ML19269D406
| ML19269D406 | |
| Person / Time | |
|---|---|
| Site: | 05000072 |
| Issue date: | 05/02/1979 |
| From: | Grimes B Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19269D405 | List: |
| References | |
| R-025-A-010, R-25-A-10, NUDOCS 7906020233 | |
| Download: ML19269D406 (30) | |
Text
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WNtT20 STATES
~
'JUCLsM REGULMORY cOMMISSloN 0,
MSHINGTCN. 3. C. 20555
'I THE INSTITUTIONAL COUNCIL OF THE UNIVERSITY OF UTAH DOCKET NO. 50-72 AMENDED FACILITY OPERATING LICENSE Amendment No.10 License No. R-25 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by The Institutional Council of the University of Utah (the licensee) dated July 8,1977, as supplemented by filings dated March 3,1978 and January 2,1979, 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; 8.
The faciltty will operate in conformity with the application, the provisions of 'he Act, and the regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations;
-0.
The licensee is technically and financially qualified to engage in the activities authorized by this operating license in accor-dance with the regulations of the Comission; E.
The licensee is a nonprofit educational institution and will use the facility for the conduct of educational activities, and has satisfied the applicable provisions of 10 CFR Part 140, " Financial Protection Requirements and Idemnity Agreements," of the Comission's regulations; F.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; 2263 20 7 90 602 0$L33.
r
. G.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and H.
The receipt, possession and use of the byproduct and special nuclear materials as authorized by this license will be in acci Mrce with the Commission's regulations in 10 CFR Parts 30 and 7'
.ncluding 10 CFR Sections 30.33, 70.23 and 70.31.
2.
Facility Operating License No. R-25 issued to The Institutional Council of the University of Utah is hereby amended in its entirety to read as follows:
A.
This license applies to the homogeneous nuclear reactor model AGN-201M, Serial No.107 (the reactor), owned by the Institutional Council of the University of Utah, located in the Merrill Engineering Building on the licensee's campus at Salt Lake City, Utah, and described in the licensee's application for license dated June 1,1957, and subsequent amendments thereto, including the application for license renewal dated July 8,1977, and supplements thereto dated March 3,1978 and January 2,1979.
8.
Subject to the conditions and requirements incorporated herein, the Commission hereby licenses The Institutional Council of the University of Utah:
(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 Salt Lake City, Utah, in accordance with the procedures and limitations set forth in this license; (2) Pursuant to the Act and 10 CFR Part 70, "Special Nuclear Material," to receive, possess and use up to 700 grams of contained uranium-245, enriched to less than 20% in uranium dioxide (UO ) embedded in radiation stabilized 2
polyethylene, in connection with operation of the reactor; and (3) Pursuant to the Act and 10 CFR Parts 30 " Rules of General Applicability to Licensing of Byproduct Material," and 70 to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the reactor.
2263
- 21
n
- C.
This license shall be deemed to contain and is subject to the conditions specified in the following Comission regu-lations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate she reactor at steady-state power levels up to a maximum of 5 watts (thermal).
(2) Technical Soecifications The Technical Soecifications contained in Acpendix A attached hereto are hereby incorporated in this license.
(These Technical Specifications supersede the Technical Specifications issued June 28, 1968, as amended.)
The licensee shall operate the reactor in accordance with the Technical Specifications.
{
(3) Physical Security Plan I
The licensee shall maintain in effect and fully implement all provisions of the Comission-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 the following documents which are withheld from public disclosur'e pursuant to 10 CFR 2.790(d):
Original dated August 4,1972, submitted with letter dated February 6,1974, and i
Revision 1, submitted with letter dated July 16, 1974.
D.
This amended license is effective as of the date of issuance and shall expire at midnight, September 12, 1997.
t FOR
': NUCLEAR REGULATO C MMISSION "A
. 4-r._
Srian K. Grimes, Assistan: Direc:ce for Engineering & projects Division of Operating Reactors r x
- Technical 2263 ;22 Scecifications
- a:s cf :ssuance:
May 2, 1979
APPENDIX A TO FACILITY OPERATING LICENSE NO. R-25 TECHNICAL SPECIFICATIONS FOR THE INSTITUTIONAL COUNCIL OF THE UNIVERSITY OF UTAH AGN-201M REACTOR (SERIAL #107)
DOCKET NO. 50-72 2263
_,23 Amendment No.10 Dated: May 2, 1979
e s
TABLE OF CONTENTS PAGE 1.0 DEFINITIONS.
......1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS.
4 2.1 Safety Limits.
4 2.2 Limiting Safety System Settings.
4 3.0 LIMITING CONDITIONS FOR OPERATION.
5 3.1 Reactivity Limits.
5 3.2 Control and Safety Systems.
6 3.3 Limitations on Experiments.
9 3.4 Radiation Monitoring, Control and Shielding..
10 4.0 SURVEILLANCE REQUIREMENTS.
. 11 4.1 Reactivity Limits.
. 11 4.2 Control and Safety System.
. 12 4.3 Reactor Structure.
13 4.4 Radiation Monitoring and Control...
. 13 5.0 DESIGN FEATURES.
. 14 5.1 Reactor..
. 14 5.2 Fuel Storage.
. 15 5.3 Reactor Room.
. 15 6.0 ADMINISTRATIVE CONTROLS.
. 15 6.1 Organization.
. 15 6.2 Staff Qualifications..
. 20 6.3 Training.
. 20 6.4 Reactor Safety Committee.
. 20 6.5 Approvals.
. 22 6.6 Procedures.
. 22 6.7 Experiments.
23 6.8 Safety Limit Violation.
23 6.9 Reporting Requirements.
. 23 6.10 Record Retention.
. 25 2263 324
t.
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 cnannel such :nat its output responds, within acceptable range and accuracy, to known values of the parameter wnich the channel Calibration shall encomcass the entire channel, including measures.
equipment, actuation, alarm, or trip.
1.2 Channel Check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior.
This verification may include comparison of the channel with other independent channels or methods measuring the same variable.
1.3 Channel Test - A channel test is the introduction of.a signal into tne enannel to verify that it is operable.
1.4 Exceriment -
a.
An experiment is any of the following:
(1) An activity utili::ing the reactor system or its components or the neutrocs or radiation generated therein; (2) An evaluatico or test of a reactor system cper"ational, surveillance, or maintencnce 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 Exceriment - 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, bouyant, or other forces which are normal to the operating environment of the experiment or which might arise as a result of credible malfunctions, c.
Unsecured Exceriment - 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 pcets of experi-ments are deemed to be unsecured when they are in motion.
2263 25 1
d.
Movable Exoeriment - A movable experiment is one which may be inserted, removed, or manipulated while the reactor is critical.
e.
Removable Exoeriment - A removable experiment is any experiment, experimental facility, or component of an experiment, other than j
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 Exoerimental Facilities - Experimental facilities art those portions l
of the reactor assemoly that are used for the introduction of experi-ments 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 Exelosive Material - Explosive material is any solid or -liquid wnfen is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in " Dangerous Properties of Industri-al Materials" 5y N. L. Sax, Third Ed. (1968), or is given an Identification of i
Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Association in its publication 704-M,1966, "Identifica-ttbn System for Fire Hazards of Materials," also enumerated in the
" Handbook for Laboratory Safety" 2nd Ed. (1971) pub'.ished by The Chemical Rubber Company.
l.7 Measurine 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 coerabl
- Operable means a component or system is capable of per-forming its intended function in its normal manner.
1.9 Ooeratino.- Operating means a component or system is performing iLts intended function in its normal manner.
1.10 Potential Reactivity Worth - The potential reactivity worth.of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position l
or configuration.
l Evaluations of potential reactivity worth of experiments also shall include effects of possible trajectories of the experiment in motion relative to the reactor, its orientation along each trajectory and circumstances which an cause internal changes such as creating or filling of void spaces or motion of mechanical For removable experiments, the potential reactivity components.
worth is eoual to or greater than the static reactivity worth.
1.11 Reactor comoonent - A reactor component is any a::aratus, device, or material :nat. is a normal part of the reactor assembly.
1.12 Reactor Oceration - Reactor coeration is any conditien wnerein
- ne reacto" is net shut dcwn.
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1.13 Reacter Safety System - The reactor safety system is that ccetination of safety cnannels and associated circuitry which forms the autcmatic protective system for the reactor or provides information wnich recuires manual protective action be initiated.
l.14 Reactor Shutdewn - The reactor shall be considered shutdown whenever a.
either:
1.
All safety and c:ntrol rods are fully withdrawn from the core, or 2.
The core fuse melts resulting in separation of the
- core, and:
b.
The reactor censole key switch is in the "off" position and the key is removed from the console and under the control of a licensed operator or the Reactor Administrator.-
1.15 Safetv Channel - A safety channel is a mea 5Uring ch_annel in the reaccor safety system.
1.16 Static Reactivity Worth - The static reactivity worth of an experimenc is the acsolute value of the reactivity change which is measurable by calibrated control or regulating red c:mparison metheds 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 r intended position.
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.i e
2.0 SAFETY LIMITS AND LMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits Aeolicability This specification applies to the =axi=um steady state power level and maximum core temperature during steady state or transient opera-tion.
Ob iective To assure that the integrity of the fuel material is maintained and all fission products are retained in the core matrix.
Soecification a.
The reactor power level shall not exceed 100 watts.
b.
The maximum core temperature shall not exceed 2000C during either steady state or transient operation.
Bases The polyethylene core caterial does not =elt below 2000C and is expected to maintain its integrity and retain essentially all of the fission products at temperatures below 2000C. The Hazards Summary Report dated February 1962 submitted on Locket F-15 by Aerojet-General Nucleonics (AGN) calculated a steady state core average temperature rise of 0.44 C/ watt.
Therefore, a steady state power level of 100 watts would result in an average core temperature rise of 440C. The corresponding maximum core temperature would be below 2000C thus assuring integrity of tha core and retention of fission products.
2.2 Limiting Saferv system Settinas Apolicability This specification applies to the parts of the reactor safety system which will limit maximum power and core temperature.
Objective To assure that automatic protective action is initiated to prevent a safety llait from being exceeded.
Seecification a.
The safety channels shall initiate a reactor scram at the following limiting safety system settings:
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Channel Ccnditien LSSS Nuclear Saferf f2 High Pcwer 110 watts Nuclear Safety #3 High Pcwer 110 watts b.
The care themal 61se shall :nelt when heated to a te=perature of about 1200C tesulting in core separation and a reactivity loss greater than Stak.
Bases Based en bst:unentaticn respcnse ti:nes sod scram tests, the ACN Hatards Report e-c W ad that reacter periods in excess cf 30-50 milliseconds would be adequately arrested by the scram system.
Since the mreism available excess reac~.2tief n the reactor is less i
than ene dollar the reactcc em;mnt bec::xne prtmpt critical and the wuc.5---Ng shtest possible period is greater than 200 ::rilli-seconds. The high pcwer LSS5 of 10 watts in conjunction with auto-
- natic safert syste=s and/or mam =1 scras capabilities will assure that the safeef limits will not be exceeded during steady state or as a re-sult of the :nost severe credible transient.
In the event of failure of the :. actor to scrsm, the self ti 4 ting characeristics due to the high negative temperature coefficient, and the :nelting of the thesal fuse at a te::perature belcw 120*C will assure safe shutdcwn without exceeding a core te=perature of 200*C.
3.0 LNITING (DiDITICNS FCR CFERATICN 3,1 Reactivier Limits Acolicabilier This specificaticn applies to the reactivirf cc:nditicn or' the reactor and the Mvirf worths of c::ntrol rods and experiments.
Cbiective To assure that the reactor can be shut d=wn at all times and that the saferf limits will not be exceedec.
Scecificatien The available excess reactivirf with all control ard saferf reds a.
6:lly inserted and including the potential reactivity worth of all experi=ents shall not exceed 0.65%.ik/k referenced to 20*C.
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b.
The shutdcwn margin with the mcst reactive saferf or c=ntrol mi fully inserted shall be at least 1%.i k/k.
c.
The reactivity worth of the centrol and safety rods shall ensure sub-criticalief en the withdrawal of the cearse centrol rod or any cne safety rod.
Bases The limitaticns en total core excess reactivity assure reactor per-iods of sufficient Icngth so that the reactor protectica system and/or cperator ac ica vill be able to shut the reactor dcwn with-out ~Wng any safety limits. The shutdcwn margin and centrol and safety mi reactivity limitaticns assure that the reactor can be brcught and maintained subcritical if the highest reactivity rod fails to scram and r-mim in its mest reac ive positicns.
3.2 Centrol and Safety Systems Are H ability These specificaticns apply to the reactor c::ntrol and safety systems.
Cbjective To specif icwest acc..ptable level of pn.La, inst = ment set f
points, and the mini:m:n number of cperable ca::penents fer the reactor centrol and safety systems.
Seecificatien a.
The total scra:a withdrawal time of the safety rods and coarse c::ntrol rod shall be less than 200 =illisec cds.
b.
The average reactivity addition rate for each control or safety rod shall not exceed 0.065% ak/k per second.
The safety cds and coarse centr:1 rod shall be interlocked c.
such that:
1.
Reactor startup cancot e-:ce unless both safety rods and coarse centrol mi are fully withdrawn.~.
the core.
2.
Only cne safety rod can be inserted at a time.
3.
The c: arse centrol md cannot be inserted unless both safety rods are 3:117 inserted.
d.
All reactor safety system instrumentation shall be operable in accordance with Table 3.1 whenever the teactor control or safety rods are not in their fully withdrawn position.
However Nuclear safety channel 11 may be bypassed for operation at power levels exceeding 0.1 watt.
2263 330 TABLE 3.1 Saferr Ch:r=nel Set Poi.:
m.cticn 3
Nuclear Saferf f1*
L::w ccunt rate t 10 cps sc am belcw 10 cps t
Nuclear Saferf #2 High pcwer
< 10 wa::
scra::: at pcwer >10 wat:
[
L:w pcwer t 1.0 x 10'l~' a=ps scram at scurce levels I
< 1.0 x 10-12 amps i
Reactor period
> S sec sc=m at periods < 5 sec
.%-Tar Saferf f 3 (T iner Pcwer)
Righ Pcwer
< 10 wat:
scram at pcwer >10 war:
L::w pcwer g 5% #""
scale scram at scu=e levels
< 5% of S.111 scale
.* m mt sc:sm scram at cperater eptien Nuclear Safety Channel dl =av ':e 'c" passed a: power levels exceeding 0.1 va::.
i i
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The shield water level inter 1cck shall be set to prevent reactor e.
startup and scram the reactor if the shield water level falls 10.5 inches belcw de highest point en de reactor shield tank manhole cpening.
f.
The shield water te=perature intericek shall be set to prevent reactor star:uo and scram the reac::: if the shield water :e=perature falls belcw 15' C.
8 The seismic displacement intericck sensor shall be installed in such a manner to prevent reactor startup and scram the reactor during a seismic disp 1'mt.
h.
A loss of eleC*ric pcwer shall cause the reac*cr to scram.
Bases ne specificatiens en scram withdrawal time in ccnjuncticn with the safety systs inst:=mentatica and set points assure safe reactor shutdcwn Smng the most severe foreseeable transients. Interlocks en cen:rol and safer /
rods assure an orderly approach to criticality and an adequate shu:dcun capabilirf. The limitations on reactivity addition rates allow only relatively slow increasas of reactivity so that ample time will be available for manual or automatic scram during any operating conditions.
The neutrcn detector chm =15 (cuclear safety channels 1 thrcugh 3) assure that reac:cr pcwer levels are adequately =cnitcred during r=.ac:cr star =m and cperaticn. Requ:trements en rni m neutrcn levels will prevent reac:cr sg unless channels are cperable and respcnding, and will cause a scram in the event of inst:tmentatica failure. The pcwer level scrams initiate redundant an:cmatic protective ac ica at pcwer level scrams icw encugh to assure safe shutdckn witcu: exceeding any saferf 1 bits. He period scram c=nservatively limits de ra:e of rise of reactor pcwer to periods which are wnmily ccntrollable and will au:cmatically s-de reac:cr in the event of unexpected large reac ivirf additicns.
The XN-101's negative ta=perature coe:ficient of reactivi:v causes a reactivir/ increase with decreasing ccre temperature. He shield water te=perature intericek will preven: reac or cperatien at ta=peratures belcw 15'C thereby lui.in associated with ta=perature decreases. g potential reactivi:7 addi:icns Water in de shield tank is an i=ccr:an: c:mpenent of de reactor shield and cperatien witcut the water =av procuce excessive radiarien levels. H e shield tank water level inter 1cck will preven:
reac:Or cperaticn widcu: adeqt. ate water levels in the shield tank.
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t
The reactor is designed to withstmi 0.6g acceleratiens and 6 cm A seisnic inst:usent causes a reactor scrm whenever disp 1=r e ts.
the instrument receives a horizontal acceleration that causes a n e seismic hori:cntal disn1=~t of 1/16 inch or greater.
displacement ihterlock assures that the reactor will be ss. M and brcught to a stteritical configurati::n during any seismic disturbance that :nay cause damage to the reactor or i:s ccc:penents.
The =anual scram allcws the operater to - m m'17 shut dcwn the reactor if an tasafe or otherwise abnorral ccndi:icn occurs dat does not A loss of electrical power de-energi:es otherwise scras the reacter.
the safety and coarse c::ntrol rod holding magnets causing a reactor scran thus assunng safe and iW4nte shutdchu in case of a pcwer cutrage.
3.3 M 4:stiens en Exceriments_
Acolicability This specificaticn applies to experi=ents installed in de reae::: and its experimental facilities.
Cbiective To prevent damage to the reac:cr or excessive release of radiocc:ive matenals in the event of an experinent 02ilf W.
Soed " catien Expenments cen"in4*g =aterials w..aive to reactor czpcnents or which c::ntain liquid er gasecus, fissicnable =aterials shall a.
be doubly encapsulated, Explosive materials shall not be inserted into experimental b.
facilities of the reactor or stored within the confines of the reactor facility.
He radioactive : material centen:, 4-et">4"g fissicn procucts of any exper bent shall be I'm4:ed so dat d e cecolete release c.
of all gasecus, particulate, or volatile c::mpenents frcm de experiment will ::ot result in doses in excess of 10% of de equivalent am 21 doses stated in 10 CFR Part 20 for perscns occucying (1) tm icted areas centi::ucusly for tw hcurs starting at time of release er (2) restricted areas during the length of ti:ne rec,uired to evacuate the restricted area.
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+
d.
The radicactive =aterial centent, !"chv"ng fissicn proch: cts of any dcubly ene'etsted experiment shall be limited so that the ccmplete release of all gasecus, particulate, or volatile c:mpenents of the experi=ent slut 11 not result in exposures in excess of 0.5 Rem *atiale body or 1.5 Rem thyroid te perscns ocMg an tmrestricted area centinucusly fer a period of tw hcurs starting at the time of release or exposure in excess of 5 Rem whole body or 30 Rem thyroid to perscas oc W M a restricted area during the length of time required to evacuate the restricted area.
Bases These specificaticns are intended to reduce the likelihood of damage to reactor ctupenents and/or radicactivity releases resulting frcm an experiment
- nihrre and to protect cperating perscnnel and the public frta excessive radiaticn deses in the event of an experiment failure.
3.4 Radiation Monitoring, Control and Shielding Apolicability 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 An operable portable and installed radiation survey instrument a.
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.
The following shielding requirements shall be fulfilled during c.
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 ther=al column shall be filled with water or graphite except during a critical experiment (core loading) or during measurement of reactivity worth of thermal column water or graphite.
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_1,_
Bases Radiation surveys perfor=ed under the supervision of a qualified health physicist have shown that the total gar =a, thermal neutron, and fast neutron radiation dose rate in the reactor room, at the closest approach to the reactor is less than 100 mres/hr at reactor power levels less than or equal to 5.0 watts and the thermal column filled with water.
"D:e fac'. lief shielding in ccnjunctica with designated restricted radiatir.n areas is designed to li=it radiaticn deses to facility persennel and to the public to.a level belcw 10 CFR 20 limits ur. der cperating ccnditicus, and to a level belcw criterien 19, Appendix A, 10 CFR 50 ret
"*tiens under ad Amt c::nditicas.
4.0 SCRVEIllANCE RECUILW.a Acticas specified in this secticn are not required to be psi.'siM if dur2ng the specified surveilinen period the reactor has not been brcught critical or is mim4ad in a shutdcwn ccnditica extending beyond the specified surm411w= period. Hcwever, the surveill w e require =ents ::mst be fulfilled prior to subsequent start:p of the reac*ar.
4.1 Reactivier Limits Acolicability his specificarica applies to the surteillmee requirements for reactivirf limits.
Cbiective To assure that reactivirf limits for Specificaticn 3.1 are ::ct exceeded.
Soecificatien Safety and c::ntal red reactivirf worths snm11 he seasured annually, a.
but at interials not to exceed 16 mcnths.
b.
Total emess reactivirt and shutdcun :mLTh shall be determined ni'y, but at interials not to exceed 16 mcnths.
The reactivirf wcrth of an experiment shall be esti=ated Or c.
measured, as appi+ihte, before or durtng the fi st sta:- :p subsequent to the experi=ent's inserticn.
2263 ;35 t
Bases The control and safety rods are inspected and their 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 cne red not fu:ct ening and that the -=M n possible reactivity i
inserticn will not result in reactor periods shcrter than these that can be adequately terminated by either cperater or aut:matic acticn.
Based en experience with AGN reactors, significant changes in reattivity or rod wrth are not expected within a 16-menth period.
4.1 Centrol and Safety Systas Acolicabilier This specificatica applies to the surveillance requirements of the reactor centrol and safety systems.
Objective To assure that the reactor control and and safety systems are operable as required by Specification 3.2.
Specification a.
Safety and control rod scram times and average reactivity insartfon rates shall be measured annually, but at intervals not t'o exceed 16 months.
b.
Safety and centrol rods and drive shall be inspected fcr deterioratica at interrals not to em=.ed 2 years.
c.
A chmal test of the fo11cwing safety channels shall be perfor=ed prior to the first reacter starn:p cf the day er pr!ce to each cperatica extM ny more than ene day:
Nuclear Safety 91, f2, and f3 Mmm1 scrsa d.
A chmnnel test of the seismic disp 1'emt inter 1cck shall be per h d se m m117 e.
A channel check of the fo11cwing safety channels shall be perfer:ned daily t enever the reactor is in cperaticn.
Nuclear Safety fl,f2, and #3 f.
Prior to each day's coeratien er prict to each cperatica extending more than ene day, sai~ety red f1 shall be insarted and swd to veriff cperabilief.
2263 336 g.
The period, count rate, and pcwer level measuring cha:mels s!all be calibrated and set points verified - n M y, but at intervals not to exceed 16 :ncnths.
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 calibrations shall be performed annually, but at intervals not to exceed 16 =onths.
Bases The ch2m *1 tests and checks required daily cr before each start:.:D will assure that the safety eb= mis and scra:n functicns are cperable.
Based en cperating experience with reactors of this type, the armm1 scra:2 :nea.muts, samal calibraticns, set point verificaticns, and insIw-*im are of sufficient frequency to assure, with a high degree of em%c=, that the safety system settings will be within acc..ptable drift tolerance for cperaticn.
4.3 Reactor Structure Aeolicability This specification applies to surveillance requirements for reactor components other than control and safety rods.
Obiective To assure integrity of the reactor structures.
Specification a.
The shield tank shall be visually inspected every two ye srs.
If apparent excessive corrosion or other da= age is observed, corree?.ive measures shall be taken prior to subsequent reactor operation.
b.
Visual inspection for water leakage from the shield tank shall be per-formed annually. Leakage shall be corrected prior to subsequent reactor operation.
Bases Based on experience with reactors of this type, the frequency of inspection and leak test require =ents of the shield tank will assure capability for radiation protection during reactor operation.
4.4 Radiation Monitoring and Centrol Aeolicabiliev This specification applies to the surveillance requirements of the radiation monitoring and control systems.
2263
'37
_u-
o Objective To assure that the radiation monitoring and control systems are operable and that all radiation areas within the reactor f acility are identified and controlled as required by Specification 3.4.
Specification All portable and installed radiation survey instruments assigned to a.
the reactor facility shall be calibrated under the supervision of the Radiation Safety Of ficer 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.
A radiation survey of the reactor room and reactor control room c.
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 cor-responding to reactor operating power levels.
Bases The periodic calibration of radiation monitoring equipment and the surycil-lance of the reactor room high radiation area alarm will assure that the radiation monitoring and control systems are operable during reactor operation.
The periodic radiation surveys will verify the location of radiation and high radiation areas and will assist reactor facility personnel in properly labeling and controlling each location in accordance with 10 CFR 20.
5.0 DESIGN FEATURES 5.1 Reactor a.
The reactor core, including control and safety rods, contains approxi-mately 660 grams of U-235 in the form of(20% enriched UO2 dispersed in approximately 11 kilograms of polyethylene. The lower section of the core is supported by an aluminum red hanging from a fuse link.
The fuse melts at a fuse temperature of about 1200C causing the lower core section to fall away from the upper section reducing reactivity by at least 57. a k/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.
b.
The core is surrcunded by a 20 cm chick high density (1.75 g=/cm3) graphite reflector followed by a 10 cm thick lead ga==a shield. The core and part of the graphite reflector are sealed in a fluid-tight alu=inum core tank designed to contain any fissien gases that =ight leak f rom the core. 2263 ;38
c.
The core, reflector, and lead shielding are enclosed in and supported 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.
The 6i foot diameter, fluid-tight shield tank is filled with water constituting a 55 cm thick f ast neutron shield. The fast neutron shield is for=ed by filling the tank with approx 1=ately 1000 gallons of water. The complete reactor shield shall limit doses to operating personnel in restricted and unrestricted areas to levels less than permitted by 10 CFR 20 under operating conditions.
e.
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 =aterial.
These rods are lifted into the core by electromagnets, driven by reversible DC motors through lead screw asse=blies. Deenergi:ing the magnets causes a spring-driven, gravity-assisted scram. The fourth red 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 expert =ents and fuel devices not in the reactor, shall be stored in locked rooms in the nuclear engineering depart =ent laboratories. The storage array shall be such that K is no greater ff than 0.8 for all conditions of moderation and reflection.
5.3 Reactor Room The reactor room houses the reactor asse=bly and accessories required a.
for its cperation and maintenance, b.
The reactor rcom is a 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 normal 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 ADMINISTRATIVE CONTROLS 6.1 Ortanization The administrative organi:ation for control of the reactor facility and 4he operation shall be as set forth in Figure 1 attached hereto The authorities and responsibilities set forth belcw are designed to comply with the intent and requirements for administrative controls of the reactor facility as set forth by the Nuclear Regulatory Ccemission.
2263 339 Institutional Council University of Utah President Vice President for Research Reactor Safere Radiological Director.N.uclear Comittee Health Cemittee Engineering Laborater).
NRC Reactor Reactor Radiation Safet.c Supervisor q
Ad::anistrat:r Officer i
l I
i l
i Reactor
-a Operators Line of Pespcnsibility
Line of Ccnruni ation Figure i "niveccity cf Jtah
.Wiinistrative hpni ation for Nuclear Reacter 'pentions 2263 a40 6.1.1 President-:he President is the chief.W.inistrative Officer responsible for tne University and is responsibin to the Institutional Counc1T fn whose name the application for licensin; is made.
6.1.2 Vice President for Research -The Vice President is the Ad:ainistrative Orticer respenstole to tne President for all research facilities at the University. In this capacity he shall represent the President in all health and safety mtters per*aining to the reactor facili:v.
6.1.3 Director Nuclear Engineering Laboratorv-The Director of the Nuclear Engineering Lacoratorv is tne Actunistrative Officer responsible for the Reactor Facility and its operation, mintenance, and safetv. In this capacity he shall have final authority and ultimte responsibility for the reactor facility and, within the li:-itations set forth by the facilitv license, =ake final policy decisions on all phases of reactor operation; appoint personnel to all positions reporting to hint as described in Section 6.1 of the Technical Specifications and as shown on Figure 1 of these specifications; be advised in all natters concerning health and safety by the Radiological Health Cct:-i::ee; and be advised in all matters concerning reactor safety by de Reactor Saf4ty Cecrtittee.
o.l.4 Reactor Administrator-The Reactor Ad:.inistrator (RA) is responsible to the Reactor Sa:ety Cemittee and the Vice President for Research for insuring regulatorv cc :pliance of the reactor facilitv. In this capacity, he shall, within the policies set forth by the Director and the facili:V license, prepare all regulations for the facility, review and approve all procedures, seek approni of all procedures and proposals for changes and experiments from the Reactor Safety and Radiological Hea'th Corttittees, and be respcnsible fer the health and saferv of all persennel in the reactor facility.
6.1.5 Reactor Supervisor-lhe Reactor Supervisor (F$) shall be a licensed Senior Reactor Operator and shall be responsible for the preparation, promulgation, and enforcement of administrative controls including all rules, regulations, instructions and operating procedures to insure that the 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 of the reactor; schedule reactor operations and =ai.:enance; be responsible for the preparation, authentication, and storage of all prescribed logs and cperating records of the facilitv; authori:e all experiments, procedures.
and cunges thereto which have firs: received approval of the Reactor Safety Cemittee, the Radiological Heal-h Comittee, and the Reactor
.Wiinistrator, and be responsible for de preparation of all instn:ctio=1 manuals and e:gerimental pIncedures involving use of the reactor.
2263 341 6.1.6 Reactor Operators-Reactor Operators shall be responsible for the manipulation or the reactor controls, r:rnitoring of instrumentation, operation of reactor related equipment, and maintenance of comolete and current records during operation of the facility. A Reactor Operator shall be in direct charge of the reactor console at all times during reactor operation and when the reactor is not secured and confom to the rules, instrxtions, and procedures established by the Reactor Administrator and Reactor Supertisor for operation of the reactor and the perfomance of experiments.
6.1.7 Reactor Safety Committee-Be Reactor Safety Comittee (RSC) shall be respons1 ole for independent reviews and audits of facility operations to insure that the reactor is operated in a safe and competent manner within the requirements of the NRC and advise the Vice President for Research in all matters related to reactor safety ard personnel safety.
I Be Reactor Safety Comittee shall hold periodic =eetings and have the authcrity to conduct reviews and audits of reacter operations.
o.l.S Radiological Health Comittee Bis Corrittee (RCSC) shall advise the
\\1ce Presicent for Researen in all mtters concerning the health and safety of personnel who might be exposed to radiation produced by University owned and/or operated samtes or equipment. This cocmittee shall review, approve, and pmmulgate a Radiatien Safety Program for the University. This comittee shall be infomed of all reportable occurrences related to radiation heald and safety and reactor safety which are reportable to any authorities outside the University, and advise the President of such occurances and make reco=nendations to the Vice President with regard to any such =atters.
2263
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-ta-
6.1.3 Radiation Safety officer "D.e Mdia ica Safety Officer (F50) shall be
- ne cnae: ac::u.:us:rstive officer of de Cemittee and represen: de cernittee in natters conce: ting de ndistien safety aspects of reactor operatien. He shall prepare de University's Radiation Safety nanual and have de authori y to enforce the regulatiens, rules, and precedures set for:n by the Radiolcgical Heal d Co. ti :ee, suspend ce opera:icn and use of radiatien prehMg devices when their use is in violation of these rules, and secure such scunes of radistica un:11 corrective action is taken. He shall also have de auderity to disapprove de acquisitien of radiatica producing scunes until satisfac crv evidence is cresented to ensure the safe storage and use of dese facili-ies.
The' Radiatica Safety Officer is also respcnsible for the reporting of all reportable cccurrences to de a:propriate regulato:-e agency and for ensuring that de appropriate felict. up action is taken.
5.1.10 Operacing Staff The mini =um operating staff during any :ine in wnich the react:r.
a.
is not shutdown shall consist of:
1.
One licensed Reactor Operator in the reac:or centrol room.
2.
One other person in the reac:or room or reactor control room certified by the Reactor Supervisor as qualified to activate manual scram and initiate emergency procedures.
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 contac:ed by telephone and who can arrive at the reactor facility within 30 minutes.
b.
A licensed Senior Reactor Operator shall supervise all reactor raintenance or sodification which could af fect the reac:ivi:v of the reactor.
~
2263 143 1 Y
6.2 Staff Qualifications Ihe Director of the Nuclear Sngineerug L1boratorf, the Reactor Supervisor, licensed Reactor Operators, and technicians perfoming reactor maintenance shall met the :::inimm qualificaticns set forth in ANS 15.4. Reactor
- afety Ccmittee meders shall have a mini
- m.ra of five (5) years experience in their profession or a baccalaureate degree and two (2) years of professicnal experience. Reactor Safety cemittee mders will generally be University faculty meters with censiderable experience in their area of expertise. The Radiation Safety Officer shall have a baccalaureate degree in biological or physical science and have at least tw (2) years experience in health physics.
6.3 Training The Ditec or of the Nuclear Engineering Laboratorf shall be responsible for directing training as set forth in.CS 15.4, " Standards for Selecticn and Truanmg of Persennel for Research Reactors". All licensed reactor operators shall participate in requalificaticn training as set forth in 10 CFR 55.
6.4 Reactor Safety Comit t ee 6.4.1 Meetings and Ouorum Reactor Safety Comittee shall met as often as deemed necessarf bv the Reactor Safety Comittee Chaira.an who is the Reactor Administrator but shall met at least once each calendar year. A quorum for the conduct of official businers shall be the chaiman, or his designated alternate, and wo (2) cther regular meders. At no tim shall the operating organization comprise a voting =ajority of the meers at any Reactor Safety cx:mittee meeting.
6.4.2 Reviews The Reactor Safety Cec =ittee shall review:
a.
Safety evaluations for changes to precedures, equipment or systems, and tests or egeriments,,c::nducted without Nuclear Regulatorf Ccmissicn approval under the provision of 10 CFR 50.59, to verify that such acticas do not ccnstitute an unreviewed safety question.
b.
Pwposed changes to procedures, equipment or systems that change the original intent or use, and are non-conservative, or those that involve an unreviewed safety questien as defined in 10 CFR 50.59.
2263
- 44 I
c.
Proposed tests or experiments which are significantly different from previously approved tests or experiments, or those that involve an tnreviewed safety question as defined i1.'0 CF2 30.3!.
d.
Preposed changes in Technical Specifications or licenses.
Violatiens of applicable statutes, codes, regulations, orders, e.
Technical Specifications, license requirements, or of intemal procedures or instructions having nuclear safety si;nificance.
f.
Significant operating abnomalities or deviations from normal and expected perfomance of facility equipment that affect nuclear safety.
g.
Reportable occurrences.
h.
Audit reports.
6.4.3. Audits Audits of facility activities shall be perfomed under the cognizance of the Reactor Safety Comittee but in no case by the persorr.el responsible for the item audited. These audits shall examine the cperating records and encompass but shall not be Hmited to the following:
a.
The confomance of the facility cperatica to the Te _.nical Specific-aticas and applicable license c::nditions, at least annually.
b.
The Facility Emergenc/ plan and i=ple=enting pitcedures, at least everf reo years.
c.
The Facility Security Plan and i=plementing procedures, at least everf reo years.
6.4.4 Authority The Reactor Safety Cemittee shall report to the Vice President and shall advise the Director of the Nuclear Engineering Laborst::re en those areas of respcnsibility outlined in sectica 6.1.
of these Tech.ical Specifications.
6.4.5 Minutes of the Reactor Safety Cormittee The Reactor.thainistrator shall direct the preparatien, =sintenance, and districution of ninutes of its activities. These minutes shall include a str=nrf of all =eetings, acticns taken, audits, and reviews.
2263 ;45 6.5 Aporovals The procedure for obtaining approval for any change, modification, or procedure which requires approval of the Reactor Safety Committee shall be as follows:
a.
The Reactor Supervisor shall prepare the proposal for review and approval by the Director of the Nuclear Engineering Laboratory, b.
The Director of the Nuclear Engineering Laboratory 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 Safaty Committee can approve the proposal by majority vote.
6.6 Procedures There shall be written procedures that ecver the fo11cwing activities:
a.
Startup, cperaticn, and shutdcun of the reactor, b.
Fuel move =ent and changes to the ccre and experments that cculd affect reactivity.
c.
Ccnduct of i.%:icns and exper ments that cculd affect the cperaticn or safety of the reacter.
d.
Preventive er u.sdve maintenance which c::uld affect the safety of the reactor.
Surve"w, testing, and "T** ratica of inst::=ents, c=ocnents, e.
c and systes as specified in sectica 4.0 cf these Tec.W. cal Specifica-&-,
f.
Implewitaticn of the Security Plan and ~..wrgency Plan.
The above listed procedures shall be approved by the Director of the Nuclear Engineering Laboratory and the Reactor Safety Committee.
Temporary procedures which do not change the Ment of previously approved procedures and which do not involve any unreviewed safety question may be employed on approval by the Reactor Supervisor or Director of the Nuclear Engineering Laboratory.
2263 '46 6.7 Experiments a.
prior to initiating any new reactor experiment an experimental procedure shall be prepared by the Reactor Supervisor and reviewed and approved by the Director of the Nuclear Engineering Laboratory and the Reactor Safety Comittee.
b.
Approved experiments shall only be performed under the cognizance of the Director of the Nuclear Engineering Laboratory and the Reactor Supervisor.
6.8 Safetv Limit Violation ne fo11cw=g ac icns shan be aken it. the even a Safety !.i=i: is nelated:
Le reac.::: vin be shut dcwn i::::erla:ely and reactor cperaticn v21 a.
not be resu:ed W.bcut althori:aden by the Nuclear Regula cry Cct=.issica (h~.t").
b.
The Safety '"t violaticn shan be recorted to de awmriate NRC Regicnal Office of Inspecticn and :nfoi---m:, the Dih.i:r of de NRR, ard 9.e Reac:cr Safety Comittee not later than the next work day, c.
A Safety i M : Violaticn Recc: shall be prepared for review by the React 0r Safety Committee. This report shall describe the applicable
&Mances preceding the violatica, the effects of de violaticn t.",cn f3ed 147f C3::pcnents, systfr.s c str*Mes, and C0rreC ive aC:icn to prevent recurrence.
d.
The Safety 7i"4: Violatien Repc-s:.all be submitted to de NRC, and Reactor Safety Co=mittee within 14 days of the violation.
6.9 Reporting Requirements In additicn to the applicable reporting requirements of Title 10, Code of Federal Regulaticus, the folicwing reports shall be submitted to the Director of the.w w g iate NRC Regicnal Office.
6.9.1 Annual Operating Report Teutine mual cperating reports shall be submitted no later than thirty (30) days foncwing the end of the operating year. Sach annual report shan include a stmaary of the folicwing activities oc:nzrring during the cperating year:
a.
Facility "rdiFcaticns.
b.
Results of majcr surveillm e-tests and inspecticns.
Correc ive maintee perfcrmi.
c.
2263 ;47
_u-
d.
&_ergy peduced by the reac:cr in watt-hcurs.
e.
Uhschtxhlled shutdcWns.
f.
Reac cr Safer / Coc:::littee action pertinent to the facility.
g.
Any activities which require reperting per 10 CFR 50.59.
h.
Any w i 41e a ; :=< as defined in sec:scn 6.9.2 of these T- % ic=1 Speem ent:.cns.
- 6. 9.2 Reportable Occurrences Reportable oc h m, b cD'A4ng causes, probable ccusequences, ccrrec:ive acticns and :nessures to prevent recLw, shall be reported to the NRC.
a.
Prca:ct Notificatien With Written Folicwue. The types of events listed sna.11.oe reported as expeca.t:.cusly as pessible by telephone and telegraph to de Director of the 4epw gilate NRC Regicnal Office, or his desig::ated representative no later than de first work day folicwing the event, with a written folicwup report within em week.s. Irhtien provided shall centain narrative :caterial to provide c:z:plete explanntien of the circ =::: stances surreurdig de event.
(1) Failure of de reactor protec-J.cn system subject to limiting safety system settings to initiate the required protective function by de time a :neni:cred parameter reached the setpoint specified as the limiting safety system setting in the Technical specificaticns.
(2) Operatica of the reactor when any para =eter or operatica subject to a limiting c::nditicn is less c::nservative than the limiting ccaditica for cperatica established in the lec%ical specificaricns.
(3) Abncr:nal.degrad=*4cn disccvered in a fissica product barrier.
(4) Reactivity h=1*e====1tes imrolving:
(a) disagat betwen expected and actual critical positicas of.pp4wdmately 0.3% ak/k; (b) W h g excess reactivir/ limit; (c) shutdcwn :nargin less c::nse:va:ive than specified in Tx.hnical specificaticns; 2263 348
=,
l (5) Failure or =T#ed:n of cne (cr :nore) c
,-mt (s) which p -i., cr c::uld prevet, by itself, the St1#411mt of de fee-dennt requ:t.9.:!ments of system (s) used to ccpe with accideits analyzed in the Safety Analysis Report.
(6) Perscanel errer er gh.a1 f and~ m y diich prevents, or c=uld prevent, by itself, the B21#ii' ant of de f.:ncticnni requ:Lrements of systems required to C pe with accidents analyzed in the Safety Analysis Recort.
(7) ". w u dis e wi d in the transi m t cr er" dant analyses er in de
- nethods used fcr such analyses as described in de Saferf Analysis Recrt er in the bases fer de Tec!nical Specificaticns dat have pe:mitted reacter cperatica in 1 =anner less c:nsertative than
= = = d in the analyses.
(8)
Ps.%-c, cf st::uctures, systas, er compcnents that requires rM4 = 1 ac icn or e..ctive measures to prevent cperatica in a
- man::er less c:nsertative than assc. sed in de =e"da,t analyses in the Safety Analysis Re. pert er Technical spee"icaricn bases; or discoverf uring plant life of canditicns not spee4 #4cally considered d
in the Safety Analysis Report cr TWim1 Specificaticns that require r=~44=1 actien or correc.ive :easures to prevent the existence or develegnent of an unsafe c=ndi-de 6.10 Record Retention 6.10.1 Records to be retained for the life of the Aci?irf:
a.
Annual reports.
b.
Records of centrolled or ure ntrolled release of radicac-Jte effluents to the envircr=ent.
c.
Puel inventories and fuel transfers.
d.
Operating legs.
e.
Maintenance logs.
f.
Updated drawngs of the reactor Acility.
g.
Perscnnel dosi: metry rec rds en file vid de Radiation Safety Officer.
h.
Mimwas of the Reactor Safety comittee meetings.
6.10.2 Reccrds to be retained fcr a period of at least three years:
Surta411=ne= ac-Jiities requirsi by Tec.%icsi Spa-"icaticns.
a.
b.
Facility rad 4=* den and c:nted-meden surteys.
6.10.3 Perscnnel r*Q'ali#dcaticn and tr149 49g recortis vill be kept at least cne year after ter:nnatica of ec:picyment.
2263 349