ML19208D382
| ML19208D382 | |
| Person / Time | |
|---|---|
| Site: | Idaho State University |
| Issue date: | 08/30/1979 |
| From: | Gammill W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19208D378 | List: |
| References | |
| R-110-A-002, R-110-A-2, NUDOCS 7909280338 | |
| Download: ML19208D382 (34) | |
Text
.-
t o
O
[
UNITED STATES 11; 9.j
- b NUCLEAR REGULATORY CCMMisSION i
WASHINGTCN, D. C. 20555
\\?...X/
IDAHO STATE 'JNIVERSITY CCCKET NO. 50 N AMENDED FACILITY OPERATING LICENSE Amendmenc No. 2 License No. R-110 1.
The Nuclear Regulatory Comission ( ?.e Comission) has found that:
A.
The application for amendment by Idaho State University (the licensee) dated August 10, 1978, as supplemented Nover2er 21,1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I; 3.
The facility will operate in confor:lity with the application, the provisiens of the Act, and the regu-lations of the Ccmission; C.
There is reasonable assurance (i) that the activities authorized by this amended license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conduc*ad in compliance with the Comissien's regulatiens; D.
The licensee is technically and financially qualified to engage in the activities authorized by this operating license in accordance with the regulaticns of the Ccmission; E.
The licensee is a nonprofit educational institution and will use the facility for the conduct of educaticnal act.ivities and has satisfied the acplicable provisiens of 10 CFR Part 140, " Financial Protection Requirements and Indemnity Agreements," of the Comission's regulations; F.
The issuanc'. of this amended operating license will not be inimical to the cer:non defense and security or to tne health and safety of the public; G.
The receipt, acssession and use of byprbduct and scecial nuclear raterials as authori:ed by this license will be in acenrdance with the Commissien's regulations in 1052 249 730928 03=ss
.=
s
d 2-10 CFR Parts 30 and 70, including 10 CFR Sections 30.33, 70.23 and 70.31; and H.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Ccmmission's regulations and all applicable requirements have been satisfied.
2.
Facility Operating License No. R-110 issued to Idaho State University is hereby amended in its entirety to read as follows:
A.
This license applies to the Model AGN-201M, Serial No.
103, nuclear reactor (the reactor) cwned by Idaho State University. The reactor is located on the licensee's campus in Pocatello, Idaho, and is described in the ifcensee's application for license dated April 3,1967, and subsequent amendments and supplements thereto, including the application for amendment dated August 10, 1978, and supplement thereto dated November 21, 1978.
B.
Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Idaho State University:
(1) Pursuant to Section 104c. of the Act and 10 CFR Part 50, " Licensing of Production and Utilization Facilities," to possess, use and operate the reactor as a utilization facility at the designated location in Pocatello, Idaho; (2) Pursuant to the Act and 10 CFR Part 70, "Special Nuclear Material,* to receive, possess and use up to 7CO grams of contained uranium-235, enriched to 20 percent in uranium dioxide (UO ) embedded in 2
radiation stabilized polyethylene, in connection with operation of the reactor; (3) Pursuant to the Act and 10 CFR Part 30, " Rules of General Applicability to Licensing of 3yproduct Material,* to possess, but not to separate, such byproduct material as may be produced by oceration of the reactor.
C.
This license shall be deemed to contain and is subject to the conditions specified in the following Conmission regulations in 10 CFR Chapter I: Part 20, Section 30.34 n5J?
250 of Part 30, Secticns 50.54 and 50.59 of Part 50 and V
t
. Section 70.32 of Part 70, is subject to all applicable provisions of the Act and to the rules, regulatiens and orders of the Ccmmission now or hereafter ir. effect; and is subject to the additional conditions specified below:
(1) Maximum Power Level The lic,ensee is authorized to cperate the reactor at steady-state pcwer levels up to a maximum of 5 watts (thermal).
(2) Technical Scecificatiens The Technical Specifications centained in Appendix A attached hereto are hereby incorporated in this license. These Technical Specifications supersede the Technical S:ecifications dated October 11, 1967, as amended April 28, 1976.
The licensee shall operate the reactor in accordance with the Technical Scecifications.
(3) Physical Security Plan The licensee shall maintain in effect and fully imolement all provisions of the Commission-approved physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p).
The approved security plan consists of a document withheld frem public disclosure pursuant to 10 CFR 2.790(d), dated June 19, 1978, enti tl ed,
Idaho State University Security Plan.
E.
This amended license is effective as of the date of issuance and shall expire at midnight, December 23,1995.
FOR T}iE NUCLEAR REGULATORY COMMISSION O
Q,v&
William
. Gaanill, Acting Director for Operating Reactor Projects Division cf C; era:ing Reactors A::acn ent:
A::encix A - Technical 5:ecificati:ns
- a:ec: August 30, 1979 1052 25i
4 APPENDIX A TO FACILITY OPERATING LICENSE NO. R-110 TECHNICAL SPECIFICATIONS FOR IDAHO STATE UNIVERSITY AGN-201 M Reactor (Serial 4103)
DOCKET NO. 50-284 4
Amendment No. 2 Date: August 30, 1979 1052 252
a TABLE OF CONTENTS P_agg a
1.0 DEFINITIONS 1
2.0 SAFETY LIMITS AND LIMITED SAFETY SYSTEM SETTINGS 4
2.1 Safety Limits 4
2.2 Limiting Safety System Settings 4
3.0 LIMITING CONDITIONS FOR OPERATION 6
3.1 Reactivity Limits 6
3.2 Control and Safety Systems 6
3.3 Limitations on Experiments 10 3.4 Radiation Monitoring, Control and Shielding 11 4.0 SURVEILLANCE RECUIREMENTS 12 4.1 Reactivity Limits 4.2 Control and Safety System 12 4.3 Reactor Structure 13 14 4.4 Radiation Monitoring and Control 14 5.0 DESIGN FEATURES 16 5.1 Reactor 5.2 Fuel Storage 16 5.3 Reactor Room 18 18 6.0 ADMINISTRATIVE CONTROLS 19 6.1 Organization 6.2 Staff Qualifications 19 6.3 Trainin9 29
.,_6.4 Reactor Safety Cemittee.
25 6.5 Approvals
~
22 6.6 Procedures 24 6.7 Experiments 24 6.8 Safety Limit Violation 25 6.9 Reporting Requirements 25 6.10 Record Retention 25 28 1052 253
I O
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 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 enccmpass the entire channel, including equipment, actuation, alarm, or trip.
1.2 Channel Check - A channel check is a qualitative verification of acceptaDie performance by observation of channel behavior. This verification may include comparison of the channel with other independent channels or methods measuring the same variable.
1.3 Chanrel Test - A channel test is the introduction of a signal into the channel to verify that it is operable.
1.4 Experiment -
a.
An experiment is any of the follcwing:
(1) An activity utilizing the reactor system or its components or the neutrons or radiation generated therein; (2) An evaluation or test of a reactor system operational, surveillance, or maintenance technique; or (3) The material content of any of the foregoing, including structural components, encapsulation or confining boundaries, and contained fluids or solids.
b.
Secured Experiment - Any experiment, 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 to overccme the expected effects of hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experi-ment or which might arise as a result of credible malfunctions.
c.
Unsecured Experiment - Any experiment, or component of an experiment is deemed to be unsecured whenever it is not secured as defined in 1.4.b above. Moving parts of experiments are deemed to be unsecured when they are in motion.
d.
Movable Experiment - A movcble experiment is one which may be inserted, removed or manipulated while the reactor is cri tical.
1 e
1()S2. 254
t e
e.
Removable Experiment - A renovable experiment is any experi-ment, experimental facility, or component of an experiment, other than a permanently attached appurtenance to the reactor system, which can reasonably be anticipated to be moved one or more times during the life of the reactor.
1.5 Experimental Facilities - Experimental facilities are those portions of the reactor assembly that are used for the introduction of experiments into or adjacent to the reactor core region or allow beams of radiation to exit from the reactor shielding.
Experimental facilities shall include the thermal column, glory hole, and access ports.
1.6 Explosive Material - Explosive material is any solid or liquid which is categorized as a Severe, Dangerous, or Very Dangercus 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 publication 704-M,1966, "Identiff-cation System for Fire Hazards of Materials," also enumerated in the " Handbook for Laboratory Safety," 2nd Ed. (1971) published by the Chemical Rubber Company.
1.7 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.
1.8 Operable - Operable means a component or system is capable of performing its intended function in its normal manner.
1.9 Coerating - Operating means a component or system is performing its intended function in its normal manner.
1.10 Potential Reactivity Worth - The potential react tity worth of an experiment is the maximum absolute value,f the reactivity change that would occur as a result of intended or anticipated changes or credtble malfunctions that alter experiment cosition or configuration.
Evaluations of potential reactivity worth of experiments also shall include effects of possible trajectories of the experiment in motion relative to the rtactor, its orientation along each trajectory, and circumstances which can cause internal changes such as creating or filling of void scacts or motion of mechanical components. For removable experiments, the potential reactivity worth is equal to or greater than the static reactivity worth.
1.11 Reactor Component - A reactor component is any apparatus, device, or material tnat is a normal part of the reactor assembly, 2
1052 255
1.12 Reactor Coeration - Reactor operation is any condition wherein the reactor is not shutdown.
1.13 Reactor Safety System - The reactor safety system is that combi-nation of safety channels and associated circuitry which forms an automatic protective system for the reactor or provides information which requires manual protective actic, be initiated.
1.14 Res: tor Shutdown - The reactor shall be considered shutdown whenever:
a.
either:
1.
all safety and control rods are fully withdrawn frem the core, or 2.
the core fuse melts resulting in separation of the core, and:
b.
The reactor console key switch is in the "off" position and the key is renoyed from the console and under the control of a licensed operator.
1.15 Safety Channel - A safety channel is a measuring channel in the reactor safety system.
1.16 Static Reactivity Worth - The static reactivity worth of an experiment is the value of the reactivity change which is measurable by calibrated control or regulating rod comparison methods between two defined terminal positions or configurations of the experiment. For removable experiments, the terminal positions are fully removed from the reactor and fully inserted or installed in the normal functioning or intended position.
3 1052 256
2.0 SAFETY LIMITS AND IIMITED 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 all fission products are retained in the core matrix.
Specification a.
The reactor power level shall not exceed 100 watts, b.
The maxi'um core temperature shall not exceed 2000C during either steady state or transient operation.
Basis The polyethylene core material does not melt below 2000C and is expected to maintain its integrity and retain essentially all of the fission products at temperatures below 2000C. The Hazards Stanary Report dated February 1962 submitted on Docket F-15 by Aerofet-General Nucleonics (AGN) calculated a steady state core average temperature rise of 0.44C/ watt. Therefore, a steady state power level of 100 watts would result in an average core temperature 0
rise of 44 C.
The corresponding maximum core temperature would be below 2000C thus assuring integrity of the core and retention of fission products.
2.2 Limiting Safety System Settings Apolicability This specification applies to the parts of the react 07 safety system which will limit maximum power and core temperature.
Objective To assure that autcmatic protective action is initiated to prevent a safety limit fecm being exceeded.
Scecification a.
The safety channels shall initiate a reactor scram at the following limiting safety system settings:
4 1052 257
Channel Condition LSSS Nuclear Safety #2 High Power
< 10 watts Nuclear Safety #3 High Power
][10 watts b.
The core thermal fuse shall melt when heated to a temperature of about 1200C resulting in core separation and a reactivity loss greater than 5% k.
Basis Based on instrumentation response times and scram tests, the AGN Hazards Report concluded that reactor periods in excess of 30-50 milliseconds would be adequately arrested by the scram system.
Since the maximum available excess reactivity in the reactor is less than one dollar the reactor cannot become prompt critical and the corresponding shortest possible period is greater than 200 milliseconds. The high power LSSS of 10 watts in conjunction with automatic safety systems and/or manual scram capabilities will assure that the safety limits will not be exceeded during steady state or as a result of the most severe credible transient.
In the event of failure of the reactor to scram, the sel '-limiting characteristics due to the high negative temperature coefficient, and the melting of the thermal fuse at a temperature below 1200C will assure safe shutdown without exceeding a core temperature of 2000C.
5 1052. 2)58
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.
Obj ective To assure that the reactor can be shut down at all times and that the safety limits will not be exceeded.
Specification a.
The available excess reactivity with all co.? trol and safety rods fully inserted and including the potential reactivity worth of all experiments shall not exceed 0.65% a k/k referenced to 200C.
b.
The shutdown margin with the most reactive safety or control rod fully inserted shall be at least 1% a k/k.
c.
The reactivity worth of the control and safety rods shall ensure sub-criticality on the withdrawal of the coarse control rod or any one safety rod.
Basis The limitations on total core excess reactivity assure reactor periods of sufficient length so that the reactor protection system and/or operator a' tion will be able to shut the reactor down without exceeding any safety limits. The shutdown margin and control and safety rod rea;tivity limitations assure that the reactor can be brought and maintained suberitical if the highest reactivity rod fails to scram c.1d remains in its most reactive position.
3.2 Control and Saf_ety 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.
1052 259 6
s.
TABLE 2.1 Safety Channel Set Point' Function Nuclear Safety #1 Low power 5% Full scale scram at source levels < 5% of full scale Nuclear Safety 42 High power 10 watt scram at power
>10 watt Low power 3.0 x 10-13 amos scram at source levels
<3 x 10-13 amos Reactor ceriod 5 sec scram at periods
<5 sec Nuclear Safety #3 (Linear Power)
High power 10 watt scram at power >10 watt Lcw power 5% full scale scram at source levels < 5% of full scale Manual scram scram at operator option Radiation Monitor alarm at or below level set to meet requirements of 10 CFR Part 20 7
1052 260
e Specification a.
The total scram withdrawal time-of the safety rods and coarse control rod shall be less than 200 milliseconds.
b.
The average reactivity addition rate for each control or safety rod shall not exceed 0.065" ak/k per second.
c.
The safety rods and coarse control rod shall be interlocked such that:
1.
Reactor startup cannot commence unless both safety rods and coarse control rod are fully withdrawn from the core.
2.
Only one safety rod can be inserted at a time.
3.
The coarse control rod cannot be inserted unlesi both safety rods are fully inserted.
d.
All reactor safety system instrumentation shall be operable in accordance with Table 3.1 with the exception that Safety Channels 1 or 3 may be bypassed whenever the reactor control or safety rods are not in their fully withdrawn position.
e.
The shield water level interlock shall be set to prevent startup and scram the reactor if the shield water level falls 10 inches below the highest poi.it on the reactor shield tank manhole opening.
f.
The shield water temperature interlock shall be set to prevent react:ir startup and scram the reactor if the shield water tempe ature falls below 150C.
g.
The s eismic displacement interlook sensor shall be installed in st ch a manner to prevent reactor startup and scram the reac1or during a seisnic displacement.
h.
.~. loss of electric power shall cause the reactot to scram.
Basis The specifications on scram withdrawal time in conjunction with the safety system instrumentation and set points assure safe reactor shutdown during the most severe foreseeable transients.
Interlocks on control and safety rods assure an orderly approach to criticality and an adequate shutdcwn capability. The limitations on reactivity addition rates allow only relatively slow increases of reactivity so that ample time will be available for manual or autanatic scram during any operating conditions.
8 1052 26i
The neutron detector channels (nuclear safety channels 1 through 3) assure that reactor power levels are adequately monitored during reactor startup and operation.
Requirements on minimum neutron levels will prevent reactor startup unless channels are operable and responding, and will cause a scram in the event of instrumentation failure. The power level scrams initiate redundant automatic protec-tive action at power level scrams 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-201's negative temperature coefficient of reactivity causes a reactivity increase with decreasing core temperature.
The shield water temperature interlock will prevent reactor operatior, at tem eratures below 150C thereby limiting potential reactivity additions associated with temcerature decreases.
Tater in the shield tank is an im=ortant component of the reactor shield and ooeration 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.
The reactor is designed to withstand 0.6g accelerations and 6 cm displacements.
A seismic instrument causes a reactor scram whenever the instrument receives a horizontal acceleration that causes a horizontal displacement of 1/16 inch or greater.
The seismic displacement interlock assures that the reactor will be scrammed and brought to a suberitical configuration during any seismic disturbance that may cause damage to the reactor or its components.
The manual scram allcws 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 control rod holding magnets causing a reactor scram and thus assuring safe and i: mediate shutdown in case of a power outage.
A' radiation monitor must alway's be available to operating cersonnel to provide an indication of any abnormally high radiation levels so that appropriate action can be taken to shut the reactor down and assess the hazards to cersonnel.
9 1052 262
4 3.3 Limitations on Experiments Applica b_il ity This specification applies to experiments installed in the reactor and its experimsntal facilities.
Objective To prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.
Specification a.
Experiments containing materials corrcsive to reactor components or which contain liquid or gaseous, fissionable materials shall be doubly en apsulated.
b.
Explosive materials shall not be inserted into experimental facilities of the reactor or stored within the confines of the reactor facility.
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 (1) unrestricted areas continuously for two hours starting at time of release or (2) restricted areas during the length of time required to evacuate the restricted area.
d.
The radioactive material content, including fission products of any doubly encapsulated experiment shall be limited so that the complete release of all gaseous, particulate, or volatile components of the experiment shall not result in exposures in excess of 0.5 Rem whole body or 1.5 Rem thyroid to persons occupying an unrestricted area continuously for a period of two hours starting at the time of release, or exposure in excess of 5 Rem whole body or 30 Rem thyroid to persons occupying a restricted area during the length of time required to evacuate the restricted area.
Basis These specifications are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting from an experiment failure and to protect operating personnel and the public from excessive radiation doses in the event of in experiment failure.
10 u;-
3.4 Radiation Monitorino, Control, and Shielding Appl icability This specification applies to radiation monitoring, control, and reactor shielding required during reactor operation.
Objective _
To protect facility personnel and the public from radiation exposure.
Soecification a.
An operable portable and installed radiation survey instrument capable of detecting gamma radiation shall be immediately available to reactor operating personnel whenever the reactor is not shutdown.
b.
The reactor room shall be considered a restricted area whenever the reactor is not shutdown.
c.
The following shielding requs.ements shall be fulfilled during reactor operation:
1 The reactor shield tank shall be filled with water to a height within 10 inches of the highest point on the manhole opening.
2.
The thermal column shall be filled with water or graphite except during a critical experimen'. (core loading) or during measurement of reactivity worth of thermal column water or graphite.
Basis Radiation surveys performed under the supervision' of a qualified health physicist have shown that the total garma, thermal neutron, and fast neutron radiation dose rate in the reactor room, at the closest approach to the reactor, is less than 100 mrem /hr at reactor power levels less than 1.0 watt, and that the total gamma, thermal neutron, and fast neutron dose rate in the accelerator room is less than 15 mrem /hr at reactor power levels less than or equal to 5.0 watts and the thermal column filled with water.
The facility shielding in conjunction with designated restricted radiation areas is designed to limit radiation doses to facility personnel and to the public to a level below 10 CFR 20 limits under operating condtions, and to a level below Criterion 19, Appendix A, 10 CFR 50 recommendations under accident conditions.
1052 264
d 4.0 SURVEILLANCE REOUIREMENTS Actions specified in t'is section are not required to be performed n
if during the specified surveillance period the reactor has not been brought critical or is maintained in a shutdcwn 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 Acplicability This specification applies to the surveillance requirements for reactivity limits.
Obj ective To assure that reactivity limits for Specification 3.1 are not exceeded.
Specification a.
Safety and control rod reactivity worths shall be measured annually, but at intervals not to exceed 16 months.
b.
Total excess reactivity and shutdcwn margin shall be determined annually, but at intervals not to e.
3d 16 months.
c.
The reactivity worth of an experiment shal'l be estimated or measured, as appropriate, before or during the first startup subsequent to the experiment's insertion.
Basis The control and safety rods are ir.spected and their reactivity worths measured annually to assure that no degradation or unexoected changes have occurred which could adversely affect reactor shutdown margin or total excess reactivity.
The shutdown margir. and total excess reactivity are determined to assure that the reactor can always be safely shutdewn 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 action. Based on experience with AGN reactors, significant changes in reactivity or rod worth are not expected within a 16-month period.
12 s
4.2 Control and Safety System Applicability This specification applies to the surveillance requirements of the reactor control and safety system.
Objective To assure that the reactor control and safety systems are operable as required by Specification 3.2.
Soecification
- ~
a.
Safety and control rod scram times and average reactivity insertion rates shall be measured annually, but at intervals not to exceed 16 months.
b.
Safety and control rods and drive shall be inspected for deterioration at intervals not to exceed 2 years.
c.
A channel test of the following safety channels shall be perfomed prior to the first reactor startup of the day or prior to each operation extending more than one day.
Nuclear Safety #1, #2, and #3 Manual scram d.
A channel test of the seismic displacement interlock shall be performed semiannually.
e.
A channel check of the following safety ci:annels shall be performed daily whenever the reactor is in operation:
Nuclear Safety #1, #2, and #3 f.
Prior to each day's operation or prior to each operation extending more than one day, safety rods #1, and #2 shall be inserted and scrarm:ed to verify operability.
g.
The period, count rate, and power level measuring channels shall be calibrated and set points verified annually, but at int 2rvals not to exceed 16 tronths.
h.
The shield tank water level interlock, shield water temperature interlock, and seismic displacement safety channel shall be calibrated by perturbing the sensing element to the appropriate set point.
These alibrations shall be perfomed annually, but at intervals not to exceed 16 months.
- i. The rad'ation monitoring instrumentation shall be calibrated annually, but at intervals not to exceed 16 months.
1052 266
Basis The channel tests and checks required daily or before each startup will assure that the safety channels and scram functions are operable. Based on operating experience with reactors of this type, the annual 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.
4.3 Reactor Structure Acolicability This specification applies to surveillance requirements for reactor components other than control and safety rods.
Objective To assure integrity of the reactor structures.
Soecification a.
The shield tank shall be visually inspected every two years.
If apparent excessive corrosion or other damage is observed, cor ;ctive measures shall be taken prior to subsequent reactor operation.
b.
Visual inspection for water leakage from the shield tank shall ba performed annually. Leakage shall be corrected prior to subsequent reactor operation.
Bas is Based on experience with reactors of this type, the frequency of inspection and leak test requirements of the shield tank will assure capability for radiation protection during reactor operation.
4.4 Radiation Mo_nitoring and Control _
,Applicabil ity This specific;tior. applies te the surveillance requirements of the
. radiation monitoring and control systems.
Objective To assure that the radiation monitoring and control systems are operable and that all radiation areas within the reactor facility are identified and controlled as required by Specification 3.4.
14 1052 267
O Specification a.
All portable and installed radiation survey instruments assigned to the reactor facility shall be calibrated under the supervision of the Radiation Safety Officer annually, but at intervals not to exceed 16 months.
b.
Prior to each day's reactor operation or prior to each reactor operation extending more than one day, the reactor room high radiation alarm shall be verified to be operable.
c.
A radiation survey of the reactor room and reactor control room shall be performed under the supervision of the Radiation Safety Officer annually, but at intervals not to exceed 16 months, to determine the location of radiation and high radiation areas corresponding to reactor operating power levels.
Basis The periodic calibration of radiation monitoring equipment and the surveillance of the reactor room high radiation area alarm will assure that the radiation monitoring and control systems are operable during reactor operation.
The periodic radiatic
- urveys will verify the location of radiation and high radiation ar as and will assist reactor facility personnel in properly labeling and controlling each location in accordance with 10 CFR 20.
1052 268 15
5.0 DESIGN FEATURES 5.1 Reactor a.
The reactor core, in_cluding cortrol and safety rods, contains approximately 670 grams of U-235 in the form of 20% enriched UO dispersed in approximately 11 kilograms of polyethylene.
2 The lower section of the core is supported by an aluminum rod hanging frcm a fuse link. The tuse melts at temperatures below 120*C ca"..ing tne lo er core section to fall away from the upper section redicing reactivity by at least 5%
ak/k. Sufficie.1t clearanca between core and reflector is provided to insure free fall of the bottcm half of the core during the most severe transient.
b.
The '.cre is surrounded by a 20 cm thick high density '(1.75 gm/cm3) graphite reflector followed by a 10 cm thick lead gamma shieTd.
The core and part of the graphite reflector are sealed in a fluid-tight aluminum core tank designed to contain any fissian gases that might leak frem the core.
c.
TF.e core, reflector, and lead shielding are enclosed in and rupported by a fluid-tight steel reactor tank. An upper or " thermal column tank" may serve as a shield tank when filled with water or a thermal column when filled with graphite.
d.
ihe 6-1/2 foot diameter, fluid-tight shield tank is filled with water constituting a 55 cm thick fast neutron shiel d.
The fast neutron shield is formed by filling the tank with approximately 1000 gallens of water. The cr plete reactor shield shall limit doses to operationg personnel in restricted and unrestricted areas to levels less than permitted by 10 CFR 20 unoc-operating conditions.
e.
Shielding is provided by a concrete wall constructed of 4" x 8" x 16" concrete blocks and 4" x 8" x 12" barytes concrete blocks for 5 watt operation. The blocks are held to close dimensicnal tolerance in manufacture and stacked in such a manner that voids in the comple ted wall are at a minimum.
Near the beam ports and glory hole, high density blocks are used between 20 inches and 112 inches above the base.
The use of these blocks further reduces radiation level in these areas. Overhead shielding is provided by 8 inch thick barytes blocks (minimum density 3.7 gcc).
As detailed in the amendment for 5 watt operation for Aerojet-General Nuclecnics, dated 11 February 1957, and on file with the Ccemissien in Docket 50-32, an 18 inch additional cencrete shield wall was sufficient to maintain sub-tolerance radiation levels external to the shield when operating at 5 watts.
Subsequent analysis by Aerojet-General Nuclecnics indicated that 16 inches of ordinary concrete snielding was sufficient. Twelve inches of barytes concrete is more effective than 18 inches of ordinary concrete.
16 1052 269
4 The radiation levels assgciated with 5 watt operation (peak thermal flux of 2.5 x 1@ n/cm2-sec) have been calculated by Aeroject-General Nucleanics. The table below gives Aerojet-General results for garma dose for an 18" concrete shield:
Eneroy (Mev) u ux _
B e-ux Dose (mrem /hr) 3.0 0.081 3.64 1.9 0.027 1.5 2.2 0.105 d.73 2.2 0.0088 1.0 Total = 2.5 mrem /hr (18" concrete)
Changing the e-ux for a 16" cencrete wall and using the same buildup factors yields the follcwing table:
Energy Energy (hev) u ux B
e-ux Dose (mrem /hr) 3.0 0.081 3.30 1.9 0.037 2.1 2.2 0.105 4.28 2.2 0.014 1.2 Total 3.3 mrem /hr s
(16" concrete)
The National Naval Medical Center reported neutron dose rates were less than 0.2 mrem /hr for 18" shield.
Although calculations and operating surveys of similar facilities show that at 5 watts power there will be no area on the reactor floor outside the concrete shield where the total radiatien exceeds tolerance levels; nevertheless, the reactor floor is a control area with restricted access, f.
Two safety rods and one control rod (identical in size) contain up to 20 grams of U-235 each in the same form as the core material.
These rods are lifted into the core by electromagnets, driven iy reversible DC motors'through lead screw assemblies. Deenergizing the magnets causes a spring-driven, grmvity-assisted scram. The fourth rod or fine control rod (approximately one-half the diameter of the other rods) is driven directly by a lead screw.
This rod may contain fraled or unfueled polyethylene.
1052 270 17
d 5.2 Fuel Storace Fuel, including fueled experiments and fuel desices not in the reactor, shall be stored in locked rooms in the nuclear engineering department laboratories.
The storage array shall be such that K.,,
is no greater than 0.8 for all conditions of m6deration and reflection.
5.3 Reactor Room a.
The reactor reem houses the reactor assembly and accessories-required for its operation and maintenance, b.
The reactor room is a secarate rocm in the Lillibridge Engineering Laboratory, constructed vich adequate shielding and other radiation protective features to limit doses in restricted and unrestricted areas to levels no greater than permitted by 10 CFR 20, under normal operatinc conditions, and to a level below Criterien 19, A pendix A, 10 CFR 50 recommendations under accident conditions.
c.
Access doors to the reactor reem are self-locking.
l052 271 is
6.0 ADMINISTRATIVE CONTROLS 6.1 Organization The administrative organization for control of the reactor facility and its operation shall be as set forth in Figure 1 of these specifications. The authorities and responsibilities set forth below are designed to comply with the intent and requirements for admin-istrative controls of the reactor facility as set forth By the Nuclear Regulatory Comission.
6.1.1 University Officer The University Officer is an administrative officer responsible for the University and in whose name the application for licensing is made.
6.1. 2 Dean, School of Engineerino The Dean of the School of Engineering is the administrative officer responsible for the operation of the School of Engineering (also Reactor Administrator).
6.l.3 Reactor Administrator The Reactor Administrator is the administrative officer responsible for the operattn of the AGN-20lM Reactor Facility.
In this capacity he shall have final authority and ultimate responsbility for the operation, maintenance, and safety of the reactor facility within the limitations set forth in the facility license. He shall be responsible for appoint'ng personnel to all positions reporting to him as described in Section 6.1 of the Technical Specifications. He shall seek the advice and approval of the Radiation Safety Committee and/or the Reactor Safety Committee in all matters concerning unresolved safety questions, new experiments and new procecures, and facility modifications which might affect safety. He shall be an ex, officio member of the Reactor Safety Committee.
6.1. 4 Reactor Sucereiser The Reactor Supervisor shall be responsible for the preparation, promulgation, and enforcement of administrative controls including all rules, regulations, instructions, and operating procedures to ensure that the reactor facility is operated in a safe, competent, and authorized manner at all times. He shall direct the activities of operators and technicians in the daily operation and maintenance of the reactor; schedule reactor operations and maintenance; be responsible for the preparation, authentication, and storage of all prescribed logs and operating records; authorize all experiments, procedures, and changes thereto which have received the approval of the Reactor Safety Committee and/or the Radiation Safety Committee and tne Reactor Administrator; and be responsible for the pre-paration of experimental procedures involving use of the reactor.
1052 272 19
FIGURE 1 Administrative Organization of the Idaho State University AGN-20lM Reactor Facility NRC License R-110 Idaho State University Officer Radiation Safety Reactor Safety Committee Committee Reactor Administrator
- Reactor Supervisor *
~
Reactor Cperators**
- Requires NRC Senior Operators License
- Requires NRC Operators License except where exempt per
- = - = -
10 CFR 55 paragraph 55.9 2
1052 273
6.1.5 Reactor Ooerators Reactor Operators shall be responsible for the manipulation of the reactor controls, monitoring of instrumentation, operation of reactor related equipment, and maintenance of complete and current records during operation of the facility. Reactor Operators who are exempt from holding an NRC license per 10 CFR 55 paragraph 55.9 shall only operate the reactor under the direct and immediate supervision of a licensed Reactor Operator.
6.1.6 Reactor Safety Committee The Reactor Safety Committee shall be responsible for, buc not limited to, reviewing and approving safety standards asso-ciated with the use of the reactor facility; reviewing and approving all proposed experiments and procedures and changes thereto; reviewing and approving all modifications to the reactor facility which might affect its safe operation; determining whether proposed experiments, procedures, or modifications involve unreviewed safety questions, as defined in 10 CFR 50 paragraph 50.59(c), and are in accordance with these Technical Specifications; conducting periodic audits of procedures, reactor operations and maintenance, equipment performance, and records; review all reportable occurrences and violations of these Technical Specifications, evaluating the causes of such events and the corrective action taken and recommending measures to prevent reccurence; reporting all their findings and reccmmendations concerning the reactor facility to the Reactor Administrator.
6.1. 7 Radiation Safety Committee The Radiation Safety Ccmmittee shall advise the University administration and the Radiation Safety Officer on all matters concerning radiological safety at University facilities.
6.1.8 Radiation Safety Officer The Radiation Safety Officer shall review and approve all pro-cedures arid experiments involving radiological safety. He shall enforce all federal, state, and university rules, regulations, and procedures relating to radiological safety.
He shall perform routine radiation surveys of the reactor facility and report his findings to the Reactor Administrator.
He shall provide personnel dosimetry and keep records of personnel radiation exposure. He shall advise the Reactor Administrator on all matters concerning radiological safety at the reactor facil ity. The Radiation Safety Officer shall be an ex officio member of the Reactor Safety Committee.
6.1.9 Ocerating Staff a.
The minimum operating staff during any time in which the-reactor is not shutdcwn shall consist of:
1052 274 2'
1.
One licensed Reactor Operator in the reactor control room.
2.
One other person in the reactor room or reactor centrol room certified by the Reactor Supervisor as qualified to activate manual scram and initiate emergency pro-cedures.
3.
One licensed Senior Reactor Operator readily available on call. This requirement can be satisfied by having a licensed Senior Reactor Operator perform the duties stated in paragraph 1 or 2 above or by designating a licensed Senior Reactor Operator who can be readily contacted by telephone and who can arrive at the reactor facility within 30 minutes, b.
A licensed Senior Reactor Operator shall supervise all reactor mainter.ance or modification which could affect the reactivity of the reactor.
6.2 Staff Qualifications The Reactor Administ,rator, the Reactor Supervisor, licensed Reactor Operators, and technicians performing reactor maintenance shall meet the minimum qualifications set forth in ANS 15.4,
" Standards for Selection and Training of Personnel for Research Reactors. " Reactor Safety Committee members shall have a minimum of five (5) years experience in their profession or a baccalaureate degree and two (2) years of professional experience. Generally, these cannittee members will be made up of University faculty, but outside experience may be sought in areas where additional experience is considered necessary by the Reactor Administrator.
6.3 Training The Reactor Administrator shall be responsible for directing training as set forth in ANS 15.4,
- Standards for Selection and Training of Personnel for Research Reactcrs." All licensed reactor operators shall participate in requalification training as set forth in 10 CFR 55.
6.4 Reactor Safety Committee 6.4.1 Meetinas and Ouorum The Reactor Safety Committee shall meet as often as deemed necessary by the Reactor Safety Ceanittee Chairman but shall meet at least once each calendar year. A quorum for the conduct of official business shall be the chairman, or his designated alternate, and two (2) other regular members. At no time shall the operating organization comprise a voting majority of the members at any Reactor Safety Committee meeting.
1052 275 22
6.4.2 Reviews The Reactor Safety Committee shall review:
a.
Safety evaluations for changes to procedures, equipment or systems, and tests or experiments, conducted without Nuclear Regulatory Commission approval under the provision of 10 CFR 50 paragraph 50.59, to verify that such actions do not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems that change the original intent or use, and are non-conservative, or those that involve an unreviewed safety question as defined in 10 CFR 50 paragraph 50.59.
c.
Proposed tests or experiments which are significantly different from previously approved tests or experiments, or those that involve an unreviewed safety question as defined in Section 50.59, 10 CFR 50 paragraph 50.59.
d.
Proposed changes in Technical Specifications or licenses.
e.
Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal p tcedures or instructions having nuclear rafety significance, f.
Significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect nuclear safety, g.
Reportable occurrences.
h.
Audit reports.
6.4.3 Audits Audits of facility activities shall be performed under the cognizance of the Reactor Safety Committee but in no case by the perscnnel responsible for the item audited. These audits shall examine the operating records and encompass but shall not be limited to the folicwing:
a.
The conformance of facility operation to the Technical Specifications and applicable license conditions, at least once per 12 months.
b.
The performance, training, and qualifications of the entire facility staff, at least once per 24 mcnths.
c.
The results of all actions taken to correce deficiencies occurring in facility ecuipment, structures, systems or method of operation that affect nuclear safety, at least once per calendar year.
1052 276 23
d.
The Facility Emergency Plan and implementing procedures at least once per 24 months.
e.
The Facility Security Plan and implementing precedures, at leas 3 once per 24 months.
6.4.4 Authori ty The Reactor Safety Cannittee shall report to the University officer and shall advise the Reactor Administrator on those areas of responsibility outlined in Section 6.1.6 of these Technical Specifications.
6.4.5 Minutes of the Reactor Safety Committee The Chairman of the Reactor Safety Comnittee shall direct the preparation, maintenance, and distribution of minutes of its activities. These minutes shall include a summary of all meetings, actions taken, audits, and reviews.
6.5 Acorovals The procedure for obtaining approval for any change, modification or procedure which requires approval of the Reactor Safety Connittee shall be as follows:
a.
The Reactor Supervisor shall prepai s the proposal for review and approval by the Dean of tae School of Engineering.
b.
The Dean of the School of Engineering shall submit the proposal to the Chairman of the Reactor Safety Committee.
c.
The Chairman of the Reactor Safety Committee shall submit the proposal to the Reactor Safety Committee members for review and comment.
d.
The Reactor Safety Committee can approve the proposal by majority vote.
6.6 Procedures There shall be w itten procedures that cover the following activities:
a.
Startup, operation, and shutdcwn of the reactor.
b.
Fuel movement and changes to the ccre and exceriments that could affect reactivity, c.
Conduct of irradiations and experiments that could affect the safety of the reactor.
1052 277 d.
Preventive or corrective maintenance which could affect the safety of the reactor.
e.
Surveillance, testing, and calibration of instruments, comoonents and systems as specified in Section 4.0 of these Technical Specifications.
2a
e f.
Implementation of the Security Plan and Emergency Plan.
The above listed procedures shall be approved by the Dean of the School of Engineering and the Reactor Safety Committee.
Temporary procedures which do not change the intent of previously approved procedures and which do not involve any unreviewed safety question may be employed on approval by the Reactor Supervisor or Dean of the School of Engineering.
6.7 Exceriments a.
Prior to initiating any new reactor experiment an experi-mental procedure shall be prepared by the Reactor Supervisor and reviewed and approved by the Dean of the School of Engineering and the Reactor Safety Committee.
b.
Approved experiments shall only be performed under the cognizance of the Dean of the School of Engineering and the Reactor Supervisor.
6.8 Safety Limit Violation The following actions shall be taken in the event a Safety Limit is violated:
a.
The reactor will be shutdown immediately and reactor operation wi?1 not be resumed without authorization by the Nuclear Regulatory Commission (NRCl.
b.
The Safety Limit violation shall be reported to the appropriate NRC Regional Office of Inspe: tion and Enfortenent, the Director of the NRR, and the Reactor Safety Committee not later than the next working day.
c.
A Safety Limit Violation Report shall be prepared for review by the Reactor Safety Committee.
This rercrt shall describe the applicable circumstances preceding the violation, the effects of the violation upon facility components, systens 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 14 days of the violation.
- 6. 9 Reportino Recuirements In additien to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the e.ppropriate NRC Regional Office.
6.9.1 Annual Oceratino Recort Routine operating reports covering the operation of the unit during the previous calendar year should be submitted prior to June 10 of each year.
25 1052 278
The annual operating reports made by licensees shall provide a comprehensive summary of the operating experience having safety significance that was gained during the year, even though some repetition of previously reported information may be involved. References in the annual operating report to previously submitted reports shall be clear.
Each annual operating report shall include:
(1) A brief narrative summary of a.
Char ues in facility design, performance characteristics, and operating pro'.edures relating to reactor safety that occurred during the reporting period.
b.
Results of major surveillance tests and inspections.
(2) A monthly tabulation showing the hours the reactor is operating.
(3) List of the unscheduled shutdowns, including the reasons therefore and corrective action uken, if any.
(4) Discussion of the major safety related corrective maintenance performed during the period, including the effects, if any, on the safe operation of the reactor and the reasons for the corrective maintenance required.
(5) A brief description of:
a.
Each change to the facility to the extent that it changes a description of the facility in the application for license and amendments thereto.
b.
Changes to the procedures as described in Facility Technical Specifications.
c.
Any new or untried experiments or tests cerformed during the reporting period.
(6) A summary of the safety evaluation made for each change, test, or experiment not submitted for NRC approval pursuant to 10 CFR 50.59 which clearly shows the reason leading to the conclusion that no unreviewed safety question existed and that no change to the Technical Specifications was required.
(7) A surnary of the nature and amour.t of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as determined at or prior to the point of such release or discharge.
a.
Liquid waste -
Total estimated quantity of radioactivity released (in curies) and total volume (in liters) of effluent water (including diluent) released.
1052 279
b.
Airborne waste -
Total estimated grantity of radioactivity released (in curies) determined by an approved sampling and counting method.
c.
Solid waste -
(1)
Total amount of solid waste packaged (in cubic meters)
(ii)
Total activity in solid waste (in curies)
(iii) The dates of shipments and disposition (if shipped off site)
(8) A descripticq of the results of any environmental radiological surveys performed outside the facility.
(9) Radiation Expcsure - A summary of radiation exposures greater than 100 mrem (50 mrea for persons under 18 years of age) received during the reporting period by facility personnel or visitors.
6.9.2 Esportable Occurrences Reportable occurrences, including causes, probable consequences, corrective actiens and maasures to prevent recurrence, s5all be reportad to the NRC.
a.
Prompt Notification with Written Follewup -
The types of events listed shall be reported as expe-ditiously as possible by telephone and telegraph to the Director of the appropriate NRC Regional Office, or his designated represent 1tive no later than the first working day following the event, with a written follewup 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 operatica subject to a limiting condition is less conservative than the limiting condition for operation established in the technical specifications.
(3) Abnormal degradation discovered in a fission product.
barrier.
1052 280 n
(4) Reactivity balance anomalies involving:
a.
disagreement between expected and actual critical positions of approximately 0.3".
ak/k; b.
exceeding excess reactivity limits; c.
shutdown margin less conservative than specified in technical specifications.
(5) Failure or malfunction of one (or more) component (s) which prevents, or could prevent, by itself, the ful-fillment of the functional requirements of systems used to cope with accidents analyzed in Safety Analysis Report.
(6) Personnel error or procedural inadequacy which prevents, or could prevent, by itself, the fulfillment of the func-tional requirements of systems required to cope with accidents analyzed in Safety Analysis Report.
(7) Errors discovered in the transient or accidentalalyses or in the metheds used for such analyses as described in the Safety Analysis Report or in the casts 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 compone'nts that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the Safety Analysis Report or Technical Specification basis, 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.
6.10 Record Retentien 6.10.1 Records to be retained for a period of at least five years:
a.
Operating logs or data which shall identify:
(1) Ccmpletion of pre-startup checkout, startup, power changes,.and shutdown of the reactor.
(2) Installation or removal of fuel elements, control rods or experiments that could affect core reactivity.
(3) Installation or removal of jumpers, special tags or notices, or other temocrary changes to reactor safety circuitry.
(.t) Rod worth measurements and othar reactivity measurements.
7C 1052 281
b.
Principal maintenance operations.
c.
Reportable occurrences.
d.
Surveillance activities required by technical specifications.
e.
Facility radiation and contamination surveys.
f.
Experiments performed with the reactor.
This requirements may be satisfied by the normal operations log book plus:
(1) Records of radioactive material transferred from the facility as required by license.
(2) Records required by the Reactor Safety Ccmmittee for the performance of new or special experiments.
g.
Changes to operating procedures.
6.10.2 Records to be retained for the life of the facility:
a.
Gaseous and liquid radioactive effluents released to the environs, b.
Appropriate off site environmental monitoring surveys.
c.
Fuel inventories and fuel transfers.
d.
Radiation exposures for all personnel.
e.' Updated as-built drawings of the facility.
f.
Records of transient or operational cycles for those components designed for a limited number of transients or cycles.
g.
Records of training and qualification for members of the facility staff.
h.
Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
i.
Records of meetings of the Reactor Safety Ccmnittee.
29 1052 282
_.