ML15071A264

University of New Mexico - Notification of Changes to the Technical Specifications Reflecting the Split of the Department Into a Stand-Alone Nuclear Engineering Department

ML15071A264
Person / Time
Site:	University of New Mexico
Issue date:	03/04/2015
From:	Busch R D, Prinja A Univ of New Mexico
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML15071A264 (41)

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4\ 1-JNM ScHOOLo{ENGINEERING Department of Nuclear Engineering U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington D.C. 20555 '2. o IS" March4,~

Based on comments from our Reactor Safety Advisory Committee, we have reviewed and revised the technical specifications for the AGN-201M reactor located at the University of New Mexico, Docket 50-252. Changes have been made to reflect the split of the department into a stand-alone Nuclear Engineering Department.

Other changes of an editorial nature have been made to make the document easier to read and use. Changes are noted by bars on the right hand side of the page and are described in the pages following this transmittal letter. An original and one copy of the revised document are submitted for approval.

If you have any questions or comments, please let us know. I declare under penalty of perjury that the foregoing is true and correct.

Executed on March 4, 2015. Robert D. Busch, Ph.D, P .E. Chief Reactor Supervisor (505) 277-8027 Fax: (505) 277-5433 Email: busch@unm.edu

( CJ 6-*bt (,g, r-N, c:J ~J ru ffll)) Anil Prinja, P Reactor Administrator (505) 277-2209 The University of New Mexico

• MSCO I 1120

• l University of New Mexico

• Albuquerque, NM 87131-000 I

• Phone 505.277.5431

• Fax 505.277.5433

• engineering.unm.edu 209 Farris Engineering Center U>NIM SCHOOL of ENGINEERING Department of Nuclear Engineering REVIEW OF TECHNICAL SPECIFICATIONS Changes are noted below:Definitions have been renumbered with the addition of 3 items and the removal of I item.Pg. 2, 1.1.9 -current definition of explosive material was vague. It has been replaced with thefollowing based on a definition from 10 CFR 61.Explosive Material

-any chemical

compound, mixture, or device, which produces a substantial instantaneous release of gas and heat spontaneously or by contact with sparks or flame.For ease of reference, definitions for Safety Limit (SL), Limiting System Safety Setting (LSSS),and Limiting Condition of Operation (LCO) have been placed in the documentdefinitions.

The following definitions based on 10CFR50.36 have been added to thedefinitions list and the introductory sentence referencing 10CFR50.36 has been modifiedto indicate that the definitions are provided for ease of reference.

Pg. 2, 1.1.11 -Limiting Conditions for Operation (LCO) -the lowest functional capability orperformance levels of equipment required for safe operation of the facility.

Pg. 2, 1.1.12 -Limiting Safety System Settings (LSSS) -settings for automatic protective devicesrelated to those variables having significant safety functions.

Pg. 4, 1.1.29 -Safety Limits (SL) -limits on important process variables that are necessary toreasonably protect the integrity of certain of the physical barriers that guard against theuncontrolled release of radioactivity.

Pg. 5, The term 'true value' is not used anywhere in the document, so that definition will beremoved.Pg. 8, 3.1 .b -The phrase "with the most reactive safety or coarse control rod fully inserted andthe fine rod fully inserted" has been moved to the definition of shutdown margin.Pg. 12, 3.3.d -English equivalent units have been added to paragraph.

"... in excess of 0.1 rem (1mSv) or ..." and "...restricted area in excess of 5 rem (0.05 Sv).Pg. 13, 3.4 Specification

-Editorial change to make it read easier. Section has been changed toread -During Reactor Operation:

a. An operable

..., and then removed the phrasesrelating to reactor operation from parts a, b, c, d, and e.The University of New Mexico

• MSCOI 1120 -1 University of New Mexico -Albuquerque, NM 87131-0001

-Phone 505.277.5431

-Fax 505.277.5433

-engineering.unmn.edu 209 Farris Engineering Center j SCHOOL of ENGINEERING Department of Nuclear Engineering Pg. 13, 3.4 Basis -radiation levels in the second paragraph now have a dose equivalent of 100jirem/hr listed rather than an exposure level.Pg. 14, 15, 16, and 17, Startup check requirements need to be defined in a consistent manner onthese pages. That is, which are required for the first startup of the day and which arerequired before any startup or operation.

Specifically, have changed 4.4.b to read, "Thereactor access control (Ref 3.4c) shall be verified to be operable prior to the first reactorstartup of the day or prior to each reactor operation extending more than one day."Pg. 17, 4.4.b and the Basis section -both referenced 3.4.d, which was incorrect.

These have beenchanged to the correct reference

-3.4.c.Due to the split in departments, department name and title corrections were required in Section6.1 and in Figure 1..Pg. 22, 6.1.2 -changed to Chair, Nuclear Engineering.

Pg. 22, 6.1.3 -changed the wording of the second sentence to read, "... is selected by the Chairof the Nuclear Engineering Department and ...". Also removed the last phrase of the lastsentence because the NE Lab Supervisor has control over the locks on the NE Lab.Pg. 23, Fig. 1 -changed the title of the Chair to Chair, NE Department, and removed the NE LabSupervisor as this position has no direct responsibility to the Reactor.Pg. 24, 6.1.4 -changed the wording of the second sentence to read, "The UNM Radiation SafetyOfficer or designee normally

..."Pg. 26, 6.1.12 -For consistency, changed reactor control room to reactor room on a. 1, and on c.Also, changed a.3 to read, "One radiation safety staff member who can ..."The RSAC agreed that these changes were editorial in nature and will not affect the intent orcoverage of the Technical Specifications.

The University of New Mexico' MSCOI 1120

• I University of New Mexico

• Albuquerque, NM 87131-0001

• Phone 505.277.5431

• Fax 505.277.5433

-engineering.unmn.edu 209 Farris Engineering Center LICENSE NUMBER R- 102TECHNICAL SPECIFICATIONS FORTHE UNIVERSITY OF NEW MEXICO AGN-201M REACTORSERIAL NUMBER 112DOCKET NUMBER 50-252REVISED NOVEMBER 2014 TABLE OF CONTENTS1.0 DEFIN ITION S ....................................................................................................................

12.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

.........................

62.1 Safety Lim its ......................................................................................................

62.2 Lim iting Safety System Settings

........................................................................

73.0 LIM ITIN G CON D ITION S FO R O PERA TION .............................................................

83.1 Reactor Core Param eters ...................................................................................

83.2 Reactor Control and Safety System s .................................................................

93.3 Lim itations on Experim ents .............................................................................

123.4 Radiation M onitoring, Control A nd Shielding

.................................................

134.0 SU RV EILLAN CE REQ U IREM EN TS ........................................................................

144.1 Reactivity Lim its ...............................................................................................

144.2 Control and Safety System s .............................................................................

154.3 Reactor Structure

...............................................................................................

164.4 Radiation M onitoring and Control ....................................................................

174.5 Conduct of Experim ents ....................................................................................

185.0 D ESIGN FEA TU RE S ...................................................................................................

195.1 Reactor ..................................................................................................................

195.2 Fuel Storage .....................................................................................................

205.3 Reactor Room ....................................................................................................

216.0 A D M IN ISTRA TIV E CON TRO LS .............................................................................

226.1 Organization

......................................................................................................

226.2 Staff Q ualifications

...........................................................................................

266.3 Training

..................................................................................................................

276.4 Reactor Safety A dvisory Com m ittee ...............................................................

276.5 Approvals

...............................................................................................................

296.6 Procedures

.........................................................................................................

296.7 Experim ents ......................................................................................................

306.8 Safety Lim it V iolations

....................................................................................

306.9 Reporting Requirem ents ....................................................................................

306.10 Record Retention

...............................................................................................

35

1.0 DEFINITIONS

The terms Safety Limit (SL), Limiting Safety System Setting (LSSS), and LimitingConditions for Operation (LCO) are as defined in 10 CFR 50.36. Those definitions areincluded here for ease of reference.

1.1 Definitions

1.1.1 Cadmium Rod -An aluminum rod wrapped with Cd and inserted into theglory hole to assure that the reactor is secured.

The rod is worth at least $7of negative reactivity.

1.1.2 Channel Calibration

-A channel calibration is an adjustment of thechannel such that its output responds, within acceptable range andaccuracy, to known values of the parameter that the channel measures.

Calibration shall encompass the entire channel, including equipment, actuation, alarm, or trip.1.1.3 Channel Check -A channel check is a qualitative verification ofacceptable performance by observation of channel behavior.

Thisverification may include comparison of the channel with otherindependent channels or methods measuring the same variable.

1.1.4 Channel Test -A channel test is the introduction of a signal into thechannel to verify that it is operable.

1.1.5 Coarse Control Rod -The control rod with a scram function that can bemechanically withdrawn/inserted at two possible speeds (40-50 secondsfull insertion time or 80-100 seconds full insertion time).1.1.6 Excess Reactivity

-The amount of reactivity above a keff = 1. This is theamount of reactivity that would exist if all control rods were moved to themaximum reactive condition from the point where the reactor is exactlycritical (k.f = 1)1.1.7 Experiment

-An experiment is any of the following:

a. An activity utilizing the reactor system or its components or theneutrons or radiation generated therein;b. An evaluation or test of a reactor system operational, surveillance, ormaintenance technique;

c. The material content of any of the foregoing, including structural components, encapsulation or confining boundaries, and contained fluidsor solids.Revised November 2014I 1.1.8 Experimental Facilities

-Experimental facilities are those portions of thereactor assembly used for the introduction of experiments into or adjacentto the reactor core region or to allow beams of radiation to exist outsidethe reactor shielding.

Experimental facilities shall include the thermalcolumn, glory hole, and access ports.1.1.9 Explosive Material

-Any chemical

compound, mixture, or device, whichproduces a substantial instantaneous release of gas and heat spontaneously or by contact with sparks or flame.1.1.10 Fine Control Rod -A low worth control rod (about 25% of the worth ofthe other control rods) used primarily to maintain an intended power level.Its position may be varied manually.

The fine control rod does not drop ona scram signal, but withdraws automatically.

1.1.11 Limiting Conditions for Operation (LCO) -The lowest functional capability or performance levels of equipment required for safe operation of the facility.

1.1.12 Limiting Safety System Settings (LSSS) -Settings for automatic protective devices related to those variables having significant safetyfunctions.

1.1.13 Major Change -Any change in reactor configuration which affects theprobability or consequences of an event.1.1.14 Measured Value -The measured value is the value of a parameter as itappears on the output of a channel.1.1.15 Measuring Channel -A measuring channel is the combination of sensor,lines, amplifiers, and output devices which are connected for the purposeof measuring or responding to the value of a process variable.

1.1.16 Movable Experiment

-A movable experiment is one that may be inserted,

removed, or manipulated while the reactor is critical.

1.1.17 Operable

-Operable means a component or system is capable ofperforming its intended function in its normal manner.1.1.18 Operating

-Operating means a component or system is performing itsintended function in its normal manner.1.1.19 Potential Reactivity Worth -The potential reactivity worth of anexperiment is the maximum absolute value of the reactivity change thatwould occur as a result of intended or anticipated changes or crediblemalfunctions that alter experiment position or configuration.

Revised November 20142 1.1.20 Reactor Component

-A reactor component is any apparatus, device, ormaterial that is a normal part of the reactor assembly.

1.1.21 Reactor Operation

-Reactor operation is any condition wherein the reactoris not secured.1.1.22 Reactor Operator

-An individual who is licensed to manipulate thecontrols of a reactor.1.1.23 Reactor Safety System -The reactor safety system is that combination ofsafety channels and associated circuitry which forms an automatic protective system for the reactor or provides information that requiresmanual protective action be initiated.

1.1.24 Reactor Secured -The reactor shall be considered secured whenever:

a. either: 1. The safety and control rods are fully withdrawn from the core; or2. The core fuse melts resulting in separation of thecore.and:b. the reactor console key switch is in the "off' position; the key isremoved from the console and under the control of a certified operator; and the Cd rod is in the glory hole.1.1.25 Removable Experiment

-A removable experiment is any experiment, experimental

facility, or component of an experiment, other than apermanently attached appurtenance to the reactor system, which canreasonably be anticipated to be moved one or more times during the life ofthe reactor.1.1.26 Research Reactor -A research reactor is a device designed to support aself-sustaining neutron chain reaction for research, development, educational,

training, or experimental

purposes, and which may haveprovisions for producing radioisotopes.

1.1.27 Safety Channel -A safety channel is a measuring channel in the reactorsafety system.1.1.28 Safety Control Rod -One of two scramniable control rods that can bemechanically withdrawn/inserted at only one speed (35 to 50 seconds fullinsertion time).Revised November 20143 1.1.29 Safety Limit (SL) -Limits on important process variables that arenecessary to reasonably protect the integrity of certain of the physicalbarriers that guard against the uncontrolled release of radioactivity.

1.1.30 Scram Time -The time for the control rods acting under gravity to changethe reactor from a critical to a subcritical condition.

In most cases, this isless than or equal to the time it takes for the rod to fall from full-in tofull-out position.

1.1.31 Secured Experiment

-Any experiment, or component of an experiment isdeemed 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 shallexert sufficient force on the experiment to overcome the expected effectsof hydraulic, pneumatic,

buoyant, or other forces which are normal to theoperating environment of the experiment or which might arise as a resultof credible malfunctions.

1.1.32 Senior Reactor Operator

-An individual who is licensed to direct theactivities of reactor operators.

Such an individual is also a reactoroperator.

1.1.33 Shall, Should and May -The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" todenote permission--neither a requirement nor a recommendation.

1.1.34 Shutdown Margin -Shutdown margin shall mean the minimum shutdownreactivity necessary to provide confidence that the reactor can be madesubcritical by means of the control and safety systems starting from anypermissible operating condition with the most reactive safety or coarsecontrol rod fully inserted and the fine control rod fully inserted, and thatthe reactor will remain subcritical without further operator action.1.1.35 Static Reactivity Worth -The static reactivity worth of an experiment isthe value of the reactivity change measurable by calibrated control orregulating rod comparison methods between two defined terminalpositions or configurations of the experiment.

For removable experiments, the terminal positions are fully removed from the reactor and fully insertedor installed in the normal functioning or intended position.

Revised November 20144 1.1.36 Surveillance Time -A surveillance time indicates the frequency of tests todemonstrate performance.

Allowable surveillance intervals shall notexceed the following:

a. Two-year (interval not to exceed 30 months)b. Annual (interval not to exceed 15 months)c. Semiannual (interval not to exceed seven and one-half months)d. Quarterly (interval not to exceed four months)e. Monthly (interval not to exceed sixweeks).

Revised November 20145 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS2.1 Safety LimitsApplicability This specification applies to the maximum core temperature during operation.

Objective To assure that the integrity of the fuel material is maintained and that all fission productsare retained in the core matrix.Specification

a. The maximum core temperature shall not exceed 2000C during operation.

BasisThe polyethylene core material does not melt below 2000C and is expected to maintain itsintegrity and retain essentially all of the fission products at temperatures below 2000C.The Hazards Summary Report dated February 1962 submitted on Docket F-15 byAerojet-General Nucleonics (AGN) calculated a core maximum temperature rise of71.3°C while the Safety Analysis Report submitted during the 1986 relicensing of theUNM AGN calculated a core maximum temperature rise of 100.70C .In either case,assuming operation at 20'C, the corresponding maximum core temperature would be120.7°C or 91.30C, both of which are well below 2000C thus assuring integrity of the coreand retention of fission products.

Revised November 20146 2.2 Limiting Safety System SettingsApplicability This specification applies to the parts of the reactor safety system which will limitmaximum power and core temperature.

Objective To assure that automatic protective action is initiated to prevent a safety limit from beingexceeded.

Specification

a. The safety channels shall initiate a reactor scram at the following limiting safetysystem settings:

Channel Condition LSSSNuclear Safety #2 High Power6 wattsNuclear Safety #3 High Power6 wattsb. The polystyrene core thermal fuse melts when heated to a temperature of about120'C resulting in core separation and a reactivity loss greater than 5% Ak/k.BasisBased on instrumentation response times and scram tests, the AGN Hazards Reportconcluded that reactor periods in excess of 30-50 milliseconds would be adequately arrested by the scram system. Since the maximum available excess reactivity in thereactor is less than one dollar, the reactor cannot become prompt critical, and thecorresponding shortest possible period is greater than 200 milliseconds.

The high powerLSSS of 6 watts in conjunction with automatic safety systems, and the maximumtemperature rise of 100.7°C, and/or manual scram capabilities will assure that the safetylimits will not be exceeded during normal operation or as a result of the most severecredible transient.

In the event of failure of the reactor to scram, the self-limiting characteristics due to thehigh negative temperature coefficient, and the melting of the thermal fuse at atemperature below 120'C will assure safe shutdown without exceeding a coretemperature of 200°C (the Safety Limit).Revised November 20147 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters Applicability This specification applies to the reactivity condition of the reactor and the reactivity worths of control rods and experiments.

Objective To assure that the reactor can be shut down at all times and that the safety limits will notbe exceeded.

Specification

a. The available excess reactivity with the coarse, fine, and safety control rods fullyinserted and including the potential absolute value of the reactivity worth of allexperiments shall not exceed 0.65% Ak/k.b. The shutdown margin shall be at least one dollar.c. The reactivity worth of the control rods shall ensure subcriticality on thewithdrawal of the coarse control rod or any one safety rod.d. The excess reactivity with no experiments in the reactor and the coarse, fine, andsafety control rods fully inserted shall not exceed 0.25% Ak/k.BasisThe limitations on total core excess reactivity assure reactor periods of sufficient lengthso that the reactor protection system and/or operator action will be able to shut the reactordown without exceeding any safety limits. The shutdown margin and control and safetyrod reactivity limitations assure that the reactor can be brought and maintained subcritical if the highest reactivity rod fails to scram and remains in its most reactive position.

Revised November 20148 3.2 Reactor Control and Safety SystemsApplicability These specifications apply to the reactor control and safety systems.Objective To specify lowest acceptable level of performance, instrument set points, and theminimum number of operable components for the reactor control and safety systems.Specification

a. The fine control rod, coarse control rod, and the two safety rods shall be operableand the carriage position of the fine and coarse control rods shall be displayed atthe console whenever any rod is above its lower limit.b. The total scram withdrawal time of the safety rods and coarse control rod shall beless than I second.c. The average reactivity addition rate for each control rod (fine, coarse, or safetyrod) shall not exceed 0.065% Ak/k per second.d. The safety rods and coarse control rod shall be interlocked such that:1. Reactor startup cannot commence unless both safety rods and the coarsecontrol rod are fully withdrawn from the core.2. Only one safety rod can be inserted at a time.3. The coarse control rod cannot be inserted unless both safety rods are fullyinserted.

4. At any operating power below 50 x 10-6 watts, none of the rods can bemoved to a more reactive position.

e. Nuclear safety channel instrumentation shall be operable in accordance withTable 3.1 whenever the reactor is in operation.

f. A manual scram shall be provided on the reactor console, and the safety circuitry.

shall be designed so that no single failure can negate both the automatic and-manual scram capability.

Revised November 20149

g. The shield water level interlock shall be set to prevent reactor startup and scramthe reactor if the shield water level falls more than 18 cm below the highest pointon the reactor shield tank manhole opening.h. The shield water temperature interlock shall prevent reactor startup or scram thereactor if the shield water temperature falls below 180C.The seismic displacement interlock shall be installed in such a manner to preventreactor startup or to scram the reactor during a seismic displacement.

j. A loss of electric power shall cause the reactor to scram.BasisThe specification on operability of the rods assures console control over reactivity conditions within the reactor.

Display of the positions of the fine and coarse control rodsassures that the positions of these rods are available to the operator to evaluate theconfiguration of the reactor.The specifications on scram withdrawal time in conjunction with the safety systeminstrumentation and set points assure safe reactor shutdown during the most severeforeseeable transients.

Interlocks on control rods assure an orderly approach to criticality and an adequate shutdown capability.

The limitations on reactivity addition rates allowonly relatively slow increases of reactivity so that ample time will be available formanual or automatic scram during any operating conditions.

The neutron detector channels (Nuclear Safety Channels

1. 2 and #3) assure that reactorpower levels are adequately monitored during reactor startup and operation.

The powerlevel scrams initiate redundant automatic protective action at power levels low enough toassure safe shutdown without exceeding any safety limits. The manual scram assures amethod of shutdown without reliance on safety channels and circuitry.

The AGN-20 l's negative temperature coefficient of reactivity causes a reactivity increasewith decreasing core temperature.

The shield water temperature interlock will preventreactor operation at temperatures below 18'C thereby limiting potential reactivity additions associated with temperature decreases.

Water in the shield tank is an important component of the reactor shield and operation without the water may produce excessive radiation levels. The shield tank water levelinterlock will prevent reactor operation without adequate water levels in the shield tank.Revised November 201410 The reactor is designed to withstand 0.6 g accelerations and 6 cm displacements.

Aseismic instrument causes a reactor scram whenever the instrument receives a horizontal acceleration that causes a horizontal displacement of 0.16 cm or greater.

The seismicdisplacement interlock assures that the reactor will be scrammed and brought to asubcritical configuration during any seismic disturbance that may cause damage to thereactor or its components.

The manual scram allows the operator to manually shutdown the reactor if an unsafe orotherwise abnormal condition occurs that does not scram the reactor.

A loss of electrical power de-energizes the safety and coarse control rod holding magnets causing a reactorscram thus assuring safe and immediate shutdown in case of a power outage.Table 3.1Nuclear Safety Channel Instrumentation Channel No.23FunctionHigh Power ScramHigh Power ScramOperating Limits120% of licensed power(6 Watts)120% of licensed power(6 Watts)Revised November 201411 3.3 Limitations on Experiments Applicability This specification applies to experiments installed in the reactor and its experimental facilities.

Objective To prevent damage to the reactor or excessive release of radioactive materials in theevent of an experimental failure.Specification

a. Experiments outside the reactivity limits defined in TS 3.1 shall not be permitted.

b. Experiments within the reactivity limits defined in TS 3.1 containing materials corrosive to reactor components or which contain gaseous or liquid fissionable materials shall be doubly encapsulated.

c. Explosive materials or materials which might combine violently shall not beinserted into experimental facilities of the reactor or irradiated in the reactor.d. The radioactive material

content, including fission products, of any doublyencapsulated experiment should be limited so that the complete release of allgaseous, particulate, or volatile components from the encapsulation could notresult in:(1) a total effective dose equivalent to any person occupying anunrestricted area in excess of 0.1 REM (0.001 Sv) or(2) a total effective dose equivalent to any person occupying a restricted area during the length of time required to evacuate the restricted area inexcess of 5 rem (0.050 Sv).BasisThese specifications are intended to reduce the likelihood of damage to reactorcomponents and/or radioactivity releases resulting from an experimental failure and toprotect operating personnel and the public from excessive radiation doses in the event ofan experimental failure.

Specification 3.3d conforms to 10 CFR 20 as of the date of thisrevision.

Revised Novemher 2014 112 3.4 Radiation Monitoring, Control And Shielding Applicability This specification applies to radiation monitoring,

control, and reactor shielding requiredduring reactor operation.

Objective The objective is to protect facility personnel and the public from radiation exposure.

Specification During Reactor Operation:

a. An operable portable radiation survey instrument capable of detecting gammaradiation shall be immediately available to reactor operating personnel wheneverthe reactor is in operation.

b. The reactor room shall be considered a restricted area according to 1OCFR20.c. The top of the reactor shall be considered a high radiation area, and the accessstairs to the top of the reactor shall be equipped with a gate and a lock for accesscontrol.

The keys for the gate shall be in control of the reactor operator duringoperation.

d. The following shielding requirement shall be fulfilled:

The thermal column shall be filled with water or graphite except during acritical experiment (core loading) or during other approved experiments thatrequire the thermal column to be empty.e. The core tank shall be sealed.BasisRadiation surveys performed under the supervision of a qualified health physicist haveshown that the total gamma, thermal neutron, and fast neutron radiation dose rate in the,reactor room, at the closest approach to the reactor but without access to reactor top, isless than 50 mrem/hr at reactor power levels of 5.0 watts.When the reactor is secured, radiation levels at all points in the reactor room are below100 jirem/hr.

The facility shielding in conjunction with radiation monitoring,

control, andrestricted areas is designed to limit radiation doses to facility personnel and to the publicto a level below 10 CFR 20 limits under all conditions.

Revised November 201413 4.0 SURVEILLANCE REQUIREMENTS Actions specified in sections 4.1, 4.2, and 4.3 are not required to be performed if duringthe specified surveillance period the reactor has not been brought critical or is maintained in a secured condition extending beyond the specified surveillance period. However, thesurveillance requirements shall be fulfilled prior to subsequent startup of the reactor.4.1 Reactivity LimitsApplicability This specification applies to the surveillance requirements for reactivity limits.Objective To assure that reactivity limits for Specification 3.1 are not exceeded.

Specification

a. Control rod reactivity worths shall be measured annually to verify 3.1 c.b. Total excess reactivity and shutdown margin shall be determined annually.

c. The reactivity worth of an experiment shall be estimated or measured, asappropriate, before or during the first startup subsequent to the experiment's first insertion.

BasisThe control and safety rod reactivity worths are measured annually to assure that nodegradation or unexpected changes have occurred which could adversely affect reactorshutdown margin or total excess reactivity.

The shutdown margin and total excessreactivity are determined to assure that the reactor can always be safely shut down withone rod not functioning and that the maximum possible reactivity insertion will not resultin reactor periods shorter than those that can be adequately terminated by either operatoror automatic action. Based on experience with AGN reactors, significant changes inreactivity or rod worth are not expected within a 12 month period.Revised November 201414 4.2 Control and Safety SystemsApplicability This specification applies to the surveillance requirements of the reactor control andsafety systems.Objective To assure that the reactor control and safety systems are operable as required bySpecification 3.2.Specification

a. A channel test of Nuclear Safety Channels

1. 2 and #3 shall be performed priorto the first reactor startup of the day or prior to each reactor operation extending more than one day.b. A channel check of Nuclear Safety Channels

1. 2 and #3 shall be performed daily whenever the reactor is in operation.

c. Prior to each day's reactor operation the rod interlock shall be checked tomake sure it is operating.

d. Prior to each day's reactor operation or prior to each reactor operation extending more than one day, safety rod #1 shall be inserted and scrammed toverify operability of the manual scram system.e. Prior to each day's reactor operation, it shall be verified that the lock on thegate for the access stairs is locked.f. Control rod scram times and average reactivity insertion rates shall bemeasured annually.

g. Control rods and drives shall be inspected for proper operation annually.

h. A channel test of the seismic displacement interlock shall be performed annually.

i. The power level measuring channels shall be calibrated and set points verifiedannually.

j. The shield water level interlock and shield water temperature interlock shallbe calibrated annually.

k. It shall be verified annually that loss of electrical power causes a scram.Revised November 201415 BasisThe channel tests and checks required daily or before each startup will assure that thesafety channels and scram functions are operable.

Based on operating experience withreactors of this type, the annual scram measurements, channel calibrations, set pointverifications, and inspections are of sufficient frequency to assure, with a high degree ofconfidence, that the safety system settings will be within acceptable drift tolerance foroperation.

4.3 Reactor Structure Applicability This specification applies to surveillance requirements for reactor components other thancontrol rods.Objective The objective is to assure integrity of the reactor structures.

Specification Visual inspection for water leakage from the shield tank shall be performed prior to eachstartup.

Leakage sufficient to activate the shield water level safety interlock shall becorrected prior to subsequent reactor operation.

BasisBased on experience with reactors of this type, the frequency of inspection and leak testrequirements of the shield tank will assure capability for radiation protection duringreactor operation.

The shield water level safety interlock is checked annually andprovides assurance that sufficient water is in the tank for adequate personnel shielding.

Revised November 201416 4.4 Radiation Monitoring and ControlApplicability This specification applies to the surveillance requirements of the radiation monitoring and control systems.Objective To assure that the radiation monitoring and control systems are operable and that allradiation and high radiation areas within the reactor facility are identified and controlled as required by Specification 3.4.Specification

a. All portable radiation survey instruments assigned to the reactor facility shallbe calibrated annually under the supervision of the Radiation Safety Office.b. The reactor access control (Ref 3.4c) shall be verified to be operable prior tothe first reactor startup of the day or prior to each reactor operation extending more than one day.c. A radiation survey of the reactor room shall be performed under thesupervision of the Radiation Safety Officer to determine the location ofradiation and high radiation areas corresponding to reactor operating powerlevels and to verify that the thermal column is providing shielding.

Thissurvey shall be performed as necessary but at least annually.

BasisThe periodic calibration of radiation monitoring equipment and the surveillance of thereactor access control (Ref 3.4c) will assure that the radiation monitoring and controlsystems are operable during reactor operation.

The periodic radiation surveys will verify the location of radiation and high radiation areas and will assist reactor facility personnel in properly labeling and controlling eachlocation in accordance with 10 CFR 20.Revised November 201417 4.5 Conduct of Experiments Applicability This specification applies to the surveillance requirements for experiments inserted in thereactor.Objective To prevent the conduct of experiments that may damage the reactor or release excessive amounts of radioactive materials as a result of experiment failure.Specification

a. The reactivity worth of an experiment shall be estimated or measured, asappropriate, before reactor operation with said experiment.

b. An experiment shall not be installed in the reactor or its irradiation facilities unless a safety analysis has been performed and reviewed for compliance withSection 3.3 by the Chief Reactor Supervisor and the Reactor Safety AdvisoryCommittee in full accord with Section 6.4.2 of these Technical Specifications.

BasisExperience has shown that experiments reviewed by the Chief Reactor Supervisor andRSAC can be conducted without endangering the safety of the reactor or exceeding thelimits in the Technical Specifications.

Revised November 201418 5.0 DESIGN FEATURES5.1 ReactorApplicability This specification applies to basic design features of the reactor.Objective To specify specific reactor design features.

Specification

a. The reactor core, including control rods, contains approximately 667 grams ofU-235 in the form of <20% enriched U02 dispersed in approximately 11kilograms of polyethylene.

The lower section of the core is supported by analuminum rod hanging from a fuse link. The fuse melts at a fuse temperature of about 120'C causing the lower core section to fall away from the uppersection reducing reactivity by at least 5% Ak/k. Sufficient clearance betweencore and reflector is provided to ensure free fall of the bottom half of the coreduring the most severe transient.

b. The core is surrounded by a 20 cm thick high density (1.75 gm/cm3) graphitereflector followed by a 10 cm thick lead gamma shield. The core and part ofthe graphite reflector are sealed in a fluid-tight aluminum core tank designedto contain any fission gases that might leak from the core.c. The core, reflector and lead shielding are enclosed in and supported by afluid-tight steel reactor tank. An upper or "thermal column tank" may serve asa shield tank when filled with water or a thermal column when filled withgraphite.

d. The 198 cm diameter, fluid-tight shield tank is filled with water constituting a55 cm thick fast neutron shield. The fast neutron shield is formed by fillingthe tank with approximately 3785 liters of water. The complete reactor shieldshall limit doses to personnel in unrestricted areas to levels less than permitted by 10 CFR 20 under operating conditions.

e. Two safety rods and one coarse control rod (identical in size) contain less than15 grams of U-235 each in the same form as the core material.

These rods arelifted into the core by electromagnets, driven by reversible DC motors throughlead screw assemblies.

De-energizing the magnets causes a spring-driven, gravity-assisted scram. The fourth rod or fine control rod (approximately one-half the diameter of the other rods) is driven directly by a lead screw. Thisrod may contain polyethylene with or without fuel.Revised November 201419 NOTE:All dimensions, masses, and densities given in the above description are nominal values.BasisThese basic design criteria are relevant to the safe operation of the reactor and should notbe changed or modified without NRC approval.

5.2 Fuel StorageApplicability This specification applies to the storage of reactor fuel at times when it is not in thereactor core.Objective To assure that fuel being stored shall be secured and shall not become critical.

Specification Fuel, including fueled experiments and fuel devices not in the reactor, shall be stored in asecured location when not in use. The storage array shall be such that ker is no greaterthan 0.9 for all conditions of moderation and reflection.

BasisThe limits imposed are conservative and assure safe storage (NUREG-1537).

Revised November 201 420 5.3 Reactor Room (065)Applicability This specification applies to the reactor location.

Objective To specify the characteristics of specific facility design features.

Specification

a. The reactor room houses the reactor assembly and accessories required for itsoperation and maintenance, and the reactor control console.b. The reactor room is a separate room (065) in the Nuclear Engineering Laboratory, constructed with adequate shielding and other radiation protective features to limitdoses in restricted and unrestricted areas to levels no greater than permitted by 10CFR 20.c. The access doors to the reactor room shall contain.

locks.,BasisThe reactor room provides a secure, controlled access area with appropriate shielding forpersonnel radiation protection.

Revised November 201421 6.0 ADMINISTRATIVE CONTROLS6.1 Organization The current administrative organization for control of the reactor facility and its operation is as set forth in Figure 1. Levels 1, 2, and 3 refer to administrative levels for whichchanges in staffing must be communicated to the Nuclear Regulatory Commission as setforth in 6.9.3. The authorities and responsibilities set forth below are designed to complywith the intent and requirements for administrative controls of the reactor facility as setforth by the Nuclear Regulatory Commission.

6.1.1 UNM Administration Has administrative responsibilities for all activities on Campus. The President (Level 1)is the chief administrative officer responsible for the University and in whose name theapplication for licensing is made. The Radiation Control Committee is a permanent committee established to act on behalf of the President of the University for control of allUniversity of New Mexico (UNM) activities involving sources of ionizing radiation.

TheCommittee consists of members from the UNM faculty/staff.

Meetings are held regularly.

Responsibilities are: to establish policy and disseminate rules for radiation safety andcontrol at UNM; to serve as the UNM liaison with the NRC in matters of registration, licensing, and radiation control; and to ensure periodic inspections and radiation surveysfor the purpose of assuring the safety of radiation operations within any UNM facility.

6.1.2 Chair, Department of Nuclear Engineering The chair (Level 1) is the administrative officer responsible for the operation of theDepartment, for its financial affairs and for appointing the Reactor Administrator.

TheChair is responsible for appointing members of the Reactor Safety Advisory Committee (RSAC) and the RSAC reports to the chair on all matters.6.1.3 Reactor Administrator Provides final policy decisions on all phases of reactor operation and regulations for thefacility.

The Reactor Administrator (Level 2) is selected by the Chair of the NuclearEngineering Department and shall hold a graduate degree in Engineering.

The ReactorAdministrator is advised on matters concerning personnel health and safety by theRadiation Safety Officer and/or the Radiation Control Committee.

The ReactorAdministrator is advised on matters concerning safe operation of the reactor by theReactor Operations Committee and/or the Reactor Safety Advisory Committee; designates Reactor Supervisors and names the Chief Reactor Supervisor; approves allregulations, instructions and procedures governing facility operation; submits the annualreport to NRC.Revised November 201422 i ~Chief Reda, SupervisI; Level 2NRC licensed~ReactorSupervisors Authorized Operators Figure 1Revised November 201423 6.1.4 Radiation Safety OfficeThe Radiation Safety Office will provide emergency direction and assistance forsituations involving radiation safety. The UNM Radiation Safety Officer or designeenormally represents the Radiation Control Committee in matters concerning the radiation safety aspects of reactor operation.

6.1.5 Reactor Safety Advisory Committee

Reviews, evaluates, and audits reactor operations and procedures to ensure that thereactor shall be operated in a safe and competent manner. There shall be at least fourmembers on the RSAC with at least two members from organizations outside theUniversity.

The Committee is available for advice and assistance on reactor operation problems.

Any major change in the facility shall be approved by the RSAC. Members ofthe RSAC are appointed by the Department Chair and shall not be members of theReactor Operations Committee.

The RSAC reports to the Chair and advises the ReactorAdministrator.

6.1.6 Reactor Operations Committee Consists of the Reactor Supervisors with the Chief Reactor Supervisor.

Other qualified persons may also be members.

They are directly responsible to the Reactor Administrator for the preparation and submission of detailed procedures, regulations, forms, and rules toensure the maintenance, safe operation, competent use and security of the facility.

TheCommittee ensures that all the activities, experiments, and maintenance involving thefacility are properly logged and are in accordance with established local and U.S. NuclearRegulatory Commission regulations.

They review all proposed changes in procedure orchanges in the facility and approve any minor change before the change is implemented.

6.1.7 Chief Reactor Supervisor Shall hold a Senior Reactor Operator's license issued by the NRC. He/she is responsible for the distribution and enforcement of rules, regulations and procedures concerning operation of the facility.

The Chief Reactor Supervisor (Level 3) is directly responsible for enforcing operating procedures and ensuring that the facility is operated in a safe,competent and authorized manmer. He/she is directly responsible for all prescribed logsand records; is the Emergency Director for emergencies not involving radiation; and hasthe authority to authorize experiments or procedures which have received appropriate prior approval by the Reactor Operations Committee, the Reactor Safety AdvisoryCommittee and/or the Radiation Control Committee (or the Radiation Safety Officer) andhave received prior authorization by the Reactor Administrator.

He/she shall notauthorize any proposed changes in the facility or in procedure until appropriate evaluation and approval has been made by the Reactor Operations Committee or theReactor Safety Advisory Committee and authorization given by the ReactorAdministrator.

Revised November 201424 6.1.8 Reactor Supervisors Shall hold valid Senior Reactor Operator's licenses issued by the Nuclear Regulatory Commission.

A Reactor Supervisor shall be in charge of the facility at all times duringreactor operation and shall witness the startup and intentional shutdown procedures.

ASenior Reactor Operator is required to be present whenever fuel is handled.

The ReactorSupervisors are directly responsible to the Chief Reactor Supervisor.

A ReactorSupervisor shall be present when the reactor is going critical, being intentionally shutdown, or when reactor experiments are loaded or unloaded.

The location of the ReactorSupervisor shall be known to the Reactor Operator at all times during operation so that itis possible to contact him/her if required.

6.1.9 Reactor Operators Shall hold a valid Reactor Operator's license issued by the NRC. They shall conform tothe rules, instructions and procedures for the startup, operation and shutdown of thereactor, including emergency procedures.

Within the constraints of the administrative andsupervisory controls outlined above, a reactor operator will be in direct charge of thecontrol console at all times that the reactor is operating.

The reactor operator shallmaintain complete and accurate records of all reactor operations in the operational logs.6.1.10 Authorized Operators Individuals authorized by the Reactor Supervisor to operate the reactor controls and whodo so with the knowledge of the Supervisor and under the direct supervision of a ReactorOperator.

6.1.11 Reactor Assistants These are individuals who are present during reactor operation to provide assistance tothe Operator as needed, with the exception that a Reactor Assistant does not operate thecontrols of the reactor.

In an emergency, or if asked, they may push the Reactor Scrambutton.Revised November 201425 6.1.12 Operating Staffa. The minimum operating staff during any time in which the reactor is not securedshall consist of all of the following:

1. One Reactor Operator or Reactor Supervisor in the reactor room.2. One other person in the reactor room or Nuclear Reactor Laboratory qualified to activate manual scram and initiate emergency procedures.

3. One radiation safety staff member who can be readily contacted by telephone and who can arrive at the reactor room within 30 minutes.4. One Reactor Supervisor readily available on call. This requirement can besatisfied by having a licensed Reactor Supervisor perform the duties stated inparagraph 1 or 2 above or by designating a licensed Reactor Supervisor whocan be readily contacted by telephone and who can arrive at the reactorfacility within 30 minutes:b. A Senior Reactor Operator shall supervise all reactor maintenance or modification that could affect the reactivity of the reactor.c. A listing of reactor facility personnel by name and phone number shall beconspicuously posted in the reactor room.6.2 Staff Qualifications The Chief Reactor Supervisor, licensed Reactor Supervisors and Reactor Operators, andtechnicians performing reactor maintenance shall meet the minimum qualifications setforth in ANSI 15.4, "Standards for Selection and Training of Personnel for ResearchReactors".

Reactor Safety Advisory Committee members shall have a minimum of five(5) years experience in a technical profession or a baccalaureate degree and two (2) yearsof professional experience.

The Radiation Safety Officer shall have a baccalaureate degree in biological or physical science and have at least two (2) years experience inhealth physics.Revised November 201426 6.3 TrainingThe Reactor Administrator shall be responsible for directing training as set forth in ANSI15.4-2007, "Standards for Selection and Training of Personnel for Research Reactors".

All licensed reactor operators shall participate in requalification training as set forth in 10CFR 55.6.4 Reactor Safety Advisory Committee 6.4.1 Meetings and QuorumThe Reactor Safety Advisory Committee shall meet as often as deemed necessary by theReactor Safety Advisory Committee Chair but shall meet at least semiannually (interval not to exceed seven and one-half months).

A quorum for the conduct of official businessshall be three members.6.4.2 ReviewsThe Reactor Safety Advisory Committee shall review:a. Safety evaluations for changes to procedures, equipment or systems, and tests orexperiments, conducted without Nuclear Regulatory Commission approval underthe provision of 10 CFR 50.59 to verify that such actions do not require a licenseamendment.

b. Proposed changes to or additional procedures, new or existing equipment orsystems that change the original intent or use, and are non-conservative, or thosethat are covered in 10 CFR 50.59.c. Proposed tests or experiments which are significantly different from previousapproved tests or experiments, or those that are covered in 10 CFR 50.59.d. Proposed changes in Technical Specifications or other license documents.

e. Violations of applicable

statutes, codes, regulations, orders, Technical Specifications, license requirements, or internal procedures or instructions havingnuclear safety significance.

f. Significant operating abnormalities or deviations from normal and expectedperformance of facility equipment that affect nuclear safety.g. Reportable occurrences.

h. Audit reports.Revised November 201427 6.4.3 AuditsAudits of facility activities shall be performed at least annually (interval not to exceed 15months) under the cognizance of the Reactor Safety Advisory Committee but in no caseby the personnel responsible for the item audited.

These audits shall examine theoperating records and encompass, but shall not be limited to, the following:

a. The conformance of the facility operation to the Technical Specifications andapplicable license conditions, at least annually (interval not to exceed 15 months).b. The Facility Emergency Plan and implementing procedures, at least every twoyears (interval not to exceed 30 months).c. The Facility Security Plan and implementing procedures, at least every two years(interval not to exceed 30 months).d. Operator requalification program and records, at least every two years (interval not to exceed 30 months).e. Results of actions taken to correct deficiencies, at least annually (interval not toexceed 15 months).f. Deficiencies uncovered that affect reactor safety shall immediately be reported toLevel 1 management.

A written report of the findings of the audit shall besubmitted to Level 1 management and the review and audit group members within3 months after the audit has been completed.

6.4.4 Authority

The Reactor Safety Advisory Committee shall report to the Nuclear Engineering Department Chair and shall advise the Reactor Administrator the Chief ReactorSupervisor on those areas of responsibility outlined in Section 6.1.5 of these Technical Specifications.

6.4.5 Minutes of the Reactor Safety Advisory Committee One member of the Reactor Safety Advisory Committee shall be designated to direct thepreparation, maintenance, and distribution of minutes of its activities.

These minutes shallinclude a summary of all meetings, actions taken, audits, and reviews.

Minutes shall bedistributed to all RSAC members, all administrative levels, and the Radiation SafetyOfficer within 2 months (interval not to exceed 10 weeks) after each meeting.Revised November 201428

6.5 Approvals

The procedure for obtaining approval for any change, modification, or procedure whichrequires approval of the Reactor Safety Advisory Committee is as follows:a. The Chief Reactor Supervisor shall prepare the proposal for review and approvalby the Reactor Administrator.

b. The Reactor Administrator shall submit the proposal to the Reactor SafetyAdvisory Committee for review, comment, and possible approval.

c. The Reactor Safety Advisory Committee shall approve the proposal by majorityvote.d. The Reactor Administrator shall provide final approval after receiving theapproval of the Reactor Safety Advisory Committee.

6.6 Procedures

There shall be written procedures that cover the following activities:

a. Startup, operation, and shutdown of the reactor.b. Fuel movement and changes to the core and experiments that could affectreactivity.

c. Conduct of irradiations and experiments that could affect the operation or safetyof the reactor.d. Preventive or corrective maintenance which could affect the safety of the reactor.e. Routine reactor maintenance.

f. Radiation Safety Protection for all reactor related personnel.

g. Surveillance, testing and calibration of instruments, components, and systems asspecified in Section 4.0 of these Technical Specifications.

h. Implementation of the Security Plan and Emergency Plan.The above listed procedures shall be approved by the Reactor Administrator and theReactor Safety Advisory Committee.

Temporary procedures which do not change theintent of previously approved procedures and which do not involve a I OCFR50.59 reviewmay be employed on approval by the Chief Reactor Supervisor.

Revised November 201429

6.7 Experiments

a. Prior to initiating any new reactor experiment, an experimental procedure shall beprepared by the Chief Reactor Supervisor and reviewed and approved by theReactor Safety Advisory Committee.

b. Experiments shall only be performed under the cognizance of the Chief ReactorSupervisor.

6.8 Safety Limit Violations The following actions shall be taken in the event a Safety Limit is violated:

a. The reactor will be shut down immediately and reactor operation will not beresumed without authorization by the Nuclear Regulatory Commission (NRC).b. The Safety Limit Violation shall be reported to the NRC Operations Center, theDirector of NRR, the Reactor Safety Advisory Committee, and ReactorAdministrator not later than the next work day.c. A Safety Limit Violation Report shall be prepared for review by the ReactorSafety Advisory Committee.

This report shall describe the applicable circumstances preceding the violation, the effects of the violation upon facilitycomponents,

systems, or structures, and corrective action to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the NRC and the ReactorSafety Advisory Committee within 14 days of the violation.

6.9 Reporting Requirements In addition to the applicable reporting requirements of Title 10, Code of FederalRegulations, the following reports shall be submitted to the Document Control Desk,USNRC, Washington D.C., 20555.Revised November 201430 6.9.1 Annual Operating ReportRoutine annual operating reports shall be submitted no later than ninety (90) daysfollowing June 30. The annual operating reports shall provide a comprehensive summaryof the operating experience having safety significance gained during the year, eventhough some repetition of previously reported information may be involved.

References in the annual operating report to previously submitted reports shall be clear.Each annual operating report shall include:a. A brief narrative summary of1. Changes in facility design, performance characteristics, and operating procedures related to reactor safety that occurred during the reporting period.2. Results of major surveillance tests and inspections..

b. A tabulation showing the hours the reactor was operated and the energy produced bythe reactor in watt-hours.

c. List of the unscheduled shutdowns, including the reasons therefore and corrective action taken, if any.d. Discussion of the major safety related corrective maintenance performed during theperiod, including the effects, if any, on the safe operation of the reactor, and the reasonsfor the corrective maintenance required.

e. A brief description of:1. Each change to the facility to the extent that it changes a description of thefacility in the application for license and amendments thereto.2. Changes to the procedures as described in Facility Technical Specifications.

3. Any new experiments or tests performed during the reporting period.f. A summary of the safety evaluation made for each change, test or experiment notsubmitted for NRC approval pursuant to 10 CFR 50.59 which clearly shows the reasonleading to the conclusion that no license amendment was required and that no Technical Specifications change was required.

Revised November 201431

g. A summary of the nature and amount of radioactive effluent released or discharged tothe environs beyond the effective control of the licensee as determined at or prior to thepoint of such release or discharge.

1. Liquid Waste (summarized for each release)a. Total estimated quantity of radioactivity released (in Curies) and totalvolume (in liters) of effluent water (including diluent) released.

2. Solid Waste (summarized for each release)a. Total volume of solid waste packaged (in cubic meters)b. Total activity in solid waste (in Curies)c. The dates of shipment and disposition (if shipped off site).h. A description of the results of any environmental radiation surveys performed outsidethe facility.

i. Radiation Exposure

-A summary of personnel exposures received during the reporting period by facility personnel and visitors.

Revised November 201432 6.9.2 Reportable Occurrences Reportable occurrences, including causes, probable consequences, corrective actions andmeasures to prevent recurrence, shall be reported to the NRC as described in Section 6.9.Supplemental reports may be required to fully describe final resolution of the occurrence.

In case of corrected or supplemental

reports, an amended licensee event report shall becompleted and reference shall be made to the original report date.a. Prompt Notification with Written Follow-up The types of events listed below are considered reportable occurrences and shallbe reported as expeditiously as possible by telephone and confirmed by facsimile transmission to the NRC Operations Center no later than the first work dayfollowing the event, with a written follow-up report within two weeks asdescribed in Section 6.9. Information provided shall contain narrative material toprovide complete explanation of the circumstances surrounding the event.1. Failure of the reactor protection system or other systems subject to limitingsafety system settings to initiate the required protective function by the time amonitored parameter reached the set point specified as the limiting safety systemsetting in the Technical Specifications or failure to complete the requiredprotective function.

2. Operation of the reactor or affected systems when any parameter or operation subject to a limiting condition is less conservative than the limiting condition foroperation established in the Technical Specifications

-without taking permitted remedial action.3. Abnormal degradation discovered in a fission product barrier.4. Reactivity balance anomalies involving:

a. Disagreement between expected and actual critical rod positions ofapproximately 0.3% Ak/k.b. Exceeding excess reactivity limit.c. Shutdown margin less conservative than specified in Technical Specifications.

d. If sub-critical, an unplanned reactivity insertion of more thanapproximately 0.5% Ak/k or any unplanned criticality.

5. Failure or malfunction of one (or more) component(s) which prevents or couldprevent, by itself, the fulfillment of the functional requirements of system(s) usedto cope with accidents analyzed in the Safety Analysis Report.Revised November 201433

6. Personnel error or procedural inadequacy which prevents, could prevent, byitself, the fulfillment of the functional requirements of system(s) used to copewith accidents analyzed in the Safety Analysis Report.7. Unscheduled conditions arising from natural or manmade events that, as adirect result of the event, require reactor shutdown, operation of safety systems, orother protective measures required by Technical Specifications.

8. Errors discovered in the analyses or in the methods used for such analyses asdescribed in the Safety Analysis Report or in the bases for the. Technical Specifications that have or could have permitted reactor operation in a mannerless conservative than assumed in the analysis.

9. Release of radiation or radioactive materials from site above allowed limits.10. Performance of structures,

systems, or components that requires remedialaction or corrective measures to prevent operation in a manner less conservative than assumed in the accident analysis in the SAR or Technical Specifications thatrequire remedial action or corrective measures to prevent the existence ordevelopment of an unsafe condition.

6.9.3 Special ReportsSpecial reports which may be required by the Nuclear Regulatory Commission shall besubmitted to the Director, Office of Nuclear Reactor Regulation, USNRC within the timeperiod specified for each report. This includes personnel changes in Level 1 (University President),

2 (Reactor Administrator) or 3 (Chief Reactor Supervisor) administration, asshown in Figure 1, which shall be reported within 30 days of such a change.Revised November 201434 6.10 Record Retention 6.10.1 Records to be Retained for a Period of at Least Five Yearsa. Operating logs or data which shall identify:

1. Completion of pre-startup check-out,

startup, power changes, and shutdown ofthe reactor.2. Installation or removal of fuel elements, control rods, or experiments thatcould affect core reactivity.

3. Installation or removal of jumpers, special tags or notices, or other temporary changes to reactor safety circuitry.

4. Rod worth measurements and other reactivity measurements.

b. Principal maintenance operations.

c. Reportable occurrences.

d. Surveillance activities required by Technical Specifications.

e. Facility radiation and contamination surveys.f. Experiments performed with the reactor.

This requirement may be satisfied by thenormal operations log book plus,1. Records of radioactive material transferred from the facility as required bylicense.2. Records required by the Reactor Safety Advisory Committee for theperformance of new or special experiments.

g. Records of training and qualification for members of the facility staff.h, Changes to operating procedures.

Revised November 201435 6.10.2 Records to be Retained for the Life of the Facilitya. Records of liquid and solid radioactive effluent released to the environs.

b. Off-site environmental monitoring surveys.c. Fuel inventories and fuel transfers.

d. Radiation exposures for all personnel.

e. Drawings of the facility.

f. Records of reviews performed for changes made to procedures or equipment or reviewsof tests and experiments pursuant to 10 CFR 50.59.g. Records of meetings of the Reactor Safety Advisory Committee, and copies of RSACaudit reports.h. Records of the review of:* Violations of any safety limit (SL)0 Violations of any limiting safety setting (LSSS)* Violations of any limiting condition of operation (LCO)Revised November 201436