ML20058A179
| ML20058A179 | |
| Person / Time | |
|---|---|
| Site: | Idaho State University |
| Issue date: | 10/24/1978 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Kegel C IDAHO STATE UNIV., POCATELLO, ID |
| Shared Package | |
| ML20058A183 | List: |
| References | |
| NUDOCS 7811130330 | |
| Download: ML20058A179 (53) | |
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UNITED STA TES
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['....h, j NUCLEAR REGULATORY COMMish.ON WASHING TON, D. C. 20555 Mha[/
October 24, 1978
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Docket flo. 3-284 Dr. Charles H. Kegel Academic Vice President Idaho State University Pocatello, Idaho 83201
Dear Dr. Kegel:
Your request of August 10, 1378, for amendment to.Jacility License flo. R-110 for the Idaho Staa University AGrl-201 reactosi-s cur r ently under review. We have noted that your current Technical Specifications (TS) are not of a format consistent with current practice. During our recent telephone conversation with your staff, you agreed to revise your TS to be consistent with the common fomat for the AGil-201 reactor. Accordingly, enclosed for your guidance, are the TS for Catholic University AGrl-201 and Texas A&M University AGil-201M which do follow the common fomat. Also enclosed for your guidance, is a marked-up copy of the proposed Americal fiational Standard AflS 15.18, Standard for Administrative Controls for Research Reactors. As each AGil-201 reactor facility has differences, we expect that some changes will be necessary, particularly in limiting safety system settings and limiting conditions of operation. However, it is anticipated that these will be minimal.
We request that you modify your request for amendment to incorporate the comon AG!1-201 TS and submit it as soon as possible for review.
Sincerely, m
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v Robert W. Reid, Chief Operating Reactors Branch d4 Division of Operating Reactors
Enclosures:
1.
TS for Catholic University 2.
TS for Texas A&M 3.
AtlS 15.18 7811130'
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TABLE OF CONTENTS PAGE 16.1 DEFINITIONS.................................................... 16-1 L'
16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS.............. 16,3 16.2.1 S a fe ty L i m i t s............., 7................................ 16 -3 16.2.2 Lie.iting Safety System Settings..........*................... 16-4 16.3 LIM ITI N G COND IT I ONS FOR OP ERATI ON.............................. 16-5 16.3.1 Reactivity Limits 16-5 E
i 16.3.2 Con t ro l a n d S a f e ty S ys t ems.................................. 16-5 16 3.3 Lim i t a t i en on Exp e ri men ts.................................,. /. 16-8 16.3.4 Shielding......................................:........... 16-8 16.4
$URV E I L LAN C E REQU I REM ENTS...................................... 16-B a 16.4.1 Re a c t i v i ty L i m i t s........................................... 16 - 8 a 16.4.2 Con t ro l and S a fe ty Sys tem................................... 16-9 16.4.3 R e a c t o r S t ru c tu re........................................... 16 - 1 1 i
i 16.5 D E S I GN F EATU RE S................................................ 16-1 1 16.5.1 Reactor..................................................... 16-11 16.5.2 Fuel Storage................................................ 16-12 16.5 3 R e a c t o r Room................................. o.............. 16-12 AD M I N S T RAT I V E CO NT RO L S......................................... 16-12 16.6.
16.6.1 organi:ation................................................ 16-12 B
16.6.2 S t a f f qu a l i f i ca t i o ns........................................ 16-15 16.6.3 Training.................................................... 16-15 16.6.4 Reactor Safety Cenni ttee Reviews., Audits,
16-15 An d A u th o r i ty................................................ 16 - 15 16.6.5 Approvals................................................... 16-17 16.6.6 Op e ra t i ng P roc e d u res........................................ 16-17 16.6.7 Exp e r i me n t s................................................. 16 - 18 16.6.8 S a f e ty L i m i t ' v i o l a t i on...................................... 16-18 16.6.9 Repo rt i ng Req u i remen ts...................................... 16-18 16.6.10 Re co r d Re t e n t i o n............................................ 16 -2 2 ed i
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16.1.0 DEFINITIONS The tern:s safety Limit (st), Limiting safety System Setting (LSSS),
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and Limiting conditions for Operation (LCO) are as defined in :0.36 l
of 10 CFR part 50.
Channel Calibration - A channel calibration is an adjustment of the cnannel sucn tnat its output responds, within acceptable range 16.1.1 and accuracy, to known values of the parameter which the channel Calibration shall encompass the entire channel, includin; measures.
equipment, acutation, alar n, or trip.
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P 16.1.2 Channel Check - A channel chec'k is a qualitative verification of acceptaole cerformance by observation of enannel benavior. This verification may include comoarison of the enannel with other independent channels or methods measuring the same variable.
16.1.3 Channel Test - A channel test is the introduction of a signal into the channel to verify that it is operable.
16.1.4 Exoeriment_ -
An experiment is any_of the following:
a.
(1) An activity utilizing the reactor system or its components
- or the neutrons or radiation generated therein; I
(2) An evaluation or test of a reactor system operational, surveillance, or maintenance technique; (3) An experimental or testing activity which is conducted with-in the confinement or containment system of the reactor; or (4) The material content of any of the foregoing, including structural components, encapsulation or confining bouncaries,
- i and contained fluids or solids.
- b. ' Secured Exceriment - Any experiment, or component of an experiment is deemec to ce 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 nor al to the operating environment of the experiment or wnich might arise as a result of credible mal-functions.
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c.
Unsecured Excerient - Any exoeriment, or ccm:enent of an experiment
'is deemed to ce unsecured wnenever it is not secured as definec in 16.1.4.b. above. Moving parts of experi=ents are deemed to be unsecured when they are in motion.
- d. Movable Exceriment - A movable experiment is one which may be insertec, re:nvec, or manipulated wnile the reactor is critical.
- e. Removable Etoerimnt - A re.anvable ex:eriment is aily experiment, experimental racility, or component of an experiment, other than a per anently attach ?d appurtenance to the reactor system, wnich carl reasonably be anticipated to.be moved one or incre times during the life of the reactor.
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I' 16.1.5 Exoerimental Facilities - Experimental facilities are those portions of the reactor assercly that are used for the introduction of experi-ments into or adjacent to the reactor core region or allow beams of radiation to exist from the reactor shielding. Experiment.al facilities i
shall include the thermal column, glory hole, and access ports.
16.1.6 Explosive Material - Explosive material is any solid or liquid wnicn is categorized as a Severe, Dangerous, or Very Dangerous
- Explosion Hazard in " Dangerous Properties of Industrial Materials" by N.1. Sax. Third Ed. (1968), or is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Association in its publication 704-M,1966, " Identification System for Fire Hazards of Materials " also enumerated in the
" Handbook for taboratory Safety" 2nd Ed. (1971) published by The Chemical Rubber Co.
.. a.
16.1.7 Measurino Channel _ - A measuring channel is the combination of sensor, lines, amniitters, and outout cevices wnich are connected for the puroose of measuring or resconcing to the value of a process variaele.
16.1.8 Ooerable - Ooerable means a comnonent or system is capable of nerforming
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it's' liitended function in its normal manner.
i 16.1.9 Ooeretine - Ooerating means a component or system is performing its intenceo runction in its normal manner.
Potential Reactivity Worth - The potential reactivity worth of 16.1.10 an experiment is tne 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 or configuration, 16.1.11 Reacter Comoenent - A reactor comoonent is any acoaratus, device, or material tnat is a normal part of the reactor assemoly.
l'6.1.12 Reactor Oceration - Reactor operation is any condition wherein the reactor is not snutdown.
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3 Reactor Safetv Svsten - The reactor safety system is that ccebination
- t of safety enannels and associated circuitry which forms the autcmatic 16.1.13 protective system for the reactor or provides information which requires manual protective action be initiated.
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jf 16.1.14 Reactor Shuteewn - The ' reactor shall be considered shutdown whenever
- a. either:
1.
All safety and' control rods are fully withdrawn frem the core, or
_3 is; 2.
The core fuse melts resulting in secaration of the core.
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and:
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- b. The reactor console key switch is in the "off" position and,
the key is removed from the console and under the control of a licensed cperator.
Safety Channel - A safety channel is a measuring channel in the reactor 16.1.15 safety system.
I 16.1.16 Static Reactivity 'Jorth - The static reactivity worth of an experi-ment is tne aosciute value of the reactivity change which is measurable by calibrated control or regulating rod comparison methods between two defined terminal positions or configuraticns of the experiment. For removable exoeriments, the terminal positions are fully removed from the reactor and fully inserted or installed,
in the normal functioning or intended position.
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16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l
16.2.1 Sa fe ty Limits Applicability This specification applies,to the maximum steady state power level and maximum core temperature during steady state or transient op-eration.
Objective 1
To assure that the integrity of the fuel material is maintained and all fission fragments are retained in the core matrix.
Specification a.
The reactor power level shall not exceed 100 watts.
b.
The maximum core temperature shall not exceed 200*C during either steady state or transient operation.
Bases The polyethylene core material does not melt below 200*C and is 'ex-pected to maintain its integrity and retain essentially all of the fission fragments at tem:gratures below 200*C, The Hazards Sunta ry Report dated February 1962 submitted on Docket F-15 by Aeroiet-General Nucleonics (AGN) calculated a steady state core average temperature rise of 0.044C/ watt. Therefore, a steady state power level of 100 watts would result in an average core temperature rise of a4CC. The corresponding maximum core temperature would be belew 200*C thus assuring integrity of the core and retention of fission frag =ents.
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Speci f i ca t ion c.
The reactor shield tank water temperature shall be maintained above 10*C, and the water level in the tank shall not be more than 10 inches bel the highest point on the manhole opening.~
Bases Low reactor shield tank water temperature may result in freezing
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of the water. The resultant ex ansien due to freezing of the water may damage the shield tank and other reactor components.
This condition would degrade core containment and snielding capability. A safety limit of 10*
provides a margin for confidence that the reactor will not be operated with frozen shielding water.
The shield tank water level of 10 inches below the highest point en the
=anhole opening provides adequate biological shielding during reactor operation.
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,l 16.2.2 Limiting Safety Systems Settings Applicability This specification applies to the parts of the reactor safety system which will limi t maximum power and core temperature.
5 Objective e
To assure that automatic protective action is initiated to prevent a safety limit from being exceeded.
Specification i
a.
The safety channels shall initiate a reactor scram at the fol-lowing limiting safety system settings:
Channel Condition LSSS Nuclear Safety fl High Power
< 0.2 watts Nuclear Safety #2 High Power 30.2 watts Nuclear Safety 13 High Power
< 0.2 watts b.
The core thermal fuse shall melt when heated to a temperature i'.
,of about 120cc resulting,in core separation and a reactivity
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't 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 maximum available excess reactivity in the reactor is less than one dollar the reactor cannot become prompt critical and the corresponding shortest possible perled is greater than 200 milli-seconds. The high power LSSS of 0.2 watts in conjunction 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 result of the most severe credible transient.
In the event of failure of the reactor to scram, the sel f-limiting characteristics due to the high negative temcerature coefficient.
and the melting of the thermal fuse at a temperature, at about 1200C will assure safe shutdown without exceeding a coref emperature of t
200*C.
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_ LIMITING CONDITIONS FOR OPERAil0N 16.3.1 Reactivity Limits Applicability This specification applies to the reactivity condition of 'the reactor and the reactivity worths of control rods and experiments, e
Objective To assure that the reactor can be shut down at all times and that the safety limits will not be exceeded.
S peci fication a.
The available excess reactivity with all control and. safety rt.ds fully inserted and including the potential reactivity w.irth of all experiments shall not exceed 0.65% ak/k re-f-renced to 20'C.
b.
The shutdown margin with the most reactive safety or control rod fully inserted shall be at least 2% Ak/k.
c.
The reactivity worth of the control and safety rods shall
.i ensure 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-leds of sufficient length so that the reactor protection system and/or operator action will be able to shut the reactor down wi th-out exceeding any safety limits. The shutdown margin and centrol and safety rod reactivity limitations assure that the reactor can be brought and maintained succritical if the highest reactivity rod falls to scram and remains in its most reactive position.
16.3.2 Control and Sa fety Systems Applicabilley These specifications apply to the rectter control and safety systems.
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j Objective i
To speci fy lowes t acceptable level of performance, Ins t rumen t se t po in ts.
l and the minimum number of operable components for.the reactor control and I-sa fety sys tems.
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Speci fica tion The reactor shall not be made critical 'unless the following specifications j
are met:
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j a.
The total scram withdrawal time of the safety rods and coarse control rod shall be less than 200 milliseconds.
G b.
The maximum reactivity addition rate for each rod shall not exceed 0.04% Ak/k/sec.
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c.
The sa fety rods and coarse control rod shall be Interlocked such l
that:
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Reactor startup cannot commence unless both safety rods and coarse control rod are fully withdrawn from the core.
2.
Only one safety rod can be Inserted at a time.
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3 The coarse control red cannot be Inserted unless both safety rods t
are fully inserted.
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All reactor safety system instrumentation shall be operable in accord-
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ance with Table 16-1 with the following allowable exceptions:
a 1.
Nuclear Safety Channel No.1 may be bypassed for a period not to l'
exceed twelve consecutive hours provided Nuclear Safety Channel Nos. 2 and 3 are verifled to be operable.
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2.
Nuclear Safety Channel No. 3 may be bypassed for a period not to exceed 12 consecutive hours provided Nuclear Safety Channel Nos.
I and 2 are verified to be operable.
3.
The seismic displacement scram may be cut of service during reactor i
' operation for no more tnan 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 3-month period.
3.
A loss of electric pcwer shall cause the reactor to scram.
Bases I
The specificatien on scram reactivity in conjunction with the safety system Instrumentation and set points assure safe reactor shutcown during the most i
. severe foreseeable transients.
The limitatiens en reactivity addition rates i
Silow caly relatively slow increases of reactivity so that amole titre will be available for manual or. automatic scram during any cperating conditions.
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.o r-s Interlocks on control and safety rods assure an orderly approach to criticality
- and an adequate shutdown capability.
.. 3 The neutron detector channels (nuclear safety channels 1 through 3 assure that reactor power levels are adequately monitored during reactor star, tup and operation. Requirements on minimum neutron levels will prevent reactor startup unless channels are operable and responding, and will cause a scram in the event of instrumentation failurg. The power level scrams initiate redundant automatic protective action at power levels low enough to assure safe shutdown without exceeding any safety limits. The period scram conser-vatively 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 reactivi ty additions. In order. to provide somei time to correct channel defects, a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed for operation with either Nuclear I
Safety Channels Nos. I or 3 bypassed i f the remalning two channels are verified to be operable. Although some redundancy in the reactor protection sy' stem is lost during the limited time Interval, all scram functions and monitoring capabilities are still available.
-The AC1-20l's negative temperature coefficient of reactivity causes a reactivity increase with decreasing core temperature. Tne shield water temperature safety channel will prevent reactor operation at temperatures below 15'c thereby limiting potential reactivity additions associated with temperature decreases.
Water in the shield tank is an important component of the reactor shield, and j
operation without the water may produce excessive radiation levels. The shield
._ tank water level safety channel will prevent reactor operation without adequate water levels in the snield tank.
j The reactor is designed to withstand 0.69 accelerations and 6 cm displacements.
A seismic instrument causes a reactor scram whenever the instrument receives a horizontal acceleration that causes a hori: ental displacement of 1/16 inch or g rea te r.
The seismic displacement safety channel assures that the reactor will be scrammed and brought to a subcritical configuration during any seismic disturbance that may cause damage to the reactor or its components. Due to the low probability of earthquake damage, the seismic instrument can be out of service during reactor operation for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 3$cnth period.
The manual scram allows the operator to manually shut down the reactor if an unsafe or otheralse 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 4
i larnediate shutdown in case of a pcwer outage.
A radiation monitor must always be available to coerating personnel to provide an Indication of any abnormally hign radiation levels so that a pp rop ri ate action can be taken to shut the reactor down and assess the hazards to personnel.
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16.3.3 1.lmitations on Exeeriments ApplI cab Il I ty This specification applies to experiments nstalled in the reactor and its experimental facilities.
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Objective To prevent damage to the reactor or excessive release of radlo-a active materials in the e/ent of an experimental failure.
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Speci fication t
a.
Experimen ts containing materials corrosive to reactor compo-nents 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 f rom the experiment will not results in doses in excess of 10% of the equivalent annual doses sta.ed in 10 CFR Part 20 for persons occupying (1) unrestricted areas continuously for two hours starting at time of release or (2) restricted areas during the length of time requi red to evacuate the restricted area, d.
The radioactive material content, including fission products of any doubly encaosulated execriment shall be limited so that the complete release of all gaseous, particulate, or volatile components of the experiment shall not results in exposures in excess of 0.5 Rem whole body or 1.5 Rem thyroid to persens 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 occuoying a restricted area during the length of time required to evacuate the restricted area.
Bases 7ese specifications are Intenced to reduce the Ilkelihood of damage to reactor components aad/or radioactivity releases resulting frem an experimental f ailure and to protect coerating personnel and the public from excessive radiation doses in the event of an experimental failure.
16.3.4 Shieldine Acelleablittv This specification applies to reactor shielding required during reactor operation.
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objective The objective is to protect facility personnel and the public frem radiation exposure.
Speelfication j,
The following shielding requirements shall be fullfilled prior to reactor startup and during reactor operation:
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- a. The reactor shield tank shall be filled with water to a height within 10 inches of the highest point on the manhole opening, i,,
- b. The thermal column shall be filled with watgr or graphite.
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Bases r
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The inher/nt reactor shielding in conjunction with designated re-stricted radiation areas is designed to limit radiation doses to facility persennel and to the pub!!c to a level below 10CFR20 Timits under operating conditions, and to a level below Criterion 19, Ap-pendix A,10CFR50 recc.mendations under accident conditions.
16.4 SURVEILLANCE REQUIREMENTS l
Actions specified in this section are not required to be performed q
if during the specified surveillance period the reactor has not been brought critical or is maintained in a shutdewn condition extending beycnd the specified surveill1nce period. Hwever, the surveillance requirements must be fulfilled prior to subsequent startup of the
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16.4.1 g,,cggyggy tgm;g, j
Aeolicability This specification applies to the surveillance requirements for reactivity limits.
Objective
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j To assure that reactivity limits for Specification 16.J.1 are not a
exceeded.
Soeelfication
- a. Safety and control red reactivity worths shall be measured annually, but at intervals not to exceed 16 months.
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i L. Total excess reactivity and shutdown margin shall be determined annually, but at intervals not to exceed 16 months.
- c. The reactivity worth of an experiment shall be estimated or measured, as appropriate, before or during the first startup subsequent to the experiment's insertion.
Bases u
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 l
margin or total excess reactivity. The shutdown margin and total excess reactivity are determined to assure that the reactor can al-ways Hbe shutdown safely with one rod not functioning and that the maximum possible reactivity insertion will not result in periods shorter than those that can be terminated adequately by either op6rator or automatic action. Based on experience with AGN reactors,
significant changes in reactivity or rod worth.are not expected within a 16 month period.
16.4.2 control and Safety System ApoIIcability t:
This specification applies to the surveillance requirements of the reactor control and safety systems.
Objective To assure that the reactor control and safety systems are operable l
as required by Specification 16.3.2 Specification
- a. Safety and control rod scram time and insertion rates shall be measured annually, but at intervals not to exceed 16 months.
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- b. Safety and control rods and drives shall be inspected for deterioration at intervals not to exceed 2 years.
- c. A channel test of the following safety channels shall be perfor=ed prior to each day's operation or prior to eacn operation extending f more than one day:
a Nuclear Safety #1, #2, and !3 Manual scram Area radiation monitor
- d. A' channel test of the following safety channels shall be per-formed quarterly:
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Shield water temperature Shield water level Seismic displacement e.
A channel check of the folicwing safety channels shall be performed daily or whenever the reactor is in cperation:
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Nuclear Safety #1, #2, and #3 Area radiation moni tor r
f.
Daily, prior to startt/p, each of the two safety rods shall be inserted and scraced to verify operabIIIty.
g.
The period, count rate, and power level' measuring channels shall i
be calibrated and set points verified annually, but at intervals not to exceed 16 menths.
t h.
The shield tank water level and temperature and seismic displacement safety channels shall be calibrated by perturbing, the sensing element to the appropriate set 'pcint.
These calibrations shall be performed annually, but at Intervals not to exceed 16 months.
l 1.
The radiation monitoring Instrumentation shall be calibrated annually, but at intervals not to exceed 16 months.
Bases The channel tests and checks required daily or before each startup will assure that the safety channels and scram functicns are operable.
Based en operating experience with reactors of this type, the annual scram measurements, channel calibrations and set point verificatiens are of sufficient frequency to assure, with a high degree of confi-e i
dence, that the safety system setting will be within acceptable i
drift tolerance for operation. The periodic surveillance and calibration of the radiation =eni toring instrumentation will assure that the radiation monitoring equipment is operab!e during reactor ope ra tion.
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,j, 36,4.3 Reactor structure
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Appitcabilltv This specification applies to surveillance requirements for reactor components other than control and safety rods.
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,,L Objective The objective is to assure,fntegrity of the reactor structures.
Specification a
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a.
The shield tank shall be visually inspected every two years.
If apparent excessive corrosion or other damage is observed, corrective measures shall be taken prior to subsequent reac-I tor operatien.
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b.
Visual inspection for. water leakage from the shield tank shall i
be performed every year. Leakage shall be corrected prior to.
.L subsequent reactor operation.
Bases
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1 Sased on experience with reactors of this type, the frequency of Inspection and leak test requirements of the shield tank will assure capability for radiation protection during reactor operation.
16.5.0 DESIGN FEATURES 16.5.1 Reactor 1
t a.
The reactor core, including control and safety rods, contains approximately 675 grams of U-235 In the form of 19.9% enriched i
U0 dispersed in approximately 11 kilograms of polethylene.
2 The lower'section of the core is supported by an aluminum ro d
, hanging from a fuse link. The fuse melts at fuse temperatures of about-120*C causing the lower core section to fall away frem tne upper section reducing reactivity by at least 5% ak/k. Sufficient clearance between core and reflector is provided to insure free I
fall of the bottom half of the core during the most severe trans-lent.
3
- b. 7The core is surrounded by a 20 cm thick high density (1.75 gm/cm )
graphite reflector followed by a 10 cm thick lead gamma shield.
The core and part of the grachite reflector are sealed in a fluid-tight aluminum core tank designed to contain any fission gases that might leak from the core.
c.
The core, reflector, and lead shielding are enclosed in and suc-ported 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 wi th grachite.
d.
The 6 foot diameter, fluid-tight shield tank is filled with water 3 6-l i..-
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constituting a 55 cm thick fast neutron shield. The fast neutron shield is formed by filling the tank with 1000 gallons of water..
The complete reactor shield shall limit doses to operating per-sonnel in restricted and unrestricted areas to levels less than permitted in 10 CF 20 under operating conditions.
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e.
Two safety rods and one control rod (Identical in size) contain less than 15 grams of U-235 each in the same form as the core ma te ri al. These rods are lif ted into the core by electromagnets, r
driven by reversible DC, motors through lead screw assemblies.
l Deenergi:Ing the magnets causes a spring-driven, gravity-assisted i
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 contains unfueled polyethylene.
4 16.5.2 Fuel Storace Fuel,' Including fueled experiments ar$d fuel devices, not In the reactor shall be stored in a locked safe in the reactor room. The* storage array shall be such that K is no greater than 0.8 for all conditions ff of moderation and reflect!8n.
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'16.5 3 Reactor Room a.
The reactor room houses the reactor assembly and accessories I
required for its operation and maintenance.
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b.
The reactor room is a separate room in the Pangborn Engineering building, constructed with adequate shielding and other radiation protective features to limit deses in restricted and unrestricted areas to levels no greater than permitted by 10 CFR 20, under normal operating conditions, and to a level below criterien 19, Appendix A, 10 CFR 50 recommendations under accident ccndi tions.
i c.
The reactor room doors are self-closing and-locking.
16.6 ADMINSTRATIVE CONTROLS 16.6.1 ORGANIZATION The administrative organization for control of the rea-tor f acili ty and its operation shall be as set forth in Figure 16-1 attached hereto. The authorities and responsibilities set forth belo.< are designed to comply with the intent and requirements for administrative controls of the reactor facility as set forth by the Nuclear Regulatory Commission.
16.6.1.1 PRESIDENT The President is the chief Administrative Officer res=on-i sible for the University and in whose name the application for licens-ing is made.
16.6.1.2 EXECUTIVE VICE PRESIDENT AND PROVOST. The executive Vice President and Provost reports to tne Presiaant and is responsible for the in-ternal administration of the University, both academic and non-academic.
In this capacity he represents the President in all matters pertaining to. the reactor except in those cases' of health and safety for which the Radiation Safety Ccemittee has authority.
16-12 i
l i
i
A g.
16.6.1.3 DEAN OF ENGINEERING AND ARCHITECTURE. The Dean o'f Engineering and t
~
Architecture is responsible to the Executive Vice President and Provost for all departments in the School of Engineering and Archi-j tecture.
In this capacity he shall have final authority and ulti-mate responsibility for the reactor facility and, within the !!mita-tions set forth by the facility license, make final policy decisions
]
with respect to reactor operation; appoint the Reactor Adml.nistrator, l
4,
'with the advice of the Chairman of the Department of Mechanical Engineering; be advised on all matters concerning reactor safety by j
the Reactor Safety Conrni tt,ee.
16.6.1.4 CHAIRMAN, DEPARTMENT OF MECHANICAL ENGINEERING. The Chairman of the Department of Meenanical Engineering is responsible to the Dean of Engineering and Architecture for the routine functioning of the rea-tor facility. He shall advise the Dean of Engineering and Architec-ture concerning the appointment of the Reactor Administrator; and I
appoint the members of the Reactor Safety Conunittee.
16.6.1 5 REACTOR ADMINISTRATOR The Reactor Administrator is responsible to the Chai rman of Meenanical Engineering Departme'nt for the daily acmin-istration of the reactor facility.
In this capacity, he shall, within the policies set forth by the Chairman and the facility license, pre-pare all regulations for the facill:y, review and approve all pro-cedures, seek approval of all procedures and proposals for changes and experiments frem the Radiation Safety Conrnittee, and be responsible for the health and safety of all personnel In the reac-tor facility. He shall be responsible for the official files of the facility, including storage of such prescribed logs and records of the facility which are no longer requi red for current operaticns. Prior to periods of scheduled absence, he shall designate an alternate and notify the Chal rman.
~
16.6.1.6 REACTOR SUPER'llSCR The R$ shall be a IIcensed SRO. He shall be responsible for tne preparation, promulgation, and enforcement of administrative centrols including all rules, regulations, instructions and cperating procedures to ensure 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 coer-ation of the ree c:or; schedule reactor operations and maintenance; he
, responsible for the preparation and authentication of all prescribed logs and operating records of the facill:y; authorize all experiments,
procedures, and changes thereto whlen have firs t received approval of the Reactor Safety Committee, the Radiation Safety Cem-mittee, and the Reactor Administrator, and be responsible for the preparation of all Instructional manuals and experimental procedures involving use of the reactor. The Reactor Sucervisor shall advise the Reactor Administrator of any scheculed perleds "of absence.
16.6.1.7 REACTOR OPERATORS Reactor Oceraters shall be responsible for the manipulation of the reactor controls, menitoring of instrumenta:Icn, operation of reactor related equicment, anc maintenance of complets and current records during operation of the faci.lity. The Reactor Operator shall be in direct enarge of the reactoi <ensole at all times during reactor operation and wnen the reactor is not' secured and cen-form to the rules, instructions, and procedures es:ablished by :ne Rea.ctor Administrator and Reactor Supervisor for cperation of the 16 -13 a '
1 e
e-e
- -e e
~, _,.,
ay r:act:r cnd tha perfarmtnce of exp3riments.
(
.5 18.6.1.'8 REACTOR SAFETY COMMITTEE The Reactor Safety Comittee (RSC) s' hall 1
be responsiole for incependent reviews and audits of facility oper-j ations to insure that the reactor is operated in a safe and compe- -
f tent manner and advise the Reactor Administrator in all matters related to reactor safety and personnel safety.
k 16.6.1.S RADI ATION SAFETY COMITTEE. The Chairman of the Radiation Safety
.)
Cormiittee snail be appointed by the Executive Vice President and i
Provost (EVPP). Additional members of the Co.mittee shall be appoint-
]*
ed by the EVPP with the advice of the Chairman. The Committee shall advise the EVPP in all matters concerning radiological aspects of the health and safety of personnel who might be exposed to radiation produced by Universi ty owned and/or operated sources or equipment.
The comittee shall review, approve, and premulgate a Radiation l
Safety Manual for the Universi ty.
The Committee shall be informed i
of all occurrences related to radiation health and safety and reactor i
safety which are reportable to any authorities outside the University
}
the EVPP with regard to such matters.
and advise the EVPP of such occurrences and make recommendations to j
16.6.1.10 RADIATION SAFETY OFFICER. The Radiation Safety Officer (RS0) shall I
be appointed by tne Executive Vice President and Provost, wi th the advice of the Chairman of the Radiation Safety Committee. He is responsible to the Chairman for the day-to-day administration of the radiation safety program and shall serve as the Secretary of the Radiation Safety Comittee and as an ex officio member of the Reactor l
Safety Comi ttee. He shall prepare the University's Radiation Safety I
Hanual and have the authority to enforce the regulations, rules and procedures set forth in the Radiation Safety Manual, suspend the oper-
.ation and use of radiation producing devices when their use is in violation of these rules, and secure such sources of radiation until corrective action is taken. He shall also have the autnority to dis-approve the ac:;uisition of radiation producing sources until satis-factory evidence is presented to ensure the safe storage and use of these facilities. The Radiation Safety Officer is also resoonsible i
for preparing, for the signature of the Chairman of the Radiation Safety Committee, reports of all reportable occurrences to the a:p.o-priate regulatory agency and for ensuring that appropriate follow-up action is taken.
3 16.6.1.11 OPERATING STAFF a.
The minimum staff durPng any time in which the reactor is not shutdcwn (1) shall consist of:
1.
A licensed reactor operator in the control room (RO).
2.
A second person present at the facility complex capable of carrying out any prescribed written instructions and instructions of operators and to summon help in the event the licensed operator becomes incapacitated.
Unexpected absence for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or less for this person is acceptable providing action is taken to l
obtain a replacement.W
~~
(1) Reactor Shutd6wft is-defined in Section 16.1 l
(1)
See p a ge' 16-15 16-14 a
l
.. ~
=
.,. a 44cuneau aantor asaccer up,trcccr unus, tr en3ra era k.
moro thrn ens licens d operators sc,ths fccility, shall.
~ t
, be racdily availablo en es11.
If th:re is caly one licensed operator, he shall be an SRO and shall also i
fulfill sub paragranh 1 above.
l a.
"Readily Available on Call: is defined to mean
.an indivi.!ual who (1) has been specifically designated and whose designation is known to
)
, the RO, (2) keeps the RO infor=ed of where he may be rapidly contacted by phone, and (3) is.
capable of getting,to the reactor facility within a reasonable time period under normal conditions (e.gt, 30 minutes or within a 20 mile radius).
t 16.6.2 STAFF Q'UAllFICATIONS The Reactor Administrator, Reactor Supervisor, Reactor Operator, and any Technicians performing work on the reactor l I
shall meet the minimum qualif.ications set forth in ANS 15.4 "Standardi-
~
~
act o,rs".
l for Selection and Training of Personnel for Research Re The qualifications of the Reactor Safety Committee members shall be five (5) years of professional experience in the field represented by the member or a baccalaureate degree plus at least two years.
experience. Generally, these ceceittee members will be made uo of
{
University facul y; but outside ex:erience may be sought in areas t
where additional experience is considered necessary by the Chairman o;;
Hechanical Engineering Department.
In this case, a baccalaureate degree plus five (5) years experience will be required.
l i
16.6.3 TRAINING The Reactor Administrator shall be responsible for the i
faci li ty retraining program.
l 16.6.4 REACTOR SAFETY COMMITTEE REVI EVS, AUDITS. AND AUTHORITY 16.6.4.1 MEETINGS AND OUORUM The Reactor Safety Committee shall mee: as necessa y but at least twice each calendar year. A quorum for 1
review shall consist of the chairman, or his designated alternate, and two other members, or alternate members as long as a majority i
of these present shall be regular members, and shall include repre-sentation in reactor cperations and radiation protection. However, the operating staff shall not be a voting majority.
16.6.4.2 ALTERNATES Alternate members may be appointed by the Chairman of Mecnanical Engineering Department to serve on a temcoran easis; each appointment shall be in writing. No more than two aliernates shall par:Icipate on a voting basis in Reactor Safety Ccemit:ee i
activities at any one time.
16.6.4.3 REVIEVS The Reactor Safety Committee shall review:
a.
Safety evaluations for (1) changes to procedures, equipment or systems and (2) tests or exoeriments, condue:ed without Nuclear f
Regulatory Cer:r.ission appreval under the provision of Sectica 3
30.59, 10 CFR, to verify that such actions do not cens ti tute an unreviewed safety question.
(2) Persons capable of performing emergency procedures shall be authori:ed by the Reactor Supervisor; be familiar with the Emergency Plan and capable of Initiating the Evacuation Alarm; knew the locations and be capable of using i emergency and radiation survey equipment.
l 16-15
,r;
.q.
.a b.
Proposed changes to procedures, equipment or systems that change W:
the original intent or use, and are non-conservative, or those that involve an unreviewed safety quesilon as defined in See:lon 50.59, IcCra.
c.
Proposed tes ts or experiments which are significantly different
.],,
from previous approved tests or experiments, or those that involve 4,
an unreviewed safety question as defined in Section 50.59, 10 CFR.
d.
Proposed changes in Tec,hnical Specifications or IIcenses.
Violations of applicable statutes, codes, regulations, orders, e.
Technical Specifications, license requirements, or of Internal procedures or instructions having nuclear safety significance.
.. L.
i 8
r f.
Significant cperating abnormalities or deviations from normal and expected performance of facility equipment that affect nuclear safety.
g.
Events which have been reported in writing w'ichin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Nuclear Regulatory Commissicn.
h.
Audit reports.
16.6.4.4 AUDITS Audits of facility activities shall be performed under the cognizance of the Reactor Safety Cornnittee but in no case by the personnel responsible for the item audited.
Individual audits may be performed by one individual who need not be an Icentified Reactor Safety Cc.mi t:ee memeer. These audits shall examine the opera:Ing records and enccmpass:
a.
The conformance of facility opera: Ten to the Technical Specifications and applicable license condi:lons, at least once per 12 months.
b.
The performance, training and qualifications of the entire facility staff, at least once per 24 =onths.
c.
The results of all actions taken to correc: deficier~cies occuring in f acili ty equipment, s tructures,,sys: ems or method
+.
of operation that affec: nuclear safety, at least once per.
calendar year.
d.
The Facility Emergency Plan and implementing procedures, at leas t once per 24 months.
The Facility Security Plan and implementing procedures, at least e.
once per 24 months.
f.
Any other area of facility cperation considered acpropriate by the Reactor Safety Comittee or the Chairman of Mechanical Engin-eering Department.
16.6.4.5 AUTHORITY The Reactor Safety Ccemi::ee shall recor: to the Chai rman l
of Mechanical Engineering Cesar: ment and advise ene Reactor Admints-trator on those areas of respcnsibility specified in see:icns 16.6.I. 3
.and 16.6.4.4
.16-16 l
~' -
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16.6.4.6 RECORDS AND REPORTS OF THE REACTOR SAFETY COMMITTEE The chairman of the Reactor Safety Commi ttee snali prepare, maintain, and distribute' records of its activities as indicated belcw:
a.
Minutes of each Reactor Safety Conrnittee meeting shall be pre-pared, and forwarded to the Chairman of Mechanical Engineering and the Committee members within 30 days following each meeting.
~
b.
Reports of all reviews and audits shall be prepared and forwarded to the Chairman of Mechinical Engineering within 30 days follow-Ing completion of the review or immediately upon completion if corrective action is requi red.
I c.
Reviews of approvals requested by the Reactor Administrator for proposed changes shall be forwarded to the Chairman of Mechanical Engineering upon completion.
16.6.5 APPROVAt.S The procedure for cbtaining approval for any change, mod-Ification, or other item which requires approval of the Reactor Saf-ety Committee shall be as follows:
1.
The Reactor Supervisor shall prepare a proposal for review of the Reactor Administrator who shall submit it for approval to I
the Reactor Safety Ccemi ttee. The Reactor Safety Ccemittee shall be responsible for review and audit as prescribed in Section 16.6.4 A copy of the findings of this committee shall 1
t be submitted to the Radiation Safety Officer for action as re-quired by the University Radiatien Safety Comittee.
2.
The Reactor Administrator shall submit copies of proposals reviewed by the Reactor Safety Ccemittee to the Chairman of Mechanical Engineering.
j 3
The Reactor Administrator shall upon receipt of the required approvals from the Reactor Safety Committee and the Radiation Safety Committee authorize the Reactor Supervisor to proceed with the preposed change or mcdi fication.
~
16.6.6 PROCEDURES There shall be written operating procedures that cover the folicwing activities. They shall be approved by the Reactor Administrator, a.
Conduct of Irradiations and experiments that could affect the cperation or safety of the reactor.
b.
Startup, cperation, and shutdcwn of the reactor.
c.
Fuel movement and changes to the core and experiments that can effect the reactivity.
d.
Preventive or corrective maintenance which could have an effect on the safety of the reactor.
-~
l e.
Surveillance, testing and calibration of ins truments, ccc enents l
l and systems involving nuclear safety.
16-17 f
w.
.w
.r-g f.
Review and approval of changes to procedures.'
y g.
Personnel radiation protection censistent with 10 CFR Part 20..
3 h.
Implementation of the Security Plan an'd Emergency Plan.
i I.
Administrative control of operation and maintenance.
. i.
?
Though substantive changes to the above procedures shall be made only wi th approval by the Reactor, Adminis trator, temporary changes to the procedures that do not chartge their original Intent may be made by the Reactor Supervisor. All such temporary changes shall be docu-a mented, and subsequently approved by the Reactor Administrator within lf 14 days.
16.6.7 EXPERIMENTS fi a.
Prior to initiating any new reactor experiment, e.g., class of experiments that could affect reactivity of the reactor or result in release of radioactive materials, an experiment plan shall be prepared, reviewed by the Reactor Safety Committee, and the Radiation Safety Committee and approved by the Reactor Administra-tor.
b.
Each experiment plan shall (1) Identify the type of experiment (previousIy approved or recently reviewed per 16.6.4), (2) identii/
the experimenters and (3) have been approved by the licensed r
I senior operator in charge of reactor operation.
I6.6.8 SAFETY LIMIT VIOLATICN The following actions shall be taken in the event a Safety Limit is violated:
The reactor will be shut dcun immediately and reactor operation a.
will not be resumed without authorization by the Nuclear Regula-tory Commission (NRC).
b.
The Safety Limit violation shall be reported to the Director of the appropriate NRC Regional Office of inspection and Enforcement (or his designee), the Chairman of Mechanical Engineering and to the Reactor Safety Committee not later than,the next work day.
c.
A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Reactor Safety Coenittee. This report shall describe (1) applicable circumstances preceding the viola-tion, (2) ef fects of the violation upon facility cceponents, systems or structures, and (3) corrective action taken to prevent recurrence, d.
The Safety Limit Violation Report shall be submitted to theNRC,
Radiation Safety Cemittee, the Chairman of Mec3anica!
l Engineering and the Reactor Administrator within 14 days of the violation.
16.6.9 REPORTING REOUIREPENTS in addition to the acclicable resorting re--
quirements or TITLE 10, Code of Federal Regulations, the folicwing reports shall be submitted to the Directo. of the appropriate NRC 16-18
- ~ -
e
~
Tech Sp:cs.
.i.
.~
i Regional Office unless otherwise noted.
16.6 9 1 RCUTINE REPORTS m.
Startuo Recort. A sumary report of plant startup and power escalation testing shall be submitted following (1) 4 I
receipt of an operating license. (2) amendment to the / license involving a planned increase in power level, (3) installation of fuel that has a different design, and (4) modifications that maj have signif t"antly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall j
address each of the tests identified in the Hazards Summary Report (hereinaf ter Safety Analysis Report) and shall in general include a descr8ption of the measured values of
?
the operating conditions or charapteristics cbtained during i
the tes t program and a comparison of these values with design predictions and specifications. Any corrective,
- actions that were required to obtain satisfactorf operation shall also be described. Any additional specific details required in license cenditions based on other commitments shall be included in this report.
Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days follcwing resumption or commencement of pcuer ope ra ti on, (3) 9 months folicwing Initial criticality, whichever is earliest.
I f the Startup Report does not Cover all three events (i.e. Initial criticality, completicn of startup test program and resumption or cccmencement of power operation), supplementary reports shall be submitted i
at least every three months untill all three events have been completed.
b.
Annual Coeratine Reecrt Routine coerating reports ccvering tne operation of tne uni: during the previous calendar year should be submitted prior to March 31 of each year.
The annual operating reports made by IIcensees shall provide a comprehensive sumarf of the operating experience having safety significance that was gained during the year, even though some repetition of previously reported Infornation may be involved.
References In the annual operating report
.to previously submitted reports shall be clear.
Each annual operating report shall include:
(1) A brief narrative sumarf of (a) Changes in facility design, performance characteris tics,
i and operating procedures related to reactor safety that
' occured during the reporting period.
\\
(b) 'Results of major surveillance tests 'and irspections.
i 16-19 l
l
--y
r c
'(2) A monthly tabulation showing the hours the reactor is operating.
(3) 1.ist of the unscheduled shutdowns, including the reasons therefore and corrective action taken, if any.
(4) Discussion of the major safety related corrective maintenance performed during the period, including the effects, if any, on the safe operation of the
~
reactor, and the seasons for the corrective maintenance required.
(5). A brief description of:
(a) Each change to the facility to the extent that it changes a description of the facility in the appl!-
J cation for license and amendments thereto.
(b) Changes to the procedures as described in Facility Technical Speci fications.
(c) Any new or untried experiments or tests performed during the reporting period.
(6) A sumary of the safety evaluation made for each change, test, or experiment not submitted for NRC approval pur-suant to 10 CFR 50.59 which clearly shows the reason leading to the conclusion that no unreviewed safety question existed and that no technical specification change was requi red.
(7) A sumary of the nature and amount of radioactive effluents released or discharged to the environs beyond tne effective control of the license as determined at or prior to the point of such release or discharge.
(a) 1.lould Vaste (sumarized on a 3 month basis)
(1) Total estimated quantity of radioactivity released (in curiss) and Total volu=e (in liters) of effluent water (including diluent) released.
(b) Airborne Vaste (Sumarl:ed en a 3 mcnth basis)'
(1) Total estimated quantity of radioactivity released (in curies) determined by an approved sampilng and counting metned.
(c) Solid Vaste (sumarized on an annual basisJ (1) Total amount of solid waste packaged (in cubic meters)
(2) Total activity in solid waste (in curies)
(3) The dates of shipments and dispcsitten (if shipped off sJte).
16-20 I
e e-
. g,r 4 (8) A description of the results of any environmental radiological
.T" F'
surveys performed outside the facility.
\\
(9) Radiation Exposure - A sumary of radiation exposures greater l'l than 100 mrem (50 mrem for persons under 18 years of age) received f,1; during the reporting period by fac! fity personnel or vistors.
-w.
M' 16.6.9.2 REPORTABLE OCCURENCES Reportable occurences, including causes probable consequences, corrective actions and measures to prevent f.J,
recurrence, shall be reported to the NRC. Supplemental reports may be required to fully descrfbe final resolution of the occurrence.
In case of corrected or supplemental reports, an amended if censee event report shall be completed and reference shall be made to the
.Q[.g',
original report date.
Promet Notification With Veltten Folicwue. The types of events l['.
a.
L listed below shall be reported as expeditiously as possible by telephone and confirmed by telegraph, mallgram, or f acsi' mile I
transmission to the Director of the appropriate NRC Regional Office, or his designated representative no* later than the firs t work day folicwing the event, with a written followup report wi thin two weeks.
Information provided shall contain narrative material to provide complete explanation of the circumstances surrounding the event.
(1) Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the
'^
required protective function by the time a monitored para-meter reached the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.
(2) Operation of the reactor or affected systems when any para-meter or operation subject to a limiting coaditlen is less conservative than the limi ting cond! tion for operation established in the technical specifications without taking permitted remedial action.
(3) Abnormal degradation discovered in a fission product barrier, lj i.e., cracked fuel disc, primarf gas-tight seals.
- a (4) Reactivity balance anomalies involving:
(a) disagreement between expected and actual critical positions of approximately 0.3% ak/k; (b) exceeding excess reactivity limit;
.T
-l (c) shutdo.n margin less conservative than spect fled in i
technical spectflcations;
' '. unexpected short-term reactivity changes that resulted In a period of 10 seconds or less; (e)
If sub-critical, an unplanned reactivity insertion of more than approximately 0.5 sk/k or any unplanned cri ti cali ty.
+-
16-21 f
f.
e a.
emm m.
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. /-
{-
(5) Failure or malfunction of one (or more) component (s) which pre-
't vents or could prevent, by itself, the fulfillment of the function.-
al requirements of system (s) used to cope with accidents analyzed in the Safety Analysis Report.
(6) Personnel error or procedural inadequacy which prevents, or could prevent, by itself, the fulfillment of the functional require-j i
ments of systems required to cope with accidents analy:ed in j
Safety Analysis Report.
F (7) Unscheduled Conditions' arising from natural or man-made events that, as a direct result of the event require reactor shutdcun, operation of safety systems, or other protective measures required bv, Technical Speciflcations.
e
{
(8) Errors discovered in the transient.or accident analyses or in the i
8 i
methods used for such analyses as described in the Safety Analysis Report or in the bases for the Technical Specifications that have or could have permitted reactor coeration in a manner less con-servative than assumed in the analyses.
4 (9) Performance of s tructures, systems, or ccmponents that requi res remedial action or corrective me3sures to prevent operation in a manner less conservative than assumed in the accident analysis q
in the Safety Analysis Report or technical specifications bases; or discovery during plant life of conditions not specifically 1
considered in the Safety Analysis Report or Technical Specifica-I tions that require remedial acticn or corrective measures to 8
prevent.the existence or development of an unsafe condition.
16.6.9 3 SPECIAl. REPORTS Special reports which may be required by the Nuclear Regulatory Connission shall be submitted to the Director of tne 4
appropriate NRC Regicnal Office within the time period specified for each report.
s.
4 16.6.10 RECORD RETENTION 16.6.10.1 Records to be P&tained for,a Period of at least five years:
a.
Operating loss or data which shall Identify:
j 1.
Completion of pre-startup -checkout, s tartup, power. changes,
and shutdown of the reactor.
't
~2.
Installation or removal of fuel elements, control rods or experiments that could af fect core reactivity.
3 Installation or removal of jumper:, special tags or notices, or other temporary changes to reactor safety ci rcui tr f.
4.
Rod worth measurements and other reactivity measurements.
b.
Principal maintenance operations.
c.
Reportable occurrences.
16-22 i
a q
f b
)
4
'I r-
.cw o,2w.
,f.
'e d.
Surveillance activities required by technical specifications.
e.
Facility radiation and contamination surveys.
f.
Experiments performed with the reactor.
This requirement may be satisfied by the nomal operations / log book plus.
1.
Records of radioaptIve material transferred from the facility as requi red by IIcense.
2.
Records required by the Reactor Safety Committee for the performance of new or special experiments.
1 g.
Changes to operating procedures.
16.6.10.2 Records to be retained for the life of the facility:
Gaseous and liquid radioactive effluents released to the environs a.
b.
Appropriate off-site environmental monitoring surveys.
c.
Fuel inventories and fuel transfers.
-l d.
Radiation exposures for all personnel.
Updated as-built drawings of the facility.
e.
f.
Records of transient or operational cycles for those cceponents designed for a limited number of transients or cycles.
g.
arcrds of training and qualifiention for members of the facility 5
i s taH.
a.
.ecords of reviews performed for changes made to procedures or alpment or reviews of tests and experiments pursuant to 10CFR W.59 I.
Records of meetings of the Reactor Safety Comittee.
t I
/
I I
i 1
16-23 I
i l
~L*
sucn ay:we.
~
i s
TABLE 16-!
LIMITING CONDITio!!S FOR OPERATION, INSTRUMENTATION
~
Sa fety Channel Set Point Functien e.
Nuclear Safety #1 Low count rate 3 120 cpm scram below 120 cpm 7
Nuclear Safety f2 (log)
~
r High power 3 0.2 watts scram at power > 0.2 watts Low power 3 0.5 x 10-I3 amps scram at soyjce levels
.< 0.5 x 10 amps Reactor period 3 8 see scram at period < 8 sec t '
~
Nuclear safety 13 (linear)
High power j 0.2 watt scram at powe r >. 0.2 wat t Low power
? 52 full scale scram at source levels
< 5% of full scale Shield water temperature
? 15'c scram at temperature < 15*C Shield water level y 10.5 inches scram at water levels > 10.5 Inches below highest point on manhold opening Seismic displacement
< 1/16" scram at displacements
~
> l/16" scram at operator option Manual scram alarm at or below level Radiation moni tor set to meet requirements of 10 CFR Part 20 4
e
/
I l
i
,l 4
I i
16-24
g
.o
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UNIVERSITY EXECUTIVE AUTHORITY o
~
PRESIDENT RADIATION EXECUTIVf VICE PRESIDENT AND PROVOST SAFETY COMMITTEE I.
DEAN, ENGINEERING AND ARCHITECTURE 9
REACTOR CHAIRMAN, MECHANICAL ENGlHEERING SAFETY COMMITTEE i
I i
I I
REACTOR ADMINISTRATOR REACTOR SUPERVISOR REACTOR OPEPATORS FIG. 16-1 ADMINIST?ATIVE ORGANIMTICN FOR REACTOR CONTROL AND SAFETY eO l
16-25 d
.. ~...
~.-
~-
l l
-r APPENDIX A LICENSE NO. R-23 TEGNICAL SPECIFICATIWS FCR TEXAS A&M UNIVERSITY AGN-20LM REACTOR (SERIAL #106)
DOCIEr NO. 50-59 DATE: August 1, 1978 AS 50DIFIED TO INCIEDE ANSI N378-1974 AND REGULATORY GUIDE 2.2 GUIDANCE
TABLE OF CONIEES 4
PAGE 1.0 DEFINITIONS........................
1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTBf SETTINGS.....
4 2.1 Safety Limits 4
2.2 Limiting Safety System Settings...........
4 3.0 LIMITING CONDITIONS FCR OPERATION.............
5 3.1 Peactivity Limits 5
3.2 Control and Safety Systems..............
6 3.3 Limitation on Experiments 9
3.4 Shielding
...................... 10 4.0 SIRVEILIXCE REQUIRBOUS................. 11 4.1 Reactivity Limits
.................. 11 4.2 Control and Safety System
.............. 12 5.0 DESIGN FEAWRES......................
13 5.1 Reactor 13 5.2 Fuel Storage...........
......... 14 5.3 Reactor Rocm
.................... 14 6.0 AD4INISTRATIVE CC6TRCLS..................
14 6.1 Organization..................... 14 6.2 Staff Qualifications..
18 6.3 Training........
18 6.4 Reactor Safety Board....
18 6.5 Procedures.........
20 6.6 Safety Limit Violaticn...........
20 6.7 Reporting Requirements.......
21 6.3 Record Retention..........
22 f
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1 1
1 1.0 DEFLNITIONS The tems Safety Limit (SL), Limiting Safety System Setting (LSSS),
and Limiting Conditions for Operation (ILO) are as defined in 50.36 of 10 CFR part 50.
l.1 Reactor Shutdcwn - The reactor shall be considered shutdown whenever 1.
either:
A.
All safety and control rods are fully withdrawn frczn j
the core, or B.
The core fuse melts resulting in separation of the core, 2.
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.
1.2 Reactor Ooeration - Reactor operation is any condition wherein the reactor is not shutdown.
1.3 Measuring Channel - A measuring channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring or responding to the value of a process variable.
1.4 Safety Channel - A safety channel is a measuring channel in the reactor safety system.
1.5 Reactor Safety System - The reactor safety system is that combination of safety channels and associated circuitry which foms an automatic protective system for the reactor or provides infomation which requires manual protective action be initiated.
1.6 Reactor Comnonent - A reactor component is any apparatus, device, or material that is a nomal part of the reactor assembly.
1.7 Ocerable - Operable means a ccmponent or system is capable of perfoming its intended function in its nonnal manner.
1.8 Crerating - Operating means a component or system is perfoming its intenced function in its nomal manner.
1.9 Channel Check - A channel check is a qualitative verification of acceptacle perfonnance by observation of channel behavior. This verification may include comparison of the channel with other independent channels or methods measuring the same variable.
1.10 Channel Test - A channel test is the introduction of a signal into the cnannel to verify that it is operable.
1
t 1.11 Channel Calibration - A channel calibration is an adjustment of the channel such that its output responds, within acceptable range and accuracy, to kncwn values of the parameter which tne channel measures. Calibration shall enccmpass the entire channel, including equipment, actuation, alam, or trip.
1.12 Experiment - An experiment is any of the following:
a.
An activity utilizing the reactor system or its components or the neutrons or radiation generated therein; b.
An evaluation or test of a reactor system operational, surveillance, or maintenance technique; c.
The material centent of any of the foregoing, including structural cceponents, encapsulation or confining boundaries, and contained fluids or solids.
1.13 Secured Experiment - Any experiment, or component of an experiment is deemed to be secured, or in a secured position, if it is held in a stationary position relative to the reactor by mechanical means. The restraint shall exert sufficient force on the experiment to 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 Exteriment - Any experiment, or ccmponent of an experiment is deemed to be unsecured whenever it is not secured as defined in 1.13 above. Moving parts of experiments are deemed to be unsecured when they are in motion.
1.15 Movable Experiment - A movable experiment is one which may be inserted, removed, or manipulated while the reactor is critical.
1.16 Removable Exteriment - 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.
1.17 Exterimental Facilities - Experimental facilities are those portions of tne reactor assecoly that are used for the introduction of experi-ments into or adjacent to the reactor core region or allow beams of radiation to exist fran the reactor shielding. Experimental facilities shall include the thermal colt =:n, glory hole, and access ports. i f
. e.
1.18 Potential Reactivity Worth - The potential reactivity worth of an experiment is the canmum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration.
The evaluation must consider possible trajectories of the experirent in motion relative to the reactor, its orientation along each trajectory, and circumstances which can cause internal changes such as creating or filling of void spaces or motion of mechanical components. For rerovable 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 :rcthods between two defined tenainal positions or configurations of the experiment.
For removable experiments, the ten:nnal positions are fully removed frcm the reactor and fully inserted or installed in the nomal functiening or intended positicn.
1.20 Explosive Material - Explosive material is any solid or liquid wruch is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in " Dangerous Properties of Industrial Materials" by N. I. Sax, Third Ed. (1068), or is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protectica Association in its publication 704-M, 1966, " Identification System for Fire Hazards of Materials," also enumerated in the
" Handbook for Laboratory Safety" 2nd Ed. (1971) published by The Chemical Rubber Co.
' L; W}
2.0 SAFETY LDHTS AND LBUTED SAFETY SYSTBf SETTINGS 2.1 Safety Limits Applicability His specification applies to the maximum steady state power level and maximu:n 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.
Soecification a.
The reactor power level shall not exceed 100 watts.
b.
ne mximum core temperature shall not exceed 200*C during either steady state or transient operation.
Bases ne 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. He Ha::ards Sumary Report dated August 1956 submitted on Docket F-15 by Aerojet-General Nucleonics (AGN) calculated a steady state core average temperature
+
<- rise of &5*C/ watt. Berefore, a steady state power level of 100 watts would result in an average core temperature rise of -50?C. The r
e corresponding maximum core temperature would be below 200*C thus assuring integrity of the core and retention of fission prod' cts.
2.2 Limiting Safety System Settings Aoplicability This specification applies to the parts cf the reactor safety system which will limit maximum pcwer and core temperature.
Objective To cssure that autcmatic 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-Icwing limiting safety system settings:
4
l 01annel Condition LSSS s
Nuclear Safety #2 High Power 110 watts Nuclear Safety #3 High Power
< 10 watts i
b.
The core therral fuse shall melt when heated to a temperature of 120*C or less resulting in core separation and a reactivity loss greater than Stak.
Bases Based on instrumentaticn 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 maximum available excess reactivity in the reactor is less than one dollar the reactor cannot became prompt critical and the corresponding shortest possible period is greater than 200 eilli-seconds. The high power LSSS of 10 watts in conjunction 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-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 te::perature of 200'C.
3.0 LBfITING GNDITIONS FCR CPERATICN 3.1 Reactivity Limits Aoplicability This specification applies to the reactivity condition om 'Se 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 The available excess reactivity with all control and safety rods a.
fully inserted and including the potential reactivity worth of all experiments shall not exceed 0.65% t.k/k referenced to 20*C.
5-
o b.
The shutdown margin with the most reactive safety or control rod fully inserted shall be at least 1% a k/k.
c.
The reactivity worth of the control and safety rods shall ensure sub-criticality on the withdrawal of the coarse control rod or any one safety rod.
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 red reactivity limitations assure that the reactor can be brought and maintained subcritical if the highest reactivity rod fails to scram and remains in its most reactive positions.
3.2 Control and Safety Systems Acplic.ibility These specifications apply to the reactor control and safety systems.
Objective To specify lowest acceptable level of perfomance, instnment set points, and the minimum number of operable components for the reactor control and safety systers.
Specification The total scram withdrawal time of the safety rods and coarse a.
control rod shall be less than 200 milliseconds.
b.
The safety rods and coarse control rod shall be interlocked such that:
1.
Reactor startup cannot cemence unless both safety rods and coarse control rod are fully withdrawn from the core.
r 2.
Only one safety rod can be inserted at a time.
I 3.
He coarse control rod cannot be inserted unless both safety l
rods are fully inserted.
Nuclear safety channel instnnentation shall be operable in c.
accordance with Table 3.1 whenever the reactor control or l
safety rods are not in their fully withdrawn positien. '
TABLE 3.1 Safety Channel Set Point Function hbclear Safety #1 Low count rate t 10 cps scram below 10 cps hbclear Safety 82 High p er
< 10 watt scram at power >10 watt
~1 im power 1 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)
High Power
< 10 watt scram at power >10 watt Low power
> 5% full scale scram at source levels
< 5% of full scale
.\\tanual scram scram at operator option
~ 7-
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 reactorstartugandscramthereactoriftheshieldwatertemperature falls below 15 C.
f.
The seismic displacement interlock sensor -hall 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. Interlocks on control and safety rods assure an orderly apprcach to criticality and an adequate shutdown capability.
The neutron detector channels (nuclear safety channels 1 through 3) assure that reactor pcwer levels are adequately monitored during reactor startup and operation. Requirements on minimum neutron levels will prevent reactor startup unless channels are operable and responding, and will cause a scram in the event of instrumentation failure. The power level scrams initiate redundant autcmatic protective action at power level scrams low enough to assure safe shutdown without exceeding any safety limits. The period scram conservatively limits the rate of rise of reactor power to periods which are manually controllable and will automatically scram the reactor in the event of unexpected large reactivity additicns.
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 te=peratuus belcw 15*C thereby limiting potential reactivity additions associated with temperature decreases.
Water in the shield tank is an 1:portant ccmponent of the reactor shield and operation without the water may produce excessive radiation levels. The shield tank water level interlock will arevent reactor operation without adequate water levels in tie shield tank.
-S-l
4 The reactor is designed to withstand 0.6g accelerations and 6 cm displacements. A seismic instrument causes a reactor scram whenever the instunent receives a horizontal acceleration that causes a hori: ental displacement of 1/16 inch or greater. The seismic displacement interlock assures that the reactor will be scramed and brought to a subcritical configuration during any seismic disturbance that may cause damage to the reactor or its components.
The manual scram allows the operator to manually shut down the reactor if an unsafe or other.<ise abnormal condition occurs that does not otheraise scram the reactor. A loss of electrical power de-energi:es the safety and coarse control rod holding magnets causing a reactor scram thus assuring safe and innediate shutdcwn in case of L pcwer outrage.
3.3 Lianitations on Exceriments Applicability This specification applies to experiments installed in the. eactor and its experic.antal facilities.
I Cbjective To prevent damage to the reactor or exce:sive release of radioactive materials in the event of an experimental failure.
Specification Experiments containing materials corrosive to reactor ccraponents a.
or which contain liquid or gasecus, fissionable materials shall be doubly encapsulated.
b.
Explosive materials shall not be inserted into experimental facilities of the reactor, The radioactive material content, including fissicn products c.
of any experiment shall be limited so that the ccmplete release of all gaseous, particulate, or volatile cceponents from the experiment will not result in doses in excess of 10% of the equivalent annual doses stated in 10 CFR Part 20 for persons occupying (1) unrestricted areas centinuously for two hours starting at time of release or (2) restricted areas during the length of time required to evacuate the restricted area.
a
.g.
d.
He 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 ccznpanents 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 6wo 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 Rese specifications are intended to reduce the likelihood of damage to reactor ccznponents and/or radioactivity releases resulting from an experimental failure and to protect operating personnel and the public frcm excessive radiation doses in the event of an experimental failure.
3.4 Shielding Apolicability This specification applies to reactor shielding required during reactor operation.
Objective The objective is to protect facility personnel and the public from radiation exposure.
Soecification ne follcwing shielding requirements shall be fulfilled during reactor i
operation:
H e reactor shield tank shall be filled with water to a height a.
within 10 inches of the highest point on the manhole opening.
l b.
The thermal column shall be filled with water or graphite except during I
a critical experiment (core loading) or during measurement of reactivity worth of therr.a1 column water or graphite.
The accelerator recm shall be censidered a restricted area.
c.
<\\
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 recomendations under accident conditions.
4.0 SURVEILIANCE RECUIRBENTS Actions specified in this section are not required to be perfonned if during the specified surveillance period the reactor has not been brought critical or is maintained in a shutdown condition extending beyond the specified surveillance period. However, the surveillance requirements must be fulfilled prior to subsequent startup of the reactor.
4.1 Reactivity Limits Acolicability This specification applies to the surveillance requirements for reactivity limits.
Objective a
To assure that reactivity limits for Specification 3.1 are not exceeded.
Specificaticn 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 shall be determined annually, but at intervals not to exceed 16 months.
i c.
The reactivity worth of an experiment shall be estimated or measured, as appropriate, before or during the first startup subsequent to the experiment's insertion.
l l
l l
l -..
N Bases
'Ihe 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 detemined 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 automatic action.
Based on experience 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 requirements of the reactor control and safety systems.
Specification a.
Safety and control rod scram times shall be measured annually, but at intervals not to exceed 16 months.
b.
Safety and control rods and drive shall be inspected for deterioration at intervals not to exceed 2 years.
c.
A channel test of the following safety channels shall be perfomed prior to the first reactor startup of the day or prior to each operation extending more than one day:
Nuclear Safety #1, #2, and #3 Manual scram d.
A channel test of the seismic displacement interlock shall be performed smiannually.
4 e.
A channel check of the following safety channels shall be performed daily whenever the reactor is in operation:
Nuclear Safety fl, #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 scramed to verify operability. i l
l l
l L
~
g.
The period, count race, and power level measuring channels shall 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 the sensing element to the apprcpriate set point. These calibrations shall be performed annually, but at intervals not to exceed 16 m;'ths.
Bases The channel tests and checks required daily or before each startup will assu: e that the safety channels and scram functions are operable.
Based on operating experience with reactors of this type, the annual scram measurements, channel calibrations, set point verifications, and inspections are cf 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 FENIURES 5.1 Reactor a.
The reactor core, including control and safety rods, contains approximately 660 grams of U-235 in the fom of 20% enriched UO2 dispersed in approximately 11 kilograms of polyethylene. The lower section of the cote is supported by an aluminum rod hanging frca a fuse link. He fuse melts at temperatures belcw 120*C causing the lower core section to fall away from 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.
3 b.
De core is surrounded by a 20cm thick high density (1.75 gm/cm )
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 aluminum core tank designed to contain any fission gases that might leak from the core.
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 them11 column when filled with graphite.
d.
The 6b foot diameter, fluid-tight shield tank is filled with water constituting a 55 cm thick fast neutron shield. The fast neutron shield is fomed by filling the tank with approximately 1000 gallons of water. He em:plete reactor shield shall limit doses to personnel in unrestricted areas to levels less than pemitted by 10 CFR 20 under operating conditions.
l i
e.
Two safety rods and one control rod (identical in size) contain up to 20 grams of U-235 each in the same fom as the core material.
These rods are lifted into the core by electrcmagnets, driven by reversible DC motors through lead screw assemblies. Dcenergizing the magnets causes a spring-driven, gravity-assisted scram. He fourth rod or fine control rod (approximately one-half the diameter of the other rods) is driven directly by a lead screw. His rod may contain fueled or unfueled polyethylene.
/.2 Fuel Storage 5
Fuel, including fueled experinents 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 ff is no greater than 0.8 for all conditions of moderation and reflect,on.
i 5.3 Reactor Room a.
The reactor room houses the reactor assembly and accessories required for its operation and maintenance.
b.
He reactor rocm is a separate room in the Zachry Engineering Center, constmeted with adequate shielding and other radiation protective features to limit doses in restricted and unrestricted areas to levels no greater than pennitted by 10 CFR 20, under nomal operating conditions, and to a level below criterion 19, Appendix A,10 CFR 50 recocuendations under accident conditions.
c.
Access doors to and frcm the reactor rooms will contain locks.
6.0 AD4INISTRATIVE CChTROIS 6.1 CRGANIZATION The administrative organization for control of the reactor facility and its operation shall be as set forth in Figure 1 attached here to.
The authorities and responsibilities set forth below are designed to comply with the intent and requirements for administrative controls of the reactor facility as set forth by the Nuclear Regulatory Cor=tission.
6.1.1 FRESIDENT ne president is the chief administrative officer responsible for the University and in whose name the application for licensing is made.
6.1.2 DEAN, COLLEGE OF ENGINEERING The Dean of Engineering is the administrative officer responsible for the operation of the College of Engineering.
FIGURE 1 Administrative Organi::ation of the Texas A&M University AGN-20LM Reactor Facility NRC License R-23 President Texas A&M University Radiological Dean Reactor Safety Board College of Safety Boani Engineering I
I I
I I
1 I
I I
l l
Head I
Department of
J Nuclear Engineering i
Reactor Supervisor
- I Reactor Operators **
1 Requires NRC Senior Operators License Requires NRC Operators License except where exempt per 10 CFR 55 paragraph 55.9..
6.1.3 HEAD, DEPARTME.Vr OF NIX'UAR ENGINEERING The Head of the Department of Nuclear Engineering is the administrative officer responsible for the operation of the Department of Nuclear Engineering, including the AGN-201M Reactor Facility.
In this capacity he shall have final authority and ultimate respcnsibility for the operation, maintenance, and safety of the reactor facility within the limitations set forth in the facility license. He shall be responsible for appointing personnel to all positions reporting te him as described in Section 6.1 of the Technical Specifications. He shall seek the advice
'~
and approval of the Radiological Safety Board and/or the Reactor Safety Board in all matters concerning unresolved safety questions, new experiments and new procedures, and facility modifications which might affect safety. He s ull be an g officio member of the Reactor Safety Board.
6.1.4 REACIER SUPERVISOR The Reactor Supertisor shall be responsible for the preparation, promulgation, and enforcenent of ad:ninistrative controls including all mies, regulations, instructions, and operating procedures to ensure that the reactor facility is operated in a safe, ccmpetent, and authori:ed manner at all times. He shall direct the activities of operators and technicians in the daily operation and maintenance of the reactor; schedule reactor operations and maintenance; be responsible for the preparation, authentication, and storage of all prescribed logs and operating records; authori:e all experiments, procedures, and changes thereto which have received the approval of the Reactor Safety Board and/or the Radiological Safety Corraittee and the Head of the Department of Nuclear Engineering; and be responsible for the preparation of experimental procedures involving use of the reactor.
6.1.5 REACTOR OPERATORS Reactor Operators shall be responsible for the manipulation of the reactor controls, monitoring of instmmentation, operation of reactor related equipment, and maintenance of complete and current records during operaticn of the facility.
Reactor Operators who are exempt from holding an NRC license per 10 CFR 55 paragraph 55.9 shall only cperate the reactor under the direct and i:rediate supervisen of a licensed Reactor Operator. e
e 9
6.1.6 REACTOR SAFELY BOARD n e Reactor Safety Board shall be responsible for, but not limited to, revicwing and approving safety standards associated with the use of the reactor facility; reviewing and approving all proposed experiments and procedures and changes thereto ; reviewing and approving all modifications to the reactor facility which might affect its safe operation; detemuung whether proposed experiments, procedures, or modifications involve unreviewed safety questions, as defined in 10 CFR 50 paragraph 50.59 (c), and are in accordance with these Technical Specifications; conducting periodic audits of procedures, reactor operations and maintenance, equipment performance, and records; review all reportable occurrences and violations of these Technical Specifications, evaluating the causes of such events and the corrective action taken and recorrnending measures to prevent reoccurrence; reporting all their findings and reccmendations concerning the reactor facility to the Head of the Nuclear Engineering Department.
6.1.7 RADIOIDGICAL SAFEIY BOARD ne Radiological Safety Board shall advise the University administration and the Radiological Safety Officer on all matters concerning radiological safety at university facilities.
6.1.8 RADIGILGICAL SAFETY OFFICER ne Radiological Safety Officer shall review and approve all procedures and experiments involving radiological safety. He shall enforce all federal, state, and university rules, regulations, and procedures relating to radiological safety. He shall perfom routine radiation surveys of the reactor facility and report his findings to the Head of the Depart =ent of Nuclear Engineering. He shall provide personnel dosimetry and keep records of persennel radiation exposure. He shall advise the Head of the Department of Nuclear Engineering on all matters conceming radiological safety at the reactor facility. The Radiological Safety Officer shall be an g officio member of the Reactor Safety Board.
6.1.9 CPERATING STAFF The miniriu operating staff during any time in which the reactor is not shutdcwn shall consist of cne licensed Reactor Operator and one other perscn certified by the Reactor Supervisor as qualified to scram the
[
reactor and initiate emergency procedures. A licensed Senior Reactor Operator shall supervise all reactor maintenance or modification which could affect the reactivity of the reactor.
i.
6.2 STAFF CUALIFICATIONS The Head of the Department of Nuclear Engineering, the Reactor Supervisor, licensed Reactor Operators, and technicians performing reactor maintenance shall meet the minimum qualifications set forth in ANS 15.4, " Standards for Selection and Training of Personnel for Research Reactors". Reactor Safety Board members shall have a minimum of five (5) years experience in their profession or a baccalaureate degree and tw (2) years of professional experience. Reactor Safety Board members will generally be University faculty members with considerable experience in their area of expertise. The Radiological Safety Officer shall have a baccalaureate degree in biological or physical science and have at least two (2) years experience in health physics.
6.3 TRAINING The Head of the Department of Nuclear Engineering 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 reactor operators shall participate in requalification training as set forth in 10 CFR 55.
6.4 REACTOR SAFETY BCARD 6.4.1 MEETINGS AND CUORUM The Reactor Safety Board shall meet as often as deemed necessary by the Reactor Safety Board Chairman but shall meet at least once each calendar year. A quorum for the conduct of official business shall be the chairman, or his designated alternate, and two (2) other regular members.
At no time shall the operating organi::ation comprise a voting majority of the members at any Reactor Safety Beard meeting.
6.4.2 REVIEWS The Reactor Safety Board shall review:
a.
Safety evaluations for changes to procedures, equipment or systems, and tests or experiments, conducted without Nuclear Regulatory Co::nission approval under the provision of 10 CFR 50 paragraph 50.59, to verify that such actions did not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems that change the original intent or use, and are non-censervative, or those that involve an unreviewed safety question as defined in 10 CFR 50 paragraph 50.59.
j 5 l l
r i
Proposed tests or experiments which are significantly different c.
from previous approved tests or experiments, or those that involve an unreviewed safety question as defined in Section 50.59, 10 CFR 50 paragraph 50.59.
d.
Proposed changes in Technical Specifications or licenses, Violations of applicable statutes, codes, regulations, orders, e.
Technical Specifications, license 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.
h.
Audit reports.
6.4.3.
AUDITS Audits of facility activities shall be perfonned at least quarterly under the cognizance of the Reactor Safety Board but in no case by the personnel responsible for the item audited. These audits shall exaznine the operating records and encompass but shall not be limited to the following:
The conformance of the facility operation to the Technical Specific-a.
ations and applicable license conditions, at least annually.
b.
The Facility Emergency plan and impicmenting procedures, at least every two years.
The Facility Security Plan and implementing procedures, at least c.
every t'ao years.
6.4.4 AIJIFCRITf The Reactor Safety Board shall report to the President and shall advise the Head of the Department of Nuclear Engineering on those areas of responsibility outlined in section 6.1.6 of these Technical Specifications.
6.4.5 MINJTES OF THE REACTOR SAEIY BOARD The Chairman of the Reactor Safety Board shall direct the preparation, maintenance, and distributien of minutes of its activities. These minutes shall include a simcary of all meetings, actions taken, audits,.
and reviews.
6.5 PROCEDURES nere shall be written procedures that cover the following activities:-
a.
Startup, operation, and shutdown of the reactor.
b.
Fuel novement and changes to the core and experiments that could affect reactivity.
c.
Conduct of irradiatic,ns and experiments that could affect the operation or safety of the reactor.
d.
Preventive or corrective maintenance which could affect the safety of the reactor.
Surveillance, testing, and calibration of instruments, components, e.
and systems as specified in section 4.0 of these Technical Specifications.
f.
Implementation of the Security Plan and Emergency Plan.
The above listed procedures shall be approved by the Head of the Department of Nuclear Engineering and the Reactor Safety Board. Temporary procedures which do not change the intent of previously approved procedures and which do not involve any unreviewed safety question may be employed on approval by the Reactor Supervisor.
6.6 SAFETY LIMIT VI0IATION The following ac-ions shall be taken in the event a Safety Limit is violated:
The reactor zill be shut down imediately and reactor operation will a.
not be resumed without authorization by the Nuc1 car Regulatory Camission (h3C).
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 not later than the next work day.
A Safety Limit Violation Report shall be prepared for review by the c.
Reactor Safety Board. His report shall describe the a?plicable circumstances preceding the violation, the effects of tie violation upon facility cccponents, systens or stmetures, and corrective action to prevent recurrence.
d.
The Safety Limit Violation Report shall be submitted to the NRC, and Reactor Safety Board within 14 days of the violation. --
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o 6.7 REPORTING REQUIRBIENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the fo11cwing 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 follcwing the end of the operating year. Each annual report shall include a summary of the following activities or. curring during the operating year:
a.
Facility modifications.
b.
Results of major surveillance tests and inspections.
c.
Corrective maintenance performed.
d.
Energy produced by the reactor in watt-hours.
e.
Unscheduled shutdowns.
f.
Reactor Safety Board action pertinent to the Acility, g.
Any activities which require reporting per 10 CFR 50 paragraph 50.59 h.
Any reportable occurrences as defined in section 6.7.2 of these Technical Specifications.
6.7.2 REPCRTABLE CCCURRENCES Reportable occurrences, including causes, probable consequences, corrective actions and measures to prevent recurrence, shall be reported to the NRC.
Promet Notification With Written Fo11cwuo. The types of events listed a.
shall.be reported as expeciticusly 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 folicwing the event, with a written followup report within two weeks. Information provided shall contain narrative material to provide ccmplete explanation of the circumstances surrounding the event.
(1) Failure of the reactor protection system subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reached the seemint 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 condition for operation established in the technical specifications.
e (3) Abnonnal degradation discovered in a fission product barrier.
I (4) Reactivity balance ancnalies involving:
(a) disagreement between expected and actual critical
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positions of approximately 0.3% ak/k; (b) exceeding excess reactivity limit; i
(c) shutdcwn margin less conservative than specified in technical specifications; i
(5) Failure or malfunction of one (or more) ccuponent (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 which prevents, or could prevent, by itself, the fulfillment 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 i
i Report or in the bases for the Technical Specifications that have permitted reactor operation in a manner less conservative than assumed in the analyses.
I (8) Perfonnance of structures, systems, or compcnents 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 during plant life of conditions not specifically considered in the Safety Analysis Report or Technical Specifications that require remedial action or corrective measures to prevent the existence or developnent of an unsafe conditicn.
j 6.8 RECCRD RETSTICN 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.
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c.
Fuel inventories and fuel transfers.
d.
Operating logs.
a.,.
i e.
Maintenance logs.
f.
Updated drawings of th,e reactor facility.
g.
Personnel dosimetry records on file with the RadioJogical Safety
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Officer.
i h.
Minutes of the Reactor Safety Board meetings.
6.8.2 Records to be retained for a period of at least 'three years:
Serveillance activities required by Technical Specifications, a.
b.
Facility radiation and contamination surveys.
6.8.3 Personnel requalification and training records will be kept at least one year after tennination of employment.
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