ML20049J378
| ML20049J378 | |
| Person / Time | |
|---|---|
| Site: | 05000112 |
| Issue date: | 03/05/1982 |
| From: | OKLAHOMA, UNIV. OF, NORMAN, OK |
| To: | |
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| ML19268A881 | List: |
| References | |
| NUDOCS 8203180021 | |
| Download: ML20049J378 (33) | |
Text
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- APPENDIX A ,
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! LICENSE NO. R-53 1
TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF OKLAHOMA l
REACTOR MODEL AGN-211P (S.N. 102)
March 5, 1982 i
- Docket No. 50-112 1
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i 8203180021 820312 i- PDR ADOCK 05000112 P PDR,
TABLE OF CONTENTS PAGE 1.0 DEFINITIONS 1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4 2.1 Safety Limits ..................... 4 3.0 LIMITING CONDITIONS FOR OPERATION 6 3.1 Reactivity Limits ................... 6 3.2 Control and Safety Systems .............. 8 3.3 Limitations on Experiments .............. 12 3.4 Shielding ....................... 13 4.0 SURVEILLANCE REQUIREMENTS 14 4.1 Reactivity Limits ................... 14 4.2 Control and Safety System ............... 15 5.0 DESIGN FEATURES 16 5.1 Reactor ........................ 16 5.2 Fue l St ora ge . . . . . . . . . . . . . . . . . . . . . . 17 5.3 Reactor Room . . . . . . . . . . . . . . . . . . . . . . 17 5.4 Safety and Control Rods ................ 18 5.5 Shielding ....................... 20 5.6 Water Purification System ............... 20 5.7 Experimental Facilities ................ 20 5.8 Instrumentation and Control s . . . . . . . . . . . . . . 20 5.9 Fu e l St o r a g e . . . . . . . . . . . . . . . . . . . . . . 21 5.10 Reactor Room . . . . . . . . . . . . . . . . . . . . . . 23 5.11 Sec u ri ty . . . . . . . . . . . . . . . . . . . . . . . . 23 6.0 ADMINISTRATIVE CONTROLS 23 6.1 Organization ..................... 23
1 1.0 DEFINITIONS The terms Safety Limit (SL), Limiting Safety System Setting (LSSS), and Limiting Conditions for Operation (LCO) are as defined in 50.36 of 10 CFR part 50.
1.1 Reactor Shutdown - The reactsr shall be considered shutdown when all safety (2) and control rods (2) are fully inserted.
1.2 Reactor Operation - Reactor operation is any condition wherein the reactor is not shutdown; i.e., at least one rod is partially or fully out.
1.3 Measuring Channel - A measuring channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring or responding to the value of a process variable.
1.4 Safety Channel - A safety channel is a measuring channel in the reactor safety system.
1.5 Reactor Safety System - The reactor safety system is that combination of safety channels and associated circuitry which forms an automatic protective system for the reactor or provides information which requires manual protective action be initiated.
1.6 Reactor Component - A reactor component is any apparatus, device, or material that is a normal part of the reactor assembly, 1.7 Operable - Operable means a component or system is capable of performing its intended function in its normal manner.
1.8 Operating - Operating means a component or system is performing its intended function in its normal manner.
1.9 Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification may include comparison of the channel with other independent channels or methods measuring the same variable.
1.10 Channel Test - A channel test is the introduction of a signal into the channel to verify that it is operable.
1.11 Channel Calibration - A channel calibration is an adjustment of the channel such that its output responds, within acceptable range and accuracy, to known values of the parameter which the channel measures.
Calibration shall encompass the entire channel, including the equipment, actuation, alarm, or trip.
1.12 Reactor Utilization - Utilization of the reactor is defined by three (3) categories:
1.12 a - Irradiation of a sample in or near the core or by a beam brought out of the reactor.
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1.12 b - Experiment is any use of the reactor or a component of the reactor for other than a sample irradiation.
1.13 c - Special operation of the reactor for the sole purpose of evaluating the reactor's performance or testing the instrumentation.
1.13 Secured Experiment or Sample - Any experiment or sample, or component of an experiment is deemed to be secured, or in a secured position, if it is held in a stationary position relative to the reactor by mechanical means. The restraint shall exert sufficient force on the experiment or sample to overcome the expected effect of hydraulic, pneumatic, bouyant, or other forces which are normal to the operating environment of the experiment or sample or which might arise as a result of credible mal functions.
1.14 Unsecured Experiment or Sample - Any experiment or sample, or component of an experiment is deemed to be unsecured whenever it is not secured as defined in 1.13 above. Moving parts of experiments are deemed to be unsecured when they are in motion.
1.15 Movable Experiment or Sample - A movable experiment .is one which may be inserted, removed, or maripulated while the reactor is critical.
1.16 Removable Experiment - A removable experiment is any experiment, experimental facility, or component of an experiment, other than a perraanently attached appurtenance to the reactor system, which can reasonably be anticipated to be moved one or more times during the life-of the reactor.
1.17 Experimental Facilities - Experimental facilities are those portions of the reactor assembly that are used for the introduction of experiments or samples into or adjacent to the reactor core region or allow beams of
.adiation to exist from the reactor shielding. Experimental facilities shati ir.clude the thermal column, glory hole, and access ports.
1.18 Potential Reactivity Worth -The potential reactivity worth of an experiment or a fample is the maximum value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment or sample position or configuration.
The evalaation must consider possible trajectories of the experiment in motion relative to the reactor, its orientation along each trajectory, and circumstances which can cause internal changes such as creating or filling of void spaces or motion of mechanical components. For removable experiments, the potential reactivity worth is equal to or greater than the static reactivity worth.
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1.19 Static Reactivity Worth -The static reactivity worth of an experiment is the value or the reactivity change which is measureable by calibrated control or regulating rod comparison methods between two defined terminal postions or configurations of the experiment. For removable experiments, the terminal postions are fully removed from the reactor and fully inserted or installed in the normal functioning or intended position.
1.20 Explosive Material - Explosive material is any solid or liquid which is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in " Dangerous Properties of Industrial Materials" by N. I. Sax, Third Ed. (1968), or is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Association in its pubicatior 704-M,1966, " Identification System for Fire Hazards of Material," also enumerated in the "Handbood for Laboratory Safety" 2nd Ed. (1971) published by The Chemical Rubber Co.
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2.0 SAFETY LIMITS AhD LIMITED SAFETY SYSTEM SETTINGS 2.1 Safety Limits Applicability This specification applies to the maximum steady state power' level and maximum core temperature during steady state or transient operation.
Objective To assure that the integrity of the fuel material is maintained and that fission products are retained in the core matrix to the extent possible.
Specification
- a. The reactor power level shall r.ot exceed 100 watts.
- b. The ..laximum core temperature shall not exceed 392*F under any conditions.
Bases ,
The polyethylene core material does not melt Selow 392"F and is expected to maintain its integrity and retain essentially all of the fission products at temperatures below 392*F. The corresponding maximum core temperature wculd be well below 392*F thus assuring integrity of the core and aiding the retention of fission products.
In addition, the OU AGN 211P fuel elements have been clad (painted) with an epoxy. This epoxy has further served to reduce gaseous fission product migration into the pool water. Prior to going to 100 watts all fuel elements will be stripped and recoated if there is any indication of epoxy degradation. After operation at 100 watts, the recoating probably will not be possible due to the fission product gamma-ray radiation emitted by the elements.
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The University of Oklahoma reactor has operated at 15 watts over the past 20 years. The Reactor Manufacturer's Operations Manual (Dec.1958)
-tates that the reactor may be operated intermittently at a power level of 1000 watts with polyethylene-UO2 fuel elements. The University of Oklahoma Reactor's fuel elements have shown no evidence of deterioration over the years. In August 1980, the epoxy clad on twelve fuel elements was totally removed and replaced with fresh epoxy. Prior experience indicates the epoxy will be good for several years. It is anticipated that at power levels over 15 watts an increase in the gaseous fission product concentrations in the pool water can be expected. At 15 watts the air monitor, which senses air drawn directly from above the pool water surface, sees no activity above background. We expect none up to powers of 100 watts in spite of the expected increase in the water concentration.
To ensure that no unknown factors of importance are present, a power increase program will be employed whereby the reactor is operated at 50 watts and 75 watts for several days prior to going to 100 watts. The concentration of gaseous fission products in the pool water will be monitored at each power level. Special emphasis will be placed on detection of fission gases in the air above the pool water.
It is also noted that detailed radiation surveys taken at 15 watts indicate the facility may be operated at 100 watts in full comformity with regulations (10 CFR-Part 20) and good radiation safety practice. We note also that the University of West Virginia operated its AGN 211 at 75 watts for a number of years without problems.
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3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity limits Applicability This specification applies to the reactivity condition of the reactor and the reactivity worths of control rods and experiments.
Objective To assure that the reactor can be shut down at all times (normal and emergency) and that the safety limits will not be exceeded.
Specifications 3.11 The total excess reactivity (p), including the static positive reactivity worth of any experiments or samples and with all safety and control rods withdrawn will not, under any conditions, exceed 0.65% at any pool water temperature. The purpose of this rigid limit is to ensure that it is not possible to put the reactor on a prompt period under any conditon.
3.12 The reactor will operate in one of two general core configurations:
- a. Standard,12 fuel element, parallelepiped with a graphite or water reflector,
- b. Flux trap, 20 or less fuel element configuration, with a one or two element void in the center of the core and with a graphite or water reflector.
3.13 To ensure safety and to greatly lessen the possibility of mistakes, a maximum of five flux trap cores (fuel and reflector) _
will be defined. The exact value of the excess reactivity for each flux trap core will be determined by a critical experiment.
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Additionally, nuclear safety channels number 1 and 2 will be checked to ensure that power calibration has not been altered and fine and course control rod worths will be measured.
3.14 The shutdown margin with the most reactive safety or control rod fully inserted shall be at least 1% ak/k.
3.15 The reactivity worth of the control and safety rods shall ensure sub-criticality on the withdrawal of the coarse control rod or any safety rod.
3.16 The reactivity worth of all new experiments will be determined before the experiment is performed.
3.17 A change in reflector requires a recheck of excess reactivity.
Bases The limitations on total core excess reactivity assure reactor periods of sufficient length so that the reactor protection system and/or operator action will be able to shut the reactor down without exceeding any safety limits. The shutdown margin and control and safety rod reactivity limitations assure that the reactor can be brought and maintained subcritical if the highest reactivity rod fails to scram and remains in its most reactive position.
The flux trap configuration adds considerably to the educational value of the reactor. This is important since the reactor is used most heavily in direct support of the undergraduate and graduate nuclear engineering program. Additionally, the flux trap offers a ne tron flux in the trap equivalent to a power level of about 150 watts.
The static reactivity limit of 0.65% p applies equally to both core configurations. At 20*C and with the glory hole empty the reactor fuel loading will be adjusted so that an excess reactivity of about 0.4% is 7
available. The most positive reactivity that can be added is to insert a pure polyethylene rod completely filling the glory hole. This adds about 0.1% giving a total excess of 0.5%. This means (for example) that the fuel temperature could be reduced by 5'C (0.02%/'C x 5*C = 0.10%) by lowering the pool water tempe.ature, thereby providing an additional 0.10% excess, for a total of 0.60%.
3.2 Control and Safety Systems Applicability These specifications apply to the reactor control and safety systems.
Objective To specify lowest acceptable level of performance, instrument set points, and the minimum number of operable components for the reactor control and safety systems.
Specification 3.21 The total scram insertion time of the safety rods and coarse control rod shall be about 250 milliseconds.
3.22 The safety rods and coarse control rod shall be interlocked such that:
- 1. Reactor startup cannot commence unless both safety rods are fully withdrawn.
- 2. Only one safety rod can be withdrawn at a time.
- 3. The coarse control rod cannot be withdrawn unless both safety rods are fully withdrawn.
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3.23 Nuclear safety channel instrumentation shall be operable in accordance with Table 3.1 whenever the reactor control or safety rods are not in their fully inserted position. A neutron source must be present in the reflector.
3.24 Loss of electric power will result in a scram.
3.25 Pool water conductivity will be measured prior to and immediately after each reactor operation. A conductivity of less that 10 micrombos/cm is to be maintained.
3.26 High Power Pool Shield -At any power level above 50 watts, quarter inch iron-plate (one inch total thickness) gamma-shields will be installed. The iron-plate gamma shields span the top of the pool tank reducing the background in the reactor room from gamma rays resulting from neutron capture in the pool water.
3.27 Pool Water Monitor (PWM) -The pool water monitor refers to the system which consists of a 2 x 2 NaI (T1) detector placed at the center of a cylindrical tank filled with water from the reactor pool. A pump circulates the pool water to the water-monitor detector and through an ion-exchanger and filter system and back to the pool . The water intake to the water monitor system is located at the bottom of the pool near the bottom edge of the graphite reflector.
The purpose of the water monitor system is to measure the buildup of radioactivity in the pool water during reactor operation. Because the fuel elements are epoxy clad polyethylene-uranium oxide, some high-yield gaseous fission products will slowly diffuse out of the fuel into the water as the fuel temperature and the concentration of the fission products in the fuel elements increase.
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TABLE 3.1 REACTOR SAFETY SYSTEMS WHICH PRODUCE SCRAMS i
DEVICE (SAFETY CHANNEL) FUNCTION LIMITING SAFETY SYSTEM SETTINGS
- 1. Log Power Channel High Power Limit 150 Watts (Nuclear Safety Channel-1)
Low Power Limit less than neutren startup source current reading Short Period Limit 5 seconds 9 Linear Power Channel High Power 120 watts 3 (Huclear Safety Channel-2)
- 3. Reactor Tank Water Level Low Pool Water 9 inches below reactor tank top
- 4. Pool Water Temperature Low Pool Water Temperature 5 C
- 5. Pool Water Circulating Pump Pump must be operating ; cram if pump power not on
- 6. Pool Water Monitor Detect excessive water activity Scram at 2 times level found after five
, hours of operation at 100 watts (this is not equilibrium)
- 7. Manual Scram Normal Shutdown Scram on operator decision
Bases The specifications on scram insertion time, in conjunction with the safety system instrumentation and set points, assure safe reactor shutdown during the most severe foreseeable transients. Interlo-ks on control and safety rods assure an orderly approach to criticality and an adequate shutdown capability.
The neutron detector channels (nuclear safety channels 1 and 2) assure that reactor power levels are adequately monitored during reactor startup and operation. Requirements on minimum neutron levels will prevent reactor startup unless the log power channel is operable and responding, and will cause a scram in the event of instrumentation failure. The power level scrams initiate redundant automatic protective action at power low enough to assure safe shutdown without exceeding any safety limits. The period scram conservatively limits the rate of rise of reactor power to periods which are manually controllable and will automatically scram the reactor in the event of unexpected large reactivity additions.
The AGN-211's negative temperature coefficient of reactivity causes a reactivity increase with decreasing core temperature. The shield water temperature interlock will prevent reactor operation at temperatures below 5*C thereby limiting potential reactivity additions associated with temperature decreases.
Water in the shield tank is an important component of the reactor shield and operation without the water may produce excessive radiation levels. The shield tank water level interlock will prevent reactor operation without adequate water levels in the shield tank, t
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e The manual scram allows the operator to manually shut down' the reactor if an unsafe or otherwise abnormal condition occurs that does not otherwise scram the reactor. A loss of electrical power de-energizes the safety and coarse contrni rod holding magnets causing a reactor scram, thus assuring safe and immediate shutdown in case of a power outages.
3.3 Limitations on Experiments Applicability This specification applies to experiments installed in the reactor and its experimental facilities.
Objective To prevent damage to the reactor or excessive release of radioactive materials in the event of an experimental failure.
Specifications
- a. Experiments or samples containing materials which might damage reactor components by corrosion shall be doubly encapsulated.
- b. Explosive materials shall not be inserted into experimental facilities of the reactor unless approved by the Reactor Safety Committee,
- c. The radioactive material content, including fission products of any experiment shall be limited so that the complete release of all gaseous, particulate, or volatile components from the experiment will not result in doses in excess of 10% of the equivalent annual doses stated in 10 CFR Part 20 for persons occupying r;stricted areas during the length of time required to evacuate the restricted area.
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- d. No part of the concrete biological shield will be removed or replaced to accommodate experiments without the written permission of the Reactor Safety Committee.
Bases These specifications are intended to reduce the' likelihood of damage to reactor components and/or radioactivity releases resulting from an experimental failure and to protect operating personnel and the public from excessive radiation doses in the event of an experimental failure.
3.4 Shielding Applicability This specification applies to reactor shielding required during reactor operation.
Objective The objective is to protect facility personnel and the public from radiation exposure.
Specification The following shielding requirements shall be fulfilled during reactor operation:
- a. The reactor shield tank shall be filled with water to a height within 9 inches of the top of the pool. This 'results in a gamma exposure rate of about 15 mr/hr at 15 watts.
- b. One inch of iron plate will be layed across the pool for operation at power levels above 50 watts. This will ensure a gamma level of about 0.1 mr/hr. at the reactor console at 100 watts.
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Bases The facility shielding in conjunction with designated restricted radiation areas is designed to limit radiation dose to facility personnel and to the public to a level below 10 CFR 20 limits under operating conditions, and to a level below criterion 19, Appendix A,10 CFR 50 recommendations under accident conditions. Because the Reactor Laboratory has moderate student use, a very conservative exposure policy 4
has always and will continue to be followed.
4.0 SURVEILLANCE REQUIREMENTS Actions specified in this .section are not required to be performed if during the specified surveillance period the reactor has not been brought critical or is maintained in a shutdown condition extending -
beyond the specified surveillance period. However, the surveillance requirements must be fulfilled prior to subsequent startup of the reactor.
4.1 Reactivity Limits Applicability This specification applies to the surveillance requirements for reactivity limits.
Objective To assure the reactivity limits for Specification 3.21 are not exceeded.
l Specification
- a. Safety and control rod reactivity worths shall be measured annually, i
but at intervals not to exceed 18 months.
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- b. Total shutdown margin shall be determined annually, but at intervals not to exceed 18 months.
- c. The reactivity worth of an experiment or a sample shall be estimated or measured, as appropriate, before or during the first startup after the experiment or sample is inserted.
- d. The total excess reactivity will be estimated prior to or during the initial phases of every startup. Usual procedure will be to go to 0.1 watt and determine the total excess from the calibrated rod worths. This is deemed necessary since reactivities will be sightly different for the two core configurations.
Bases The control and safety rod reactivity worths are to be measured to assure that no degradation or unexpected changes have occurred which could adversely affect reactor shutdown margin or total excess reactivity. The shutdown margin and total excess reactivity are determined to assure that the reactor can always be safely shutdown with one rod not functioning and that the maximum possible reactivity insertion will not result in reactor periods shorter than those that can be adequately terminated by either operator or automatic actions. Based on known experience with AGN reactors, significant changes in reactivity or rod worth are not expected with an 18-month period (we have experienced no change in rod worths over the past 20 years).
4.2 Control and Safety System Applicability This specification applies to the surveillance requirements of the reactor control and safety systems.
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Specification
- a. Safety and control rod scram times shall be measured at intervals not to exceed 12 months.
- b. Safety and control rods and drive shall be inspected for deterioration at intervals not to exceed 12 months.
- c. Channel tests of all safety channels listed in Table 3.1 (except safety channels 3 and 4) will be performed prior to the first reactor startup each day.
- d. The period, count rate, and power level measuring channels shall be calibrated and set points verified at intervals not to exceed 12 months,
- e. The shield water level interlock, shield water temperature and pool activity interlocks shall be calibrated by perturbing the sensing element to the appropriate set point. These calibrations shall be performed at intervals not to exceed 12 months.
Bases The channel tests and checks required will assure that the safety channels and scram functions are operable. Based on operating experience with reactors of this type, scram measurements, channel calibrations, set point verifications, and inspections are of sufficient frequency to assure, with a high degree of confidence, that the safety system settings will be within acceptable drift tolerance for operation.
5.0 DESIGN FEATURES 5.1 General Design The AGN 211P nuclear reactor consists of the reactor unit and the control console. The reactor unit consists of a core assembly in a 16
water-filled tank suspended by means of an aluminum frame attached to a steel I-beam bridge mounted at the top of the tank. The control rod actuator assemblies are mounted on the bridge with the absorber control rods mounted beneath. The reactor tank is a 3/8' thick 40" x 60" x 8' high tank shielded by a number of concrete blocks 20" x 20" x 40". A platform and railing are secured to the top of the reactor. The reactor tank is empty below floor level (sunk into the ground).
5.2 Core Assembly The core assembly consists of 2-7/8" x 3-1/16" x 20" fuel elements.
The base of the element has an aluminum guide which fits into the i
aluminum grid plate. The grid plate is essentially identical to the one used in the Bulk Shielding Facility. The grid plate is supported from the I-beam bridge by an aluminum frame. The bridge also acts as a support for the ion chambers.
The elements are removable from above by grasping the element handle with the proper removal tool. The control rods fit into the core assembly in such a manner that a control rod cannot be removed without removing a fuel element.
The lattice is arranged so that each fuel element in the assembly is surrounded by 1/8" of water. One watt operation results in a temperature rise at the center of the element of approximately 0.07'F, hence, at 100 watts a fuel temperature equilibrium of approximately 7*F over ambient is expected (this temperature increase will reduce the excess reactivity by about 0.1%).
5.3 Fuel Elements Each fuel element is 2-7/8" x 3-1/16" x 20". The center 10" of each element is composed of the homogeneous mixture of UO2 in polyethylene 17
(56 mg U-235/cm 3 ) in the form of 20 micron diameter particles. The density of uranium dioxide is 318 mg/cm3 The elements are coated with i
an epoxy to ensure that the U02 particles are not exposed to the water.
A fuel element is shown in Figure 1 and is constructed as follows:
- 1. Bottom aluminum mounting section l 2. Five inches of carbon reflector
- 3. Ten inches of fuel 4 Five inches of carbon reflector
- 5. Two inches of lead shadow shield
- 6. The top aluminum removal fitting 5.4 Safety and Control Rods The AGN 211P has a fine and a coarse control rod plus two safety rods. The safety rods are made of Boral, the coarse control rod of aluminum clad Cadmium and the fine control rod of stainless steel. The l rods are in the form of 2-3/4" wide blades which fit into slots in the edge of four specially constructed fuel elements.
During operation, the rods are held in the out position by eletro-magnets. A scram signal deenergizes the holding magnets allowing the safety and coarse control rods to be accelerated into the core by both gravity and spring loading. The spring constant provides an initial acceleration of Sg which gives a total injection time of approximately 300 milliseconds. The fine rod is rigidly mounted to its carriage and is driven into the core by the automatic return system which drives the magnets to their initial postion. All rod operation is manual (no automatic control).
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The reactivity controlled by the rods is approximately 1.25% for the coarse and safety rods and 0.45% for the fine rod.
5.5 Shielding Shielding in the radial direction is provided by 40" of ordinary concrete. Shielding in the vertical direction is provided by 5" of graphite, 2" of lead attached to the top of the fuel elements, and 4-1/2 feet of water.
5.6 Water Purification System The water purification system consists of a demineralizer and circulating punp to maintain the water purity. The water loop has a radiation monitor and conductivity meter, both outputs being displayed at the console. The radiation monitor is connected to a scram circuit so that the reactor cannot be operated if activity in the water system exceeds the scram setting.
5.7 Experimental Facilities The AGN 211P is equipped with a 7/8" diameter horizontal glory hole through the core. Samples may be placed in the glory hole at varying positions in the core and reflector.
A horizontal access port through the reflector, and a horizontal beam port which faces the reflector is present.
The variable core and reflector configurations make it possible to test different types of reflector elements or to leave water filled voids in the core in t ich large samples may be placed for irradiation.
5.8 Instrumentation and Controls The neutron flux in the reactor is monitored by two neutron detectors located near the reactor core and corresponding flux indicators 20
on the console. These detectors are connected respectively to a logarithmic micro-microammeter and a linear microammeter. Strip chart recorders record the output of both channels.
Each of the indicators is connected to a scram circuit so that if any indicator reads above or below the respective limiting safety system settings, it de-energizes the holding magnets, and shuts down the reactor. A rate, or period scram, is connected to the logarithmic channel to prevent too rapid a power rise. Additional safety interlocks provide for reactor shutdown if the water level drops below 9 inches, if the water temperature falls below 5'C, if the radioactivity level of the water rises above a preset level, or if the circulating pump loses power.
The rods are controlled by manual hold-down switches, the two control rods having a normal and slow speed. The positions of the control rods are indicated to 0.01 cm on the console. Sequenced interlocks of the control switches require that only one safety rod be moved at a time, and that both safety rods be fully withdrawn before the coarse rod can be moved.
5.9 Fuel Storage The University of Oklahoma is authorized a total of 24 standard AGN 211 fuel elements. A standard core required 12 elements chosen to provide an excess reactivity of about 0.4% at 20'C (glory hole empty). A flux trap care requires 16 elements which are selected to provide about 0.4% excess at 20'C.
All fuel elements not in the reactor are stored in the fuel storage pit immediately adjacent to the reactor. Each fuel elecent is stored in a steel cylinder rigidly held in a subcritical configuration when in an infinite water reflector. See Figure 2.
21
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5.10 Reactor Room The reactor is located in Room 107A, in the Nuclear Engineering Laboratory, 905 Asp Avenue. This room has 20,000 square feet and is a ground ficor room. The reactor pool is partially sund into the floor.
The arrangement of the room is shown in Figure 3.
5.11 Security 1
The reactor room has two entrances, A & B. Both have magnetic door latches which are energized. Entrance B opens to a fire lane.
Three four inch steel pipes are cemented into the ground to prevent door B from being rammed by a vehicle. A microwave detector is located at point C which will detect movement near the reactor or full storage i
pit.
i The microwave and magnetic door latches are activated at 5:00 p.m.
each weekday and all day Saturday and Sunday. Both systems alarm at the University Police Center. Officers are able to respond within two minutes.
Doors A and B are under restricted control. Five keys are currently issued.
6.0 ADMINISTRATIVE CONTROLS 6.1 Organization of Responsibility for the University of Oklahoma Nuclear Reactor - Administrative control of the reactor is diagrammed in Figure 4.
6.11 President - Chief Executive Officer of the University 6.12 Provost - Chi.ef Academic Officer i 6.13 V.P. for Administrative Affairs - Chief Administrative Officer 23
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i N L__________3 FIGURE 3. FLOOR PLAN OF REACTOP AREA IN g NUCLEAR ENGINEERING d~;!LDING L
AY
PRESIDENT UNIVERSITY l OF l OKLAHOMA l r
VP ADMINISTRATIVE PROVOST AFFAIRS RADIATION ENGINEERING SAFETY 4 DEAN OFFICER g r J
! I
' I RADIATION l DIRECTOR SAFETY -- q g SCHOOL COMMITTEE I I AktiE
, I I a j i l l l I I REACTOR I I- REACTOR SAFETY DIRECTOR COMMITTEE - - - - - -
(
REACTOR SUPERVISOR FIGURE 4. ADMINISTRATIVE ORGANIZATION OF UNIVERSITY OF OKLAHOMA REACTOR AGN 211P REACTOR STAFF 25
6.14 Dean of Engineering - Administrative Officer in charge of the College of Engineering 6.15 Director, School of Aerospace, Mechanical and Nuclear Engineering
( ANNE) - The Director is the head of the three degree programs, Aerospace, Mechanical and Nuclear Engineering. The reactor, while operated as a facility available to all appropriate university personnel, is deemed to be a facility for which the Director is-responsible. The Director will provide proper financial and technical support to ensure operation of the facility in accordance with all laws.
6.16 Reactor Director - Responsible for the general administration of the facility. In this capacity the Director shall have the authority and responsibility for the facility and, within the limitations set forth by the facility license, shall make policy decisions on all phases of the reactor operation, appoint personnel and be advised in all matters concerning reactor safety by the Reactor Safety Committee, the Radiation Safety Committee and the University Radiation Safety Officer.
6.17 Reactor Supervisor - Has over-all responsibility for the daily operation and maintenance of the reactor facility and training and supervision of reactor staff personnel. The Reactor Supervisor is responsible to the Reactor Director.
All operations which include physical alteration of the fuel will be directly supervised by the Reactor Supervisor or the Reactor Director.
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6.18 Reactor Safety Committee -The Reactor Safety Committee shall be composed of not less than five members, of whom no more than three nor less than one are members of the reactor operating organization. The Committee members are appointed by the VP for Administrative Affairs and shall meet on call of the Chairman and they shall meet at least quarterly.
Three members shall constitute a quorum and written minutes will be kept. The Committee shall be responsible for, but not limited to the following: (1) reviewing and approving safety standards associated with the use of the facility, (2) reviewing and approving all proposed experiments and procedures and changes thereto, and modifications to the reactor and its associated components, (3) determining whether proposed experiments, procedures or modifications involve unreviewed safety questions, as defined in 10 CFR 50, Part 50.59(c), and are in accordance with these Technical Specifications, (4) conducting periodic audits of procedures, reactor operations and maintenance, equipment performance, and records, (5) reviewing all reported abnormal occurrences and violations of these Technical Specifications, evaluating the causes of such events and the corrective action taken and recommending measures to prevent reoccurrence and, (6) reporting their findings and recommendations concerning the 'bove items to the Reactor Director and, where appropriate, to higher aut Jrities.
University Radiation Safety Officer (RS0) - Has over-all responsibility for radiological safety at the University and shall see that necessary radiatD+ surveys are conducted in the Reactor and other associated laboratories.
The Radiological Safety Officer shall review the radiological safety aspects of reactor experiments and shall advise the Reactor Director in 27
i all matters relative to radiological safety of the reactor and its operations. The RSO may be a member of the Reactor Safety Committee and of the Radiation Safety Committee.
6.19 Reactor Staff Oualifications - All personnel directly associated with the reactor shall meet the minimum standard qualifications set forth in ANS 15.4 " Standard for Selection and Training of Personnel for Research Reactors." The Reactor Director shall see that Training and requalification programs are conducted as required. <
6.20 Procedures - Detailed written procedures shall be provided for:
- a. Normal Start-up
- b. Normal Operation
- c. Normal Shut-down
- d. Refueling or rearrangement of fuel
- e. Emergency procedures involving radiation
- f. Special maintenance which could conceivably affect reactor safety.
6.21 Record Retention - Records to be retained for a period of not less than six years are:
- a. Operating logs
- b. Check-out and shut-down forms
- c. All data relevant to reactor performance
- d. Principal maintenance operations
- e. Facility contamination surveys
- f. Records of experiments performed 9 Production of radioisotopes 28
6.22 Records to be retained for the life of the facility -
- a. Gaseous and liquid radioactive effluents intentionally released to the environs
- b. Any off-site environmental mor-itoring surveys
- c. Fuel inventories and transfers
- d. Facility drawings
- e. Records on training of personnel
- f. Reactor Safety Committee minutes 6.23 Safety Limit Violation The following actions shall be taken in the event a Safety Limit is violated:
- a. The reactor will be shut down immediately and reactor operation will not resume without the concurrence of the Reactor Safety Committee, the Reactor Director and the Director of AMNE.
- b. The Safety Limit violation shall be reported to the appropriate NRC Regional Office of Inspection and Enforcement, the Director of the NRC, and the Reactor Safety Committee no later than the next work day.
- c. A Safety Limit dolation Report shall be prepared for review by the Reactor Safety Committee. Tnis report shall describe the applicable circumstances preceding the violation, the effects of the violation upon facility components, systems or structures, and corrective action to prevent recurrence.
- d. The Safety Limit Violation Report shall be submitted to the NRC, and Reactor Safety Committee within 30 days of the violation.
29
r 6.24 Reportable Occurrences Reportable occurrences, including causes, probable consequences, corrective actions and measures to prevent recurrence, shall be reported to the NRC.
- a. Prompt Notification With Written Followup. The types of events listed shall be reported as expeditiously as possible by telephone or telegraph to the Director of the appropriate NRC Regional Office, or his designated representative no later than the first work day following the event, with a written followup report within two weeks. Information provided shall contain narrative material to provide complete explanation of the circumstances surrounding the event.
(1) Failure of the reactor protection system subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reached the setpoint specified as the limiting safety system setting in the technical specifications.
(2) Operation of the reactor when any parameter or operation subject to a limiting condition is less conservative than the-limiting condition for operation established in the technical specifications.
(3) Abnormal accumulation of fission products in pool water.
(4) Reactivity balance anomalies involving:
- a. exceeding excess static reactivity limit;
- b. shutdown margin less conservative than specified in these technical specifications; 30
(5) Failure or malfunction of one (or more) component (s) which prevents, or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents.
(6) Personnel error or procedural inadequacy which prevents, or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents.
(7) Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the Safety Analysis Report or in the bases for the Technical Specifications that have permitted reactor operation in a manner less conservative than assumed in the analyses.
(8) Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the Safety Analysis Report to technical specifications bases; or discovery during plant life of conditions not specifically considered in the Safety analysis Report or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.
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