ML20054J726

Amended Tech Specs for License R-71

ML20054J726
Person / Time
Site:	05000142
Issue date:	04/30/1982
From:	CALIFORNIA, UNIV. OF, LOS ANGELES, CA
To:
Shared Package
ML20054J718	List: ML20054J717 ML20054J723 ML20054J726
References
NUDOCS 8206290494
Download: ML20054J726 (46)

Text

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O AeeEn0ix v TECHNICAL SPECIFICATIONS l FOR THE UNIVERSITY OF CALIFORNIA AT LOS ANGELES TRAINING REACTOR LICENSE N0. R-71 DOCKET NO 50-142 O

February 1980 Q 4-30-82 8206290494 820623 PDR ADOCK 05000142 P PDR

{} APPENDIX Y TECHNICAL SPECIFICATIONS CONTENTS Chapter 1 Definitions . . . . . . . . . . . . . . . . . . . .

1.1 Safety Channel . . . . . . . . . . . . . . . . V/1-1 1.2 Reactor Safety System . . . . . . . . . . . . V/1-1 1.3 Operable . . . . . . . . . . . . . . . . . . . V/1-1 1.4 Channel Check . . . . . . . . . . . . . . . . V/1-1 1.5 Channel Text . . . . . . . . . . . . . . . . . V/1-1 1.6 Channel Calibration . . . . . . . . . . . . . V/1-1 1.7 Unscheduled Shutdown . . . . . . . . . . . . . V/1-1 1.8 Reactor Shutdown . . . . . . . . . . . . . . . V/1-1 1.9 Reactor Operating . . . . . . . . . . . . . . V/1-2 1.10 Reactor Secured . . . . . . . . . . . . . . . V/1-2 1.11 Measuring Channel . . . . . . . . . . . . . . . V/1-2 1.12 Reportable Occurrence . . . . . . . . . . . . V/1-2 1.13 An Experiment . . . . . . . . . . . . . . . . V/1-2 1.14 Experiment Facilities . . . . . . . . . . . . V/1-3 1.15 Control Rod . . . . . . . . . . . . . . . . . V/l-3 1.16 Readily Available On Call . . . . . . . . . . V/1-3 1.17 Rod Drop Time . . . . . . . . . . . . . . . . V/1-3 1.18 Drop-Rod Scram . . . . . . . . . . . . . . . . V/1-3

(]) 1.19 Ful l Sc ram . . . . . . . . . . . . . . . . . . V/ l-4 1.20 Inhibit . . . . . . . . . . . . . . . . . . . V/l-4 1.21 Sa fety Limi ts. . . . . . . . . . . . . . . . . V/l-4 1.22 Closures . . . . . . . . . . . . . . . . . . . V/1-4 Chapter 2 Safety Limits and Limiting Safety System Settings .

2.1 Safety Limits of Reactor Operation . . . . . . V/2-1 2.1.1 Appl i ca bi l i ty . . . . . . . . . . . . . V/ 2-1 2.1.2 Obj ec t i ve . . . . . . . . . . . . . . . V/ 2- 1 2.1.3 Speci fications . . . . . . . . . . . . V/2-1 2.1.2 Bases. . . . . . . . . . . . . . . . . V/2-1 2.2 Limiting Safety System Settings. . . . . . . . V/2-1 2.2.1 Safety Channel Set Points. . . . . . . V/2-1 2.2.1.1 Applicability . . . . . . . . V/2-1 2.2.1.2 Objective . . . . . . . . . . V/2-2 2.2.1.3 Specifications. . . . . . . . V/2-2 2.2.1.4 Bases . . . . . . . . . . . . V/2-2 O

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h Chapter 3 Limiting Conditions for Operation . . . . . . . .

3.1 Reactivity Limitations . . . . . . . . . . . V/3-1 3.1.1 Shutdown Margin. . . . . . . . . . . V/3-1 3.1.2 Excess Reactivity. . . . . . . . . . V/3-1 3.1.3 Experiments . . . . . . . . . . . . V/3-1 3.1. 4 Control Rods . . . . . . . . . . . . V/3-1 3.2 Control and Safety Systems . . . . . . . . . V/3-1 3.2.1 Scram Time . . . . . . . . . . . . ..V/3-1 3.2.2 Measuring Channels . . . . . . . . . V/3-1 3.2.2.1 Bases . . . . . . . . . . . V/3-2 3.2.3 Safety Channels. . . . . . . . . . . V/3-2 j 3.2.3.1 Bases . . . . . . . . . . . V/3-3 3.3 Radiation Monitoring Systems . . . . . . . . V/3-4 3.3.1 Bases. . . . . . . . . . . . . . . . V/3-4 3.4 Engineered Safety Features . . . . . . . . . V/3-6 O 3.4.1 Sefety wish tevei Redietion Monitor. v/3-6 3.4.1.1 Specification . . . . . . . V/3-6 3.4.1.2 Bases . . . . . . . . . . . V/3-6 3.4.2 Confinement System . . . . . . . . . V/3-6 3.4.2.1 Specification . . . . . . . V/3-6 3.4.2.2 Bases . . . . . . . . . . . V/3-6 3.5 Limitations on Experiments . . . . . . . . . V/3-7

! 3.5.1 Experiments . . . . . . . . . . . . V/3-7 3.5.1.1 Applicability . . . . . . . V/3-7 3.5.1.2 Objective . . . . . . . . . V/3-7 3.5.1.3 Specification . . . . . . . V/3-7 3.5.1.4 Bases . . . . . . . . . . . V/3-7 3.6 Fuel . . . . . . . . . . . . . . . . . . . . V/3-9 3.6.1 Applicability. . . . . . . . . . . . V/3-9 3.6.2 Objective. . . . . . . . . . . . . . V/3-9 3.6.3 Specifications, Operational

! Limitation . . . . . . . . . . . . V/3-9 O 3.6.4 8 eses. . . . . . . . . . . . . . . . ves-9

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O 3.7 P ri ma ry Wa te r Qu al i ty. . . . . . . . . . . . V/ 3-10 3.7.1 Appl icabili ty. . . . . . . . . . . . V/3-10 3.7.2 Objective . . . . . . . . . . . . . V/3-10 l 3.7.3 Specifications . . . . . . . . . . . V/3-10 l 3.7.4 Bases. . . . . . . . . . . . . . . . V/3-10 3.8 Radioactive Releases . . . . . . . . . . . . V/3-11 3.8.1 Applicabili ty. . . . . . . . . . . . V/3-11 3.8.2 Objective . . . . . . . . . . . . . V/3-11 3.8.3 Specifications,0perational Limitation V/3-ll 3.8.4 Bases. . . . . . . . . . . . . . . . V/3-11 3.8.5 Liquid Effluent Releases . . . . . . V/3-12 3.8.5.1 Applicability . . . . . . . V/3-12 3.8.5.2 Objective . . . . . . . . . V/3-12 3.8.5.3 Specifications. . . . . . . V/3-12 3.9 Radiological Environmental Monitoring. . . . V/3-13 3.9.1 Applicability. . . . . . . . . . . . V/3-13 3.9.2 Objective . . . . . . . . . . . . . V/3-13 3.9.3 Specifications . . . . . . . . . . . V/3-13 3.9.4 Bases for Environmental Specificatior,s V/3-13 Q

Chapter 4 Surveillance Requirements . . . . . . . . . . . . V/4-1 4.1 Applicability. . . . . . . . . . . . . . . . V/4-1 4.2 Objective . . . . . . . . . . . . . . . . . V/4-1 4.3 Specifications . . . . . . . . . . . . . . . V/4-1 4.3.1 Frequency of Testing . . . . . . . . V/4-1 4.3.2 Safety Channel Calibration . . . . . V/4-1 4.3.3 Reactor Surveillance . . . . . . . . V/4-1 4.3.4 Control and Safety System Surveillance V/4-1 4.3.5 Radiation Monitoring System. . . . . V/4-2 4.3.6 Engineered Safety Features . . . . . V/4-2 4.3.6.1 Safety High Level Stack Moni tor . . . . . . . . . . V/4-2 4.3.6.2 Confinement System. . . . . V/4-2 4.3.7 Reactor Fuel . . . . . . . . . . . . V/4-2 4.3.8 Primary Coolant Water. . . . . . . . V/4-2 4.4 Bases. . . . . . . . . . . . . . . . . . . . V/4-2 O

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Chapter 5 Design Features . . . . . . . . . . . . . . . . . .

5.1 Appl ica bil i ty. . . . . . . . . . . . . . . . .V/ 5-1 5.2 Objective ..................V/5-1 5.3 Speci fications . . . . . . . . . . . . . . . .V/5-1 5.3.1 Rea c to r Fuel . . . . . . . . . . . . . V/ 5-1 5.3.2 Control and Safety Features. . . . . .V/5-2 5.3.2.1 Power Level (Normal Channels)V/5-2 5.3.2.2 Log Power Level Channel . . .V/5-2 5.3.2.3 Count Rate (Startup Channel).V/5-2 5.3.2.4 Neutron Source. . . . . . . .V/5-3 5.3.3 Rod Control System . . . . . . . . . .V/5-3 '

5.3.3.1 Shim (Control) Rods . . . . .V/5-3 5.3.3.2 Regulating Rod. . . . . . . .V/5-3 5.3.4 Cooling System 5.3.4.1 Primary Coolant System. . . .V/5-4 5.3.4.2 Secondary Cooling System. . .V/5-4 5.3.5 Confinement System . . . . . . . . . .V/5-5 5.3.5.1 Physical Features . . . . . .V/5-5 C' 5.3.5.2 Emergency Sequence. . . . . .V/5-5 5.3.5.3 Exhaust Duct Monitor (" Stack Monitor") . . . . . . . . . .V/5-5 5.3.6 Fuel Storage . . . . . . . . . . . . V/5-6 5.3.6.1 New Fuel . . . . . . . . . . .V/5-6 5.3.6.2 Irradiated Fuel . . . . . . .V/5-6 5.4 Fire Protection. . . . . . . . . . . . . . . .V/5-6 5.4.1 Applicabili ty. . . . . . . . . . . . .V/5-6 5.4.2 O bj ec t i ve . . . . . . . . . . . . . . . V/ 5-6 5.4.3 Specifications . . . . . . . . . . . .V/5-6 5.4.4 Bases. . . . . . . . . . . . . . . . .V/5-6 Chapter 6 Administrative Controls . . . . . . . . . . . . . .

6.1 Organization . . . . . . . . . . . . . . . . .V/6-1 6.1.1 Structure. . . . . . . . . . . . . . .V/6-1 6.1.2 Responsibili ty . . . . . . . . . . . .V/6-1 6.1.3 Staffing . . . . . . . . . . . . . . .V/6-1 O

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, f'-) 6.1.3.1 6.1.3.2 Minimum Staffing. . . . . . .V/6-1 Presence of Senior Reactor Ope ra to r. . . . . . . . . . . V/ 6-2 6.1.4 Selection and Training of Personnel. .V/6-4 6.1.5 Review and Audit . . . . . . . . . . '.V/6-4 6.1.5.1 Composition & Qualifications.V/6-4 6.1.5.2 Charter and Rules . . . . . .V/6-4

6.1.5.3 Review Function . . . . . . .V/6-4 4

6.1.5.4 Audit Function. . . . . . . .V/6-5 6.2 Procedu res . . . . . . . . . . . . . . . . . . V/6-6 6.3 Experiment Review and Approval . . . . . . . .V/6-7 j 6.4 Required Actions . . . . . . . . . . . . . . .V/6-8 6.4.1 Action to be Taken in Case of a Reportable Occurrence. . . . . . . . .V/6-8 6.5 Reports. . . . . . . . . . . . . . . . . . . .V/6-9 6.5.1 Operating Reports. . . . . . . . . . .V/6-9 6.5.2 Special Reports (Reportable 0ccurrence).V/6-9

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6.5.3 Wri tten Reports. . . . . . . . . . . .V/6-10

'() 6.6 Reco rds . . . . . . . . . . . . . . . . . . . . V/6-l l 6.6.1 Records to be Retained for a Period of at least Five Years. . . . . . . . . .V/6-ll 6.6.2 Records to be Retained for at least i One Requalification Cycle or for the Length of Employment of the Individual,

, Whichever Is Smaller . . . . . . . . .V/6-ll

! 6.6.3 Records to be Retained for the Lifetime of the Reactor Facility. . . . . . . .V/6-ll I

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APPENDIX V TECHNICAL SPECIFICATIONS sl ,

I LIST OF FIGURES i i

i Figure V/6-1 Organizational Relations . . . . . . . . . . . . .V/6-3 i

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,I LIST OF TABLES ~

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! Table V/3-1 Radiological Environmental Monitoring System . . .V/3-14 b

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1.0 DEFINITIONS

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The ter ns Safety Limit (SL), limiting Safety System Setting (LSSS), and

)- Limiting Condition of Operation (LCO) are as defined in 50.36 of 10 CFR Part 50.

11 SAFETY CHANNEL A Safety Channel is a measuring or protective channel in the reactor s .tfety system.

1.2 REACTOR SAFETY SYSTEM The Reactor Safety System is a combination of safety channels and associated circuitry which forms the automatic protective system for the reactor, or ,

provices infonnation which requires the initiation of manual protective action.

1.3 OPERABLE Cperable means a component or system is capable of performing its intended function in its required manner.

1.4 CHANNEL CHECK A Channel Check is a qualitative verification of acceptable performance by observation of channel benavior.

1.5 CHANNEL TEST O A Channel Test is the introduction of a calibration or test signal into the channel to verify that it responds in the specific manner.

l l 1.6 CHANNEL CALIBRATION 1

A Channel Calibration is an adjustment of the channel components such that its output responds, within specified range and accuracy, to known values of the parameter which the channel measures. Calibration shall enccmpass the entire channel, including readouts, alarm, or trip.

1.7 UNSCHEDULED SHUTOOWN An Unscheduled Shutdown is any unplanned shutdown of the reactor, after startup has been initiated.

1.8 REACTOR SHUT 00'4N The reactor is shut down when the reactor, at ambient temperature and xenon free condition and including the reactivity worth of all experiments, is subcritical by at least one dollar.

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1.9 REACTOR OPERATING O The reector is considered to de operatin, whenever it is not secured nor shut down.

1.10 REACTOR SECURED The reactor is secured when:

A. The core contains insufficient fuel to attain criticality under optimum conditions of moderation and reflection, or B. The moderator has been removed, or C. 1. All control rods fully inserted as required by Technical Specifications, and

2. The console key switch is in the off position and the key is removed from the lock, and

3. No work is in progress involving core fuel, core structure, installed control rods or control rod drives unless they are physically decoupled from the control rods, and

4. No in-core experiments are being moved or serviced with a reactivity worth exceeding the maximum value O aiio ed for a sin 9 e 1 experimeat-1.11 MEASURING CHANNEL A Measuring Channel 'is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring the value of a parameter.

1.12 REPORTABLE OCCURRENCE A Reportable Occurrence is any of those conditions described in Section 6.5.2 of this specification.

1.13 AN EXPERIMENT An Experiment is an apparatus, device or material, placed in the reactor core, in an experiment facility, or in If ne with a beam of radiation emanating from the reactor, excluding devices designed to measure reactor characteristics such as detectors and foils.

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A. Secured Experiment. Any experiment, experiment facility, or O comnonent or ea experiment is deemea to he secured. or ia a secured position, if it is held in a stationary position relative to the reactor core.

B. Movable Experiment. A movable experiment is one which may be inserted, removed, or manipulated while the reactor is c ri tical .

C. Untried Experiment. An Untried Experiment is a single experi-ment or class of experiments that has not been previously evaluated and approved by the Reactor Use Committee.

1.14 EXPERIMENT FACILITIES An Experiment Facility is any structure, device or pipe system which is intended to guide, orient, position, manipulate, control the environment or otherwise facilitate a multiplicity of experiments of similar character.

1.15 CONTROL R00 A Control Rod is a semaphore-type blade fabricated with cadmium as the neutron absorbing material which is used to compensate for fuel burnup, temperature, and poison effects. A control rod is magnetically coupled to its drive unit allcwirg it to perform the safety function when the magnet is de-energized.

1.16 READILY AVAILABLE ON CALL Readily available on Call means an individual who:

A. has been specifically designated and the designation known to the operator on duty, B. keeps the operator on duty infomed of where he may be rapidly contacted (e.g., by phone, etc.),

C. is capable of getting to the reactor facility within a reasonable time under nomal conditions (e.g.,1 hr.

or within a 30 mile radius).

1.17 R00 DROP TIME l Rod Drop Time is the elapsed time between the instant a limiting safety system set point is reached and the instant that the rod is fully inserted.

1.18 DROP-ROD SCRAM All four control rods fall by gravity into the core. Cooling water circulation continues.

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O i.i9 rutt Scaan The cooling water moderator is dumped in addition to the Drop-Pod Scram, and cooling water circulation ceases.

1.20 INHIBIT Inhibit prevents the withdrawal of control rods under a potentially unsafe condition. When under auto-control, inhibit is effected by decoupling the rod drive from the auto-controller. The decoupling initiates a drive reg-rod-down sequence.

1.21 SAFETY LIMITS Safety Limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against release of radioactivity. The principal physical barrier is the fuel cladding.

1.22 Ct05URES l Six micro-switches, wired in series, are recessed in the fixed concrete biological shield. If the removable concrete shield blocks are properly p' seated, the switches will all be closed. If the co'ncrete shield blocks v are not properly seated, and the power level is below 1 watt, the reactor will receive an " inhibit"; if the power is above I watt, the reactor will scram.

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS CF REACTOR OPERATION 2.1.1 APPLICABILITY This specification applies to the variables that affect thermal and hydraulic performance of the core. They are:

A. Power in KW, B. Flow in GPM, C. Maximum coolant temperature in *F.

2.1.2 OBJECTIVE To assure fuel cladding integrity 2.1.3 SPECIFICATIONS A. The maximum steady power level under various flow conditions shall not exceed 100 kw. The maximun power level that shall initiate a scram shall be 125 kw.

B. The coolant flow rate shall be at least 10 gpm at all power O/ levels greater than one watt.

C. The core water outlet temperature shall not exceed 200*F.

D. The specific r,esistivity of the primary water shall not be less than 0.5 megohm centimeters.

2.1.4 BASES Operating experience shows that specifications (A), (B), and (D) suffice to maintain cladding temperatures below 212*F. Specification (B) is known to be conservatively safe from tests conducted under a temporary amendment that permitted brief periods of operation at 500 kw.

2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 SAFETY CHANNEL SET POINTS 2.2.1.1 Applicability This specification applies to the set-points of the safety channels.

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2.2.1.2 Objective O To insure that automatic action is initiated that will prevent a safety limit from being exceeded.

2.2.1.3 Specification The limiting safety system settings are:

A. Power level at any flow rate shall not exceed 125 kw.

B. The primary coolant flow rate shall be greater than 10 gpm.

C. The average primary coolant outlet temperature shall not exceed 180*F.

D. The primary coolant shall be demineralized light water with a specific resistivity not less than 1 megohm centimeters.

2.2.1.4 Bases These limits arise fron operating experience. The trip-points A and B initiate automatic scram. Trip-point C is signalled by a horn and light, trip-point 0 hy a light. Points C and D can be approached and exceeded only slowly and timing will permit operator intervention. At power levels below one watt, the primary coolant flow trip may be bypassed and the primary circulation cut off. This would nonnally occur during fuel transfer operations, or when trying to isolate a breach of cladding in a fuel bundle.

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3.0 LIMITING CONDITIONS FOR OPERATION O 3.i aE4Cr m tv LIsitarIONS 3.1.1 SHUTDOWN MARGIN The minimum shutdown margin provided by control rods in the cold, xenon-free condition with the highest worth control rod fully withdrawn shall not be less than *,1.00 This specification ensures that the reactor can be shut down from any operating condition and remain shut down af ter cool-down even if the highest-worth control rod is stuck in its fully withdrawn condition.

3.1.2 EXCESS REACTIVITY The total clean, xenon free core excess reactivity with or without experiments shall not exceed $3.00.

3.1.3 EXPERIMENTS Reactivity limits on experiments are specified in 3.5 below.

3.1.4 CONTROL RODS 1

The reactivity insertion rate for a single control blade shall not exceed $0.077/sec.

3.2 CONTROL AND SAFETY SYSTEMS 3.2.1 SCRAM TIME The scram time shall not exceed I second (rod drop time).

3.2.2 MEASURING CHANNELS The minimum number and type of measuring channels operable and providing information to the control room operator required for reactor operation are given as follows:

Channel No. Operable l Safety Amplifier 2 Linear w/ auto controller 1 Log N and Period Channel 1 a

Start up 1 0

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Channel No. Operable Rod Position 4 Coolant Flow (Primary and Secondary) 2 Coolant Temperature 2 (primary)

Core Level 1 NOTE: a. Operable below 0.02W 3.2.2.1 Bases The normal power level instruments (" Level Safeties") provide redundant infomation on reactor power in the range 5% - 150% of the normal operating power level of 100 kw. The linear power channel presents the reactor power in a linear manner by decades. The linear recorder also serves as part of the feedback network for the auto controller servo.

The log power instrument (" Log N") provides usable reactor power information in the logarithmic range of 0.01 U to 1 Mwt. By differentiating the log of the power, the reactor period information is also displayed.

qU The count rt.te channel covers the neutron flux range from the source level

(-1 cps) to 105 cps on a digital scale. It enables the operator to start the reactor safely from a shutdown condition and to bring the power to a level that can be measured by the Log N instrument.

Coolant flow rate and temperature instruments allow the operator to calculate reactor power and calibrate the neutron flux channels in terms of power.

The primary water outlet temperature is monitored.

Rod position indicators show the operator the relative position of control rods, and enable rod reactivity calibrations to be made.

3.2.3 SAFETY CHANNELS The minimum number and type of channels providing automatic action that are required for reactor operation are as follows:

l Channel No. Operable Function Power Level (Safety) 2 Full Scram 0 > 125 kw Power level (Log N 8 Period) 1 Full Scram (a < 3 sec. period Inhibit 0 < 6 sec. period G

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Channel No. Operable Function a

O Count Rate 1 Inhibit 9 < 2 cps Core Water level 1 Scram 9 < 45 in.

b Primary Coolant Flow 1 Scram 9 < 10 gpm Manual Button 1 Full Scram Keyswitch [or loss of power to console 31 Full Scram b

Closures 6 Full Scram > 1 watt Notes: a. Operable below 0.02 W and bypassed above,

b. May be bypassed at power levels below 1 watt.

3.2.3.1 Bases The power level scram provides redundant automatic protective action to prevent exceeding 125% of the license limit on reactor power.

The period scram, assisted by the intermediate rod inhibit, limits the rate of increase in reactor power to values that are controllable without excessive power levels or temperature. These functions are not limiting safety system response.

O The inhibit on the count rate channel prevents inadvertent criticality

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during cold startup that could arise from lack of source neutrons and the neutron instrument response.

! Reactor core low water level scrams the reactor. This scram also prevents startup until the minimum core water level is reached.

If the primary coolant outlet temperature exceeds 180*F, a high tenperature alarm annunciates at the control room annunciator panel.

The coolant flow scram ensures adequate coolant flow to prevent boiling in the core.

The keyswitch scram prevents unauthorized operation of the reactor.

Bypass is permitted on non-power parameters for experiments, tests , and special purposes only [ refueling].

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3.3 RADIATION MONITORING SYSTEMS The minimum acceptable radiation monitoring and annunciating instrumentation required for reactor operation is as follows:

Max. Alarm Type No. Operable Setpoint Function Safety high level 1 25 mR/hr Detect high radiation radiation and/or high stack gas concentration Alann and isolate at

> 25 mR or 1.8x10-5 uCi/mi Exhaust Duct Monitor 1 1.8x10 Detect Ar-41 and other

(" Stack Monitor") uCi/ml* radioactive gases via ion chamber

, Alarms with displays in the control room i

• In the event that the limits for Argon 41 contained in 10 CFR Part 20, Appendix B, Table II, with a reduction factor of 460 are exceeded in n

v the stack, the ventilation fans shall be shut down and the automatic damper system closed to limit natural circulation from i.he reactor room to the external environment and the reactor is automatically scrammed.

Fixed Area Monitors 2 5 mR/hr Detect radiation (gamma) in key locations; alarm in control room Evacuation Switch 1 ---

Alarm to initiate evacuation sequence.

(Manual)

NOTE: For maintenance or repair, required radiation monitors may be replaced by portable or substitute instruments for periods up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided the function will still be accomplished. Interruption for brief periods to pennit checking or calibration is pemissible.

3.3.1 BASES The radiation monitoring system components are located and have set points to ensure that 10 CFR Part 20 requirements are not exceeded for restricted and unrestricted areas.

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! A. The Safety High level Radiation alarm scrans the reactor by iO i

.i shetti=2 do # the ve"ti' tio" s'stc=-  :

B. The Exhaust Duct Monitor annunciates at the control room panel indicating to the operator to shut down the venti-lation fan. This action closes the automatic damper and

automatically scrams the reactor.

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3.4 ENGINEERED SAFETY FEATURES

'O These specifications apply to required equipment for the confinement of activity through controlled release of reactor building air to the atmosphere.

3.4.1 SACETY HIGH LEVEL RADIATION MONITOR 3.4.1.1 Specification See 3.3 3.4.1.2 Bases This monitor senses excessive radiation in the reactor room and/or high concentrations of radioactive gases in the exhaust stack and automatically initiates the reactor room ventilation shutdown sequence, which consists of the shutdown of the reactor room supply and exhaust fans, closure of dampers in the building ventilation system and causing a drop rod scram if the reactor is in operation.

3.4.2 CONFINEMENT SYSTEM 3.4.2.1 Specification A. The exhaust fan shall have a capacity of 14,000 CFM and shall maintain a negative pressure in the reactor building and an Q exhaust rate from the reactor room of approximately 8000 CFM.

B. The high bay ventilation exhaust and intake fans are interlocked to shut off simultaneously when the ventilation system is shut down.

C. Spring loaded, air operated damper motors automatically close the intake and exhaust dampers.

D. All doors to the reactor high bay shall be nomally closed l while the reactor is operating. Transit is not prohibited I

under proper supervision.

l l E. The safety rods shall automatically scram when the ventilation fan is shut down.

3.4.2.2 Bases i

t To effect controlled release under normal conditions of gaseous activity l present in the building atmosphere, a negative pressure is required so that the air flow to the reactor room is non-radioactive air from " cold" areas in the building. This serves to dilute reactor room ambient air prior to discharge, and to prevent the flow of reactor room air out of the reactor room to other parts of the building. Under emergency conditions, the reactor room dampers will close, and the reactor will be scrammed.

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l 3.5 LIMITATIONS ON EXPERIMENTS l f')

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! 3.5.1 EXPERIMENTS 3.5.1.1 Applicability This specification applies to those experiments installed in the reactor and its experimental facilities.

3.5.1.2 Objective The objective is to prevent damage to the recctor, excessive release of radioactive material in the event of experiment failure, and to prevent the safety limits from being exceeded.

3.5.1.3 Specifications c

Experiments installed in the reactor shall meet the following conditions:

A. The reactivity worth of any single experiment shall be .

limited to - $0.90to +$0.28.

B. In no event shall the sum of the available excess reactivity of the cold critical core plus any experiment exceed $3.00.

C. An experiment shall not be inserted or removed unless all O the control blades are fully inserted or its absolute V reactivity worth is less than that which would cause a 20-second positive stable period (50.28).

D. Experimental apparatus, material, or equipment to be inserted in the reactor shall be reviewed to insure non-interference with the safe operation of the reactor. '

E. Each type of experiment utilizing the reactor cust have been previously reviewed and approved by the Reactor Use Comittee and/or the Supervisor and Health Physicist for routine experiments.

F. No explosive materials shall be irradiated.

3.5.1.4 Bases A. Specification A prevents pro @ t criticality.

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B. Spert and Borax tests indicated that inadvertent reactivity'-

insertions of $3.90will not cause the MTR-type fuel or cladding to melt or rupture.

C. The insertion of all rods assures that the reactor will not go critical upon sample insertion or removal .

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'D.

.O Ensures that no physical or nuclear interference with the safe operation of the reactor will occur.

E. Ensures that all experiments are evaluated by an independer.t group knowledgeable in the appropriate fields.

F. These Technical Specifications prohibit the irradiation of explosives.

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3.6 FUEL 3.6.1 APPLICABILITY These specifications apply to the number and condition of the fuel elements present in the core.

3.6.2 OBJECTIVE To geometrically constrain excess reactivity, and to avoid excessive release of fission products.

3.6.3 SPECIFICATIONS, OPERATIONAL LIMITATION 3.6.3.1 The maximum fuel loading shall consist of 24. bundles, each containing 11 plates. The plates consist of enriched uranium-aluminum sandwiched between high purity aluminum clad.

3.6.3.2 Fuel element loading and distribution in the core shall comply with the fuel handling procedures.

3.6.3.3 Fuel elements exhibiting release of fission products due to cladding rupture shall, upon positive identification, be removed from the core. Fission product contamination of the primary water shall be treated as evidence of fuel element failure, and positive identification shall be determined by gamma-ray spectroscopy.

3.6.3.4 A non-operating period of not less than 21 days shall precede l core entry for the purpose of fuel handling.

3.6.4 BASES 3.6.4.1 The original and present configurations.

3.6.4.2 There are written procedures for fuel handling and loading.

3.6.4.3 Migration of fission products within the fuel meat will be strongly inhibited by the large volume fraction of aluminum (nearly 98% aluminum) in the alloy. Thus a relatively small defect cannot release large quantities of fission products originating at locations remote from the defect. Small releases will be largely captured by the primary water system. (See 3.7 Primary Water Quality.) A major failure would be readily evident via one or more of the inhouse radiation monitors.

3.6.4.4 ANSI N398-1974.

3.6.4.5 Experience has shown that a 21 day holding period suffices to limit personnel fuel handling exposures to acceptable levels.

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3.7.1 APPLICABILITY This specification applies to primary cooling system water in contact with fuel elements.

3.7.2 OBJECTIVE i To minimize corrosion of the aluminum cladding of fuel plates and activation of dissolved materials.

3.7.3 SPECIFICATIONS 3.7.3.1 Primary water temperature will not exceed 200*F.

3.7.3.2 Primary water specific resistance is to be not less than 500,000 ohm-cm. Primary water conductivity is checked prior to each reactor start-up.

3.7.33 Primary water shall be sampled, concentrated, and the gross radioactivity of the residue measured with a thin window GM counter.

Radioisotope identification shall be by gamma-ray spectroscopy. This specification and procedure shall prevail:

7 A. upon the appearance of any unusual radioactivity in the (V primary water or the primary water demineralizers and B. prior to the release of any primary water from the site.

C. for frequency'of analyses refer to 4.3.8 3.7.4 BASES 3.7.4.1 Specification A is designed to protect the fuel element integrity and is based upon operating experience.

3.7.4.2 At this water quality, the activation products (of trace minerals) do not exceed acceptable limits and corrosion rates of fuel elements is acceptably low.

3.7.4.3 This specification is designed to:

A. detect and identify fission products resulting from fuel failure, B. to fulfill reportability requirements pertaining to liquid wastes, and C. detect cladding failure due to corrosion.

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l 3.8 RADI0 ACTIVE RELEASES j O 38i ^aetic^ai'irr Applies to all radioactive releases to the atmosphere via the discharge stack.

3.8.2 OBJECTIVE To assure that the discharge of Ar-41 is within the limits established by 10 CFR Part 20, Appendix B.

3.8.3 SPECIFICATIONS, OPERATIONAL LIMITATION A. Unrestricted Areas - The concentration of Argon-41 released to the atmosphere, multiplied by a reduction factor of 1/460 shall not exceed the limits of 10 CFR Part 20 Appendix B.

Table II. The reduction factor is defined as the product of the reactor use factor, occupancy factor, and meteorological dilution factor.

B. Restricted Areas - The concentration of Argon-41 released to the atmosphere, multiplied by a reduction factor equivalent to the reactor use factor shall not exceed the limits of 10 CFR Part 20, Appendix B, Table I. All " restricted areas shall be posted accordingly. A list of restricted areas shall be maintained with one copy sent to I&E. Included as " restricted O are snaii de the roor of the buiioias coata$aias tne reactor exhaust system; other roof top areas shall be designated unrestricted areas.

C. The Commission shall be notified if, over any one year period, the reactor use factor,, the occupancy factor, or the dilution factor change so as to increase the effective reduction factor.

The reactor use factor shall be based upon an annual average use equivalent to 8.5 hrs. per week with not more than 20 equivalent full power hours of reactor operation in any one week peri D. In the event that the limits for Argon-41 contained in 10 CFR Part 20, Appendix B, Table II, with a reduction factor of 460 are exceeded in the stack, the ventilation system shall be secured and shall cause the automatic system to isolate the reactor room, and the reactor shall be shut down.

3.8.4 BASES A conservative emissions reduction factor of 460 was used to establish the exhaust monitor alarm setting of 1.8x10-5 u Ci/cc the reduction factor is composed of a average reactor utilization factor of 0.18 [ based on 8.5 hrs.

per 45 hour5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> week], an atmospheric dilution factor of 0.115, and an occupancy factor of 0.10.

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. The conservative reduction factor of 460 coupled with the stack monitor set point of 1.8 x 10-Sci /cc to shut down the reactor ensures that the dose at the uncontrolled area boundary from gaseous effluents from reactor operation will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas, which is (,500 mrem /yr. to the total body.

3.8.5 LIQUID EFFLUENT RELEASES 3.8.5.1 Applicability This applies to all radioactive liquid releases to the sewer systems.

3.8.5.2 Objective To assure compliance with 10 CFR Part 20, Appendix B.

3.8.5.3 Specifications A. Before releases from the hold tank or sump, the liquid waste shall be sampled and the activity level measured.

B. Liquid wastes discharged to sewers shall be diluted to concentrations less than or equal to those specified in 10 CFR Part 20, Appendix B.

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3.9 RADIOLOGICAL ENVIRONMENTAL MONITORING 3.9.1 APPLICABILITY This specification refers to the number of samples, frequency of sampling, type of sampling, type of analysis and frequency of analyses.

3.9.2 . OBJECTIVE To conduct a radiological environmental monitoring program that will ensure l cort.pliance with 10 CFR 20, 3.9.3 SPECIFICATIONS The radiological environmental monitoring grJagram shall be conducted as specified in Table 15-3.9-1. The results of analyses performed on the radiological environmental monitoring samples shall be summarized in the Annual Report to the Nuclear Regulatory Commission.

3.9.4 BASES FOR ENVIRONMENTAL SPECIFICATIONS 3.9.4.1 Specification 3.8.1 is provided to ensure that the dose at the uncontrolled area boundary from gaseous effluents from the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas, which is < 500 mrem /yr. to the total body.

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O O O Table V/3-1 RADIOLOGICAL Ef1VIR0flMEfiTAL f10filTORIfiG PROGRAM Exposure Pathway Number of Samples Sampling and Type and Frequency and/or Sample and Sample Locations Collection Frequency of Analysis

1. Airborne

a. Gaseous 1 sample from exhaust Continuous during Flow-through ion chamber stack operations and core al terations

b. Particulates 1 sample from exhaust ' Continuous operation of Particulate sampler.

stack i sampler with sample Analyze for gross beta roml collection as required radioactivity following

) by dust loading but at filter change. Perform re c or b 1 I intake air least once per 7 days gamma isotopic analysis on y each sample when gross beta g activity is > 10 times the mean of control samples for any medium.

2. Direct Radiation Annual survey of all Annually Gamma and neutron, with internal areas and portable instruments extending beyond the controlled perimeter of the facility

3. Other Area Wipes Weekly Gross alpha-beta-gamma 14 within the

, controlled area L, 4 external to the

? controlled area 8

4.0 SURVEILLANCE REQUIREMENTS 4.1 APPLICABILITY This specification refers to the frequency and type of sampling, testing and inspections that will be conducted on surveillance systems.

4.2 OBJECTIVE To prescribe periodic samples, tests and inspections that verify that the performance of required systems is in accordance with specifications given in Table 15-3.9-1 and Sections 2 and 3.

4.3 SPECIFICATIONS 4.3.1 FREQUENCY OF TESTING In all instances where the specified frequency is annual, the interval between tests is not to exceed 15 months; and when semiannual, the interval should not exceed 71/2 months. However, the interval shall average out to be twelve months + 1 month or 6 months, over any five year period 4.3.2 SAFETY CHANNEL CAllBRATION A channel calibration of each safety channel shall be perfomed annually (see Section 3.2.3).

4.3.3 REACTOR SURVEILLANCE A. The reactivity worth of each control rod (including the regulating rod) and the shut-down margin shall be detemined whenever var'ation requires a reevaluation of core physics parameters, or annually, whichever occurs first. The rod worth will be determined by observing the rod position changes necessary to compensate for samples of known worth.

B. The reactivity worth of an experiment shall be measured at i low power, before conducting the experiment. A reactivity measurement on a rabbit experiment may be perfomed at high power if the experiment has been performed previously and there is good reason to believe the reactivity worth is less t than 30.28.

4.3.4 CONTROL AND SAFETY SYSTEM SURVEILLANCE A. The rod drop times shall be measured annually. The reactivity worth of each control blade, reactivity insertion rate of I each control, blade, and shutdown margin shall be detemined l

annually.

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B. A channel test of each measuring channel in the reactor safety system shall be performed following related main-O teaence to thet caennei-C. A channel check of each measuring channel in the reactor safety system shall be performed prior to each startup.

It is not perfomed for startup from temporary shutdown.

4.3.5 RADIATION MONITORING SYSTEM A. The safety high level monitor, stack monitor, and area monitors shall be calibrated semi-annually.

B. The area monitors shall receive a channel test prior to each startup.

4.3.6 ENGINEERED SAFETY FEATURES 4.3.6.1 Safety High Level Stack Monitor The safety high level stack monitor will shut down the reactor room ventilation system and dampers to isolate the room from the surrounding environment if the trip point is exceeded.

4.3.6.2 Confinement System The operability of the evacuation alam and confinement isolation system O (fan and damper shutdown) shall be tested and verified semiannually.

4.3.7 REACTOR FUEL A. Upon receipt from the fuel vendor, all fuel elements shall be visually inspected and the accompanying quality control docu-ments checked for compliance with specifications.

B. Each new fuel element will be inspected for damage and flow obstructions prior to insertion into the core.

4.3.8 PRIMARY COOLANT WATER A. The specific aesistance of the primary coolant water shall be measured priar to startup.

B. An analysis of the primary water for radioactive material shall be done annually, and prior to the release of any primary water from the site. An analysis shall also be made upon the appearance of any unusual activity in the water or the demineralizer.

4.4 BASES g The frequency of inspection calibration and testing have been verified by twenty years of operation.

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5.0 DESIGN FEATURES 5.1 APPLICABItITY This specification applies to the design of equipment relevant to operational safety.

5.2 OBJECTIVE To ensure that those design features which are relevant to operational safety and to limits that have been previously specified and described below shall not be changed without appropriate review.

5.3 SPECIFICATIONS 5.3.1 REACTOR FUEL Fuel elements shall be of the general MTR type with thin plates containing uranium fuel enriched to about 93% U-235 and clad with aluminum. The fuel matrix may be fabricated by alloying high purity aluminum-uranium or by the powder metallurgy method where the starting ingredients (uranium-aluminum) are in fine powder formi.

Elements shall conform to these nominal specifications:

A. Overall size (bundle) 2.845 in. x 2.14 in. x 25.625 in.

B. Clad thickness: 0.015 in.

C. Plate thickness: 0.070 in.

D. Water channel width: 0.137 in.

E. No. of plates: standard element - 11 fueled plates F. Plate attachment: bolted with spacers G. Fuel content per plate: 14g U-235 nominal H. Fuel bundles: 24 bundles, each containing 11 plates v

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5.3.2 CONTROL AND SAFETY SYSTEMS O~

Design features of the components of this system (3.2.2, 3.2.3) that are important to safety are given below.

5.3.2.1 Power Level (Nonnal Channels)

For this function two independent measuring channels are provided, both are required for the reactor to be operable. Each channel covers reliably the range from about 5*. to 1507, (of 100 kw). Each channel is comprised of an uncompensated boron-coated ion chamber feeding an amplifier that controls electronic switches in the DC current that flows through each control rod electromagnetic clutch. Each channel controls and drops all control rods.

Each channel is fail-safe. Scram from each channel is accomplished through interruption of poer to the electromagnetic clutches that couple the rod drives to the rods. Each channel indicates power level on a panel meter.

Each chamber can be changed in position, over a limited range, so as to allow the channel reading to be standardized against reactor thermal power.

5.3.2.2 Log Power Level Channel For this function a single channel is provided, covering reliably the range 10 to 1.0 Mw with a logarithmic output indication on both a panel meter and a chart recorder. To cover the range under all : ore conditions a gamma-compensated boron-ion chamber is used to supply a logarithmic amplifier. The chamber can be changed in position, over a limited range, so as to allow the O channel reading to be standardized against reactor thennal power. Rate of change of power information is also derived, in the form of a period, that can initiate full scram. This channel also provides control and inhibit actions, viz. bypassing of the start-up channel functions, bypassing of

" closures" scram below 1.0W, and inhibit of control rod.

5.3.2.3 Count Rate (Start-up Channel)

A BF proportional chamber is used to supply pulses to a scaler and digital count rate circuitry. Pulse height discrimination selects pulse amplitudes that correspond to neutron events and rejects those from gamma events. Count rate on a digital counter is displayed on a panel meter. The channel covers a range of 1 - 105 cps, corresponding roughly from scurce to .02 W. To prevent control-rod withdrawal when the neutron count rate information may not be reliably indicated, an inhibit is provided when the count rate is less than 2 per second. This inhibit is bypassed at a power of >.02 W. The scaler can also be used for obtaining accurate values at low c.ount rates if needed (e.g., approach to critical with new fuel or for instructional purposes).

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O 5.3.2.4 Neutron Source for obtaining reliable neutron information necessary to startup from a cold shut-down condition, a radium-beryllium neutron source is permanently installed.

5.3.3 R00 CONTROL SYSTEM 5.3.3.1 Shim (Control) Rods Three control rods are provided for the control of core reactivity. These rods are cadmium-tipped magnesium (see 4.3.4). Individual integral worths l vary from about $2.40 - $2.70, depending on position and individual character-istics. The rods are coupled to drive shafts through electromagnetic clutches that allow release of the rods within 12 ms after receiving a scram signal.

Position indicators on the control console show the extent of withdrawal for each rod. To limit the rate of reactivity increase upon startup, the rod drive speeds are limited to 7.7x4/sec. and only one rod can be withdrawn at a time. These rods are not otherwise automatically controlled, but are used to compensate for seasonal and long-term reactivity changes.

5.3.3.2 Regulating Rod One regulating rod is provided to aid in fine control and maintenance of constant reactor power for long periods. The rod is limited to a total O worth of about $1.80 and can be either manually or servo-controlled. The drive speed is approximately 1% (of full range) per second. In the commonly used range, this amounts to approximately $0.03/second.

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5.3.4 COOLING SYSTEM O 5.3.4.1 erimerv Cooient Svstem In normal operation, the primary water is pumped from the dump tank to the bottom of the fuel boxes, upward past the fuel plates, to overflow pipe weirs above the fuel plates, and returned to the dump tank. The dump tank is fitted with tube bank through which secondary water is circulated to effect heat removal. There is a bypass from the fuel box fill line to the dump tank, that is controlled by the dump valve. Under the full scram condition, the dump valve is opened and drains the core water, by gravity, to the dump tank. The removal of the core water moderator introduces approximately $25.00 of negative reactivity into the core. The dump valve is spring loaded open, closed by air pressure, and will open upon f ailure of the air supply or electrical f ailure.

Primary water quality is maintained by the circulation of a slip-stream (30 gph) from the pump discharge through cartridge filters and demineralizers to the dump tank. The primary water flow rate (10 gpm minimum, normally 16 gpm) is automatically controlled and indicated at the control console.

Core inlet and outlet water temperatures are monitored. The combination of temperature rise and ficw rate yield the heat removal rate that defines the reactor thermal power.

5.3.4.2 Secondary Cooling System 7u Q Reactor power transferred through the heat exchanger is dissipated to the storm drain by once-through cooling water. To minimize corrosion, the exchanger has stainless steel tubes. To prevent water from entering the secondary system should a tube leak occur, the static pressure in the secondary is made higher than that of the primary through the relative elevations of the two systems and the back pressure of the city water system.

Potential leakage from the secondary cooling system into the primary system is noted by analyzing for rate of conductivity change of the primary coolant water.

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] 5.3.5 CONFINEMENT SYSTEM Physical Features 5.3.5.1 1

The confinement structure consists of the reacter building, with a free air volume of about 3000 m . This building houses the reactor, the primary cooling system including the dump tank / heat exchanger and the hold-up tanks. Personnel access is via specially keyed doors. Ventilation access to the building is through pneumatically operated damper valves that can be used to prevent natural through-flow when the fans are turned off. These dampers are fail safe and close upon loss of air pressure.

5.3.5.2 Emergency Sequence The emergency sequence is initiated either automatically by the safety high level radiation stack monitor or manually by the console operator. The sequence is that the reactor room ventilation supply and exhaust fans are shut down, the dampers are closed, and the reactor is shut down.

5.3.5.3 Exhaust Duct Monitor (" Stack Monitor")

Air in the exhaust duct is continuously sampled for particulate activity.

During reactor operations or core alterations, exhaust air drawn from the reactor room is continuously monitored for gross concentrations of radioactive gases.

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O s36 rut' stoa ^ct 5.3.6.1 New Fuel Unirradiated new fuel elements are stored in a vault-type room security area equipped with intrusion alarms in accordance with the Security Plan. Elements are stored in a steel, fireproof safe in which a cadmium plate separates each layer of bundles. With such an arrangement, subcriticality is assured.

5.3.6.2 Irradiated Fuel Irradiated fuel is stored upright in a dry storage pit within the reactor building in criticality-safe holes. Holes are located with center-to-center '

square spacing of 13 inches, each can accommodate 1 to 3 elements.

5.4 FIRE PROTECTION 5.4.1 APPLICABILITY This specification applies to providing for the detection of a fire or smoke in the reactor room.

5.4.2 OBJECTIVE To detect the presence of a fire in the reactor room at the earliest time O possible.

5.4.3 SPECIFICATIONS A smoke detection device shall be installed in the reactor room. The device shall have the capability of sounding an alann that is audible to routine security forces or is connected directly to the UCLA security forces facility or appropriate fire departments.

5.4.4 BASES Early detection of a fire is necessary for rapid response and confinement of any fire.

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O 6 o ^ont"ts'a^'tvt co" Tao's 5.1 ORGANIZATION 6.1.1 STRUCTURE The organization structure for the management and operation of the reactor facility is shown in Figure 6.1.1-1. Job titles are shown for illustration and may vary. Four levels of authority are provided, as follows:

Level 1: Individual responsible for the facility license and site administration.

Level 2: Individual responsible for the reactor facility operation and management.

Level 3: Individual responsible for daily reactor operations.

Level 4: Reactor operating staff.

The Reactor Use Committee is appointed by, and shall report to, the Dean of the School of Engineering and Applied Science. Radiation safety personnel shall report to level 2 or higher through a separate and distinct organizational channel.

Q 6.1.2 RESPONSIBILITY Responsibility for the safe operation of the reactor facility shall be within the chain of command shown in Figure 6.1.1-1. Management levels in addition to having responsibility for the policies and operation of the reactor facility shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating ifcense and technical specifications. 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 6.1.3.1 The minimum staffing when the reactor is not secured shall be:

A. A licensed Reactor Operator in the control room, and a health physics instructed individual in the facility.

a B. A licensed Senior Reactor Operator shall be readily available on call.

C. A Health Physicist shall be readily available on call.

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l 6.1.3.2 Events requiring the presence of a Senior Operator:

All fuel-element or control-rod alterations within the i A.

'! reactor core region.

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B. Recovery from unplanned or unscheduled shetdowns.  !

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E LEVEL 2 DEN 4, SCK)OL OF AS$1STANT 4 ENGINEERING #O VICE CHANCELLOR APPLIED SCIENCE COtt O41TY SAFETY D REC EAR P REACTOR USE DIRECTOR, M) CLEAR C0441 TTEE ENERGY LABORATORY m p gp , , , ,

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OPERATIOr6 gy e 37 LEVELS 3&4 FLP4CTIONAL RESPONSIBILITIES EXPERIMENTAL USE APPROVAL STAFFING OVERALL SAFETY Bt.DGET PRC COrtV41CATIOf6 TEOt41 CAL O W4GES UNIVERSITY PUBLIC ATIN G PREPARE LICENSE ATtOMENTS ORGANIZATIONAL RELATIONS FIGURE V/6-1 O

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] 6.3 EXPERIMEpi REVIEW AND APPROVAL All new experiments or classes of experiments that could affect 6.3.1 reactivity or result in release of radioactive materials shall be reviewed bf the Reactor Use Commitee. This review shall assure that compliance with the requirements of the license, technical specifications, and applicable regulations has been satisfied, and shall be documented.

6.3.2 The reactor supervisor and the resident health physicist shall review and approve in writing all proposed experiments prior to their performance.

6.3.3. The following conditions shall govern the performance of experiments:

A. the reactivity worth of any single experiment shall be limited to -50.90 to +50.28. (Refer to 3.5.1.3.A)

B. An experiment shall not be inserted or removed unless all the control blades are fully inserted or its absolute reactivity worth is less than that which would cause a 20-second positive stable period (50.28). (Refer to 3.5.1.3.C) i C. No explosive materials shall be irradiated. (Refer to 3.5.1.3.F) p D. In no event shall the sum of the available excess reactivity v of the cold critical core plus any experiment exceed $3.00 (Refer to 3.5.1.3.B) l l

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6.4 R_EOUIRED ACTIONS "O

\/ 6.4.1 ACTION TO BE TAKEN IN CASE OF A REPORTABLE OCCURRENCE Reportable Occurrences are described in 6.5.2 A. Corrective action shall be taken to return conditions to normal;

! otherwise, the reactor shall be shut down and reactor operation shall not be resumed unless authorized by the Level 2 authority or designated alternates.

i B. All such occurrences shall be promptly reported to the Level 2 authority or designated alternates.

l C. All such occurrences where applicable shall be reported to the Commission in accordance with section 6.5.3.

D. All such occurrences including action taken to prevent or reduce the probability of a recurrence shall be reviewed by the Reactor Use Committee.

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6.5 REPORTS In addition to the requirements of applicable regulations, reports shall be made to the Commission as follows:

6.5.1 OPERATING REPORTS A routine report (in writing to the Director, Division of Licensing, USNRC, Washington, D.C. 20555) at the end of each 12-month period providing the following information:

A. A narrative summary of reactor operating experience, including the energy generated by the reactor (in megawatt-hours).

B. A discussion of unscheduled shutdowns, including the corrective action taken to preclude recurrence.

C. A summary of the preventive and corrective maintenance operations performed having safety significance.

D. A discussion of the changes in the facility and procedures, and the tests and experiments, carried out, without prior approval by the U.S. Nuclear Regulatory Commission pursuant to 10 CFR Part 50, Section 50.59.

E. A summary of the nature and amount of radioactive material O released to the environs.

F. The results of any environmental surveys performed outside the facility.

G. A summary of significant (above 500 mrem) radiation exposures received by facility personnel and visitors in any one year, including the dates and times of significant exposures.

6.5.2 SPECIA!. REPORTS (REPORTABLE OCCURRENCES) 6.5.2.1 There shall be a report not later than the following working day (by telephone or telegraph to the Director, NRC Region V Inspection l 8 Enforcement Office) and a report within 10 working days (in writing j to the Director, Division of Licensing, USNRC, Washington, D.C. 20555) of:

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A. Release of radioactivity from the reactor above allowed limits, as provided by section 3.8.1 of this specification.

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C B. Any of the following:

B.1 a violation of the Technical Specification or the facility license; B.2 e in an unanticipated excess or total of $0.90 or uncontrolled reactivityreactivity chang in excess of $3.00; B.3 an uncontrolled or unanticipated release of radioactivity from the site; B.4 a safety system component malfunction or other system or component malfunction which renders or threatens to render the safety systt:m incapable of performing its intended safety function; B.5 an observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy cuases or could have caused the existence or development of an unsafe condition with regard to reactor operation; B.6 abnormal degradation of reactor fuel as revealed by periodic inspection; B.7 if the power level exceeds 135 kw.

6.5.3 WRITTEN REPORTS A written report within 30 days to the Commission of:

A. Permanent changes in the facility organization structure; B. Significant changes in the transient or accident analysis as described in the Safety 8 :alysis Report; C. Substantial variances of safety related operating characteristics from previously predicted or measured values.

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6,6 RECORDS Records of the following activities shall be maintained and retained for the periods specified below. The records may be in the form of logs, data sheets, or other suitable forms. The required information ,

may t,e contained in single, or multiple records, or a combination thereof. Recorder charts showing operating parameters of the reactor (i.e., power level, temperature, etc.) for unscheduled shutdown and significant unplanned transients shall be maintained for a minimum period of two years.

6.6.1 RECORDS TO BE RETAINED FOR A PERIOD OF AT LEAST FIVE YEARS A. Reactor operations, including unscheduled shutdowns and tests and experiments performed. Note: Supporting documents such as checklists, log sheets, chart recordings, etc., shall be maintained for a period of at least two years.

B. Principal maintenance operations.

C. Reportable occurrences.

D. Reviews of changes made to the facility or procedures and reviews of tests and experiments performed without prior approvals by the U.S. Nuclear Regulatory Commission pursuant to 10 CFR 50.59.

E. Shipments of radioactive materials.

F. Releases of gaseous and liquid wastes to the environs.

G. Facility radiation and environmental surveys required by the Technical Specifications.

H. Fuel inventories and fuel transfers.

6.6.2 RECORDS TO BE RETAINED FOR AT LEAST ONE REQUALIFICATION CYCLE OR FOR THE LENGTH OF EMPLOYMENT OF THE INDIVIDUAL, WHICHEVER IS SMALLER A. Retraining and requalification of licensed operations personnel.

However, records of the most recent complete cycle shall be maintained at all times the individual is employed.

6.6.3 RECORDS TO BE RETAINED FOR THE LIFETIME OF THE REACTOR FACILITY (NOTE: ANNUAL REPORTS MAY BE USED WHERE APPLICABLE AS RECORDS IN THIS SECTION.)

A. Radiation exposure for all personnel monitored.

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1 6.1.4 SELECTION AND TRAINING Of PERS0'iNEL q

The selection, training, and requalification of personnel shall meet or exceed the requirements of ANS-15.4/N 380 and Appendix A of 10 CFR Part 55 and be in accordance with the requalification plan approved by the Comission.

6.1.5 REVIEW AND AUDJT There is a Reactor Use Comittee which rev'O.?s and approves new experi-ments and proposed alterations to the reat. Lor. The Comittee shall review and audit reactor operations for safety.

6.1.5.1 Composition and Qualifications This comittee shall be corposed of the reactor supervisor and radiation health physicist, both ex officio (voting), and at least 3 other members having expertise in reactor technology. Conr.ittee members shall be appointed by the Dean of the School of Engineering and Applied Science.

6.1.5.2 Charter and Rules The comittee shall function under the following operating rules: 1 A. The Reactor Use Comittee shall meet at least semiannually and shall keep writte.. records of its meetings. The Comittee O shall report directly to the Dean of the School of Engineering and Applied Science.

B. A quorum shall be three members.

C. Any action recomended by the Reactor Use Comittee, which may affect the operations and/or safety of the University comunity beyond the Nuclear Energy Laboratory facility, shall be brought to the attention of the Campus Radiation Safety Comittee which shall have veto power to such a recommendation.

D. The Comittee may appoint one or more qualified individuals to perform the Audit Function.

6.1.5.3 Review Function The following items shall be reviewed by the review group or a subgroup thereof:

A. Determinations that proposed changes in equipment, systems, tests, experiments, or procedures do not involve an unreviewed safety question.

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

C. Tests and experiments in accordance with section 6.3.

D. Proposed changes in technical specifications or license.

E. Violations of technical specifications, or license. Violations of internal procedures or instructions having safety significance.

F. Operating abnonnalities having safety significance, and audit reports.

G. Reportable occurrences listed in section 6.5.3.

6.1.5.4 Audit Function The audit function shall include selective (but comprehensive) examination of operating records, logs, and other documents. Where necessary, dis-cussion with cognizant personnel shall take place. In no case shall an individual audit in an area for which he has operational responsibility.

The following items shall be audited:

(a) Facility operations for confonnance to the technical specifications and applicable license or charter conditions, at least annually.

(b) The retraining and requalification program for operating staff, at least biennially (interval between audits not to exceed 30 months).

(c) The results of action taken to correct those deficiences that may occur in the reactor facility equipment, systems, structures, or method of operations that affect reactor safety, at least annually.

(d) The reactor facility emergency plan, and implementing procedures at least biennially (interval between audits not to exceed 30 months).

Deficiencies uncovered that affect reactor safety shall immediately be reported to Level 2 management. A written report of the findings of the audit shall be submitted to the review and audit group members within three months after the audit has been completed.

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6.2 PP.3_C E D'JP.E S The facility shall be operated and maintained in accordance with approved written procedures. All procedures and major changes thereto shall be reviewed and approved by the Director of the Nuclear Energy Labora-tory prior to being effective. Changes which do not change the origi-nal intent of a procedure may be approved in writing by the reactor supervisor. Such changes shall be recorded and submitted to the Director for routine review. The following types of written procedures shall be maintained:

A. Normal startup, operation and shutdown procedures for the reactor. These procedures shall include applicable checkoff lists and instructions.

B. Procedeces which delineate the operator action required in the event of specific malfunctions and emergencies.

C. Radiological control procedures for all facility personnel.

D. A laboratory emergency procedure to guide the behavior and action of all personnel in the event of an emergency condition.

E. Procedures for the installation, operation and removal of l

experiments where reactor safety is concerned.

F. Procedures for handling irradiated and unirradiated fuel elements.

G. Procedures for operation of the Pneumatic Sample Transfer System.

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