ML20042F031

Proposed Tech Specs,Increasing Max Licensed Steady State Power from 1,000 Kw to 1,100 Kw,Allowing Implementation of Fuel Follower Control Rods & Future Installation of Microprocessor Based Instrumentation & Control Sys

ML20042F031
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	03/31/1990
From:	ARMED FORCES RADIOBIOLOGICAL RESEARCH INSTITUTE
To:
Shared Package
ML20042F028	List: ML20042F027 ML20042F029 ML20042F031
References
NUDOCS 9005070128
Download: ML20042F031 (21)

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TEC}lNICAL SPECIF] CATIONS 1011 Tile AFRR1 REACTOR FACILITY LICENSE NO. R 64 DOCKET # 50-170 TABLE Oi CONTENTS Parc 1.0 DEFINITIONS .

1.1 ALARA 1 12 Channel Calibration 1 1.3 Channel Check 1 1.4 Channel Test 1 1.5 Cold Critical J 1.6 Core Grid Position 1 1.7 Experiment i 1.8 Experimental Facilities 1 19 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 Facility Director 3 1.16 Reactor Operation 3 1.19 Reactor Safety Systems 3 1.20 Reactor Secured 3 1.21 Reactor Shutdown 3 1.22 Repcetable Occurrence 3 1.23 Safety Channel 4

. 1.24 Safety Limit 4 l.25 Shutdown Margin 4 1.26 Standard Control Rod 4 1.27 Steady State Mode 4 1.28 Transient Rod 4 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 S afet > Limit - Fuel Element Tem perat ure 5 22 Limiting Safety System Settings for Fuel Temperat ure 5 3.0 LIMITING CONDITIONS FOR OPERATIONS 3.1 Reactor Core Paramet ers 7 31 1 Steady Stat <- O perst mn 7 3 .l.2 Pulse Mode Opera t non 7 313 Reactivity Limitations 8 3.1.4 ram Time 8

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c. Retetor Pool' ,

. d. Core Experiment Tub-

e. Portable Beam Tubes

f. Pneumatic Transfer System 4

g. ' incore Locations 1.9 FUEL, ELEMENT ,

L A fuel element is a single TRIG A fuel rod, or the fuel portion of a fuel follower control rod.

1.10 INSTRUMENTED ELEMENT An instrumented element is a special fuel element in which sheathed chromal/alumel or ;

equivalent thermocouples are embedded in the fuel.

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

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

channel.  ;

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

1.14 ON CA1L A person is considered on call if i

a. The individual has been specifically deugnated and the operator knows of the -

designation; r

b. The individual keeps the operator posted as to hia her whereabouts.and telephone

.i number; and ,

y c The individual. is capable of getting to the reactor fa Ety withia 30 minutes under -

, g normal circumstances.'

1.15 OPERAHLE ,

,, A system channel, device, or Vompcment shall be considered opeable when it is capable of ~- w'$

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performing its intended functi>n(s) in a normal manner 3:

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4 I 1 10 Pl'ISE MODE Operation in the pulse inode 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.  !

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! 1.17 REACTOR PACILITY DIRECTOR .

k The Reactor Facility Director (RFD) is the senior licensed operator responsible for administration and operation of the Reactor Facility and for determination of applicability of procedures, experiment authoritations, and maintenance operations. _ The RFD's

• designee *, a '

person app inted orally or in writing by the RFD and meeting the requirements of Section 6.1.31, may_ perform any. functions of the RFD in his absence.

l. 1.18 REACTOR OPER ATIONI Reactor operation .is any condition wherein the reactor is not shut down, or any core n innintenance is being performed, or there is movement of any control rod, l

1.19' REACTOR J A f'ETY SYSTEMS I

Reactor safety systems are those systems, including their associated ;rput circuits, that are  ;

designated to initiate a reactor scram for the primary purpose of protecting the reactor or to 1 provide information that may require manual protective action to be initiated.

l 1.20 REACTOR SECURED i

The reactor is secured when all the following conditions are satisfied:

a. The reactor is shut down.

b. The console key switch is in the *off' position, and the key is removed form the console

.and is under the control of a licensed operatcr, or is stored in a locked storage area.  !

c. No work is in progress involving in core fuel handling or refueling operations, maintenance d the reactor or its control mechanisms, or insertion or withdrawal of in.

E core experiments. 'unless sufficient reactivity is removed to insure a 80.50 (or greater) shutdown margin with the most reactive control rod removed.

1.21 REACTOR SHUTDOWN i l

The reactor is shut down when the reactor is sul. critical by at least 60.50 of reactivity.

[ 1.22 IlEPORTABLE OCCURRENCE

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! A reportable occurrence is any of the following that occurs during reactor operation: a v

y c n. Operation wiu any safety system setting less conser ative than specified in Section-2.2,

, Limiting Safety System Settings.

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c Malfun: tion of a rsquirrd re:ctor or experiment s,afety system component th:t could render the system incapable of performing its intended safety function unless the

-- rnalfunction is discovered during tests.

d. Any unantleipated or uncontrolled positive change in reactivity greater than $1.00,-

e. An observed inadequacy in the implementstion of either administrative or procedural controls, so that the inadequacy could have caused the existence or development of a condition that could result in operation of the reactor in a mr.nner less safe than t conditions covered in the Safety Analysis Report (SAR),

f. The release of fission products from a fuel element through degradation of the fuel '

cladding. Possible degradstion may be determined through an increase in the background activity level of the reactor pool water. '

g. An unplanned nr uncontrolled release of radioactivity that exceeds or could have  ;

exceeded the limits allowed by' Title 10, Part 20 of the Code of Federal Regulations (10 CFR 20), or these technical specifications.

l 1.23. S AFETY Cll ANNFL .

A safety channel is a measuring channel in the reactor safety system that provides a reactor ,

protective function.

l 1.24 S AFETY LIMIT .

Safety limits are limits on'important process varichles that are found to be neensary to .,

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reasonably protect the integrity of certain physical barriers that guard against te uncontrolled release of rsdioactivity.

l 1.25 l JilUTDOWN M ARGIN Shutdown rnargin shall mean the minimum shutdown reactivity considered necessary to provide confidence that the reactor can be made suberitical by means 'of the control and safety rystems, starting from any permissible operating conditions,' an'd that the reactor will ,

remam subtritical without further operator action.- 5

!- 1.20 STANDARD CONTROL ROD l'

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

! 1.27 STE ADY ST, ATE MODE ll Operation ir the steady state mode shall mean the steady state operation of the reactcr

! either by manaal operation of the control rods or by- automatic operation of one or most control rod (servocontrol) at power levels not exeteding 1.1 megawatts, utilizing the appropriate scrams in Table 2 and the appropriate it.terlocks in Table 3. >

! l- I.28 ,TR ANSIENT ROD -

The tr, nsis, rod is a control rod eith scram capabilities that can be rapidly ejected from the reactor core tr produce h. pulse. It is activated b) applying compressed air to a piston.

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r M.0 yk_11 TING CONDITIONS FOR OPER ATIONS 4 3.1- HEACTOR CORE PAR AMETERS 3.1.1 STEADY STATE OPFR ATION j Antilicability This specifia. tion applies tc. the maximurn reactor power attained during steady state operation.

i Ohiective ,

To assure that the reactor safety litnit (fuel temperature) is not exceeded, and to i provide for a set point for the high flux limiting safety systems, so' that I automatic protective action will prevent. the safety limit from being reached '

during steady state operations.

Specifications ,

The reactor steady state power level shall not execed 1.1 megawatts. The normal steady state operating power of limit of the reactor shall be.1.0 megawatt. For purposes of testing - and calibration, the reactor may be operated at power;les els

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not to exceed 1.1 megawatts during the testing period. '

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

3.1.2 PULSE MODE OPER ATION Avnlical.ili'r i This specification applies to the maximum thermal energy produced in the )

reactor as .a result of a piorapt critical ins:rtion 'of reactivity.

1 Obiective The objective is to assure that. the fuel temperature safety limit will not be exceeded.

Snecification The maximum step insertion of reactivity shall be 2.651 Ak/k ($4.00) in the- [

pulse mode, Rasis J

Based upon the Fuchs-Nordheim mathematical model (eited by C.E. Clifford et al. in the April 1901 G A Report = 2119J

• Model of the AFRRI-THIG A Heactor"). an insertion of 2Fi 6L/L results iri a maximum everage fuel temperature of less than 550'C thereti) staying within the limiting safety settings that protect the safety bnni The 50'C- maron to the Limiting Safety i

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= System Setting cnd the' 450*C mcrgin to tho sciety, limit omply allow for uncertainties due to extrapolation of measured data, accuracy of measured data, and location of instrumented fuel elements in the core.

3.1.3 REACTIVITY LIMITATIONS Appli ability

.l These specifications apply to the_ reactivity condition of the reactor and the I reactivity worths of control rods and experiments. They apply for all modes of operation.

Objective -

The objective is to gr.arantee thst the reactor can be shut down at all times and that the fuel temperature safety limit will not be exceeded. >

SpeEifications

a. The reactor shall not be operated with the maximum available excess .

. reactivity above ccid 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 conttol rods with the most reactive control rod fully withdrawn or removed shall be >

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

Basis

a. The limit on available excess reactivity est'ablishes_ the1 maximum power if-

. all control elements are removed.

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

3.1.4 SCH AM TIME' Applicability The specification applies to the time required to fully insert'any control rod to a 4 full down. position from a full up position.

L Obiective The objective is to achieve rapid shutdown of the reactor .to prevent fuel dama~ge.

Specification The time for scram initiation to the full insertion of any control rod from -a full up position shall be less than 1 second 8

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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, 3.2.2 REACTOR SAFETY SYSTEMS -

Applicabiljn, This specification applies to the reactor safety system, Objective The objective is to specify the minimum number of reactor safety system -

channels that muct be operable for: safe operation.

Specification The reactor.shall not be operated unless the safety. systems described in Tables 2 L 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, Iligh Flux 1,1 MW 2' O Console Manual Scram Bar Closure switches 1 1 High Voltage Losa to Safety Channels 20% loss 2 1 Pulse Time 15 seconds - 0 ;1 E:nergency Stop (1 each exposure room, 1 on console) Closure switch 1 1.

Pool Water Level 14 feet from top of core 1 1-Watchdog (DAC to CSC) . On digital console ' 1 1-Basis The fuel temperature and power level scrams provide protection to assure that the reactor can be shut down before.the safety limit on the fuel element temperature will be exceeded. The manual scram allows the operator 'to shut down. the system at any time if an unsafe or abnormal condition occurs. In the event of failure of the power supply for the safety chanibers, operation of the -

reactor without adequate instrumentation is prevented The preset timer ; insures that the reactor power level will rsduce to a low level after pulsing. The 10 d

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A emergency stop allows' personnel tr:pped in a potentillly hasardoIs expc;ure roon. or the reactor operator to stop actions through the interlock system The

+: pool water level insures that a. loss of biological shielding would result in a reactor shutdown. 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 Effective Mode Action Prevented Steady State . Pulse Pulse initiation at power levels greater X than L kilowatt

-Withdrawal of any control rod expect transient X Any rod withdrawal with count rate in .X X operational-channel below 0.5 eps Simultaneous manual l withdrawal of two X standard control rods q

f 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 j temperature to' approach the safety limit. The interlock that prevents movement of standard control rods'in pulse mode will prevent the inadvertent placing of 2

the reactor on a positive period while in pulse mode. Requiring a count rate to E!

be seen by the operational channels insures sufficient source neutron;n to' bring the i

3 reactor critical .under controlled conditions.~ The interlock that prevents the- 1 simultaneous manual withdrawal of two standard control rods limits thetamount  !

of reactivity added per unit time.

3.2.3 FACILITY INTERLOCK SYSTEM 1

Applicability This specification applies to the interlocks 'that prevent the accidental exposure of - l an individual in either exposure room.

Objective  ;

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The objective is to provide sufficient warning and inter:ocks to prevent movement--.  !

of the reactor core to the exposure room in which someone may be working, or -  ;

prevent the inadvertent movement of the core into the lead shield doors. j.

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. Functional' checks shall be made annually,' but not to exceed 15 months, to insure the fellowing:

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

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

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

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These functional checks will ver fy operation:of the interlock system; Experience at AFRRI indicates that this is adequate to insure operability.

4.2.5 REACTOR' FUEL El,EMENTS Applicability This specification applies to the: surveillance requirements for the fuel elements.-

q Ob_iective j 4

The objective is to verify the integrity of the fuel element cladding. ' l t

Specificat ions All the fuel elements present in the reactor core, to _ include fuel follower control j rods, shall be inspected for damage or deterioration, and measured for length and

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l bow at intervals separated by not more than 500 pulses of inowtion greater than }

$2.00 or annually ' (not to exceed 15 months), - whichever occur first. ' Fuel-  !

elements in long term storage need not be measuttd: until returned. to core; ,

however fuel elements routinely moved to temporary. storage shall be measured .  !

every 500 pulses of insertion greater than $2.00 'or tr nually (not to exceed 15  ;

months), whichever occurs first.- {

l Basis I The frequency of inspection and measurement -h based on the: parameters most likely to affect the fuel cladding of a pulse reactor, and the utilization of fuel- ,

elements whose characteristics are well known s 1 The limit of transverse bend has1been shown to result in no difficulty in {

disaswmbling the core. Analysis of a wont case scenarifin which. two adjacent 4

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fuel elements suffer sufliciently severe transverse bends to result in the touching of the fuel elements has shqwn tha:Dao damage io the fuel elements will result -  !

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• Appliesbility This specification applies to the surveillance requirements for monitoring the pool water s and the water conditioning system.

Objective The objective is to assure the integrity of the water purification system, thus maintaining the purity of the reactor pool water, eliminating possible tsdiation hasards

, from activated impurities in the water system, and limiting the potential corrosion of' fuel cladding and other components in the primary water system.

Specifications  !

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

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

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

44 VENTil,ATION SYSTEM '

Applicability This specification applies to the facility ventilation system isolation.

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

Specification 4 The operating mechanism of the positive sealing dampers in the reactor 7oom ventilation ~

system shall be verified to be operable and visually inspected at 'let.st monthly, not to l exceed six weeks.

Basit Experience accumulated over years of operation has demonstrated that the tests of the  :

ventilation system on a monthly basis are sufficient to assure proper operation of the  !

system and ' control of the release of radioactive material. .

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5.0 - DESIGN FEATURES .

5.1 SITE AND FACII,lTY DESCRIPTION 'i Applicability This specification applies to the building that houses the reactor.

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

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Specifications l

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 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 - AFRR1 complex. }

b. The reactor room shall contain a minimum free volume of 22,000 cuble 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 1 automatically close off ventilation to the reactor room upon a signal from the reactor room air particulate monitor. l!

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

Dani, The fac~ility 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 emcegency shutdown of the ventilation system. Building construction and gaskets around doorways help restrict leakage of air into or out. of the reactor room. The stack height insures an adequate dilution of effluents -well above ground level. The separate .

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

5.2 REACTOR CORF AND F11EL t

5.2.1 RE ACTOR FI'El.

Applicability These specifications apply to the fuel elements. to include fuel follower control l rods, used in the reactor core.

Object ive 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) 25 i

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assure that the fuel elements used in the core are substanti:lly those taalysed in the Safety Analysis Iteport.

Specifications The individual nonirradiated standard TRIG A fuel elements shall have the following characteristics:

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

b, Ilydrogen to-airconium ratio (in the Zril ): Nominal 1.711 atoms .to 1.0 Zr atoms with a range between 1.0 and 18. l

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.

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A maximum uranium content of 9 weight percent in' a standard TRIG A element - J is greater than the design value of 8.5 weight percent, and encompasses the j maximum probable variation in individual elements. Such an increase in loading j would result in an increase in power density of less than OE An increase in i local power density of 0% in individual fuel element reduces the safety margin )

by 10% at most. The hydrogen.to airconium ratio of 1.7 will produce a maximum pressure within the cladding well below the rupture strength of the cladding.

The power density of a 12.0 weight percent fuel follower element of the same diameter as a control rod, which is smaller than the standard TRIGA element, f

will produce the same power density in the . local area as the standard -8.5 weight . '

percent TRIG A elements due to'its increased hydraulic diameter. I 5.2.2 REACTOR CORE i Annlienbilit y These specifications apply to the configuration of fuel and in core experiments.

i obienive f

The objective is to restrici the arrangement' of fuel elements and experiments so j as to provide assurance that excessive power densities will not be produced.  !

i Snecifie nt ions  !

a. The reactor core shall consist of standard TRIGA reactor fuel elements in a close packed array and a minimum of two thermocouple l instrumented TitlG A reactor fuel elements )

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b; There shall b2 four single core pccitions occupied by the three standard control rods and transient rod, e neutron start-up source with holder, ,

- and positions for possible in-core experiments. 3

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

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

e. Fuel elements indicating an. elongation greater than 0.100 inch, a -i lateral bending greater than 0.0625 inch, or significant visible damage shall be considered damaged, and shall not be used in the reactor core.

Basis Standard TRIGA cores have been in use for years,.and their safe operational characteristics are well documented. Experience with TRIGA reactors has shown

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that fuel element bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified to (a) assure that the cladding material will not be subjected to stresses that could cause a loss of.

integrity in the fuel containment, and (b) assure adequate coolant flow. ,

5.2.3 CONTROL HODS A pplicability These specifications apply to the control rods used in the reactor core.

Obiective 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 l- characteristics.

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l. Specificationa l a. The standard control rods shsll have scram capability,. and shall contain borated graphite, B4C powder, or boron and'its compounds in solid form as

~l a poison in altiminum or stainless-steel cladding. These rods msy have an- .i aluminum, air, or fuel follower, if fuel followed, the fuel region will conform to the Specifications of 5.2.1.

b. The transient control rod'shall have scram capability, and shall contain

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l borated graphite, B C powder, or boron and its compounds in solid form as 4

a poison in aluminum or stainless-steel cladding. This rod may incorporate .

an aluminum, poison, or air follower.

' Basi.

The poison requirements for the control rods are satisfied by usinF neutron-absorbing borated graphite, B C _ powder, or boron. and its compounds. These 4

materials must be contained in a suitable cladding material. such as aluminum or-stainless steel, to insure mechanical stability during movement ' and to isolate the - i l- poison from the pool water environment Scram' capabilities are provided by the i rapid insertion of the control rods. which is the primary operational safety  ;

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feature of the reactor. The transient control rod is designed for use in a pulsing

. TRIG A reactor.

5.3 SPECIAL NUCLEAR MATERI AL STORAGE  ;

Apolieability ,,

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, f Specification All fuel elements not in the reactor core shall be stored and handled in accordance with 3 applicable regulations. Irradiated fuel elements and fueled devices shall be stored in an .!

array that will permit sufficient natural convective cooling by water or air, so that the ,{

fuel element or fueled device temperature will not exceed design values. Storage shall be I such that groups of stored fuel elements will remain subcritical under all conditions of s moderation.

f EA!!iP.

The limits imposed by this specification are conservative and ' assure safe storage and handling. Experience shows that approximately 67 fuel elements are required, of the i

design used at AFRRI, in a closely packed. array to achieve criticality. Calculations show l

that in the event of a full storage rack failure with all 12 elements falling in the most reactive nucleonic configuration, the mass would be less than that required for criticality.

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

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6.0 ADMINISTRATIVE CONTROLS l

. 6.1 ORGANIZATION l 1

6.1.1 STRUCTURE l

The organisation of personnel for the management and operation of the AFRRI reactor facility is shown in Figure 1. Organisation 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 l Chain. In which the Reactor Facility Director has direct responsibility to the-Director, ' AFRRI.

Director, AFRR!

Chairman Adataiareueo p8"8"YW. Reactor and Radiation safety e4 m Facility Safety Committee Health Dept. w

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Radiation .

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Ad*arr Adeory l Chief, Reactor i

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............ Reactor Facility Director ..............j i

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Rosetoropersuan statr-ylgure 1. Organisation of Personnel for Management and Operation of the AFRRI Rosetor Facility.

Any reactor staff member hae access to the Director for matters of safety, 29

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6.1.2 R ESPONSIBILITY ,

The Director, AFRRI, 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 Spf licability of procedures, experiment sunwrisations, maintenance, and operations. The RFD may designate an individual who meets the requirements of Section 6.1.3.a to discharge the RFD's responsibilities in the RFD's absence. During brief absences of the Reactor Facility Director and-his designee, the Reactor Operations Supervisor shall discharge these raponsibilities.

6.1.3 STAFFING 6.1.3.1 Selection of Personnel 7

a. Reactor Facility Director I At the time of appointment to this position, the Reactor Facility Director shall have 6 or more years of nuclear experience. Higher education in a scientific or nuclear engineering field may fulfill up to 4 years of ' experience on a one-for-one basis. 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.

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b. Reactor Operations Supervisor (ROS)

At the time of appointment to'this position, the ROS shall have's years nuclear experience. Higher education in a science or nuclear

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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 at AFRRI or at a similar facility before the appointment 7 to this position,

c. Reactor Operators / Senior Reactor Operators At the time of appointment to this position, an individual shall have l

a high school diploma or equivalent, and shall possess the appropriate USNRC license.

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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. -

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i 6.1.3.2 Operations

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

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

2. Radiation control technician'on call i 30 i

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3l At'leut one licensed Resetor Operator (RO) or Senior Re:ctor 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 incapacitated. .

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b. hiaintenance 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. hianagement personnel (Reactor Facility Director, AFRR1 Director)  ;

2. Radiation- safety personnel (llead, Safety and Health Department) ,

3. Other operations personnel (Reactor Staff, ROS) 0.1.4 TR AINING 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.-

0.2 REVIEW AND AUDIT Tile REACTOR AND R ADIATION FACILITY S AFETY COMMITTEE (RRFSCl 6.2.1 COMPOSITION AND QUAblFICATIONS 6.2,1.1 Composition

a. Regular RRFSC hiembers (Permanent hiembers) r (t) The following shall be members of the RRFSC:

(a) Chairman, Safety and llealth Department, AFRRI (b) Reacter licility Director, AFRRI (2) The following shall be appointed to the RRFSC by the Director, AFRRl:

(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 Spe-iahst, or a llealth Physics Specialist.

b. Special RRFSC Alembers (Temporar.s hiembers)

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

(2) Voting y1 & members inuted by the Director of AFRR1, to assist in review of a particular problem

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c. Nonvoting members as invited b> - the Chairm'an RRPSC.

l 6.2.1.2 Qualifications '

The minimum qualifications for a person on the RRFSC shall be 6 years of L professional experience in the discipline or specific field represented. A .  ;

baccalaureate degree may fulfill 4 years of experience.

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.6.2.2 FUNCTION ~ AND AUTilORITY 6.2.2.1 Function i 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 Authoritv 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 CII ARTER AND RUI,ES 6.2.3.1 Alternates Alternate members may be appointed in writing by- the RRFSC Chairman l 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 Meetine Freauency The RRFSC or a subcommittee thereof shall meet at least four times a ,

calender year. The full RRFSC shall rneet at least-semi annually.

l 6.2.3.3 Quorum A quorum of the RRFSC for review shall consist of the' Chairman -(or. .

g 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.

l' 6 2.3 4 Votine R ule.

L Each regular RHFSC member shall have one vote. Each special appointed member shall have one vote. The majority is 517e or more of the regular l and special members present and voting.

l 6.2.3.5 Min utes Minutes of the previous meeting shall be available to regular members at least one week before a regular scheduled meeting.

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f. - Any ,other crea of Fccility oper:tions considered cppropri:te by the RRTSC or the Director /AFRRI.

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. g. Reactor facility ALARA Program.- This program' rnay be a section of the total APRRI program.

6.3 PROCEDURES i

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 j '

Committee (RRFSC). The procadures 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:

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a. Conduct of irradiations and experiments that could affect the operation 'and ,

safety of the reactor,

b. Reactor staff training program,

c. Surveillance, testing, and calibration of instruments, components, and systems involving nuclear safety. ,

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d. Personnel radiation protection consistent with' 10 CFR 20. i e, implementation of required plans such as the Security Plan and Emergency Plan, i

f. Reactor core loading and unloading.

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

6.3.2 Although substantive changes to the above_ procedures shall be made onlyi 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 reactor authorization, new experiments shall be reviewed for l radiological safety and approved by the following:

a. Reactor Facility Director lx Safet> and Ilealth Department 'l

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:

a. Special Renctor Aut horization for new experiments or experiments not {'

included in a Routine Reanor Authorization. These experiments shall be -

performed under the direct supervision of the Reactor Facility Director or designee 34

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I b. Routine Reactor Authoritation 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 authorisations do not require additional RRFSC +

review. _-

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

I 6.4.3 ' Substantive (reactivity worth more than *$0.25) changes to previously approved -

p experiments shall be made only after review by the RRFSC and after approval L (in: writing) by' the Reactor Facility Director or designated alternate. Minor

} changes that do not significantly alter the experiment (reactivity worth of less l -l, than *80.25) may be approved by the ROS. Approved experiments shall' be ,

carried out in accordance with established- procedures. -

6.5 REQUIRED ACTIONS t 6.5.1 ACTIONS .TO DE TAKEN IN CASE OF SAFETY LIMIT VIOLATION; 1

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a. The reactor shall be shut'down immediately, and reactor operation shall not be resumed without authorisation by the NRC.

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b. The safety limit violation shall be reported to the Director of NRC Region 'i i_

1, Office of Inspection and Enforcement (or designate); the Director,- AFRRl; i l and the RRFSC not later than -the next working day.

1 c, A Safety Limit Violation Report shall be prepared. This report shall be-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 redoce the probability of recurrence.

l d. The Safety Limit Violation Report shall be submitted to the NRC; the-Director, AFRRI, and % RRFSC within 14 days of the violation.

6.5.2 REPORTABLE OCCURRENCES I Reportable occurrences as defined m 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. Prom pt Notifiention With Written Followup. The types of events listed below shall be reported as soon as possible by telephone and confirmed by 1