ML16278A111

Us Dept. of Defense, Armed Forces Radiobiology Research Institute (Afrri), Response to Request for Additional Information Regarding Application for License Renewal

ML16278A111
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	09/30/2016
From:	Miller S US Dept of Defense, Armed Forces Radiobiology Research Institute
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME1587
Download: ML16278A111 (85)

Text

~o-/10 ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE 8901 WISCONSIN AVl!:NUI!!:

BE:THl!:SDA, MARYLAND 20889-6603 September 30, 2016 Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE APPLICATION FOR LICENSE RENEWAL (TAC NO. MEl 587)

Sir:

In response to conversations on 30 September 2016 with NRC staff, additional information is enclosed.

1. DRAFT Technical Specifications dated 30 Sept 2016 with miscellaneous edits

2. Additional information for RAI 16 June 2016.

3. Minor correction to Chapter 7, Instrumentation and control If you need further information, please contact Mr. Steve Miller at 301-295-9245 or stephen.miller@usuhs.edu.

I declare under penalty of perjury that the foregoing

• true and correct to the best of my knowledge. Executed on September 30, 201 Stephen Mill r Reactor Faci ity Director

DRAFT TECHNICAL SPECIFICATIONS FOR THE AFRRI REACTOR FACILITY 30 September 2016 LICENSE R-84 DOCKET 50-170

Preface Included in this document are the Technical Specifications and the Bases for the Technical Specifications. These bases, which provide the technical support for the individual Technical Specifications, are included for information purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

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TECHNICAL SPECIFICATIONS FOR THE AFRRI REACTOR FACILITY LICENSE NO. R-84 DOCKET 50-170 TABLE OF CONTENTS 1.0. DEFINITIONS 1.1. ALARA ....................................................................................................................... l 1.2. Channel ....................................................................................................................... 1 1.3. Channel Calibration .................................................................................................... 1 1.4. Channel Check ............................................................................................................ 1 1.5. Channel Test ............................................................................................................... 1 1.6. Confinement ................................................................................................................ 1 1.7. Control Rod ................................................................................................................. 1 1.8. Core Configuration ..................................................................................................... 2 1.9. Core Grid Position ...................................................................................................... 2 1.10. Emergency Stop .......................................................................................................... 2 1.11. Excess Reactivity ........................................................................................................ 2 1.12. Experiment ...........................................................................*....................................... 2 1.13. Experimental Facilities ............................................................................................... 2 1.14. Fuel Element ............................................................................................................... 3 1.15. High Flux Safety Channel.. ......................................................................................... 3 1.16. Initial Startup and Approach to Power.. ...................................................................... 3 1.17. Instrumented Fuel Element ......................................................................................... 3 1.18. Long-Term .................................................................................................................. 3 1.19. Measured Value .......................................................................................................... 3 1.20. Movable Experiment. .................................................................................................. 3 1.21. On Call ........................................................................................................................ 3 1.22. Operable ..................................................................................................................... , 4 1.23. Operational Channel ................................................................................................... 4 1.24. Operating ..................................................................................................................... 4 1.25. Power Level Monitoring Channel.. ............................................................................. 4 1.26. Protective Action ........................................................................................................ 4 1.27. Pulse Mode .................................................................................................................. 4 1.28. Reactivity Worth of an Experiment ............................................................................ 4 1.29. Reactor Operating ....................................................................................................... 4 1.30. Reactor Operator ......................................................................................................... 5 1.31. Reactor Safety Systems ............................................................................................... 5 1.32. Reactor Secured .......................................................................................................... 5 1.33. Reactor Shutdown ........................................................................................................ 5 1.34. Reference Core Condition ........................................................................................... 5 1.35. Safety Channel ............................................................................................................ 6 1.36. Scram Time ................................................................................................................. 6 1.37. Secured Experiment .................................................................................................... 6 11

1.38. Senior Reactor Operator ............................................................................................. 6 1.39. Shall, Should, and May ............................................................................................... 6 1.40. Shutdown Margin ........................................................................................................ 6 1.41. Standard Control Rod ................................................................................................. 6 1.42. Steady State Mode ...................................................................................................... 6 1.43. Surveillance Intervals ................................................................................................. 7 1.44. Transient Rod .............................................................................................................. 7 1.45. True Value .................................................................................................................. 7 1.46. Unscheduled Shutdown .............................................................................................. 7 2.0. SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING 2.1. Safety Limit: Fuel Element Temperature .................................................................. 8 2.2. Limiting Safety System Setting for Fuel Temperature ............................................... 8 3.0. LIMITING CONDITIONS FOR OPERATIONS 3.1. Reactor Core Parameters ........................................................................................... 10 3.1.1. Steady State Operation .................................................................................... 10

• 3.1.2. Pulse Mode Operation ..................................................................................... 10 3.1.3. Reactivity Limitations ..................................................................................... 11 3.2. Reactor Control and Safety Systems ......................................................................... 12 3.2.1. Reactor Control System .................................................................................. 12 3.2.2. Reactor Safety Systems ................................................................................... 13 3.2.3. Facility Interlock System ................................................................................ 15 3.3. Coolant Systems ........................................................................................................ 16 3.4. Ventilation System .................................................................................................... 17 3.5. Radiation-Monitoring System and Effluents ............................................................ 19 3.5.1. Monitoring System .......................................................................................... 19

3.5.2. Effluents

Argon-41 Discharge Limit ............................................................ 20 3.6. Limitations on Experiments ...................................................................................... 21 3.7. Fuel Parameters ......... ;............................................................................................... 24 4.0. SURVEILLANCE REQUIREMENTS 4.1. Reactor Core Parameters ........................................................................................... 25 4.2. Reactor Control and Safety Systems ......................................................................... 26 4.2.1. Reactor Control Systems ................................................................................. 26 4.2.2. Reactor Safety Systems ................................................................................... 27 4.2.3. Fuel Temperature ............................................................................................ 28 4.2.4 Facility Interlock System ................................................................................ 29 4.3. Coolant Systems ........................................................................................................ 30 4.4. Ventilation System .................................................................................................... 31 4.5. Radiation-Monitoring System and Effluents ............................................................ 31 4.5.1. Monitoring System .......................................................................................... 31 4.5.2. Effluents ......................................................................................................... 32 4.6. Reactor Fuel Elements .............................................................................................. 33 iii

5.0. DESIGN FEATURES 5 .1. Site and Facility Description ..................................................................................... 34 5.2. Reactor Core and Fuel .............................................................................................. 35 5.2.l. Reactor Fuel .................................................................................................... 35 5.2.2. Reactor Core ................................................................................................... 36 5.2.3. Control Rods ............................................................................................ :...... 37 5.3. Fuel Storage .............................................................................................................. 38 6.0. ADMINISTRATIVE CONTROLS 6.1. Organization .............................................................................................................. 39 6.1.1. Structure .......................................................................................................... 39 6.1.2. Responsibility ................................................................................................. 40 6.1.3. Staffing ............................................................................................................ 40 6.1.3.1. Selection of Personnel.. ..................................................................... 40 6.1.3.2. Operations ......................................................................................... 41 6.1.3.3. Training of Personnel. ....................................................................... 42 6.2. Review and Audit - The Reactor and Radiation Facilities Safety Subcommittee (RRFSS) ..................................... :****************************************************** 43 6.2.1. Composition and Qualifications ..................................................................... 43 6.2.1.1. Composition ...................................................................................... 43 6.2.1.2. Qualifications ........................................................................ ;........... 43 6.2.2. Function and Authority ................................................................................... 44 6.2.2.1. Function ............................................................................................ 44 6.2.2.2. Authority ........................................................................................... 44 6.2.3. Rules ............................................................................................................... 44 6.2.3.1. Alternates .......................................................................................... 44 6.2.3.2. Meeting Frequency ........................................................................... 44 6.2.3.3. Quorum ...................... :...................................................................... 44 6.2.3.4. Voting Rules ..................................................................................... 44 6.2.3.5. Minutes ............................................................................................. 45 6.2.4. Review Function ............................................................................................. 45 6.2.5. Audit Function ................................................................................................ 45 6.3. Procedures ................................................................................................................. 46 6.4. Review and Approval of Experiments ...................................................................... 47 6.5. Required Actions ...................................................................................................... 48 6.5.1. Actions to be Taken in Case of Safety Limit Violation .................................. 48 6.5.2. Reportable Occurrences .................................................................................. 48 6.5.3. Actions to be Taken in Case of Reportable Occurrences ............................... 49 6.6. Operating Reports ..................................................................................................... 50 6.7. Records ..................................................................................................................... 53 iv

6.7.l. Records to be Retained for a Period of At Least Five Years .................... 53 6.7.2. Records to be Retained for At Least One Certification Cycle .................. 53

6. 7 .3. Records to be Retained for the Life of the Facility ............................... 53 v

1.0. DEFINITIONS 1.1. ALARA The ALARA program (As Low As Reasonably Achievable) is a program for maintaining occupational exposures to radiation and release of radioactive effluents to the environment as low as reasonably achievable.

1.2. 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.

1.3. 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 that the channel measures.

Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and shall be deemed to include a channel test.

1.4. 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.

1.5. CHANNEL TEST A channel test is the introduction of a signal into the channel for verification that it is operable.

1.6 CONFINEMENT Confinement is an enclosure of the overall facility that is designed to limit the release of effluents between the enclosure and its external environment through controlled or defined pathways.

1.7 CONTROL ROD A control rod is a device fabricated from neutron absorbing material or fuel, or both, that is used to establish neutron flux changes and to compensate for routine reactivity losses.

Scrammable control rods can be quickly uncoupled from their drive units to rapidly shutdown the reactor if needed.

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1.8. CORE CONFIGURATION The core configuration includes the number, type, or arrangement of fuel elements and standard control rods/transient rod occupying the core grid.

1.9. CORE GRID POSITION The core grid position refers to the location of a fuel element, control rod, or experiment in the grid plate. It is specified by a letter indicating the specific ring in the grid plate and a number indicating a particular position within that ring.

1.10. EMERGENCY STOP Emergency Stop is a scram designed to prevent or cease reactor operations.

Emergency stop buttons are provided in Exposure Room 1, Exposure Room 2 and on the console 1.11. EXCESS REACTIVITY Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical (keff = 1) at reference core conditions or at a specific set of conditions.

1.12. EXPERIMENT Any operation, hardware, or target (excluding devices such as detectors, foils, etc.)

that is designed to investigate nonroutine reactor characteristics or that is intended for irradiation within an experimental facility. Hardware rigidly secured to the core or shield structure so as to be a part of its design to carry out experiments is not normally considered an experiment.

1.13. EXPERIMENTAL FACILITIES The experimental facilities associated with the AFRRI TRIGA reactor shall be:

a. Exposure Room #1

b. Exposure Room #2

c. Reactor Pool

d. Core Experiment Tube (CET)

e. Portable Beam Tubes

f. Pneumatic Transfer System 2

g. In-core Locations 1.14. FUEL ELEMENT A fuel element is a single TRIGA fuel rod or the fuel portion of a fuel follower control rod (FFCR).

1.15 HIGH FLUX SAFETY CHANNEL A high flux safety channel is a power measuring safety channel in the reactor safety system, NP and NPP.

1.16. INITIAL STARTUP AND APPROACH TO POWER Intentionally left blank 1.17. INSTRUMENTED FUEL ELEMENT An instrumented fuel element is a fuel element in which one or more thermocouples have been embedded for the purpose of measuring fuel temperatures.

1.18. LONG-TERM STORAGE Long-term storage of fuel applies to fuel that has been taken out of service with no plans for use for more than one fuel measurement cycle.

1.19. MEASURED VALUE The measured value is the value of a parameter as it appears on the output of a channel.

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

1.21. ON CALL A person is considered on call if:

a. The individual has been specifically designated and the operator knows of the designation;

b. The individual keeps the operator posted as to their whereabouts and telephone number; 3

c. The individual remains at a reachable location and is capable of getting to the reactor facility within 60 minutes under normal circumstances; and

d. The individual remains in a state of readiness to perform their duties.

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

1.23. OPERATIONAL CHANNEL Operational Channel: The Operational Channel is a power measuring channel used during steady state and square wave operations 1.24. OPERATING Operating means a component or system is performing its intended function.

1.25. POWER LEVEL MONITORING CHANNEL A power level monitoring channel is defined to be a channel that is intended to provide real time power level readings to the operator.

1.26. 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 set point.

1.27. PULSE MODE Operation in the pulse mode shall mean that the reactor is intentionally placed on a prompt critical excursion by making a step insertion of reactivity above critical with the transient rod. The reactor may be pulsed from a critical or subcritical state.

1.28. 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.

1.29. REACTOR OPERATING The reactor is operating whenever it is not secured or shutdown.

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1.30. REACTOR OPERATOR A reactor operator is an individual who is licensed to manipulate the controls of a reactor.

1.31. REACTOR SAFETY SYSTEMS Reactor safety systems are those systems, including their associated input channels that are designed to initiate a reactor scram for the primary purpose of protecting the reactor or to provide information for initiation of manual protective action.

1.32. REACTOR SECURED The reactor is secured when:*

a. Either there is insufficient moderator available in the reactor to attain criticality or there is insufficient fissile material in the reactor to attain criticality under optimum available conditions of moderation and reflection; or,

b. All of the following conditions exist:

1. All control rods are fully inserted into the core;

2. The console key switch is in the off position and the key is removed;

3. No work is in progress involving fuel movement, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods; and

4. No experiments are being moved or serviced that have, on movement, a reactivity worth exceeding $1.00.

1.33. REACTOR SHUTDOWN The reactor is shutdown when it is subcritical by at least $1.00 of reactivity in the reference core condition with the reactivity worth of all installed experiments included.

1.34. REFERENCE CORE CONDITION The reference core condition is when the core is at ambient temperature and the reactivity worth of xenon is negligible (<$0.01).

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1.35. SAFETY CHANNEL A safety channel is ahigh flux safety channel with scram capability.

1.36. SCRAM TIME Scram time is the elapsed time between the initiation of a scram signal and the full insertion of the control rod.

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

1.38. SENIOR REACTOR OPERATOR A senior reactor operator is an individual who is licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.

1.39. 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, neither a requirement nor a recommendation.

1.40. 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 rod in the most reactive position, and that the reactor will remain subcritical without further operator action.

1.41. STANDARD CONTROL ROD A standard control rod is a control rod having electromechanical drive and scram capabilities. It is withdrawn by an electromagnet/armature system.

1.42. STEADY STATE MODE Operation in the steady state mode shall mean operation of the reactor either by manual operation of the control rods or by automatic operation of one or more 6

control rods at power levels not exceeding 1.1 MW. Square wave mode is a subset of the steady state mode of operation.

1.43. SURVEILLANCE INTERVALS Allowable surveillance intervals shall not exceed the following:

a. Biennial - interval not to exceed 30 months

b. Annual- interval not to exceed 15 months

c. Semi-annual- interval not to exceed 7.5 months

d. Quarterly-interval not to exceed 4 months

e. Monthly - interval not to exceed 6 weeks

f. Weekly-interval not to exceed 10 days 1.44. TRANSIENT ROD The transient rod is a control rod with scram capabilities that can be rapidly ejected*

from the reactor core to produce a pulse. It is activated by applying compressed air to a piston.

1.45. TRUE VALUE The true value is the actual value of a parameter.

1.46. UNSCHEDULED SHUTDOWN An unscheduled shutdown is defined as any unplanned shutdown of the reactor caused by actuation of the reactor safety systems, 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.

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2.0. SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING 2.1. SAFETY LIMIT: FUEL ELEMENT TEMPERATURE Applicability This specification applies to the temperature of the reactor fuel.

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

Specification The maximum temperature in a TRIGA fuel element shall not exceed l ,000°C under any mode of operation.

The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well suited as a single specification because it can be measured. A foss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel-moderator and cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the fuel-moderator. The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium in the alloy.

The safety limit for the TRIGA fuel is based on data which indicates that the stress in the cladding will remain below the ultimate stress, provided that the temperature of the fuel does not exceed 1,000°C and the fuel cladding is water cooled.

2.2. LIMITING SAFETY SYSTEM SETTING FOR FUEL TEMPERATURE Applicability This specification applies to the scram settings which prevent the safety limit from being reached.

Objective The objective is to prevent the safety limit from being reached.

Specification 8

The limiting safety system setting shall be equal to or less than 600°C, as measured in the instrumented fuel elements. There shall be two fuel temperature safety channels. One channel shall utilize an instrumented fuel element in the B ring, and the second channel shall utilize an instrumented fuel element in the Cring.

Basis The limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated, preventing the safety limit from being exceeded. A setting of 600°C provides a safety margin of 400°C for TRIGA fuel elements. Part of the safety margin is used to account for the difference between the true and the measured temperatures resulting from the actual location of the thermocouple. If the instrumented fuel element is located in the hottest position in the core, the difference between the true and measured temperatures will be only a few degrees. There are two fuel temperature monitoring channels within the reactor core (one in the B ring and one in the C ring). The highest power density occurs in these two rings, and therefore provides temperature monitoring in the hottest locations of the reactor core. Table 4-14 of the AFRRI Safety Analysis Report identifies the rod power factors for each fuel location in the reactor core. Within the B ring, the highest and lowest power factors are 1.552 and 1.525, respectively. Assuming the instrumented fuel element is located in the lowest power density position (B-1 ), a temperature indication of 600°C would yield a peak temperature at the highest power density location (B-4) of 611°C. Within the C ring, the highest and lowest power factors are 1.438 and 1.374, respectively. Assuming the instrumented fuel element is located in the lowest power density position (C-12), a temperature indication of 600°C would yield a peak temperature at the highest power density loc~tion (C-9) of 628°C.

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3.0. LIMITING CONDITIONS FOR OPERATIONS 3.1. REACTOR CORE PARAMETERS 3.1.1. STEADY STATE OPERATION Applicability This specification applies to the maximum reactor power attained during steady state operation.

Objective The objective is to ensure that the fuel temperature safety limit shall not be exceeded during steady state operation.

Specification The reactor steady state power level shall not exceed 1.1 MW.

The thermal-hydraulic analysis of steady state operation using the RELAP5 computer code, as detailed in the AFRRI Safety Analysis Report, indicates that the reactor may be safely operated with TRIGA fuel at a power level of 1.1 MW.

3.1.2. PULSE MODE OPERATION Applicability This specification applies to the maximum thermal energy produced in the reactor as a result of a prompt critical insertion of reactivity.

Objective The objective is to ensure that the fuel temperature safety limit shall not be exceeded during pulse mode operation.

Specification The maximum step insertion of reactivity shall be $3.50 (2.45% ~k/k) in pulse mode.

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Based upon calculations detailed in the AFRRI Safety Analysis Report, an insertion of $3.50 (2.45% ~k/k) results in a peak fuel temperature of less than 830°C.

3.1.3. REACTIVITY LIMITATIONS Applicability These specifications apply to the reactivity condition of the reactor and the reactivity worth of control rods and experiments. They apply for all modes of operation.

Objective The objective is to guarantee that the reactor can be shut down at all times and that the fuel temperature safety limit shall not be exceeded.

Specifications

a. The reactor shall not be operated with the maximum available excess reactivity greater than $5.00 (3.5% ~k/k).

b. The shutdown margin provided by the remaining control rods with the most reactive control rod in the most reactive position shall be greater than $0.50 (0.35% ~k/k) with the reactor in the reference core condition, all irradiation facilities and experiments in place, and the total worth of all non-secured experiments in their most reactive state.

a. The limit on available excess reactivity establishes the maximum achievable power should all control rods be in their most reactive positions.

b. The value of the shutdown margin ensures that the reactor can be shut down from any operating condition, even if the most reactive control rod remains in its most reactive position.

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3.2. REACTOR CONTROL AND SAFETY SYSTEMS 3.2.1. REACTOR CONTROL SYSTEM Applicability This specification applies to the channels monitoring the reactor core which shall provide information to the reactor operator during reactor operation. It also specifies the minimum number of operable control rod drives.

Objective The objective is to require that sufficient information be available to the operator as well as a sufficient number of operable control rod drives to ensure safe operation of the reactor.

Specifications

a. The reactor shall not be operated unless the measuring channels listed in Table 1 are operable for the specific mode of operation.

b. The reactor shall not be operated unless the four control rod drives are operable except:

a. the reactor may be operated at a power level no greater than 250kw with no more than one control rod drive inoperable with the associated control rod drive fully inserted.

c. The time from scram initiation to the full insertion of any control rod from a full up position shall be less than 1 second.

Table 1. Minimum Measuring Channels Effective Mode Measuring Channel Steady State Pulse

!Fuel Temperature Safety Channel 2 2

.__,inear Power Channel 1 0

....,og Power Channel 1 0 High-Flux Safety Channel 2 1 (1) Any Linear Power, Log Power, High-Flux Safety or Fuel Temperature Safety Channels may be inoperable while the reactor is operating for the purpose of performing a channel check, test, or calibration.

(2) If any required measuring channel becomes inoperable while the reactor is operating for reasons other than that identified in the previous footnote ( 1) above, 12

the channel shall be restored to operation within five minutes or the reactor shall be immediately shutdown.

Fuel temperature displayed at the control console gives continuous information on this parameter, which has a specified safety limit. The power level channels ensure that radiation-indicating reactor core parameters are adequately monitored for both steady state and pulsing modes of operation.

The specifications on reactor power level indication are included in this section because power level is related to the fuel temperature. The four control rod drives must be op*erable or the control rods inserted for the safe operation of the reactor. This specification ensures that the reactor will be promptly shut down when a scram signal is initiated. Experience and analysis indicate that, for the range of transients in a TRIGA reactor, the specified scram time is adequate to ensure the safety of the reactor.

At 25% power, the peak fuel temperature has been measured to be approximately 200 degrees C, 50% of the maximum expected fuel temperature at full power. Even if 30% of the core were to be effectively disabled by the inoperable control rod, the remaining elements would stay well below the established safety margins. This provides sufficient safety margin to ensure safe operations.

For footnote (1), taking these measuring channels off-line for short durations for the purpose of a check, test, or calibration is considered acceptable because in some cases, the reactor must be in operation in order to perform

• the check, test or calibration. For footnote (2), events which lead to these circumstances are self-revealing to the operator.

3.2.2. REACTOR SAFETY SYSTEMS Applicability This specification applies to the reactor safety systems.

Objective The objective is to specify the minimum number of reactor safety system channels that shall be operable for safe operation.

Specification The reactor shall not be operated unless the safety systems described in Tables 2 and 3 are operable for the specific mode of operation.

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Table 2. Minimum Reactor Safety System Scrams Channel Maximum Set Point Effective Mode Steady State Pulse fuel Temperature 600°C 2 2

[Percent Power, High Flux 1.1 MW 2 0 Console Manual Scram Button Closure switch 1 1 High Voltage Loss to Safety Channel 20% Loss 2 1

!Pulse Time 15 seconds 0 1

~mergency Stop Closure switch 3 3

( 1 in each exposure room, 1 on console) 14 feet from the top of Pool Water Level ~he core 1 1

!Watchdog (DAC to CSC) On digital console 1 1 The fuel temperature and power level scrams provide protection to ensure that the reactor can be shut down before the fuel temperature safety limit is exceeded. The manual scram allows the operator to shut down the system at any time if an unsafe or abnormal condition occurs. In the event of failure of the power supply for the safety channels, operation of the reactor without adequate instrumentation is prevented. The preset pulse timer ensures that the reactor power level will return to a low level after pulsing. The emergency stop allows personnel trapped in a potentially hazardous exposure room, or the reactor operator, to scram the reactor through the facility interlock system. The pool water level ensures that a loss of biological shielding would result in a reactor scram. The watchdog scram ensures reliable communication between the Data Acquisition Computer (DAC) and the Control System Computer (CSC).

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Table 3. Minimum Reactor Safety System Interlocks Effective Mode Action Prevented Steady State Pulse Pulse initiation at power levels greater than 1 kW x Withdrawal of any control rod except transient x Any rod withdrawal with count rate below 0.5 cps as measured by the operational channel x x Simultaneous manual withdrawal of two standard rods x Any rod withdrawal if high voltage is lost to the operational channel x x Withdrawal of any control rod if reactor period is less than 3 seconds x Application of air if the transient rod drive is not fully down.

This interlock is not required in square wave mode. x

• Reactor safety system interlocks shall be tested daily whenever operations involving these functions are planned The interlock preventing the initiation of a pulse at a power level above 1 kW ensures that the pulse magnitude will not allow the fuel element temperature to exceed the safety limit. The interlock that prevents movement of standard control rods in pulse mode will prevent the inadvertent increase in steady state reactor power prior to initiation of a pulse. Requiring a minimum count rate to be measured by the operational channel ensures sufficient source neutrons to bring the reactor critical under controlled conditions. The interlock that prevents the simultaneous manual withdrawal of two standard control rods limits the amount of reactivity added per unit time. Correct high voltage to the operational channel ensures accurate power indications. Preventing the withdrawal of any control rod if the period is less than 3 seconds minimizes the possibility of exceeding the maximum permissible power level or the fuel temperature safety limit.

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3.2.3. FACILITY INTERLOCK SYSTEM Applicability This specification applies to the interlocks that prevent the accidental exposure of an individual in either exposure room.

Objective The objective is to provide sufficient warning and interlocks to prevent movement of the reactor core to the exposure room in which someone may be working, or prevent the inadvertent contact between the core and the lead shield doors.

Specifications Facility interlocks shall be provided so that:

a. The reactor cannot be operated unless the lead shield doors within the reactor pool are either fully opened or fully closed;

b. The reactor cannot be operated unless the exposure room plug door adjacent to the reactor core position is fully closed and the lead shield doors are fully closed; or if the lead shield doors are fully opened, both exposure rooms plug doors must be fully closed; and

c. The lead shield doors cannot be opened to allow movement into the exposure room projection unless a warning horn has sounded in that exposure room, or unless two licensed reactor operators have visually inspected the room to ensure that no personnel remain in the room prior to securing the plug door.

These interlocks prevent the operation and movement of the reactor core into an area until there is assurance that inadvertent exposures or facility damage will be prevented.

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3.3. COOLANT SYSTEMS Applicability This specification refers to operation of the reactor with respect to the temperature and condition of the pool water.

Objective

a. To ensure the effectiveness of the resins in the water purification system;

b. To prevent activated contaminants from becoming a radiological hazard; and

c. To protect the integrity of the reactor core Specifications

a. The reactor shall not be operated if the bulk water temperature exceeds 60°C;

b. The reactor shall not be operated if periodic measurements taken IAW TS 4.3 show conductivity of the bulk water greater than 5 micromhos/cm; and

c. Both audible and visual alarms shall be provided to alert the AFRRI security guards and other personnel to any drop in reactor pool water level greater than 6 inches.

d. The reactor shall not be operated if the measurement required by TS 4.3 shows concentrations of radionuclides above the values in 10CFR part 20 appendix B table 2 are found in the primary coolant until the source of the activity is determined and appropriate corrective actions are taken.

Manufacturer data states that the resins in the water purification system break down with sustained operation in excess of 60°C. Based on experience, activation of impurities in the bulk water at power levels below 5 kW does not pose a significant radiological hazard. The conductivity limits are established to provide acceptable control of corrosion and are consistent with the fuel vendor recommendation and experience at similar reactors. The water level monitoring system provides prompt notification of a potential loss of primary coolant.

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3.4. VENTILATION SYSTEM Applicability This specification applies to the operation of the facility ventilation system.

Objective The objective is to ensure that the ventilation system is operable to mitigate the consequences of possible releases of radioactive material resulting from reactor operation.

Specification

a. The reactor shall not be operated unless the facility ventilation system is operating, except for periods of time not to exceed two continuous hours to permit repair, maintenance, or testing. The ventilation system is designed such that if operable there is negative pressure in the reactor room. In the event of a release of airborne radioactivity in the reactor room above routine reactor operation and normal background values, the ventilation system to the reactor room shall be automatically secured via closure dampers by a signal from the reactor deck continuous air particulate monitor.

b. The reactor shall not be operated in exposure room 1or2

.1. If the relative air pressure in the exposure room in use is greater than the reactor prep area (room 1105) except for periods of time not to exceed two continuous hours to permit repair, maintenance, or testing when the dampers shall be closed.

or;

2. The prep area RAM E3 or E6 is alarming.

During normal operation of the ventilation system, the concentration of argon-41 in unrestricted areas is below the limits allowed by 10 CFR Part 20. In the event of a fuel cladding rupture resulting in a substantial release of airborne particulate radioactivity, the ventilation system dampers shall be closed, thereby isolating the reactor room. Therefore, operation of the reactor with the ventilation secured for short periods of time ensures the same degree of control of release of radioactive materials. Moreover, radiation monitors within the building independent of those in the ventilation system provide warning if high levels of radiation are detected with the ventilation system secured.

18

3.5. RADIATION-MONITORING SYSTEM AND EFFLUENTS 3.5.1. MONITORING SYSTEM Applicability This specification applies to the functions and essential components of the radiation-monitoring system which shall be operable during reactor operations.

Objective The objective is to ensure that adequate radiation monitoring channels shall be available to the operator to ensure safe operation of the reactor.

Specifications The reactor shall be secured unless the following radiation monitoring systems are operable:

a. Radiation Area Monitoring System:

L 2 RAMS on the reactor Deck (Room 3160) are operable ii. If operating in an exposure room (ERl or ER2) the RAM adjacent to the exposure room in use shall be operable

b. Stack Gas Monitor: The stack gas monitor (SGM) shall sample and measure the gaseous effluent in the exhaust system;

c. Continuous Air Particulate Monitor: The continuous air particulate monitor (CAM) shall sample the air above the reactor pool. This unit shall be sensitive to radioactive particulate matter. Alarm of this unit shall initiate closure of the ventilation system dampers, restricting air leakage from the reactor room; and

d. Table 4 specifies the alarm and readout system for the above monitors.

19

Table 4. Locations of Radiation Monitoring Systems Location(s) of readouts Sampling Location Audible alarms and visual Indicators eactor Room (2 required) eactor and Control Rooms xp. Room 1 Area rep Area and Control Room x . Room 2 Area re Area and Control Room eactor and Control Rooms eactor and Control Rooms This system is intended to characterize the normal operational radiological environment of the facility and to aid in evaluating abnormal operations or conditions. The radiation monitoring system provides information to the operating personnel of any existing or impending danger from radiation. The automatic closure of the ventilation system dampers restricts the flow of airborne radioactive material to the environment.

3.5.2. EFFLUENTS

ARGON-41 DISCHARGE LIMIT Applicability This specification applies to the quantity of argon-41 that may be

• discharged from the AFRRI TRIGA reactor facility.

Objective The objective is to ensure that the radiation dose to members of the public due to the discharge of argon-41 from the AFRRI TRIGA reactor facility shall be below the value specified in 10 CFR Part 20.

Specifications

a. An environmental radiation monitoring program shall be maintained to determine the effects of the facility on the environs; and 20

b. If calculations, which shall be performed at least quarterly but not to exceed 20 MWh of operation, indicate that argon-41 release in excess of 313.5 curies to the unrestricted environment could be reached during the year as a result of normal reactor operations, reactor operations that generate and release significant quantities of argon-41 shall be curtailed for the remainder of the year as needed to ensure adherence with the 10 Rem constraint.

As described in the AFRRI Safety Analysis Report, COMPLY analysis indicates that the release of 313 .5 curies from the stack to the unrestricted environment in one calendar year yields a dose to the maximally exposed member of the public of 9.9 mrem. Therefore, limiting argon-41 release to less than 313.5 curies ensures that 10 CFR Part 20 limits on doses to the public are not exceeded. The upper limit of 20 MWh of reactor operation between gaseous effluent analyses ensures it is not possible to exceed 15% of the 10 mrem limit between reports.

3.6. LIMITATIONS ON EXPERIMENTS Applicability This specification applies to experiments installed in the reactor and its experimental facilities.

Objective The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment malfunction such that airborne concentrations of activity averaged over a year do not exceed 10 CFR Part 20, Appendix B.

Specifications The following limitations shall apply to the irradiation of experiments:

a. If the possibility exists that a release of radioactive gases or aerosols may occur;

1. The amou.nt and type of material irradiated shall be limited to ensure yearly compliance with Table 2, Appendix B, of 10 CFR Part 20, assuming that 100% of the gases or aerosols escape;

2. The ventilation system shall be operational while the samples are being transferred from the pool or the reactor core.

21

b. Each fueled experiment shall be limited such that the total inventory of iodine isotopes 131through135 in the experiment is not greater than 1.0 curies, and the maximum strontium-90 inventory is not greater than 5.0 millicuries;

c. Known explosive materials shall not be irradiated in the reactor in quantities greater than 25 milligrams. In addition, the pressure produced in the experiment container upon detonation of the explosive shall have been determined experimentally, or by calculations, to be less than half the design failure of the container;

d. Samples shall be doubly contained when release of the contained material could cause corrosion of the experimental facility or damage to the reactor;

e. The sum of the absolute reactivity worth of all experiments in the reactor and in the associated experimental facilities shall not exceed $3.00 (2.1 % ~k/k). This includes the total potential reactivity insertion that might result from experiment malfunction, accidental experiment flooding or voiding, and accidental removal or insertion of experiments. The absolute reactivity worth of any single secured experiment shall not exceed $3.00 (2.1 % ~k). The absolute reactivity worth of any single moveable or unsecured experiment shall be less than $1.00 (0.70%

~k). The combined absolute reactivity worth of multiple moveable or unsecured experiments in the reactor and associated experimental facilities at the same time shall be less than $1.00 (0.70% ~k);

f. In calculations regarding experiments, the following assumptions shall be made:

1. If the effluent exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of the particles produced can escape; and

2. For a material whose boiling point is above 55°C and where vapor formed by boiling the material can escape only through an undisturbed column of water above the core, up to 10% of the vapor can escape;

g. If an experimental container fails and releases materials that could damage the reactor fuel or structure by corrosion or other means, physical inspection of the reactor fuel and structure shall be performed to identify damage and potential need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Reactor Facility Director and shall be determined to be satisfactory before operation of the reactor is resumed; and

h. Experiments shall be designed such that failure of one experiment shall not contribute to the failure of any other experiment. All operations in an experimental facility shall be supervised by a member of the reactor operations staff. All experiments shall be either secured or observed for mechanical 22

stability to ensure that unintended movement will not cause an unplanned reactivity change in excess of $1.00.

a. This specification is intended to provide assurance that airborne activities in excess of the limits of Appendix B of 10 CFR Part 20 will not be released to the atmosphere outside the facility boundary.

b. The 1.0 curie limitation on iodine isotopes 131 through 135 and 5.0 millicurie limitation on strontium-90 ensures that, in the event of malfunction of a fueled experiment leading to total release of radioactive material including fission products, the dose to any individual will not exceed the limits of 10 CFR Part 20.

c. This specification is intended to prevent damage to reactor components resulting from malfunction of an experiment involving explosive materials.

d. This specification is intended to provide an additional safety factor where damage to the reactor and components is possible if an experiment container fails.

e. The maximum worth of experiments is limited such that their removal from the reactor at the reference core condition will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. The $3.00 limit is less than the authorized pulse magnitude. Limiting moveable or unsecured experiments to a worth less than $1.00 will prevent unintended pulsing of the reactor and unnecessary fuel mechanical stress.

f. This specification is intended to ensure that the limits of 10 CFR Part 20, Appendix B, are not exceeded in the event of an experiment malfunction.

g. This specification is intended to ensure that operation of the reactor with damaged reactor fuel or structure is prevented.

h. This specification ensures that unintended movement will not cause an unplanned reactivity change, physical damage or contribute to the failure of any other experiment.

23

3.7. ,FUELPARAMETERS Applicability This specification applies to all fuel elements.

Objective The objective is to maintain integrity of the fuel element cladding.

,

Specification

1. The reactor shall not operate with damaged fuel elements, except for the purpose of locating damaged fuel elements. A fuel element shall be considered damaged and removed from the core if:

a. The transverse bend exceeds 0.0625 inches over the length of the cladding;

b. The length exceeds its original length by 0.100 inches;

c. A cladding defect exists as indicated by the release of fission products; or

d. Visual inspection identifies bulges, gross pitting, or corrosion.

2. The burnup of uranium-235 in the UZrH fuel matrix shall not exceed 50 percent of the initial concentration.

Gross failure or obvious visual deterioration of the fuel is sufficient to warrant declaration of the fuel element as dam_aged. The elongation and bend limits are the values found acceptable to the USNRC (NUREG-1537).

24

4.0. SURVEILLANCE REQUIREMENTS Applicability These specifications apply to the surveillance requirements for reactor systems.

Objective The objective is to allow for variation in the execution of surveillance when maintenance issues prevent the timely completion surveillance items.

Specifications Surveillance requirements may be deferred during reactor shutdown (except TS 4.4, TS 4.5.1 and TS 4.5.2) however; they shall be completed prior to reactor startup unless reactor operation is required for performance of the surveillance. Such surveillance shall be performed as soon as practical after reactor startup. Scheduled surveillance which cannot be performed with the reactor operating may be deferred until a planned reactor shutdown.

TS Possible to defer Required prior to during shutdowns? routine operations?

1. 4.1 Reactor core parameters Yes Yes

2. 4.2.1 Reactor Control Systems Yes Yes

3. 4.2.2 Reactor Safety Systems Yes Yes

4. 4.2.3 Fuel Temperature Yes Yes

5. 4.2.4 Facility Interlock System Yes Yes

6. 4.3 Coolant Systems Yes Yes

7. 4.4 Ventilation Systems No NIA

8. 4.5 .1 Monitoring System No Yes

9. 4.5.2 Effluents No NIA

10. 4.6 Reactor Fuel Elements Yes Yes

11. 4.2.2 Low Pool Water Scram Yes No The surveillances items listed in the table above cannot be executed during times when required equipment are out of service. During these times, reactor operations are also suspended; therefore there is no decrement in safety deferring these surveillance items.

25

4.1. REACTOR CORE PARAMETERS Applicability These specifications apply to the surveillance requirements for reactor core parameters.

Objective The objective is to verify that the reactor does not exceed the authorized limits for power, shutdown margin, core excess reactivity, and verification of the total reactivity worth of each control rod.

Specifications

a. The reactivity worth of each standard control rod/transient rod and the shutdown margin shall be determined annually, not to exceed 15 months, or following any significant (>$0.25) changes to core configuration (excluding in-core experiments).

b. The reactivity worth of an experiment shall be estimated before reactor power operation with the experiment the first time it is performed. If the absolute reactivity worth is estimated to be greater than $0.25, the worth shall be measured at a power level less than 1 kW.

c. The core excess reactivity shall be measured each day of operation involving the movement of control rods, or prior to each continuous operation exceeding more than a day, and following any significant (>$0.25) core configuration changes.

At a minimum excess reactivity shall be measured annually, not to exceed 15 months. This measurement is also a complete channel test of the linear power channel and log power channel.

d. The power coefficient of reactivity at 100 kW and 1 MW shall be measured annually, not to exceed 15 months.

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

Excess reactivity measurements ensure that core configuration remains unchanged.

Knowledge of power coefficients allows the operator to accurately predict the reactivity necessary to achieve required power levels.

26

4.2. REACTOR CONTROL AND SAFETY SYSTEMS 4.2.1. REACTOR CONTROL SYSTEMS Applicability These specifications apply to the surveillance requirements for reactor control systems.

Objective The objective is to verify the operability of system components that affect the safe and proper control of the reactor.

Specifications

a. The standard control rods/transient rod shall be visually inspected for damage and deterioration annually, not to exceed 15 months.

b. The control rod drop times of all rods shall be measured semiannually, not to exceed 7 .5 months. After work is done on any rod or its rod drive mechanical components, the drop time of that particular rod shall be verified.

c. On each day that pulse mode operation of the reactor is planned, the transient rod system is channel tested to verify that the system is operable. Semiannually, not to exceed 7 .5 months, the transient rod drive cylinder and the associated air supply system shall be inspected, cleaned, and lubricated as necessary.

Visual inspection of the standard control rods/transient rod is made to evaluate corrosion and wear characteristics caused by operation in the reactor. Channel tests along with periodic maintenance ensure consistent performance.

Measurement of the rod drop times on a semiannual basis or after mechanical maintenance is a verification of the scram system and provides an indication of the capability of the control rods to perform properly.

27

4.2.2. REACTOR SAFETY SYSTEMS Applicability These specifications apply to the surveillance requirements for measurement, test, and calibration of the reactor safety systems.

Objective The objective is to verify the performance and operability of the systems and components that are directly related to reactor safety.

Specifications.

a. A channel test of the scram function of the high-flux safety channels shall be made each day that reactor operations are planned.

b. A channel test of each of each of the reactor safety system channels for the intended mode of operation shall be performed weekly, whenever operations are planned.

c. Channel calibration shall be made of the NP, NPP, NMlOOO, NLW, NMP or any other console instrumentation designated to provide direct power level information to the operator, annually not to exceed 15 months.

d. A thermal power calibration shall be completed annually not to exceed 15 months.

e. The emergency stop scram shall be tested annually, not to exceed 15 months.

f. The low pool water scram shall be tested weekly not to exceed 10 days whenever operations are planned.

g. The console manual scram button shall be tested weekly not to exceed 10 days whenever operations are planned.

TRIGA system components have proven operational reliability. Daily tests ensure reliable scram functions and ensure the detection of channel drift or other possible deterioration of operating characteristics. The channel checks ensure that the safety system channel scrams are operable on a daily basis or prior to an extended run. The power level channel calibration will ensure that the reactor is operated within the authorized power levels.

• 28

4.2.3. FUEL TEMPERATURE Applicability These specifications apply to the surveillance requirements for the safety channels measuring the fuel temperature.

Objective The objective is to ensure operability of the fuel temperature measuring channels.

Specifications

a. A channel check of the fuel temperature scrams shall be made each day that the reactor is to be operated.

b. A channel calibration of the fuel temperature measuring channels shall be made annually, not to exceed 15 months.

c. A weekly channel test shall be performed on fuel temperature measuring channels, whenever operations are planned.

d. If a reactor scram caused by high fuel element temperature occurs, an evaluation shall be conducted to determine whether the fuel element temperature exceeded the safety limit.

Operational experience with the TRIGA systems demonstrates that annual calibration and weekly channel tests provides reliable fuel temperature measurements. The daily scram channel check ensures scram capabilities.

29

4.2.4. FACILITY INTERLOCK SYSTEM Applicability This specification applies to the surveillance requirements that ensure the integrity of the facility interlock system.

Objective The objective is to ensure performance and operability of the facility interlock system.

Specifications Functional checks shall be made annually, not to exceed 15 months, to ensure the following:

a. With the lead shield doors open, neither exposure room plug door can be electrically opened.

b. The core dolly cannot be moved into region 2 with the lead shield doors closed.

c. The lead shield doors cannot be opened to allow movement into the exposure room projection unless a warning horn has sounded in that exposure room, or unless two licensed reactor operators have visually inspected the room to ensure that no personnel remain in the room prior to securing the plug door.

These functional checks will verify operation of the interlock system.

Experience at AFRRI indicates that this is adequate to ensure operability.

30

4.3. COOLANT SYSTEMS Applicability This specification applies to the surveillance requirements for monitoring the pool water and the water conditioning system.

Objective The objective is to ensure the integrity of the water purification system, thus maintaining the purity of the reactor pool water, minimizing possible radiation hazards from activated impurities in the water system, and limiting the potential corrosion of fuel cladding and other components in the primary water system.

Specifications

a. The pool water temperature, as measured near the input to the water purification system, shall be measured daily, whenever operations are planned.

b. The conductivity of the bulk water shall be measured monthly, not to exceed 6 weeks.

c. The reactor coolant shall be measured for radioactivity quarterly, not to exceed 4 months.

d. The audible and visual reactor pool level alarms shall be tested quarterly, not to exceed 4 months.

Based on experience, observation at these intervals provides acceptable surveillance of limits that ensure that fuel cladding corrosion and neutron activation of dissolved materials are minimized. Testing of the audible and visual alarms ensures that personnel will be able to detect and respond to pool water loss in a timely manner.

The pool water temperature is continuously displayed on the reactor console and is manually recorded at the beginning of each day of reactor operations. The conductivity of the bulk pool water is monitored to help minimize the activation of impurities in the water system and monitor the possibility of corrosion in the fuel cladding or reactor system components.

31

4.4. VENTILATION SYSTEM Applicability This specification applies to isolation of the facility ventilation system.

Objective The objective is to ensure the proper operation of the ventilation system in controlling the release of radioactive material into the unrestricted environment.

Specification

1. The operating mechanism of the ventilation system dampers in the reactor room shall be verified to be operable and visually inspected monthly, not to exceed 6 weeks.

2. The relative air pressure in the reactor room and exposure room to be. used shall be verified to be negative each day operations in the affected exposure room are planned.

3. The reactor exhaust damper flow failure closure system shall be tested each day that reactor operations are planned.

Experience accumulated over years of operation has demonstrated that tests of the ventilation system dampers on a monthly basis are sufficient to ensure proper operation of the system and control of the release of radioactive material.

32

4.5. RADIATION-MONITORING SYSTEM AND EFFLUENTS 4.5.1. MONITORING SYSTEM Applicability This specification applies to surveillance requirements for the radiation monitoring system.

Objective The objective is to ensure that the radiation munitoring equipment is operating and to verify the appropriate alami settings.

Specification The radiation area monitoring, continuous air particulate monitoring, and stack gas monitoring systems shall be channel tested quarterly, not to exceed 4 months. A channel check of these systems shall be performed daily to verify operability when operations are planned. These systems shall be calibrated annually, not to exceed 15 months.

Experience has shown that quarterly verification of radiation area monitoring, continuous air particulate monitoring, and stack gas monitoring systems set points in conjunction with a quarterly channel test is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span.

33

4.5.2. EFFLUENTS Applicability This specification applies to surveillance requirements for environmental monitoring.

Objective The objective is to ensure the health and safety of the public through detection of the release of radioactive material to the environment.

Specifications

a. The unrestricted area outside of AFRRI shall be monitored by dosimeters that shall be analyzed quarterly, not to exceed 4 months.

b. Samples of soil, vegetation, and water in the vicinity of the reactor shall be collected and tested for radioactivity quarterly, not to exceed 4 months.

c. A gaseous effluent release report shall be generated quarterly or every 20 MW hours of reactor operations (whichever comes first) to ensure radioactive effluents will not exceed the annual dose limits to the public.

Experience has shown that quarterly environmental monitoring is sufficient to detect and quantify any release of radioactive material from research reactors. The requirement for gaseous effluent release reports will ensure that Ar-41 production from normal reactor operations does not exceed 10CFR20 annual dose limits to the public.

34

4.6. REACTOR FUEL ELEMENTS Applicability This specification applies to the surveillance requirements for the fuel elements.

Objective The objective is to verify the specifications for fuel elements are met.

Specifications Fuel elements shall be inspected visually for damage or deterioration and measured for length and bend in accordance with the following:

a. Before being placed in the core for the first time or following long-term storage;

b. Every two years, not to exceed 30 months, or at intervals not to exceed 500 pulses of insertion greater than $2.00, whichever comes first, for fuel elements in the B, C, and D rings;

c. Every four years (not to exceed 54 months), or at intervals not to exceed 500 pulses of insertion greater than $2.00, whichever comes first, for fuel elements in the E and Frings; and

d. If damage, deterioration, or unacceptable length and bend measurements are found in one or more fuel elements, all fuel elements in the core shall be inspected for damage or deterioration and measured for length and bend.

The frequency of inspection and measurement is based on the parameters most likely to affect the fuel cladding of a pulse reactor. Inspecting fuel elements in rings with higher power factors more frequently will provide early indication of fueJ damage while significantly reducing the amount of fuel movement required.

35

5.0. DESIGN FEATURES 5.1. SITE AND FACILITY DESCRIPTION Applicability This specification applies to the reactor building.

Objective The objective is to restrict the amount of radioactivity released into the environment.

Specifications

a. The reactor building, as a structurally independent building in the AFRRI complex, shall have its own ventilation system branch. The effluent from the reactor ventilation system shall exhaust through absolute filters to a stack having a minimum elevation that is 18 feet above the roof of the highest building in the AFRRI complex.

b. The reactor room shall contain a minimum free volume of 22,000 cubic feet.

c. The ventilation system air ducts to the reactor room shall be equipped with dampers which automatically close off ventilation to the reactor room upon a signal from the reactor room continuous air particulate monitor.

d. The reactor room shall be designed to restrict air leakage when the ventilation system dampers are closed.

e. The reactor areas exhausting through the reactor ventilatio.n system shall include the Controlled Access Area (CAA) and the Reactor Control Area (RCA). The specific rooms included in each of those areas shall be listed in the Physical Security Plan for the AFRRI TRIGA Reactor Facility.

f. The reactor is housed in building #42 of the AFRRI complex and the restricted areas are located within that structure. The restricted areas are described in the SAR for the AFRRI reactor facility section 1.~.1 including figures 2-2 through 2-4 which describe the location of the reactor in the AFRRI complex.

Figures 3-1 through 3-4 are the floor plan layouts which identify the reactor areas The facility is designed so that the ventilation system will normally maintain a negative pressure with respect to adjacent areas, limiting personnel exposure. The free air volume within the reactor building is confined when there is an emergency shutdown of the ventilation system. Building construction and gaskets around 36

doorways help restrict leakage of air into or out of the reactor room. The stack height ensures an adequate dilution of effluents well above ground level. The separate ventilation system branch ensures a dedicated air flow system for reactor effluents and shall exhaust from all reactor spaces.

5.2. REACTOR CORE AND FUEL 5.2.1. REACTOR FUEL Applicability These specifications apply to the fuel elements including fuel follower control rods used in the reactor core.

Objective The objectives are to (1) ensure that the fuel elements are designed and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics, and (2) ensure that the fuel elements used in the core are comparable to those analyzed in the Safety Analysis Report.

Specifications The individual non-irradiated TRIGA fuel elements shall have the following characteristics:

a. Uranium content: Maximum of 9.0 weight percent enriched to less than 20% uranium-235. In the fuel follower, the maximum uranium content shall be 12.0 weight percent enriched to less than 20% uranium-235.

b. Hydrogen-to-zirconium atom ratio (in the ZrHx): Nominal 1.7 H atoms to 1.0 Zr atoms with a range between 1.6 and 1.7.

c. Cladding: 304 stainless steel, nominal 0.020 inches thick.

d. Any burnable poison used for the specific purpose of compensating for fuel bumup or long-term reactivity adjustments shall be an integral part of the manufactured fuel elements.

A maximum uranium content of 9.0 weight percent in a !RIGA element is greater than the design value of 8.5 weight percent and encompasses the maximum probable variation in individual elements. Such an increase in loading would result in an increase in power density of less than 6%. The 37

hydrogen-to-zirconium ratio of 1.7 will produce a maximum pressure within the cladding that is well below the rupture strength of the cladding. The local power density of a 12.0 weight percent fuel follower is 21 % greater than an 8.5 weight percent TRIG A fuel element in the D ring. The volume of fuel in a fuel follower control rod is 56% of the volume of a TRIGA fuel element.

Therefore, the actual power produced in the fuel follower rod is 33% less than the power produced in a TRIGA fuel element in the D ring.

5.2.2. REACTOR CORE Applicability These specifications apply to the configuration of fuel and in-core experiments.

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

Specifications

a. The reactor core shall consist of TRIGA reactor fuel elements in a close packed array with a minimum of two thermocouple instrumented TRIGA reactor fuel elements.

b. There shall be four single core positions occupied by the three standard control rods and transient rod, a neutron startup source with holder, and positions for possible in-core experiments.

c. The core shall be cooled by natural convection water flow.

d. In-core experiments shall not replace B ring, C ring, and/or D ring fuel elements within the reactor core.

TRIGA cores have been in use for decades and their safe operational characteristics are well documented. Analysis has shown that natural convection water flow provides sufficient cooling to ensure that the fuel temperature safety limit is not exceeded during reactor operations in accordance with the Technical Specifications. Placement of in-core experiments in the B ring, C ring, and/or D ring is restricted to ensure safe power peaking in adjacent fuel element positions.

38

5.2.3. CONTROL RODS Applicability These specifications apply to the control rods used in the reactor core.

Objective The objective is to ensure that the control rods are designed to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications

a. The standard control rods shall have scram capability, contain borated graphite, B4C powder, or boron and its compounds in solid form as a poison in aluminum or stainless steel cladding. These rods may have an aluminum, air, or fuel follower. If fuel followed, the fuel region will conform to Technical Specification 5.2.l.

b. The transient control rod shall have scram capability and contain borated graphite, B4C powder, or boron and its compounds in solid form as a poison in aluminum or stainless steel cladding. This rod may incorporate an aluminum, poison, or air follower.

The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, B4C powder, or boron and its compounds. These materials must be contained in a suitable cladding material, such as aluminum or stainless steel, to ensure mechanical stability during movement and to isolat~ the poison from the pool water environment. Scram capabilities are provided by the rapid insertion of the control rods, which is the primary operational safety feature of the reactor. The transient control rod is designed for use in a pulsing TRIGA reactor.

39

5.3. FUEL STORAGE Applicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Objective The objective is to ensure that stored fuel does not become critical and does not reach an unsafe temperature.

Specification All fuel elements not in the reactor core shall be stored and handled in accordance with applicable regulations. Irradiated fuel elements and fueled devices shall be stored in an array that permits sufficient natural convective cooling by water or air and that prevents the fuel element or fueled device temperature from exceeding design values. Storage shall be such that stored fuel elements and fueled devices remain subcritical under all conditions of moderation and reflection in a configuration where keff is not greater than 0.90.

The limits imposed by this specification are conservative and ensure safe storage and handling. Experience shows that approximately 67 TRIGA fuel elements in a closely packed array are required to achieve criticality. Calculations show that in the event of a full storage rack failure with all 12 elements falling in the most reactive nucleonic configuration, the mass would be less than that required for criticality.

Therefore, under normal storage conditions, criticality cannot be reached.

40

6.0. ADMINISTRATIVE CONTROLS 6.1. ORGANIZATION 6.1.1. STRUCTURE The organizational structure of the reactor facility is depicted below.

Figure 1. Organization of Personnel for Management and Operation of the AFRRI Reactor Facility Level 1

.A.FRRI Licensee Level2 AFRRI Head, Radiation Reactor and A:ii.S*:O:")' *Radiation Radiation Safety

• Officer

-- - - - Sciences Department.

- - - -*-

A::*.*'is.:.:-v Facilities Safety Reactor Facility Subcommittee Director Level 3 Reactor Operations Supervisor Level4 Reactor Operations Staff*

"'Any reactor staff member has direct access to the AFRRl Licensee for matters concerning safety.

41

MANAGEMENT LEVELS Level 1: AFRRI Director: Responsible for the facility license.

Level 2: Reactor Facility Director: Responsible for reactor facility operations and administration shall report to Level 1.

Level 3: Reactor Operations Supervisor: Responsible for the day-to-day operation of the reactor and shall report to Level 2.

Level 4: Reactor Operating Staff: Licensed reactor operators and senior reactor operators and trainees. These individuals shall report to Level 3 for matters involving reactor operations.

6.1.2. RESPONSIBILITY The AFRRI Licensee shall have license responsibility for the reactor facility.

The Reactor Facility Director (RFD) shall be responsible for administration and operation of the reactor facility and for determination of applicability of procedures, experiment authorizations, maintenance, and operations. The Reactor Facility Director may designate an individual who meets the requirements of Technical Specifications 6.1.3.1.a to discharge these responsibilities during an extended absence. During brief absences (periods less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) of the Reactor Facility Director and his designee, the Reactor Operations Supervisor shall discharge these responsibilities. The Radiation Safety Officer shall be responsible for implementing the radiation safety program for the AFRRI TRIGA reactor. The requirements of the radiation safety program are established in 10CFR20. The program shall comply with the requirements in 10CFR20. Additional guidelines from ANSl/ANS-15.11-1993;R2004 "Radiation Protection at Research Reactor Facilities" should be considered.

6.1.3. STAFFING 6.1.3.1. Selection of Personnel

a. AFRRI Licensee The AFRRI Licensee is the AFRRI Director. The AFRRI Director has management responsibility for adhering to the terms and conditions of the AFRRI reactor license R-84, the AFRRI 02 byproduct license, the AFRRI Technical Specifications and for protecting the health and safety of the facility staff and members of the public.

b. Reactor Facility Director 42

At the time of appointment to this position, the Reactor Facility ,

Director shall have six or more years of nuclear experience.

The individual shall have a baccalaureate or higher degree in an engineering or scientific field. The degree may fulfill up to four years of experience on a one-for-one basis. The Reactor Facility Director shall have held a USNRC Senior Reactor Operator license on the AFRRI reactor for at least one year before appointment to this position.

c. Reactor Operations Supervisor At the time of appointment to this position, the Reactor Operations Supervisor shall have three years nuclear experience. Higher education in a scientific or engineering field may fulfill up to two years of experience on a one-for-one basis. The Reactor Operations Supervisor shall hold a USNRC Senior Reactor Operator license on the AFRRI reactor. In addition, the Reactor Operations Supervisor shall have one year of experience as a USNRC licensed Senior Reactor Operator at AFRRI or at a similar facility before the appointment to this position. ,

d. Reactor Operators/Senior Reactor Operators At the time of appointment to this position, an individual shall have a high school diploma or equivalent, and shall possess the appropriate USNRC license.

e. Additional reactor staff as required for support and training At the time of appointment to the reactor staff, an individual shall possess a high school diploma or equivalent.

6.1.3.2. Operations

a. Minimum staff when the reactor is not secured shall include:

1. A licensed Senior Reactor Operator on call, but not necessarily on site;

2. Radiation control technician on call, but not necessarily on site;

3. At least one licensed Reactor Operator or Senior Reactor Operator present in the control room; and 43

4. Another person within the AFRRI complex who is able to carry out written emergency procedures, instructions of the operator, or to summon help in case the operator becomes incapacitated.

5. One licensed Senior Reactor Operator may fill both the on call and control room positions simultaneously. In that case, the minimum staff is three persons.

b. A Senior Reactor Operator shall be present at the reactor during the following operations:

1. All fuel or control rod relocations within the reactor core region (control rod movement associated with routine reactor operation is not considered to be a relocation);

2. Initial reactor startup and approach to power;

3. Recovery from an unplanned or unscheduled shutdown.;

and

4. Relocation of any experiment with reactivity worth greater than $1.00.

c. A list of the names and telephone numbers of the following personnel shall be readily available to the operator on duty:

1. Management personnel (Reactor Facility Director, AFRRI Licensee) or designee;

2. Radiation safety personnel (AFRRI Radiation Safety Officer) or designee; and

3. Other operations personnel (Reactor Staff, Reactor Operations Supervisor) 6.1.3.3. Training of Personnel Training and retraining program shall be maintained to ensure adequate levels of proficiency in persons involved in the reactor and reactor operations. The training and retraining program will be consistent with the NRC-approved reactor requalification plan.

44

. 6.2. REVIEW AND AUDIT - THE REACTOR AND RADIATION FACILITIES SAFETY SUBCOMMITTEE CRRFSS) 6.2.1. COMPOSITION AND QUALIFICATIONS 6.2.1.1. Composition

a. Regular RRFSS Members (Permanent Members)

1. The following shall be members of the RRFSS:

a. AFRRI Radiation Safety Officer

b. AFRRI Reactor Facility Director

2. The following shall be appointed to the RRFSS by the AFRRI Licensee:

a. Chairman

b. One to three non-AFRRI members who are knowledgeable in fields related to reactor safety. At least one shall be a Reactor Operations Specialist or a Health Physics Specialist.

b. Special RRFSS Members (Temporary Members)

1. Other knowledgeable persons to serve as alternates in section 6.2.1.1.a.2.b above as appointed by the AFRRI Licensee.

2. Voting ad hoc members, appointed by the AFRRI Licensee to assist in !eview of a particular problem'.

c. Nonvoting members as appointed by the AFRRI Licensee.

6.2.1.2. Qualifications The minimum qualifications for a person on the RRFSS shall be six years of professional experience in the discipline <;>r specific field represented. A baccalaureate degree may fulfill four years of experience.

45

6.2.2. FUNCTION AND AUTHORITY 6.2.2.1. Function The RRFSS shall be directly responsible to the AFRRI Licensee.

The subcommittee shall review all radiological health and safety matters concerning the reactor and its associated equipment, the structural reactor facility, and those items listed in Section 6.2.4.

6.2.2.2. Authority The RRFSS shall report to the AFRRI Licensee and shall advise the Reactor Facility Director in those areas of responsibility specified in Section 6.2.4.

6.2.3. RULES 6.2.3.1. Alternates

• Alternate members may be appointed in writing by the RRFSS Chairman to serve on a temporary basis. No more than two alternates shall participate on a voting basis in RRFSS activities at any one time.

6.2.3.2. Meeting Frequency The RRFSS shall meet at least two times during a calendar year.

Any member of the RRFSS may submit a written request to the RRFSS Chairman to convene a special meeting of the RRFSS to discuss urgent matters.

6.2.3.3. Quorum A quorum of the RRFSS for review shall consist of a minimum of four members that can vote and occupy the following positfons; the Chairman (or designated alternate), the Reactor Facility Director (or designated alternate), the Radiation Safety Officer (or designated alternate), and one or more non-AFRRI member. A majority of those present shall be regular members. The operating staff shall not constitute a majority. A member may occupy two positions but may only vote once.

46

6.2.3.4. Voting Rules Each regular RRFSS member shall have one vote. Each special RRFSS member shall have one vote. The majority is 51 % or more of the regular and special members present and voting and concurrence between the Radiation Safety Officer and the Reactor Facility Director.

6.2.3.5. Minutes

a. Draft minutes of the previous meeting should be available to regular members at least one week before a regular scheduled meeting.

b. Once approved by the subcommittee, final minutes shall be submitted to level one management for review within 3 months.

6.2.4. REVIEW FUNCTION The RRFSS shall review:

a. Safety evaluations for (1) changes to procedures, equipment, or systems having safety significance and (2) tests or experiments conducted without NRC approval under provisions of Section 50.59 of 10 CFR.

b. Changes to procedures, equipment, or systems that change the original intent or use, are non-conservative, or those that meet any of the applicable criteria in Section 50.59 of 10 CFR;

c. Proposed tests or experiments that could affect reactivity or result in the uncontrolled release of radioactivity, or those that might meet any of the applicable criteria in Section 50.59 of 10 CFR;

d. Proposed changes in Technical Specifications, the Safety Analysis Report, or other license conditions;

e. Violations of applicable statutes, codes, regulations, orders, technical specifications, license requirements, or of internal procedures or instructions having safety significance;

f. Operating abnormalities having safety significance;

g. Events that have been reported to the NRC; and

h. Audit reports of the reactor facility operations.

47

6.2.5. AUDIT FUNCTION Audits of reactor facility operations shall be performed under the cognizance of the RRFSS, but in no case by the personnel responsible for the item audited. The audits shall be performed annually, not to exceed 15 months. A report of the findings and recommendations resulting from the audit shall be submitted to the AFRRI Licensee within three months after the report has been received. Deficiencies uncovered that affect. reactor safety shall immediately be reported to level one management. Audits may be performed by one or more individuals who need not be RRFSS members. These audits shall examine the operating records and the conduct of operations, and shall encompass the following:

a. Conformance of facility operation to the Technical Specifications and the license;

b. Performance, training, and qualifications of the reactor facility staff;

c. Results of all actions taken to correct deficiencies occurring in facility equipment, structurf'.S, systems, or methods of operation that affect safety;

d. Facility emergency plan and implementing procedures;

e. Facility Physical Security Plan;

f. Any other area of facility operations considered appropriate by the RRFSS or the AFRRI Licensee; and

g. Reactor Facility ALARA Program. This program may be a section of the total AFRRI program.

6.3. PROCEDURES Written procedures for certain activities shall be approved by the Reactor Facility Director and reviewed by the RRFSS. The procedures shall be adequate to ensure safe operation of the reactor, but shall not preclude the u~e of independent judgment and action as deemed necessary. Operational procedures shall be used for the following items:

a. Conduct of irradiation and experiments that could affect the operation and safety of the reactor;

b. Surveillance, testing, maintenance, and calibration of instruments, components, and systems involving nuclear safety; 48

c. Personnel radiation protection consistent with 10 CFR Part 20;

d. Implementation of required plans such as the Physical Security Plan and the Emergency Plan, consistent with restrictions on safeguards information;

e. Fuel loading, unloading, and movement within the reactor core; and

f. Reactor startup checklist, standard operations, and securing the facility.

Although substantive changes to the above procedures shall be made only with approval by the Reactor Facility Director, temporary changes to the procedures that do not change their original intent may be made by the Reactor Operations Supervisor. All such temporary changes shall be documented and subsequently reviewed and approved by the Reactor Facility Director.

6.4. REVIEW AND APPROVAL OF EXPERIMENTS Before issuance of a reactor authorization, new experiments shall be reviewed for radiological safety and approved by the following:

a. Reactor Facility Director

b. Health Physics Department

c. Reactor and Radiation Facilities Safety Subcommittee (RRFSS)

Prior-to its performance, an experiment shall be included under one of the following types of authorizations:

a. Special Reactor Authorization for new experiments or experiments not included in a Routine Reactor Authorization. These experiments shall be performed under the direct supervision of the Reactor Facility Director or designee.

b. Routine Reactor Authorization for approved experiments safely performed at least once. These experiments may be performed at the discretion of the Reactor Facility

, Director and coordinated with the Health Physics Department, when appropriate.

These authorizations do not require additional RRFSS review.

c. Reactor Parameters Authorization for routine measurements of reactor parameters, routine core measurements, instrumentation and calibration checks, maintenance, operator training, tours, testing to verify reactor outputs, and other reactor testing procedures. This shall constitute a single authorization. These operations shall be performed under the authorization of the Reactor Facility Director or the Reactor Operations Supervisor.

49

Substantive (> $0.25) changes to previously approved experiments shall be made only after review by the RRFSS and after approval (in writing) by the Reactor Facility Director or designated alternate to ensure that the change does not impact compliance with TS 3.6, LIMITATIONS ON EXPERIMENTS. Minor changes that do not significantly alter the experiment (<$0.25) may be approved by the Reactor Operations Supervisor. Approved experiments shall be carried out in accordance with established procedures.

50

6.5. REQUIRED ACTIONS 6.5.l. ACTIONS TO BE TAKEN IN CASE OF SAFETY LIMIT VIOLATION

a. The reactor shall be _shut down immediately, and reactor operation shall not be resumed without authorization by the USNRC.

b. The safety limit violation shall be reported to the USNRC, the AFRRI Licensee, and the RRFSS not later than the next working day.

c. A Safety Limit Violation Report shall be prepared. This report shall be reviewed by the RRFSS, and shall describe (1) applicable circumstances preceding the violation, when known, the cause and contributing factors (2) effects of the violation on facility components, structures, or systems, the health and safety of personnel and the public and (3) corrective action taken to prevent or reduce the probability of recurrence.

d. The Safety Limit Violation Report shall be submitted to the USNRC, the AFRRI Licensee, and the RRFSS within 14 days ofthe violation.

6.5.2. REPORTABLE OCCURRENCES The types of events listed below shall be reported as soon as possible by telephone and confirmed in writing by facsimile, e-mail, or similar transmission to the USNRC no later than the following working day after confirmation of the event, with a written follow-up report within 14 days.

The report shall include (as a minimum) the circumstances preceding the event, current effects on the facility, and status of corrective action. The report shall contain as much supplemental material as possible to clarify the situation. Supplemental reports may be required to fully describe the final resolution of the occurrence.

a. Operation with any safety system setting less conservative than specified in Section 2.2, Limiting Safety System Setting for Fuel Temperature.

b. Operation in violation of any Limiting Condition for Operation, Section 3, unless prompt remedial action is taken as permitted in section 3. *

c. Malfunction of a required reactor safety system component during operation that renders or could render the system incapable of performing its intended safety function unless the malfunction or condition is caused by maintenance.

d. Any unanticipated or uncontrolled change in reactivity greater than $1.00.

Reactor trips resulting from a known cause are excluded.

51

e. An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy causes or could have caused the existence or development of a condition that could result in operation of the reactor in a manner less safe than conditions covered in the Safety Analysis Report.

f. The release of fission products from a fuel element through degradation of the fuel cladding. Possible degradation may be determined through an increase in the background activity level of the reactor pool water.

g. Abnormal and significant degradation of the reactor coolant boundary or confinement boundary (excluding minor leaks).

h. A release of radioactivity that exceeds or could have exceeded the limits allowed by 10 CFR Part 20, or these Technical Specifications.

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

J. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the Safety Analysis Report or in the bases for the Technical Specifications that have or could have permitted reactor operation with a smaller margin of safety than in the original analysis.

k. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the Safety Analysis Report or Technical Specifications bases, or discovery during facility life of conditions not specifically considered in the Safety Analysis Report or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

6.5.3. ACTIONS TO BE TAKEN IN CASE OF REPORTABLE OCCURRENCES

a. Reactor conditions shall be returned to normal, or the 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 the Reactor Facility Director or designated alternate.

b. The occurrence shall be reported to the Reactor Facility Director or designated alternate and to the USNRC.

c. The occurrence shall be reviewed by the RRFSS at its next scheduled meeting.

52

6.6. OPERATING REPORTS In addition to the applicable reporting requirements of Title 10 of the Code of Federal Regulations, the following reports shall be submitted to USNRC Document Control Desk.

a. Annual Operating Report: Routine operating reports covering the operation of the reactor during the previous calendar year shall be submitted by March 31 of each year. The Annual Operating Report shall provide a comprehensive summary of the operating experience having safety significance during the year, even though some repetition of previously reported information may be involved. References in the Annual Operating Report to previously submitted reports shall be clear.

Each Annual Operating Report shall include:

1. A brief narrative summary of:

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

b. Results of surveillance test and inspections;

2. A tabulation showing the energy generated by the reactor on a monthly basis, the cumulative total energy since initial criticality, and the number of pulses greater than $2.00;

3. List of the unscheduled shutdowns for which corrective action was required to ensure safe operation of the reactor, including the reasons and the corrective actions taken;

4. Discussion of the major safety-significant corrective and/or preventive maintenance performed during the period, including the effects (if any) on the safe operation of the reactor, and the reasons for the corrective maintenance required;

5. A brief description of:

a. Each change to the facility to the extent that it changes a description of the facility in the Safety Analysis Report;

b. Changes to the procedures as described in the Safety Analysis Report;

c. Any new experiments or tests performed during the reporting period that is not encompassed in the Safety Analysis Report;

6. A summary of the safety evaluation made for each change, test, or experiment not submitted for Commission approval pursuant to Section 53

50.59 of 10 CFR Part 50. The summary shall show the reason leading to the conclusion that the criteria in paragraph (c)(2) of that Section were not met and that no change to the Technical Specifications was required;

7. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as determined at or prior to the point of such release or discharge. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed, a statement to this effect is sufficient.

a. Liquid Waste (summarized on a monthly basis)

i. Radioactivity discharged during the reporting period:

Total radioadivity released (in curies);

Concentration limits used and isotopic composition for fission and activation products Total radioactivity of each nuclide released during the reporting period and, based on representative isotopic analysis, average concentration at point of release during the reporting period; IL Total volume of effluent water (including diluents) during periods of release;

b. Gaseous Waste (summarized on a quarterly basis)

Radioactivity discharged during the reporting period for:

Argon-41; Particulates with half-lives greater than eight days;

c. Solid Waste (summarized on a quarterly basis)

Total cubic feet and combined activity in curies of materials in solid form disposed of under license R-84;

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

9. A list of exposures greater than 25% of the allowed 10 CFR Part 20 limit received by reactor personnel or visitors to the reactor facility;

b. Other Reports: A report shall be submitted to the USNRC within 30 days describing:

54

1. Any permanent change of either the AFRRI Licensee or the Reactor Facility Director; or

2. Significant changes in the transient or accident analysis described in the SAR.

6.7. RECORDS 6.7.1 RECORDS THAT SHALL BE RETAINED FOR A PERIOD OF AT LEAST

' FNEYEARS

a. Normal reactor operations;

b. Principal maintenance operations;

c. Reportable occurrences;

d. Surveillance activities required by Technical Specifications;

e. Reactor facility radiation and contamination surveys;

f. Experiments performed with the reactor;

g. Changes to operating procedures;

h. Fuel inventories and fuel transfers;

i. Records of meetings of the RRFSS.

6.7.2. RECORDS TO BE RETAINED FOR AT LEAST ONE CERTIFICATION CYCLE Records of retraining and requalification of licensed reactor operators and senior reactor operators shall be retained at all times the individual is employed or until the license is renewed.

6.7.3. RECORDS THAT SHALL BE RETAINED FOR THE LIFE OF THE FACILITY

a. Gaseous and liquid radioactive effluents released to the environs;

b. Appropriate offsite environmental monitoring surveys;

c. Radiation exposures for all reactor personnel monitored; and

d. Drawings of the reactor facility.

55

e. Reviews and reports pertaining to a violation of the safety limit, limiting safety system setting (LSSS) or limiting condition of operation (LCO).

56

24. C. If three control rods can be withdrawn simultaneously, analyze this rod withdrawal accident and discuss how the analysis affects the results of the analysis of a ramp insertion accident.

The following information is offered in addition to the information submitted April 20, 2012.

This analysis is related to a rod withdrawal accident in which three standard control rods: shim ,

regulating , and safety, can be withdrawn simultaneously. There are 2 redundant, independent power and scram channels . If the first scram channel failed , the second channel would terminate the reactivity insertion at the same power level, 1.09 MW(t). For this analysis, the 3 second period RWP is also ignored. There is a delay from the initiation of a scram to the insertion of a control rod of no more than 0.5 seconds. This is the time necessary to close relay contacts, and bleed the magnetic field from the rod drive magnetic coupling of a standard control rod. The average insertion rate of the three standard control rods is 0.2285 ($/sec).

Starting at 1.0 MW(t) , the reactor would initiate a scram at 1.09 MW, and the three standard control rods are assumed to continue driving out for 0.5 seconds, resulting in an additional reactivity insertion of $0.11 , which would produce a positive period of 75 seconds.

At scram+ 0.5 seconds when the insertion would be terminated and the rods all scrammed ,

the peak power would be 1,097,290.94 watts. The maximum temperature would be 416.98 °C assuming that peak power is reached immediately without delay before scramming the reactor.

In Table 1, the second column represents the total rod worth of each standard control rod and the sum of the total rod worth of all three standard control rods combined. The third column represents the withdrawal time of each standard control rod and the average withdrawal time of three standard control rods . The fourth column represents the rod worth insertion rate of the standard control rods which is the quotient of rod worth divided by withdrawal time.

Standard Control Rod Rod Worth ($) Withdrawal Time (sec) Insertion Rate ($/sec)

Safety 2.65 39.4 0.0673 Shim 2.74 36.1 0.0759 Regu lating 3.01 34.8 0.0865 Safe/Shim/Reg 8.40 36.8 0.2285 Table 1. Rod worth , withdraw time and insertion rate of the standard control rods.

In Table 2, the second column represents the reactivity insertion 0.5 seconds after the reactor scrams from each standard control rod and the three standard control rods being withdrawn simultaneously. This is the product of the insertion rate from Table 1 multiplied by 0.5 seconds.

The third column represents the period resulting from the reactivity insertion in first column based on the In-Hour equation . The fourth column represents reactor power in watts 0.5 seconds following a scram starting at 1.09 MW, based on the period in the third column . The fifth column represents the fuel temperature in degrees Celsius corresponding to the reactor power in the fourth column , wh ich was calculated based on empirical data fitted to a fifth degree polynomial.

Standard Control Rod $ (O.S sec) Period (sec) P (0.5 sec) Watts T (0.5 sec) °C Safety 0 .0336 320 1,091, 704.46 415.48 Sh im 0 .0380 280 1,091,948.17 415.53 Regulating 0.0432 230 1,092,372.14 415 .60 Safe/Shim/Reg 0.1142 75 1,097,290.94 416.98 Table 2. Rod worth , period , power, and temperature of the standard control rods.

This analysis results in a negligible effect the on the results of the analysis of a ramp insertion accident. As a result of this analysis, the maximum power reached by withdrawing all three standard control rods simultaneously for 0.5 seconds following a scram that was initiated at 1.09 MW would result on a reactor power level of 1,097,290.94 watts or 1.097 MW which remains 2,709.06 watts below the AFRRI TRIGA licensed power limit of 1.1 MW. The fuel temperature calculated at this power level corresponds to 416.98 °C. This temperature is 583.02 °C below the safety limit of 1000 °C and 183.02 °C below limiting safety system setting of 600 °C.

The following additional information is offered in response to discussions with NRC staff.

The following additional analysis was requested .

Starting from 100 watts, what is the peak power, 1.0 second following a SCRAM initiated at 1.09 MW with a 3.0 second Period?

Standard Control Rod Period (sec) P (1.0 sec) Watts Safe/Shim/Reg 3 1,521,217.54 Table 3. Period and power of the standard control rods.

Having started from 100 watts , the peak power, 1.0 second following a SRCAM initiated at 1.09 MW with a 3.0 second Period would be 1,521 ,217.54 watts.

7 INSTRUMENTATION AND CONTROL SYSTEMS 7.1

SUMMARY

DESCRIPTION The reactor is operated from a Control System Console (CSC) located in the control room. The Data Acquisition Cabinet (DAC) is located in the reactor room along with cabinets that house the digital neutron linear/log power channel and the driver modules for the control rod stepping motors.

The operating mode of the reactor is determined by four mode selector switches on the console.

In Automatic and Steady-State modes, the reactor can operate at demand power levels up to 1.0 MW. In Square Wave mode, a step insertion of reactivity rapidly raises reactor power to a steady-state a demand power level up to 1.0 MW. In the Pulse mode, a large-step insertion of reactivity results in a short duration reactor power pulse.

The reactor control system is all solid-state circuitry with a mixture of analog and digital instrumentation. .

7.2 DESIGN OF INSTRUMENTATION AND CONTROL SYSTEMS Three independent power measuring channels provide for a continuous indication of power from the source level to peak power resulting from the maximum allowed pulse reactivity insertion.

Trips are provided for over power and loss of detector high voltage on the two analog safety channels. Fuel temperature is measured for display as well. Other parameters not used by the reactor protection system are also monitored and displayed.

The instrumentation and control system is designed to provide the following:

• complete information on the status of the reactor and reactor-related systems

• a means for manually withdrawing or inserting control rods

• automatic control of reactor power level

• automatic scrams in response to over power, loss of detector high voltage, or high fuel temperatures

• automatic scrams in response to a loss of operability of the digital computer system

• monitoring of radiation and airborne radioactivity levels 7.2.1 Design-Basis Requirements The primary design basis for the AFRRI Reactor is the safety limit on fuel temperature. To prevent exceeding the safety limit, design features, operating limitations, and automatic scrams are provided for over power conditions. Interlocks limit the magnitude of transient reactivity insertion.

7 .2.1.1 Reactor Power Measurements 7-1

Reactor power is measured by three separate detectors; a fission or ion chamber serving the operational channel and either ion chambers or fission chambers serving the safety channels.

The signal from the operational channel(s) provide wide range log power from 10-8 % to 100%

reactor power and period indication from -30 seconds to +3 seconds. One ion chamber or fission chamber is connected to the NP-1000 safety channel. A second ion chamber or fission chamber is used by the NPP-1000 percent power and pulsing channel. Both the NP-1000 and NPP-1000 provide indication of linear reactor power from 0 % to 120 % steady state reactor power and the NPP-1000 also provides indication of reactor power for pulsing operations. Figure 7-1 shows the relative ranges of the channels and the detectors.

The fission chamber for the operational channel(s) wide range instrument is connected to analog circuitry in a NEMA Preamplifier box mounted on the wall of the reactor room. This box contains a high voltage power supply, low voltage power supplies, preamplifier, campbelling module, counter/transmitter module and other circuitry. The high voltage power supply also monitors the high voltage to the fission chamber. If a loss of high voltage to the operational channel's fission chamber is sensed, an interlock is generated. The analog output from the preamplifier and campbelling module box described above is sent to the counter/transmitter which digitizes the signal and communicates this signal to a NEMA Microprocessor box that is also mounted on the wall of the reactor room. This box contains low voltage power supplies and microprocessor circuitry to convert the detector signal to useable digital values and transmit that signal to the DAC computer system. The Microprocessor box also contains circuitry for trip setpoints that provide interlock functions and indications on the console.

The operational channel(s) log display provides a continuous indication from 10-8 % to 100% of full power for the console display, analog bar graph display, and the console chart recorder.

The reactor period signal is generated by the microprocessor assembly of the operational channel(s). Reactor period is displayed on the console display and analog bar graph display. A bistable circuit provides a visual warning and rod withdrawal interlock when the period is less than a predetermined limit. The period\ signal is also used by the AUTO control system.

The NP-1000 safety channel provides a linear power signal to the console display and analog bar graph display. These displays are scaled at 0 to 120% of full power. A bistable circuit provides scram and alarm functions if the high power setpoint is exceeded. The detector input to the NP-1000 safety channel is disabled during pulse mode operations. A separate bistable circuit provides a scram signal to the reactor protection system upon a loss of detector high voltage.

7-2

2500 MW 1000MW 100 10 1.lMW

______ L_ __ _

100%

100 kW 10%

10 1%

NP-1000 NPP-1000

-1 I.AU ----------~----------------------- 1n  %

1 kW Interlock 100 -2 10  %

10 -3 w Operational 1n  %

channel 1W -4 10  %

0.1 Source Level 0.01 -6

---

~------------------------

1n  %

0.001 --------- -7

---

~------------------------- 1n  %

Source Interlock -8 1n o.1n Operational channel(s)- Fission Chamber NP-1000- Uncompensated Ion Chamber NPP-1000- Uncompensated Ion chamber Figure 7-1 AFRRI Power Instrument Ranges 7-3

The NPP-1000 power pulsing channel functions as a redundant NP-1000 safety channel during steady state operations. During pulsing operations, it displays peak power from a pulse on the scale of 0 to 3300 MW on the analog bar graph and a scale of 0 to 3300 MW on the console display. An analog bar graph display of integrated energy is also provided with a scale of 0 to 30 MW-s. A graphical display of a pulse is available on the console display, along with text information on the pulse number, pulse time and date, full-width at half-maximum power, peak power, integrated power, minimum period, and peak fuel temperature. These data are recorded and may be stored and recalled at a later date. The pulsing channel is enabled when the pulse mode switch is pressed, as long as all interlock conditions are met. The pulse data collection is performed by the DAC computer and begins when the pulse rod "Fire" button is depressed. This also enables the peak hold circuit and starts a one minute timer. The peak power and energy displays are reset at the end of the one minute period. The peak power is also recorded on the console data recorder.

The NPP-1000 channel contains bistable circuits that will produce a scram in steady state mode and alarm output for the conditions of the high power setpoint being exceeded and for loss of high voltage.

7.2.1.2 Temperature Measurements As illustrated in Figure 7-2, fuel temperature is measured by a thermocouple embedded in an instrumented fuel element. There are two fuel temperature channels in the reactor instrumentation system, and therefore two thermocouples may be connected at one time. Fuel temperature is displayed on the console display and console analog bar graphs. A high temperature scram is sent to the reactor protective system for high power pulsing operations.

Temperature of the bulk pool water is measured by a resistance temperature detector (RTD).

The RTD is mounted to the top of the reactor tank and the probe extends about 18 inches (45.72 cm) below the top of the tank. It sends a signal to the console that displays as the pool water temperature. A temperature alarm circuit on the pool water channel will annunciate an audible and visual alarm on the console if the water temperature exceeds a preset temperature. Two additional RTDs are located in the primary piping, one on the inlet to the heat exchanger and one on the outlet of the heat exchanger. The temperature signals from these detectors are sent to the console for display as the pool water outlet temperature and the pool water inlet temperature.

These primary piping RTDs may not display accurate temperatures for the primary cooling water if the primary pump is not operating.

7-4

ST1\INLESS STEa LE*"'D*OlJT TUBE --~

Figure 7-2 Instrumented Fuel Element 7-5

7.3 REACTOR CONTROL SYSTEM 7.3.l Control Rod Drives The four control rods are positioned by control rod drives mounted on the reactor top center channel.

As illustrated in Figure 7-3, the 3 standard control rod drives (Reg, Shim and Safe) rod drives are rack-and-pinion linear actuators. The regulating rod drive uses a stepper motor that is able to operate at variable speeds when operated by the servo system. The regulating rod drive operates at its maximum speed when controlled in the manual mode by the reactor operator. The shim and safe rod drives use the same type stepping motor as the regulating rod drive but operate at a single speed. An electromagnet is secured to the bottom of the draw tube to which the rack is mounted. The magnet is moved up or down in response to rotation of the pinion shaft. The control rod is attached to the armature by a long connecting rod. When the magnet is energized, the armature is magnetically coupled to the draw tube. De-energizing the magnet causes the rod to drop. A dash pot is incorporated into the armature section to decelerate the rod near the bottom following a scram. Limit switches sense when the magnet is fully withdrawn, the magnet is fully down, and the armature (and thereby the rod) is fully down. A ten-turn potentiometer is coupled to the pinion shaft to provide for rod position indication. The pinion shafts are chain-and-sprocket coupled to a DC stepper motor.

The transient rod (also called the pulse rod) is operated by a pneumatic/electric drive. A connecting rod couples the transient rod to a piston rod assembly. As illustrated in Figure 7-4, the piston resides within an externally threaded cylinder. A ball screw nut acts on these external threads to raise or lower the cylinder. Rotation of the ball screw nut is accomplished by a worm gear coupled to a motor. A potentiometer is gear-driven by the worm gear shaft to provide rod position indication. A hydraulic shock absorber is incorporated into the top of the cylinder. Air from a compressor is connected to a normally-closed port of a three-way air solenoid valve. The common port is connected to the transient control rod drive cylinder below the piston. The normally-open port is vented. When the air solenoid valve is energized, air pressure is placed on the bottom of the piston causing the piston to be brought in contact with the shock absorber. The resulting reactivity insertion is dependent on the position of the cylinder prior to applying air.

With air applied, energizing the motor in the up or down direction will cause the cylinder, piston, and control rod to move up or down as a unit. Scram of the transient rod is accomplished by de-energizing the air solenoid valve. This vents the air pressure under the piston and results in the control rod dropping. As illustrated in Figure 7-5, limit switches provide for sensing cylinder up, cylinder down, and rod down. A bracket extends over the top of the cylinder. A switch on the bracket opens a contact in the up circuitry when the shock absorber assembly contacts it. The bracket itself is substantial enough to stall the motor should the switch contact fail to open.

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O-RI *G Figure 7-3 Standard Control Rod Drive and Limit Switches 7-7

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ro~T R Figure 7-4 Transient Rod Drive 7-8

rf - CV.'ER .. FCR il.#J'I' HOUtU lJlRJl u S\..fl0-1 Figure 7-5 Transient Rod Drive Limit Switches 7-9

7 .3.2 Servo System In the Automatic and Square-Wave modes of operation, the regulating rod is controlled by the servo system to control reactor power based on input signals from a power channel, the reactor period signal, and the power demand control.

In Automatic mode, the reactor power is compared against the power demand setting to obtain power error. The period signal is monitored by the controller to limit the reactor period to a minimum of +8 seconds when power is being increased. To reduce hunting of the regulating rod, a deadband is incorporated in the system. The power error signal is used by the DAC computer to determine which direction (if any) the regulating rod needs to move to correct the power error. The regulating rod speed is ".ariable and it will move slowly for small errors and it will move fast for large errors. The regulating rod speed cannot exceed the travel speed that is used in manual control. The variable speed ability of the servo system reduces power overshoot during transients.

To perform a Square Wave, the reactor must be configured in Steady-State mode. First, the reactor power is raised to some nominal low power (less than 1000 W) with the air to the transient rod off. Second, the transient rod cylinder is raised to the position corresponding to the*

desired reactivity insertion. Finally, the square wave mode switch is depressed to change the console mode from Steady-State to Square Wave and the transient rod fire button pressed.

Reactor power will increase to the desired power level and then switch to the Automatic mode to maintain a constant power level.

7 .3.3 Interlocks The following are the interlocks utilized by the reactor console:

• Th.e 1-kW permissive interlock to prevent pulsing when wide range log power is above 1 kW

• Interlock to prevent the shim, safety and regulating rods from being withdrawn in pulse mode

• Interlock to ensure that only one control rod can be manually withdrawn at a time in the steady state mode

• Rod withdrawal prevent (RWP) interlock, activated by a low count rate on the operational channel(s)when the log power is not greater than 10-7 % power. An indication is provided on the console low resolution monitor to indicate when a source level rod withdrawal interlock is present

• Interlock to prevent the application of air to tpe transient rod drive mechanism in the steady state mode unless the drive cylinder is fully inserted;

• Interlock to ensure that only one control rod can be manually withdrawn at a time in the square wave mode, excluding the transient rod.

7.4 REACTOR PROTECTION SYSTEM 7-10

The scram circuits function to shut down the reactor by dropping all four control rods to their fully inserted positions. Scram is accomplished by de-energizing the magnets for the safety, shim, regulating rods and by de-energizing the air solenoid valve for the transient rod.

A reactor scram will result under any of the following conditions:

• Operator-initiated manual scram

• Fuel element temperature in excess of the setpoint

• Safety channels measuring power in excess of the setpoint

• Loss of high voltage to the safety channels

• Drop in pool level below setpoint

• Activation of emergency stop circuit

• Pulse timer initiated scram

• DAC-to-CSC Watchdog failure scram 7.5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS There are no engineered safety feature actuation systems.

7.6 CONTROL CONSOLE AND DISPLAY INSTRUMENTS 7.6.1 High Resolution Monitor Display I

• I The high resolution graphic monitor displays the following reactor parameters:

• Reactor wide range linear power from operational channel(s)

• Reactor wide range log power from operational channel(s)

• Reactor period from operational channel( s)

• Reactor linear power from NP-1000 and NPP-1000

• Reactor fuel temperatures

• Reactor pool temperature

• Rod position and reactor graphic 7 .6.2 Reactor Status Display The reactor status display monitors parameters in text format for reactor power channels, temperatures, interlock status, etc. The items displayed are determined by the facility management. This display also can be changed to the SCRAM and WARNING window display by depressing the computer keyboard SPACE bar, and will display any items in a SCRAM or WARNING trip mode.

7.6.3

• Hardwired Analog Bargraph Displays Analog bargraphs on the control console are hardwired separately from the computer system. In the event of a computer malfunction, this allows observance of reactor conditions. These bargraphs include indications of:

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• NP-1000 linear power level

• NPP-1000 linear power level

• operational channel(s)log power level

• operational channel(s)period

• Fuel Temperature channel 1

• Fuel Temperature channel 2

• NVT for pulsing operations

• NV peak power for pulsing operations 7.6.4 Reactor Mode Control Panel The Mode Control Panel has switches the operator uses to select various modes of operation, test functions, automatic mode demand power level, and main power on/off.

7.6.5 Console Chart Recorder The console chart recorder records and displays wide-range log power.

7.6.6 Annunciator Function When an alarm or warning is received at the control console, an audible signal will sound and a message will be displayed on the monitor. When the operator presses the acknowledge button, the audible signal will be silenced and the message display will remain until the alarm or warning has cleared. If the alarm or warning condition has cleared, the message will clear.

7.7 RADIATION MONITORING SYSTEMS The radiation monitoring systems associated with reactor operations at AFRRI are maintained as a means of ensuring compliance with radiation limits established under 10 CFR 20. These systems consist of remote area monitors, continuous air monitors, reactor stack monitors, and AFRRI perimeter monitoring. Detailed information (such as alarm setpoints for the various monitors, appropriate reactor operator responses to radiation alarms, and procedures involving monitor data evaluation and archiving) can be found in References 7-1 and 7-2.

The radiation monitoring systems associated with AFRRI reactor operations provide readouts and radiation alarms at key locations in the AFRRI complex. These locations are:

• Reactor Room (Room 3161)

• Reactor Control Room (Room 3160)

• • Emergency Response Center

• Annunciator panel in Hallway 3101 The radiation alarms in the reactor room and the radiation alarm readouts in the reactor control room provide the reactor operators with information necessary for the safe operation of the AFRRI-TRIGA reactor. The audible and visual alarms on the annunciator panel in Hallway 7-12

3101 alert the Security Watchman (during nonduty hours) of unusual reactor conditions when the reactor is secured. When reactor personnel are present in the reactor administration/control area, the audible alarm on the annunciator panel in Hallway 3101 is turned off.

7.7.1 Remote Area Monitors The remote area monitors (RAMs) in the remote area monitoring system of primary concern to the reactor are R-1, R-2, E-3, and E-6. These units are placed in various areas of the reactor building where potential radiation hazards may exist due to reactor operation.

The monitors utilize scintillation detectors which measure gamma radiation with energies greater than 20 keV. The units have a range of 1 mrem/hr to 105 mrem/hr and a nominal accuracy of

+/-15 percent at all levels. The units have a time constant of 2 seconds and a meter and alarm response time of less than 1 second. The monitors activate radiation alarms at various locations within AFRRI; the alarm set points are variable. The monitors also activate visual alarms in the control room and the Emergency Response Center (room 3430).

The RAMs are calibrated at regular intervals using a radiation source of known intensity. The locations of the RAMs, the readouts, and the audible and visual radiation alarms are given in Table 7Table 7 1 and Figures 7-6 through 7-8. The alarm setpoints can be found in AFRRI internal documents (References 7-1 and 7-2).

7. 7.2 Continuous Air Monitors The continuous air monitors (CAMs) of primary importance to the reactor are two CAMs located in the reactor room. Three additional CAMs, which monitor the exposure rooms and the prep area, are discussed in Section 10. *The CAMs provide continuous air sampling and monitoring (gross beta-gamma activity) primarily of airborne particulate matter.

The CAMs draw air with an air pump (-7 cfm) through a shielded filter assembly, which traps any particulate matter greater than 0.3 microns in diameter. A G-M detector measures any radioactive particulates trapped by the filter. The count rate (counts per minute) is recorded by a three-cycle, logarithmic, strip-chart recorder mounted on the CAM itself. The units have a sensitivity range of 50 cpm to 50 x 103 cpm and a nominal accuracy of+/- 10 percent. The units have a time constant which is inversely proportional to the count rate, being 200 seconds at 50 7-13

cpm and 1 second at 50,000 cpm. The units have the capability of actuating alarms at two adjustable radiation levels.

Table 7-1 Reactor Remote Area Monitors RAM Location Readout Radiation alarm R-1 Approximately 7 Meter in reactor Activates audible and visual alarm in feet above the floor control room the reactor room and in the reactor on the reactor room and Emergency control room; activates visual alarm in east wall Response the Emergency Response Center; Center activates visual and optional audible alarm on annunciator panel in Hallway 3101 R-2 Approximately 7 Same as R-1 Activates visual alarm in the reactor feet above the floor control room and in the Emergency on the reactor room Response Center west wall E-3 6 feet above the Same as R-1 Same as R-2. In addition, there are a floor on the west visual and audible local alarm in the wall prep area prep area near ER #1, and a red light opposite ER #1 plug at the front desk.

door E-6 6 feet above the Same as R-1 Same as E-3, except the visual and floor on the west audible local alarm is in the prep area wall prep area nearER#2 opposite ER #2 plug door The primary reactor room CAM is located in the southwest corner of the reactor r9om and is visible from Room 3156. The air sampled by this CAM is taken from approximately 36 inches above the reactor pool surface inside the core support structure. The air is passed through a hose to the CAM. The air is exhausted by the CAM back to the reactor room. The reactor room CAMs form an integral part of the reactor room containment capability, in that when either CAM's high-level alarm is activated, the supply and exhaust dampers to the reactor room in the ventilation system are automatically closed to isolate the reactor room air volume.

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The backup reactor room CAM is located along the west wall of the reactor room and its alarms are visible from the control room and Room 3158. The air sampled by this CAM is taken from a point near the warm drain located along the west side of the reactor pool. The air is exhausted by the backup CAM back to the reactor room.

A description of the CAMs' alarms, locations and read-out is given in Table 7-2 and Figures 7-6 through 7-8. The alarm setpoints can be found in the appropriate AFRRI internal documents (Reference 7-2). Additionally a flashing visual light on the reactor auxiliary instrumentation console in the reactor control room will be illuminated when either reactor room CAM is set in

• the TEST mode during testing.

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....

....

FIRST LEVEL e LOCATION OF SENSOR

• LOCATION OF READOUTS OR ALARMS Figure 7-6 AFRRl Radiation Monitors Associated with AFRRI-TRIGA Reactor First Level 7-16

.....

EQUJP ROOM SECOND LEVEL NO REACTOR MONITORS ON SECOND LEVEL

'""

RHS8 Figure 7-7 AFRRI Radiation Monitors Associated with AFRRI Reactor Second Level 7-17

EMERGENCY RESPONSE CENTER THIRD LEVEL e

D

• LOCATION OF SENSOR LOCATION OF READOUTS OR ALARMS RHS9 Figure 7-8 AFRRI Radiation Monitors Associated with AFRRI Reactor Third Level 7-18

Table 7-2 Reactor Room Continuous Area Monitors CAM Location of Readouts High-level alarm Low-level air intake Alarm< 1l Approximately 36 Meter in reactor Activates audible and Activates inches above reactor control room visual alarm on unit visual alarm on Primary pool inside core itself the unit itself carriage Strip chart Activates audible and recorder located visual alarm on on the unit itself reactor control room annunciator panel Activates visual alarm on reactor room wall panel

• Activates audible and visual alarm on annunciator panel in Hallway 3101 Causes the reactor room ventilation dampers to close Near the warm drain Strip chart Identical to primary Activates along the west side recorder located CAM alarm visual alarm on Alternate of the reactor tank on the unit itself indications, when the unit itself Backup and meter in connected reactor control room (ll If the low-level alarm is being used 7.7.3 Stack Monitoring Systems The stack monitoring systems consist of the stack flow monitor and the stack gas monitor. These systems provide data about the radioactive effluents discharged through the reactor stack. The stack flow monitor measurements are recorded by a strip chart recorder. Stack gas monitor measurements of Ar-41 emissions are recorded on a strip chart recorder and can be viewed at the end of each day by an operator to verify that no unusual Ar-41 releases have occurred.

7.7.3.1 Stack Flow Monitoring System The stack flow monitoring system measures the average flow rate of air exhausted through the reactor stack. The system consists of a pair of pitot tubes and Magnehelic pressure gauges which mechanically measure the dynamic pressure in the stack and produce a proportional electrical signal. A strip chart recorder located in the reactor control room records the stack flow. There 7-19

are no level alarms associated with this system, except when exhaust fan EF5 fails, in which case a visual alarm is activated in the reactor control room.

7.7 .3.2 Stack Gas Monitoring System The stack gas monitor (SGM) system is a Nal scintillation detection system which samples exhaust air from the reactor stack. The air is passed through a filter to remove particulates before being analyzed. This system will detect those effluents which have been released into the reactor stack, and are set to alarm at the limit currently specified in the AFRRI Reactor Emergency Plan.

The stack gas monitor system is capable of activating alarms at two levels. Additionally, a flashing visual light on the reactor auxiliary instrumentation console in the reactor control room will be illuminated when the stack gas monitoring system pump motor is turned off. The locations of the system readouts and alarms are listed in Table 7-3. The setpoints for the radiation alarms can be found in the appropriate AFRRI internal documents.

Table 7-3 Stack Monitoring Systems System Readout Radiation Alarm Stack Flow Strip chart recorder in (Not applicable) However, Monitoring System reactor control room EF5 failure gives a visual alarm in reactor control room Stack Gas Meter in reactor control Activates audible and visual Monitoring System room alarm in reactor control room 7.7.4 Perimeter Monitoring An environmental monitoring program is conducted by AFRRI primarily to measure environmental doses received from radionuclides produced by the AFRRl-TRIGA reactor, particularly Ar-41. The environmental monitoring program shall consist of an NRC/EPA approved reporting method.

7.8 REFERENCES

7-1. Armed Forces Radiobiology Research Institute, Health Physics Procedures (HPPS),

Safety and Health Department.

7-2. Armed Forces Radiobiology Research Institute, Reactor Operational and Administrative Procedures, Radiation Sciences Department, Reactor Division.

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