ML20092G539

Tech Specs for Armed Forces Radiobiology Research Inst Reactor Facility

ML20092G539
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	06/30/1984
From:	DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY
To:
Shared Package
ML20092G520	List: ML20092G514 ML20092G539
References
NUDOCS 8406250149
Download: ML20092G539 (46)

Text

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Technical Specifications for the AFRRI Reactor Facility i

i Docket 50-170 l License R-84 j i

June 1984 i

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DEFENSE NUCLEAR AGENCY i ARMED FORCES RADI0 BIOLOGY RESEARCH lNSTITUTE BETHESDA, MARYLAND 20814 l APPROVED FOR PUBLIC RELEASE; OlSTRIBUTION UNLIMITED h$R DOC 000 0 p PDR l

TECHNICAL SPECIFICATIONS FOR THE AFRRI REACTOR FACILITY DOCKET 50-170 LICENSE R-84 JUNE 1984 4

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Preface Included in this document are the Technical Specifications and the " Bases" for the Technical Specifiestions. 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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4 TECHNICAL SPECT!ICATIONS FOR THE AFRRI REACTOR FACILITY LICENF2 NO. R-84 ,

DOCF ET #50-170 ,

TABLE OF CONTENTS 1.0 DEFINITIONS 1.1 ALARA 1 1.2 Channel Calibration 1 1.3 Channel Check 1 1.4 Channel Test 1 1.5 Cold Critical 1 1.6 Core Grid Position 1 1.7 Experiment 1 1.8 Experimental Facilities 1 1.9 Fuel Element 2 1.10 Instrumented Element 2 1.11 Limiting Safety System Setting 2 1.12 Measured Value 2 1.13 Measuring Channel 2 1.14 On Call 2 1.15 Operable 2 1.16 Pulse Mode 3 1.17 Reactor Operation 3 1.18 Reactor Safety Systems 3 1.19 Reactor Secured 3 1.20 Reactor Shutdown 3 1.21 Reportable Occurrence 3 1.22 Safety Channel 4 1.23 Safety Limit 4 1.24 Shutdown Margin 4 1.25 Standard Control Pad 4 1.26 Steady State Mode

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1.27 Transient Rod 4 2.0 SAFETY LIMITS AND LIMITINO SAFETY SYSTEM SETTINGS 5 2.1 Safety Limit - Fuel Element Temperature 5 2.2 Limiting Safety System Setting for Fuel Temperature 5 3.0 LIMITING CONDITIONS FOR OPERATIONS 7 3.1 Reactor Core Parameters 7 T

l 3.1.1 Steady State Operation 7 3.1.2 Pulse Mode Operation 7 3.1.3 Reactivity Limitations 8 3.1.4 Scram Time 8 l 1 l

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page 3.2 Reactor Control and Safety Systems 9 3.2.1 Reactor Control System 9 3.2.2 Reactor Safety Systems 10 3.2.3 Facility Interlock System 11 ,

3.3 Coolant Systems 12 3.4 Ventilation System 13 3.5 Radiation-Monitoring System and Effluents 13 3.5.1 Monitoring System 13

3.5.2 Effluents

Argon-41 Discharge Limit 15 3.6 Limitations on Experiments 15 3.7 System Modifications 17 3.8 ALARA 18 4.0 SURVEILLANCE REQUIREMENTS 19 4.1 Reactor Core Parameters 19 4.2 Reactor Control and Safety System 20 4.2.1 Reactor Control Systems 20 4.2.2 Reactor Safety Systems 20 4.2.3 Fuel Temperature 21 4.2.4 Facility Interlock System 21 4.2.5 Reactor Fuel Elements 22 4.3 Coolant Systems 23 4.4 Ventilation Sys, tem 23 4.5 Radiation-Monitoring System 24 5.0 DESIGN FEATURES 25 5.1 Site and Facility Description 25 5.2 Reactor Core and Fuel 25 .'

5.2.1 Reactor Fuel 25

• 5.2.2 Reactor Core 26 5.2.3 Control Rods 27 5.3 Fissionable Material Storage 28 11

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fag 6.0 ADMINISTRATIVE CONTROLS 29 6.1 Organization 29 6.1.1 Structure 29 6.1.2 Responsibility 30 6.1.3 Staffing 30 6.1.3.1 Selection of Personnel 30 6.1.3.2 Operations 30 6.1.4 Training of Personnel 31 6.2 Review and Audit - The Reactor and Radiation Facility Safety Committee (RRFSC) 31 6.2.1 Composition and Qualifications 31 6.2.1.1 Composition 31 6.2.1.2 Qualifications 32 6.2.2 Function and Authority 32 6.2.2.1 Function 32 6.2.2.2 Authority 32 6.2.3 Charter and Rules 32 6.2.3.1 Alternates 32 6.2.3.2 Meeting Frequency 32 6.2.3.3 Quorum 32 6.2.3.4 Voting Rules 32 6.2.3.5 Minutes 32 6.2.4 Review Function 33 6.2.5 Audit Function 33 6.3 Procedures 34 6.4 Review and Approval of Experiments 34 6.5 Required Actions 35 6.5.1 Actions To Be Taken in Case of Safety Limit Violation 35 6.5.2 Reportable Occurrences 35 111 i

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6.6 Reports 36  :

6.6.1 Operating Reports 36 L 6.7 Records . 38 ,

6.7.1 Records To Be Retained For A Period of At Least 5 Years or '

As Required by 10 CFR Regulations 38 ,l 6.7.2 Records To Be Retained For At Least One Complete Training  ;

Cycle 39 1 6.7.3 Records To Be Retained For The Life of The Facility 39 i IV r.

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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 environ-

- ment as low as reasonably achievable.

1.2 CHANNEL CALIBRATION A channel calibration consists of using a known signal to verify or adjust a channel to produce an output that corresponds with acceptable accuracy to known values of the parameter that the channel measures. Calibration shall encompass the entire channel including equipment activation, alarm, or trip, and shall be deemed to include a channel test.

1.3 CH ANNEL CHECK A channel check is a verification of acceptable performance by observation of channel behavior.

1.4 CH ANNEL TEST A channel test is the introduction of a signal into the channel to verify that it is operable.

1.5 COLD CRITICAL The reactor is in a cold critical condition when it is critical at a power level less than 100 watts, with the fuel and bulk water temperature equal and less than 400C.

1.6 CORE GRID POSITION The core grid position refers to the location of a fuel or control element in the grid structure.

1.7 EXPERIMENT Experiment shall mean (a) any apparatus, device, or material that is not a normal part of the core or experimental facilities, but that is inserted in these facilities or is in line with a beam of radiation originating from the reactor core; or (b) any operation designed to measure nonroutine reactor parameters or characteristics.

1.8 EXPERIMENTAL FACILITIES

- The experimental or exposure facilities associated with the AFRRI TRIGA reactor shall be

- a. Exposure Room #1

b. Exposure Room #2 NOTE: Exposure facilities protective barriers shall be differentiated from the primary protective barrier (fuel element cladding) for purposes of placement of experiments within these barriers.

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c. Reactor Pool

d. Core Experiment Tube

e. Portable Beam Tubes

f. Pneumatic Transfer System

g. Incore Locations 1.9 FUEL ELEMENT -

A fuel element is a single TRIG A fuel rod.

1.10 INSTRUMENTED ELEMENT An instrumented element is a special fuel element in which sheathed chromal/alumel or equivalent thermocouples are embedded in the fuel.

1.11 LIMITING SAFETY SYSTEM SETTING Limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions.

1.12 MEASURED VALUE A measured value is the magnitude of a variable as it appears on the output of a measuring channel.

1.13 MEASURING CHANNEL A measuring channel is that combination of sensor, interconnecting cables or lines, amplifiers, and output device that are connected for the purpose of measuring the value of a variable.

1.14 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 his/her whereabouts and telephone number; and

c. The individual is capable of getting to the reactor facility within 30 minutes under ,

normal circumstances.

1.15 OPERABLE -

A system channel, device, or component shall be considered operable when it is capable of performing its intended function (s)in a normal manner.

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1.16 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, utilizing the appropriate scrams in Table 2 and the appropriate interlocks in Table 3. 'Ihe reactor may be pulsed from a critical or suberitical state.

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1.17 REACTOR OPERATION Reactor operation is any condition wherein the reactor is not shut down, or any core maintenance is being performed, or there is movement of any control rod.

1.18 REACTOR SAFETY SYSTEMS Reactor safety systems are those systems, including their associated input circuits, that are designed to initiate a reactor scram for the primary purpose of protecting the reactor or to provide information that may require manual protective action to be initiated.

1.19 REACTOR SECURED The reactor is secured when all the following conditions are satisfied

a. The reactor Is shut down.

b. The console key switch is in the "off" position, and the key is removed from the console and is under the control of a licensed operator, or is stored in a locked storage area.

c. No work is in progress involving in-core fuel handling or refueling operations, maintenance of the reactor or its control mechanisms, or insertion or withdrawal of in-core experiments, unless sufficient fuel is removed to insure a $0.50 (or greater) shutdown margin with the most reactive control rod removed.

1.20 REACTOR SHUTDOWN The reactor is shut down when the reactor is suberitical by at least $0.50 of reactivity.

1.21 REPORTABLE OCCURRENCE A reportable occurrence is any of the following that occurs during reactor operation:

a. Operation with any safety system setting less conservative than specified in Section 2.2, Limiting Safety System Settings.

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b. Operation in violation of any Limiting Condition for Operation, Section 3.

c. Malfunction of a required reactor or experiment safety system component that could render the system incapable of performing its intended safety function unless the malfunction is discovered during tests.

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

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e. An observed inadequacy in the implementation of either administrative or pro-cedural controls, so that the inadequacy 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 (SAR).

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. An unplanned or uncontrolled release of radioactivity that exceeds or could have exceeded the limits allowed by Title 10, Part 20 of the Code of Federal Regulations -

(10CFR20), or these technical specifications.

1.22 SAFETY CHANNEL A safety channel is a measuring channel in the reactor safety system that provides a reactor protective function. .

l 1.23 SAFETY LIMIT Safety limits are limits on important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. ,

1.24 SHUTDOWN MARGIN Shutdown margin shall mean the minimum shutdown reactivity considered necessary to "

provide confidence that the reactor can be made suberitical by means of the control and safety systems, starting from any permissible operating conditions, and that the reactor will remain suberitical without further operator action.

1.25 STANDARD CONTROL ROD A standard control rod is a control rod having an electro-mechanical drive and scram capabilities. It is withdrawn by an electromagnet / armature system.

1.26 STEADY STATE MODE Operation in the steady state mode shall mean steady state operation of the reactor either by manualoperation of the control rods or by automatic operation of one or more control rod (servocontrol) at power levels not exceeding 1.1 megawatts, utilizing the appropriate scrams in Table 2 and the appropriate interlocks in Table 3.

1.27 TRANSIENT ROD ,

The transient rod is a control rod with scram capabilities that can be rapidly ejected i from the reactor core to produce a pulse it is activated by applying compressed air to a l piston.

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT: FUEL ELEMENT TEMPERATURE Applicability

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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 will result.

Specification The maximum temperature in a standard TRIGA fuel element shall not exceed 10000C under any condition of operation.

Basis The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well suited as a single specification, especially since it can be measured. A loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The 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 standard TRIGA fuel is based on data that includes the large mass of experimental evidence obtained during high-performance reactor tests on this fuel. These data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress, provided that the temperature of the fuel does not exceed 10000C while immersed in water.

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

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

Specification There shall be two fuel temperature safety channels. 'The limiting safety system setting for these instrumented fuel elements' temperature shall not exceed 6000C. One channel shall utilize an instrumented element in the "B" ring, and the second channel shall utilize an instrumented element in the "C" ring.

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' The limiting safety system setting is a temperature which, if exceeded, shall cause a

. reactor serain to be initiated, preventing the safety limit from being exceeded. A r

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if' ' setting of 6000C provides a safety margin of at least 4000C for standard TRIGA stainless-ste& clad 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

. > ht location of the thermocouple. If the thermocouple element is located in the hottest

• 1 position in the core, the difference between the true and measured temperatures will be only a few degrees, if the thermocouple element is located in a region of lower U.

temperature, the measured temperature will differ by a greater amount from that *

, c 6 g actually occurring at the core hot spot. To lessen this difference, the requirement is to locate the element in the hottest region of the core. These margins are sufficient to account for the remaining suheertainty' in the accuracy of the fuel temperature measurement channel arid: any ' overshoot in reactor power resulting from a reactor p transient during steady state mode operation.

5 i ; ;j i t., In the pulse mode of operation, the same limiting safety system setting shall apply.

(a  ! However, the temperature channel will have no effect on limiting the peak power j ,i ? ,("' generatLd, because of its relatively long time constant (seconds), compared with the i- , width of the pulse (milliseconds). In this mode, however, the temperature trip will act to reduce the amount of energy generated in the entire pulse transient, by cutting the " tail" of the power transient if the pulse rod remains stuck in the fully withdrawn position with enough reactivity to exceed the temperature-limiting safety system setting.

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3.0 LIMITING CONDITIONS FOR OPER ATIONS 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 i To assure that the reactor safety limit (fuel temperature) is not exceeded, and to provide for a set point for the high flux limiting safety systems, so that automatic protective action will prevent the safety limit from being reached during steady state operations.

Specifications The reactor steady state power level shall not exceed 1.1 megawatts. The normal staady state operating power limit of the reactor shall be 1.0 megawatt.

For purposes of testing and calibration, the reactor may be operated at power levels not to exceed 1.1 megawatts during the testing period.

Basis Thermal and hydraulic calculations and operational experience indicate that TRIGA fuel may be safely operated up to power levels of at least 1.5 megawatts with natural convective cooling.

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 assure that the fuel temperature safety limit will not be exceeded.

Specification

- The maximum step insertion of reactivity shall be 2.8% A k/k ($4.00) in the pulse mode.

. Basis Based upon the Fuchs-Nordheim mathematical model(cited by C. E. Clifford et al. In the April 1961 GA Report #2119, model of the AFRRI-TRIGA reactor), an insertion of 2.8% ok/k results in a maximum average fuel temperature of less than 5500C, thereby staying within the limiting safety settings that protect the 7

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safety limit. The 500C margin to the Limiting Safety System Setting and the 4500C margin to the safety limit amply allow for uncertainties due to extrapolation of measured data, accuracy of measured data, and location of instrumentated fuel elements in the core.

3.1.3 REACTIVITY LIMITATIONS Applicability .

These specifications apply to the reactivity condition of the reactor and the reactivity worths of controls 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 will not be exceeded.

Specifications

a. The reactor shall not be operated with the maximum available excess reactivity above cold critical with or without all experiments in place greater than $5.00 (3.5%Ak/k).

b. The minimum shutdown margin provided by the remaining control rods with the most reactive control rod fully wi'.hdrawn or removed shall be $0.50 (0.35%4k/k) for any conditions of operations.

Basis

a. The limit on available excess reactivity establishes the maximum power if all control elements are removed.

b. The shutdown margin assures that the reactor can be shut down from any operating condition even if the highest worth control rod remains in the fully withdrawn position or is completely removed.

3.1.4 SCRAM TIME Applicability The specification applies to the time required to fully insert any control rod to a full down position from a full up position.

Objective The objective is to achieve rapid shutdown of the reactor to prevent fuel damage.

Specification -

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

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Basis This specification assures that the reactor will be promptly shut down when a scram signal is initiated. Experience and analysis indicate that, for the range of transients for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.

. 3.2 REACTOR CONTROL AND SAFETY SYSTEMS l 3.2.1 REACTOR CONTROL SYSTEM  ;

Applicability This specification applies to the channels monitoring the reactor core, which  ;

must provide information to the reactor operator during reactor operation.

t Objective The objective is to require that sufficient information be available to the operator to assure safe operation of the reactor.  ;

Specification The reactor shall not be operated unless the measuring channels listed in Table 1 are operable.

TABLE 1. MEASURING CHANNELS ,

l Minimum Number Operable in Effective Mode Steady State Pulse Fuel Temperature Safety Channel 2 2 '

Linear Power Channel 1 1 Log Power Channel 1 0 High-Flux Safety Channel 2 1*

i Pulse Energy Integrating Channel 0 1*

(* NOTE: Same channel as linear power in this mode)

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. Fuel temperature displayed at the control console gives continuous information on this parameter, which has a specified safety limit. The power level channels assure that radiations indicating reactor core parameters are adequately monitored for both steady state and pulsing modes of operation. The specifica- i 9 r t.

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tions on reactor power level indication are included in this Section, since the power level is related to the fuel temperature.

3.2.2 REACTOR SAFETY SYSTEM Applicability This specification applies to the reactor safety system.

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

Specification The reactor shall not be operated unless the safety systems describcd in Tables 2 and 3 are operable.

TABLE 2. MINIMUM REACTOR SAFETY SYSTEM SCRAMS Minimum Number in Mode Maximum Channel Set Point Steady State Pulse Fuel Temperature 6000C 2 2 Percent Power, High Flux 1.1 Mw 2 0 Console Manual Scram Bar Closure switches 1 1 i liigh Voltage toss to Safety Channels 20% loss 2 1 Pulse Time 15 seconos 0 1 Emergency Stop (1 each exposure room, 1 on console) Closure switch 1 1 t

Pool Water Level 14 feet from top of core 1 1 Basis .

The fuel temperature and power level scrams provide protection to assure that the reactor can be shut down before the safety limit on the fuel element temperature will be exceeded. The manual scram allows the operator to shut down the system at any time if an unsafe or abnormal condition occurs. In the 10 s r. -

event of failure of the power supply for the safety chambers, operation of the reactor without adequate instrumentation is prevented. De preset timer insures that the reactor power level will reduce to a low level after pulsing.

De emergency stop allows personnel trapped in a potentially hazardous l exposure room or the reactor operator to stop actions through the interlock system. The pool water level insures that a loss of biological shielding would

- result in a reactor shutdown.

TABLE 3. MINIMUM REACTOR SAFETY SYSTEM INTERLOCKS Effective Mode Action Prevented Steady State Pulse ,

Pulse initiation at power levels greater than 1 kilowatt X Withdrawal of any control rod except transient X Any rod withdrawal with count rate in operational channel below 0.5 cps X X Simultaneous manual withdrawal of two standard rods X Basis The Interlock preventing the initiation of a pulse at a critical level above 1 kilowatt assures that the pulse magnitude will not allow the fuel element temperature to approach the safety limit. The interlock that prevents movement of standard control rods in pulse mode will prevent the inadvertent placing of the reactor on a positive period while in pulse mode. Requiring a count rate to be seen by the operational channels insures sufficient source neutrons to bring the reactor critical under controlled conditions. 'rhe interlock that prevents the simultaneous manual withdrawal of two standard control rods limits the amount of reactivity added per unit time.

l 3.2.3 FACILITY INTERLOCK SYSTEM l Applicability i

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This specification applies to the interlocks that prevent the accidental exposure

• l of an individualin either exposure room.

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

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Specification Facility interlocks shall be provided so that

a. De reactor cannot be operated unless the shielding doors within the

. eactor 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 shielding doors are fully closed; or if the lead shielding doors are fully opened, both exposure room plug doors must be fully 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 operators have visually inspected the room to insure that no personnel remain in the room prior to securing the plug door.

Basis These interlocks prevent the operation and movement of the reactor core into an area until there is assurance : hat inadvertent exposures will be eliminated.

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

Objective

a. To insure the effectiveness of the resins in the water purification system

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

c. To help preclude corrosion of fuel cladding and other components in the primary system.

Specifications

a. The reactor shall not be operated above a thermal power of 5 kilowatts when the purification system input water temperature exceeds 600C.

b. The reactor shall not be operated if the conductivity of the water is greater than 2 micromhos/em (or less than 0.5 x 106 ohms-em resistance) at the output of the purification system, averaged over one week.

c. The reactor shall not be operated if the conductivity of the bulk water is greater than 5 micromhos/cm (or less than .2 x 106 ohms-cm resistance) averaged over 1 week.

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Basis Manufacturer's data state that the resins in the water purification system break down with sustained operation in excess of 600C. 'Ihe 2 micrombos/em is an acceptable level of water contaminants in an aluminum / stainless-steel system of the type at AFRRI. '

Based on experience, activation at this level does not pose a significant radiological

. hazard. Also, the conductivity limits are consistent with the fuel vendor's experience and with similar reactors.

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

Objective The objective is to assure that the ventilation system is operable.

Specification The reactor shall not be operated unicss the facility ventilation system is operable, excapt for periods of time necessary fup to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) to test or permit minor repair of the system. In the event of a sigr.if' cant release of airborne radioactivity in the reactor room, the ventilation system to the reactor room shall be secured via closure dampers automatically by a signal from the reactor deck air particulate monitor. ,

Basis

'During normal operation of the ventilation system, the concentration of argon-41 in unrestricted areas is below the MPC. In the event of a clad rupture resulting in a substantial release of airborne particulate radioactivity, the ventilation system shall be shut down, thereby isolating the reactor room automatically by spring-loaded, positive sealing dampers. Therefore, operation of the reactor with the ventilation system shut down for short periods of time to test or make repairs insures the same degree of control of release of radioactive materials. Moreover, radiation monitors within the building independent of those in the ventilation system will give warning of high levels of radiation that might occur during operation with the ventilation system secured.

3.5 RADIATION-MONITORING SYSTEM AND EFFLUENTS I

3.5.1 MONITORING SYSTEM l

Applicability I . This specification applies to the functions and essential components of the area radiation-monitoring equipment and the system for continuously monitoring '

radioactivity and radiation levels, which must be available during reactor

- operations.

Objective The objective is to assure that adequate radiation-monitoring equipment and radiation information are available to the operator to assure safe operation of the reactor.

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f Specification '

The reactor shall not be operated unless the following radiation-monitoring systems are operable:

, s. Area Radiation-Monitoring System. 'Ihe area radiation-monitoring (ARM) '

system shall have two detectors located in the reactor room, and one .

detector placed near each exposure room plug door so that streaming 4

radiation will be detected.

b.

Gas Stack Monitor. The gas stack monitor (GSM) will sample and measure  ;

the gaseous effluent in the building exhaust system.  !

c. Air Particulate Monitor. The air particulate monitor (APM) will sample the air above the reactor pool. This unit will be sensitive to particulate '

matter from decayed fission products. Alarm of this unit will cause "

closure of the positive sealing dampers, causing reactor room isolation. +

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

TABLE 4. LOCATIONS OF RADIATION MONITORING SYSTEMS Location of Alarm Readout  !

Monitor (A = Audible; V = Visual) Location ARM R1, Reactor Room Control Room A&V Control Room i

,. R2, Reactor Room Control Room V Control Room E3, Exp. Room 1 Area Control Room V Control Room E6, Exp. Room 2 Area . Control Room V Control Room '

GSM - Reactor exhaust Control Room V Control Room i ,

APM - Reactor room Control Room A&V Control Room i Basis r

This system is intended to characterize the normal operational radiological environment -

of the facility and to aid in evaluating any at' normal operations or conditions. The radiation monitors provide information to the operating personnel of any existing or impending danger from radiation, to give sufficient time to evacuate the facility and ,-

take necessary steps to prevent the spread of radioactivity to the surroundings. The l automatic closure of the ventilation system dampers provides reactor room isolation i

from the outside environment, in the event of airborne radioactivity within the reactor l room from fission products decay.

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3.5.2 EFFLUENTS

ARGON-41 DISCH ARGE LIMIT Applicability

'Ihis specification applies to the concentration of argon-41 that may be discharged from the TRIGA reactor facility.

Objective To insure that the health and safety of the public are not endangered by the discharge of argon-41 from the TRIG A reactor facility.

Specification

a. An environmental radiation-monitoring program shall be maintained to determine effects of the facility on the environs,

b. If a dosimeter reading for any designated environmental monitoring station indicates that a probable exposure of 400 millirem above background has been reached during the year as a result of reactor operations, then reactor operations that generate and release to the unrestricted environment measurable quantities of argon-41 shall be curtailed to 2 megawatt-hours per month for the remainder of the calendar year.

c. If a dosimeter reading for any designated environmental monitoring station indicates that an exposure of 500 millirem above background has been reached during the year as a result of reactor operations, reactor operations that generate and release measurable quantities of argon-41 shall be ceased for the remainder of the calendar year.

Basis Since argon-41 does not represent an uptake or bioaccumulation problem, only the direct exposure modality is pertinent with regard to limiting reactor operations. Since direct plume shine may be more controlling than immersion conditions, cumulative exposure is the more appropriate quantification of this limit than maximum permissible concentration values of 10 CFR 20.

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

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Objective I

! The objective is to prevent damage to the reactor or excessive release of radioactive l

• materials in the event of an experiment malfunction, so that airborne concentrations of activity averaged over a year do not exceed 10 CFR 20, Appendix B.

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Speelfications

'Ihe following limitations shall apply to the irradiation of materials (other than air):  ;

a. If the possibility exists that a release of radioactive gases or aerosols may occur, the amount and type of material irradiated shall be limited to assure the yearly compliance with Table II, Appendix B, of 10 CFR 20, assuming that 100% of the -

gases or aerosols escape.

b. Each fueled experiment shall be limited so that the total inventory of iodine -

isotopes 131 through 135 in the experiment is not greater than 1.3 curies and the maximum strontium-90 inventory is not greater than 5 millicuries.

c. F.nown 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 experi-mentally, or by calculations, to be less than the design pressure of the container.

d. Samples shall be doubly contained when release of the contained material could cause corrosion of the experimental facility.

e. The sum of the absolute reactivity worths of all experiments in the reactor and in the associated experimental facilities shall not exceed $3.00 (2.1% Ak/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.

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.  !

2) For a material whose boiling point is above 550C and whose vapor (formed by boiling the material) can escape only through a column of water above the core, up to 10% of the vapor is permitted to escape.

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

h. All experiments placed in the reactor exposure environment shall be either firmly '

secured or observed by a Senior Reactor Operator for mechanical stability, to insure that unintended movement will not cause an unplanned reactivity change or physical damage. All operations in any experimental area shall be supervised by a member of

• the reactor operations staff.

Basis

a. This specification is intended to provide assurame that airborne activities in excess of the limits of Appendix B of 10 CFR 20 wit ;t be released to the atmosphere i outside the facility boundary.

16

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b. 'Ihe 1.3 curie limitation on iodine-131 through -135 assures that, in the event of malfunction of a fueled experiment leading to total release of the iodine, the i particulate iodine trapped by the absolute filtering system will present a minimal  !

hazard to staff personnel should a release occur.

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 a capsule fails.

e. The maximum worth of experiments is limited so that their removal from the cold critical reactor will not result in the reactor achieving a power level high enough to  !

exceed the core temperature safety limit. The three (3.00) dollar limit is less than the SAR analyzed authorized pulse magnitude.

f. This specification is intended to insure that the limits of 10 CFR 20, Appendix B, are not exceeded if an experiment malfunctions.

g. To assure that operation of the reactor with damaged reactor fuel or structure is ,

prevented, the release of fission products to the environment is limited. l

?

h. All experiments placed in the reactor environment shall be either firmly secured or observed for mechanical stability to insure that unintended movement will not cause an unplanned reactivity change or physical damage. ,

3.7 SYSTEM MODIFICATIONS Applicability This specification applies to any system related to reactor safety.

Objective i The objective is to verify the proper operation of any system modification related to reactor safety.

Specification  !

Any additiorc or modifications to SAR stated systems including the ventilation system, the core and its associated support structure, the pool, coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with ,

the specifications to which the systems were originally designed and fabricated, or to specifications approved by the Reactor and Radiation Facilities Safety Committee. A system shall not be considered operable until after it is successfully tested. l l '

Basis l

- This specification is related to changes in reactor systems that could directly affect the safety of the reactor. As long as changes or replacements to these systems continue to

, meet the original design specifications, they meet the presently accepted operating l criteria.

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3.8 ALARA Applicability

'Ihis specification applies to all reactor operations that could result in significant personnel exposures.

Objective .

To maintain all exposures to ionizing radiation to the staff and the general public as low as is reasonably achievable. .

Specification As part of the review of all operations, consideration shall be given to alternative operational profiles that might reduce staff exposures, release of radioactive materials to the environment, or both.

Basis Experience has shown that experiments and operational requirements can, in many cases, be satisfied with a variety of combinations of facility options, core positions, power levels, time delays, and other modifying factors. Many of these can reduce radioactive effluents or staff radiation exposures. Similarly, overall reactor scheduling achieves significant reductions in staff exposures. Consequently, ALARA must be a part of both the overall reactor scheduling and the detailed experiment planning.

18 s to '~

4.0 SURVEILLANCE REQUIREMENTS 4.1 REACTOR CORE PARAMETERS Applicability

- These specifications apply to the surveillance requirements for reactivity control of experiments and systems affecting reactivity.

- Objective The objective is to measure and verify the worth, performance, and operability of those systems affecting the reactivity of the reactor.

Specifications

a. The reactivity worth of each control rod and the shutdown margin shall be determined annually but at intervals not to exceed 15 months,

b. The reactivity worth of an experiment shall be estimated or measured as appropriate, before reactor power operation with an experiment, the first time it is performed.

c. The control rods shall be visually inspected for deterioration annually, not to exceed 15 months.

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

Semlannually, at intervals not to exceed 7.5 months, the transient (pulse) rod drive cylinder and the associated air supply system shall be inspected, cleaned, and lubricated as necessary.

e. The core excess reactivity shall be measured at the beginning of each day of operation involving the movement of control rods, or prior to each continuous operation extending more than a day.

f. The power coefficent of reactivity at 100 kilowatts and 1 megawatt will be measured annually, at intervals not to exceed 15 months.

Basis The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worths of experiments inserted in the core.

1

- Past experience with TRIGA reactors gives assurance that measurement of the reactivity worth, on an annual basis, is adequate to insure that no significant changes in c

the shutdown margin have occurred. Visual inspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation in the reactor.

Functional checks along with periodic maintenance assure repeatable performance.

Excess reactivity measurements assure that core configuration is the same, with no fallen material of reactive value near the core. Knowledge of power coefficients allow 19 u.

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the operator to accurately predict the reactivity necessary to achieve required power levels.

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.

Specification The control rod drop times shall be measured semiannually, but at intervals not to exceed 7.5 months.

Basis Measurement of the scram time on a semiannual basis is a verification of the scram system, and is an indication of the capability of the control rods to perform properly.

4.2.2 REACTOR SAFETY SYSTEMS Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibrations 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 check of the scram function of the high-flux cafety channels shall be made on each day that the reactor is to be operated.

b. A Channel test 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 power level-monitoring channels annually, at intervals not to exceed 15 months.

20 A he

Basis TRIGA system components have operational proven reliability. Daily checks insure accurate scram functions. Weekly channel testing is sufficient to insure the detection of possible channel drift or other possible deterioration of operating characteristics. The channel checks will assure that the safety

. system channel scrams are operable on a daily basis or prior to an extended run.

'The power level channel calibration will assure that the reactor is to be operated at the authorized power levels.

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

Objective To insure operability of the fuel temperature-measuring channels.

Specifications

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

b. A calibration of the fuel temperature-measuring channel shall be made annually, at intervals 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 actually exceeded the safety limit.

Basis Operational experience with the TRIGA system assures that the thermocouple measurements have been sufficiently reliable as an indicator of fuel tempera-ture with proven reliability. The weekly channel test assures operability and indication of fuel temperature. The daily scram check assures scram capabilities.

4.2.4 FACILITY INTERLOCK SYSTEM

? .

Applicability This specification applies to the surveillance requirements that insure the integrity of the facility interlock system.

Objective To insure performance and operability of the facility interlock system.

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Specification

~

Functional checks shall be made annually, but not to exceed 15 months, to inr:ure 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 position 2 with the lead shield doors closed.

c. The warning horn shall sound in the exposure room before opening the lead shield door, which allows the core to move to that exposure room unless cleared by two licensed operators.

Basis These functional checks will verify operation of the interlock system. Experi- i-ence at AFRRIindicates that this is adequate to insure operability.

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

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

Specifications All the fuel elements shall be inspected for damage or deterioration, and measured for length and bow at intervals separated by not more than 500 pulses of insertion greater than $2.00 or annually (not to exceed 15 months), whichever occurs first.

Basis i

I The frequency of inspection and measurement is based on the parameters most likely to affect the fuel claddir.g of a pulse reactor, and the utilization fuel elements whose characteristics are well known.

'Ihe limit of transverse bend has been shown to result in no difficulty in ,

disassembling the core. Analysis of the removal of heat from touching fuel -

elements shows that there will be no hot spots that result in damage to the fuel (caused by this touching). .

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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 assure the integrity of the water purification system, thus maintaining the purity of the reactor pool water, eliminating 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 water at the output of the purification system shall be measured weekly, whenever operations are planned.

Basis Based on experience, at these intervals provides acceptable surveillance of limits that assure that fuel clad corrosion and neutron activation of dissolved materials will not occur.

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

Objective The objective is to assure the proper operation of the ventilation system in controlling the release of radioactive materialinto the unrestricted environment.

Specification The operating mechanism of the positive sealing dampers in the reactor room ventilation system tall be verified to be operable and visually inspected at least monthly.

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

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4.5 RADIATION-MONITORING SYSTEM Applicability This specification applies to surveillance requirements for the area radiation-monitoring equipment and the air particulate monitoring system.

Objective The objective is to assure that the radiation-monitoring equipment is operating and to .

verify the appropriate alarm settings.

Specification The area radiation-monitoring system and the air particulate monitoring system shall be channel tested quarterly, but at intervals not to exceed 4 months. They shall be verified to be operable by a channel check daily when reactor is in operation, and shall be calibrated annually, not to exceed 15 months.

Basis Experience has shown that quarterly verification of area radiation-monitoring and air-monitoring system 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. Annual calibration insures that units are within the specifications demanded by extent of use.

24 s s. -

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

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

Specifications 1

a. The reactor building, as a structurally in'iependent 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 level 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 ventilating system air ducts to the reactor room shall be equipped with positive sealing dampers that are activated by fail-safe controls, which will automatically close off ventilation to the reactor room upon a signal from the reactor room air particulate monitor.

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

P Basis The facility is designed so that the ventilation system will normally maintain a negative ,

pressure with respect to the atmosphere, so that there will be no uncontrolled leakage to the environment. The free air volume within the reactor building is confined when there is an emergency shutdown of the ventilation system. Building construction and gaskets

. around doorways help restrict leakage of air into or out of the reactor room. The stack height insures an adequate dilution of effluents well above ground level. The separate l

ventilation system branch insures a dedicated air flow system for reactor effluents.

l 5.2 REACTOR CORE AND FUEL i

5.2.1 . REACTOR FUEL Applicability These specifications apply to the fuel elements used in the reactor core.

l .

l Objective

! The objectives are to (1) assure that the fuel elements are designed and fabricated in such a manner as to permit their use with a high degree of 25 I

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reliability with respect to their physical and nuclear characteristics, and (2) ,

assure that the fuel elements used in the core are substantially those analyzed in the Safety Analysis Report.

Speelfications The individual nonirradiated standard TRIGA fuel elements shall have the following characteristics: -

i

a. Uranium content: Maximum of 9.0 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 inch thick,

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

Basis A maximum uranium content of 9 weight percent in a standard TRIGA 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%. An increase in local power density of 6% in an individual fuel element reduces the safety margin by 10%, at most. He hydrogen-to-zirconium ratio of 1.7 will produce a maximum pressure within the cladding well below the rupture strength of the cladding.

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 so as to provide assurance that excessive power densities will not be produced.

Specifications

n. He reactor core shall consist of standard TRIGA reactor fuel elements in

• a close packed array and 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 start-up source with holder, and positions for possible in-core experiments.

26 A fe

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

d. In-core experiments shall not be placed in adjacent fuel positions of the B-ring and/or C-ring.

e. Fuel elements indicating an elongation greater than 9.100 inch, a lateral bending greater than .0625 inch, or significant visible damage shall be considered damaged, and shall not be used in the reactor core.

Basis Standard TRIGA cores have been in use for years, and their safe operational characteristics are well documented. Experience with TRIG A reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified to (a) assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment, and (b) assure adequate coolant flow.

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

Objective The objective is to assure 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, and shall 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 or air follower.

5. The transient control rod shall have scram cepability, and shall 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.

Basis The poison requirements for the control rods are satisified by using neutron-absorbing borated graphite, B4 C powder, or boron and its compounds. These materials must be contained in a suitable cladding material, such as aluminum or stainless steel, to insure mechanical stability during movement and to isolate the poison from the pool water environment. Scram capabilities are provided for 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 e pulsing TRIG A reactor.

27 s..

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

Objective The objective is to assure that stored fuel will not become critical and will 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 will permit sufficient natural convective cooling by water or air, so that the fuel element or fueled device temperature will not exceed design values. Storage shall be such that groups of stored fuel elements will remain suberitical under all conditions of moderation.

Basis The limits imposed by this specification are conservative, and assure safe storage and handling. Experience shows that approximately 67 fuel elements are required, of the design used at AFRRI, in a closely packed array 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.

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6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION 6.1.1 STRUCTURE The organization of personnel for the management and operation of the AFRRI reactor facility is shown in Figure 1. Organization changes may occur, based on Institute requirements, and they will be depicted in internal documents.

However, no changes may be made in the Operation, Safety, and Emergency Control Chain, in which the Reactor Facility Director has direct responsibility to the Director, AFRRL Director, AFRRI Chairman. Administrative Operation, Radiation Control Safety, and Reactor & Radiation Ufety Emergency Facility Safety Committee Department Control i

I I I I Advisory Chairman. Advdory 8 Radiation s l Sciences Department l l l 1 1 I I I Chief, Radiation l I Sources Division l 1 1 I I I I L------ Resetor Facility Director ----l Reactor Operations Supervisor i

l l O Reactor Staff

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

j ' Any resetor staff member has access to the Director for matters nf safety.

1 I

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6.1.2 RESPONSIBILITY

'Ihe Director, AFRRI, shall have responsibility for the reactor facility. 'Ihe t Reactor Facility Director (RFD) shall be responsible for administration and  !

operation of the Reactor Facility and for determination of applicability of procedures, experiment authorizations, and maintenance operations. During the -l l'

absence of the Reactor Facility Director, the Reactor Operations Supervisor t shall discharge these responsibilities.

6.1.3 STAFFING

(

6.1.3.1 Selection of Personnel

a. Reactor Facility Director (Reactor Branch Chief) '

I At the time of appointment to this position,' the Reactor Facility Director '

shall have 6 or more years of nuclear experience. Higher education in a scientific or nuclear engineering field may fulfill up to 4 years of

, experience on a one-for-one basis. 'Ihe Facility Director must have held a USNRC Senior Reactor Operator license on the AFRRI reactor for at least  :

1 year before appointment to this position.  ;

b. Reactor Operations Supervisor (ROS)

At the time of appointment to this position, the ROS shall have had 3 years nuclear experience. Higher education in a science or nuclear engineering  ;

field may fulfill up to 2 years of experience on a one-for-one basis. The  !

ROS shall have held a USNRC Senior Reactor Operator license on the '

AFRRI reactor for at least 1 year before the appointment to this position.

c. Reactor Operators / Senior Reactor Operators ,

l 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. t

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

6.1.3.2 Operations 1

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

l

1. A licensed Senior Reactor Operator (SRO) on call but not necessarily -

'l on site

2. Radiation control technician on call i

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3. At least one licensed Reactor Operator (RO) or Senior Reactor Operator (SRO) present in the control room ,

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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.

b. Maintenance activities that could affect the re' ttivity of the reactor shall

. be accomplished under the supervision of an SRL.

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

1. Management personnel (Reactor Facility Director, AFRRI Director) l

?

2. Radiation safety personnel (Head, Radiation Safety Dept) i

3. Other operations personnel (Reactor Staff, ROS). ,

6.1.4 TRAINING OF PERSONNEL A training and retraining program will be maintained, to insure adequate levels  ;

of proficiency in persons invol*ted in the reactor and reactor operations.

6.2 REVIEW AND AUDIT - THE REACTOR AND RADIATION FACILITY SAFETY COMMITTEE (RRFSC)  ;

6.2.1 COMPOSITION AND QUALIFICATIONS 2

6.2.1.1 Composition

,. a. Regular RRFSC Members (Permanent Members) l (1) The following shall be members of the RRFSC: , ,

(a) Chairman, Radiation Safety Department, AFRRI ,

(b) Reactor Facility Director, AFRRI (2) The following shall be appointed to the RRFSC by the Director, AFRRI: .

(a) Chairman as appoint'ed by the AFRRI Directorate.

(b) One to three non-AFRRI members who are knowledgeable in fields i related to reactor safety. At least one shall be a Reactor  ;

Operations Specialist, or a Health Physics Specialist.

b. Special RRFSC Members (Temporary Members)  ;

(1) Other knowledgeable persons to serve as alternates in item a(2)(b) l above as appointed by the AFRRI Director.

(2) Voting ad hoc members, invited by the Director of AFRRI, to assist in review of a particular problem.

31

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, _ _ _ _ _ _ . _ _ . . _ _ , _ _ _ _ _ - . . _ _ . _ _ . , _ _ _ _ , _ , _ _ _ , _ _ . _ . . _ . . , _ . , , _ , i

c. Nonvoting members as invited by the Chairman, RRFSC.

6.2.1.2 Qualifications The minimum qualifications for a person on the RRFSC shall be 6 years of professional experience in the discipline or specific field represented. A baccalaureate degree may fulfill 4 years of experience.

6.2.2 FUNCTION AND AUTHORITY 6.2.2.1 Function -

The Reactor and Radiation Facility Safety Committee is directly responsible to the Director, AFRRI. This committee shall review all radiological health and safety matters concerning the reactor and its associated equipment, the structural reacter facility, and those items listed in Section 6.2.4.

6.2.2.2 Authority The RRFSC shall report to the Director, AFRRI, and shall advise the Reactor Faciny l>irector in those areas of responsibility specified in Section 6.2.4.

6.2.3 CHARTER AND RULES 6.2.3.1 Alternates Alternate members may be appointed in writing by the RRFSC Chairman to ,

serve on a temporary basis. No more than two alternates shall participate on a voting basis in RRFSC activities at any one time.

6.2.3.2 Meeting Frequency The RRFSC or a subcommittee thereof shall meet at least four times a calendar year. The full RRFSC shall meet at least semi-annually.

6.2.3.3 Quorum A quorum of the RRFSC for review shall consist of the Chairman (or designated alternate) and two other members (or alternate members), one of which must be a non-AFRRI member. A majority of those present shall be regular members.

6.2.3.4 Voting Rules Each regular RRFSC member shall have one vote. Each special appointed member shall have one vote. The majority is 51% or more of the regular and special memoers present and voting. -

6.2.3.5 Minutes

.s Minutes of the previous meeting shall be available to regular members at least I week before a regular scheduled meeting.

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6.2.4 REVIEW FUNCTION l

'the RRFSC shall review

a. Safety evaluations for (1) changes to procedures, equipment, or systems and (2) tests or experiments conducted without NRC approval under provisions of Section 50.59 of 10 CFR Part 50, to verify that such actions did not constitute an unreviewed safety question.

b. Changes to procedures, equipment, or systems that change the original intent or use, and are non-conservative, or those that involve an unreviewed safety question as defined in Section 50.59 of 10 CFR Part 50.

c. Proposed tests or experiments that are significantly different from previously approved tests or experiments, or those that might involve an unreviewed safety question as defined in Section 50.59 of 10 CFR Part 30.

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 instrue-tions having nuclear safety significance.

f. Significant variations from normal and expected performance of facility equipment that might affect nuclear safety.

g. Events that have been reported to the NRC.

h. Audit reports of the reactor facility operations.

6.2.5 AUDIT FUNCTION Audits of reactor facility activities shall be performed under tt'e cognizance of the RRFSC, but in no case by the personnel responsible for the i+.em. audited, annually not to exceed 15 months. A report of the findings and recommenda--

tions resulting from the audit shall be submitted to the AFRRI Director. Audits may be performed by one individual who need not be an RRFSC member.. These, audits shall examine the operating records and the conduct of operations, and shall encompass the following: ,

. i, \

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

b. Performance, training, and qualifications of the reactor facility operations <'

, staff, t e  :

I .

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c. Results of all actions taken to correct deficiencies occurri'ig in facility

. equipment, structures, systems, or methods of operation that affect safety. , ,

d. Facility emergency plan and implementing precedures.

e. Facility security plan and implementing procedures.

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f. Any other area of Facility operations considered appropriate by the RRFSC or the Director /AFRRI.

g.' . Rdactor' Facility ALARA Program. lThis program may be a section of the

(.

total AFRRI program.

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b. 6.3 ,

PROCEDURES . f .

T 6.3.1 Writteni instructions for certain activities shall be approved by the Reactor Facility < Director and reviewed by the Reactor and Radiation Facility Safety

• Committee (RRFSC). The procedures shall be adequate to assure safe operation of the reactor, but shall not preclude the use of independent judgment

,and' action as deemed necessary. The activities are as fouow:

a. . Conduct of irradiations and experiments that could affect the operation j;

/ .' and safety of the reactor.

4, b. Reactor staff-training program.

, 3 c. Surveillance, testing, and calibration of instruments,' components, and

systems involving nuclear safety. )

c 1

d. Personnel radiation protection consistent with 10 CFR 20.

i- '

4 e. Implementation of required plans such as the Security Plan and Emergency Plan, t-S f. Reactor core loading and unloading. ,

g. Checkout startup, standard operations, and secur'ing facility.

6.3.2 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 ROS. All such i' temporary changes shall be documented 'and subsequently reviewed and approved lDy the Reactor Facility Director.

g 6.ds REVIEW AND AP' PROVAL OF EXPENhMENTS 6.4.1 Behore issua'nce of a reactor authorization, new experiments shall be reviewed

/ for radiological safety and approved by the following:

t

a. Reactor Facility Director, Reactor Branch

, ,y' ,

b. Radiation Safety Department - ,

-s iC c. Reactor and Radiation Facility Safety Committee (3RFSC) k 6.4.2 Prior to its performance, an experiment shall be includhd under one of the following types of authorizations: ,

a. Special Reactor Authorization for new experiment $. cc experiments not included in a Routhe Reactor Authorization. These experiments shall be l r

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performed under the direct supervision of the Reactor Facility Director or designee.

b. Routine Reactor Authorization for experiments safely performed at least l once. These experiments may be performed at the discretion of the Reactor Facility Director and coordinated with the Radiation Safety Department when appropriate. 'These authorizations do not require additional RRPSC 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 may be performed under the authorization of the Reactor Facility Director or Reactor Operations Supervisor.

6.4.3 Substantive (reactivity worth of more than i$0.25) changes to previously approved experiments shall be made only citer review by the RRFSC and after approval (in writing) by the Reactor Facility Director or designated alternate.

Minor changes that do not significantly alter the experiment (reactivity worth of less than i$0.25) may be approved be the ROS. Approved experiments shall be carried out in accordance with established procedures.

6.5 REQUIRED' ACTIONS 6.5.1 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 NRC.

b. The safety limit violation shall be reported to the Director of NRC Region I, Office of Inspection and Enforcement (or designate); the Director, AFRRI; and the RRFSC not later than the naxt working day.

c. A Safety Limit Violation Report shall be prepared. This report shall be i reviewed be the RRFSC, and shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation on facility components, structures, or systems, and (3) corrective action taken to prevent or reduce the probability of recurrence.

d. 'Ihe Safety Limit Violation Report shall be submitted to NRC; the Director, AFRRI; and the RRFSC within 14 days of the violation.

6.5.2 REPORTABLE OCCURRENCES Reportable occurrences as defined in 1.21 (including causes, actual or prooable '

. consequences, corrective actions, and measures to prevent recurrence) shall be reported to the NRC. Supplemental reports may be required to fully describe ,

the final resolution of the occurrence.

a. Prompt Notification With Written Followup. The types of events listed below shall be reported as soon as possible by te'.ephone and confirmed by 35 s.

~

telegraph, mailgram, or similar transmission to the Director of the appropriate NRC Regional Office (or designate) no later than the first workday following the event, with a written followup report as per 10 CFR.

'Ihe 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. .,

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

(2) 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 base 3 for the Technical Specifications that have or ,

could have permitted reactor operation with a smaller margin of t

safety than in the erroneous analysis.

(3) Performance of structures, systems, or components that requires remedial actiori 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 plant 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.

6.6 REPORTS In addition to the epplicable reporting requirements of Title 10 of the Code of Federal Regulations, the following reports shall be submitted to the Director of the appropriate NRC Regional Office unless otherwise noted.

6.6.1 OPERATING REPORTS

a. Startup Report: A summary report of planned startup and power escalation

! testing shall be submitted following (1) receipt of an operating license; (2) l amendment of the license involving a planned increase in power level;

- (3) installation of fuel that has a different design; and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the reactor. The report shall address each of the tests identified in the Safety Analysis Report and shall, in general, include a .

description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these l

values with design predictions and specifications. Any corrective actions ,

that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on other commitments shall be included in this report, ,

Startup Reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all 36

i l

i three events (i.e., initial criticality, completion of startup test program, i and resumption or commencement of power operation), supplementary l reports shall be submitted at least every 3 months until all three events  ;

have been completed.  ;

b. Annual Operating Report: Routine operating reports covering the opera- l

. tion of the unit during the previous calendar year shall be submitted prior to March 31 of each year, covering the previous calendar year's operation. i The Annual Operating Report made by license shall provide a comprehen- ,

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

Each annual operating report shall include l (1) A brief narrative summary of (a) Changes in facility design, performance characteristics, and oper-ating procedures related to reactor safety, that occurred during the reporting period (b) Results of surveillance test and inspections f

(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, including the reasons and the l corrective action taken, if applicable (4) Discussion of the major safety-related corrective maintenance _per- ,

formed during the period, including the effects (if any) on the safe operation of the reactor, and the reasons for the corrective mainte-nance 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 i Report

~,- (c) Any new experiments or tests performed during the reporting period that are 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 puwjant to  ;

Section 50.59 of 10 CFR Part 50. The summary shall cleady show t'ae reason leading to the conclusions that no unraviewed safety question existed and that no change to the Technical Specifications was  ;

required.

37

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---v.,, ,-.-m- - . ~ . - - - - - . . - - - , ~

f (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 j is less than 25% of the concentration allowed, a statement to this  ;

effect is suffichnt.

(a) Liquid Was. ;ummarized on a quarterly basis) -

(i) Radioactivity discharged during the reporting period , l Total radioactivity released (in curies) l MPC used and isotopic composit!on if greater than 3 x 10-6 microcuries/ml for fission and activation products Total radioactivity (in curies), released by nuclide during the f reporting period, based on representative isotopic analysis j Average concentration at point of release (in microcuries/cc) during the reporting period {

i (ii) Total volume (in gallons) of effluent water (including diluent) during periods of release (b) Gaseous Waste (summarized on a quarterly basis)

Radioactivity discharged during the reporting period (in curies) for:

r Argon-41 Particulate with half-lives greater than 8 days.

(c) Solid Waste (Summarized on a quarterly basis)

Total cubic feet of 3 to 83 material in solid form disposed of under R-84.

(8) A description of the results of any environmental radiological surveys performed outside the facility P

(9) A list of exposures greater than 25% of the ellowed value (10CFR 20) received by reactor personnel or visitors to the reactor facility 6.7 RECORDS l 6.7.1 RECORDS TO BE RETAINED FOR A PERIOD OF AT LEAST 5 YEARS [

OR AS REQUIRED BY 10 CFR REGULATIONS ,

a. Operating logs or data that shall identify  !

(1) Complation of pre-startup checkout, startup, power changes, and shut-  :

down of the reactor (2) Installation or removal of fuel elements, control rods, or experiments )

that could affect core reactivity 38 1

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-.. .. ,, ,,,, __.._, ..- -_.. ,.-- . -.3

-w - . ..-,w._m,,. .__% , _ . , , . _ , m

(3) Installation or removal of jumpers, special tags, or notices of oth.ir temporary changes to bypass reactor safety circuitry (4) Rod worth measurements and other reactivity measurements

b. Principal maintenance operations

c. Reportable occurrences

d. Surveillance activities required by Technical Specifications

e. Facility radiation and contamination surveys

f. Experiments per.' armed with the reactor This requirement may be satisfied by the normal operations log book plus (t) Records of radioactive material transferred from the Reactor Facility as required by license (2) Records required by the RRFSC for the performance of new or special experiments

g. Changes to operating procedures

h. Fuel inventories and fuel transfers

i. Records of transient or operational cycles for those components designed for limited number of transients or cycles

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

k. Records of reviews performed for changes made to procedures or equip-ment, or reviews of tests and experiments pursuant to Section 50.59 of 10 CFR Part 50

1. Records of meetings of the RRFSC 6.7.2 RECORDS TO BE RETAINED FOR AT LEAST ONE COMPLETE TRAINING CYCLE

a. Training exams

b. Requalification records

- 6.7.3 RECORDS TO BE RETAINED FOR THE LIFE OF THE FACILITY

a. Gaseous and liquid radioactive effluents released to the environs

b. Appropriate off-site environmental monitoring surveys B

c. Radiation exposures for all personnel

d. Updated as-built drawings of the facility.

39 .

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