ML19259B260
| ML19259B260 | |
| Person / Time | |
|---|---|
| Site: | 05000072 |
| Issue date: | 12/01/1978 |
| From: | UTAH, UNIV. OF, SALT LAKE CITY, UT |
| To: | |
| Shared Package | |
| ML19259B259 | List: |
| References | |
| NUDOCS 7901180192 | |
| Download: ML19259B260 (25) | |
Text
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APPENDIX A LICENSE NO. R-25 TEONICAL SPECIFIG N_S mR LNIVEttSITY OF LTTAH AGN-201M REACIUR (SERIAL #107)
DOCET NO. 50-72 DATE: DECDBER 1,1978 AS 50DIFIED 1D INCWDE ANSI N378,1974 AND REGUIA70RY GUIDE 2.2 GJIDANCE i
0 7901180/92'
1 TABLE OF CONTEhTS PAGE 1.0 DEFINITIONS........................
1 2.0 SAFETY LIMITS AND LDfITING SAFETY SYSTBf SEITINGS.....
4 2.1 Safety T.imits....................
4 2.2 Limiting Safety System Settings 4
3.0 LDf1 TING CONDITIONS FOR OPERATION.............
5 3.1 Reactivity Limits Control and Safety Systems............
., 5 3.2
.6 3.3 Limitation on Experiments 3.4 Shielding 9
...................... 10 4.0 SURVEILLANCE REQUIRBfENIS.................
11 4.1 Reactivity Limits..................
11 4.2 Control and Safety System
.............. 12 5.0 DESIGN FEKIURES......................
13 5.1 Reactor 5.2 Fuel Storage.........
13 5.3 Reactor Room 14 14 6.0 ADfINISIRATIVE CONTROLS..................
14 6.1 Organization.....................
14 6.2 Staff Qualifications..
6.3 Training........
18 Reactor Safet 18 6.4 Procedures. y. Committee a..............
18 6.5 20 6.6 Safety Limit Violation........
20 6.7 Reporting Requirements............
21 6.8 Record Retention................
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1.0 DEFINITIONS The tems Safety Limit (SL), Limiting Safety System Setting (LSSS),
and Limiting Conditions for Operation (140) are as defined in 50.36 i
of 10 CFR part 50.
1.1 Reactor Shutdown - The reactor shall be considered sh down whenever 2
I 1.
either:
A.
All safety and control rods are fully withdrawn frmi j
j the core, or l
B.
The core fuse melts resulting in separation of the core, and:
2.
The reactor console key switch is in the "off" position and the key is removed fro:a the console and under the control of a licensed operator.
1.2 Reactor Operation - Reactor operation is any condition wherein the reactor is not shutdown.
I 1.3 Measunng 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 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 cmbination of safety channels and associated circuitry which foms an aute:atic protective system for the reactor or provides information which requires mmm1 protective action be initiated.
1.6 Reactor Cm:ponent - A reactor cmponent is any apparatus, device, or material that is a nomal part of the reactor asserrbly.
1.7 Operable - Operable means a component or system is capable of perfoming its intended function in its nomal manner.
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1.8 Operating - Operating means a co=ponent or system is perfoming its I
intenced function in its nomal mner.
1.9 Channel Check - A channel check is a qualitative verification of acceptable perfomance by obseIvation of channel behavior. 'Ihis verification may include comparison of the channel with other independent channels or methods measuring the same variable.
l.10 Channel Test - A channel test is the introduction of a signal into the channel to verify that it is operable.
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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 equipment, actuation, alam, or trip.
1.12 Experiment - An experiment is any of the following:
a.
An activity utilizing the re ctor system or its components or the neutrons or radiation generated therein; b.
An evaluatica or test of a reactor system operational, surveillance, or maintenance technique; c.
The material content of any of the foregoing, including structural ccs::ponents, encapsulation or confining boundaries, and contained fluids or solids.
1.13 Secured Experiment - Any experiment, or cc:::ponent 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 overcome the expected effects of hydraulic, pneumatic, bouyant.
or other forces which are nomal to the operating environment of the experiment or which might arise as a result of credible mal-functions.
1.14 Unsecured Experiment - Any experiment, 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 moticn.
1.15 Movable Experiment - A movable experiment is one which may be inserted, removed, or manipulated 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 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.
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1.17 Experimental Facilities - Experimental facilities are those portions of the reactor assembly that are used for the introduction of experi-ments into or adjacent to the reactor core region er allow beams of radiation to exist from the reactor shielding. Experimental facilities shall include the themal column, glory hole, and access ports.
1.18 Potential Reactivity Worth - The potential reactivity worth of an experament is the mmun absolute value of the reactivity change that would occur as a re;sult of intended or anticipated changes or credible malfunctions that alter experiment position m-configuration.
The evaluation must consider possible trajectories of the experiment in motion relative to the reactor, its orientation along esdt 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.
1.19 Static Reactivity Worth - The static reactivity worth of an experi-ment is the value of the reactivity change which is measurable by calibrated control or regulating rod comparison meth~1s between two defined teminal positions or configurations of the experiment.
For removable experiments, the teminal positions are fully re:oved from the reactor and fully inserted or installed in the normal functioning or intended position.
1.20 Explosive Material - Explosive material is arr/ 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) inder of 2, 3, or 4 by the National Fire Protection Association in its publication 704-M, 1966, " Identification System for Fire Hazards of Materials," also enumerated in the
" Handbook for Iaboratory Safety" 2nd Ed. (1971) published by The Chemical Rubber Co.
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2.0 SAFEIY LIMITS AND LIMITED SAFEIY SYSIBf SEITINGS 2.1 Safety Limits Applicability This specification applies to the maxima steady state power level and maxima core temperature during steady state or transient op-eration.
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.
I b.
The maximrm core temperature shall not exceed 200*C during either steady state or transient operation.
l Bases Ihe polyethylene core material does not melt below 200*C and is ex-pected to maintain its integrity and retain essentially all of the fission products temperatures below 200*C. The Ha::ards Sumary Report dated August 1956 submitted on Docket F-15 by Aerojet-C. 'eral Nucleonics (AGN) calculated a steady state core average temperature rise of 0.5*C/ watt. Therefore, a steady state power level of 100 watts would result in an average core te::perature rise of 50*C.
The corresponding maxmm core temperature would be below 200*C thus assuring integrity of the core and retention of fission products.
2.2 Limiting Safety System Settings Applicability This specification applies to the parts of the reactor safety system which will limit maximum pcwer and core temperature.
Objective To assure that automatic protective action is initiated to prevent a safety limit from being exceeded.
Specification a.
The safety channels shall initiate a reactor scram at the fol-lowing limiting safety system settint,s:
Channel Condition LSSS Nuclear Safety #2 High Power 1
1 0 watts Nuclear Safety #3 High Power 110 watts i
b.
The core thermal fuse shall melt when heated to a temperature of 120*C or less resulting in core separation and a reactivity loss greater than 5%Ak.
Bases Based on instrumentation response times and scram tests, the AGN Hazards Report concluded that reactor periods in excess of 30-50 milli-seconds would be adequately arrested by the scram system.
Since the maximm available excess reactivity in the reactor is less than one dollar the reactor cannot becme prompt critical and the corresponding shortest possible period is greater than 200 milli-seconds. The high power LSSS of 10 watts in ccnjunction with auto-matic safety systems and/or manual scram capabilities will assure that the safety limits will not be exceeded during steady state or as a re-l sult of the most severe credible transient.
In the event of failure of the reactor to scram, the self-limiting characteristics due to the high negative temperature coefficient, and the melting of the thermal fuse at a temperature below 120'C will assure safe shutdown without exceeding a core temperature of 200*C.
3.0 LIMITING QNDITIONS 10R OPERATION 3.1 Reactivity Limits Applicability e
Tras 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 and that the safety limits will not be exceeded.
Specification i
a.
The available excess reactivity with all control and safety rods fully inserted and including the potential reactivity worth of all experiments shall not exceed 0.65% ak/k referenced to 20*C.
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b.
The shutdown =argin with the most reactive safety or control rod fully inserted shall be at least 1% a k/k.
The reactivity worth of the control and safety rods shall ensure c.
sub-criticality on the withdrawal of the coarse control rod or any one safety rod.
Bases The limitations on total core excess reactivity assure reactor per-iods of sufficient length so that the reactor protection system and/or operator action will be able to shut the reactor down with-out exceeding any safety limits. The shutdown margin and control and safety rod reactivity limitations assure that the reactor can i
be brought and maintained subcritical if the highest reactivity rod fails to scram and remains in its most reactive positions.
3.2 Control and Safety Systens Applicability These specifications apply to the reactor control and safety systems.
Objective To specify lowest acceptable level of perfomance, instrument set points, and the minimum number of operable ccraponents for the reactor control and safety systems.
Specification i
The total scram withdrawal time of the safety rods and coarse a.
I control rod shall be less than 200 milliseconds.
I b.
The safety rods and coarse control rod shall be interlocked such that:
1.
Reactor startup cannot comence 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 unless both safety rods are fully inserted.
Nuclear safety channel instrumentation shall be operable in c.
accordance with Table 3.1 whenever the reactor control or safety rods are not in their fully withdrawn position. Ilowever Nuclear safetv channel ma" be bypassed for operation at power levels exceedin- 0.1 watt.
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TABLE 3.1 Safety Channel Set Point Function Nuclear Safety i1*
Law count rate 1 10 cps scram below 10 cps l
Nuclear Safety #2 i
High power
.1 10 watt scram at power >10 watt
-12 Iow power t 1.0 x 10 amps scram at source levels
< 1.0 x 10-12 amps Reactor period
> 5 sec scram at periods <5 sec Nuclear Safety #3 (Linear Power) i High Power
<.10 watt scram at power >10 watt I
Law power
> 5% full scale scram at source levels
< 5% of full scale Manual scram scram at operator option Nuclear Safety Channel #1 may be bypassed at power levels exceeding 0.1 watt. I
d.
The shield water level interlock shall be set to prevent reactor startup and scram the reactor if the shield water level falls 10.5 inches below the highest point on the reactor shield tank manhole opening.
e.
The shield water temperature interlock shall be set to prevent reactor startug and scram the reactor if the shield water temperature falls below 15 C.
f.
H e seismic displacement interlock sensor shall be installed in such a manner to prevent reactor startup and scram the reactor during a seismic displacement.
g.
A loss of electric power shall cause the reactor to scram.
Bases 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.
Intmlocks on control and safety rods assure an orderly approach to criticality and an adequate shutdown capability.
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 minin n 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 re<hmdet automatic protective action at power level scrams low enough to assure safe shutdown without exceeding any safety limits. He period scram ccnservatively 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.
i 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 operation at temperatures below 15'C thereby limiting potential reactivity additions i
associated with temperature decreases, i
j Water in the shield tank is an important component of the reactor shield and operation without the water may produce excessive radiation
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1evels. The shield tank water level interlock will prevent reactor operation without adequate water levels in the shield tank.
4 The reactor is designed to withstand 0.6g accelerations and 6 cm displacements. A seismic instnraent causes a reactor scram whenever the insttraent receives a horizontal acceleration that causes a horizontal disp 1nemt of 1/16 inch or greater. The seismic displacement interlock assures that the reactor will be so M and brought to a subcritical configuration during any seismic disturbance l
that may cause damage to the reactor or its components.
f The manual scram allows the operator to manually shut down the reactor l
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 thus assuring safe and imediate shutdown in case of a power outrage.
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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.
Specification Experiments containing materials corrosive to reactor ccrapenents a.
or which contain liquid or gaseous, fissionable materials shall be doubly encapsulated.
b.
Explosive materials shall not be inserted into experimental facilities of the reactor.
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 frcxn 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 j
length of time required to evacuate the restricted area.
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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.
Bases
'Ihese specifications are intended to reduce the 1.kelihood of damage to reactor ccc:panents and/or radioactivity relear,es resulting frcm an experimental failure and to protect operating personnel and the public frcra 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 frcn radiation exposure.
l Specification The following shielding requirements shall be fulfilled during reactor i
operation:
a.
The reactor shield tank shall be filled with water to a height within 10 inches of the highest point on the manhole opening.
[
b.
The ther::a1 colu::n shall be filled with water or graphite except during a c-itical experiment (core loading) or dur1ng measurement of reactivity war n of thennal colts:n water or g c.phite.
c.
The reactor rocxn shall be considered a restricted area.
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Bases 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 conditions, and to a level below criterion 19, Appendix A, 10 CFR 50 recernmdntions under accident conditions.
I 4.0 SURVEIIJANCE REQUIREMEBTIS Actions specified in this section are not required to be perfomed if during the specified surveillance period the reactor has not bsen brought critical or is maintained in a shutdown condition extending i
beyond the specified surveillance period. However, the surveillance requirements must be fulfilled prior to wbsequent startup of the reactor.
4.1 Reactivity Limits Applicability This specification applies to the surveillance requirements for reactivity limits.
Objective To assure that reactivity limits for Specification 3.1 are not exceeded.
Specification i
Safety and contol rod reactivity worths shall be measured annually, a.
but at intervals not to exceed 16 months.
b.
Total excess reactivity and shutdown margin chall be determined annually, but at intervals not to exceed 16 months.
c.
The reactivity worth of an experment shall be estimated or measured, as appwyriate, before or during the first startup I
subsequent to the experiment's insertion.
M Bases The control and safety rod reactivity worths measured annually 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 teminated by either operator or autmatic action.
Based on exaerience with AGN reactors, significant changes in reactivity or rod worti are not expected within a 16-month period.
4.2 Control and Safety System Applicability This specification applies to the surveillance requiremnts of the reactor control and safety systems.
Specification a.
Safety and control red scram times shall be measured annuntly, but at intervals not to exceed 16 months.
b.
Safety and control rods and drive shall be inspected for deterioration at intervals noc to exceed 2 years.
c.
A chnnnal test of the following safety channels shall be perfomed pnor to the first reactor startup of the day or prior to each operation extending more than ene day:
Nuclear Safety #1, #2, and #3 Manual scram d.
A channel test of the seismic displacement interlock shall be performed seminnnually.
e.
A channel check of the following safety channels shall be perfomed 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 rod #1 shall be inserted and scrammed to verify operability.
The period, count rate, and power level measuring channels shall g.
be calibrated and set points verified annually, but at intervals not to exceed 16 months, h.
The shield water level interlock and shield water temperature interlock shall be calibrated by perturbing tJ1e sensing element to the appropriate set point. These calibrations shall be perfomed annually, but at intervals not to exceed 16 months.
I Bases 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 e:oerience with reactors of this type, the annual scram measurements, ciannel 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 FEKIURES 5.1 Reactor The reactor core, incinAing control and safety rods, contains a.
appeuumately 660 grams of U-235 in the fom of 20% enriched 002 The dispersed in approximately 11 kilograms of polyethylene.
lower section of the core is supported by an aluminum rod hanging frcm a fuse link. The fuse melts at temperatures below 120'C causing the lower core section to fall away fram the upper section reducing reactivity by at least 5% ak/k. Sufficient clearance between core and reflector is provided to insure free fall of the bottom half of the core during the most severe transient.
The core is surrounded by a 20cm thick high density (1.75 gm/cm3) b.
graphite reflector followed by a 10 cm thick lead gama shield.
The core and part of the graphite reflector are sealed in a fluid-tight aluainum core tank designed to contain any fission gases i
that might leak from the core.
I The core, reflector, and lead shielding are enclosed in and supported c.
l by a fluid-tight steel reactor tank. An upper or " thermal column l
tank" may serve as a shield tank when filled with water or a themal column when filled with graphite.
d.
The 6.5 foot diameter, fluid-ugnt shield tank is filled with water constituting a 55 cm thick fast neutron shield. The fast neutron shield is formed by filling the tank with approxmately 1000 The ccr:plete reactor shield shall limit doses gallons of water.
to personnel in unrestricted areas to levels less than permitted by 10 CFR 20 under operating conditions.
Two safety rods and one control rod (identical in size) contain e.
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 by reversible DC motors through lead screw assemblics. Decnergizing The the magnets causes a spring-driven, gravity-assisted scram.
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 fueled or unfueled polyethylene.
5.2 Fuel Storage _
Fuel, including fueled experimei ts and fuel devices not in the reactor, shall be stored in locked rooms in the nuclear engineering department laboratories. The storage array shall be such that K i
t than 0.8 for all conditions of moderation and reflectTdd. s no grea er 5.3 Reactor Room The reactor room houses the reactor assembly and accessories a.
required for its operation and maintenance.
b.
The reactor room is separate room in the Merrill Engineering Building constructed with 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 nomal operating conditions, and to a level below criterion 19, Appendix A, 10 CFR 50 recommendations under accident conditions.
Access doors to and from the reactor rooms will contain locks.
c.
6.0 ADMINISTRATI1'E CO.\\TROLS 6.1 ORGANIZATION The administrative organization for control of the reactor facility and The its operation shall be as set forth in Figure 1 attached here to.
authorities and responsibilities set forth below are designed to comply with the intent and requirements for administrative controls of the l
reactor facility as set forth by the Nuclear Regulatory Commission.
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I Institutional I
Council University of Utah President l
Vice President for Research Reactor Safcty Radiological Director Nuclear IIealth Courtittec Committee Engineering Laboratory 1
NRC I
Reactor Reactor Radiation Safety Administrator Officer Supervisor
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i Reactor
_d_.---------------J Operators Line of Responsibility
Line of Communication Figure 1 University of !Itah Admnistrative Organization for Nuclear Reactor Operations i
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6.1.1 PRESIDENT The President is the chief Adninistrative officer responsibic for the University and is responsible to the Institutional Council fn whose name the application for licensing is mde.
6.1.2 VICE PRESIDENT FOR RESEARGI The Vice President is the Administrative Officer responsible to the President for all research facilities at the University.
In this capacity he shall represent the President in all health and safety matters pertaining to the reactor facility.
6.1.3 DIRECIDR NUCLilAR ENGINEERING LABORATORY The Director of the Nuclear Engineering Laboratory is the Administrative Officer responsible for the Reactor Facility and its operation, maintenance, and safety.
In this capacity he shall have final authority and ultimte responsibility for the reactor facility and, within the linitations set forth by the facility license, make final policy decisions on all phases of reactor operation; appoint personnel to all positions reporting to him as described in Section 6.1 of the Technical Specifications and as shown on Figure 1 of these specifications; be advised in all maters concerning health and safety by the Radiological Health Comittee; and be advised in all matters concerning reactor safety by the. Reactor Safety Conni~ttee.
6.1.4 REACIUR ADMINISTRATOR The Reactor Adninistrator (RA) is responsible to the Reactor Safety Committee and the Vice President for Research for insuring regulatory compliance of the reactor facility.
In this capacity, he shall, within the policies set forth by the Director and the facility license, prepare all regulations for the facility, review and approve all procedures, seek approval of all procedures and proposals for changes and experiments from the Reactor Safety and Radiological Health Committees, and be responsible for the health and safety of all personnel in the reactor facility.
6.1.5 REACIOR SUPERVISOR The Reactor Supervisor (RS) shall be responsible for the preparation, promulgation, and enforcement or adninistrative controls including all rules, regulations, instructions and operating procedures to ensure that the facility is operated in a safe, conpetent, and authorized manner at all times. He shall direct the activities of Operators and Technicians in the daily operation of the reactor; schedule reactor operations and maintenance; be responsible for the preparation, authentication, and storage of all prescribed logs and operating records of the facility; authorize all experiments, procedures, and changes thereto which have first received approval of the Reactor Safety Committee, the Radiological Health Committee, and the Reactor Administrator, and be responsible for the preparation of all instructional muuals and experimental pmcedures involving use of the reactor.
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Reactor Operators shall be responsible for the REACIDR OPERNIURS_
manipulation of the reactor controls, monitoring of instrumentation, 6.1.6 operation of reactor related equipment, and mintenance of complete A Reactcr and current records during operation to the facility.
Operator shall be in direct charge of the reactor cons to the rules, instmetions, and procedures established by the Reactor Administrator and Reactor Supervisor for operation of the reactor and the performnce of experiments.
The Reactor Safety Committee (RSC) shall be REACIOR SAFL'IY OLT4tITTEE resix>nsible for independent reviews and audits of facility operations 6.1.7 to insure that the reactor is operated in a safe and competent nanner within the requirements of the NRC and advise the Vice President for Research in all matters related to reactor safety and personnel safety.
We Reactor Safety Comittee shall hold periodic meetings and have the authority to conduct reviews and audits of reactor operations.
This Comittee (RCSC) shall advise the RADIOLOGICAL IIEALTII C0!4fITTEE Vice President for Research in all matters concerning the health and 6.1.8 safety of personnel who might be exposed to radiation produced by This comittee University owned and/or operated sources or equipment.
shall review, approve, and promulgate a Radiation Safety Program for This committee shall be informed of all reportable occurrences related to radiation health and safety and reactor safety the University.
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which are reportable to any authorities outside the University, and advise the President of such occurrences and make recommendatio j
the Vice President with regard to any such matters.
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6.1.9 RADIATION SAFETY OFFIG R The Radiation Safety Officer (RS0) shall be the chief admnistrative officer of the Cornittee and represent the comittee in matters concerning the radiation safety aspects of reactor l
operation. lie shall prepare the University's Radiation Safety manual and have the authority to enforce the regulations, rules, and procedures set forth by the Radiological IIcalth Comittee, suspend the operation and use of radiation producing devices when their use is in violation of these rules, and secure such sources of radiation unitl corrective action is taken. Ile shall also have the authority to disapprove the acquisition of radiation producing sources until satisfactory evidence is presented to ensure the safe storage and use of these facilities.
The Radiation Safety Officer is also responsible for the reporting of all reportable occurrences to the appropriate regulatory agency and for ensuring that the appropriate follow up action is taken.
6.1.10 OPERATING STAFF The ninimun operating staff during any time in which the reactor is not shutdown shall consist of one licensed Reactor Operator and one other person approved by the Reactor Supervisor. A licensed Senior Reactor Operator shall supervise all reactor maintenance or nodification which could affect the reactivity of the reactor.
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6.2 STAFF QUALIFICATIONS The Director of the Nuc1 car Engineering Laboratory, the Reactor Superivsor, licensed Reactor Operators, and technicians performing reactor maintenance shall meet the minimum qualifications set forth in ANS 15.4. Reactor jafety Committee members shall hrve a minimura of five (5) years experience in their profession oc a baccalaumste degree and two (2) years of professional experience. Reactor Safety committee members will generally be University faculty members with considerable experience in their ama of expertise. 7he Radiological Safety Officer shall have a baccalaureate degree in biological or physical science and have at least two (2) years e>perience in health physics.
6.3 TRAINLNG The Director of the Nuclear Engineering Laboratory shall be responsible for directing training as set forth in ANS 15.4, " Standards for Selection and Training of Personnel for Research Reactors".. All licensed mactor operators shall participate in requalification training as set forth in 10 CFR 55.
6.4 REACIOR SAFETY. COMMITTEE 6.4.1 MEETINGS AND QUOR31 Reactor Safety Committee shall meet as often as deemed necessary by the Reactor Safety Committee Chairman who is the Reactor Administrator 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 mgular members. At no time shall the operating organization comprise a voting majority of the members at any Reactor Safety committee meeting.
6.4.2 REVIElG The Reactor Safety Committee shall review:
Safety evaluations for changes to procedures, equipment or systems, and a.
tests or experiments, conducted without Nuclear RegulatoIy Comctission approval under the prevision of 10 CFR 50.59, to verify that such actions did not constitute an unreviewed safety question.
b.
Proposed changes to pmcedures, equipment or systems that ciunge the original intent or use, and are non-conservative, or those that involve an unreviewed safety question as defined in 10 CFR 50.59.
Proposed tests or experiments which are significantly different c.
fmm previous approved tests or experiments, or those that involve an unreviewed safety question as defined in 10 CFR 50.59.
d.
Proposed changes in Technical Specifications or licenses.
Violations of applicable statutes, codes, regulations, orders, e.
Technical Specifications, licnese requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect nuclear safety.
g.
Reportable occurrences.
i h.
Audit reports.
6.4.3. AUDITS Audits of facility activities shall be perfonned at least quarterly under the cognizance of the Reactor Safety Committee but in no case by the personnel responsible for the item audited. These audits shall examine the operating records and encompass but shall not be limited to the following:
The confonnance of the facility operation to the Technical Specific-a.
ations and applicable license conditions, at least annually, b.
'Ihe Facility Emergency plan and implementing procedures, at least every twa years.
The Facility Security Plan and implementing procedures, at least c.
every two years.
6.4.4 AUITORITY The Reactor Safety Comraittee shall report to the Vice President and shall advise the Director of the Nuclear Engineering Laboratory on those amas of responsibility outlined in section 6.1.7 of these Technical Specifications.
6.4.5 MIhUTES OF TIE REACTOR SAFETY BOARD The Reactor Administrator shall direct the preparation, maintenance, and distribution of minutes of its activities. These minutes shall include 3
I a summary of all meetings, actions taken, audits, and reviews.
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I 6.5 PROCEDURES
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There shall be written proceduies that cover the following activities:
1 Startup, operation, and shutdown of the reactor, a.
b.
Fuel move:: lent and changes to the core and experiments that could affect reactivity, Conduct of irradiations and experiments that could affect the c.
operation or safety of the reactor.
i d.
Preventive or corrective maintenance which could affect the safety of the reactor.
e.
Surveillanca, testing, and calibration of instruments, components, and systems as specified in section 4.0 of Wese Technical Specifications.
f.
Implementation of the Security Plan and Emergency Plan.
Tne above listed procedures shall bc appmved by the Director cf the Nuclear 1:n;;uleerin;; Lacoratory and the Reactor Comittee. Tempora:v urocedures which do not change the intent of previously anproved pmcedu'res and which do not involve anv unreviewed safety avestion nav be encloved on approval ny the Reactor Supernsor or Director of the ' Nuclear Erigineering Lat'.
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I 6.6 SAFETY LIMIT VIOLATION The follcwing acticns shall be taken in the event a Safety Limit is violated:
The reactor will be shut down i=ediately and reactor operation will a.
not be resumed without authorizaticn by the Nuclear Regulatory Cnmmusion (NRC).
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 Board net later tnan the next work day.
i A Safety Limit Violatien Repcrt shall be prepared for review by the j
c.
l Reactor Safety Board. This report shall describe the applicable circumstances preceding the violation, the effects of the violation l
upon facility ccc:ponents, systems er structures, and corrective action to prevent recurrence.
The Safety limit Violation Repcrt shall be submitted to the NRC, and d.
Reactor Safety Board within 14 days of the violation. k
6.7 REPORTING REQUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the appropriate NRC Regional Office.
6.7.1 ANNUAL OPERATING REPORT Routine annual operating reports shall be submitted no later than thirty (30) days following the end of the operating year. Each annual report shall include a sumary of the following activities occurring during the operating year:
a.
Facility modifications.
b.
Results of major surveillance tests and inspections.
c.
Corrective maintenance performed.
d.
Energy pr~hrM by the reactor in watt-hours.
e.
Unscheduled shutdowns.
f.
Reactor Safety Board action pertinent to the facility.
g.
Any activities which require reporting per 10 CFR Shparamir 50.59 h.
Any reportable occurrences as defined in section 6.7.2 of these Technien1 Specifications.
6.7.2 REFUKIABLE OCCURRENCES l
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 and 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.
Infcrmation provided shall contain narrative material to provide cocplete explanation of the circumstances surrounding the event.
(1) Failure of the reactor prosection system subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reached the sermint specified as the limiting safety system setting in the tecmical specifications. :
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(2) Operation of the reactor when any parameter or operation subject to a limiting condition is less conservative than the limiting i
condition for operation established in the technical specifications.
(3) Abnormal degradation discovered in a fission product barrier.
(4) Reactivity balance anomalies involving:
(a) disagreement between expected and actual critical positions of approximately 0.3% ak/k; (b) exceeding excess reactivity limit; (c) shutdown margin less conservative than specified in technical specifications; (5) Failure or malfunction of one (or more) component (s) which prevent, or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in Safety Analysis Report.
(6) Personnel error or procedural inadequacy wilich prevents, or could prevent, by itself, the fulfillmmt of the functional requirements of systems required to cope with accidents analyzed in Safety Analysis Report.
(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 fer the Technical Specifications that have pemitted reactor operation in a manner less conservative than assmed in the analyses.
(8) Perfomance of struccures, systas, 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 or technical specifications bases; or discovery dur1ng 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.
l 6.8 RECORD RETENTION 6.8.1 Records to be retained for the life of the facility:
a.
Annual reports.
b.
Records of controlled or uncontrolled release of radioactive effluents to the environment..
A c.
Fuel inventories and fuel transfers.
d.
Operating logs.
e.
Maintenance logs.
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f.
Updated drawings of the reactor facility.
g.
Personnel dosimetry records on file with the Radiological Safety
/
Officer.
h.
Minutes of the Reactor Safety Board meetings.
6.8.2 Records to be retained for a period of at least three years:
a.
Surveillance activities required by Technical Specifications.
b.
Facility radiation and contamination surveys.
6 Personnel requalification and training records will be kept at least one yen after termination of employment.
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