ML20199K141

Proposed Tech Specs

ML20199K141
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Site:	Reed College
Issue date:	06/30/1986
From:	Cronyn M, Kay M REED COLLEGE, PORTLAND, OR
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ML20199K132	List: ML20199K128 ML20199K141
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NUDOCS 8607090051
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% R=d R$ctor Fccility Technical Specifications Page 1 '

< l SAFETY RELATED DOCUMENT '

• l Reed Reactor Facility ,

Docket # 50-288 Ucense # R-112 ,

TECHNICAL SPECIFICATIONS FOR THE REED COLLEGE TRIGA MARK 1 REACTOR .

' June,1986 ,

Reed Reactor Facility The Reed Institute dba ' '

Reed College 3203 SE Woodstock Boulevard

~ Portland, Ore ~gon 97202 -

(503) 771-1112 c  :-

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Marshall W. Cronyn, Michael A. Kay .

Vice President-Provost Director, Reed Reactor Facility

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Revised edition incorporating July 2,1968 edition and changes issued July 28, 1969, October 3,1972, August 22,1973, January 17,1974,and.

State of Oregon ~

This is to certify that -Marshall Cronyn and Michael Kay appeared before me and signed in my presence this document on July 1, 1986.

W InV b/ v OM.B Tracy Frantel, Notary Public Commissicn expires 11/13/87 June 1986 edition

' 8607090051 860622 PDR ADOCK 05000288 P PDR

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. R;ed Reactor Facility Technical Specifications Page 2 ,

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TABLE OF CONTENTS 1.0 DEFINITIONS- 4 1.1 AirConfinement 4 1.2 Certified Operators 4 1.3 Channel, Instrumentation 4 1.4 Core Excess 4 1.5 Expenment 5 1.6 Fuel Element, Standard 5 1.7 Limiting Conditions for Operation 5 ,

1.8 Operable 5 1.9 Operating 5

_ 1.1.0 Protective Action 6 1.11 Reactivity, Excess 6 1.12 Reactivity, Limits 6-1.13 ReactorBay 6 1.14 Reactor Com, Standard 6 1.15 Reactor Com, Operational 6 '

1.16 Reactor Facility -

7 1.17 Reactor Operating 7 1.18 Reactor Safety System .

7 1.19 ReactorSecured 7 1.20 Reactor Shutdown 7 1.21 Reference Core Condition 8 1.22 Research Reactor '

8 1.23 Rod, Control 8 1.24 Safety Limit 8 1.25 Scram 8 1.26 ScramTime 9 1.27 Shall, Should, and May -

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1.28 Shutdown Margin 9 - - ~.

1.29 Shutdown, Unscheduled 9 1.30 Startup 9 1.31 Surveillance' Activities 9 1.32 Time Intervals 9 1.33 Value, Measured 10-1.34 Value,True 10 1.35 Zero Power Critical 10 m

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 11 2.1 Safety Limit 11 2.2 Limiting Safety System Setting 12 3.0 LIMITING CONDITIONS FOR OPERATION 13 3.1 Reactor Core Parameters 13 3.2 Reactor Control and Safety System 15 ,

3.3 Operational Support Systems 17.

3.4 Limitations On Experiments 20 June 1986 edition

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. Reed Reactor Facility Technical Specifications Page 3 4.0 SURVETT LANCE REQUIREhENTS . 24

_ 4.1 Reactor Core Parameters 24 -

4.2 Reactor Control and Safety System 25 4.3 ' Operational Support Systems , 27 4.4 Limitations On Experiments 29 5.0 DESIGN FEATURES _ 31 5.1 Site an,d Facility Description 31 5.2 Reactor Coolant Sy~s tem 33 5.3 Reactor Core and Fuel -

33 5.4 Reactor FuelElement Storage 35 6.0 ADMINISTRATIVE STRUCTURE '

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• 6.1 Orgamzation -

36 6.2 Review and Audit ~-

37 6.3 Operating Procedures - 40 6.4 Experiment Review and Approval 41 -

6.5 Requid Actions 41

' 6.6 Reports 42 6.7 Records . -

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7.0 El+ECTIVE DATE '46 I

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Reed Reactor Facility Technical Specifications Page 4 1.0 DEFINITIONS . 4 '

1.1 Air Confinement' _. ,

Confinement means a closure on the overall facility which controls the movement of . -

air into it and out through a controlled path.

1.2 Certified Operators

, An individual authorized by the chartering or licensing organization to carry out the -

responsibilities associated with the position requiring the certification.

1.2.1 Class A Reactor Operator a

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An individual who is certified to direct the activities of Class B Reactor Operators' Such an individual is also a reactor operator and is commonly referred to as a ,

Senior Reactor Operator.

1.2.2 Class B Reactor Operator , ,,. a An individual who is certified to manipulate the controls of a reactor. Such an individual is commonly referred to as a Reactor Operator.

1.3 Channet Instrumentation A channel is the combination of sensor, line, amplifier or other electronics, and output device which are connected for the purpose of measuring the value of a

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parameter, or performing a safety related function.

.. 1.3.1 ChannelTest '

Channel test is the introduction of an appropriate signal (ie. nuclear for'a nuclear

. channel, physical activation for a level sensor) into the channel sensor and

measurement of channel output for verification that the entire channel is operable. -

1.3.2 ChannelCheck Channel check is a qualitative verification of acceptable performance of a channel or --

portion of a channel by observation of channel behavior (eg. comparison of? -- '

mdependent channels, introduction of electronic signals into the channel).

! 1.3.3 ChannelCalibration i

! Channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment _ ;

. actuation, alarm, or trip and shall be deemed to include a channel test.

1.4 - Core Excess -

The RRF term for excess reactivity measured at zero power critical (see Reactivity, .

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Excess and Zero Power Critical) as part of the startup procedure. Also used to denote the reactivity available at a specified power level (eg. at 150KW the core excess is $1.82).

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- , 1.10 Pmtective Action -

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Protective action is the initiation.of a signal or the operation of equipment within the 7y reactor safety system iri response to a variable or condition of the reactor facility having reached a specifiedlimit.

, ,1.10.1 Instrument CliannelLevel . .

At the protective instrument channel level, protective action is 'the generation and .

- transnussion of a trip signal indicating' that a reactor variable has reached the ..

specified limit. - . r '

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1.10.2 ' Instrument Subsystem level' '

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At the protectiveins ~ t subsystem level, protective action-is the' generation and , .

transmission of a tri)isi ~ indicatin~g that a specified limithas been;eached. ,

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.. , , . , 1.10.3 Instrument System Level . ,

At the pr6tective instrument system level, protective action is the generatiori and* '

transmission of the command signal for the safety shutdown equipment 'to operate.. .

1.10.4 Reactor Safelr5ystem,Ievel .

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. At the reactor safBtysystem level, protective action is the operation of sufficient

,- equipment to immediately shut "

d'o'wn th8Wa'etdr; o a ,

' 1.11 -Reactivi~tv. Excess ~

Excess reactivity is that amount of reactivity that would exist if all the control rods ~ .

were. moved to the maximum reactive condition from the point where~ the reactor is exactly critical (k(eff) = 1)-(see Zero PowerCritical). '

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1.12 Retivity Limits - * -

The reactivity limits are those limits imposed on the reactor core excess reactivity. ' '

. Quantities are defined under reference core conditions. ,

-+ - 1.13- Reactor Bay t -

M The reactor bay is th'e enclosure in the Reactor Facility containing the, pool, reactor, and other equipment. ..

1.14 Reactor Core. Standard -

A standa'rd core is an arrangement of standard TRIGA Mark I fuel elements in 'the .

{ reactor grid plate and may include installed experiments. ,

1.15 Reactor Core. Ooerational *

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, An opeutional core is a standard core for which the gore parameters of excess reactivity, thutdown margin, power calibration, and reactivity worths of control .

.. m rods and exp riments have been determined to s'atisfy the requirements set forth in ,

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the Technical Specifications. ,

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1.0

. ,,, DEFINITIO. NS . -

1.1 ' Air Confinement ,

Confinament means a closure on the overall facility which controls the movement of ~

air into it and out through a controlled path #w

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1.2 Certified O.oerators "

An individual authoiized by the chartering or licensing organization to carry oct the ,~ ~

responsi,bilities ass'ociated with the'li osition requiring the certification.

-1.2.1 Class A Reactor Operator ^ .

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'An individual who is certified to direct the activities of Class B Reactor Operators.

. Such an individual is also a reactor operator and is commonly referred to as a

. , Senior Reactor Opentor. .

1.2.2 Class B Reactor Operator -

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An individual who is certified to manipulate the controls of a hactor. Such an individual is commonly referred to as a Reactor Operator; - -

1.3 . Channel Instrumentatidn , -

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A channel is' the combination of sensor, line, amplifier or other electrodies, and ' -

output device which are connected for the purpose of measuri,ng .the value of a - -

parameter, or performing a safety related function.

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1.3.1 ' Channel Tdst -

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%. Channel test is the introduction 'of a'n appropriate signal (ie. nuclear for a'miclear e a level sensor) into the channel sensor and -

i ~ . channel,ofphysical measurement channel output activation f@ for verification that the entire channel ~

1.3.2 ChannelCheck' - '

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Channel check is a qualitative verification of acceptabidperformance of a channel &*

portion of a chan,nel by observation of channel behavior (eg. comparison of mdependent channels, introduction of electronic signals into the channel). .

1.3.3 ChannelCalibration .

Channel calibration is an adjustment of the channel such that its output corresponds-with acceptable xcuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire chamel, including equipment ~

actuation, alarm, or trip and shall be deemed to include .a channer test; -

1s4 Core Excess .

The RRF term for excess reactivity ' measured at zero power critical (see Reactivity, Extess and Zero Power Critical) as.- part of the.startup procedure. Also used to -

denote the reactivity available at a specified power level.(eg. at 150KW the core excess is $1.82). .

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1.5 Experiment e " ~

a) Any apparatus, device, or material installed in the core or experimental facilities (except for underwater lights, fuel element storage racks, and the lilie) which is,not a design component of these facilities, or -

b) Any operation designeti to measure reactor parameters or characteristics.

1.5.1 Expenment, Movable . -

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, . A movable expenment is one where it is intended that 'the' entire exhriment may be moved in or near the core or into and out of the reactor while the reactor is operating (eg. pneumatic tube irradiations). ,

1.5.2 Experiment, Secured  :( "

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A secured experiment is any experiment, experiment facility, or component of an

• experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining. force must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating enviromhent of the experiment, or by forces which can arise as a r6 ult of credible conditions (eg. rotary specimen rack irradiations). .

f, 1.5.34 ExperimentalFacilities - "

Experimental facilities shall mean rotary specimen rack, pneumatic transfer tube, centrhl thimble, and irradiation facilities m, the core or in the pool.

1.6 Fuel Element. Standard . +

l A fuel element is a single TRIGA Mark I aluminum clad or stainless steel clad fuel- ,* .

moderator element. Fuel is U-ZrH,8.5 weight-% uranium enriched to a nominal

[ 19.7% U-235. Zirconium to hydrogen ratio is nominally 1:1 in. aluminum clad I . elements and 1:1.6 in stainless steel clad elements. .

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1.7 Limitine Conditions for 06eration

.. Limiting Conditions for, Operation (LCO) are those administratively established

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constraints on eqlffpment anii 6*p erationhl characteristics which shall be adhered to

. during openition of the' facility.~ The LCO's are the lowest functional capability or performance level required for safe operation of the fa'cility. ,

1.8 Ooerable -

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Operable means a component or system is capable of performing its intended

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function.

1.9 Ooeratine

.. Operating means a component or system is performing its inSnded fQnctiori.

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1.1(f Protective Action

,', Protective action is the initiation of a signal or the operation of equipment witn'n the s -

reactor safety system in res mnse to a variable or condition of the reactor facility having reached a specified .imit. *

, ,- 1.10.1 Instrument Channel Level -

At the protective instrument channel level, protective action is the generation and .

,, transmissi6n of a trip signalindicating that a reactor variable has reached the specifiedlimit. .

1.10.2 Instrument Subsystem Level' At the protective instrument subsystem level, protective action is the generation and Transnussion of a trip signalindicating that a specified limit has been reached.

1.10.3 ~ Instrument Systemlevel .

At the protective instrument system level, protective action is the generation and transmission of the command signal for the. safety shutdown equipment to operate. '

3 1.}0.4 Reac r Safety System Level ,

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At the reactor safety system level, protective action is the operation of s'ufficient -

equipnient to immediately shut down the reactor.. .

1.11; Reactivity. Excess '

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Excess reactivity is that amount of reactivity that would exist if all the control rods

,'were moved to the maximum reactive condition from the point where the reactor is exactly critical (k(eff) = 1) (see Zero Power Critical).

1.12[ Reactivity Limits .

. The reactivity limits am those limi.ts imposed on the reactor core excess reactivity. '

Quantities are defined under referdnce core conditions.

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.JL 1.13 Reactor Bay -

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The reactor bay is the enclosure in the Reactor Facility containing the pool, reactor,'

and other equipment.

1.14 Reactor Core. Standard ,

A standard core is hn arrangement of standard TRIGA Markl fuel elements in the .

rehctor grid plate and may include installed experiments.

,. 1.15 Reactor Core. Ooerational - ~

,. . 4 An operational core is a standard core for which the core parameters df excess reactivity', shutdown margin, power calibration, and~ reactivity worths of. control .

rods and experiments have been determined to satisfy the requirements set forth iii the Technical Specifications.

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1.r6 . Reactor Facility .

4 Reactor Facility refers to the specially designed and constructed addition to the Reed ,

College Chemistry Building in which the pobl, reactor, associated mechanical and electrical equipment, laborato,ry, counting rooms, and storage rooms are located.

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1.17 Reactor Ooerating The reactor is operating whenever it is not secured or shutdown. -

1.18,. Raetor SafEtv'Svstem ~ ~~ '-

Reactor safety systems are those systems, including their associated input channels,

, which are designed to initiate automatic reactor protection or to provide information for the initiation of manual protective action.

- 1,49, Reactor Secured '

The reactoris secured when either: ,,

.--- 1.19.1

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4 It contains insufficient fissile material or moder1ttorpresent in the reactor, control rods, or adjacent experiments to attain criticality under optimum available conditions of moderation and reflection, or 4

1.19.2

. .. All of the following conditions are met

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'The minimum number of neutron absorbidg control rods are fully insert'ed such that the reactor is subcritical by a margin greater *than 0.75% Ak/k ($1.00) in -

the reference corecondition with all experiments accounted for.

b, The console key switch is in thepff position, and the key is remoheh from

, the console and under the control of a certified operator or stored in a locked storage -

l area. .

c. No work is in progress involving core fuel, core structure, installed control

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. rods, orYontrol rod dnves.unless thb'y are physically decoupled from the control

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.. rods. .

, d. No experiments inIir near the' reactor are being moved or serviced that nye, on move;nent, a reactivity worth exceeding the maximum allowed ~focaeingle l

experiment or 0.75% Ak/k ($1.00)which eve ^rls,synaller.' "

1.20 Reactor Shutdown 1

... Th'e' reactor is sh'utdown when it is subcritical~by a'~ margin greater than 0.75% Ak/k

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l ($1.00) in the reference core condition with all experiments accounted for. .

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.e Reed Reactor Facility Technical Specifications Page 8 1.21 Reference Core Condition .

The condition of the core whin it is at ambient temperatum,(cold) and the reactivity worth of xenon is. negligible (less than 0.05% Ak/k (<$0.07)).

' 1.22 Research Reactor -

'A research reactor is a device designed to support a self-stustaining neutron chain reaction for research, development, educational, training, or experimental purposes, and which may have provisions for the production of radionuclides.

Rod. Control 1.23 A control rod is a device fabricated from neutron absorbing ma'terial which is used '

to establish neutron flux changes and to compensate for routine reactivity losses. A control rod may be coupled to its drive unit allowing it to perform a safety function when the couplingis disengaged. ,

143.1 Regulating Rod ,

A regulating rod is a control rod used to maintain an intended power level and may be varied manually or by a servo-controller. The regulating rod shall have scram

. capability. ,

1.23.2 Safety Rod A safety rod is a control rod having an electric motor drive and scram capabilities.

i 1.23.3 ShimRod

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A shim rod is a control rod having an electric motor drive and scram capabilities. A shirri' rod may be' varied manually or by a servo-controller.

1.24 Safety Limit , , ,

1 Safety limits are limits on important pr6 cess variables which are found to be j necessary to protect reasonably the integrity of the principal barriers which guard c , against the uncontrolled release of radioactivity. The principal banier is the fuel element cladding., '-

1.25 &Iam k scram is any condition or event causing interruption of the magnet current to the control rods'immediately. shutting down the' reactor.

,1.25.1 Inadvertent Scram. .

An inadvertent scram is a unscheduled shutdown when the reason for the unscheduled' shutdown is known (eg. missed a range switch operation).

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1.25.2 Unexplained ~ Scram An unexplained scram is a unscheduled shutdown the cause of which cannot be immediatel ' determined. .

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1.26 Scram Time

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Scram time is the elapsed time between reaching a limiting safety system setting and a specified controlrod movement.

1.27 Shall. Should. and May The word shall is used to denote a requirement. The word should is used to denote a recommendation. The word may is used to denote permission, neither a requirement nor a recommendation.

1.28 Shutdown Marcin Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made or maintained subcritical by means of the control and safety systems starting from any permissible operating condition

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

1.29 Shutdown. Unscheduled ..

An unscheduled shutdown is any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operation, not including shutdowns which occur during testing or check-out operations. .1

, 1.30. Startup -

Startup is the sequence of procedures and operations to be completed whenever the reactor is to be taken from a Reactor Secured condition.

' 1.31' Surveillance Activities -

Surveillance activities will usually have a prescribed frequency and scope to demonstrate performance of systems required under Limiting Conditions for Operations.

In general, two types of surveillance activities are specified, operability checks'and calibrations. Operability checks are generally specified as monthly to quarterly.

Calibrations are generally specified as annually to biennially.

1.32 Time Intervals ' ,

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To provide operational flexibility, where time intervals for surveillance and audit activities are specified in the document, maximum intervals shall not exceed 125%

of the specified interval. Established frequencies shall be maintained over the long term.

June 1986 edition P

R;ed Reactor Facility Technical Specifications Page 10 Surveillance activities (except those specifically required for safety when'the reactor is secured) may bp deferred when the reactor is secured, however, they shall be completed prior to reactor startup. Surveillance activities scheduled to occur during an operating cycle which cannot be performed with the reactor operating may be deferred to the end of the cycle.

1.33 Value. Measured The measured value is the value of a parameter as it appears on the output of a channel.

1.34 Value. True The true value is the actual value of a parameter. .

1.3S Zero Power Critical ,

The reactor is zero power critical when the reactor is exactly critical and the reactor linear power channel reads less than or equal to ten (10) watts. This is the operational point where the excess reactivity, or core excess, is measured during reactor operations.

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

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2.1 Safety Limit -

Applicability .

- This specification applie,s to the reactor power.

~ Objective -

! The objective is to define the maximum reactor power that can be permitted with confidence that no damage to the fuel element cladding will result.

Specification (s) ,

The maxiinum reactor po'wer shall not exceed 250 kilowatts. However, for the

, purpose of testing the 110% full power safety scrams, an exception shall be made to allow the reactor to be operated at power levels not to exceed 287.5 kilowatts during the testing period. -

t Basis t

The safety limit for the standard TRIGA fuel is based on calclulations and expenmental eyidence. The results indicate that the stress in the cladding due to 1

hydrogen pressure from the dissociation of zirconium hydride will remain below l _

the ultimate stress provided that the temperature of the fuel does not exceed 1150 C and the fuel cladding does not exceed 500*C.

, ,4, Appendix E of"THE REED COLLEGE REACTOR FACILITY (TRIGA MARId)

SAFETY ANALYSIS REPORT, April 15,1967" (SAR) gives 225*C as the .

I approximate maximum fuel temperature for operation at 250 kilowatts. This

. conservative limit will assure a cladding temperature less than 500 C under all '

design basis accident conditions (step insertion of all available excess reactivity or instantaneous less Of Cooling Accident). Sec. tion 2.1 of the SAR gives 150 C as l

the maximum fuel temperatum for an instantaneous Loss of Cooling Accident after operation at 250 kilowatts for infinite time prior to the accident. The equilibrium pressure resulting from fission gases, entrapped air, and hydrogen at 150 C is less A

than 30 psi. Section 7.1 of the SAR gives a maximum measured fuel temperature less than 500 C for a 2.25% Aluk ($3.00) step insertion.

Thermal and hydraulic calculations indicate that standard TRIGA fuel elements may be safely operated at po'wer levels in exc~ess of 1500 kilowatts with natural-convection cooling. Details on the performance of TRIGA fuel are given in the c SAR and in papers available as " Fuel Elements for Pulsed TRIGA Research Reactors", Simnad, er al., Nuclear Technology,2.1, 31 (January,1976), and "The ,

U-ZrH xAlloy: . Its Properties and Use'in TRIGA Fuel", GA Project No. 4314 Report E-117-833, General Atomic Company, P.O. Box 81608, San Diego, CA 92138,1980.

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2.2 Limiting Safety System Setting ,

2.2.1 Powerlevel +. .-

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Applicability - - '

This specification applies to the protective action for the reactor during operation. ~

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Objective #

The objective is to specify the maximum reactor power ihat can be permitted with confidence that no damage to the fuel cladding will result. '

Specification (s) , ,

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a. The maximum operating power level for the continuous operation of the reactof slitill be 250 kilowatts as measured by the linear or % power channelst -

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b. The linear and % power channels shall be calibrated so that the measured g value is within 10% of the true value as determined by calorimetry.

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c. The linear power and % power channels shall initiate a scram at 110% of 250 kilowatts (275 kilowatts).- . ~,

d. For the ?urpose of testing the 110% full power safety system set points, an exception shall x made to allow the reactor to be operated at powerlevels not to -

exceed 287.5 kilowatts'during the testing period. .

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Basis . ,-

See basis for Section 2.1. .,

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3.0 uMmNG CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters ,.- -

3.1.1 Excess Reactivity Applicability .

This specification applies to the reactivity of the reactor core in terms of the available excess reactivity above the cold, xenon free, zero power critical conditio_n.

Objective .

The objective is.to prevent the reactor safety limit from being reached by limiting the potential reactivity availabig in the reactor.

Specifigations(s)

- Maximum excess reactivity shall be 2.25% Ak/k ($3.00) with experiments in Pl ace. ,.

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. Basis ,

Maximum excess cose reactivity is sufficient to provide the core rated power, xenon compensation, and reactivity for shutdown. Analysis of the reactor core demonstrates that no single pomponent represents sufficient potential reactivity to reach tiie reactor safety limit during any condition of operation. (SAR Sections 2.2 Reactivity Insertion and 7.1 Reactor Power Transients)

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3.1.2 Shutdown Margin. '

, Applicability This speification applies to the reactivity margin by which the reactor core will.be contricered shutdown.

Objective

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The objective is 'to ass'ure that the reactor can be shut down safely by a margin that is sufficient to compensdte for the failure of a control rod or the movement of an experiment. . ...

Specification (s) .

The' reactor shali not be operated upless the shutdown margin provided by control- 5 .

rods is gre'ater than'0'418 Ak/k ($0.53) with:

a. The reactor in the reference core conilition. -

'b.. '.The most reactive control rod fully withdrawn. . -

c.- The highest worth, movable experiment in its most reactive sta'te.

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Basis

' 'Ihe value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the highest worth contml rod should remain in the fully withdrawn position and a movable experiment is in a high reactivity state.

3.1.3. FuelElements -

Applicability" This specification applies to the fuel elements.

Objective

. The objective is to ensure the physicalintegrity of the fuel element cladding.

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Specification (s) ,.

The reactor shall not be operated with damaged fuel. A fuel element shall be cotisidered damaged and must be removed from the core and stored in accordance with Section 5.4 if:

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a. A visual inspection reveals deterioration of fuel element cladding.

b. The fuel element does not enter the fuel inspection tool.

c. A clad defect exists as indicated by release of fission products.

Basis De performance of TRIGA fuel elements under RRF operating conditions has been evaluated in the documents refercaped in the Basis for Section 2.1.

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3.1.4 Core Configuration Applicability This spe9i fication applies to the configuration of fuel elements, control rods, experiments and other reactor grid plate components.

Objective The objective is to anure that provisions are made to restrict the arrangement of fuel elements and experiments to provide assurance that exc'essive power den.sities will

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not be produced. .

Specification (s)

He core shall be an assembly of TRIGA Mark I aluminum clad and/or stainless-steel clad fuel moderator elements arranged in a close-packed array except for:

a. replacement of single individual elements with in core irradiation facilities or 2 -

control rods. .

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b. two (2) separated experiment locations in the D through F rings, each occupying a maximum of three fuel element positions.

c. unoccupied grid plate positions may contain graphite filled dummy elements to increase moderation and reflection.

d. the reflector (excluding experiments and experimental facilities) which shall be water or a combination of graphite (clad in aluminum) and water.

e. the startup source may occupy an F ring position.

Basis Standard TRIGA cores have been in use for years, and their characteristics are well -

documented in the publications referenced in the basis for Section 2.1. The Specific RRF configuration has been evaluated in the SAR.

3.2 Rem tor Control and Safety System 3.2.1 Control Assemblies Applicability This specification applies to control rods.

Objective The objective is to ensure that the control rods are operable.

Specification (s) ,

, The reactor shall not be operated unless the control rods are operable, and

a. Control rods shall not be operable if damage is apparent to the drive assemblies, or if the cladding has been breeched.

b. The scram time measured from the instant a simulated signal reaches the value of a limiting safety system setting to the instant that the slowest scrammable control rod reaches its fully inserted position shall not exceed 1 second.

c. Maximum reactivity insertion rate of a control rod.shall be less than *

. 0.12% Ak/k ($0.16) per second. .' .

Basis ' .

. The apparent cond'i tion of the control rod assemblies willprovide assurance that the rods will continue to perform reliably and as designed. The specification for rod scram time assures that the reactor will shut down promptly when a signal initiating' e a scram is generated. The specification for rod reactivity msertion rates assures that the reactor will~ start up controllably when rods are'withdra'wn. Analysis has indicated that'for the range of transients anticipated forit 'l'RIGA reactor the June 1986 edition ,

• 6 .

Reed Reactor Facility Technical Specifications Page 16 specified scram time and insertion rate is adequate to assure the safety of the reactor. (SAR Section 7.1 Reactor Power Transients) 3.2.2 ReactorControlSystem .

Applicability These specifications apply to logic of the reactor control system.

Objective The objective is to specify the minimum control system interlocks that shall be operable for operation of the reactor.

Specification (s)

The following control system safety interlocks shall be operable:

a. CountRate Interlock Withdrawal of any control rod shall be prevented if there are less than 2 neutron counts per second in the Count Rate Channel.

b. Rod Raising Interlock Simultaneous withdrawal of 2 or more control rods shall be prevented.

Basis Interlocks are specified to prevent function of the control, rod drives unless certain ,-

specific conditions exist. The interlock to prevent startup of the reactor at power

! levels less than 2 neutron counts per second assures that sufficient neutrons are available for controlled reactor startup. The interlock to limit the maximum positive reactivity insertion rate prevents simultaneous withdrawal of most than one control rod.

3.2.3 Reactor Safety System .

l Applicability

These specifications apply to operation of the reactor safety syst,em. ,

l

, , Obje.ctive

^

De objective is to specify the minimum safety system scrams '*

whith shall be '~

operable for the operation of the reactor. *

' ~

Specification (s) . ,- '

He following contr'ol rod scram safety channels shall be operable:, -

a. The Linear Power channel. ,

b. The % Power channel. ,, .,

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. l Reed Reactor Facility Technical Specifications Page 17 .

c. He Manual Scram Bar on the control console shall initiate a scram on demand.

Basis Manual operation of the reactor safety system is considered part of the protective action of the reactor safety system. Automatic control rod insertion assures compliance with the limiting safety system setting in Section 2.2.

3.2.4 ReactorInstrument System Applicability These specifications apply to measurements of reactor operating parameters.

Objective

l The objective is to specify the minimum instrument system channels that shall be operable for operation of the reactor.

Specification (s)

The following minimum reactor parameter measuring channels shall be operable:

a. Linear Power Level

b. Percent PowerIzvel l

\ l

c. Neutron Count Rate.

l Basis

he minimum measuring channels are sufficient to provide signals for reactor control and automatic safety system cperation. Measurements of the same or different parameters provide redundancy.

3.3 Ooerational Supoort Systems 3.3.1 WaterCoolant Systems

,' .. Applicability -

This specification applies to the operating conditions for the reactor pool and

- coolfht water systems. , , .

. Objective . -. .

= -,

The' objective is to provide shielding 6f the reactor radiation, protection against "

o corrosion of the reactor components, cooling of the reactor fuel, and to prevent . .

. ' leakage from the primary coolant system. . , _

~

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R:ed Reactor Facility Technical Specifications Page 18 Specification (s) -

Corrective action shall be taken or the reactor shut down if the following reactor coolant water conditions are observed:

a. The bulk pool water temperature exceeds 48*C (120 F).

b. The water depth is less than 6.0 meters (20 feet) measured from the top grid plate to the pool water surface, or 30 cm (1 foot) measured from the bottom of the bridge to the pool water surface.

c. The water electrical conductivity is greater than 2.0 mho/cm <

averaged for measurement periods of one month.

d. During heat exchanger operation, the pressure in the secondary system (measured at the secondary basket filter outlet) is less than 35 kPa (5 psi differential) greater than the pressure in the primary system (measured at the primary filterinlet).

Basis

a. The bulk water temperature constraint assures that sufficient core cooling exists under all anticipated operating conditions and protects the resin of the water purification system from degradation or deterioration.

b. A water depth of 6.0 meters (20 feet) above the top of the core grid plate is sufficient so that radiation levels above the reactor pool are at reasonable levels.

c. Average measurements of pool coolant water conductivity of 2.0 pmho/cm assure that water purity is maintained to control the effects of corrosion and activation of coolant water impurities.

d. A pressure difference at the secondary basket filter outlet and the primary .

filter inlet of 35 kPa (5 psid) will be sufficient to prevent loss of pool water from the lower pressure primary reactor coolant system to the higher pressure secondary water system in the event of a leak in the heat exchanger.

. 3.3.2 AirConfinementSystems Applicability L

This specification applieg to the' air. ventilation conditions in the reactor bay or'

. expenmental facilities during reactor operation.- -

.. Qbjective"

~

The objective is to control the release of air from the reactor bay or experimental facilities. *

. Specification (s)- ,

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Reed Reactor Facility Technical Specifications Page 19 ,

The reactor shall not be operated unless minimum conditions for air confinement are functional. 'Ihe following mimmum conditions shall exist:

a. Equipment shall be operable to isolate the reactor bay by closure of ventilation supply and exhaust dampers.

b. The double doors shall be closed and barred; the emergency exit i

door shall be closed and locked (the door shall be equipped with an emergency release mechanism); and the door to the control room shall be closed except forpersonnel access.

c. Upon detection of elevated radioactivity levels by the Continuous Air Monitor or the Gaseous Stack Monitor, the ventilation system shall automatically close supply air dampers and initiate restricted air exhaust from the reactor bay in order to maintain a negative pressure relative to ambient conditions. Air released during this restricted air exhaust shall be filtered through high efficiency particulate adsorption filters. --

Basis Tlie specifications for exhaust ventilation and confinement of the reactor bay provide control for airborne radioactive releases during both routine and non-routine operation.

3.3.3 Radiation Monitoring Systems Applicability This specification applies to the radiation monitoring conditions in the reactor bay during reactoroperation.

Objective '

The objective is to monitor the radiation and radioactivity conditions in the area of the reactor.

I Specification (s)

The reactor shall not be operated unless minimum conditions for radiation

, measurement are operable. The following minimum conditions shall exist:

h a. A Continuous Air Monitor capable of detecting beta and gamma

, radiatio'n in the air above the. pool shall be operable with readout and audible alarm..

~

~

b. An Area Radiation Monitor capable of' detecting gamma radiation

• above the pool shall be operable with readodt and audible alarm. -

l c. A portable surv- l meter capable of detecting Ib kBq (microcurie) levels of, beta or gamma radiation shall be operable. ,

d. A portable ion chamber monitoring <!evice or'e4uivalent non- ,

. saturating personnel dosimetry instrum'ent capable of determining beta and gamma exposure dose rate shall be operable.

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R:ed Reactor Facility Technical Specifications Page 20

e. The portable ion-chamber type radiation monitor may be substituted for the Area Radiation Monitor during periods of maintenance or repair.

f. The Gaseous Stack Monitor may be substituted for the Continuous Air Monitor during periods of maintenance or repair.

Basis The radiation monitors provide information to operating personnel of impending or existing hazards from radiation. This should provide sufficient time to evacuate the facility or take the necessary steps to maintain the exposure of personnel as low.as practicable and to control the release of radioactivity. The Gaseous Stack Monitor mitiates confinement upon alarm as does the Continuous Air Monitor. Therefore, substitution during maintenance or repair provides the same capability to initiate

. confinement without operator intervention. Personnel exposure dose rates should only be measured with an ion-chamber or equivalent type monitor. A survey meter should be used only for detection of contammation.

3.4 Limitations on Exneriments 3.4.1 Approval and Conductof Experiments Applicability This specification applies to all experiments involving the reactor. ,

Objective The objective is to ensure the safety of the reactor and its components during the .--

performance of any experiment.

Specification (s)

a. Prior to performing any experiment, the proposed experiment or class of experiments shall be approved as provided in Section 6.4.

b. All experiments shall be carried out in accordance with established and approved written procedures. Minor changes to written procedures that do not significantly alter the experiment may be made by a Class A Operator provided these changes are documented.

Basis

,' The overriding consideration of reactor safety requires a thorough review and approval of proposed experiments prior to performing them.-

, 3.4.2 Reactivity.

~

Applicability This specification applies to the reactivity associated with experiments.

Objective .

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Reed Reactor Facility Technical Specifications Page 21 .

The objective is to control the amount of reactivity associated with experiments to values that will prevent the reactor safety limit from being exceeded.

! Specification (s)

The reactor shall not be operated unless the following conditions goveming experiment reactivity exist:

a. Any movable experiment shall have a reactivity worth less than 0.75% Ak/k ($1.00).

b. Any secured experiment shall have a reactivity worth less than

, 1.01% Ak/k ($1.35).

c. The total reactivity worth ofin-core experiments shall not exceed 1.50% Ak/k ($2.00). This shallinclude the potential reactivity which might result from malfunction, flooding, voiding, or removal and insertion of the experiments.

d. No experiment shall be performed if failure of such experiment could lead to a failure of a fuel element or of other expenments and these associated failures could result in a measurable increase in reactivity or a i measurable release of radioactivity.

Basis

a. The worth of a single movable experiment is limited so that sudden removal movement of the experiment will not cause prompt criticality. The limited worth of a single movable experiment will not allow a reactivity insertion that would exceed the reactor safety limit.

b. The maximum worth of a secured experiment is limited so that the reactor safety limit will not be exce:ded by removal of the experiment. Since these experiments are secured in place, removal from the reactor operating at full power .

would result in a relatively slow power increase such that the reactor protective systems would act to prevent excessive power levels from being attained.

c. The maximum worth of experiments is limited so that removal of the total worth of all experiments will not exceed the reactor safety limit.

d. The interaction of all experiments in the reactor is to be considered to assure the safety of the reactor under all anticipated operating condidons.

3.4.3 Materials .

Applicability These specifications apply to experiments (as defined in Section 1.5.a) installed in the rextor and its experimental facilities. .

Objective June 1986 edition

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R:ed Reactor Facility Technical Specifications Page 22 The objective is to prt; vent the release of radioactive material in the event of:an experiment failure, either by failure of the experiment or subsequent damage to the reactor components.

STecification(s)

I The reactor shall not be operated unless the following conditions goveming expenment materials exist:

a. Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials, and liquid fissionable materials shall be doubly encapsulated.

b. Each experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 Curies and the maximum strontium-90 inventory is no greater than 5 millicuries.

c. Explosive materials shall not be irradiated in the reactor or j , experimental facilities.

. - d. - Experiment materials, except fissionable materials, whicli could off-

! gas, sublime,'volatize, or produce aerosols under:

1) normal operating conditions of the experiment or reactor,

! 2) credible accident conditions in the reactor, 1

3) possible accident conditions in the experiment - '

j , shall be limited in activity such that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor bay or the irradiation facility atmosphere, the airborne concentration of radioactivity released

,

• averaged over a year would not exceed the limits of Appendix B of 10CFR20.

In calculations pursuant to the above, the following assumptions shall be used:

-(1) If the effluent from an experimental facility exhausts through a system which closes automatically on high radiation level, at least 10% of . .

the gaseous activity or aerosols produced will escape.

(2) If the effluent from.an experimental facility exha'usts through a

~

filter installatiod'desighed for great 6r than 99% efficiency for 0.25 micron

' particles, at least.10% of these particles can escape. - -

1 (3) For materials whgse boiling point is above 55*C (130'E) and where vapors formed by boiling this thaterial can escape only through an -

~

. undis,turbed column of water above the core, at least 10% of these vapors can escape. ~ ' --

June 1986 edition . .

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. .-v.,. - - -_,, ,..-- _,.  %,. ,-,_.___,_____,._m__y_ _._ ,, ,,__-., , , . -_,___ .,,. . _ . e_- , _ _ , _ - . - . - . . _ , . _ _ . _ _ , . ,

, Reed Reactor Facility Technical Specifications Page 23 .

Basis .

a' Double encapsulation is required to lessen the experimental hazards of some '

types of materials. *

b. The 1.5-Curie limitation on iodines 131 through 135 assures that in the -

event.of failure of an experiment leading to total release of the iodine from the ,

experiment, the exposure dose at the exclusion ama boundary from iodine-131 does not exceed the limits of Table II, Appendix B,10CFR20 averaged over one year.

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

, msulting from failure of an expenment involving explosive materials.

d. This specification is intended to reduce the likelihood that airborne activities in excess of the maximum allowable limits will be released <o the atmosphere outside the facility boundary. Guidance for the calculations is provided.

3 3.4.4 Failures and Malfunctions of Experiments Applicability ~

These specifications apply to the design of experiments and to actions to be taken ,

upon experiment failure or malfunction.

Objective The objective is to limit the consequences of experiment failure or malfunction. it '

Specification (s)

a. Credible failure of any experiment shall not result in releases or exposures in excess of established limits nor in excess of the limits established in Table II, Appendix B,10CFR20 averaged over one year.

b. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, removal of the capsule and physical -

inspection of the reactor shall be performed to determine the consequences and need .

for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Director and determined to be satisfactory befom operation of the reactoris resumed. ~

' Basis

' ' a.

' Experiments shall be designed to liniit release of ratiioactivity under all -

credible accident conditions:' '* .

, b. Operation'of the reactor with the reactor fuel or structure daniaged is -

,prohibitqd to atoid release of fission products. -.

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4.0 SURVEILLANCE REQUIREMENTS 4.1 Renctor Core Parameters -

4.1.1 Excess Reactivity Applicability -

This specification applies to the measurement of reactor excess reactivity or core Sxcess.,

Objective The objective is to periodically determine the changes in core excess reactivity available forpower generation. -

Specification Excess reactivity shall be determined at zero power critical as part of the startup procedure.

Basis This specification assures determination of excess reactivity after all reactor core or control rod changes and after experiment installations. This specification monitors changes in the core excess reactivity as an indicatjon of the condition of the reactor ,

core and to insure compliance with excess reactivity limits in the Techobal Specifications.

4.1.2 Shutdown Marg'in Applicability This specification applies to the measurement of reactor shutdown margin.

Objective The objective is to periodically determine the core shutdown reactivity available for reactor shutdown.

Specification (s) ,

O Shutdown margin shall'lx: determined samiannually, after fuel inoveme,nt, or ,

f, . control rod removal and replacement.- .

Basis -

Semiannual determination of shutdown margin and measurements after 'reactor core or control, rod changes are, sufficient to monitor significant changes in the core - -

shutdown margin. . . . .

4.1.3 FuelElements ' '

Applicability .

~

Juile1986 edition -

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Reed Reactor Facility Technical Specifications Page 25 . .

His specification applies to the inspection requirements for the fuel elements.

l ' Objective ..

l He objective is to visually inspect the physical condition of the fuel element i cladding.

Specification (s) i At least 1/5,of all the fuel elements in the core shall be visually inspected each year with the fuel elements to be inspected rotated such that each fuel element shall be

~

' inspected at least once every five (5) years.

Basis .

t He frequency ofinspection is based on the parameters most likely to affect the fuel

,i cladding of a reactor operated at moderate power levels and udlizing fuel elements .

whose characteristics are well known as given in the references for Section 2.1.

4.1.4 Core Configuration .-

1 Applicability I his specification applies to the inspection requirements of the core configuration.

r .

! Objective -

l The objective is to ensure proper core configuration prior to operating the reactor.

Spegi,fication(s) .

He reactor core configuration shall be visually inspected as part of the startup

! procedures prior to reactor operation. ,

i Basis i Inspection for changes in core configuration and determination of proper core l . configuration for operation are accomplished as part of the startup procedures.

4.2 Reactor Contml and Safety Svitam ,,

, 4.2.1 Control Assemblies , .x . , . _ . ~.

l Applicability .

. This specification ' applies 40 the surveillance'of the control fods/ .' -

~ .

l '. Objective .

The objectives are to dicasure the control rod wo'rths, to inspect the physical

. condition of the reactor control rods, and to' establish the operable condition of_the -

p. . , control rods by periodic measurement of the scrarh times and insertion rates. , . .

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' Reed Reactor Facility 7echnical Specifications Page 26 ,

~

Specification (s) , N ,

. Control rod. worths shall be determined semiannually or after significant core or controlrod changes, and ,

, a. Each control rod shall be dsually inspe'cted at biennial intervals.

b. The scram time of each control rod shall be measured semiannually.

, t. The reactivity insertion rate of each control rod shall be measured

.. , annually. ~

.. Basis ,

Semiannual determination of control rod worths or measurements after significant core changes provide information'about changes in reactor total re' activity and

, individual rod worths. 'Ihe fiequency ofinspection for the control rods will

• provide periodic verification of the condition of the control rod assemblies.

Verification will.be by measurement and visual observation of absorber sections plus examination of linkages and drives. The specification intervals for scram time and ipsertion rate assure operable performance of the rods. ,

4.2.2 ReactorControlSystem -

Applicabilit - ' '~ .

'This sp'ecification applies to the tests'of the logic of the reactor controfsystem. '

Objective -

3 -

The objective ~is to specify intervals for tests of the minimum control system

' interlocks

].

Specification (s)

The minim.um safety interlock channels shall be tested prior to startup as part of the

startup procedure.

Basis The routine test of the interlock logic at startup provides adeqeate information that Jhe control system interlocks are operable. -

,, ' 4.1.3 Reactor Safety System '

Applicability

.. This specNication applies to test and calibration of the reactor safety system. .

.g

• Objective '

, The objectivp is to-specify intervals for test and calibration of the' minimum safety -

system scrams. -

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I Reed Reactor Facility Technical Specifications Page 27 ' ,

Specification (s)

The minimum safety channels shall be calibrated annually and tested prior to each startup as part of the startup procedure.

Basis .

The periodic calibration at annual intervals provides adequate information that the setpoints of the safety system scrams are accurate. Tests of the safety system prior to each planned operation assure that each intended scram function is operable. .

4.2.4 ReactorInstrument System ,

Applicability These specifications apply to calibrations and tests of reactor measurement channels. -

Objective The objective is to specify intervals for calibrations and tests of the minimum instrument channels. . -

Specification (s)

The minimum instrument channels shall be calibrated annually. Calibration of the linear and %-power channels shall be by the calorimetric method. A test of each channel shall be made prior to each startup as part of the startup procedure.

Basis Annual calibration ofinstrument channels is scheduled to allow adjustments for -

changes in reactor and instrumentation parameters. Tests are applied prior to reactor operation to verify each system is operable.

4.3 Onerational Suonort Systems 4.3.1 Water Coolant Systems Applicability This specification applies to surveillance of the reactor pool and coolant water systems. .

Objective

~

'Ihe objective is to maintain the reactor coolant conditions within acceptable specifications.

Specification (s)

The following measurements shall monitor the reactor coolant conditions:

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Reed Reactor Facility Technical Specifications Page 28

a. The water temperature channel shall be calibrated annually and monitored continuously during reactor operation.

b. The pool level channel shall be tested bimonthly, and monitored ,

continuously during operation of the reactor.

c. The pool water conductivity channel shall be calibrated annually and the electrical conductivity shall be measured weekly.

d. The secondary low pressure channel shall be tested semiannually and monitored contmuously during operation, c .  !

Basis Periodic calibrations and tests of measurement devices for the reactor coolant system parameters assure that the coolant system will perform its intended function.

4.3.2 AirConfinementSystems Applicability This specification applies to surveillance of the air confinement system in the reactor bay .

Objective -

The objective is to demonstrate that the air confinement system is operable and that airborne releases of radioactive material are properly quantified.

Specification (s)

. The following actions shall demonstrate the air confinement conditions: -

a. Annual visual examination of isolation dampers,

b. Bimonthly tests of air confinement system operation.

c. Bimonthly visual examination of facility doors and closing - '

mechanisms.

d. . Annual calibration of the Gaseous Stack Monitor and air confinement trip points using Argon-41 and semiannual tests.

e. Annual calibration of the Continuous Air Monitor.

f. Weekly tests of the alarm set points of the Continuous Air Monitor.

, Basis ,

Periodic evaluations of air confinement criteria are determined by examination, test, and calibration of the appropriate ventilation functions. The air confinement system provides control for radioactive releases during both routine and non-routine operating conditions.

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. -Reed Reactor Facility Tech'nical Specifications Page 29 .

4.3.3 Radiation Monitoring Systems .

Applicability This specification applies to the, surveillance of the radiation monitoring chantiels. ,

Objective -

l The objective'is to assure the radiation monitoring systems are operable. .

Specification (s)

Surveillance of the minimum-radiation monitors specified to be operable during reactor opention shall be performed as follows: '

a. The Air Particulate Monitor and Radiation Area Monitor shall be calibrated at annualintervals'.

~

b. The portable ion chamber (s) and ponable survey meter (s) shall be

' calibrated at semiannual intervals.

l c. . - The alarm set points of the Radiation Area Mon'itor shall be tested at weekly intervals.

d. The portable ion chamb'er(s) and portable survey meter (s) shall be tested as part of the startup. procedure.

Basis r Periodic calibrations and frequent tests am specified to maintain reliabl'e -

performance of the radiation monitoring instruments.

4.4 Limitations on Experiments' *-

4.4.1 Approval .

Applicability

, This specification applies to surveillance of prior approval for all experiments involving the reactor.

Objective '

' The objective is to ensure no experiment is performed without prior review and approval as given in Section 6.4.

Specification (s')

~

No experiment using the mactor shall be performed without a copy of a procedure approved as given in Section 6.4 in the control room.

Basis June 1986 edition

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R:ed Reactor Facility Technical Specifications Page 30

,The Reactor Super

• visor and Reactor Operators shall only use an approved procedure for conduct of an experiment.

4.4.2 Reactivity Applicability ,

. This specification applies to surveillance of the reactivity of experiments.

Objective The objective is to assure the reactivity of an experiment does not exceed the allowable specification.

' Specification (s)

The reactivity of any experiment designed to be performed with the reactor operating shall be measured at zero power critical before the experiment is performed. This specification may not apply to pneumatic tube experiments at the '

discretion of the Director with the concurrence of the Reactor Safety Committee.

9.4tJ';

~

Basis .

The measured reactivity or determidation that the reactivity is not significant will

~

provide data that the' configuration of the experiment or experiments is allowable.'

4.4.3 Materials a : .-

Applicability This specification applies to the surveillance requirements for materials inserted into the reactor.

Objective .

.. ; seu The objective is to prevent the introduction of materials that could damage the reactor orits components.

Specification (s) ,

Any surveillance conditions or special requireinents shall be specified as a part of the experunent approval.

Basis An evaluation of all experiments is performed to classify the experiment as an approved experiment.

4 Juna1986 edition

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R:ed Reactor Facility,, Technical Specificatipns Page 31 '

5.0 DESIGN FEATUFFf

^

5.1 Site and Facility Description

~

5.1.1 Location Applicability J '-

'This specification applies to the Reed Reactor Facility location and specific facility design features. "

l ..

, Objective ,

The objective is to specify those features which are related to the Safety Andlysis evaluation.

s ., . .

l ~ Specification (s) ,

~

- a. The Reed Reactor Facility is in'the northeast part of the Reed College *

• campus in the city of Portland, Multnomah County, Oregon. The 90 acre campus

.,., property.is ownec by the Reed Institute.

bNAThe TRIGA Mark I research reactor is installed in thepactor bay.

c. The feactor core'is assembled in a below ground shield and pool structure -

with vertical access to the core. -

. d. The restricted acc'es's area of the Reed Reactor Facility shall consist of the reactor bay, the mechanical room, and the reactor control room.  ;

. m . Basis

. a. The Reed Reactor Facility site is located in an area owried and controlled by the Reed Institute.

l b. The Reed Reactor Facility addition has been ' designed with characteristics related to the safe operation of the reactor. -

..,' c. The shield and pool structure has been designed for structural integrity I

below ground and for radiation levels approximately 1 mrem /hr at locations adjacent

.- to the reactor pool in the reactor bay. .

d.' The restricted access to specific facility areas assures that proper controls are *

. , established for the safety of the public and for the security of special nucleir

~ materials.

5.1.2 ' AirConfinement Applicability ~ '

This specification applies to the design features which control air released fro,m the ,

reactor bay. .

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4 Reed Reactor Facility Technical Specificatin Page 32 -

Objective " '

De objective is to assure that provisions'are made to contro1Ir mstrict the airbome release of radioactivity to the environ' ment. -

, Specification (s)

a. The reactor bay shall be designed to mstrict leakage and shall have a - .

minimum enclosed air volume of 340 cubic meters (12,000 cubic feet).

i

b. Under normal operating conditions, the ventilation system shall provi.de two (2) air changes per hour and shall maintain a slight negative pressQre infhe reactor bay relative to ambient conditions. -

e. Upon detection of a limit signal related to the radiation level, the air confinement system shall automatically restrict unfiltered air dxhaust as described in Section 3.3.2.b.

, d. All air or other gas exhausted from the reactor bay and from ass *ociated experimental facilities during reactor operation shall be released to the environment . .

at a minimum of 3.7 meters (12 feet) above ground level.

Basis

a. The enclosed air volume determines the concentration of airbome radionuclides in the reactor bay. .

b. Exchange of air in the reactor bay prevents the buildup of gaseous radioactivity. Maintaining a slight negative pressure in the reactor bay ensures that ,

air leaving the bay passes through monitoring systems and is released through the -

stack. 'y ,

c. Elevated radiation levels automatically prevent the uncontrolled release of ..

unfiltered air from the reactor bay as described in Section 3, Limiting Conditions of Operation. . . .

~

d. Release of air from the facility at a minimum of 3.7 meters above the ground -

surface provides for dispersion and dilution of releases.

5.1.3 Safety Related Systems Applicability .

This specification applies to any addition, modification, and non-routine modifying .

maintenance to any system related to reactor safety. ,

Objective

'* .. {

. \

The objective is to' assure the proper function of any system related to reactor

, safety.

l

. 1 Specification (s)  !

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R:ed Reactor Facility Technical Specifications Page 3a .

Any addition, modification, oi non-routine modifying maintenance to the core and

. its aspeiated support structure, the pool structure, the control rod drive mechanisms, the reactor safety system, the air confinement system, and the water coolant system shall be made and tested in accordince with the specifications to

, which the systems or components were originally designgd and fabricated, or to '

specifications approved by.the Reactor Safety Committee as suitable and not - ,

involving an unreviewed safety ques' tion. The reactor shall not be placed in ' .,

• ~~

bperation until the affected system has been verified to be operable. -

Basis

, . - Changes to the above systems could affect the safe operation of the reactor and ,

must be approved by the Reactor Safety-Committee including an analysis of any ' , ,

, . unreviewed safety questions (10CFR50.59).

5.2 Reactor Coolant System '

Appliegbility

.~ .

'Ihis specificatidn applies to the reactor coolant system.' -

~

Objective

' l *.

~

The objective is to assure that adequate water is available for cooling and shielding .

i during reactoroperation. , .

. Specification (s) -

a. Th$ reactor core shall be cooled by natural cohvectio.n of wa.ter.

b.- Pool water level shall be prdtected by holes in pool water system pipe iines # '

which act as siphon breaks..

. Basis , ,

. ' a.

Thermal and hydraulic calculations which show that a standard 85 element TRIGA core can operate in a safe manner at power levels specified for the Reed Reactor Facility are presented in the references given for Section 2.1, Safety Limits.

b. Siphon breaks prevent the loss of coolant \ vater caus d by inadvertant -

pumping or accidental siphoning. .

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5.3 Reactor Core and Euei ,

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5.3.l' FuelElements .

Applicability * -

This specification applies to the fuel elements used in the reactor core.-

, ' Objective -

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' Reed Reactor Facility Technical Specifications Page 34

. The. objective is to' assure that the fuel elements are designed and fabricated to

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._ permit their use with a high degree of reliability with respect to their physical and .

. nuclear characteristics.

Specification (s) # ..

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'Ihe standard TRIGA fuel' element at fabrication shall have the following .

characteristics:" - -

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a. -

Uranium content: 8.5 weight percent (wt%) uranium enriched to a nominal 19.7% Uranium-235. .

b. Aluminum Clad Standard TRIGA Fuel Elements: -

Zirconium to hydrogen atom ratio nominally 1:1

- Cla41ing: 0.030 inches of Aluminum. ... r-

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c. Stainitss Steel Clad Standard TRIGA Fuel Elements:

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. Zirconium to hydrogen ato'm ratio nominally 1:1.6 ,

. Cladding: .0.020 inches of stainless steel type 304.

/-

' d.' The length of a fuel element shallte '28.37 inches. .-

, ( e. - The diameter of a.fgel element shall be 1.47 inches. ,. -

.. Basis

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The Design Basis 6f the standard TRIGA core demonstrates that 250 kilowatt Jteady state operation represents a conservative safety limit for the maximum

, te.mperature generated in the fuel as presented in 'the references to Section 2.1 Safety. .

.g Limits. .

5.3.2 ControlRods. .

. , , ,' - Applicability 4 ,

[ This specification applies to the control rods.

Objective '

. . ,- - The objective is to assure that the control rods are designed to permit their use as i neutfon absorbers with a high degree of reliability and safety. ,

. 4 .

.- Specification ('s), "t -

, The. safety, shim, and regulating codtrol rods shall have scram capability, and shall

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co'ntain borated graphite, B4 C powder, or boron and its compounds in solid form . ;:

• as a neutron absorber which is encased in' aluminum cladding.

B.asisi , . . ..

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, he neutron absorbing requirements for the control rods are satisfied by using .

borated graphite, B4 C powder, or boron and its compounds. Rese materials must -

' be contained in a suitable clad material, such as aluminum, to insure mechanical stability during movement and to isolate the neutron absorber from the pool water .

' environment. Scram capabilities are provided for rapid insertion of the control rods whibliis.the pnmary safety feature of the reactor. "

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. 5.4 Reactor Fuel ElementStorage e -

Applicability '

This specification applies to the storage of reactor fuel at times when it is not in th'e reactor,qore.  ;

_ 05jective The objective is to assure that stored fuel will not become critichl and will not exceed design temperatures. _,

Specification (s)

a. All fuel elements shall be stored in a geometrical array where tee effective

-multiplication is less than 0.8 for all conditions of moderation.

b. Irradiated fuel eleme~nts and fueled devices shall be stomd in an array which ~

will permit sufficient natural convection. cooling by water or air such that the fuel

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element or fueled device temperature will not exceed design values.

Basis g The limits imposed by these specificitions am given in the " Technical Specifications for the Reed Reactor Facility,1968 edition", and are more conscryative than the Americah National Standards Institute ANSI 15.1_" Technical Specifications for Research Reactors,1982 edition."

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Reed Reactor Facility Technical Specifications Page 36 -

' 6.0 ADMINISTRATIVE CONTROLS ,

6.1 Oreani7ation 6.1.1 Structum i

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The Reed Reactor Facility (RRF) shall bc under the direct coritrol of the Facility Director (hereafter referred to as the Director). The management for operation of RRF shall consist of the organizational structure established as given Figure 6.1.1 RRF Organization Chart. ~

6.1.2 Responsibility ,

2

'Ihe Director shall be responsible to the President of Reed College through the RRF Management for the safe operation and maintenance of the reactor and its associated equipment. The Director's staff shall include a Reactor Supervisor, Class A Operators, and Class B Operators. The Director, or a designated appointee, shall review and approve all expenments and experimental procedures prior to their use in the reactor. Individuals of the RRF Management'and the Director shall be responsible for the policies and operation of the facility, and shall be msponsible for safeguarding the public and facility personnel from unnecessary radiation exposures and for adhering to the Operating License and Technical Specificati6ns.

In all instances responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications.

6.1.3 Staffing *

. The minimum staffing when the reactor is not secured shall be:

a. A Certified Operator in the control room. ..

b. A second person in the Reactor Facility who can perform prescribed written instructions such as initiation of the first stages of the. emergency plan including evacuation and initial notification procedures. In the event the absence.of the second person exceeds 15' minutes, the reactor shall be shutdown.

c. A designated Class A operator shall be readily available on call. The available operator shall be on the Reed Campus within ten (10) minutes of

reaching the Reactor Facility and shall keep the operator on duty informed

. of phone number for contact.

f Events requiring the direction of a Class A Reactor Operator shall be:

a. '

All fuel elements or control rod relocations within the reactor core region.

, b. Relocation of any in-core experiment with a reactivity. worth greater than 0.75% Ak/k ($1.00).

c. Recovery from an inadvertent scram. -

June 1986 edition

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Reed Reactor Facility Technical Specifications Page 36a RRF Manaaement Reed College President Vice President-Treasurer Vice President-Provost l

Reactor Safety Reed Reactor Facility

• Reactor Health Committee 4-* 4 Physicist Operations and D.irector Systems Subcommittee and RRF Health' Emergency, and Reactor ,

Superv,sor i

'. Security Subcommittee l RRF Class A Operators (Senior Reactor Operators) 4 Responsibility RRF -

Class B

< y Operators Communication  ;, (Reactor Operators) l 4

Figure 6.1.1 RRFORGANIZATION CHART

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June,1986 edition '..

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s- a a . _ _ _ .2 Re@d Reactor Facility Technical Specifications Page 37 A liet of RRF personnel by name and telephone number shall be madily available in the control room for use by the operator. This list shall include:

a.' Management Personnel

b. Health Physics Personnel

c. AllCertified Operators. .

6.1.4 Selection and Training of Personnel The selection, training, and requalification of operators shall meet or exceed the requirements of American National Standard for Selection and Training of Personnel for Research Reactors ANSI /ANS - 15.4. Qualification and requalification of certified operators shall be subject to a program approved by the Nuclear Regulatory Commission (NRC).

6.2 Reactor Safety Committee (Review and Audit)

. The Reactor Safety Committee (RSC) is established as a metitod for the independent review and audit of the safety aspects of Reed Reactor Facility (RRF) operations and to advise the President of Reed College regarding these matters.

6.2.1 Composition and Qualifications The RSC shall be composed of a minimum of five (5) members. The members, appointed by and reporting to the President of Reed Collei;e, shall collectively represent a broad spectrum of expertise in the appropriate reactor technology and, in addition, represent community interests in safe operation of the RRF. Individuals may be either from within or outside the operating organization. Qualified and formally approved alternates may serve in the absence of regular members. The Reactor Health Physicist shall be,a member of the RSC. The Chair of the RSC shall be responsible for:

w Calling and Leading Meetings Establishing the Meeting genda Disseminating Minutes to Members of the RSC, RRF Staff, and RRF Management.

Two subcommittees of the RSC shall be used to assess safety aspects of the RRF.

~ At least two members from each sSbcommittee shall serve on the RSC, including the subcommittee chairs. Each subcommittee shall consist of at least five members; no more than two members of each may be a current student at Reed College. The Chairs of the Subcommittees shall be responsible for:

Calling andleading meetings Establishing the meeting agenda Disseminating minutes of meetings to the Subcom:. ee, RSC, and RRF .

Staff. .

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R:ed Reactor Facility Technical Specifications Page 38 ,

The subcommittees am 6.2.1.a The Ooeratlons and Systems Subcommittee (OSS) shall deal with day-to-day operations of the reactor, its maintenance, reactor safety, and operator trainmg and requalification. Persons serving on this subcommittee shall have a background in reactor, mechanical, or electrical engineering, nuclear physics, nuclear chemistry, or other similar technical field. The subcommittee shall ensure that the technical concems of federal, l state, and private insurance agencies are answered in a timely and technically correct manner. The Operations and Systems Subcommittee shallroutinely review

OperationalProblems .

Maintenance, Plumbing, and Electrical Problems New Experiments FuelMovement Core Configuration Changes l Unexplained Scrams l- Startup Procedures OperatorTraming OperatorRequalification.

The following sh'all be routincly audited by the OSS:

l MainLogbook .

Maintenance Log

, OperatorLog Problem Log.

6.2.1.b The Health. Emergency. and Security Subcommittee

- (HESS) shall deal with emergency preparedness, health physics, radiation l safety, security, environmental impact of the RRF, and the interface t

between the RRF and the Reed Campus and surrounding communities.

• Members shall have a background dealing with emergencies, health care, l

environmental issues, or health physics, or be representative of surrounding community issues. The Health, Emergency, and Security Subcommittee shallroutinely review:-

Radiation Exposure Records

. ' Radiation Safety Security Emergency Drills Emergency Preparedness Interface Between RRF and External Regulating Agencies Radioactive MaterialTransfer Radioactive Waste Disposal .

Radioactive Material Releases from RRF l Community Affairs.

l The following shall be routinely audited by HESS:

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Health Physics Log i

Wipe-Test Log l -

Monitor Calibration Logs  ;.,

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R:ed Reactor Facility Technical Specifications Page 39 Security Log.

6.2.2 RSC Charter and Rules The review and audit functions shall be conducted by the RSC in accordance with

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an established charter including the following:

a. Meeting Frequency The RSC shall meet a least once per calendar quarter and mom fmc uently as circumstances warrant, consistent with effective monitoring of RR P activities. Each subcommittee shall meet at least once per calendar quarter.

b. Quorums A quorum for action by the RSC shall be not less than one half of the members where the operating staff, including the director, does not constitute a majority. The majority vote of the full RSC will be its official decision regarding safety aspects of the RRF. A quorum of each _

subcommittee shall consist of three members or one-half of the current members, whicheveris larger.

c. Use of subgroups Each subcommittee shall report to the RSC through its chair.

d. Dissemination, review, and approval of minutes in a timely manner

, RSC Meeting minutes shall be disseminated to members and to the President of Reed College for review in a timely manner after each meeting and approved by the RSC within the calendar quarter after each meeting.

6.2.3 Review Function The following items shall be reviewed for adequacy by the RSC:

a. Determinations that proposed changes in equipment, systems, tests, .

experiments, or procedures do not involve an unreviewed safety question (10CFR50.59 Review).

, b. All new procedures and major revisions thereto having safety significance, and proposed changes in reactor facility equipment or systems having safety significance.

c. All new experiments or classes of experiments that could affect reactivity or result in the micase of radioactivity.

d. Proposed changes in technical specifications or facility license.

e. Reports of violations of technical specifications or facility license, or violations of intemal procedures or instructions having safety significance.

f. Reports of: Operating abnormalities having safety significance;

reponable occurrences (violation of safety limits); release of radioactivity June 1986 edition

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i Reed Reactor Facility Technical Specifications Page 40

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from the site above allowed limits; operation with actual safety-system

~ settings less conservative than allowed in the Technical Specifications; operation in violation of Technical Specifications unless prompt remedial action is taken; reactor safety system component malfunctions which render or could render the associated system incapable unless the malfunction is 1

discovered during maintenance or pe:iods of reactor shutdown; unanticipated or uncontrolled change in reactivity greater than 0.75% Ak/k

($1.00); abnormal and significant degradation in reactor fuel, cladding,

. coolant boundary, or confinement boundary which could result in exceeding prescribed radiation exposure limits of personnel or release to the environment; and observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an ur.,afe condition.

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g. Audit reports.

6.2.4 A'uditFunction The subcommittee chairs shall perform or arrange for comprehensive examination of selected operating records, logs, and other documents. Discussions with cognizant personnel and observation of operations shall be used as appropriate. In no case shall the individualimmediately responsible for the area audit that area. The following items shall be audited:

1 . <.

, a. Facility operations for conformance to the Technical Specifications and applicable Facility License conditions, at least once per calendar year.

b. " The requalification program for certified operators at least once every other calendar year.

c. ' Results of actions to correct those deficiencies that may occur in reactor facility equipment, structures, systems, or methods of operation that affect reactor safety, at least once per calendar year.

d. <The RRF Emergency Plan, Physical Security Plan, and implementing procedures at least once every other calendar year.

Deficiencies uncovered in audits that affect reactcr safety shall immediately be reported to the President of Reed College. A written report of the findings of the

audit shall be submitted to the President and RSC members within three months

! after the audit has been completed.

6.3, Oceratine Procedures Written Standard Operating Procedures shall be prepared, reviewed, and approved by the Director, or a designated alternate, and the Reactor Safety Committee prior to initiation of the following activities,:

a. Startup, operation, and shutdown of the reactor.

b. Fuel loading, unloading, and movement within the reactor.

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, Rnd Reactor Facility Technical Specifications Page 41 c Routine maintenance of major components of systems that could have an

, effect on reactor safety. .

d. Surveillance tests and calibrations required by.the technical specifications or those that could have an effect on reactor safety.

e. Personnel radiation protection, consistent with applicable regulations.

f. Administrative controls for operations, maintenance, and the conduct of irradiations and experiments that could affect reactor safety. .

g. Implementation of required plans such as the Emergency Plan or Physical Security Plan. e

h. The Reactor Health Physicist shall be.resh[nsible for the development and implementation of appropriate Radiation Safety Procedures and Practices at the RRF. Such xocedums and practices shall encompass all operations and materials within the R RF and the adjacent radiochemistry laboratory. The interface between

, RRF Radiation Safety Procedures and Practices and those implemented by the Reed College Radioisotope Committee shall be through material transfer procedures and '

letters of agreement, where specific services may be performed between'the two' groups.

i. Additions, modifications, or non-routine modifying maintenance of reactor safety systems.

i Substantive changes to the above procedures shall be made effective after approval of the Director, or a designated alternate, and the Reactor Safety Committee: Minor modifications to the original procedures which 'do'not change the original intent may be made by a Class A operator, but the modifications must be approved by the Director or a designated alternate

! within 14 days. Temporary deviations from the procedures may be made by a Class A operator in order to deal with special or unusual circumstances or conditions. Such deviations shall be documented and reported to the Director or the designated alternate.-

6.4 Exneriment Review and Anoroval i All new experiments or classes of experiments shall be approved by the Director, or a i designated altemate, and the Reactor Safety Committee.

i

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

b. Substantive changes to previously approved experiments shall require the same review as a new experiment.

c. Minor changes to an experiment that do not significantly alter the experiment

, may be made by a Class A operator.

6.5 Reanired Actions 6.5.1 Actions to be Taken in Case of a Safety Limit Violation In the event of a 250 KW reactor power limit violation, the following actions shall be taken:

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a. . The reactor shall be secured and reactor operation shall not be resumed until a report of the violation is prepared and authorization is received from the Nuclear Regulatory Commission (NRC).

b. The 250 KW reactor power limit violation shall be promptly reported to the Director or a designated altemate.

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c. The NRC shall be notified by the Director or a designated alternate within one (1) working day of the violation bWelephone (see Section ..

6.6.2). ,

d. A 250 KW re. actor power limit violation report shall be submitted to the NRC within 14 days (see Section 6.6.2). The report shall describe the following:

(1) Applicable circumstances leadin to the violation including, when known, the cause and contributin factors.

(2) Effect of the violation upon reactor facility components, systems, or structures and on the hedith and safety ofpersonnel and the public. -

(3) Corrective actions taken to prevent recurrence.

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The report shall be revie.wed by the RSC and any follow-up report shall be

. . submitted to the NRC when authorization is sought to resume operation of the reactor.

6.5.2 Actions to be Taken in,the Event of a Reportable Oc,currence In the event of an occurrence which must be reported to the NRC according to Section 6.6.2, the following actions shall be'taken: ,

a. Reactor conditions shall be returned to normal or the reactor shutdown. Ifit is necessary to shut down the reactor to correct the .

occurrence, operations shall not be resumed unless authorized by the

. Director or designated-altemate.

b. The occurrence shall be reported to the Director or designated altemate immediately, and to the N1(C within one (1) working day by telephone (see Section 6.6.2). .

c. A written report describing the occurrence shhll be submitted to the NRC within 14 days (see Section ,6.6.2). ,

d. The occurence shall be reviewed by the Reactor Safety Committee at ' ,

the next regularly scheduled meeting. -

6.6 Reoorts

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All written reports shall be sent within the prescribellinterval to the Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the Jun31986 edition

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, R::ed Reactor Facility Technical Specifications Page 43

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Regional Administrator, Region V,1450 Maria Lane, Sdite 210, Walnut Creek, CA 1 94596-5368. ~. - -

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6.6.1 Operating Reports Routine annual reports covering the activities of the reactor facility d'uring the

, previous twelve months shall be submitted within three mohths following the end

.. of each prescribed year. These reports shall cover the.same period as the Reed

• College Admmistrative Cycle. Each annual operating r'eport shallinclude the following information:

a. A narrative summary of reactor operating expefiesce including the ~

energy produced by the reactor or the hoursthe reactor was critical, or both.

b. The unscheduled shutdowns including, where applicable, correctivs action taken to preclude recurrence. .

c. Tabulation of major preventive and corrective maintenance operations having safety significance. .-

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- d. Tabulation of major changes in the reactor facility,and procedures, and tabulation of new tests or experiments o'r both, that are significantly different from those performed previously and are nat described in the Safety Analysis Report, including conclusi6ns that no unreviewed safety questions were involved.

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e. A summary of the nature and amounf of radioactive effluents released or discharged to the environs beyond' the effective control of the owner-operator as determined at or before the point of such release or discharge. The summary shall include t.o the extent practicable an estimate ofindividual radionuclides present in the e'ffldent. If the estimated average release after dilution or diffusion is less than 25% 6f the concentration

>- allowed or recommended, a statement to this effect is sufficient.

", f. A summarized result of environmental surveys performed outside the facility.

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g. A summary of exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed or -

recommended.

6.6.2 SpecialReports

[ .' a. A report shall be submitted to Region V by telephone not later than the following working day and confirmed in writing by telegraph or similar conveyan&

to be followed by a written report within 14 days that describes the circumstances of any of the following events:

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(1) Violation of the reactor power safety limit (see Section 6.5.1).

1 (2) Release of radioactivity from the site above allowed limits.

(3) Other Reportable Occurrences (see Section 6.5.2).

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, ', (i) Operation with actual safety-s stem settings for required -

. . systems less conservative than the limiting safety-system settings. .

specified in the technical specifications.

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. (ii) .Operatioii in violation oflimiting con.ditions for operation -

established in the technical specifications unless prompt remedial

, action is taken. . .

. (iii) . A reactor safety system component malfunction which . -

renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or.. ..

• ' condition is discovered during maintenance tests or periods of . '

reactorshutdowns. (NOTE: Wheie components or systems are provided in addition tcrthose re. quired by the technical specifications, the failure of the extra components or systenis is not consi.dered reponable proyided that the minimum number of components or -

systems specified or required perform their intended reactor safety function.) ..

(

(iv) An unanticipated or uncontrolled ch'an'ge in reactivity greater ,

than 0.75% Ak/k ($1.00).-Reactor scrams resulting from a known

.' cause are excluded. -

(v) Abnormal and significant degradation in reactor fuel, or

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cladding, or both, coolant boundary, or' confinement boundary .

(excluding minoileaks) where applicable, which could result in exceeding prescribed radiation exposure limits of personnel or .

environment, or Both. .

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(vi) An observed inadequacy in the implementation of . c j . administritive or procedural controls such that the inadequacy ~

. . causes or could have caused the e' x istence or development of an ' .

unsafe conditionw. ith regard to reactor operations.

b., , A written report shall'be submitted to the NRC within 30 days of: *

(1) Permanent changis in the Faceity Organization at the levels of RRF

. Management orFacility Director. -

(2) Significant c.hanges in transient or accident analysis as described in

• the Safety Analysis Repon. , _

6.7 , Records - -

Facility records may be in the forni oflogs, qlata sheets, or other suita'ble forms. The .

tequired information may be contained in single or multiple records or a c~ombination ,

thfreof. + -

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6.7.1 Records to be Retained for the Lifetinie of RRF

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. (NOTE: Applicable annual reports,if they contain all,of the~ required information,

. may be used as records in this section.) ' -

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,, -Ried Reactor Facility Technical Specifications Page 45

. a.- G.aseous and liquid radioactive effluents released to the environs.

b.

Off-site envir'onmental-monitoring surveys required by the Technical Specifications.- -

c. Radiation exposum for all personnel monitored.

d. Drawings of the reactor facility. *

e. RRF radiation and contamination surveys where required by applicable regulations.

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f. Fuel inventories, mceipts, and shipments.

6.7.2 Records to be Retained for a Period of At Least Five Years or for the Life

' Of The Component Involved if less Than Five Years

- a. Normal RRF operations.

'b .

Principalmaintenance operations.

c. ' Reportable occurrences.

d. Surveillance activities required by technical specifications..

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e. - . Experiments performed with the reactor.

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f. ~ Approveg changes in operating procedures.

g. Records of meeting and audit reports of the Reactor Safety Committee.

6.7.3- Records to be Retained for at Least One Training Cyc'le Retraining and requalification'of certified operations personnel. Records of the e

inost recent complete cycle shall be maintained at all times the individual is a Certified Operator at RRF.

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k R:ed Reactor Facility Technical Specifications Page 46-

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7.0 EFFECTTVE DATE The effective date of these technical specifications shall be upon notification of approval by the NRC. These technical specifications, including'all applicable administrative and procedural changes required, shall be implemented within one year of notification of approval by the NRC. -

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