ML100710497

University of Florida Training Reactor (Uftr) Revised Technical Specifications in Support of Application for Renewal of Facility Operating License R-56

ML100710497
Person / Time
Site:	05000083
Issue date:	03/11/2010
From:	Haghighat A Univ of Florida
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME1586
Download: ML100710497 (55)

Text

UJFUNIVERSITY of UFFLORIDA College of Engineering Department of Nuclear & Radiological Engineering University of Florida Training Reactor 202 Nuclear Science Bldg.PO Box 118300 Gainesville, FL 32611-8300 352-392-1401 x306 352-392-3380 Fax haghighat@ufl.edu March 11, 2010 Document Control Desk U.S, Nuclear Regulatory Commission Washington, DC 20555 Attn: Mr. Duane Hardesty

Subject:

University of Florida Training Reactor (UFTR) License Renewal (TAC NO. ME 1586), DOCKET NO. 50-83 Enclosed is a revised Technical Specifications (TS) in support of our application for renewal of Facility Operating License No. R-56 for the University of Florida Training Reactor (UFTR). Enclosed is also a spreadsheet providing justification for the changes considered in the revised TS.If you need further information, please do not hesitate to contact me at haghighat(&ufl.edu or (352)392-1401 x306.I declare under penalty of perjury that the foregoing is true and correct.Executed on March 11, 2010 Sincerely, IL4 -/k Lt& -4w -&- a&CAROLE ROSE REVELLE Notary Public-Maryland Montgomery County My Commission Expires July 02, 2013 Alireza Hagtllghat, PtliD FP&L Endowed Chair Professor Director of UFTR Cc Jack Donohue, NRC Gabriel Ghita, UF David Hintenlang, Interim Chair Brian Shea, UFTR Manager Glenn Sjoden, RSRS Chair UFTR -NRC fileý\O2O The Foundation for The Gator Nation An Equal Opportunity Institution 00- '\'

APPENDIX 14.1 TECHNICAL SPECIFICATIONS Table of Contents

1.0 INTRODUCTION

.........................................................................................................

1 1.2 D E FIN IT IO N S ......................................................................................................................

1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ......................

4 2.1 SAFETY LIMIT (SL) ......................................................................................................

4 2.2 LIMITING SAFETY SYSTEM SETTINGS (LSSS) ..............................................................

5 3.0 LIMITING CONDITIONS FOR OPERATION (LCO) ...............................................

5 3.1 REACTOR CORE PARAMETERS

........................................................................................

5 3.2 REACTOR CONTROL AND SAFETY SYSTEMS ....................................................................

6 3.2.1 R EACTOR C ONTRO L SYSTEM ...........................................................................................

6 3.2.3 REACTOR CONTROL AND SAFETY SYSTEMS CHANNELS .................................................

8 3.3 COOLANT SYSTEMS ..............................................................................................................

8 3.4 RADIATION MONITORING SYSTEMS AND RADIOACTIVE EFFLUENTS

...........................

9 3.4.1 AREA RADIATION MONITORS AND AIR PARTICULATE DETECTORS

..............................

9 3.4.2 A RGON-41 (A R-41)DISCHARGE

......................................................................................

9 3.4.3 REACTOR VENT/STACK MONITORING SYSTEM .............................................................

10 3,4.4 LIQUID EFFLUENTS D ISCHARGE ....................................................................................

10 3.5 REACTOR VENT SYSTEM .................................................................................................

11 3.6 LIMITATIONS ON EXPERIMENTS

.....................................................................................

12 3.7 REACTOR BUILDING EVACUATION ALARM ...................................................................

14 3.8 FUEL AND FUEL HANDLING ............................................................................................

14 3.9 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ......................................

15 4.0 SURVEILLANCE REQUIREMENTS

............................................................................

20 4.1 REACTOR CORE PARAMETERS

......................................................................................

20 4.4 RADIATION MONITORING SYSTEMS AND RADIOACTIVE EFFLUENTS

.........................

22 4.5 REACTOR VENT SYSTEM .................................................................................................

23 4.6 LIMITATIONS ON EXPERIMENTS

....................................................................................

24 4.7 REACTOR BUILDING EVACUATION ALARM .................................................................

24 4.8 SURVEILLANCE PERTAINING TO FUEL .................

.............................

24 4.9 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ......................................

25 5.0 DESIGN FEATURES ...................................................................................................

29 5.1 SITE AND FACILITY DESCRIPTION

................................................................................

29 5.2 REACTOR COOLANT SYSTEM ........................................................................................

29 5.2.1 PRIM ARY C OOLIN G SYSTEM ..........................................................................................

29 5.2.2 SECONDARY COOLING SYSTEM ....................................................................................

30 5.3 REACTOR CORE AND FUEL ............................................................................................

30 5.4 FISSIONABLE MATERIAL STORAGE ...................................................................................

31 i

6.0 ADMINISTRATIVE

CONTROLS ...............................................................................

31 6.1 O RGANIZATION

...................................................................................................................

31 6 .1 .1 S T R U C T U R E ........................................................................................................................

3 1 6 .1 .2 R E S PO N S IB IL IT Y ................................................................................................................

3 1 6 .1 .3 S T A F F IN G ...........................................................................................................................

3 2 6.1.4 SELECTION AND TRAINING OF PERSONNEL

.......................................................................

33 6.2 R EVIEW AND A UDIT ............................................................................................................

33 6.2.1 COM POSITION AND Q UALIFICATIONS

.............................................................................

33 6.2.2 C HA RTER AN D R U LES .....................................................................................................

33 6 .2 .3 R EV IEW F U N C T IO N ............................................................................................................

34 6 .2 .4 A U D IT F U N C T IO N ...............................................................................................................

3 5 6.3 RADIATION SAFETY AND ALARA (AS Low As REASONABLY ACHIEVABLE)

...............

35 6.4 P ROCEDURES

.......................................................................................................................

36 6.5 EXPERIMENT REVIEW AND APPROVAL .........................................................................

36 6.6 REQUIRED ACTIONS ........................................................................................................

37 6.6.1 ACTION TO BE TAKEN IN CASE OF SAFETY LIM IT VIOLATION

....................................

37 6.6.2 ACTION TO BE TAKEN IN THE EVENT OF AN OCCURRENCE OF THE TYPE IDENTIFIED IN SECTION 6.7.2(2) AN D 6.7.2(3) ..................................................................................................

38 6.7 R EPO RTS ..............................................................................................................................

38 6.7.1 O PERATIN G R EPO RTS ...........................................

.........................................................

38 6 .7 .2 S PEC IA L R EPO R T S ..............................................................................................................

39 6.7.3 O THER SPECIAL R EPORTS ..............................................................................................

40 6.8 R ECO RDS .............................................................................................................................

40 6.8.1 RECORDS TO BE RETAINED FOR APERIOD OF AT LEAST FIVE YEARS ................

40 6.8.2 RECORDS TO BE RETAINED FOR AT LEAST ONE TRAINING CYCLE ..............................

41 6.8.3 RECORDS To BE RETAINED FOR THE LIFETIME OF THE REACTOR FACILITY ....................

41 INDEX OF TABLES TABLE 3-1 SPECIFICATION FOR REACTOR SAFETY SYSTEM TRIPS 17 TABLE 3-2 MINIMUM NUMBER AND TYPE OF CHANNELS OPERABLE 18 TABLE 3-3 RADIATION MONITORING SYSTEM SETTINGS 19 TABLE 4-1 SAFETY SYSTEM OPERABILITY TESTS 27 TABLE 4-2 CONTROL BLADE WITHDRAWAL INHIBIT INTERLOCKS CHECKS_28 INDEX OF FIGURES FIGURE 6-1 UFTR ORGANIZATIONAL CHART 42 ii

1.0 INTRODUCTION

1.1 SCOPE

This document describes the Technical Specifications (TS) for the University of Florida Training Reactor (UFTR). Areas addressed are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls.It is important to note that for the UFTR, the "reactor safety" or "safety-related" terminology does not have the same safety significance of a power reactor. According to ANSI/ANS 15.1 standard, the Reactor Safety Systems are defined: "those systems, including their associated input channels that are designed to initiate automatic protective action or to provide information for initiation of manual protective action." 1.2 DEFINITIONS Blade-Drop Time: The blade-drop time is the measured time it takes, once power is removed from the control blade clutch, for the control blade to travel from the upper limit to the bottom limit..Channel: A channel is the combination of sensor, line, amplifier, and output devices that are connected for the purpose of measuring the value of a parameter.

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.

Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and shall be deemed to include a channel test.Channel Check: A channel check is a qualitative verification of acceptable performance by observation of channel behavior, or by comparison of the channel with other independent channels or systems measuring the same parameter.

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

Confinement means a closure on the reactor room air volume such that the movement of air into and out of the reactor room is through a controlled path.Excess Reactivity:

Excess reactivity is that amount of reactivity that would exist if all reactivity control blades were moved to the maximum reactive conditions from the point where the reactor is exactly critical (keff=l) at reference core conditions or at specified set of conditions.

1 Experiment:

Any operation, hardware, or target (excluding devices such as detectors, foils, etc.) that is designed to investigate non-routine reactor characteristics or that is intended for irradiation within the pool, on or in a beam port or irradiation facility.Hardware rigidly secured to a core or shield structure so as to be a part of its design to carry out experiments is not normally considered an experiment.

Inhibit: An Inhibit is an action that prevents the withdrawal of control blades under a potentially unsafe condition.

License: The written authorization, by the responsible authority, for an individual or organization to carry out the duties and responsibilities associated with a personnel position, material, or facility requiring licensing.

Licensee:

An individual or organization holding a license.Measured Value: The measured value is the value of a parameter as it appears on the output of a channel.Movable Experiment:

A movable experiment is one where it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.

Operable:

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

Operating means a component or system is performing its intended function.Protective Action: Protective action is the initiation of a signal or the operation of equipment within the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.Reactivity Worth of an Experiment:

The reactivity worth of an experiment is the value of the reactivity change that results from the experiment, being inserted into or removed from its intended position.Reactor Operating:

The reactor is operating whenever it is not secured or shutdown.Reactor Operator:

An individual who is licensed to manipulate the controls of a reactor.Reactor Safety System: Reactor safety systems are those systems, including their associated input channels that are designed to initiate automatic protective action or to provide information for initiation.of manual protective action.Reactor Secured: The reactor is secured when 2 (1) Either there is insufficient moderator available in the reactor to attain criticality or there is insufficient fissile material present in the reactor to attain criticality under optimum available conditions of moderation and reflection; (2) Or when all the following conditions exist: a. The reactor is shutdown;b. The console key switch is in the off position, and the key is removed from the console;c. No work is in progress involving core fuel, core structure, installed control blades or control blade drives unless they are physically decoupled from the control blades;d. No experiments are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment.

Reactor Shutdown:

The reactor is shut down if it is subcritical by at least one dollar in the reference core condition with the reactivity worth of all installed experiments included.Reactor Trip: A reactor trip is considered to occur whenever one of the following two actions takes place: (1) Blade-Drop Trip -a gravity drop of all control blades into the reactor core as a result of terminating electrical power to the blade drive magnetic clutches.(2) Full-Trip

-the water is dumped from the reactor core by the safety actuation of the dump valve in addition to the blade-drop trip.Reference Core Condition:

The condition of the core when it is at ambient temperature (20 C / 68 F) and the reactivity worth of the xenon is zero.Reportable Occurrence:

A reportable occurrence is any of the conditions described in Section 6.7.2 of this specification.

Research Reactor: A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research, developmental, educational, training, or experimental purposes, and that may have provisions for the production of radioisotopes.

Safety Channel: A safety channel is a channel in the reactor safety system.Secured Experiment:

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

Senior Reactor Operator:

Any individual who is licensed to direct the activities of Reactor Operators.

Such an individual is also a reactor operator.3 Shall, Should, and May: The word "shall" is used to denote a requirement; the word"should" is used to denote a recommendation; and the word "may" is used to denote permission, neithera requirement nor a recommendation.

Shutdown Margin: Shutdown margin is the minimum shutdown reactivity necessary to provide confidence that.the reactor can be made subcritical by means of the control and safety systems starting from any permissible operating condition and with the most reactive blade in the most reactive position and that the reactor will remain subcritical without further operator action.Surveillance Intervals:

Allowable surveillance intervals shall not exceed the following:

(1) 10 years- interval not to exceed 12 years (2) 5 years -interval not to exceed 6 years (3) Biennial -intervals not to exceed 2 /2 years (4) Annual -intervals not to exceed 15 months (5) Semiannual

-intervals not to exceed 8 months (6) Quarterly

-intervals not to exceed 4 months (7) Monthly -intervals not to exceed 6 weeks (8) Weekly -intervals not to exceed 10 days (9) Daily -completed prior to reactor startup 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 that could adversely affect safe operation, not including shutdowns that occur during testing or checkout operations.

2.0 SAFETY

LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit (SL)Applicability:

This specification applies to the temperature of the reactor fuel.Objective:

To ensure fuel cladding integrity.

Specification:

The fuel and cladding temperatures shall be < 986°F (530 0 C).4 Bases: Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.

The principal physical barrier shall be the fuel cladding.

Operating experience and detailed calculations of Argonaut reactors and for the HEU to LEU conversion have demonstrated that the Specification ,suffices to maintain the fuel and fuel cladding below temperatures at which fuel degradation would occur.2.2 Limiting Safety System Settings (LSSS)Applicability:

These specifications are applicable to the reactor safety system set points.Objective:

To ensure that automatic protective action is initiated before exceeding the safety limit.Specifications:

The limiting safety system settings shall be (1) Power level shall be < 119 kWth.(2) The primary coolant flow rate shall be > 41 gpm.(3) The primary coolant, (a) Inlet temperature shall be < 99°F.(b) Outlet temperature shall be < 155°F when measured at any fuel box outlet.(4) The reactor period shall be > 3 sec.Bases: Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant protective functions.

The UFTR LSSS's are established based on operating experience and safety considerations, which are established for the protection of the fuel, the fuel cladding, and the reactor core integrity.

3.0 LIMITING

CONDITIONS FOR OPERATION (LCO)3.1 Reactor Core Parameters Applicability:

These specifications apply to the parameters which describe the reactivity condition of the core.Obiectives:

To ensure that the reactor cannot achieve prompt criticality, that the fuel temperature does not reach the melting point, that the reactor can be safely shutdown under any condition and to limit the reactivity insertion rate to levels commensurate with efficient and safe reactor operation.

5 Specifications:

The reactor shall not be critical unless the following conditions exist: (1) Shutdown Margin: The minimum shutdown margin, with the most reactive control blade fully withdrawn, at core reference conditions, and with installed experiments at their most reactive state, shall be > one dollar.(2) Excess Reactivity:

The core excess reactivity at core reference conditions shall be< 1.4% Ak/k.(3) Maximum Single Blade Reactivity Insertion Rate: The reactivity insertion rate for a single control blade shall not exceed 0.06% Ak/k/sec, when determined as an average over any 10 sec of blade travel time from the characteristic experimental integral blade reactivity worth curve.Bases: Limiting conditions for operation are the lowest functional capabilities or performance levels required of equipment for safe operation of the facility.

Specification (1) ensures that a reactor shutdown can be established.

Specification (2) is based on analysis documented in SAR Chapter 4, Section 4.1.2 and Chapter 13 to prevent the possibility that an inadvertent sudden excess of reactivity insertion could release significant energy to damage the fuel or cladding.

Specification (3) limits the reactivity insertion rate to levels commensurate with efficient and safe reactor operation.

These limits are also established based on UFTR operating experience.

3.2 Reactor

Control and Safety Systems 3.2.1 Reactor Control System Applicability:

These specifications apply to the reactor control systems.Objectives:

To specify minimum acceptable capability and level of equipment for the reactor control system, range of reactivity insertion rate and interlocks to assure safe operation of the reactor.Specifications:

(1) The safety blades shall not be used to raise reactor power simultaneously with the regulating blade when the reactor control system is in the automatic mode of operation.

(2) The reactor shall not be started unless the reactor control system is operable.(3) The control blade drop time shall be < 1.5 seconds.(4) The following control blade withdrawal inhibit interlocks shall be operable for reactor operation:

6 (a) A source (startup) count rate < 2 cps (as measured by the wide range drawer operating on extended range).(b) A reactor period < 10 sec.(c) Safety channels 1 and 2 and wide range drawer calibration switches not in OPERATE condition.(d) Attempt to raise any two or more blades simultaneously when the reactor is in manual mode or two or more safety blades simultaneously when the reactor is in automatic mode.(e) Power is raised in the automatic mode at a period > 30 sec.Bases: Specification (1) limits the reactivity insertion rate to levels commensurate with efficient and safe reactor operation.

Specification (2) ensures the reactor control system operability for startup and that the automatic controller is operable if needed.Specification (3) ensures that the reactor can be shut down promptly when a scram signal is initiated.

Specifications (4) (a), (b), (d) and (e) ensure that blade movement is performed under proper monitoring with assured source count rate and safe period either under manual or automatic control. Specification (4) (c) ensures that the operator is capable of monitoring power changes during blade movement 3.2.2 Reactor Safety System Applicability:

These specifications apply to the reactor safety systems.Objective:

To ensure that sufficient information is available to the operator allowing safe operation of the reactor.Specifications:

(1) The reactor shall be shut down when the main alternating current (ac) power is not available.

(2) The high voltage applied to Safety Channels 1 and 2 neutron chambers shall be >90% of established normal value.(3) The reactor shall not be started unless the reactor safety system is operable in accordance with Table 3-1.Bases: Specification (1) ensures that the reactor is shut down during power outages.Specification (2) ensures that the operator has proper indication of power while operating the reactor. Specification (3) ensures that no operation will be performed under abnormal conditions as listed in Table 3-1 and that the necessary reactor control system trip 7 functions are operable in case of occurrence of any of these conditions.

These limits were set based on operating experience and safety considerations.

3.2.3 Reactor

Control and Safety Systems Channels Applicability:

These specifications apply to the channels for reactor control and safety systems.Objective:

To specify the minimum number and type of acceptable channels for the reactor safety system and safety related instrumentation.

Specification:

The minimum number and type of channels operable and providing information to the control room operator required for reactor operation are presented in Table 3-2.Bases: Table 3-2 specifies the minimum number of acceptable components for the reactor safety system and related instrumentation to assure the proper functioning of the reactor safety systems as specified in SAR Chapter 7.3.3 Coolant Systems Applicability:

These specifications apply to the reactor cooling system and water in contact with fuel plates or elements.Objective:

To specify minimum operating equipment for the coolant systems and to minimize corrosion of the aluminum cladding of fuel plates and activation of dissolved materials.

Specifications:

The reactor shall not be operated unless the following conditions exist: (1) The primary coolant pump and the dump valve shall be energized during reactor operations.

(2) The primary coolant flow rate shall be monitored at the return line.(3) The primary coolant core level shall be > 2 inches above the fuel.(4) The secondary coolant flow shall satisfy the following conditions when the reactor is being operated at power levels > 1 kW;(a) Power shall be provided to the well pump, and (b) Well water flow rate shall be > 60 gpm.(5) Primary water shall be demineralized, light water with a specific resistivity

> 0.5 MO-cm.8 (6) The primary equipment pit water level shall be < 1 inch above the equipment pit floor level.(7) Primary water pH shall be < 7.0.Bases: Specification (1) ensures that the reactor is shut down when the primary coolant pump is not energized.

Specification (2) and (3) ensure core is covered with water for reactor operations.

Specification (4) ensures there is heat removal capability when needed. Specifications (5) and (8) are designed to protect fuel element cladding integrity and are based upon operating experience.

At the specified quantity, the activation products (of trace minerals) do not exceed acceptable limits. Specification (6) is designed to detect and identify fission products resulting from fuel failure. Specification (7) is designed to alert the operator to potential loss of primary coolant, to prevent reactor operations with a reduced water inventory and to minimize the possibility of an uncontrolled release of primary coolant to the environs.3.4 Radiation Monitoring Systems and Radioactive Effluents 3.4.1 Area Radiation Monitors and Air Particulate Detectors Applicability:

These specifications apply to the area radiation monitors and air particulate detectors.

Objective:

To specify the minimum equipment or the lowest acceptable level of performance for the area radiation monitors and air particulate detectors.

Specifications:

The reactor shall not be operated unless the area radiation monitors and air particulate detectors are operated in accordance with Table 3-3 to ensure the operator has sufficient knowledge of the area radiation and air contamination within the cell.Bases: The area radiation monitoring system and air particulate detector(s) provide information to the operator indicating radiation and airborne contamination levels under the full range of operating conditions.

Audible indicators and alarm lights indicate (via monitored parameters) when corrective operator action is required, and (in the case of the area radiation monitors) a warning light indicates situations recommending or requiring special operator attention and evaluation.

3.4.2 Argon-41 (Ar-41) Discharge Applicability:

These specifications apply to the limits on radioactive gaseous effluents.

Objective:

To specify the limits for gaseous effluents.

Specifications:

The following operational limits are specified for the discharge of Ar-41 to the environment:

9 (1) The concentration of Argon-41 in the gaseous effluent discharge of the UFTR is determined by averaging it over a consecutive 30-day period.(2) The dilution resulting from the operation of the stack dilution fan and atmospheric dilution of the stack plume may be taken into account when calculating this concentration.

(3) When calculated as above, the discharge concentration of Argon-41 shall not exceed 1.0 x 10 8tCi/ml. Operation of the UFTR shall be such that this maximum concentration (averaged over a month) is not exceeded.Bases: Argon-41 discharges are limited to a monthly average which is less than the effluent concentration limit in 10 CFR 20, Appendix B, Table 2.3.4.3 Reactor Vent/Stack Monitoring System Applicability:

These specifications apply to the gaseous effluent monitoring system.Objective:

To specify the minimum equipment or the lowest acceptable level of performance for the gaseous effluent monitoring systems.Specifications:

(1) The reactor shall not be operated if the reactor vent system is not operating.

The air drawn through the reactor vent system shall be continuously monitored for gross count rate of radioactive gases. The output of the monitor shall be indicated and recorded in the control room. Operable functions and alarm settings shall be as delineated in Table 3-3.Bases: The vent/stack monitoring system provides information to the operator on the condition of the air entering the stack from the reactor vent system. Audible indicators and alarm lights indicate (via monitored parameters) when corrective operator action is required, and (in the case of the area radiation monitors) a warning light indicates situations recommending or requiring special operator attention and evaluation.

3.4.4 Liquid

Effluents Discharge Applicability:

These specifications apply to the limits on radioactive liquid effluents released to the environment.

Objective:

To specify the limits for liquid effluents.

Specifications:

The following operational limits are specified for the discharge of liquid effluents to the environment:

10 (1) The liquid effluent from the waste water holdup tank shall be sampled and the radioactivity measured before release to the sanitary sewage system which is allowed in conformance with 10 CFR 20.1301.(2) Releases of radioactive effluents from the waste water holdup tank shall be in compliance with the limits specified in 10 CFR 20, Appendix B, Table 2, Column 2, as specified in 10 CFR 20.1302.Bases: Liquid and solid radioactive wastes are regulated and controlled to assure compliance with legal requirements.

3.5 Reactor

Vent System Applicability:

These specifications apply to the equipment required for controlled release of gaseous radioactive effluent to the environment via the stack or its confinement within the reactor cell.Objective:

To limit the amount and concentration of radioactivity in the effluent from the reactor cell and reduce the back leakage of radioactivity into the reactor cell under normal operations and from the cell under emergency conditions.

Specifications:

(1) The reactor vent system shall be operated at all times during reactor operation.

In addition, the vent system shall be operated until the stack monitor indicates less than 10 counts per-second (cps) unless otherwise-indicated by facility conditions to include loss of building electrical power, equipment failure or maintenance, cycling console power to dump primary coolant or to conduct tests and surveillances and initiating the evacuation alarm for tests and surveillances including emergency drills and demonstrations.

The reactor vent system shall be immediately secured upon detection of the following:

a failure in the monitoring system, a failure of the absolute filter, or an unanticipated high stack count rate.(2) The reactor vent system shall be operable whenever the reactor is operating.

(3) The air conditioning/ventilation system and reactor vent system shall automatically shut off whenever the reactor building evacuation alarm is automatically or manually actuated (4) The diluting fan shall be operated whenever the reactor is operating and as otherwise specified in these Technical Specifications.

(5) All doors to the reactor cell shall be locked while the reactor is operating.

Transit is permitted through the air lock and control room doors.II (6) The reactor vent system shall have a backup means for quantifying the radioactivity in the effluent during abnormal or emergency operating conditions where venting could be used to reduce cell radionuclide concentrations for ALARA considerations.

Bases: Under normal conditions, to affect controlled release of gaseous activity through the reactor vent system, a negative cell pressure is required so that any building leakage will be inward. Under normal shutdown conditions with significant Argon-41 inventory in the reactor cavity, operation of the core vent system prevents unnecessary exposure from gas leakage back into the cell. Under emergency conditions, the reactor vent system will be shut down and the damper closed, thus minimizing leakage of radioactivity from the reactor cell unless venting is required.3.6 Limitations on Experiments

3.6.1 Reactivity

Limits Applicability:

These specifications apply to all experiments or experimental devices installed in the reactor core or its experimental facilities.

Objectives:

To assure operational safety and prevent damage to the reactor facility, reactor fuel, reactor core, and associated equipment; to prevent exceeding the reactor safety limit; and to minimize potential personnel and equipment hazards from experimental devices.Specifications:

(1) The absolute reactivity worth of any single experiment shall be < 0.6% Ak/k.(2) The total absolute reactivity worth of all experiments shall be < 1.4% Ak/k.(3) When determining the absolute reactivity worth of an experiment, no credit shall be taken for temperature effects.(4) An experiment shall not be inserted or removed unless all the control blades are fully inserted or its absolute reactivity worth is known to be less than that which could cause a positive 20-see stable period.Bases: These specifications generally ensure that an adequate review process is followed to assure the safe operation, proper conditions, and adherence to procedures for all experiments.

The classification of experiments clearly delineates the responsibility for approving experiments according to their potential hazards, to ensure that potentially hazardous experiments are analyzed for their safety implications, and that appropriate procedures are established for their execution.

The reactivity limitations on experiments are established to prevent prompt criticality by limiting the worth of movable experiments, to prevent a reactivity insertion larger than the stipulated maximum step 12 reactivity insertion in the accident analysis, and to allow for reactivity control of experiments within the reactor control system capabilities (20-sec positive period limitation).

3.6.2 Materials

Applicability:

These specifications apply to all experiments or experimental devices installed in the reactor core or its experimental facilities.

Objectives:

To assure operational safety and prevent damage to the reactor facility, reactor fuel, reactor core, and associated equipment; to prevent exceeding the reactor safety limit; and to minimize potential personnel and equipment hazards from experimental devices.Specifications:

(1) Explosive materials shall not be irradiated (2) Thermal-Hydraulic Effects -Experiments shall be designed so that during normal operation, or failure, the thermal hydraulic parameters of the core do not cause the safety limit to be exceeded.(3) Chemical Effects -Experiments shall be designed so that during normal operation, or failure, the physical barrier described in Section 2.1 will not be compromised by either chemical or blast effects from the experiment.

(4) Fueled Experiments

-A limit should be established on the inventory of fission products in any experiment containing fissile material, according to its potential hazard and as determined by the RSRS.(5) Radioactive Releases from Experiments

-Class III and Class IV experiments (see Section 6.5) shall be evaluated for their potential release of airborne radioactivity and limits shall be established for the permissible concentration of radioisotopes in the experiments, according to the 10 CFR 20 limitations for exposure of individuals in restricted and unrestricted areas.Bases: These specifications generally ensure that an adequate review process is followed to assure the safe operation, proper conditions, and adherence to procedures for all experiments.

The classification of experiments clearly delineates the responsibility for approving experiments according to their potential hazards, to ensure that potentially hazardous experiments are analyzed for their safety implications, and that appropriate procedures are established for their execution.

These specifications limits the type of materials and in the case of fissile materials, the amount that can be irradiated in the reactor, to their potential hazard and the reactor system's capability to handle a potential release to the cell environment.

13

3.7 Reactor

Building Evacuation Alarm Applicability:

These specifications apply to the systems and equipment required for the evacuation of the reactor cell and the reactor building (including the reactor annex).Objective:

To specify conditions to actuate the evacuation alarm.Specifications:

(1) The reactor cell and the reactor building shall be evacuated when any of the following conditions exist: a Two area radiation monitors alarming high, causing an automatic actuation of the evacuation alarm.b An air particulate detector is in a valid alarm condition, causing the operator to manually actuate the evacuation alarm.c A potentially hazardous radiological condition exists, causing the operator to manually actuate the evacuation alarm.Bases: To provide early and orderly evacuation of the reactor cell and the reactor building and to minimize radioactive hazards to the operating personnel and reactor building occupants.

3.8 Fuel and Fuel Handling Applicability:

These specifications apply to the arrangement of fuel elements in core and in storage, as well as the handling of fuel elements.Obiectives:

To establish the maximum core loading for reactivity control purposes, to establish proper fuel storage conditions and to establish fuel performance and fuel handling specifications with regard to radiological safety considerations.

Specifications:

(1) Fuel elements exhibiting release of fission products because of cladding failure shall, upon positive identification, be removed from the core.(2) The reactor shall not be operated if there is evidence of fuel element failure other than to locate the failed fuel (3) Prior to the removal of the last two levels of concrete shielding above the core, the reactor will be shut down for at least three days.14 (4) Any change in Core Configuration has been analyzed.(5) Fuel Inspections are up-to-date based on surveillance requirements (see Section 4.8).Bases: The reactor systems do not have adequate engineering safeguards to continue operating with a detectable release of fission products into the primary coolant.Specification (3) limits the possible/potential consequences of fuel handling accidents.

Specification (4) states that the UFTR core design is fixed and any changes require the necessary analysis and documentations according to NUREG 1537. Specification (5)Fuel inspection is necessary to assess its condition/health.

3.9 Radiological

Environmental Monitoring Program Applicability:

This specification applies to the environmental radioactivity surveillances and surveys conducted by UFTR personnel and Radiation Control and Radiological Services Department personnel.

Objectives:

To ensure that the radiological environmental impact of reactor operations is as low as reasonably achievable (ALARA); it is conducted in addition to the radiation monitoring and effluents control specified under Section 3.4 of these Technical Specifications.

Specifications:

The Radiological Environmental Monitoring Program shall be conducted as specified below and under the supervision of the Radiation Control Officer.The reactor shall not be considered operable if the following radiation surveys, using portable radiation monitors, are not complete: (a) Surveys measuring radiation dose rates in the restricted area shall be conducted.

Dose rates shall be maintained within 10 CFR 20 limits for radiation workers.(b) Surveys measuring the radiation dose rates in the unrestricted areas surrounding the UFTR complex shall be conducted.

Dose rates shall be within 10 CFR 20 limits for the general public.Bases: The bases for establishing the Radiological Environmental Surveillance Program are the established limits for internal and external radiation exposure and requirements that radiation doses be maintained ALARA and the necessity to confirm and document that UFTR operations are conducted to be within the established limits.15 3.10 UFTR Shield Tank Applicability:

This specification applies to the shield tank water level during operation.

Objectives:

To specify the minimum water level required in the shield tank.Specifications:

The reactor shall not be operated if the water level in the shield tank is > 6 inches below the established normal value.Bases: This specification is established to protect reactor personnel from potential external radiation hazards caused by loss of biological shielding.

16 Table 341 Specification for Reactor Safety System Trips Specification Type of safety system trip Automatic Trips Period < 3 sec Full Power > 119% of full power (Safety Channel 1 and 2) Full Loss of chamber high voltage (> 10%) Full Loss of electrical power to control console Full Primary cooling system Blade-drop Loss of primary pump power Low water level in core (< 42.5")No outlet flow Low inlet water flow (< 41 gpm)Secondary cooling system (> 1 kW) Blade-drop Loss of flow (well water < 60 gpm)Loss of secondary well pump power High primary coolant inlet temperature

(> 99°F) Blade-drop High primary coolant outlet temperature

(> 155°F) Blade-drop Shield tank low water level (6" below established normal level) Blade-drop Ventilation system Blade-drop Loss of power to stack dilution fan Loss of power to core vent fan Manual Trips Manual scram bar Blade-drop Console key-switch OFF (two blades off bottom) Full 17 Table 3-2 Minimum Number and Type of Channels Operable Channel Number of operable Safety 1 and 2 power channels 2 Linear Channel (with auto controller as appropriate) 1 Log N and period channel* I Startup channel* I Blade position indicator 4 Coolant flow indicator I Coolant temperature indicator Primary 7 Secondary 1 Core level 1 Ventilation system Core vent annunciator 1 Dilute fan annunciator I Dilute fan rpm 1*Subsystems of the wide range drawer 18 Table 3-3 Radiation Monitoring System* Settings Type No. of Required Alarm(s) Setting Purpose Operable Functions Area Radiation 3 detecting 5 mR/hr low level Detect/alarm/record Monitors 2 audio alarming 20 mR/hr high level low and high level 2 recording external radiation Air Particulate 1 detecting Range adjusted according Detect/alarm/record Monitors 1 audio alarming to APD** type airborne radioactivity 1 recording (according.

to monitoring in the reactor cell requirements)

Stack Radiation I detecting (1) Fixed alarm at 4000 Detect/alarm/record Monitor I audio alarming cps release of gaseous 1 recording (2) Adjustable alarm per radioactive effluents power level in the reactor vent duct to the environs*Note: For maintenance or repair, the required radiation monitors may be replaced by suitable portable instruments provided the intended function is being accomplished."Service, calibration, and testing interruptions for brief periods are permissible when the reactor is not in operation**Air Particulate Detector 19

4.0 SURVEILLANCE

REQUIREMENTS

4.1 Reactor

Core Parameters Applicability:

These specifications apply to the surveillance activities required for reactivity parameters.

Objective:

To specify the frequency and type of testing to assure that reactor core parameters conform to specifications in Section 3.1.Specifications:

The reactivity worth and reactivity insertion rate of each control blade, the shutdown margin and excess reactivity shall be measured annually or whenever physical or operational changes create a condition requiring reevaluation of core physics parameters.

Bases: The measurements specified are sufficient to provide assurance that the reactor core parameters are maintained within the limits specified in Section 3.1.4.2 Reactor Control and Safety Systems 4.2.1 Reactor Control System Applicability:

These specifications apply to the surveillance activities required for the reactor control systems.Objective:

To specify the frequency and type of testing or calibration to assure that reactor control system operating parameters conform to specifications in Section 3.2.1.Specifications:

(1) Control blade drop times, from the fully withdrawn position, shall be measured semiannually.

If maintenance is performed on a blade, the drive mechanism, or associated electronics, the blade-drop time shall be measured before the system is considered operable.(2) The control blade full withdrawal and controlled insertion times shall be measured semiannually.

(3) The control blade withdrawal inhibit interlock checks shall be performed as listed in Table 4-2 to ensure the system is operable.(4) The mechanical integrity of the control blades and drive system shall be inspected during each in core inspection but shall be fully checked at leastonce every 10 years.20 (5) Following maintenance or modification to the control blade system, a calibration of the affected portion of the system, including verification of control blade drive speed, shall be performed before the system is to be considered operable.(6) The reactor shall not be started unless (a) The weekly checkout has been satisfactorily completed within 7 days prior to startup, (b) A daily checkout is satisfactorily completed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to startup, and (c) No known condition exists that would prevent successful completion of the weekly or daily check.(7) The limitations established under Specification (6) (a) and (b) can be deleted if a reactor startup is made within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of a normal reactor shutdown on any one calendar day.Bases: The frequency and type of test or calibration are defined based on operating experience and/or in accordance with ANSI/ANS-15.1-2007 to assure proper functioning of the systems and equipment that comprise the reactor control system.4.2.2 Reactor Safety System Applicability:

These specifications apply to the surveillance activities required for the reactor safety systems.Objective:

To specify the frequency and type of testing or calibration to assure that reactor safety system operating parameters conform to specifications in Section 3.2.2.Specifications:

(1) Safety system scram functions or components shall be determined to be operable in accordance with Table 4-1.(2) The following channels shall be calibrated annually: (a) Log N -period channel (b) Power level safety channels (2)(c) Linear power level channel (d) Primary coolant flow measuring system 21 (e) Primary coolant temperature measuring system (3) Following maintenance or modification to the reactor safety system, a channel test and calibration of the affected channel shall be performed before the reactor safety system is considered operable.Bases: The frequency and type of test or calibration are defined based on operating experience and/or in accordance with ANSI/ANS-15.1-2007 to assure proper functioning of the systems and equipment that comprise the reactor safety system.4.3 Coolant Systems Applicability:

These specifications apply to the surveillance activities required for the reactor coolant system.Objective:

To specify the frequency and type of testing or calibration to assure the reactor coolant system conforms to the specifications presented in Section 3.3 Specifications:

(1) Safety system scram functions or components shall be determined to be operable in accordance with Table 4-1.(2) The primary coolant dump valve shall be tested during the weekly checkout.(3) The primary equipment pit water level sensor shall be tested during the weekly checkout.(4) The primary water resistivity shall be measured during the daily and weekly checkouts.

(5) The primary water pH value shall be measured during the weekly checkout.(6) The primary water radioactivity shall be measured during the weekly checkout for gross P3-y and gross a activity.Bases: These specifications assure that necessary limits are maintained on fission products and other activated materials in primary and secondary coolant samples to provide assurance that the facility is operating in a safe and effective manner.Specifications (1) and (2) ensure all trip signals are tested for the reactor coolant system.The frequency and type of monitoring are based on operating experience.

4.4 Radiation

Monitoring Systems and Radioactive Effluents 22 Applicability:

These specifications apply to the surveillance activities required for the radiation monitoring system and effluents released from the facility.Objective:

To specify frequency and type of testing to assure that the radiation monitoring system and effluent releases conform to the specifications in Section 3.4.Specifications:

(1) The area radiation monitor channels, the stack monitor, and the air particulate monitor shall be verified to be operable before each reactor startup as required by the daily checkout.

Calibration of radiation monitoring channels shall be performed quarterly.

(2) The Ar-41 concentration in the stack effluent shall be measured semiannually.

(3) Releases of liquid effluents from the waste water holdup tank shall be sampled and the radioactivity measured before release to the sanitary sewage system which is allowed in conformance with 10 CFR 20 regulations.

Bases: Specification (1) assures the monitors are operable.

Specification (2) provides the basis for limiting energy generation to assure Ar-41 releases are in accordance with 10 CFR 20, Appendix B, Table 2. Specification (3) ensures compliance with 10 CFR 20 for liquid releases from the site.4.5 Reactor Vent System Applicability:

These specifications apply to the surveillance requirements for the reactor vent system.Objective:

To specify the frequency and type of testing to assure the reactor vent system conforms to the specifications presented in Section 3.5.Specifications:

(1) The reactor vent system flow rates shall be measured annually.(2) The interlock of the core vent system damper being closed if the diluting fan is not operating shall be tested as part of the weekly checkout.(3) All doors to the cell will be verified locked prior to operation of the reactor.Bases: These specifications assure the reactor vent system is operating as specified.

The frequency and type of monitoring are based on operating experience and ANSI/ANS-15.1- 2007.23

4.6 Limitations

on Experiments Applicability:

This specification applies to the surveillance requirements for experiments installed in the UFTR core.Objective:

To prevent the performance of experiments or irradiations that could damage the reactor or release an excessive amount of radioactivity.

Specifications:

(1) Surveillance to ensure that experiments meet the requirements of Section 3.6 shall be conducted before inserting each experiment into the reactor.(2) The reactivity worth of an experiment shall be determined at approximately 1 W power level or as appropriate within limiting conditions for operation, before continuing reactor operation with the experiment.

Bases: Measurements of the reactivity worth of an experiment shall verify that the experiment is within the authorized reactivity limits.4.7 Reactor Building Evacuation Alarm Applicability:

These specifications apply to the surveillance requirements for the reactor building evacuation alarm.Objective:

To specify the frequency and type of testing to assure the building alarm evacuation conforms to the specifications presented in Section 3.7.Specifications:

(1) The automatic actuation of the building evacuation alarm in coincidence with actuation of the high level alarm on two area monitors and the manual actuation of the evacuation alarm shall be tested as part of the weekly checkout.(2) The automatic shutoff of the air handling system and the reactor vent system in coincidence with the building evacuation alarm shall be tested as part of the weekly checkout.Bases: Specification (1) ensures that the actuation of the building evacuation alarm is operable to alert occupants of the need to evacuate.

Specification (2) ensures that the system responds correctly to a known input to assure isolation of the cell atmosphere upon actuation of the evacuation alarm.4.8 Surveillance Pertaining to Fuel Applicability:

These specifications apply to fuel installed in the core.24 Objective:

To verify integrity of the fuel.Specifications:

(1) The in-core reactor fuel elements shall be inspected every 10 years, in a randomly chosen pattern, as deemed necessary.

At least 8 elements will be inspected.

Bases: Specification (1) ensures the integrity of the fuel and Specification (2) assures that reactor and support staff is properly qualified to perform fuel handling and related activities.

4.9 Radiological

Environmental Monitoring Program Applicability:

This specification applies to the environmental radioactivity surveillances and surveys conducted by UFTR personnel and Radiation Control and Radiological Services Department personnel.

Objectives:

To ensure that the radiological environmental impact of reactor operations is as low as reasonably achievable (ALARA); it is conducted in addition to the radiation monitoring and effluents control specified under Section 3.4 of these Technical Specifications.

Specifications:

The Radiological Environmental Monitoring Program shall be conducted as specified below and under the supervision of the Radiation Control Officer.(1) Monthly environmental radiation dose surveillance outside the restricted area shall be conducted by Radiological Services Department personnel.

(2) Radioactivity surveillance of the restricted area (reactor cell) shall be conducted during the weekly checkout.(3) The radiation surveys, using portable radiation monitors, shall be conducted quarterly and at any time a change in the normal radiation levels is observed.Bases: The bases for establishing the Radiological Environmental Surveillance Program are to establish limits for internal and external radiation exposure and requirements that radiation doses be maintained ALARA and the necessity to confirm and document that UFTR operations are conducted to be within the established limits.4.10 UFTR Shield Tank Applicability:

This specification applies to the shield tank water level during operation.

25 Objectives:

To assure a minimum amount of water in the shield tank to minimize dose rates to the operators.

Specifications:

The water level detector in the shield tank shall be checked periodically based on Table 4-1 to ensure it remains operable.Bases: This specification is established to protect reactor personnel from potential external radiation hazards caused by loss of biological shielding.

26 Table 4-1 Safety System Operability Tests Component or Scram Function Frequency*

Log-N period channel Before each reactor startup following a Power level safety channels shutdown in excess of 6 hr, and after repair or deenergization caused by a power outage.10% reduction of safety channels high voltage Quarterly Loss of electrical power to console Quarterly Loss of primary coolant pump power Quarterly Loss of primary coolant level Quarterly Loss of primary coolant flow Quarterly High primary coolant inlet temperature Daily checkout High primary coolant fuel box outlet Daily checkout temperature Loss of secondary coolant flow (at power levels Daily checkout above I kW)Loss of secondary coolant well pump power Quarterly Loss of shield tank water level Quarterly Loss of power to vent system and dilution fan Quarterly Console key to OFF position Daily checkout Manual scram bar Daily checkout*Surveillance may not be performed with the specified frequency because of physical or administrative limitations, including equipment failure and maintenance activities; however, surveillance shall be performed before resuming normal operations.

Whenever an unscheduled shutdown occurs, an evaluation shall be conducted to determine whether the safety limit was exceeded.27 Table 4-2 Control Blade Withdrawal Inhibit Interlocks Checks Inhibit Limit Frequency Reactor Period < 10 sec Daily Checkout Safety Channels and Wide Range -Daily Checkout Drawer not in OPERATE position Multiple blade withdrawal Any 2 or more blades Daily Checkout simultaneously in Manual Any 2 safety blades in Automatic Source count rate < 2 cps Verification only when.count rate < 2cps during daily checkout 28

5.0 DESIGN

FEATURES 5.1 Site and Facility Description The UFTR is located on the University of Florida campus, in Gainesville, Florida, in the immediate vicinity of the buildings housing the College of Engineering and the College of Journalism.

The Nuclear Science Center, which houses the Department of Nuclear and Radiological Engineering, is annexed to the reactor building.The reactor is housed in a reinforced concrete cell in the reactor building.

The reactor building is a "vault-type" building as defined in 10 CFR 73.2. The reactor building is divided into two distinct parts based upon the difference in utilization and their structure.

The overall reactor building measures approximately 60 ft by 80 ft inside. The reactor cell area is 30 ft by 60 ft with 29 ft of head room, located at the north end of the building.The rest of the building is used for research laboratories, faculty offices, and graduate study areas. For further information, refer to FSAR, Section 2.5.2 Reactor Coolant System 5.2.1 Primary Cooling System The primary coolant is demineralized light water, which is normally circulated in a closed loop. The flow is from the 200-gal storage (dump) tank to the primary coolant pump;water is then pumped through the primary side of the heat exchanger and to the bottom of the fuel boxes, upward past the fuel plates to overflow pipes located about 6 in. above the fuel, and into a header for return to the storage tank. A purification loop is used to maintain primary water quality. The purification loop pump circulates about 1 gpm of primary water, drawn from the discharge side of the heat exchanger, through mixed-bed ion-exchange resins and a ceramic filter. The purification loop pump automatically shuts.off when the primary coolant pump is operating, since flow through the purification system is maintained.

Primary coolant may be dumped from the reactor fuel boxes by opening an electrically operated solenoid dump valve, which routes the water to the dump tank. A pressure surge of about 2 psi above normal in the system will also result in a water dump by breaking a graphite rupture disc in the dump line. This drains the water to the primary equipment pit floor actuating an alarm in the control room. The primary coolant system is instrumented as follows: (1) Temperature sensing device at each fuel box and the main inlet and outlet (eight total), alarming and recording in the control room;(2) A flow sensing device in the main inlet line, alarming and displayed in the control room;(3) A flow sensing device (no flow condition) in the outlet line, alarming in the control room;29 (4) Resistiyity probes monitoring the inlet and outlet reactor coolant, alarming and displayed in the control room; and (5) An equipment pit water level monitor, alarming in the control room.5.2.2 Secondary Cooling System The well secondary cooling system is the main system used for removal of reactor generated heat to the environment.

A deep well furnishes about 200 gpm of cooling water to the shell side of the heat exchanger, removing primary heat and rejecting it to the storm sewer. Flow indications in the control room are 140 gpm as a warning and 60 gpm to initiate a trip at or above I kW after an approximate 10-sec warning. The secondary coolant system inlet and outlet temperatures are monitored by temperature sensing devices, with recording and alarm functions in the control room.5.3 Reactor Core and Fuel The UFTR fuel assemblies are placed in six aluminum boxes arranged in two parallel rows of three boxes each, separated by about 30 cm of graphite.

The fuel boxes are surrounded by a 5 ft x 5 ft x 5 ft reactor grade graphite assembly.The present UFTR license is approved for a core comprised of 22 full fuel assemblies, one partial fuel assembly, and one dummy assembly.

Currently, the core operates with 22 full fuel assemblies and two dummy assemblies.

Each fuel bundle contains 14 MTR-type fuel plates. A partial assembly can contain up to 13 plates. Each fuel plate is comprised of fuel meat and cladding.

Fuel meat is made of uranium silicide and aluminum, and uranium has a nominal enrichment of 19.75%.The core is cooled with water, which also provides neutron moderation.

The graphite provides both moderation and reflection.

UFTR is shutdown via two mechanisms including:

control blade drop and water dump. There are 4 control blades including three safety blades and one regulating blade. The regulating blade is used for change of power.The tops of the fuel boxes are covered during operations at power above 1 kW, by the use of the shield plugs and/or aluminum covers secured to the top of the fuel boxes. These covers function to prevent physical damage of the fuel, to minimize evaporation/leakage of water from the top of the fuel boxes, and to minimize entrapment of Ar-41 in the coolant water for radiological protection purposes.The UFTR design yields negative coefficients of reactivity for coolant void, coolant temperature and fuel temperature.

The core has an excess reactivity of 0.925 %Ak/k, which is designed to operate for 20 years at 4 four hours per day at full power.The operating region for the UFTR (i-e., combination of flow rate, inlet coolant temperature, power) is set to prevent the occurrence of the Onset of Nucleate Boiling (ONB), thereby avoiding any flow instability.

30

5.4 Fissionable

Material Storage New and irradiated fuel is stored in locations specified in the UFTR Physical Security Plan. Fuel elements and fueled devices shall be stored and handled out of core in a geometry such that keff < 0.9 under optimum conditions of moderation and reflection.

Irradiated fuel elements and fueled devices shall be stored so that temperatures do not exceed the Safety Limit.6.0 ADMINISTRATIVE CONTROLS 6.1 Organization

6.1.1 Structure

The organization for the management and operation of the reactor facility shall include the structure indicated in Figure 6-1. Job titles are shown for illustration and may vary.Four levels of authority are provided.Level 1 -Individuals responsible for the reactor facility's licenses, charter, and site administration.

Level 2 -Individual responsible for reactor facility management.

Level 3 -Individual responsible for reactor operations, and supervision of day-to-day facility activities.

Level 4 -Reactor operating staff (Senior Reactor Operator, Reactor Operator and trainees).

The Reactor Safety Review Subcommittee (RSRS) is appointed by, and shall report to, the Chairman of the Radiation Control Committee.

The Chairman of the Radiation Control Committee reports to the Director of Environmental Health and Safety, who reports to the level 1 of the UFTR organization chart. Radiation safety personnel shall report to Level 2 or higher.6.1.2 Responsibility Responsibility for the safe operation of the reactor facility shall be with the chain of command established in Figure 6-1. In addition to having responsibility for the policies and operation of the reactor facility, individuals at various management levels shall be responsible for safeguarding the public and facility personnel from undue radiation exposures, and for adhering to all requirements of the operating license, charter, and technical specification.

In all instances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications.

Functions delineated in Figure 6-1 may be fulfilled.

by combinations of personnel when a position is unfilled.31

6.1.3 Staffing

The minimum staffing when the reactor is not secured shall be as follows: (1) A Reactor Operator shall be in the control room.(2) A second person shall be present at the facility complex able to carry out prescribed written instructions including instructions to initiate the first stages of the emergency plan, including evacuation and initial notification procedures.

Unexpected absence for two hours is acceptable provided immediate action is taken to obtain a replacement.

(3) A designated Senior Reactor Operator shall be readily available on call. "Readily Available on Call" means an individual who: (a) Has been specifically designated and the designation is known to the operator on duty, (b) Keeps the operator on duty informed of where he/she may be rapidly contacted and the phone number or other means of communication available, and (c) Is capable of getting to the reactor facility within a reasonable time under normal conditions (e.g., 30 min or within a 15 miles radius).A list of reactor facility personnel by name and telephone number shall be readily available in the Control Room for use by the operator.

The list shall include: (1) Management personnel, (2) Radiation safety personnel, and (3) Other operations personnel.

Events requiring the presence of a Senior Reactor Operator are: (1) Initial startup and approach to power, (2) All fuel or control-blade relocations within the reactor core region, (3) Relocation of anyin core experiment with a reactivity worth > 1 dollar, and (4) Recovery from unplanned or unscheduled shutdown or a significant power reduction.

32

6.1.4 Selection

and Training of Personnel The selection, training, and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection and Training of Personnel for Research Reactors, ANSI/ANS-l15.4-2007 Section 4.7.6.2 Review and Audit A method for the independent review and audit of the safety aspects of reactor facility operations shall be established to advise management.

The review and audit functions of the UFTR operations are conducted by the Reactor Safety Review Subcommittee (RSRS).6.2.1 Composition and Qualifications The RSRS shall be composed of a minimum of five members, including the Reactor Manager and Radiation Control Officer (both ex-officio voting members), the Chairman of the Nuclear and Radiological Engineering Department and two others familiar with the operation of reactors and with the design of the UFTR and radiological safety, at least one of whom should be from outside the Department of Nuclear and Radiological Engineering.

6.2.2 Charter

and Rules The review and audit functions shall be conducted in accordance with the following established charter: Designation:

The name of the Subcommittee is Reactor Safety Review Subcommittee (RSRS).Accountability:

The RSRS is a Subcommittee of and reports to the University Radiation Control Committee (URCC). The URCC provides radiological safety recommendations to the Director of Environmental Health and Safety.Scope: The RSRS shall be responsible for the review of safety-related issues pertaining to the University of Florida Training Reactor (UFTR).Purpose: The purpose of the RSRS is to ensure the safe operation of the UFTR through the discharge of the Subcommittee review and audit function.Membership:

(1) The two technical personnel should be recommended to the Chairman of the URCC by the Chairman of the Department of Nuclear and Radiological Engineering.

Any member may designate a duly qualified representative from a standing URCC approved list to act in their absence.33 (2) An Executive RSRS Committee will consist of the Reactor Manager, University Radiation Control Officer and Chairman of the RSRS.(3) The Chairman of the RSRS will be appointed by the Chairman of the URCC. The Chairman of the RSRS is an ex-officio voting member of the URCC and will serve as liaison between the RSRS and the URCC.(4) Members appointed to the RSRS shall be reviewed annually, and as appropriate, new appointments made.Meetings (1) At least one meeting shall be held quarterly.

Meetings may be held more frequently as circumstances warrant, consistent with the effective monitoring of facility operations as determined by the RSRS Chairman.(2) Review of draft minutes will be completed before subsequent meetings, at which time they will be submitted for approval.

Responsibility to ensure that this is done falls upon the RSRS Chairman.

The RSRS Chairman is charged with the responsibility to assure that the minutes are submitted for approval in a timely manner.(3) A quorum shall consist of at least three members (and 50% or more of the RSRS membership) and at least three members must agree when voting, regardless of the number present, and the operating staff should not constitute a majority.6.2.3 Review Function The following items shall be reviewed: (1) Determination that proposed changes in equipment, systems, tests, experiments, or procedures do not have safety significance; i.e., no prior NRC approval is needed as indicated in 10 CFR 50.59 rule.(2) All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment or systems having safety significance; (3) All new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity; (4) Proposed changes in technical specifications of the license;(5) Violations of technical specifications of the license;34 (6) Violations of internal procedures or instructions having safety significance; (7) Operating abnormalities having safety significance; (8) Reportable occurrences (see Section 6.7.2); and (9) Audit reports and annual facility reports.A written report or minutes of the findings and recommendations of the RSRS or its executive committee shall be submitted in a timely manner (no later than the subsequent RSRS meeting) to the Chair of the Radiation Control Committee.

6.2A Audit Function The audit function shall include selective (but comprehensive) examination of operating records, logs, and other documents.

Where necessary, discussions with cognizant personnel shall take place. In no case shall the individual immediately responsible for the area audit in the area. The following items shall be audited: (1) Facility operations for conformance to the technical specifications and applicable license or charter conditions, at least once per calendar year;(2) The requalification and recertification program for the operating staff, at least once every other calendar year;(3) The results of action taken to correct those deficiencies in reactor facility equipment, systems, structures, or methods of operations that could affect reactor safety, at least once per calendar; and (4) The reactor facility emergency plan, and implementing procedures at least once every other calendar year.Deficiencies uncovered that affect reactor safety shall immediately be reported to the Chair of Radiation Control Committee and the Dean of the College of Engineering.

A written report of the findings of the audit shall be submitted to the Dean of the College of Engineering, Chair of Radiation Control Committee, and the review and audit group members within three (3) months after the audit has been completed.

6.3 Radiation

Safety and ALARA (As Low As Reasonably Achievable)

The Radiation Control Committee and the Radiation Control Officer shall be responsible for the implementation of the Radiation Control Program for the UFTR. The primary purpose of the program is to assure radiological safety for all University personnel and the surrounding community.

35

6.4 Procedures

The UFTR facility shall be operated and maintained in accordance with approved written procedures.

All procedures and major revisions thereto shall be reviewed and approved by the Director of the UFTR before going into effect.The following types of written procedures shall be maintained:

(1) Normal startup, operation and shutdown procedures for the reactor to include applicable check-off lists and instructions; (2) Fuel loading, unloading, and movement within the reactor;(3) Procedures for handling irradiated and un-irradiated fuel elements;(4) Routine maintenance of major components of systems that could have an effect on reactor safety;(5) Surveillances required by the technical specifications; (6) Personnel radiation protection, consistent with applicable regulations; (7) Administrative controls for operations and maintenance and for the conduct of irradiations and experiments that could affect reactor safety or core reactivity; (8) Implementation of the Emergency Plan;(9) Procedures that delineate the operator action required in the event of specific malfunctions and emergencies; and (10) Procedures for flooding conditions in the reactor facility, including guidance as to when the procedure is to be initiated and guidance on reactivity control.Substantive changes to the above procedures shall be made effective onlyafter documented review by the RSRS and approval by the facility director (Level 2) or designated alternates.

Minor modifications to the original procedures which do not change their original intent may be made by the reactor manager (Level 3) or higher, but modifications must be approved by Level 2 or designated alternates within 14 days.Temporary deviations from the procedures may be made by a senior reactor operator, in order to deal with special or unusual circumstances or conditions.

Such deviations shall be documented and reported to Level 2 or designated alternates by the nfext working day.6.5 Experiment Review and Approval (1) Experiment review and approval shall be conducted as specified under Section 3.6, "Limitations on Experiments", of these Technical Specifications.

36 (2) Substantive changes to previously approved experiments with safety significance shall be made only after review by the RSRS, approval in writing by Level 2 or designated alternates.

Minor changes that do not significantly alter the experiment may be approved by Level 3 or higher.(3) Approved experiments shall be carried out in accordance with established approved procedures.

(4) Classification of Experiments Class I --Routine experiments, such as gold foil irradiation.

This class shall be approved by the reactor manager; the radiation control officer may be informed if deemed necessary.

Class II- Relatively routine experiments that need to be documented for each new group of experimenters performing them, or whenever the experiment has not been carried out for one calendar year or more by the original experimenter, and that pose no hazard to the reactor, the personnel, or the public. This class shall be approved by the reactor manager and the radiation control officer.Class III- Experiments that pose significant questions regarding the safety of the reactor, personnel, or the public. This class shall be approved by the reactor manager and the radiation control officer, after review and approval by the Reactor Safety Review Subcommittee (RSRS).Class IV- Experiments that have a significant potential for hazard to the reactor, the personnel, or the public. This class shall be approved by the reactor manager and radiation control officer after review and approval by the RSRS and specific emergency operating instructions shall be established for conducting the experiments.

6.6 Required

Actions 6.6.1 Action To Be Taken, in Case of Safety Limit Violation (1) The reactor shall be shut down, and reactor operations shall not be resumed until authorized by the Nuclear Regulatory Commission (NRC).(2) The safety limit violation shall be promptly reported to Level 2 or designated alternates.

(3) The safety limit violation shall be reported to the Nuclear Regulatory Commission.

37 (4) A safety limit violation report shall be prepared.

The report shall describe the following: (a) Applicable circumstances leading to the violation including, when known, the cause and contributing factors;(b) Effect of the violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public; and (c) Corrective action to be taken to prevent recurrence.

The report shall be reviewed by the RSRS and any follow-up report shall be submitted to the NRC when authorization is sought to resume operation of the reactor.6.6.2 Action To Be Taken in the Event of an Occurrence of the Type Identified in Section 6.7.2(2) and 6.7.2(3).(1) Reactor conditions shall be returned to normal in case of a non-reportable malfunction, otherwise reactor shall be shut down. If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by Level 2 or designated alternates.

(2) Occurrence shall be reported to Level 2 or designated alternates and to the NRC as required.(3) Occurrence shall be reviewed by the RSRS at their next scheduled meeting.6.7 Reports In addition to the requirements of the applicable regulations, reports shall be made to the NRC as follows: 6.7.1 Operating Reports Routine annual reports covering the activities of the reactor facility during the previous calendar year shall be submitted to the NRC within six (6) months following the end of each prescribed year. The prescribed year ends August 31 for the UFTR. Each annual operating report shall include the following information:

(1) A narrative summary of reactor operating experience including the energy produced by the reactor and the hours the reactor was critical;(2) The unscheduled shutdowns and reactor trips including any corrective actions;(3) Tabulation of major preventive and corrective maintenance operations having safety significance; 38 (4) Tabulation of major changes in the reactor facility and procedures, and a tabulation of new tests or experiments, that are significantly different from those performed previously and are not described in the FSAR, including reports on activities conducted under 10 CFR 50.59 rule.(5) A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the facility operators as determined at or before the point of such release or discharge. (The summary shall include to the extent practicable an estimate of individual radionuclides present in the effluent.

If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed, a statement to this effect is sufficient.);

(6) A summarized result of environmental surveys performed outside the facility; and (7) A summary of exposure received by facility personnel and visitors where such exposures are greater than 25% of that allowed.The annual report shall be submitted with a cover letter to: U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555 6.7.2 Special Reports There shall be a report not later than the following working day by telephone and confirmed in writing to the NRC, 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 limit (see Section 6.6.1);(2) Release of radioactivity from the site above allowed limits (see Section 6.6.2);(3) Any of the following: (see Section 6.6.2)(a) Operation with actual safety-system, settings for required systems less conservative than the LSSS specified in the Technical Specifications;(b) Operation in violation of LCO established in the Technical Specifications unless prompt remedial action is taken;(c) Operation with a reactor safety system component malfunction that renders the reactor safety system incapable of performing its intended safety function.39 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

> I dollar (reactor trips resulting from a known cause are excluded);(e) Abnormal and significant degradation in: (i) reactor fuel, cladding, or both;and/or (ii) coolant boundary or confinement boundary (excluding minor leaks), where applicable, which could result in exceeding prescribed radiation exposure limits of personnel or environment or both;(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.

6.7.3 Other

Special Reports There shall be a written report sent to the Commission within 30 days of the following occurrences:

(1) Permanent changes in the facility organization involving Level 1 or 2 personnel, (2) Significant changes in the transient or accident analyses as described in the Final Safety Analysis Report.6.8 Records Records of the following activities shall be maintained and retained for the periods specified below. The records may be in the form of logs, data sheets, computer storage media, or other suitable forms. The required information may be contained in single, or multiple records, or a combination thereof. Records showing operating parameters of the reactor (i.e., power level, temperature, etc,) for unscheduled shutdowns and significant unplanned transients including trips shall be maintained for a minimum period of 2 years.6.8.1 Records To Be Retained for a Period of at Least Five Years The following records, unless indicated otherwise, are to be retained for a period of at least five (5) years: (1) Normal reactor facility operation (supporting documents such as checklists, log sheets, etc. shall be maintained for a period of at least 3 years);(2) Principal maintenance operations; 40 (3) Reportable occurrences, except for Safety Limit (LS), Limiting Safety System Settings (LSSS), Limiting Conditions of Operation (LCO) violations which have to be maintained until termination of the facility license;(4) Surveillance activities required by the Technical Specifications; (5) Reactor facility radiation and contamination surveys where required by applicable regulations; (6) Experiments performed with the reactor;(7) Fuel inventories, receipts, and shipments; (8) Approved changes in operating procedures; and (9) Records of meetings and audit reports of the RSRS.6.8.2 Records To Be Retained for at Least One Training Cycle Record of retraining and requalification of operations personnel shall be maintained at all times the individual is employed or until the license is renewed.6.8.3 Records To Be Retained for the Lifetime of the Reactor Facility The following records are to be retained for the lifetime of the facility: (1) Gaseous and liquid radioactive effluents released to the environs;(2) Offsite environmental monitoring surveys required by the Technical Specifications; (3) Radiation exposure for all personnel monitored; and Current drawings of the reactor facility.(5) Reports on SL, LSSS, and LCO.Applicable annual reports, if they contain all of the required information, may be used as records in this section.41 Level 1-UF Board of Trustee-Dean, College of Engineering

-Chair, Nuclear& Radiological Engineering Department 4 Director, Environmental Health and Safety-Chair, Radiation Control Committee 1T 1k Facility Director I I Level 3 Reactor Manager k Radiation Control Officer A t Level 4 Operating Staff Class A and B Operator Radiation Safety Specialists Legend: Communication line: Reporting Responsibility:

Figure 6-1 UFTR Organizational Chart 42 3/11/2010 The below spreadsheet provides justification for the changes considered in revised Technical Specifications in support of our application for renewal of Facility Operating License No. R-56.Item Location Location Add/ Justification New TS Old TS Remove/Change/Format I Whole F Formatted overall spacing and indentions to resemble that of the document ANSI document 2 Whole F Changed certain lists that were labeled "(a), (b), (c)" to "(1), (2), (3)." document This is in compliance with the format of the ANSI document.3 Whole C Fixed grammatical errors (i.e. comma usage, spelling mistakes, and document fluency problems.)

NOTE: Most of these changes are documented below, but some may have been overlooked and not added to this Justification chart.4 TOC TOC C Cleaned table of contents to reflect revised tech specs.5 1.0 A Added an introduction which remains consistent with ANSI 15.1.6 1.0 R Remove star at end of "Definitions" and "The word "shall" is used..." footnote moved to definitons section 7 1.0 R Remove the definition "Abnormal Occurrence", redundant with section 6.7.2 8 1.1 A Add a new definition for "Blade-Drop Time" that reflects UFTR method for measurement.

9 1.1 A Add a new definition "Channel", to remain consistent with ANSI 15.1.10 1.1 1.0 C Modify the definition "Channel Calibration", and includes a channel test." to remain consistent with ANSI 15.1.11 1.1 1.0 C Modify the definition "Channel check" to remain consistent with ANSI 15.1.12 1.1 1.0 C Modify the cliefinition "Channel test" to remain consistent with ANSI 15.1.13 1.1 A Add a new definition for "Confinement" to remain consistent with ANSI 15.1 14 1.1 A Add a new definition for "Excess Reactivity" to remain consistent with ANSI 15.1 15 1.1 A Add a new definition for "experiment" to remain consistent with ANSI 15.1 16 1.0 R Remove "Independent Experiment", because "Non-secured Experiment" covers this definition 17 1.1 1.0 C Modified the definition "Inhibit" to reflect multiple methods of implimenting the definition in a system.18 1.1 A Add a new defintion for "License", to remain consistent with ANSI 15.1.19 1.1 A Add a new definition for "Licensee", to remain consistent with ANSI 15.1.20 1.1 1.0 C Modified "Measured value" to remain consistent with ANSI- 15.1.21 1.0 R Removed "Measuring channel", adding the definition of "Channel" removes the need for this definition.

Removed "measuring channel" from entire document and replaced with "channel".

Page 1 of 9 3/11/2010 22 1.1 1.0 C Modified the definition of "Movable Experiment" by adding " to remain consistent with ANSI 15.1.23 1.0 R Removed definition "Non-secured" due to conflicting with other experiment definitions and its difficultly to interpret.

24 1.1 1.0 C Modified definition "Operable" to remain consistent with ANSI 15.1.25 1.1 1.0 C Modified definition "Operating" to remain consistent with ANSI 15.1.26 1.1 A Add definition "Protective action" to remain consistent with ANSI 15.1.27 1.1 A Add definition "Reactivity Worth of an Experiment" to remain consistent with ANSI 15.1.28 1.1 A Add a new definition for "Reactor Operator" to remain consistent with ANSI 15.1 29 1.1 1.0 C Modified definition "Reactor Safety System" to remain consistent with ANSI 15.1.30 1.1 1.0 C Modified the definition of "Reactor Secured" for readability, to reflect the system in place at the facility, and to remain consistent with ANSI 15.1.31 1.1 1.0 C Modified definition "Reactor Shutdown" to allow for operational and experimental flexability while maintaining significant safety margin and to remain consistent with ANSI 15.1.32 1.0 R Removed definition "Reactor Startup", meaning is inherent to all reactors and is.governed by facility procedures.

33 1.1 A Add a new definition for "Reference Core Condition" to remain consistent with ANSI 15.1.34 1.1 1.0 C Modified "Reportable Occurrence", changed 6.5.2 to 6.7.2 35 1.1 1.0 C Modified definition "Research Reactor" to remain consistent with ANSI 15.1.1.1 1.0 C Removed "measuring" from definition "Safety Channel" to remain consistent with previous changes.36 1.1 1.0 C Modified "Secured Experiment" to remain consistent with ANSI 15.1 (used the definition as expressed in the standards).

37 1.0 R Removed the definition "Secured Experiment with Movable Parts", definition is covered under "Movable Experiment" 38 1.1 A Add a new definition for "Senior Reactor Operator" to remain consistent with ANSI 15.1 39 1.1 1.0 C Made the footnote on page I to a defintion for "Should, Shall and May" to remain consistent with ANSI 15.1, switched order based on operational importance.

40 1.1 1.0 C Modified definition "Shutdown Margin," to include "with most reactive blade in the most reactive position.41 1.1 A Add a new definition for "Surveillance Intervals" to remain consistent with ANSI 15.1 Section 4.42 1.1 1.0 C Modified definition "Unscheduled Shutdown" to remain consistent with ANSI 15.1.43 2.1 R Remove "s" from Safety Limits, add abbreviation

"(SL)" 44 2.1 2.1 C Move SL preamble to Bases section, preamble not requirements 45 2.1 2.1 C Reword the statement in the Applicability Section to reflect the new SL from the HEU to LEU conversion 46 2.1 2.1 C Change Specification "not exceed" to "be <". Symbols will used throughout TS document for consistency.

Page 2 of 9 3/11/2010 47 2.1 R Remove Specification (2) since it is not a parameter that reflects the basic physical condition of the fuel cladding, it remains an LCO to ensure primary water purity and the minimization of corrosion buildup 48 2.1 R Remove in Bases, "the core flow conditions to assure no onset of nucleate boiling within the core and", since these parameters are not part of the SL 49 2.2 2.2 C Move LSSS preamble to Bases section, not requirements 50 2.2 2.2 C Change Objective to reflect the only one SL 51 2.2 Spec 2.2 Spec C Remove "at any flow rate", trip is not conditional on flow. Replace (1) (1) "not exceed" with "be <" 52 2.2 2.2 R Remove Spec (a) since not used for UFTR fuel. Combined (b) with Spec(2) Spec(2) (2), removed "fuel coolant channel spacing tolerance is < 20 mils" and added inequalities.

Also removed "at power levels greater than I watt" because there is no bistable for that I watt requirement, flow is greater than 41 gpm or scram is initiated.

53 2.2 Spec 2.2 Spec C Remove "average" from (3). Remove Spec (a) since not used for (3) (3) UFTR fuel. Spec (b) and (c) changed to (a) and (b), removed mil spec from new (a) and added inequalities to (a) and (b).54 2.2 Spec 2.2 Spec C Changed the wording with inequality (4) (4)55 2.2 R Remove and place in section 3.2.1(6), since parameter does not Spec(5) directly impact SL. Several power monitors exist, even if all fail this way, operator has direct means to ensure SL not violated via temperature and flow.56 2.2 R Remove and place in section 3.3(1), parameter does not directly Spec(6) impact SL. Primary pump running is indirect parameter whereas, primary flow is direct.57 2.2 R Remove and place in section 3.3(2), parameter does not directly Spec(7) impact SL. Parameter shows possible outlet pipe leakage and/or flow blockage in outlet pipe. Still protected from exceeding SL with temperature, flow, and reactor power.58 2.2 R Remove and place in section 3.3(3), parameter does not directly Spec(8) impact SL. Parameter shows moderation in core, if moderation goes away, reactor will shut down due to negative void coefficient of reactivity inherent in UFTR design.59 2.2 R Remove and place in section 3.3(4), parameter does not directly Spec(9) impact SL. Parameter shows heat removal capability.

Reactor protected by primary temperature indication as LSSS.60 2.2 R Remove and place in section 3.2.1(5), parameter does not directly Spec(1 0) impact SL. Without power, control system is not energized and reactor will be in shutdown condition no matter what condition is was in just prior to power loss.61 2.2 R Remove and place in section 3.5 (1) and (3), parameter does not Spec(1 1) directly impact SL. Parameter more for confinement.

62 2.2 R Remove and place in Section 3.10, parameter does not directly Spec(12) impact SL. Parameter for dose levels to the operators and public.63 2.2 Bases 2.2 Bases C Adjust Bases to explain Spec 1-4. All others will be in Section 3 with respective new LCO.64 3.0 A Add "(LCO)" to section title Page 3 of 9 3/11/2010 65 3.0 3.0 C Move LCO preamble to Bases section, preamble not requirements 66 3.0 3.0 A Add Applicability, Objective, Specifications, and Bases to each of the sections to remain consistent with ANSI 15.1 67 3.1 3.1 C Change title to Reactor Core Parameters to remain consistent with ANSI 15.1 and easily differentiate between core and experimental reactivity limits 68 3.1 Spec 3.1 Spec C Add core reference conditions and installed experiment reactivity (1) (1) requirements to shutdown margin, added inequality symbol.Replaced 2% with one dollar to remain consistent with definitions in previous sections and ANSI 15.1.69 3.1 Spec 3.1 Spec C Change "cold critical, without xenon poisoning" to "reference core (2) (2) conditions", and introduce inequality 70 3.1 Spec R Remove Spec (3), these reactivity coefficients are inherent to the (3) UFTR design. Do we know the value for temp?71 3.1 Spec 3.1 Spec C Shifted Spec (4) to (3).(3) (4)72 3.1 Spec R not a spec, experimental specs can be found later in section 3 (5)73 3.1 Bases 3.1(6) C Bases changed to accomodate changes in Specs Bases 74 3.2.1(1) R remove, because it not a spec; it discusses the design of the reactivity control mechanism 75 3.2.1(1) 3.2.1(2) C Removed first sentence since redundant with control blade withdrawal interlock.

76 3.2.1(3) 3.2.1(4) C Blade drop time was change from 1.0 sec to 1.5 sec based on analysis.

Still ensures prompt shutdown of reactor, but allows for operational flexibility 77 3.2.1(3) 3.2.1(4) C change "rod-drop" to "blade-drop" to be consistent with control system in place, remove last part of sentence beginning with "...from initiation..." since defined in "Blade-drop time" definition.

78 3.2.1(6) R removed; it is redundant, and discussed in Section 4.2.2 (10)79 3.2.2 (1) A added; it was LSSS,from Section 2.2(10); surveillance found in old table 3.2, new table 4-1 80 3.2.2(2) A added; it was LSSS from Section 2.2(5)81 3.2.2(2) R removed; it is redundant, and discussed in Section 4.1(2)82 3.2.3 R removed minimum number of operating channel table and renamed it to Table 3-2 and put it at the end of Section 3; Table 3-1 also moved to the end of Section 3, and the old Table 3-2 renamed to Table 4-1.and placed at the end of Section 4, for readability.

83 3.2.4 R Removed entire section, the informational material does not belong in bases or tech specs.84 3.3 3.8 C Changed section to remain consistent with ANSI 15.1 and to accommodate LSSS's moved to LCO's 85 3.3(1) A From Section 2.2(6), reason is expressed above, added dump valve as well since the two work in conjunction with eachother.

86 3.3(2) A From Section 2.2(7), reason is expressed above 87 3.3(3) A From Section 2.2(8), reason is expressed above 88 3.3(4) A From Section 2.2(9), reason is expressed above 89 3.3(4) C Removed mention of City Water Cooling, city water will be valved out.Page 4 of 9 3/11/2010 90 3.8(1)R Spec is covered in an LSSS.91 3.3(5) 3.8(2) C Added inequality and removed time requirement, check SAR 3.8(3) R Removed since primary sampling will occur when other routine checks indicate a problem, and spec c is already a well defined procedure for liquid effulent waste removal.93 3.3(6) 3.8(4) C Reworded to clarify, no change in meaning.94 3.3 Bases 3.8 Bases C Modified Bases to reflect added specifications.

Bases moved with specifications from section 2.2.95 3.4 3.4 C Table 3.3 moved to end of section 3 and all sections now contain proper parts, no data changed, section 3.4.2(2) dilution factor 96 3.4.1 3.4.1 C Adjusted to remain consistent with ANSI 15.1 97 3.4.1 3.4.4 C Added section 3.4.4 here since referenced same table.98 3.4.1 3.4.7 A Added applicable bases from old 3.4.7.Bases 99 3.4.2(2) 3.4.2(2) C Remove volumetric flow rate since it is not the limiting condition for operation, the release concentration in (3) is used as the LCO.Remove the atmospheric dilution factor since not a LCO and a researched calculated value will be used in the procedure for AR-41 concentration calcs.100 3.4.3(2) R Removed, not an LCO, part of emergency procedure.

101 3.4.6 R Section removed because not an LCO, compliance is accomplished through UF EH&S processes.

102 3.4.4(1) 3.4.5 C "waste water" added to specify the correct tanks and (2) , 103 3.5 3.3 C Section 3.5 is the old section 3.3 with proper headings 104 3.5(2) 3.3.1(2) C Adjusted spec to include system operable since there is surveillance requirement.

Removed volumetric flow rate requirement since true LCO is Ar-41 concentration limit and dilution can change.105 3.5 (3) A Added damper spec since there is surveillance in Section 4.5.106 3.5(6) 3.3.1(5) C Reworded to clarify, no change in meaning.107 3.6 3.5(1) and R Moved spec 1 and 2 to Section 6.5 to remain consistent with ANSI (2) 15.1 108 3.6.1 3.5(3) C Created section 3.6.1 to be consistent with ANSI 15.1 Section 3.8.109 3.6.1(1) 3.5(3)(a)

C Removed "movable or non-secured" to include all experiment types.Added inequality.

110 3.6.1(2) 3.5(3)(b)

C Added inequality.

111 3.6.2 3.5(4-8) C Created section 3.6.2 to be consistent with ANSI 15.1 Section 3.8.112 3.6.2(3-5) 3.5(3-5) C Changed numbering.

113 3.6.2(2) 3.5(5) C Reworded last part of sentence to say "... do not cause the safety limit to be exceeded." for ease of reading.114 3.7 3.6 R Reworded for clarification, no change in meaning.115 3.7(1) 3.6(1) C Removed material in parenteses since section 3.4.1(3) specifies more conservative setpoints could be set.116 3.7(1) R Removed because it is a design feature and can be found in Section 5.3 first sentence.117 3.7(2) R Removed because it is redundant based on Section 6.4.118 3.8(1) 3.7(3) C Removed last sentencent, elemental analysis of water performed when other factors indicate a possible fuel cladding breach.Page 5 of 9 119 3.8(2)3.7(4)C 3/11/2010 Added to the end of the sentence "...other than to locate the failed fuel" to allow the operators to locate the failed fuel assembly.120 3.7(5) R Removed because it is redundant based on Section 6.4.121 3.8(3) A Added to ensure operator safety while removing reactor shielding.

122 3.7(6)(7)

R Removed because they are redundant with statement in Section 5.4 123 3.8(4) A Added core configuration, to remain consistent with ANSI 15.1 for section 3.1 in standard,need surveillance 124 3.8(5) A Added Fuel Inspections, to remain consistent with ANSI 15.1 for section 3.1 in standard, need surveillance 125 3.8 Bases 3.7 Bases C Kept second sentence only, all others pertained toremoved specifications.

Adjusted for the added specs (4 and 5).126 3.9.2(1) R Removed and placed as surveillance in Section 4.9.127 3.9.2(2) R Removed due to part (a) being part of the universities radition control (a) and (b) program. Part(b) provides no useful information, an elevated reading outside the normal is investigated and answered regardless of a percentage above the mean normal value.128 3.9 3.9.2(3) C Reworded to only include the LCO portion, the surveillance portions (a) and (b) was moved to Section 4.9 129 3.10 A From Section 2.2(12), reason is expressed above, bases followed.130 Table 3-1 C Cleaned up and added inequalities.

131 Table3-3 C Changed Alarm setting for area radiation monitoring to 5 mR/hr warning level and 20 mR/hr alarm level to ensure conformance with 0 OCFR70.24 a(2).132 Table 4-1 4.0 C Remove title and incorporate 4.1(1) into 4.0 paragraph.

Moved to bottom of Table 4-1 to ensure reason for missing surveillances in proper place where it will have value.133 4.1 4.2 C Change Title to remain consistent with ANSI 15.1 and numbering scheme changed as well to fit ANSI 15.1.134 4.1 4.2.1(1) C Remove information in parenthese since intervals are defined in specificati definitions.

Will remove all intervals.

on 135 4.2.1(2) R Removed since no longer in LCO section and are inherent parameters and (3) to the core design.136 4.2 4.2.2 C Title changed to remain consistent with ANSI 15. 1.137 4.2.1 A Title added to remain consistent with LCO section.138 4.2.1(3) 4.2.2(3) C Reworded to remove operability and use operable which is defined.139 4.2.1 (5) 4.2.2(5) C Removed "operability test" since it is not defined, calibration of system is all inclusive for testing.140 4.2.1 4.2.2 R Removed Specifications (8) and (9) and relocated them to next section 4.2.2 Reactor Safety System for consistency.

4.2.2(1) 4.1(2) C Moved spec and reworded for readability.

141 4.2.2 4.2.2 C Incorporated Spec (8) and (9) into new section and added old spec 4.1 (2).142 4.3 4.2.8 C Shifted section for consistency with ANSI 15.1.Page 6 of 9 4.2.8(3)R 3/11/2010 Removed since check provides no useful information.

We have check for fuel failure and calcs to prove below release limits if have primary to secondary leak with no fuel failure.143 4.3 (1-3) A Added Specs to ensure surveillances were had for new specs in Section 3.3.144 4.3 (4), 4.2.8 (1) C Adjusted specs to include only surveillance and not extraneous and (6) information or LCO information.

92 4.2.8(2) (a R Removed since staff has several other means to determine fuel b) condition, primary will be sampled and analyzed for fission products when other indicaters are present.145 4.3 Bases A Added to remain consistent with ANSI 15.1 and to include new, surveillances.

146 4.4 4.2.4 C Change title, remove Surveillances.

147 4.4 4.2.4(4-5)

R Removed since redundant.

148 4.4 Note Section 4.4 covers surveillances for all of Section 3.4.149 4.2.3 (1) R Removed (a) and (b) because they located in Section 3.5.150 4.5 4.2.3 (2) C Moved spec 2(a), and removed spec (2)(b) since redundant with new 4.7(2).151 4.5(3) A Added spec due to LCO in section 3.5(6).152 4.6 4.2.5 C Change numbering 153 4.7 4.2.6 C Change numbering and added objective.

154 4.2.6(3) R Removed from tech specs, if necessary, should be part of E-plan.155 4.8(1) 4.2.7(1) C Remove portion about block shielding since determined to be an LCO, located in Section 3.8(3).156 4.2.7(2) R Removed because redundant, it is stated in Section 6 that UFTR will follow procedures, those checks are part of fuel handling procedures.

157. 4.9 3.9.2 C Surveillance information relocated to here.158 Table 4-1 Table 3-2 C Adjust table for consistency with definitions.

159 5.0 S Remove preamble from section 5.0, not requirement 160 5.1 5.1 C Change title to "Site and Facility Description", to remain consistent with ANSI 15.1 161 5.1 A Add to end of 5.1 the first paragraph from 5.2 and remove (o) from 10 CFR 73.2(o) in second sentence of moved paragraph.

162 5.1 A Added sentence at end of 5.1 to refer to UFSAR for more details 163 5.2 R Remove all of 5.2 to remain consistent with ANSI 15.1 164 5.2 5.6 A Rename 5.2 Reactor Coolant System to remain consistent with ANSI 15.1, moved 5.6 to 5.2 for same reason 165 5.2.1 5.6.1 C "Temperature sensing device" replaces thermocouples for operational flexability.

166 5.2.1 5.6.1 R Remove last sentence, "The reactor power ...",this belongs in section 4 and is at 4.2.2 spec (8)1*67 5.2.2 5.6.2 R Remove first 2 sentences, UFTR will not be using city water for secondary cooling.168 5.2.2 5.6.2 R Remove sentences describing city water system, will not be used at IUFTR.169 5.2.2 5.6.2 look into sending to, sanitary sewer 170 5.3 5.3 C Change title to Reactor Core and Fuel, added just enough useful information to tickle your fancy.Page 7 of 9 171 5.3 5.3 and 5.4 C 3/11/2010 Merge sections 5.3 and 5.4 into new 5.3 with core description first, to remain consistent with ANSI 15.1 172 5.3 C Adjusted first sentence to clarify fuel enrichment 173 5.3 5.3 Marked in yellow.., remove because in license?174 5.5, 5.6, R Removed sections, found in UFSAR, to remain consistent with ANSI 5.7 15.1 175 5.4 5.8 C old 5.8 now 5.4 to remain consistent with ANSI 15.1, change keff to 0.9 176 5.4 5.8.1 and C All content removed and replaced with statement the fuel stored lAW 5.8.2 UFTR Physical Security Plan.177 5.4 4.2.7(1) A Added, it came from section 4.2.7(1) because it was determined to be part of facility design.178 6.1 R Remove definitions, RO and SRO are in definitions, to remain consistent with ANSI 15.1 179 6.1.1 6.2.1 C Replaced "Vice-President.." with level 1 admin.180 6.1.2 6.2.2 A Add sentence at end to ensure all scenerios are covered.181 Figure 6-1 C Moved Figure 6-1 to end of section 6, ease of use.182 6.1.3 6.2.3 C Removed "certified" from spec 1 of minimum staffing, Reactor operator is defined in section 1 as being licensed.183 6.1.3 6.2.3 C Removed "Class A" and replaced with Senior in spec 3 of minimum staffing, SRO defined in section 1 184 6.1.3 6.2.3 C Cleanup Spec 3 of minimum staffing for ease of use.185 6.1.3 6.2.3 C Add Spec I to "Events requiring the presence of SRO" to remain consistent with ANSI 15.1 186 6.1.3 6.2.3 C change units and added inequality for new spec 3 of"Events requiring the direction..." 187 6.1.3 6.2.3 C Old spec 3 of "Events requiring

..." new spec 4, removed material in parentheses and added "or significant power reduction", to remain consistent with ANSI 15.1.188 6.1.4 6.2.4 C Update date and section for ANSI standard 189 6.2 6.2.5 C Numbering change to remain consistent with ANSI 15.1, includes all of old 6.2.5 to new 6.2 and subsections.

190 6.2.1 6.2.5(1) C Added wording for clarification.

191 6.2.2 6.2.5(2) C Sentence reworded to clarify and removed redundant information membershi membersh found in new 6.2.1.p (d) ip (d)192 6.2.2 6.2.5(2) C Remove surveillance interval, intervals defined locally in procedure meetings meetings (a) spec a 193 6.2.2 6.2.5(2) C Added sentence fragment at end to remain consistent with ANSI 15.1 meetings meetings spec c spec c 194 6.2.3(a) 6.2.5(3) C Remove "unreviewed safety question" and reworded since does not spec (a) exist in IOCFR50.59 determination 195 6.2.3 C Last paragraph restructured for clarity.196 6.2.4(a-d) 6.2.5(4)(a-C Remove surveillance interval, intervals defined locally in procedure, d) woring changed for clarification.

197 6.3 7.0 C Moved to remain consistent with ANSI 15.1 and changed title.198 6.3 7.0 C All content removed and replaced with statement that is directly from ANSI 15.1. All content in 7.0 were statements, not specifications, therefore do not belong.199 6.4 6.3 C Section shifted to 6.4 to remain consistent with ANSI 15.1 Page 8 of 9 200 6.4 6.3 C 3/11/2010 In first para. Remove "Nuclear Facilities, replace with "the UFTR", directors title now 201 6.4 (5) 6.3 (5) C Replaced with "surveillances required by the technical specifications", statement is all encompassing 202 6.4 6.3 C Last paragraph, last sentence, added "by the next working day." to remain consistent with ANSI 15.1 203 6.5(1) 6.4(1) C change to "Section 3.6", for consistency 204 6.4(2) R Removed, redundant with new Section 3.6 spec(1), shifted other specs up in number 205 6.6, 6.7, 6.5, 6.6, C Shifted numbering, for consistency, wording changed for clarity.6.8 6.7 206 6.7.1 6.6.1 C Clean up wording.207 6.7.1(4) 6.6.1(4) C removed "unreviewed safety", reworded since phrase no longer exists 208 6.7.2 6.6.2 C switched specs (1) and (2) and added "(see section .." to new (2) and (3) to remain consistent with ANSI 15.1 209 6.7.2(3) a 6.6.2(3)a C Abbreviated and b and b 210 6.7.2.(3)c 6.6.2(3)c C Added "Operation with " to allow for problem indentification while shutdown and not having it be reportable as long as the reactor was not operated in that condition.

211 6.7.2.(3)c 6.6.2(3)c C Moved note out of parentheses for easy reading.212 6.7.2.(3)d 6.6.2(3)d C units and inequality to remain consistent throughout specs 213 6.6.2(3)g R remove, redundant with all in section'6.6.2 and section 4 214 6.7.3(1) 6.6.3(1) C remove level 3, to remain consistent with ANSI 15.1 215 6.7.3(2) 6.6.3(2) C change end of sentence, assuming SAR and conversion report will be merged 216 6.8 6.7 C Beginning of last sentence "Recorder charts" replaced with"Records", to clarify 217 6.8.1 6.7.1 C (1) change time to 3 years to ensure NRC inspection cycle included for oversight, (3) additions made for clarity 218 6.8.2 6.7.2 C Change to remain consistent with ANSI 15.1 219 6.8.3(4) 6.7.3(4) C replace "updated" with "current", clarify the statement 220 6.8.3(5) A Added (5) to remain consistent with new 6.8.1(3).221 Figure 6.1 Figure 6.1 C clean and clarify, no change in responsibilities Page 9 of 9