ML20115C601

Forwards Updated TS for Afrri Reactor Facility

ML20115C601
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	09/30/1992
From:	Maria Moore ARMED FORCES RADIOBIOLOGICAL RESEARCH INSTITUTE
To:	Mendoza M NRC
References
NUDOCS 9210200011
Download: ML20115C601 (1)

Text

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DEFENSE NUCLEAR AGENCY

i. / - "' d ARMED FORCES 9ADIOBIOLoGY RESEARCH INSTITUTE 3901 WISCONSIN AVENUE BETHESDA, Mr"YL AND 20809-5603 RSD 30 September 1992

SUBJECT:

Inform the new publishing of Technica! Specifications fu. AFRRI Facility License No. R-84.

Mr. Marvin Mendoza, Project Manager 11B-20 United States Nuclear Regulatory Commission Washmgton D. C. 20555

Dear Mr. Mendoza:

The Armed Forces Radiobiology Research Institute (AFRRI) reactor staff updated the Technical Specifications for AFRRI facility license R-84, Docket No. 50 170. This update was appr ved by the U.S. Nuclear Regulatory Commission.

Enclosed are copies of our newly published document.

Please destroy all old Technical Specification you may have.

Should you have any questions concerning this document, please fell free to contact me at (301) 295-1290.

Sincerely,

'k b m[

Mad Moore Reactor Facility Director 160082 bx i 9210200011 920930 PDR ADOCK 03000170 P

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Specifications

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for The AFRRI Reactor Facility Mark Moore Docket 50-170 License R-84 November 1991 Defense Nuclear Agency Armed Forces Radiobiology Research. Institute Bethesda, Maryland 20889-5603 Approved For Public Release -- Distribution Unlimited i __g,

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'I TECHNICAL SPECIFICATIONS FOR THE AFRRI REACTOR FACILITY November 1991 LICENSE R-84 DOCKET 50-170 The original of this document.

f is on file with the

- U.S. Nuclear Regulatory Commission Amendment No. 22-

Reviewed and Approved 1

19 NOV 1991 4

L hi ARK L. MOORE (

Date Reactor Facility Director 1

a Approved for Release

- Fj i Y / 9 \\

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ROBERT L. BUMpARNER Date Captain, MC, Ua"N Director Amendment No. 22 l

Preface included in this document are the Technical Specifications and the " Bases" for the Technical Sp*cifications. These bases, which provide the technical support for the individual technical specifications, are includedfor information purposes only. They are not part of sne Technical Specifications, and they do not cunsti-tute limitations or requirements to which the licensee must adhere.

E Amendment No. 22

TECHNICAL SPECIFICATIONS FOR THE AFRRI REACTOR FACILITY LICENSE NO. R-84 DOCKET # 50170 TABLE OF CONTENTS 1.0 DEFINITIONS

f. age 1.1 ALARA 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 Rod 4

1.26 Steady State Mode 4

1.27 Transient Rod 4

2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING 2.1 Safety Limit - Fuel Element Temperature 5

2.2 Limiting Safety System Setting for Fuel Temperature 5

3 0 LIMITING CONDITIONS FOR OPERATIONS 3.1 Reactor Core ' Parameters 7

3.1.1 Steady State Operation 7

Amendment No. 22

hu 3.1.2 Pulse Mode Operation 7

3.1.3 Reactivity Limitations 8.

3.1.4 Scram-Time 8

3.2 Reactor Control and Safety Systems 9

3.2.1 Reactor Control System 9

3.2.2 Reactor Safety System 10 3.2.3 Facility Interlock System 11 3.3 Coolant Systerr.s 12 3.4 Ventilation System 13 3.5 Radiation Monitoring System and Effluents 3.5.1 Monitoring '.atem 13

3.5.2 Effluents

Argon-41 Discharge Limit 15 3.6 Limitations on Experiments 16 3.7 System Modifications 18 3.8 ALARA 18 4.0 SURVEILLANCE REQUIREMENTS 4.1 Reactor Core Parameters 20 4.2 Reactor Control and Safety Systems 21 4.2.1 Reactor Control Systems 21 4.2.2 Reactor Safety Systems 21 4.2.3 Fuel Temperature 22 4.2.4 Facility Interlock System 22 4.2.5 Reactor Fuel Elements 23 4.3 Coolant Systems 24 4.4 Ventilation System 24 4.5 Radiation-Monitoring System 25 5.0 DESIGN FEATURES 5.1

' Site and-Facility Description 26 5.2 Reactor Core and Fuel 26 5.2.1 Reactor Fuel 26 5.2.2 Reactor Core -

27 5.2.3 Control Rods 28-5.3 Special Nuclear Material Storage 29 Amendment No. 22

6.0 ADMINISTRATIVE CONTROLS bgt 6.1 Orge,ization 30 6.1.1 Structure 30 6.1.2 Responsibility 31 6.1.3 Staffing 31 6.1.3.1 Selection of Personnel 31 6.1.3.2 Operations 31 6.1.4 Training of Personnel 32 6.2 Review and Audit The Rector and Radiation Facility Safety Committee (RR FSC) 32 6.2.1 Composition and Qualifications 32 6.2.1.1 Composition 32 6.2.1.2 Qualifications 33 6.2.2 Function and Authority 33 6.2.2.1 Function 33 6.2.2.2 Authority 33 6.2.3 Charter and Rules 33 6.2.3.1 Alternates 33 6.2.3.2 Mecting Frequency 33 6.2.3.3 Quorum 33 6.2.3.4 Voting Rules 33 6.2.3.5 Minutes 34 6.2.4 Review Function 34 61.5 Audit Function 34 6.3 Procedures 35 6.4 Review and Approval of Experiments 35 6.5 Required Actions 36 6.5.1 Actions To Be ~ rhe.n in Case of Safety Limit Violation 36 6.5.2 Reportable Occurrences 37 6.6 Reports 37 6.6.1 Operating Reports 37 67 Records 40 6.7.1 Records To Be Retained For A Period of At Least Years or As Required by 10 CFR Regulations 40~

67.2

. Records To Be Retained For At Least One Complete

_ Training Cycle -

41 6.

D c'ds To Be Retained For The Life of The Facility 41-lii Amendment No. 22

!.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 tc the envi'ronment as low as reasonably achievable.

1.2 Cil ANNEL, CALIBR ATION A channel calibration consists of using a known signal to verify or adjust a channel to produce an output that corresponds with receptable accur cy 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 CllANNEL C11ECK A channel check is a verification of acceptable performance by observation of channel behavior.

1.4 Q1ANNEL 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 40'C.

1.6 CORE GRID POSITION The core grid position refers to the locatiot. of a fuel or control element in the grid StruClu?e.

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 AFRR1 TRIGA reactor shall be a.

Exposure Room #1 b.

Exposure Room #2 hendmens No. 22 1

NOTE: Exposure fuellities protective barriers shall be differentiated from the primary protective barrier (fuel element cladding) for purposes of placement of experiments within these barriers, c.

Reactor Pool d.

Core Experiment Tube e.

Portable Beam Tubes f.

Pneumatic Transfer System

g. Incore Lacations 1.9 FUFL Ff FMENT A fuel element is a single TRIGA fuel rod, or the fuel po. tion of a fuel follower control rod.

1.10 INSTRUMENTED ELEMENT An instrumenteh the most reactive control rod fully withdrawn or removed shall be

$0.50(0.35% Ak/k) for any condition of operation.

ILuis

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

b. The shutdown margin assures tnat 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 SCR AM 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.

8

DMik This specification assures that the reactor will be promptly shut down when a scram signalis 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 3.2.1 REACTOR CONTROL SYSTEM Applicabilitv This specification applies to the channels monitoring the reactor core, which must provide information to the reactor operator during reactor cperation.

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

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

TABLE 1. MEASURING CHANNELS Minimum Number Ooerabh in Effective Mode Steady State Pulse Fuel Temperature Safety Channel 2

2 Linear Power Channel 1

1 Log Power Channel 1

0 i

High Flux Safety Channel 2

l' Pulse Energy Integrating Channel 0

1*

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

Ilub 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 specifications on reactor power level indication are included in this Section, since the power level is related to the fuel temperature, 9

Amendment No. 22

i l

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

Objective The objective is to spccify 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 described in Tables 2 and 3 are operable.

TABLE 2. MINIMUM REACTOR SAFETY SYSTEM SCRAMS Maximum Minimum Number in Mode Channel Set Point Steady State Pulse Fuel Temperature 600 C 2

2 Percent Power, High Flux 1.1 MW 2

0 Console Manual Scram Bar Closure switches 1

1 High Voltage Loss to Safety Channels 20% loss 2

1 Pulse Time 15 seconds 0

1 Emergency Stop Closure switch 1

1 (1 each exposure room, 1 on console)

Pool Water Level 14 feet frorn top 1

1 of core Watchdog (DAC to CSC)

On digital console 1

1 Ilash 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. Ir.

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

10 Amendment No. 22

The emergency stop allows personnel trapped in a potentially hazardous exposure room or the reactor operator to stop actions through the intertvk system. The pool water level insures that a loss of biological shielding would result in a reactor shutdown. The watchdog scram will insure adequate communication between the Data Acquisition Computer (DAC) and the Control System Computer (CSC; units.

TABLE 3. MINIMUM REACTOR SAFETY SYSTEM INTERLOCKS i

Effective Mode Action Prevented Steady State Pulse i

Pulse initiation at power levels greater X

than 1 kilowatt Withdrawal of any control rod except transient X

Any rod withdrawal with count rate in X

X operational channel below 0.5 cps Simultaneous manual withdrawal of two X

standard rods 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 movument 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 conaitions. The interlock that prevents the simultaneous manual withdrawal of two standard control rods limits the amount of reactivity added per unit time.

3.2.3 FACILIYY INTERLOCK SYSTEM Apolicability This specification applies iv the interlo;ks that prevent the accidental exposure of an individual in either exposure room.

Objectivt The objective is to provide sufficient warning and interlocks to prevent movement of the reactor core to the exposure room in which someone may 11 Amendment No. 22

i l

be working, or prevent the inadvertent movement of the core into the lead shield doors.

Specifica. tion.

Facility interlocks shall be provided so that a.

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

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

c. The lead shield doors cannot be opened to allow movement into the exposure rc,om 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.

D2Si1 These interlocks prevent the operation and movement of the res.ctor ct..

mo an area until there is assurance that inadvertent exposures will be eliminated.

3.3 COOLANT SYSTEMS APdGhili1L This specification refers to 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 c:tivated contaminants from becoming a radiological hazard.

c.

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

Specifications.

a. The reactor shall not be operated above a therma; power of 5 kilowatts when the purification system input water temperature exceeds 60 C.

b. The reactor shall not be operated if the conductivity of the water is greater than 6

2 micrombos/cm (or less than 0.5 x 10 ohms cm resistance) at the output of the purification system, averaged over one week.

The reactor shall not be operated if the conductivity of the bulk water is c.

greater than 5 micrombos/cm (or less than 0.2 x 10 ohms cm resistance) averaged over 1 week, Amendment No. 22 g

Dmia Manufacturer's data state thst the resins in the water purification system break down with sustained operation in excess of 60 C. The 2 micrombos/cm is an acceptable level of water contaminants an aluminum / stainless steel system of the type at AFRRI. Based on experience, activation at this lesel 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.

Ohiective ne objective is to assure that the ventilation aystem is operable.

Spssifica11a The reactor shall not be operated unless the facility ventilation system is operable, except for periods of time necessary (up 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 significant release of altborne 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.

lhili During normal operation of the ventilaticn system, the concentration of argon 41 in unrestricted areas is below the limits allowed by 10 CFR 20. 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 recctor 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 buildin;; independent of those in the ventilation system will give waming of high levels of radiation that might occur during operation with the ventilation system secured.

3 5 B ADIATION MONITORING SYSTEM AND EFFLUENTS 3.5.1 MONITORING SYSTEhi Apelicability.

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.

Amendment No. 22 13 J

Objective The objective is to asr"re. that adequate radiation monitoring equipment and radiation information are available 'o the operator to assure safe operation of the reactor.

Specification The reactor shall not be operated unless it'e following radiation monitoring systems are operable:

Area Radiation hionitoring System. The area radiation monitoring a.

(ARhi) system shall have two detectors located in the reactor room, and one detector placed near each exposure room plug door so that streaming radiation will be detected.

b.

G as Stack hionitor. The gas stack monitor (GShi) will sample and measure the gaseous effluent in the building exhaust system.

Air Particulate hionitor. The air particulate monitor (APhi) will sample c.

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 menitors.

TABLE 4. LOCATIONS OF RADIATION hiONITORING SYSTEhiS Location of Alarm Readout hionitor (A = Audible; V = Visual)

Location l

ARhi.

R1, Reactor Room Control Room A&V Control Room j R2, Reactor Room Control Room V Control Room !

E3, Exp. Room 1 Area Control Room V Control Room E6, Exp. Room 2 Area ControlRoom V Control Room GShi. Reactor exhaust Control Room V Control Room APhi - Reactoc room Control Room A&V Centrol Room Enis This system is intended to characterize the normal operational radiological environment of the facility and to aid in evaluating any abnormal operations or conditions lae radiation monitors provide information to the operating Amendment No. 22

)

14

1 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 surrcundings. The automatic closure of the ventilation system dantpers provides reactor room isolation from the outside environment, in the evant of airborne radioactivity within the reactor room from fission products decay.

3.5.2 EEf LUENTS: ARGON-41 DISCHAtlGE LIMIT Applicability This specification applic4 to the concentration of argen 41 that may be discharged from the TRIGA reactor facility.

Obituite.

To i.1sure that the health and safety of the public are not endangered by the discharge of argon 41 from the TRIGA reactor facility.

Spssincation.

a.

An environmental radiation monitoring program snall be maintained to determine eft'cets 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 duriag 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 millitem above background has been reached during the year as a retu't of reactor operations, reactor operations that generate and relense measurable quantities of argon-41 shall be ceased for the remainder of the calendar year.

Ilasis Since argon 41 does not represent an pake or bioaccumulation problem, only the direct expoe modality is pertinent with regard to limiting reactor operations. Since direct plu.ne shine may be more controlling than

+

immersion conditions, cumulative exposure is the more appropriate quantification of this limit that. the concentration limit values in Appendix B, 10 CFR 20.

15 Amendnent No. 22

3.6 L1hilTATIONS ON EXPERIMENTS Applicability This specification applies to experiments instailed in the reactor and its exp"imental facilities.

Dhitstic The objective is to prevent damage to the reactor or excessive release of radioac*ive materials in the event of an experinient malfunction, so that airborne concentrations of activity averaged over a year do not exceed 10 CFR 20, Appendix B.

Spscifications.

The following liniitations shall apply to the irradiation of materials (other than ait):

If the poss.bility exists that a release of radioactive gases or aerosols may a.

occur, the amount and typt. of material irradiated shall be limited to assure the yearly compliance with Table 2, 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 totalinventory 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.

Known explosive materials shall not be irradiated in the r: actor in c.

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 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, c.

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%

A k/k).

This includes the total potential reactivity insertion that might result from expenment 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.

16 I?'andment No. 22

2)

For a material whose boiling point is above 55"C and whose vapor (formed by boiling the material) can escape only through a column of water above the core, up to 10cc of the vapor is permitted to escape.

g.

If a capsule fails and releases materials that could damage the reactor fuel or structure by corrosien or other means, physical inspection shall be performed to determine the consequences and need for corrective action.

The results of the inspeion and any corrective action taken shall be reviewed by the Reactor Facility Director, and shall be determined to be satisfactory before operation of the r. tor is resumed.

h.

All experiments placed in the reactor exposure environment shall be eithei firmly secured or observed by a Senior Reac'or Operatr r for mechanical a

stability, to insure tha unintended movement will not cause an unplanned reactivity chnge or physical damage. All operations in any experimental area shall be supervised by a member of the reactor operations staff.

Unis This specification is intended to provide assurance that airborne activities a.

in excess of the limits of Appendix B of 10 CFR 20 will not be released to the atmosphere outside the facility boundary.

b. The 1.3 c1:rie 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 particulate iodine trapped by the absolute filtering system will present a minimal hazard to staff personnel should a release occur.

This specification is intended to prevent damage to reactor components c.

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 possibk if a capsule fails, The maximum worth of experiments is limited so that '.i.eir removal from c.

the cold critical reactor will not result in tU teactor achieving a power level high enough to exceed the core temperature safety limit. The three (3.00) dollar limit is less than the SAR ana'yzed 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 re.netor with damaged reactor fuel or structure is prevented, the relens.

fission products to the environment is timited.

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 nhysical damage.

17 Amendment No. 22

.l

3.7 SUTEM MODIElCATIONS ApplicabilitL This specification applies to any system related to reactor safety.

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

Specification.

Any additions or modifications to SAR stated systems including the ventilation system, the core and its associated support structure, :he pool, ecolant system, the rod drive meclianism, 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 Facility Safety Committee. A system she.ll not be considered operable until after it is successfully tested.

Dails 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 systemt.

continue to meet the original desip specificatians, they meet the presently accepted operating criteria.

3.8 ALARA APPECAbili1L This specification applies to all reactor operations that coulo result in significant personnel exposures.

Obhcliya To mamtain all expo ares to ionizing radiation.o he staff ad the general public as low as is seasonably achievable.

Specificatintt 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 boin.

Dmit 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 18 Amendment No. 22

scheduling achieves significant reductions in staff exposures. Consequently, ALARA must be a part of both the overall reactor scheduling and the detailed experiment planning, 19 Amendment t4o. 22

4.0 SUR VEILLANCE REOUIREMENTS 4.1 REACTOR _ CORE PAR AMETERS Applicability _

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

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

Specifications The reactivity worth of each control rod and the shutdown margin shallbe.

a.

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 powre 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 tenctionai performance check of the transient (pulse) tod system shall be performed.

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

Th: 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 coefficient 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 reevired shutdown margin is available and to provide an accurate means for determining the reactivity worths o' experiments inserted in the core.

Past experience with TRIGA reactors gives assurance that measurement of the reactivity worth, on an annual basis, is adequate to insure that no significant changes in the shutdown margin have occurred. Visualinspection 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 20 Amendment No. 22

)

i 1

coefficients allow 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 SYS" EM S Aeolicability, 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 renner.

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

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

4.2.? REACTOR SAFETY SYSTFM Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibiations 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.

l Specifications a.

A check of the scram function of the high flux safety 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 calibra*. ion shall be made of the power level monitoring annually, ct intervals not to exceed 15 months.

Halii y

TRIG A system components have operational proven reliability. Daily checks insure accurate scram functions. Weekly channel testing is sufficient 21 Amendment No.-22

to insure the detection of possible channel drift or other possible l

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 TEMPERAT_URE 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 check of the fuel temperature scrams shall be made on each day that a.

the reactor is operated, b.

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

A weekly channel test shall be performed on fuel temperature measuring c.

channels, whenever opera' ions 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 syste'n assures that the thermo-couple measurements have been sufficiently reliable as an indicator of fuel temperature 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 INTERLOCKf YSTEM Apnlicability This specification applies to the surveillance requirements that insure the integrity of the facility interlock system.

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

Specification Functional checks shall be made annually, but not to exceed 15 months, to insure the fo' lowing:

22 Amendment No. 22

a, With the lead anie;d 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 doors, which allows the core to move to that exposure room unless cleared by two licensed operators.

Ihsis These functional checks will verhy operation of the interlock system.

Experience at AFRR1 indicates that this is adequate to insure operability.

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

DMt.cliy.c.

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

Soecifications All the fuel elements present in the reactor core, to include fuel follower control rods, shall be inspected for damage or deterioration, and measured for length and bow at intervals separated by not more tha. 500 pulses of insertion greater than $2.00 or annually (not to exceed 15 months),

whichever occurs first. Fuel elements in long term storage need not be measured until returned to core; however fuel elements routinely moved to temporary storage shall be measured every 500 pulses of insettlon greater than $2.00 or annually (not to exceed 15 months), whicheve occurs first.

lhili The frequency of inspe.ction and measurement is based on the parameters most likely to affect the fuel cladding of a pulse reactor, ar.d the utilization of fuel elements whose characteristics are well known.

t The limit of transverse bend has been shown to resuh in no difficulty in disassembling the core. Analysis of a worst case scenario in which two

-!jacent fuel elements suffer sufficiently severe transverse bends to result in the touching of the fuel elements has shown that no damage to the fuel elements will result via a hot spot or any other known mechanism.

23 Amendment No. 22

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

D.bitt111t The objective is to assure the integrity of the water purification system, thus maintaining the purity of the reactor pool water, climinating possible radiation hazards from activated impurities in the water system, and limiting the potential corrosion of fuel cladding and other components in the prirrary water system.

Specifications The pool water temperature, as measured near the input to the water a.

purification system, shall be measured daily, whenever operations are planned.

b.

The conductivity of the water at the output of the purification system shallbe measured weekly, whenever operations are planned.

Euis Based on experience, observation 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 SYSTFM Acolicability 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 material into the unrestricted environment.

Snecificatia The operating mechanism of the positive sealing dampers in the reactor room centilation system shall be verified to be operable and visually inspected at least monthly (interval not to exceed six weeks).

Euis 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.

24 Amendment No. 22 N -- -_- _ _- -__

4.5 BADIATION-MONITORING SYSTEh1 Applicability This specification applies to surveillance requirements for the area radiation monitoring equipment and the air particulate monitoring system.

Oldtslitc.

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'. e channel tested quarterly, but at intervals not to exceed 4 months. They shall be verified to be operable by a channel check daily when the reactor is in l

operation, and shall be calibrated annually, not to exceed 15 months.

D.nis Experience has shown that quarterly verification of area radiation monitoring and air monitoring system set points in conjunction with a quanerty 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 the units are within the specifications demanded by the extent of use.

)

25 Amendment No. 22 t

5.0 DESIGN FEATURES 5.1 SITE AND FACILITY DFSCRIPT10fi Applicability, This specification applies to the building that houses tb-reactor.

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

Sncifications

a. The reactor building, as a structurally independent building in the AFRR1 complex, shall have its own ventilation system branch. The effluent from the reactor ventilation systern shall exhaust through absolute filters to a stack having a minimum elevation that is 18 feet above the roof of th highest building in the AFRR1 complex.

b.

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

The ventilation system air ducts to the reactor room shall be equipped with c.

positive sealing dampers tha, are activated by fail safe controls, which will automatically close off ventilation to the r-actor room upon h 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.

Dasis The facility is designed so that the ventilation 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 ventila: ion 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 ventilation system branch insures a-dedicated air flow system for reactor effluents.

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

26 Amendment No. 22

Objective These 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 reliability with respect to their physical ano nuclear characteristics, and (2) assure that the fur! elements used in the core are substantially those analyzed in the Safety Analysis Report.

Specifications The individual nonirradiated standard 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 will 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 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.

Hails A maximum unnium 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 fun element reduces the safety margin by 10%, at most. The hydrogen.to-zirconium ratio of 1.7 will produce a maximum pressure within the cladding 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 standard TRIG A fuel element in the D Ring.

The volume of fuel in a fuel followed rod is 56% of the volume of a standard TRIG A fuel element. Therefore, the actual power produced in the fuel followed rod is 33% less than the power produced in a standard TRIGA fuel element in the D r'.ng 5.2.2 R EACTOR ___COR E Applicqhiiity,.

These specifications apply to the configuration of fuel and in-core experiments.

27 Amendment No. 22

Obitstiy.L The objective is to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will no.. produced.

Specifications a.

The reactor core shall consist of standard TRIG A reactor fuel elements in a close packed array and a minimurn of two thermocouple instrumented TRIGA reactor fuel elements, b.

There shall be four single core positbns occupied by the three standard control rods and transient rod, a neutron start up source with *aolder, and positions for possible in. core experiments, The core shall be cooled by natura) convection water flow.

c.

d.

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

Fuel elements indicating an elongation greater than 0.100 inch, a lateral e.

bending greater than 0.0625 inch, or significant vi ible damage shall be s

considered damaged, and shall not be used in the reactor core, q

Ihiis Standard TRIGA cc es have been in use for years, and & ' safe operational characteristics are well documented. Experi-ice with 'I cenctors has shown that fuel element bowing that could result in touc...g 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 3

cause a loss of integrity in the fuel containment, and (b) assure adequate coolant flow.

5.2.3 CONTRO!E E Applicability 3ese specifications apply te the control rods used in the reactor core.

Obiectivr.

He objective as to assure tnat the control rods are designed to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications The standard control rods shall have scram capability, and sh:.ll contain a.

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 the Spec'.ncations of 5.2.1.

28 Amendment No. 22

b.

The transient control rod 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. This rod may incorporate an aluminum, poison, or air follower.

Batis 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 suainum or stainless steel, to insure mechanical stability during movement and to l

isolate the poison from the pool water environmnt. Scram capabilities are I

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 j

for use in a pulsing TRIGA reacter.

5.3 SPECI AT NUCI EAR M ATERI AL STOR 4f __

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.

Soecification All fuel elements not in the reactor core shall be stored and handled in accordance with applicable regulations. Irradiated fuel elements and fueled devices snall 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 l

values. Storage shall be such that groups of stored fuel elements will remain j

suberitical under all conditions of moderation.

I Basis l'

The limits imposed by this specification are conservative and a_,are safe storage and handling. Experience shows that approximately 67 fuel elements are required, i-of the design sed at AFRRI, in a closely packed array to achieve criticality.

L Calculations show that in the event of a full storage rack failure with all 12 elements fallbg 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.

29 Amendment No. 22

i 6.0 ADMINISTR ATIVE CONTROLS 6,1 OR G A NIZATION 6.1.1 STR U CTU R E The organization.of personnel for the management and operation of the AFRR1 reactor facility is shown in Figure 1. Organization changes may occur, based on Institute requirements, and they will be depicted on 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, AFRRI.

Dinctor. AFRR1 AFRRI gw, oPermW Safe.

Reactor and Radlauon

$7 Tacihty Safety Commiue.

Radiation Protection w

Officer Cha m a Radiauon Sowces Dept.

Adnar7 Acu=7 l

l l

l l

Reacter facuity Director

..............s' l

Reactor Operidons Supemsos I

Ranctor Opersucos Staff

• t Figure 1.

Organization of Perso..'el for Management and Operation of the AFRRI Reactor Facility.

• Any reactor staff member has access to the Director for matters of safety 30 Amendment No. 22

6.1.2 RESPONSIBILITY The Director, AFRR1, shall inave license responsibility for the reactor facility. The Reactor Facility Director (RFD) shall be responsible for administration and operation of the Reactor Facil' y and for determination of applicability of procedures, experiment authorizations, maintenance, at operations. The RFD may designate an individual who meets the requirements of Section 6.1.3.1.a to dischargs ths RFD's responsibilities in the RFD's absence. During brief absences (periods less than four hours) of the Reactor Facility Director and his designee, the Reactor Operations Supervisor shall discharge these responsibilities.

6.1.3 STAFFING 6.1.3.1 3 election of Personnel

a. Reactor Facility Director 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. The Facility Director must have held a USNRC Senior kcactor Operator license on the AFRR1 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 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 hold a USNRC Senior Reactor Operator license on the AFRR1 reactor. in addition, the ROS shall have 1 year of experience as a USNRC licensed Senior Reactor Operator at AFRR1 or at a similar f2cility before the appointment to this position.

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

d. Additional 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 Oprations

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

1. A licensed Senior Reactor Operator (SRO) on call but not necessarily on site Amendment No. 22 31 l

w__._

+

J L 2. Radiation control technician on call

3. At least one licensed Reactor Operator (RO) or Senior Reactor

.. Operator (SRO)-present -in the control room

4. Another person within the.AFRR1 complex who is able to carry out written emergency procedures, instructions of the operator, or to summon help.. in. case the operator becomes hicapacitated.

b. Maintenance activities that could affect the reactivity of the reactor shall j

be accomplished under the supervision of an SRO.

c. A list of the names and telephone numbers oa the following personnel l

shall be readily available to the operator on duty:

1. Management personnel (Reactor Facility Director, AFRRI Director) 2.- Radiation safety personnel (AFRRI Radiation Protection Officer)

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 involved in the reactor and reactor operations.

6.2 REVIEW AND AUDIT - THE REACTOR AND R ADIATION FACILITY SAFETY COMMITTEE (RRFSC) 6.2.1 COMPOSITION AND OU ALIFICATIONS 6.2.1.1 Comoosition

n. Regular RRF5C Members (Permanent Members)

(1) The following shall be members of the RRFSC:

(a) AFRRI Radiation Protection Officer (b) Reactor "Fac!Iity Director, AFRRI (2) The following shall be appointed.to the RRFSC by the Director,

~

AFRRI:

(a) Chairman as appointed by the AFRRI Directorate.

(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 RRFSC Members (Temporary Members)

(1) Other knov.iedgeable persons.to serve as alternates in item a(2)(b):

above as appointed by the URRI Director.

i Amendment No. 22 32

.. = _... - - -

'(2)? Voting ad hnc members, invited by the Director of AFRRI, to assist

-in review of a particular problem.

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

6.2.1.2 Oualifications The minimum qual:fications for a person on the RRFSC shall be_6:y' ears of professional experience in t.he 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 Comatittee-is directly responsible to the Director, AFRRI. The caramittee 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 T

The RRFSC shall report to the Director, AFRRI, and shall advise the Reactor Facility Director 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 Frecuency The RRFSC or a subcommittee thereof shall meet at least four times a calendar year. The full.RRFSC shall meet at _least semiannually.

6.2.3.3' Ouorum 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.1 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 members present and voting.

33 Amendment No. 22 a

=

6.2.3.5 Minutes Minutes of the previous meeting shall be available to regu!ar members at least I week before a regular scheduled inecting.

6.2.4 REVIEW FUNCTION The RRFSC shall review Safety evaluations for (1) changes to procedures, equipment, or 6 ystems a.

and (2) tests or experiments condue;ed 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 Pat 50.

Proposed tests or experiments that are t,ignificantly different from c.

previously approved tests or experiments, or those that might involve an unreviewed safety question as defined in Section 50.59 of 10 CFR Part 50.

d.

Ptoposed changes in technical specifications, the Safety Analysis Report, or other license conditions.

Vidations of applicable statutes, codes, regulations, orders, technical c.

specifications, license requirements, or of intemal procedures or instructions 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 the cognizance of the RRFSC, but in no case by the personnel responsible for the item audited, annually not to exceed 15 months. A report of the findings and recommendations 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:

Conformance of facility operation to the Technical Specifications and a.

the license.

34 Amendment No. 22

b.

Performance, training, and qualifications of the reactor facility operations staff, c.

Results of all actions taken to correct deficiencies occurring in facility equipment, structures, systems, or methods of operation that affect safety.

d.

Facility emergency plan and implementing procedures.

e.

Facility security plan and implementing procedures.

f.

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

g.

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

6.3 PROCEDURES 6.3.1 Written instructions for certain activities shall be approved by the Reactor Facility Director and reviewed by the Reactor and Radiation Facility Safety Committee (RRFSC). De 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. These activities are as follows:

a.

Conduct ofirradiations and experiments that could affect the operation and safety of the reactor, b.

Reactor staff training program.

c.

St.rvelilance, testing, and calibration of instruments, components, and systems involving nuclear safety.

d.

Personnel radiation protection consistent with 10 CFR 20.

Implementation of required plans such as the Security Plan and c.

Emergency Plan,

f. Reactor core loading and unloading.

g. Checkout startup, standard operations, and securing the facility.

l 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 temporary changes shall be documented and subsequently reviewed and approved by the Reactor Facility Director.

6.4 REVIEW AND APPROVAL OF EXPERIMENTS 6.4.1 Before issuance of a reactor authorization, new experiments shall be reviewed for radiological safety and approved by the following:

a. Reactor Facility Director 35 Amendment No. 22

b. Safety and Health DepartmerK

c. Reactor and Radiation Facility Safety Committee (RRFSC) 6.4.2 Prior a its performance, an experiment shall be included under one of the following types of authorizations:

Special Reactor Authorization for new experiments or experiments not a.

included in a Routine Reactor Authorization. These experiments shall be performed under the direct supervision of the Reactot Facility Director or designee, b.

Routine Reactor _ Authorization for experiments safely performed at least once. These experiments may be performed at the discretion of the Reactor Facility Director an coordinated with the Safety and Health Department when appropriate. These authotizations do not require additional RRFSC review.

c.

Reactor Parameters Authorization for routine measurements of rea e

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

6.4.3 Substantive (reactivity worth more than +/- $0.25) changes to previously approved experiments shall be made only after review by the RRFSC and j

after approval (in writing) by the Reactor Facility Director or designated P

alternate. Minor changes that do not significantly alter the experiment (reactivity worth of less than +/- $0.25) may be approved by the ROS.

Approved experiments shall be carried out in accordance with established procedures.

6.5 REOUIRED 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 withcut authorization by the NRC.

b. The safety limit violation shall be reported to the Regional Administrator of NRC Region I (cr designate); the Director, AFRRI; and the RRFSC not later than the next working day, A Safety Limit Violation Report shall be preparei This report shall be c.

reviewed by 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 recurren:e.

36 Amendment No. 22 l

=.,

,. - -. ~ -.

d. The Safety Limit Violation Report shall be rubmitted to the NRC; the -

-l Dirw:or, AFRRI;.and the-RRFSC within 14 days of the violation _

l 6.5.2 REPORTABLE OCCURRENCES Reportable occurrences as defined in 1.21 (including causes, actual or probable consequences, corrective actions, and measures to prevent recurrence) shall be reported to the NRC. Supplemental reports may be regiired to fully describe the final resolution of the occurrence,

a. Prompt Notification With Written Followuo. The types of events listed.

below shall be reported as soon as possible by telephone and confirmed-by telegraph, mailgram, or similar transmission to;the Regional Administrator 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. 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.

(1) 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.

(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 bases for the Technical Specifications that have or could have permitted reactor operation with a.

smaller margin of safety than in_ the erroneous analysis.

(3) 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 p_lant: 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 applicable reporting requirements of Title 10 of the Code of-Federal Regulations,:the following reports shall be submitted to the Regional-Administrator of the appropriate NRC? Regional Office unless. otherwise noted.-

6.6.1 OPER ATING REPORTS-37 Amendment'No. 22 b

~

a.

Startuo Report: A summary report of planned startup and power escalation testing shall be submitted following (1) receipt of ar operating license; (2) 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 th; operating conditions-or characteristics obtained during the test program and a comparison of these 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 cther 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 three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of power operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

b.

Annual Operating Repor'; Routine operating reports covering the operation of the unit during previous calendar year shall be submitted prior to March 31 of each year, covering the previous calendar year's operation. 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 subraitted reports shall be clear.

Each annual operating report shallinclude (1) A brief narrative summary of (a)

Changec in facility design, performance characteristics, and operating procedures related to reactor safety, that occurred during the reperting' 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, including the reasons and the corrective action taken, if applicable 38 Amendment No. 22

(4) Discussion ot'the major safety related corrective maintenance performeo during the period, including the effects (if any) on the safe operation of the reactor, and the reasons for the corrective maintenance required (S) A brief description of (a)

Each change to the facility to the extent thn it changes a dccription of the facilitv 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 are not encompassed ir. the Safety Analysis Report (6) A summary of the safety eva..ation made for each change, test, or experiment not submitted for Commission approval pursuant to Section 50.59 of 10 CFR Part 50. The summary sha'l clearly show the reason leading to the conclusions that no unreviewed safety question exist:d and that no change to the Technical Specifications vias required.

(7) A summary of the nature and amount of rndioactive 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 diecharge. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed, a statement to this effect is suf'icient.

(a) Liquid Waste (summarized on a quarterly basis)

(i)

Radioactivity discharged during the reporting neriod Total sadioactivity re.':md (in curies)

Concentration limits used and isotopic composition if greater than 3 x 10 6 m crocuries/mi for fission and i

activation products Total radioactivity (in curies), released by nuclide during the reporting period, based on representative isotopic analysis Average concentration at point of release (in microcu-ries /cc) during the reporting period (ii) Total volume (in gallons) of effluent water (including -

diluent) during periods of release (b) Gaitous Waste (summarized on a quarter!> basis)'

39 Amendment No. 22

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

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 ar.y environmental radiological surveys performed outside the facility (9) A list of exposures greater than 25% of the allowed alue (10 CFR

20) received by reactor personnel or visitors to the reactor facility 6.7 RECw 6.7.1 RECOAD3 JO BE RETAINED FOR A PERIOD OF AT LEAST 5 YEARS OR AS REOf f1 RED BY 10 CFR REGULATIONS Operating logs or data that shallidentify a.

(1) Completion of pre-startup checkout, startup, power changes, and shutdown of the ree.ctor (2)

Installation or removal of fuel elements, controi rods, or experiments that could affect core reactivity (3)

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

Principal maintenance operasons c.

Reportable occurrences d.

Surveillance activities required by Technical Specifications -

e.

Facility _ radiation and contamination surveys f.

Experiments performed with the reactor This requirement may be satisfied by the normal operations log book plus (1)

Records of radioactive materia' 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. Fuelinventories and fuel trtnsfers 40 Amendment No. 22

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 equipment, 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 LE AST ONE COMPLETE TRAINING CYCLE a.

Training exams b.

Requalification records p

6.7.3 RECORDS TO BE RETAINED FOR THE LIFE OF THE FACILITY Gasecus and liquid radioactive effluents released to the environs a.

b.

Appropriate offsite environmental monitoring surveys Radiation exposures for all personnel c.

d.

Updated as built drawings of the facility.

41 Amendment No. 22

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