ML20086S662

Proposed Tech Specs Standardizing Type Style & Correcting Typos

ML20086S662
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	11/19/1991
From:	Bumgarner R, Maria Moore ARMED FORCES RADIOBIOLOGICAL RESEARCH INSTITUTE
To:
Shared Package
ML20086S659	List: ML20086S657 ML20086S662
References
NUDOCS 9201030271
Download: ML20086S662 (49)

Text

{{#Wiki_filter:. _ _ _ _ _ _ _ _ _. ~ PROPOSED TECHNICAL SPECIFICATIONS CHANCES PAGE CHANGE Pages 1-41 Because of the standardization of type font throughout the document, some text is now on different page numbers than before and page numbers in the Table of Contents have boon adj usted. Page i In Section 2.0 title chango 'SETTINOS" to ' SETTING' and 8 LIMITS

• to ' LIMIT
• Page 11 In Section 3.2.2 title - change ' Systems" to 'Sys*,am' In titles for Sections 4.2.1

& 4.2.2 - change ' System" to " Systems" Page 2 In Sactio'i 1.10,line 2 - change 'chromal" to 'chromel' Page 3 In Sectic n 1.21a,line 2 - chango " Settings' to ' Setting' Page 4 In Section 1.25,line 1 change 8 electro-mechanical

• to

'electromechs.nical' Page S In titles for Sections 2.0 & 2.2 change ' SETTINGS' to ' SETTING" and " LIMITS

• to ' LIMIT
• Page 11 In Tsble 3,line 3 - change " expect' to 'except*

Page 12 In Section 3.2.7, Specification c,llne 1 - change ' shield" to ' shielding

• In Section 3.3, Specification c,line 2 - change 8.28 to 80.28 Page 13 In Section 3.4 Basis,line 2 - change 'MPC' to ' limits allowed by 10 CFR 20" Page IS In Section 3.S.2 Basis,line 4 - change als tho' to 'Is a*

In Section 3.S.2 Basis,line S - change ' maximum permissible concentration values of 10 CFR 20' to "the concentration limit values in Appendix B, 10 CFR 20' Page 16 In Section 3.6, Specification a,line 3 - change 'II" to *2' Page 18 In Section 3.7 Specification,line 6 - change ' Facilities" to

• Facility" Page 23 In Section 4.2.4, Specification c,line 2 + chango ' door" to

' doors' In Section 4.2.5 Specification,line S change ' occur' to Soccurs' M01030271 911119 PlJR ' ADOC.K 05000170 P P[iR 1 __ j

e s Page 30 In FI ure 1 - change ' Chairman, Safety and Health Dept.' to 'AFRR Radiation Protection Officer" Page 32 In Sections 6.1.3.2(c)(2) and 6.2.1.la(1)(a) - change ' Head, Safety and Health Department' to 'AFRRI Radiation Protvetion 9fficer' change ' calender" to ' calender' Page 33 In Section 6.2.3.2,line 2 and change ' semi annually

• to "somiannually' change ' Office of Inspection and Page 36 In Section 6.5.lb,line 2 Enforcement
• to 'Diviolon of Radiation Safety and Sefer;uards' Page 3g In Section 6.6.1b(7) (a) (1), l i ne 3

change

• MPC" to

" Concentration limits" NOTES:

a. The changes on Pages 13, 15, and 39 replace the toon to be obsolete term

'MPC" with a generic equivalent that is compatible with both the old and the new versions of 10 CFR 20,

b. The changes on pages 30 and 32 clarify that radittlon protection decisions are made by the Radiation Protection Officer.

c, The change on Page 36 reflects the current Region I organizational strut.ture.

d. All other changes are spelling corrections only.

2

i 9 i i TECilNICAL SPECIFICATIONS FOR TiiE i 1 AFRR1 REACTOR FACILITY I November 1991 R e LICENSE R 84 DOCKET 50-170 i i i f 3 = ATTACl(MEllT 2 - ,. _, _ ;.u.,_,

4 i i Reviewed and npproved 10 NOV 1991 Su t - w N ARK L. AIOORE (j Date Reactor Facility Directot Approved for Release -{ne w ( o s,v,.n ~ - l:j ( f f { C gea.- ROBERT L. BUti%ARNER Date Captain, MC, U!iN Director r v s-- +r ,,-,n-

Preface included in this document are the Technical Specifications and the " Bases"for the Technical Specifications. These bases, which provide the technical sup;> ort for the indis idual technical specifications, are mcludedfor information purposes only. They are not part of the Technical Specifications, and they do not consti-tute li'nitations or re.purements to winich the licensee musr adhere. I l i l l l l l l l

4 5 TECllNICAL SPECIFICATIONS FOR Tile AFRR1 REACTOR FACILITY LICENSE NO. R 84 DOCKET # 50170 TABLE OF CONTENTS l.0 DEFINITIONS hge 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 l 1.16 Pulse Mode 3 l 1.17-Reactor-Operation 3 l 1.18 Reactor Safety Systems 3 I 1.19 Reactor Secured 3 L20 Reactor Shutdown 3 1.21 Reportable Occurrence 3 1.22 Safety Chane.el 4 l 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 l 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINO 2.1 ~ Safety Limit. Fuel Element Temperature 5 1 2.2 Limiting Safety System Setting for Fuel Temperature-5- 3.0 LIMITING CONDITIONS FOR PPERATIONS 3.1 Reactor Core Parameters -7 3.1.1 Steady State Operation-7 i: = .s.. .m

i e f.agt 3.1.2 Pulse hiode 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 Systern 11 i 3.3 Coolant Systems 12 l 3.4 Ventilation System 13 3.5 Radiation hfonitoring System and Effluents-13 3.5.1-hionitoring System 13-l

3.5.2 Effluents

Argon +41 Discharge Limit 15 ~ 3.6 Limitations on Experiments 16 3.7 System biodifications 18 3.8 ALARA 18 l \\ 4.0 SU R VEILLANCB REQUIREhiENTS 4.1 Reactor Core Parameters 20 4.2 Reactor Control and Safety Systems 21 l 4.2.1 - Reactor Control Systems 21

4.2.2 Reactor Safety Systems 21

-l 4.2.3 Fuel Temperat'ure 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 hionitoring System 25 5,0 DESIGN FEATURES o - 5.1 Site and Facility

Description:

26-5.2 Reactor Core and Fuel -26 ~ 5.2.1 Reactor Fuel 26 1 5.2.2 -Reactor Core - 27 5.2,3 Control Rods 28-5.3 - Special Nuclear hiaterial Storage _ .29 - 11

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6.0 ADMINISTR ATIVE CONTROLS Page 1 6.1 Organization 30 6.1.1 Structure 30 6.1.2 Responsibility 31 6.1.3 Staffing 31 { i 6.1.3.1 Selection of Personnel 31 6.1.3.2 Operations 31 ] 6.1.4 Training of Personnel 32 l 6.2 Review and Audit The Reactor and Radiation Faellity Safety l Committee (R R FSC) 32 l 6.2.1 - Composi' ion and Qualifications 32 6.2.1.1 Composition 33 l 6.2.1.'t Qualifications 33 6.2.2 Function and Authority 33 6.2.2.1-Funetlon 33 I 6.2.2.2 Authority 33 6.2.3 Charter and Rules 33 f 6.2.3.1 Alternates 33 6.2.3.2 Meeting Frequency 33 l 6.2.3.3 Quorum 33 6.2.3.4 Voting Rules 33 6.2.3.5 Minutes 34 l 6.2.4 Review Function 34 .6.2.5-Audit Function 34 l 4 6.3 Procedures 35 6,4 Review and Approval of Experiments 35 6.5 Required Actions 36 i 6.5.1 Actions To Be Taken in Case of Safety Limit Violation 36 6.5.2 Reportable Occurrences 37 6.6 Reports 37 6.6.1 - Operating Reports - 37 6.7 Records - 40 l 6.7.1 Records To Be Retained For A Period of At least Years or As Required by 10 CFR Regulations - 40 6.-7,2 Records To Be Retained For At least One Complete. Training Cycle _ 41 6.7.3 Records To Be Retained For The Life of The Facility = 41 ill veg c ..wo +c.,m.,, .+.m % +w e e me e. 3

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1.0 DEFINITION.S 1.1 ALARA [ The ALARA program (As Low As Reasonably Achievable) is a program for maintaining occupational c, 'sures to radiation and release of radioactive effluents to the environma - low as reasonably achievable. I 1.2 CllANNEL CAlllllMTjpli 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 r, hall encompass the entire channel including equipment activation, alarm, or trip, and shall be deemed to 1 include a channel test. 1.3 CllANNEL CllECK A channel check is a verification of acceptable performance by observation of channel behavior. 1.4 Cll 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 i vel less than 100 watts, with the fuel and bulk water temperature equal and less than 40"C. 1.6 CQ1Ui ORID POSITION The core grid position refeis to the location of a fuel or control element in the grid structure. 1.7 IlXPERIhiENI i Experiment shall mean (a) any apparatus, device, or material that is not a normal part of the core or experimental fr,cilities, 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. 1A . EXPERIhiENTAL FACILITJJiS The experimental or exposure facilities associated with the AFRR1 TRIGA reactor shall be a. Exposure Room # 1 i b. Exposure Room #2 - 1 ~ -... -

NOTE: Exposure facilities protective barriers shall be differentiated from the primary protective battler (fuel element cladding) for purposes of placement of experiments within these barriers. c. Reactor Pool d. Core Experiment Tube I e. Portable Beam Tubes f. Pneumatic Transfer System

g. Incore Locations 1,9 ElJEL IllFhiENT r

A fuel element is a single TRIGA fuel rod, or the fuel portion of a fuel follower control rod. 1.10 INSTRUhiENTED ELEhtENT An instrumented element is a special fuel element'in which sheathed chromel/alumel or equivalent thermocouples are embedded in the fuel. 1.11 LlhilTING SAFETY SYSIFhi SETTWO Limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions. 1.12 hiEASURED VALUE A measured value is the magnitude of a variable as it appears on the output of a measuring channel. 1.13 AIEASURING CllANNEL 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-The individual has been specifically designated and the operator knows of a. the designation;

b. The individual keeps the operator' posted as to_his/her whereabouts and ~

l' telephone number; and

c. The individualis 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. 2

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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. The reactor may be pulsed from a critical or suberitical state. 1.17 REACTOR OPERATION Reactor operation is any condition whereln the reactor is not shui down., or any core maintenance is belag performed, or there is movement of any control rod. 1.18 REACTOR SAFETY SYSTEhiS Reactor r.afety systems are those systems, including their associated input circuits, that are designed to initiate a reactor scram for the primary purpose of protecting i the reactor or to provide !nformation that-may require manual protective action to be initiated. l.19 REACTOR SECUREQ The reactor is secured wnen 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.

No work is in progress involving in core fuel handling or refueling c. 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 50.50 (or greater) shutdown margin with the most - reactive control rod removed. 1.20 REACTOR SilDTDOWN e The reactor is shut down when the reactor is suberitical by at least 50.50 of reactivity. 1.21 REPORTA11LE OCCURRENCE A reportable occurrence is my 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 Setting.1

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

Malfunction of a required reactor or experiment safety system component c. that could render the system incapable of performing its intended safety function unless the malfunction'is discovered during tests. - i 3 _ =

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d. Any unanticipated or uncontrolled positive change in reactivity greater than $1.00. e. An observed inadequacy in the implementation of either administratiw or procedural controls, so that the inadequacy could have caused the existence j or development of a condition that coald result in operation of the reactor in a manner less safe than conditions covered in th Safety Analysis Report (SAR). f. The release of11ssion products from a fuel element through degradation of j the fuel cladding. Possible degradation ruay be determined through an increase in the background activity level of the reactor pool water. I 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 i Regulations (10 CFR 20), or these technical specifications. ] 1.22 SAFETY CilANNEL A safety channel is a measuring channel in the reactor safety system that provides a reactor protective function. l 1.23 SAFETY L1Mll Safety limits are limits on important process variables thio. are found to be necessary to reasonably protect the integrity of certain physical barriers that gt.ard apinst 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 subcritical by means of the control and safety systems, starting from any permissible operating l conditions, and that the reactor will remain suberitical without further operator action. 1.25 STANDARD. CONTROL ROD 1 A standard control rod is a control rod-having an electromechanical-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 the steady state-operation of the reactor either by manual opesation 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 capalilities that can be rapidly ejected from the reactor core to produce a pulse. It is activated by applying compressed air to a piston. 4

2.0 SAFETY LIMIT AND Llh1111NH_S.AFETY SYSTEM SETT1MG 2.1 SAEliTY Llh11L_ E11EL ELEMENT TEhiPERATU1W MulinitillE This specification applies to the temperature of the reactor fuel. Objectiv.c. 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. Sneciucautti. The maximum temperature in a standard TRIGA fuel element shall not exceed 1000 C under any condition of operation. Huis The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well sulted as a single specification, especially *,ince it can be measured. A loss in the integrity of the fuel eternent cladding could arise frorn a - buildup of excessive pressure between the fuel. moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caur.ed by the presence of air, fission product gases, and hydrogen from the dissociation of thn hydrogen and zirconium in the fuel moderator. The magnitude of this pressure is dete mined by the fuel moderator ternperature 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 piessure from the dissociation of zirconium hydride will remain below the ultimate stress, provided that the temperature of the fuel does not exceed 1000 C while immersed in water. 4 2.2 LlhnIIND. SAFETY:,_ SYSTEM Sji1IING EQlLEUEL TEATERAllum AplicAh3it This specification applies to the scram settings that prevent the safety limit from being reached. QbitsdYt. The objective is to prevent the safety limit from being reached. SucsHicadon. There shall be two fuel temperature safety channels. The limiting safety system setting for these estrumented fuel elements' temperature shall not exceed 600 C. -5

One channel shall utilize an instrumented element in the "lA" ring, and the second channel shall utilize an instrumented element in the "C" ring. Ilusih The limiting safety system setting is a trmperature which, if exceeded, shall cause a reactor sciam to be initiated, preventing the safety limit from being execeded. A setting of 600"C provides a safety margin of at least 400"C for standard TRIGA stainless steel 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 actuallocation of the thermocouple. If the thermocouple element is located in the hottest 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 temperature, the measured temperature will differ by a greater amount from that actually occurriur at the core hot spot. To lessen this difference, the requirement is to locate the element in the hottest region of the core. These marg!ns are suffic!cnt to account for the remaining uncertainty in the accuracy of g the fuel temperature measurement channel and any overshoot in reactor power resuhing from a reactor transknt during steady state mode operation, in the pulse mode of operation, the same limiting safety system setting shall apply, llowever, the temperature channel will have no effect on limiting the peak power generated, because of its reladvely long time constant (seconds), compared with the width of the pulse (milliseconds). In this mode, however, the temperature trip will act to reduce the amount of energy generated in the entire pulse transient, by cutting the " tail" of the power transient if the pulse rod remains stuck in the fully withdrawn position with enough reactivity to exceed the temperature limiting safety system setting. 4 4 6.

3.0 Lih1LTING CON 1'lTIONS FOR OPER ATIONS 3.1 REACTCRJOIE EARAMETERS 3.1.1 SIlhW'o7/\\TE OPERATION Apht,iL%. Thb spW&ation applies to the maximum reactor power attained during stedy fdht0 Opo,tation. Oldusha To assure that the reactor safety limit (fuel temperature) is not exceeded, and to p<ovik for a set point for the high flux limiting safety systems, so that auwntatAc pmtective action will prevent the safety limit from being reached during steady state operations. S2ttifktdon The reactor steady state power level shall not exceed 1.1 megawatts. The normal steady state operating power limit of the reactor should be 1.0 ) megawatt. For purposes of testing and calibration, the reactor may be operated at power levels net to exceed 1.1 megawatts during the testing period. Iltadt Thermal an(, 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 ILULSL.AfDDE OPE R ATlON Applicahility, r) This specification applies to the maximum thermal energy produced in the renctor as a result of a prompt critical inseition of' reactivity. Obiettive. The objective is to assure that the fuel temperature safety limit will not be exceeded. Spulficationc The maximum step insertion of reactivity shall be 2.8% Ak/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% Ak/k results in a maximum average fuel temperature of less than 550 C, thereby staying within the limiting safety. -7 I ,..._m

f 5cttings that protect the safety limit. The 50"C margin to the Limiting Safety System Setting and the 450"C margin to the safety limit amply allow for uncertainties due to extrapolation of measured data, accuracy of measured data, and location of instrumented fuel elements in the coie. 3.1.3 REACTIVITY l_lh11TATIONS Applicabillt These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods and experiments. They apply for all modes of operation. WeC11YE. 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. Sgrifications

a. The reactor shall not be operated with the maximum available execu 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 withdrawn or removed shall be

$0,50(0.35% Ak/k) for any condition of operation, Buh

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

b. The shutdown margin assures that the reactor can be shut down from any -

operatiny condition even if the highest worth control rod remains in C e 4 fully withdrawn position or is completely removed. 3.1,4 SCRAM TIME Applicabilig The specification applies to the time required to fully insert any control rod to a full down poshion from a full up position. Objective The objective is to achieve rapld shutdown of the reactor to prevent fuel damage. Sgeification - The time from scram initiation.to the full insertion of any-control roit from a full up position shall be less than 1 second. 8

ilaib i This specification assures that the reactor will be prompt'e shut down when a scram signalis initiated. Experience and analysis indicah : hat, fc,r the range of transients for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor. l 3.2 REACTOR CONTROL AND SAFliTX_SYSTJLhtS f 3.2.1 ILEACTOR CONTROL SYSTEhi Applicability. This specification applies to the channels monitoring the reactor core, wh!ch must provide information (o the reactor operator during reactor operation. Objective The objective is to require that r.ufficient information be available to the operator to assure safe operation of the reactor. ~ Specification The reactor shall not be operated unless the masuring channels listed in Table 1 are operable. TABLE 1. MEASURINO CHANNELS hijnimmp Number Onerable in Effectl"e hiode Steady State Pulse Fuel Temperature Safety Channel 2 2 l l Linear Power Channel 1 1 Log Power Channel 1 0 i High Flux Safety Channel 2 1* Pulse Energy Integrating Channel -0 -1* . _-.a (* NOTE: Same channel as linear power in this mode) l lhib Fuel temperature displayed at the control console gives continuous information < a this paretten, which has a specified safety limit The power - level channels assure.that radiations indicating reactor. core parameters are adequately n:nnitored 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 tehted to the fuel temperature. 9 i n- ,, +. -,., ~

j 3.2,2 RFACTOR SAFETY SYSTEh! ApplicabillE This specification applies to the reactor safety system. Oldeclits. I The objective is to specify the minimum number of reactor safety system channels that must be operable for safe operation. Specificaflen. The reactor shall not be operated unless the safety systems described in i Tables 2 and 3 are operable. l TA13LE 2. MINIMUM REACTOR SAFETY SYSTEM SCRAMS - - M aximum Minimum Number in Mode Channel Set Point Steady State Pulse r Fuel Temperature 600 C 2 2 Percent Power, liigh Flux 1,1 MW 2, -Console Manual Secam llar Closure switches 1 1 liigh Voltage Loss to a Safety Channels - 20% loss 2 1 [ Pulse Time 15 seconds 0 1 Emergency Stop Closure switch 1 1 i (1 each exposure room, 1 on console); Pool Water Level 14 feel from top 1 1 of core Watchdog-(DAC to CSC) Or digital console. 1 1 lhubi 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 1 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 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 will reduce to a low level after pulsing.- 10 . -. - - -. ~, _..

The emergency stop allows personnel trapped in a potentially hazardous 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 i rc, ult in a reactor shutdown. The watchdog scram will insure adequate dmmunication between the Data Acquisition Computer (DAC) and the Control System Computer (CSC) units. l TABLE 3. MINIMUM REACTOR SAFETY SYSTEM INTERLOCKS Effgr11ve Mods Action Prevented Steady State Pulse Pulse initiation at pt,ver levels greater X than 1 kilowatt Withdrawal of any control rod except transient X Any rod withdrawal with count tr.te in X X operational channel below 0.5 eps Simultaneous manual withdrawal of two X standard rods Dash The interlock preventing the initiation of a pulse at a critical level above-1 3 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 raic to be seen by the operationa.1 channels insures sufficient source neutrons to bring the reactor critical under controlled-conditions The interlock that prevents ".he simultaneous manual withdrawal of two standard control rods limits the amovnt of reactivity added per unit-

time,

. 3.2.3 FACILITY INTERLOCK SYSTEh1 Applienbility i This specification applies to the interlocks that prevent the accidental exposure of an individual in either exposure room, DMeclim-The objective is to provide sufficient' warning and interlocks to prevent movement of the reactor core to the exposure room in which someone may 4 vv-'-W4ya --y nyv~ry- ?^N P*"$tt-+

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e be working, or prevent 'he inadvertent movement of the core into the lead shield doors. Sptdfication. Facility interlocks shall be provided so that a. The reactor enimot be operated unless the shielding doors within the i reactor pool are either fully opened or fully closed.

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

c. The lead shielding doors cannot be opened to allow movement into the I

exposure room projection unless a_ warning horn has sounded in that exposure loom, or unless two licensed operators have visually inspected =i the room to insure that no personnel remain in the room prior to securing the plug door. l L Ihuh These interlocks prevent the operation and movement of the reactor core into an area until there is assurance that inadvertent exposures will be eliminated. l 3.3 COOL ANT SYSTEMS i Annlicabilltv f This specification refers to operation of the reactor with respect to temperature and I 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.- e. To help preclude corrosion of fuel cladding and other components in the primary system. Sptcl0tallant I a - The reactor shall not be operated above a thermal 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-C

~ .2 micrombos/em (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 conductivit of the bu3 water is greater than 5 micrombos/cm (or less than 0.2 x 10[ohris cm resista c. a averaged over 1 week. L 12 vg ausrerg-r k-W t weep ?9t-gi- - gpr

i Ilmh Manufacturer's data state that the resins in the water purification system break down with sustained operation in excess of 6ffC. The 2 micrombos/em is an acceptable level of water contaminar.is !n 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. 4 3.4 YENTILATION SYSTEh1 Apalisablills This specification applies to the operation of the facility ventilation system. Dbjectlye The objective is to assure that the ventilation system is operable. SEtsificallan. The reactor shall not be operated unless the facility ventilation system is operable, except for periods of time necessary (up to 48 hours) to test or permh minor repair of the system. In the event of a significant release of airborne radioactivity in the reactor room, the ventilation system to the reactor room shall be secured via closure dampers autematically by a signal from the reactor deck air particulate

monitor, llash During normal operatior, of th: ventilation 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 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 i materials. Moreover, radiation monitors within the buildingindependent 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 - B ADI ATION-MONITORING SYS*[EM AND EEFLUENTS 3.5.1 MONITORING SYSTEM - Aonlicability This specification applies to the functions and essential components of the area radiation monitoring equipment and the system' for continuouslyL - monitoring radioactivity and radiation levels,.which must be a.ailable < luring reactor operations.' 13- -i r

Oldetths. The objective is to assure that adequale radiation monitoring equipment and radiation infornation are available to the operator to assure safe operatloa of the reactor. Saccification The reactor shall not be operated unless the following radiation monitoring systems are operable: Arca Radiation hf onitoring System. The area radiation monitoring a. (ARht) systein 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. 7

b. Gas Stack hionitor. 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 seaHug dampers, causing reactor room-isolation.

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

TABLE 4.1.OCATIONS OF RADIATION MONITORING SYSTEMS Location of Alarm Readout Monitor _ (A = Audiblei V = Visual)' Locat_lon AlBl. R1, Reactor Room Control Room A&V Control Room R2, Reactor Room Control Room V - Co itrol Room E3, Exp. Room 1 Area ControlRoom V Control Room l E6, Exp, Room 2 Area Control Room V Control. Room L-(IShi. - Reactor exhaust - - Control Room V-Control Room i AB.i. Reactor room Control Room A&V Control Room - 4 DJtsh This system is intended t_o characterize the normal operational radiological-environment of the facility and to aid in evaluating any abnormal operations or conditions. The radiation monitors provide information to the operating - 1 -14 s 4..... -,%..%-mw,---~,-.. ,J_-_,-.-.v-- v c.-%. n ,w ,.w-.,

} a personnel of any existing or impending danger from radiation, to give sufficient time to evacuoe the facility and take r.ecessray steps to prevent the spread of radioactivity to the surroundings. The automatic slosure of the-ventilation system dampers prov:Jes reactor room isolation from the outside environment, m the event of r.Irborne radiot:tivity within the reactar room from fission products dec..;/.

3.5.2 EFFLUENTS

AROQN 41 Q1SC11ARGFL.L1hliT Applicabilne This specification applies to the concentiation of argon-41 that nay be discharged from the TRIGA reactor facilky. Obiective To insure that the het.lth and safety of the public are not' endangered by the discharge of argon 41 trom thc TklOA reactor facility. Specification a. An environmental radiation-monitorhtg program shall be maintained to determine effects of the facility on tha environs, b. If a dosimeter reading foi any designated environmental monhoting station indicates that a probable exposure of 400 millitem above background has been reached during the year as a result of reac.or-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 raonth for the remainder of the calendar year. If a dosimeter reading for any designated environmental monitoring - c. station indicates that an exposure of 500 millirem above backgro.md has been reached during the year as a asult of reactor oper0tions; raacGar-operations that generate and releas measurable quan:ities of argon-41 s shall be ceased for the remainder of the calendar yeu, Basis Since argon-41 does not represent an uptake or hinaccumulation problem,- 'g only the direct exposure modality is pertinent with~ regad to limiting reactor operations. Since direct plume shine may be more controuing than immersion conditions, cumulative exposure is a' more appropriate - quantification of this limit than the concentration limit values in Anpendix B, 3 10 CFR 20 1 i 15

3.6 L1hilIATIONS ON EXEEB1MEEl.S Applicability. This specification applie:, to experiments installed in the reactor and its experimental facilities. v OldEthA The objective is to prevent damal,e to the rea: tor or excessive release of radioactive [ materials in the event of ar experiment malfunction, to that airborne i concentrations of activit" averaged over a yeai do not exceeu 10 CFR 20, Appendi': B. 3 Spsifinaimni The foPowing limitations shall apply to the irradiation of materials (other than air): It the possibility exists that a release of radioactive gases or aerosols may a. occur, the amount and type of materic irradiated shall be limited to assure the yearly compliance with Table 2, Appendix B, of 10 CFR 20, assuming that 100% of the gases cs 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 inv'.utory is not greater than 5 millicurin. c. Known explosive matenals shall not be irradiated in the reactor in quantities greater than 25 milligrras. In addition, the pressure produced in the experiment cc'itainer upm. detonation of the explosive shall have been determined experimentally, or by calculations, to be less than the design pressure of tne container. d. Samples shall be doubly contained when release of the contained material could cause corrosion of the experimental facility, The sum of the absolute reactivity worths of all experiments in the reactor e. and in the associated experimental facilities shall not exceed $3.00 (2.lre Ak/k). This inchides the total potential reactivity insertion that might result from experiment mahunction, accidental experiment tiooding or voiding, and accidental removal or insertion of experiments. f. In ulculations regarding experiments, the following assumptions shall be made: 1) If the effluent exhausts through a filter installation designed for gicater than 99% efficiency for 0.3 micron particles, at least 10% of the particles produced can escapc. 16

4 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 107c of the vapor is permitted to e:, cape. g. If a capsule rails 1.nd releases materials that could damage the reactor fuel or structure by corrosion or other meaas, physical inspection shall be performed to determine the consequences and neco for corrective action. The results of the insoection and any corrective action taken shall be reviewed by the Reactar Facility Director, and shall be deteimined to be satisfactory before operation of the reactor is resumed, h. All experiments placed in the reactor exposure environment shall be either firmly secured c c 3 served by a Senior Reactor Operator for meenanical stability, to insure mat unintended movement will not cause an unplanned reactivity change or physical damagv. All operations in any experimental area shall oc supervised by a member of the reactor operations staff. Iluh This specification is intended to provide assurance that airbotne activities n. in excess of the limits of Appendix B c.f 10 CFR 20 will not be released to the atmosphere outside the facility boundary,

b. The 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 particchite !odine trepped 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 un additional safety factor where i

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 ccid 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 relcase of fission products to the environment is limited. h. All experiments placed in the reactor environment shall be either firmly secured or observed for mechanical stability to insure that unintended movem:nt will not cause an unplanned reactivity change or physical damage. 17

i 3.7 SYSTEM MODIEJfCAllONS l Applicacility. This specification applies to any system related to reactor safety. D_hitcliEc. The objective is to verify the proper operation of any systen modification related to reactor safety. Sassificatina. Any additions 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 sately 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 Safcty Committee. A system shall not be considered operable until after it is successfully tested. 7 Ihsis This specification is related to changes in reactor systems that conld 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 criteria. 3,8 ALARA ApplifADili1Y This specification applies to all reactor operations that could result in.significant personnel exposures. Objective To maintain all exposuies to ionizing. radiation to'the staff and the general public as low as is reasonably achievable, Snecificalica. 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 (mvironment, or both. Ilasir Experience has shown that experiments and operational requirements can, in many L 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 l radioactive effluents or staff radiation exposures. Similarly, overall reactor l l l 18

g. scheduling achieves significant reductions in staff exposures. Consequently, ALARA must be a part of both the overall reactor scheduling and the detailed - experiment planning. 9 6 ( f \\ I 19 l

j 4.0 SURVEILLANCE REOUIREMENTS l 4.1 REACTE_CHER_ PAR AM EIEILS Acolicability. These specifications apply to the surveillance requirements for_ reactivity control of experiments and systems affecting reactivity, Obiective f 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 fir >t 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 plannedi a functional-performance check of the transieat (pulse) rod 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, cleaned, and lubricated as necessary, The core excess reactivity sh'all be measured at the beginning of each day of _ e. operation involving the movement of control rods, or prior to each continuous operation extendirg more than a day, f. The power coefficient of reactivity a: 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 expetiments 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 1 20

coefficients allow the operator to accurately predict the reactivity necessary to achieve required power levels, 4.2 EEACTOR CONTROL ARD SAFETY SYSIEhiS 4.2.1 BJ! ACTOR CONTILQL3Y3ffdjS Applicability _ ( These specifications apply to the surveillance requirements for reactor cortrol systems. Objective N The objective is to verify the operability of system components that affect the safe and proper control of the reactor. Specifica.ticit ) 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 B FJCTOR SAFETY SYSTEMS Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibra' ions of the reactor safety systems. Ob,iecti';t The objeeiive is to verify the performance and operability-of the systems and components thet a e directly related to reactor safety. SnecificatioJ13. A check of the scram function of the high-flux safety channels shall be a. mr.de on each day that the reactor is to be operated. b.- A Channel test of each of the reactor safety system chr.nnels for the intended mode of operation shall be performed weekly, whenever : operations - are pir.nned, Channe! calibration shall be made of the power level monitoring-c. annually, at intervals not to exceed 15 months. Basis TRIGA system components have operational proven reliability. Daily checks insure accurate scram functions. Weekly channel testing is sufficient 21

I l to insure the detection of possible channel drift or other possible deterioration of operating characteristics. The channel checks will assur e 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 E.U EL TEM PERK 1J).RE These specifications apply to the surveillance requirements for the safety channels measuring the fuel temperature. DMtc1b'.t To insure operability of the fuel temperature measuring channels. Sge;Jications A check of the fuel temperature sciams shall be made on each day that a. the reactor is operated. A calibration of the fuel temperature-measuring channel shall be made c. annually, at intervals not to exceed 15' months, - A weekly channel test shall be performed on fuel temperature-measuring c. channeh, 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 thermo- -couple measurements have been sufficiently reliable as an indicator of fuel temperature with proven reliability. The weekly channel test assures i operability and indication of fuel temperature. The daily scram check assures - scram capabilities. 4.2.4 FACIIITY I'NTERLOCK SYSTEM Anolicability, This specification applies to the surveillance requirements that insure the integrity of the facility interlock system. Q.DitCliEt To insure performance and operability of the facility interlock _ system. Snecification Functional checks shall be made annually, but not to exceed 15 months, to insure the following: 22

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.

4 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 verify operation of the interlock system. Experience at AFRRI indicates that this is adequate to insure operability. 4.2.5 REACTQILEITL ELEMENTS AnolicabHiry This specification applies to the surveillance requirements for the fuel elements. DMeslin. The objective is to verify the integrity of the fuel element cladding. Specifications All the fuel elements present in the reactor core, to include fuel follower control rods, shall be inspected for damage or thrioration, 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. Fuel elements in long term storage need not be-measured until returned to core; however. fuel elements routinely moved to l temporary storage shall be measured every 500 pulses of insertion greater l than $2.00 or annually (not to exceed 15 months), whichever occurs first Basis The frequency of inspection and measurement 'is based on the parameters most likely to affect the fuel. cladding of a pulse reactor, and the utilization of fuel elements wnose characteristics are well known.- The limit of transverse bend has-been shown.to result in no difficulty in - disassembling the core, Analysis of a. worst case scenario in which two adjacent fuel elements suffer sufficiently severe transverse bends to result in-the touching of the fuel elements has shown'thet no damage to the fuel elements will result via a hot spot or any other known mechanism. 23

4.3 C00LAN.T_5XSTEM S Amliabilhy - This specificatien applies to the surveillance requirements for monitoring the pool water and the water conditioning system. .QhierliYt 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. Sguifications 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.

D2 sis Based on experience, observation at these intervals provides acceptable surveillance of limit that assure that fuel clad corrosion and neutron activation of dissolved materials will not occur. 4.4 VENTILATION SYSTEh1 Amlicabilits This specification applies to the facility ventilation system isolation; Objective The objective is to assure the proper operation of'th'e ventilatida system in - controlling the release of radioactive __ material into the unrestrieted environment. Specification. The operating mechanism of the positi_ve scaling dampers in the reactor room ventilation system shall be verified to be operable and visually inspected at least . monthly (interval not to exceed six. weeks). Halls. I

lxperience accumulated over years of. operation has demonstrated that;the tests of the ' entilation system on a monthly basis are sufficient to assure proper operation -

v of the system and control of the release of radioactive material. i. 24l L . = -

3 1 4.5 R A1)G_ TION-MONITORIR.G. SYSTEM - i Annlicability This specification applies to surveillance requirements for the area radiation monitoring quipment and the air particulate monitoring system, Objective i The objective 1. e 'he, r6.tica-monitoring equipment is operating and to verify the app "" % gs, Snecification The area radiation moni. , the sit particulate monitoring system shall be channel tested qua.,..y, t u,. ttervals 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 annuttily, not to exceed 15 months, llnbi 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 charteteristics over a long time span. -Annual calibration insures that units are within the specifications demanded by extent of use, e ~ \\ 3. q i ~ 5. 2 ' l

j i I 5.0 DESIGN FE ATURES j 5.1 SLIE AND FACILITY DESC11EllOR Applicability This specification applies to the building that houses the reactor. Obittlim The objective is to restrict the amount of radioactivity released into the environment. Spscifications

a. The reactor building, as a structurally independent building in the AFRRI complex, shall have its own ventilation system branch. The effluent from the i

reactor ventilation system shall exhaust through absolute filters to a stack having a minimum elevation that is 18 feet above the roof of the highest building in the AFRRI complex. b. The reactor room shall contain a minimum free volume of 22,000 cubic feet,

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

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 Icactor room air particulate monitor.

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

lhiis 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 ventilation system. Building construction and gaskets around doorways he!p restrict ' leakage of air into or out of the reactor room. The stack height insures an adequate dilution of effluer.ts well above ground level. The separate ventilation system branch insures a dedicated air flow system _for reactor effluents. 5.2 EIMf'TDR-COR E AN D.F U E L 5.2.1_ R E ACTOR. F U EL Applicabiluy These specifications apply to the fuel elements, to include fuel follower control rods, used in the reactor core. 4 26

Obhrlitt 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 and nuclear characteristics, and (2) assure that the fuel elements used in the core are substantially those analyzed in the Safety Analysis Report, o Spfgifications The individual nonitradiated 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-7irconium atom ratio (la the UH:_): Nominal 1.7 H atoms to 1.0 Zr atoms with a range between 1.6 and 1.7. c. Cladding: 304 s'ainless steel, nominal 0.020 inch thick d. Any burnable poison used for the specific purpos: of compensating for fuel burnup e long term reactivity adjustments shall be an integral part of the manufactured fuel elements. Hasis 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 i an increase in loading would result in an increase in power density of less 3 than 6%. An increase in local power density of 6% in an individual fuel element reduces the safety margin by 10%, at most. The hydrogen to-zirconium ratio of 1.7 will_ produce a m'aximum 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 j than an 8.5 weight percent standard TRIGA 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-ring 5.2.2 REACTOR CORE Applicabilitv These specifications apply to the configuration of fuel and in. core experiments. 27

Obitclire. 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. SNdficalintls. i

a. The reactor core shall consist of U.cndard TRIG A reactor fuel elements in a close packed ruray 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, c. The core shall be cooled by natural convection water flow. d. In-core experiments shall not be placed in adjacent fuel por.itions of the B ring and/or C ring. c. Fuel elements indicating an clongation preater than 0.100 inch, a lateral q bending greater than 0.0625 inch, or sig.ificant visible damage shall be considered damaged, and shall not be ured in the reactor core. Ilusi'i Standaro TRIGA cores have been in use for years, and their safe operational characteristics are well documented. Experience with TRIGA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified to (a) assure that the cladding material w.ill not-be subjected to stresses that could + cause a loss of integrity in the fuel containment, ar.d (b) assure adequate coolant flow. 5.2.3 CONTROL RORS Applicability These specifications apply to the control rods used in the reactor core. DhlectiYr. i 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. Smcifications a. The standard control rods shall have scram capability, and shall contain l borated graphite. B4C powder, or boron and its compounds in solid form L as a poison in aluminum or stainless--steel cladding. These rods may have L an aluminum, air, cr fuel follower. If fuel followed, the fuel region will conform to the Specifications of 5.2.1. P 28

b. The transient control rod shall have scram capability, and shall contain borated giaphite, B4C powder, or boron and its compounds in solid form as a poisen 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 satisfied by using neutron-absorting borated graphite, B4C powder, or boron and its compounds. These materials must be contained in a suitable cladding material, such as aluminum or stah:less steel, to insure mechanical stability during rnovement and to isolate the poison from the pool water environment. Scram capabilities are provided by the rapid insertion of the control rods, which is the primary operr.tional safety feature of the reactor. The transient contiol rod is designed for use in-a pulsing TRIGA reactor. 5.3 SPECIAL NUCl,EMLMATERIAL STORAGE Apalitahility_ This specification applies to the storage of reactor fuel at times when it is not in the reactor core. OMettirt The objective is to assure that stored fuel will not become critical and will not reach an unsafe temperature. Sp_tcification All fuel elements not in the reactor core shall be stored and handled in accordance with applicable regulations. Irradiated fuel clernents and fueled devices shall be stored in an array that will per.mit 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 subcritical under all conditions of moderation. Basis The limits imposed by this specification are conservative and_ assure safe storage and handling. Experience shows that appmximately 67 fuel elements are required, of the desiga used at AFRRI, in a closely packed array to achieve criticality. 1 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. 1 29

. ~.... -. 6:0 ADMINISTRATIVE CONTROLS 6.1 OR G AN12ATIOli 6.1.1 STRU CTUR E 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 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. Directer, AFRR1 7 c- ' safey. Reactor and Radia:Aon F Tacility Safety Committee Radiati n rotection Ofticer -l -l t cimman. - I Radiacon .l { Sources Dept. t u,, Metsory e l I t .l l Rt.ector Facility Director ..............J -I ~ Reactor Operations Supervisos l.- ReactorOperudons Staf(* [ Figure 1. Organization of Personnel.for1 Management and Operation of the - AFRRIL Reactor Facility. l

• Any_ reactor staff member has access to the Director for-matters of safety:

30-

6.1.2 EliSPONSIBILITY The Director, AFRR1, shall have license responsibility for the reactor facility. The Reactor Facility Director (RFD) shall be responsible for administration and operation of the Reactor Facility and for determination of applicability of procedutes, experiment authorizations, maintenance, and operations. The RFD may designate an individual who meets the requiremeMs of Section 6.1.3.1.a to discharge the 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 Selection of Personnel

a. Reactor Facility Director At the time of appointment to this position, the Reactor Facilhy 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 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 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 AFRRI reactor, in addition, the ROS shall have 1 year of experience as a USNRC licensed Senior Reactor Operator at AFRR1'or_ at a similar - facility 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 diploma 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 Ooerations

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

1. A licensed Senior Recctor Operato'r (SRO) on call but not necessarily ca. site 31-

i

e

2. Radiation controltechnician 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 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 reactivity of the reactor shall be accomplished under the supervision of an SRO.

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

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

2. Radiation safety personnel (AFRRI Radiation Protection Officer)

3. Other operations personnel (Reactor Staff, ROS) 6.1.4 T.B AINING OF PERSONNEL A training and retraining program will be maintained, to insure adequate levels of proficiency in per;ons involved in the reactor and reactor operations.

6.2 REVIEW AND AUDIT - THE REACTOR AND RADIATION FACILITY SAFETY COhlhflTTEE (R R FSC) 6.2.1 C.Ohf POSITION AND OU AI IFICATIONS - 6.2. L 1 Comnoskinn

a. Regular RRFSC Members (Permanent Members)

(1) The following shall be members of the RRFSC: (a) AFRRI Radiation Protection Officer (b) Reactor Facility 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 Operanons Specialist, or a Health-Physics Specialist,

b. Special RRFSC Members (Temporary Members)

(1) Other knowledgeable persons to serve as alternates in item a(2)(c) t above as appointed by the AFRRI Director. 32

i (2) Voting hd hn.c members, invited by the Director of AFRR1, to assist in teview of a particular problem,

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 E.URCTION AND AUTHORITY 6.2.2.1 Enns11en The Reactor and Radiation Facility Safety Committee is directly responsible to the Director, AFRRI. The committee 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 Au;hority 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 CH ARTER AND RULES 6.2.3.1 Alternatts 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 hlecting Frecy.ency The RRFSC or a subcommittee thereof shall meet at least four times a - cale :oz year, The full RRFSC shall meet at least semiannually. 6.2.3.3 Ouoruni 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 Maling Ruhs Each regule.r 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

6.2.3.5 Minutes Minutes of the previous meeting shall be available to regular members at least 1 week before a regular scheduled meeting. 6.2.4 REVIEW FUNCTION 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 previcasly 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. Proposed changes in technical specifications, the Safety Analysis Report,

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

f. Significant variations from normal and expected performance of facility 1

quipment that might affect nuclear safety. e 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 tolthe AFRRI ~ Director. Audits may be perform'ed 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

~. b. Performance, training, and qualifications of the reactor facility operations staff. Results of all actions taken to correct deficiencies occurring in facility c. 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 RRFSC or the Director /AFRRI. g. Reactor Facility ALARA Program This program may be a section of the total AFRRI program. 6.3 PROCEDUIGS 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). 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. These activities are as follows:. 4 Conduct ofirradiations and experiments that could affect the operation a. and safety of the reactor, b. Reactor staff-training program. Surveillance, testing, and calibration of instruments, components, and c. systems involving nuclear safety,

d. Perso'nnel radiation protection consistent with 10 CFR 20.

Implementation of required plans such as the Security Plan and e. Emergency Plan.

f. Reactor core loading and unloading,

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

6.3.2 Although substantive changes to the above procedures shall be made only with approval by the Reactor Facility Directer, 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: i

a. Reactor Facility Director 35'

b. Safety and Health Department

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

Special Reactor An.thorization for new experiments or experiments not a. included in a Routine Reactor Authorization. These experiments shall be performed under the direct supervision of the Reactor Facility Director or designee. b. Routine Reactor Authorization for experiments safely performed at least once. These experiments may be performed at the discretion of the Reactor Facility Director and coordinated with the Safety and Health Department when appropriate. These authorizations do not require additional RRFSC review, c. Reactor ParametersAuthorization for routine measurements of reactor parameters, routine core measurements, instrumentation and calibration checks, maintenance, operator training, tours, testing to verify reactor outpute, and other reactor testing procedures. This shall constitute a i single authorization. These operations may be performed under the authorization of the Reactor Facility Director or the Reactor Operations Supervisor. 6.4.3 Substantive (reactivity worth more than +/- 50.25) changes to previously approved experiments shall be made only after 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 +/- 50.25) may be approved by the ROS. Approved experiments shall be carried out in accordance with established procedures. 6.5 REOUIRED ACTLDES 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 NBC,

b. The safety limit violation shall be reported to the Director of NRC Region I, Division of Radiation Safety and Safeguards (or designate);

l the Director, AFRRl; and the RRFSC not later than the next working day. A Safety Limit Violation Report shall be prepared.This teport 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 r.ecurrence. A 36 1

d. The Safety Limit Violation Report shall be submitted to the NRC; the Director, AFRRl; 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 probable 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. Promot Notification With Written Follomtp. 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 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 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 baseifor the Technical' Specifications that have or could.have permitted reactor operation with a. smaller margin of safety than in the erroneous analysis.. R (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 plant life of conditions not specifically considered in the Safety Analysis Report or Technical Specifications that require remedial action.or correctiven measures to prevent the existence or development of an unsafe condition. 6.6 REPORTS In addition to the applicable reporting requirements of Title 10of the Code of Federal Regulations, the following reports shall be submitted to the Director of the appropriate NRC Regional Office unitis otherwise noted. 6.6.1 QPER ATING REPORTS i 37

( a. Stuttip Report: A summary report of planned startup and power escalation testing shall be submitted following (1) receipt of an operating license; (2) amendment of th.: 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 measuced values of the 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 requited 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 exiticality, 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 1: ports shall be submitted at least every 3 months until all three events have been completed. b. Annual Operating Report: 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 made by license shall provide a comprehensive summary of the operating experience having safety - significance that was gained during the year, even though some repetition of previously reported information may be involved. References in the annuti operating report to previously submitted reports shall be clear. Each annual operating report shallinclude (1) A brief narrative summary of (a) Changes in facility design, performance characteristics, and operating procedures related to reactor safety, that occurred during the reporting period (b) Results of surveillance test and inspections --(2) A tabulation showing the energy generated by the reactor on a monthly basis, the cumulative total energy since initial criticality, and the number of pulses greater than $2.00 (3) List of the unscheduled shutdowns, including the reasons and the corrective action taken, if applicable 38 .1

(4) Discussion of the major safety related corrective maintenance performed during the period, including the effects (if any) on the safe operation of the reactor, and the reasons for the corrective maintenance required (5) A brief description of (a) Each cha: to the facility to the extent that it changes a dercription of the facility in the Safety Analysis Report (b) Changes to the procedures as described in the Safety Analysis Report (c) Any new experiments or tests performed during the teporting period that are not encompassed in the Safety Analysic-Report (6) A summary of the safety evaluation made for each change, test, or experiment not submitted for Commission' approval pursuant to Section 50.59 of 10 CFR Part 50. The summary shall clearly show the reason leading to the conclusions that no unreviewed safety question existed and that no change to the Technical Specifications was required, (7) A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as determined at or prior to the point of such release or discharge. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed, a statement to this effect is sufficient. (a) Lialtid Waste -(summarized on a quarterly basis)- (i) Radioactivity discharged during the reporting period Total radioactivity released (in curies) Concentration limits used and isotopic cornposition ii 4 greater than 3 x 10 microcuries/ml for fission and 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) Gaseous Waste (summarized on a quarterly basis) l 39 L

. _ - - -. ~. - - - - L Radioactivity discharged during the reporting period (in curies) for: Argon 41 Particulate with half lives greater than 8 days. (c) Solid %gg (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 1 performed outside the facility (9) A list of exposures greater than 25% of the allowed value (10 CFR

20) received by reactor personnel or visitors to the reactor facility-6.7 RECO_RDS 6.7.1 RECORDS TO BE RETAINED FOR A PERIOD OF AT LEAST 5 XEARS OR-AS REOUIRED BY 10 CFR REGULATIONS a.

Operating logs or data that shallidentify (1) Completion of pre-startup checkout, startup, power changes,.and shutdown of the reactor (2) Installation or removal-of fuel elements, control 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. measurements and other reactivity measurements-b. Principal maintenance operations c. Reportable occurrences q

d. Surveillance activities required by Technical Specifications I

c. 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 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 j

h. Fuelinventories and fuel-transfers 40-i

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 LEAST ONE COMPLETE Illt\\] SING CYCLE a. Training exams b. Requalification records 6.7.3 RECORDS TO BE RETAINED FOR THE LIFE OF THE FACILITY-a. G aseous and liquid radioactive effluents released to the environs b. Appropriate offsite environmental monitoring surveys - Radiation exposures for all personnel c. d. Updated as built drawings of the facility. ( 41 ..}}