundary of the NSC.

b) These assumptions are used to evaluate the potential aisborne radioactivity l release due to an experiment failure.

c) Operation of the reactor with reactor fuel or structure damage is prohibited to avoid release of fission products. Potential damage to reactor fuel or structure l must be brought to the attention of the Director (NSC) or his designated alternate for review to assure safe operation of the reactor.

3.6.4 Xenon Irradiation for Iodine Production Annlicability This specification applies to the experiments that produce I-125 from the activation of enriched '24Xe and the decay of i25Xe.

Obiective The objective is to prevent excessive release of radioactivity by limiting the quantity and radioactive material inventory of the experiment. j Snecifications i l

'2d (a) Xe activation experiments shall be controlled such that the total single i23 experiment activity produced is limited to no more than 2000 Ci of Xe.

5 (b) The total facility ' Xe inventory of all experiments shall not exceed 3500 Ci.

Bases (a) The 2000 Ci limitation on Xenon-125 produced in any one experiment assures that in the event of a failure of an experiment leading to the accidental release of xenon, the exposure to the general public would be less than 0.5 rem per year (10 CFR 20.1301).

(b) Xenon-125 production in excess of this limit is not necessary.

3.7 As Low As Reasonably Achievable (ALARA) Radioactive Effluents Released Annlicability i

This specification applies to the measures required to ensure that the radioactive effluents released from the facility are in accordance with ALARA criteria.

Obiective The objective is to limit the annual population radiation exposure resulting from operation of the reactor to a small percentage of the normal local background exposure.

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NSC Technical Specifications, Version 09/01/1999 Snecifications

1) In addition to the radiation monitoring specified in Section 5.4, an environmental radiation monitoring program shall be conducted to measure the integrated radiation exposure in and around the environs of the facility on a quarterly basis.

2) The annual radiation exposure due to reactor operation, at the closest off-site point of extended occupancy, shall not exceed twice the average local off-site background radiation.

3) The total annual discharge of Argon-41 into the environment may not exceed 30 Ci per year unless permitted by the RSB.

4) In the event of a significant fission product leak from a fuel rod or a significant airborne radioactive release from a sample being irradiated, as detected by the continuous air monitos, the reactor shall be shut down until the source of the leak is located and eliminated. However, the reactor may be continued to be operated on a short-term basis as needed to assist in determining the source of the leakage.

5) Before discharge, the facility liquid effluents collected in the holdup tanks shall be analyzed for the nature and concentration of radioactive effluents. The total annual quantity ofliquid eriluents (above background) shall not exceed 1 Ci per year.

Hiu!ia ne simplest and most reliable method of ensuring that ALARA release limits are accomplishing their objective of minimal facility-caused radiation exposure to the general public is to actually measure the integrated radiation exposure in the environment on and off the site.

3.8 Primary Coolant Conditions Apolicability This specification applies to the quality of the primary coolant in contact with the fuel cladding.

Obiective The objectives are (1) to minimize the possibility for corrosion of the cladding on the fuel i elements and (2) to minimize neutron activation of dissolved materials.

Snecifications

1) Conductivity of the bulk pool water shall be no higher than 5 x 104 mhos/cm for a period not to exceed two weeks.

2) The pH of the bulk pool water shall be between 5.5 and 8.0. Deviations ofpH values outside this range shall not exceed a period of two weeks.

Bases A small rate of corrosion continuously occurs in a water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.

By limiting the concentrations of dissolved materials in the water, the radioactivity of neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease 22

NSC Technical Specifications, Version 09/01/1999 the inventory of radionuclides in the entire coolant system, which will decrease personnel exposure during maintenance and operations.

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i NSC Technical Specifications, Version 09/01/1999 4.0 Surveillance Requirements 4.1 General Anolicability his specification, applies to the surveillance requirements of any system related to reactor safety.

Obiective The objective is to verify the proper operation'of any system related to reactor safety.

Snecifications l Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated or to specifications approved by the

. Reactor Safety Board. A system shall not be considered operable until after it is successfully ,

tested. i Basis l l

This specification relates to changes in reactor systems, which could directly affect the safety of l the reactor. As long as changes or replacements to these systems continue to meet the original  !

design specifications, then it can be assumed that they ineet the presently accepted operating j criteria.

4.2 Reactor Core Parameters -

4.2.1 Steady State Operation Anolicability This specification applies to the surveillance requirement of the power level monitoring channels Objective The objective is to verify that the maximum power level of the reactor meets the license requirements.

Soecification i A channel calibration shall be made of the power level monitoring channels by the calorimetric method annually but at intervals not to exceed 14 months.

Basis ne power level channel calibration will assure that the reactor will be operated at the proper power level.

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NSC Technical Specifications, Version 09/01/1999 4.2.2 Pulse Mode Operation Anolicability This specification applies to the surveillance requirements for operation of the reactor in the pulse mode.

Obiective The objective is to verify that operation of the reactor in the pulse mode is proper and safe and to determine if any significant changes in fuel characteristics have occurred.

Soecification The reactor shall be pulsed semiannually to compare fuel temperature rneasurements and core pulse energy with those of previous pulses of the same reactivity value or the reactor shall not be declared operational for pulsing until such pulse measurements are performed.

D.iulia The reactor is pulsed at suitable intervals to make a comparison with previous similar pulses and to determine if changes in fuel or core characteristics are taking place.

4.2.3 Shutdown Margin Aeolicability This specification applies to the surveillance requirement of control rod calibrations and shutdown margin.

Obiective The objective is to verify that the requirements for shutdown margins are met for operational cores.

Specification The reactivity worth of each control rod and the shutdown margin shall be determined

. annually but at intervals not to exceed 14 months.

Basis The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worth of experiments inserted in the core. Past experience with TRIGA reactors gives assurance that measurement of the reactivity worth on an annual basis is adequate to insure no significant changes in the shutdown margin.

4.2.4 Reactor Fuel Elements

. Aeolicability This specification applies to the surveillance requirements for the fuel elements.

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Obiective The objective is to verify the continuing integrity of the fuel element cladding and to j ensure that no fuel damage has occurred.

i Snecification a) The following fuel elements shall be inspected visually for damage or l deterioration and measured for length and bend annually, not to exceed 15 months.

1) At least four fuel elements which occupy the highest pulse temperature l positions in the core, j

2) At least one-fifth of the fuel elements used in operation of the reactor over the previous inspection year.

3) The four elements in (1) above may be included in the inspection of fuel elements of(2) above.

b) If any element is found to be damaged, the entire core will be inspected.

c) The reactor shall not be operated knowingly with damaged fuel.

d) A fuel element shall be considered damaged and must be removed from the core if:

1) In measuring the transverse bend, the bend exceeds 0.125 inch over the ,

length of the cladding.

2) In measuring the elongation, its length exceeds its original length by l 0.125 inch, or 1

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

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'Ihe frequency of inspection and measurement schedule is based on the parameters most likely to affect the fuel cladding of a pulsing reactor operated at moderate pulsing levels and utilizing fuel elements whose characteristics are well known. Experience has shown that temperature is the major contributor to fuel damage. Inspection of the foor fuel  ;

elements which occupy the highest pulse temperature positions in the core provides  !

surveillance for detection of the most probable fuel element damage should it occur. '

l Inspection of one fifth of elements used in operation of the reactor provides surveillance of the lower temperature elements and over a five year period provides for inspection of all elements.

The limit of transverse bend has been shown to result in no difficulty in disassembling fuel bundles. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching.

Experience with TRIGA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified to assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow.

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NSC Tecimical Specifications, Version 09/01/1999 4.3 Reactor Control And Safety Systems 4.3.1 Reactor Control Systems  !

Annlicability These specifications apply to the surveillance requirements for reactor control systems.

Obiective The objective is to verify the condition and operability of system components affecting safe and proper control of the reactor. l Snecifications a) The control rods shall be visually inspected for deterioration at intervals not to exceed 2 years.

b) The transient rod drive cylinder and associated air supply system shall be inspected, cleaned, and lubricated as necessary semiannually at intervals not to exceed 8 months.

Luis The visual inspection of the control rods is made to evaluate corrosion and wear I characteristics caused by operation of the reactor. Inspection and maintenance of the traaient rod drive assembly reduces the probability of failure of the system due to moisture induced corrosion of the pulse cylinder and piston rod assembly.

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4.3.2 Reactor Safety Systems Apolicability These specifications apply to the surveillance requirements for measurements, tests, and  ;

calibrations of the control and safety systems. j Obiective The objective is to verify the performance and operability of the systems and components which are directly related to reactor safety.

Specifications ,

a) A channel test of each of the reactor safety system channels for the intended mode of operation shall be performed prior to each day's operation or prior to each operation extending more than one day, except for the pool level channel which shall be tested weekly.

b) Whenever a reactor scram caused by high fuel element temperature occurs, an l evaluation shall be conducted to determine whether the fuel element l

temperature safety limit was exceeded.

c) A calibration of the temperature measuring channels shall be performed semiannually but at intervals not to exceed 8 months.

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NSC Technical Specifications, Version 09/01/1999 d) A channel check of the fuel element temperature measuring channel shall be made daily whenever the reactor is operated by recording a measured value of a meaningful temperature indication.

lluis Channel tests will assure that the safety system channels are operable on a daily basis or prior to an extended run. Operational experience with the TRIGA system gives assurance that the thermocouple measurements of fuel element temperatures have been sufficiently reliable to assure accurate indication of this parameter.

4.3.3 Scram Time Annlicability This specification applies to the surveillance of control rod scram times.

Obiective The objective is to verify that all scrammable control rods meet the scram time requirement.

1 Snecification The scram time shall be measured annually but at intervals not to exceed 14 months.

Dalit j Measurement of the scram time on an annual basis is a check not only of the scram system electronics, but also is an indication of the capability of the control rods to perform properly.

4.4 Equipment to Achieve Confinement: Ventilation System Arnlicability This specification applies to the building confinement ventilation system.

Obiective The objective is to assure the proper operation of the ventilation system in controlling releases of radioactive material to the uncontrolled environment.

Snecification l It shall be verified weekly that the ventilation system is operable.

Basis l

Experience accumulated over scveral years of operation has demonstrated that th *ests of the l ventilation system on a weekly basis are sufficient to assure the proper operation a the system and control of the release of radioactive material.

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NSC Technical Specifications, Version 09/01/1999 4.5 Radiation Monitoring Systems and Effluents Annlicability This specification applies to the surveillance requirements for the area radiation moniering equipment and the continuous air monitoring system.

Obiective The objective is to assure that the radiation monitoring equipment is operating and to verify the appropriate alarm settings.

Specification The area radiation monitoring system and the continuous air monitoring system shall be calibrated annually but at intervals not to exceed 14 months and shall be verified to be operable at weekly inten als.

Basis Experience has shown that weekly verification of area radiation and air monitoring system set points in conjunction with annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span. l 4.6 Experiments Annlicability l

i This specification applies to the surveillance requirements for experiments installed in the reactor and its experimental facilities and foi irradiations performed in the irradiation facilities.

l Obiective The objective is to prevent the conduct of experiments or irradiations which may damage the reactor or release excessive amounts of radioactive materials as a result of failure.

Snecifications a) A new experiment shall not be installed in the reactor or its experimental facilities until a hazard analysis has been performed and reviewed for compliance with the Limitations on ,

Experiments, Section 3.9, by the Reactor Safety Board. Minor modifications to a j reviewed and approved experiment may be made at the discretion of the senior reactor i operator responsible for the operation provided that the hazards associated with the modifications have been reviewed and a determination made and documented that the modifications do not create a significantly different, a new, or a greater safety risk than  ;

the original approved experiment. I 1

b) The performance of an experiment clamfied as an approved experiment shall not be j performed until it has been reviewed for compliance by a licensed senior operator and a person qualified in health physics.

c) The reactivity worth of an experiment shall be estimated or measured, as appropriate, before reactor operation with said experiment.

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i NSC Tecimical Specifications, Version 09/01/1999 Dmit it has been demonstrated over a number of years of experience that experiments and irradiations reviewed by the Reactor Staff and the Reactor Safety Board as appropriate can be conducted without endangering the safety of the reactor or exceeding the limits in the Technical Specifications, a

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NSC Technical Specifications, Version 09/01/1999 5.0 Design Features 5.1 Reactor Fuel Appheability  ;

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This specification applies to the fuel elements used in the reactor core, j Obiective The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and i nuclear characteristics.

Snecifications 1

a) TRIGA-FLIP Fuel i The individual unirradiated FLIP fuel elements shall have the following characteristics:

1) Uranium content: maximum of 9 Wt% enriched to nominal 70% Uranium-235,

2) liydrogen-to-zirconium atom ratio (in the Zrli x): nominal 1.611 atoms to 1.0 Zr atoms.

3) Natural erbium content (homogeneously distributedP nominal 1.5 Wt%.

4) Cladding: 304 stainless steel, nominal 0.020 inch thick.

5) Identification: Top pieces of FLIP elements will have characteristic markings to allow visual identification of FLIP elements employed in mixed cores.

b) Standard TRIGA fuel The individual unirradiated standard TRIGA fuel elements shall have the following characteristics:

1) Uranium content: maximum of 9.0 Wt% enriched to a nominal 20% Uranium- l 235.

2) liydrogen-to-zirconium atom ratio (in the Zril x): nominal 1.711 atoms to 1.0 Zr atoms.

3) Cladding: 304 stainless steel, nominal 0.020 inch thick.

D.AKE a) A maximum uranium content of 9 Wt% in a TRIGA-FLIP element is about 6% greater than the design value of 8.5 Wt%. Such an increase in loading would result in an increase in power density of about 2%. Similarly, a minimum erbium content of 1.1% in an element is about 30% less than the design value. This variation would result in an increase in power density of only about 6%. An increase in local peuver density of 6%

reduces the safety margin by at most ten percent. The maximum hydrogen-to-zirconium

[

ratio of 1.65 could result in a maximum stress under accident conditions in the fuel element clad about a factor of two greater than the value resulting from a hydrogen-to-zirconium ratio of 1.60. Ilowever, this increase in the clad stress during an accident would not exceed the rupture strength of the clad.

When standard and FLIP fuel elements are used in mixed cores, visual identification of r

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NSC Technical Speci0 cations, Version 09/01/1999 types of elements is necessary to verify correct fuel loadings. The accidental rotation of fuel bundles containing standard and FLIP clements can be detected by visual inspection.

Should this occur, however, studies of a single FLIP element accidentally rotated into a standard fuel region indicate an insubstantial increase in power generation in the FLIP clement.

b) A maximum uranium content of 9 Wr% in a standard TRIGA element is about 6%

greater than the design value of 8.5 Wt%. Such an increase in loading would result in an increase in power density ofless than 6%. An increase in local power density of 6%

reduces the safety margin by at most 10%. The maximum hydrogen-to-zirconium ratio of 1.8 will produce a maximum pressure within the clad during an accident well below the rupture strength of the clad.

5.2 Reactor Core Anolicability This speci0 cation applies to the configuration of fuel and in core experiments.

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

Snecifications a) The core shall be an arrangement of TRIGA uranium-zirconium hydride fuel-moderator bundles positioned in the reactor grid plate.

b) The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite and water or D,0.

Bases a) Standard TRIGA cores have been in use for years and their characteristics are well documented. FLIP cores have been operated at General Atomics and the Puerto Rico Nuclear Center and their operational characteristics are available. General Atomics has also performed a series of experiments using standard and FLIP fuel in mixed cores. In addition, studies performed at Texas A&M for a variety of mixed core arrangements and operational experience with mixed cores indicate that such loadings would safely satisfy all operational requirements.

b) The core will be assembled in the reactor grid plate which is located in a pool of hght I water. Water in combination with graphite reflectors can be used for neutron economy and the enhancement of experimental facility radiation requirements.

5.3 Control Rods Annlicability This specification applies to the control rods used in the reactor core.

Obiective  ;

The objective is to assure that the control rods are of such a design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics. l l

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NSC Technical Specifications, Version 09/01/1999 Snecifications a) He shim-safety control rods shall have scram capability and contain borated graphite, B.C powder or boron and its compounds in solid form as a poison in aluminum or stainless steel cladding. These rods may incorporate fueled followers which have the same characteristics as the fuel region in which they are used.

b) The regulating control rod need not have scram capability and shall be a stainless rod or contain the materials as specified for shim-safety control rods. This rod may incorporate a fueled follower.

c) The transient control rod shall have scram capability and contain borated graphite or boron and its compounds in solid form as a poison in an aluminum or stainless steel clad.

The transient rod shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may incorporate an aluminum or air follower.

Ilaici The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, D4C powder or boron and its compounds. Since the regulating rod normally is a low worth rod, its function could be satisfied by using a solid stainless steel rod. These materials must be contained in a suitable clad material, such as aluminum or stainless steel, to insure mechanical stability during movement and to isolate the poison from the pool water environment. Control rods that are fuel followed provide additional reactivity to the core and increase the worth of the control rod. The use of fueled followers in the FLIP region has the additional advantage of reducing flux peaking in the water filled regions vacated by the withdrawal of the control rods.

Scram capabilities are provided for rapid insertion of the control rods which is the primary safety feature of the reactor. The transient control rod is designed for a reactor pulse. The nuclear behavior of the air or aluminum follower which may be incorporated into the transient rod is similar to a void. A voided follower may be required in certain core loadings to reduce flux peaking values.

5.4 Radiation Monitoring System Anolicability This specification describes the functions and essential components of the area radiation monitoring equipment and the system for continuously monitoring airborne radioactivity.

Obiective The objective is to describe the radiation monitoring equipment that is available to the operator to  !

assure safe operation of the reactor.

Snecification The radiation monitoring equipment listed in the following table will have these characteristics.

Radiation Monitoring Channel and Function Area Radiation Monitor (gamma sensitive instruments)

Function: Monitor radiation fields in key locations, alarm and readout at control ,

console and readout in reception room. l I

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NSC Technical Specifications, Version 09/01/1999 Continuous Air Radiation Monitor (beta-gamma sensitive detector with air collection capability)

Function: Monitor concentration of radioactive particulate activity in building, alann and readout at control console, and readout in reception room.

Gas and Particulate Stack Radiation Monitors (gamma and beta-gamma sensitive detectors with air collection capability)

Function: Monitor concentration of radioactive particulate activity and radioactive gases in building exhaust, alarm and readout at control console and readout in reception room.

Ilaiis The radiation monitoring system is intended to provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.

5.5 Fuel Storage Annlicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Obiective The objective is to assure that fuel which is being stored will not become critical and will not reach an unsafe temperature.

Snecifications a) All fuel elements shall be stored in a geometrical array for which the k-effective is less than 0.8 for all conditions of moderation.

b) Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.

Basis The limits imposed by Specifications 5.5.a and 5.5.b are conservative and assure safe storage.

5.6 Reactor Building and Ventilation System Annlicability This specification applies to the building which houses the reactor.

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

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NSC Technical Specifications, Version 09/01/1999 Snecificatiom j a) The reactor shall be housed in a facility designed to restrict leakage. The minimum free volume in the facility shall be 180,000 cubic feet. 1 l

b) The reactor building shall be equipped with a ventilation system designed to filter and exhaust air or other gases from the reactor building and release them from a stack at a minimum of 85 feet from ground leve'l.

c) Emergency shutdown controls for the ventilation system shall be located in the reception room and the system shall be designed to shut down in the event of a substantial release of fission products.

Bases The facility is designed such that the ventilation system will normally maintain a negative pressure with respect to the atmosphere so that there will be no uncontrolled leakage to the environment. The free air volume within the reactor building is confined when there is an emergency shutdown of the ventilation system. Controls for startup, emergency filtering, and l normal operation of the ventilation system are located in the reception room. Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reception room with a minimum of exposure to operating personnel.

1 5.7 Reactor Pool Water Systems l Annlicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.

Obiective The objective is to assure that coolant water shall be available to provide adequate cnoting of the i reactor core and adequate radiation shielding.  !

Snecifications a) The reactor core shall be cooled by natural convective water flow.

b) The pool water inlet and outlet pipe to the demineralizer shall not extend more than 15 feet below the top of the reactor pool when fuel is in the core.

c) Diffuser and skimmer pumps shall be located no more than 15 feet below the top of the reactor pool.

d) Pool water inlet and outlet pipes to the heat exchanger shall have emergency covers within the reactor pool for manual shut offin case of pool water loss due to external pipe system failure, e) A pool level alarm shall indicate loss of coolant if the pool level drops approximately 10% below operating level.

Bases a) This specification is based on thermal and hydraulic calculations which show that the TRIGA-FLIP core can operate continuously in a safe manner at power levels up to 2,700 kW with natural convection flow and sufficient bulk pool cooling. A comparison of 35 a_____

NSC Technical Specifications, Version 09/01/1999 operation of the TRIGA-FLIP and standard TRIGA Mark 111 has shown them to be safe for the above power level. Thermal and hydraulic characteristics of mixed cores are essentially the same as that for TRIGA-FLIP and standard cores.

b) In the event of accidental siphoning of pool water through inlet and outlet pipes of the demineralizer system, the pool ater level will drop no more than 15 feet from the top of the pool.

c) In the event of pipe failure and siphoning of pool water through the skimmer and diffuser water systems, the pool water level will drop no more than 15 feet from the top of the pool.

d) Inlet and outlet coolant lines to the pool heat exchanger terminate at the bottom of the pool. In the event of pipe failure, these lines must be manually sealed from within the reactor pool. Covers for these lines will be stored in the reactor pool. The time required to uncover the reactor core due to failure of a single pool coolant pipe system is 17 minutes.

e) Loss of coolant alarm after 10% loss requires corrective action. This alarm is observed in the reactor control room and the reception room.

5.8 Physical Security The licensee shall maintain in effect and fully implement all provisions of the NRC staff approved physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54 (p). The approved security plan consists of documents withheld from public disclosure pursuant to 10 CFR 2.70, collectively titled " Texas A&M University System, Nuclear Science Center Reactor Security Plan."

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l NSC Techrdcal Specifications, Version 09/01/1999 6.0 Administrative Controls 6.1 Organization 6.1.1 Structure 2 a) A line management organizational structure shall provide for the administration and operation of the reactor facility. The Deputy Director of the Texas Engineering Experiment Station (TEES) and the Director of the Nuclear Science Center (NSC) shall have line management responsibility for adhering to the terms and conditions of the Nuclear Science Center Reactor (NSCR) license and technical specifications and for safeguarding the public and facility personnel from undue radiation exposure. The facility shall be under the direct control of the Director (NSC) or a licensed senior reactor operator designated by him to be in direct control.

b) Management Levels Level 1: Deputy Director TEES (Licensee): Responsible for the NSCR facility license.

Level 2: Director (NSC): Responsible for reactor facility operation and shall report to Level 1.

Level 3: Senior Reactor Operator on Duty: Responsible for the day-to-day operation of the NSCR or shift operation and shall report to Level 2.

Level 4: Reactor Operating Staff: Licensed reactor operators and .

y senior reactor operators and trainees. These individuals shall report to Level 3.

c) Radiation Safety A qualified, health physicist shall be assigned responsibility for implementation of the radiation protection program at the NSCR. The individual shall report to Level 1 management.

d) Reactor Safety Board (RSB)

Responsible to the Licensee for providing an independent review and audit of the safety aspects of the NSCR.

6.1.2 Responsibility Responsibility for the safe operation of the reactor facility shall be in accordance with the line organization established in 6.1.1.

6.1.3 Staffing a) The minimum staffing when the reactor is not secured shall be as follows:

1) At least two individuals will be present at the facility complex and will consist of a licensed senior reactor operator and either a licensed reactor operator or operator trainee. During periods of reactor maintenance as specified in 1.27 (b) the reactor operator or the operator trainee may be replaced by maintenance personnel.*

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NSC Technical Specifications, Version 09/01/1999

2) A licensed reactor operator or senior reactor operator will be in the control room.

3) The Director (NSC) or his designated alternate will be readily available for emergencies (i.e., capable of getting to the reactor facility within a reasonable time).

4) At least one member of the health physics support group will be readily available to provide advice and technical assistance in the area of radiation protection.

The licensed senior reactor operator may be permitted as the only operations person present at the facility to perform a pre-startup check of the reactor or perform general reactor maintenance not specified in 1.27 (b).

b) A list of reactor facility personnel by name and telephone number shall be readily available for use in the control room. The list shall include: I

1) Administrative personnel

2) Radiation safety personnel j I

3) Other operations personnel l c) The following designated individuals shall direct the events listed:

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1) The Director (NSC) or his designated attemate shall direct any loadmg j of fuel or control rods within the reactor core region. l i

2) The Director (NSC) or his designated alternate shall direct any loading of an in-core experiment with a reactivity worth greater than one dollar.

3) The senior reactor operator on duty shall direct the recovery from an unplanned or unscheduled shutdown other than a safety limit violation.

6.1.4 Selection and Training of Personnel The selection and training of operations personnel shall be in accordance with the l following:

a) Responsibility: The Director (NSC) or his designated alternate is responsible for the training and requalification of the facility reactor operators and senior reactor operators, b) Requalification Program

Purpose:

To insure that all operating personnel maintain proficiency at l 1) a level equal to or greater than that required for initial licensing.

2) Scope: Scheduled lectures, written examinations and evaluated console manipulations will be used to insure operator proficiency is maintained. 1 s

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1 NSC Technical Specifications, Version 09/01/1999 6.2 - Review and Audit 6.2.1 Reactor Safety Board j The Reactor Safety Board (RSB) shall consist of at least three 3) voting members knowledgeable in fields which relate to nuclear safety. The RSB shall review, evaluate and make recommendations on safety standards associated with the operational use of the facility. NSCR operations and health physics shall be represented as ex-officio members on the RSD. The review and advisory functions of the RSB shall include NSCR operations, radiation protection, and the facility license. )

6.2.2 RSB Charter and Rules

.The operations of the RSB shall be in accordance with a written charter, including provis.ons for:

a) Meeting frequency: not less than once per calendar year and as frequent as circumstances warrant consistent with effective monitoring of facility activities.

. b) Voting rules c) Quorums d) Use of subcommittees e) Review, approval and dissemination of minutes 1

6.2.3 RSB Review Function l The responsibilities of the RSB or a designated subcommittee thereofinclude, but are not limited to the following:

a) Review and evaluation of whether a proposed change, test, or experiment would  ;

constitute an unreviewed safety question or a change in Technical  :

Specifications.

b) Review of new procedures, major revisions of procedures, and proposed

hanges in reactor facility equipment or systems having safety significance.

c) Review of new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity, d) Review of proposed changes in technical specifications, license, or charter, e) Review of the NSCR radiation protection program.

I f) Review of violations of technical specifications, license, or charter, and violations ofintemal procedures or instructions having safety significance.

g) Review of operating abnonnalities having safety significance.

h) Review of reportable occurrences listed in 6.6.2.

i) Review of audit reports.

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NSC Technical Specifications, Version 09/01/1999 i 6.2.4 RSB Audit Function The RSB or a subcommittee thereof shall audit reactor operations and radiation l protection programs at least quarterly, but at intervals not to exceed four months. Audits shall include but are not limited to the following: j a) Facility operations, including radiation protection, for conformance to the technical specifications, applicable license conditions, and standard operating procedures at least once per calendar year (interval between audits not to exceed l I

15 months).

b) The retraining and requalification program for the operating staff at least once per calendar year (interval between audits not to exceed 15 months).

c) The facility security plan and records at least once per calendar year (interval between audits not to exceed 15 months).

I d) The reactor facility emergency plan and implementing procedures at least once per calendar year (interval between audits not to exceed 15 months).

6.2.5 Audit of ALARA Program The licensee or his designated alternate (excluding anyone whose normaljob function is  ;

within the NSCR) shall conduct an audit of the reactor facility ALARA program at least '

once per calendar year (interval between audits not to exceed 15 months). The results of the audit shall be transmitted by the licensee to the RSB at the next scheduled meeting.

6.3 Operating Procedures Written operating procedures shall be prepared, reviewed, and approved prior to initiating any of the activities listed in this section. The procedures shall be reviewed and approved by the Director ,

(NSC), or his designated alternate, the Reactor Safety Board, and shall be documented in a timely 1 manner. Procedures shall be adequate to assure the safe operation of the reactor but shall not preclude the use ofindependentjudgment and action should the situation require such. Operating procedures shall be in effect for the following items:

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

b) Fuel loading, unloading, and movement within the reactor.

c) Control rod removal or replacement.

d) Routine maintenance of the control rod, drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety.

c) Testing and calibration of reactor instrumentation and controls, control rod drives, area radiation monitors, and facility air monitors.

f) Administrative controls for operations, maintenance, and conduct ofirradiations and experiments, that could affect reactor safety or core reactivity.

g) Implementation of required plans such as emergency or security plans.

h) Actions to be taken to correct specific and foreseen potential malfunctions of systems, including responses to alarms and abnonnal reactivity changes.

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1 NSC Technical Specifications, Version 09/01/1999 l Substantive changes to the above procedures shall be made effective only after documented review and approval by the Director (NSC) and the Reactor Safety Board. Minor modifications or temporary changes to the original procedures which do not change their original intent may be made by the Director (NSC) or his designated alternate. All such temporary changes shall be documented and subsequently reviewed by the Reactor Safety Board.

6.4 Experiments Review and Approval Approved experiments shall be canied out in accordance with established and approved procedures, a) All new experiments or class of experiments shall be reviewed by the RSB (Section 6.2.3) and implementation approved in writing by the Director (NSC) or his designated alternate.

b) Substantive changes to previously approved experiments shall be made only after review by the RSB and implementation approved in writing by the Director (NSC) or his l designated attemate. Minor changes that do not significantly alter the experiment may be l approved by the Director (NSC) or his designated alternate.

6.5 Required Actions 6.5.1 Action to be Taken in the Event a Safety Limit is Exceeded in the event a safety limit is exceeded: I 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, Reactor Safety Board, and reports shall be made to the NRC in accordance with Section 6.6.2 of these specifications, and  ;

l c) A report shall be prepared which shall include an analysis of the cause and l extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submined to the Reactor Safety Board for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.

i 6.5.2 Action to be Taken in the Event of a Reportable Occurrence I

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

a) Reactor conditions shall be returned to normal or the reactor shall be shut down.

Ifit is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by the Director (NSC) or his designated alternate.

b) The Director (NSC) or his designated alternate shall be notified and corrective action taken with respect to the operations involved.

c) The Director (NSC) or his designated alternate shall notify the Chairman of the Reactor Safety Board.

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NSC Technical Specifications, Version 09/01/1999 d) A report shall be made to the Reactor Safety Board which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence, and c) A report shall be made to the NRC in accordance with Section 6.6.2 of these specifications.

f) Occurrence shall be reviewed by the RSB at their next scheduled meeting.

'6.6 Reporting Requirements 6.6.1 Annual Report An annual report covering the operation of'the reactor facility during the previous calendar year shall be submitted to the NRC prior to March 31 of each year providing the following information:

a) A brief narrative sununary of(1) operating experience (including experiments performed), (2) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections; b) Tabulation of the energy output (in megawatt days) of the reactor, hours reactor was critical, and thw eumulative total energy output since initial criticality; c) The number of emergency shutdowns and inadvertent scrams, including reasons therefore; d) 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; e) 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; f) A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed or recommended, a statement to this effect is sufficient.

1) Liquid Waste (summarized on a monthly basis)

(a) Radioactivity discharged during the reporting period.

(1) Total radioactivity released (in Curies).

(2) The Effluent Concentration used and the isotopic composition if greater than 1 x 10-7 pCuries/cc for fission and activation products.

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

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NSC Technical Specifications, Version 09/01/1999 (4) Average concentration at point of release (in pCuries/cc) during the reporting period.

(b) Totet volume (in gallons) of effluent water (including dilutent) during periods of release.

2) Gaseous Waste (summarized on a monthly basis) )

'(a) Radioactivity discharged during the reporting period (in Curies) for (1) Argon-41 4

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

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3) Solid Waste (a) The total amount of solid waste transferred (in cubic feet).

(b) The total activity involved (in Curies).

(c) The dates of shipment and disposition (if shipped off site).

g) A summary of radiation exposures received by facility personnel and visitors, including dates and time where such exposures are greater than 25% of that l

allowed or recommended.

h) A description and summary of any environmental surveys performed outside the facility.

6.6.2 Special Reports In addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the NRC Document Control Desk and special telephone reports of events should be made to the Operations Center as follows:

1 a) There shall be a report not later than the following working day by telephone i and confirmed in writing by telegraph or similar conveyance to be followed by a written report that describes the circumstances of the event within 14 days of any of the following:

(1) Violation of safety limits (See 6.5.1). 1 (2) Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property (

damage, personal injury, or exposure; (3) Any reportable occurrences as defined in Section 1.29 of these Specifications. The written report (and, to the extent possible, the preliminary telephone or telegraph report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent reoccurrence of the event; 43

NSC Technical Specifications, Version 09/01/1999 (b) A written report within 30 days of:

(1) Personnel changes in the facility organization involving Level 1 and j Level 2.

(2) Significant changes in the transient or accident analysis as described in the Safety Analysis Report. j l

6.7 ' Records Records of facility operations in the form oflogs, data sheets or other suitable forms shall be retained for the period indicated as follows:

6.7.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component involved a) Normal reactor facility operation I b) Principal maintenance operations c) Reportable occurrences d) Surveillance activities required by the Technical Specifications e) Reactor facility radiation and contamination surveys where required by applicable regulations f) Er.periments performed with the reactor g) Fuel inventories, receipts, and shipments h) Approved changes in operating procedures i) Records of meeting and audit repo:ts of the RSB 6.7.2 Records to be Retained for at 1. cast One Training Cycle Retraining and requalification of certified operations personnel. Records of the most recent complete cycle shall be maintained at all times the individual is employed.

6.7.3 Records to be Retained for the Lifetime of the Reactor Facility a) Gaseous and liquid radioactive efiluents released to the environs.

b) Off-site environmental monitoring surveys required by the Technical Specifications.

c) Radiation exposure for all personnel monitored.

d) Drawings of the reactor facility.

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