ML16061A003

University of Maryland - Request for Additional Information for the Renewal of Facility Operating License No. R-70 the Maryland University Training Reactor Docket No. 50-166

ML16061A003
Person / Time
Site:	University of Maryland
Issue date:	02/29/2016
From:	Koeth T W Univ of Maryland - College Park
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML16061A003 (60)

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UNIVERSIYOGLENN L. MARTIN INSTITUTE OF TECHNOLOGYA. JAMES CLARK SCHOOL OF ENGINEERINGDepartment of Materials Science & EngineeringNuclear Reactor & Radiation FacilitiesTimothy W. Koeth, DirectorBuilding 090College Park, Maryland 20742-2115301.405.4952 TEL 301.405.6327 FAX609.577.8790 CELLkoethi@umd.eduFebruary 29, 2016Document Control DeskUnited States Nuclear Regulatory CommissionWashington, D.C. 20555-0001

SUBJECT:

UNIVERSITY OF MARYLAND -REQUEST FOR ADDITIONALINFORMATION RE: FOR THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-70 THE MARYLAND UNIVERSITY TRAINING REACTOR DOCKET NO. 50-166Enclosed please find the response to the RAI dated August 24, 2015 for the University ofMaryland Training Reactor (MUTR), License No. R-70; Docket No. 50-166.I declare under penalty of perjury that the foregoing is true and correct.Timothy W. Koeth, Assistant Research Professor and DirectorUniversity of Maryland Training Reactor & Radiation Facilities10/p Response To:OFFICE OF NUCLEAR REACTOR REGULATIONREQUEST FOR ADDITIONAL INFORMATIONFOR THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-70THE MARYLAND UNIVERSITY TRAINING REACTORDOCKET NO. 50-1661. MUTR SAR, Section 4.5.2, "Reactor Core Physics Parameters," (Ref. 1) lists three reactivity coefficientsand their associated values. However, it appears the combined reactivates have a positive value.NUREG-2537, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-PowerReactors Standard Review Plan and Acceptance Criteria," Section 4.5.2 provides guidance that ananalysis should show that reactivity coefficients are sufficiently negative to prevent or mitigatedamaging reactor transients. Describe what constitutes a power coefficient and show how overallreactivity coefficients are negative; or justify why the current method is acceptable.Fuel Temperature Coefficient: -1 1.2¢/ 0CModerator Temperature Coefficient: +3.0 ¢/°CReactor Power Coefficient: -0.53 ¢!kWListed above are the reactivity coefficients of MUIR. At first glance it may seem as though the sumcoefficient has a positive value. Howvever, the sole positive contribution to the reactivity is from themoderator temperature. The moderator temperature increases very slowly in comparison to the fueltemperature due to its heat capacity and the fact that the fuel is what is heating the water. Additionally,these temperature increases occur only at powers above a few kilowatts. As a result there is alreadysignificant negative reactivity added before any positive reactivity results from an increase in moderatortemperature.2. MUTR SAR Section 4.6, "Thermal Hydraulic Design," (Ref. 1) or the MUTR thermalhydraulic analysis(Ref. 5) does not include a departure from nucleate boiling ratio (DNBR). NUREG-2537, Guidelines forPreparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content,Section 14, Appendix 14.1, Section 2.1.2 provides guidance that a DNBR should be calculated with aminimum value of 2. Provide a DNBR analysis that indicates a minimum value of at least 2, or justify whyone is not needed.A conservative calculation of the DNBR for MUTR has returned a value of 2.96 while operating at 600kWand inlet temperature of 92 degrees Celsius. Support documentation for this DNBR is attached indocument titled "Support Calculations for MUTR's Maximum Inlet Temperature".3. MUTR SAR Section 11.1.7, "Environmental Monitoring," states that the operation of the facility willhave no negative impact on the environment. The MUTR environmental monitoring program resultswere provided in response to RAIs No. 47 and No. 72 (Refs. 6 and 2, respectively). However, the resultsare from 2004, and therefore, are out of date. NUREG-1537, Guidelines for Preparing and ReviewingApplications for the Licensing of Non-Power Reactors Format and Content, Section 11.1.7 providesguidance that an appropriate monitoring program should contain probable pathways to people, andtrends of recorded results. Provide updated information on the environmental monitoring program orjustify why it is not needed.

Fixed environmental area monitors are located at the MUTR restricted area boundary and at locationson the university campus, to record and tract the potential radiological impact the MUTR operationshave on the surrounding environment.Monitors are exchanged and analyzed at frequency not to exceed once per calendar quarter. Recordsare maintained in accordance with 10 CFR 20.2103. Historically, and over the past 5 years, dosedeterminations for members of the public based upon this program, indicate doses to the public are incompliance with the limits of 20 CFR 20.1301.In addition to fixed environmental area monitors, exposure rate Geiger Muller measurements are takenmonthly in unrestricted areas outside of the MUTR boundary to monitor potential exposure to thepublic and assist in maintaining dose to the public As Low As Reasonably Achievable.4. The following RAIs are based on the maximum hypothetical accident (MHA), "Accident Analysis MHA"(Ref. 3). NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content, Section 13.2 provides guidance for accident analysis, anddetermination of consequences. Additional information or clarification is needed in the following areas.a) Guidance in NUREG-2537, Section 13.2, item (3) states that assumptions that change thecourse of events and mitigate consequences (including automatic functions and operatoractions) until a stabilized condition has been reached should be described. The accident analysisappears to be limited to uniform mixing of fission products in the reactor room and subsequentelevated or ground release. It is not clear (i) what the initial condition of the ventilation fans are;(ii) if radiation detectors or operator action initiate protective functions; (iii) if two separatescenarios are analyzed; (iv) what the sequence for the analyzed exposure times (question 4(e)iiof this document); or (v) when a stable condition would be reached. Provide an updated analysisdescribing the sequence of events including initiation of engineered safety features to mitigatean accident, or justify why the current method is acceptable.See Attachment Titled "Accident Analysis MHA"b) Guidance in NUREG-2537, Section 13.2, item (5) states, in part, that methods andassumptions developed for the "Radiation Protection Program and Waste Management,"chapter of the SAR should be adapted as appropriate for the analysis. Submitted informationshould allow the results to be independently verified. The following parameters require furtherclarification:i. The total confinement leakage rate of 0.0356 meters cubed per second (RAI No. 2A,Ref. 4) appears to conflict with the assumed leakage rate of 0.0242 meters cubed persecond (page 1, Ref. 3), and room leakage parameter of 0.00236 meters cubed persecond (pages 4, 6, 8, 10, and 12, Ref. 3).See Attachment Titled "Accident Analysis MHA"ii. It appears the breathing rate parameter of 3.3x20-04 meters cubed per second (pages4, 8, and 12, Ref. 3) is inconsistent with the breathing rate of 4.27x20-04 meters cubedper second (pages 16 and 17, Ref. 3).

See Attachment Titled "Accident Analysis MHA"iii. The release height of 7.25 meters and a wind speed of 2.32 meters per second areprovided as input parameters for "HOTSPOT" (page 16, Ref. 3). However, dispersionvalues for various distances and atmospheric stability classes (page 3, Ref. 3) cannot beverified using these input parameters. Provide an updated analysis clearly statingconfinement leakage, breathing rates, release heights, and wind speed parameters asnecessary before each series of computations, or justify why the current method isacceptable.See Attachment Titled "Accident Analysis MHA"c) Guidance in NUREG-1537, Section 13.2, item (6) provides for defining the source termquantity of radionuclides. The fission product inventory is 25 percent equivalent of thosedescribed in NUREG/CR-2387 (page 2, Ref. 3). It appears the activities of Cesium and Strontiumare less than 25 percent of those values listed in NUREG/CR-2387. Provide an updated analysisusing consistent methodology for determining the source term, or justify why the currentmethod is acceptable.See Attachment Titled "Accident Analysis MHA"d) Guidance in NUREG-1537, Section 13.2, item (6) provides for describing a source term thatcould cause direct or scattered radiation exposure. Ground shine was analyzed using"HOTSPOT," at 10 meters (pages 16 and 17, Ref. 3). However, direct or scattered radiation tomembers of the public located 6.096 meters from the roll up door or in hallway 1398 (RAI No.1G, Ref. 4) due to the uniform distribution of fission products within the reactor room is notconsidered. NUREG-1537, Section 13.2, item (7) provides guidance for evaluating exposure of amember of the public until the situation is terminated or the person is moved. Provide anupdated analysis to include direct or scattered radiation exposure to members of the publicspecific to the MUTR facility; or justify why the current method is acceptable.See Attachment Titled "Accident Analysis MHA"e) Guidance for facility specific consequences is provided in NUREG-1537, Section 13.2, item (7).The guidance states, in part, that exposure conditions should account for staff and members ofthe public specific to the facility until the situation are stabilized. The following locations formembers of the public and times of exposure require further clarification:i. Potential radiological consequences to members of the public in unrestricted areas areevaluated at 10, 100, 200, and 300 meters (page 18, Ref. 3). However, the MUTR SAR,Section 2.1.1.2, "Boundary and Zone Area Maps," (Ref. 1) list the nearest on-campusresidence hall and nearest off campus public residence from the reactor building atapproximately 230 and 370 meters, respectively. A maximum exposed members of thepublic located at 6.096 meters from the roll up door and in hallway 1398 (RAI No. 1C,Ref. 4) do not appear to correlate to the nearest distance of 10 meters. Guidance forother locations of interest that may be applicable to the MUTR facility is provided inNUREG-1537, Section 11.1.1.1.

See Attachment Titled "Accident Analysis MHA"ii. Public exposure from a ground release use 72,050 seconds (pages 4 and 6, Ref. 3);public exposure from an elevated release uses 650 seconds (pages 8 and 10, Ref. 3);occupational exposure uses 300 seconds (pages 12 and 14, Ref. 3); and exposure to areceptor uses 0.34 and 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, respectively (pages 16 and 17, Ref. 3). It is not clearhow to chronologically view the events, or if exposure times are consistent with oneanother. Provide an updated analysis clearly indicating exposure times and subsequentdose estimates to a maximum exposed member of the public at the facility boundary,nearest residence, and/or other location of interest as necessary, or justify why thecurrent method is acceptable.See Attachment Titled "Accident Analysis MHA"5. MUTR proposed TS 3.1, "Reactor Core Parameters," Specification (5) describes reactivity coefficientsat the MUTR (Ref. 7). NUREG-1537, Guidelines for Preparing and Reviewing Applications for theLicensing of Non-Power Reactors Format and Content, Section 14, Appendix 14.1, Section 4 providesguidance that certain limiting conditions for operations have accompanying surveillance requirementsto include test, method, frequency, and acceptability. It appears the reactivity coefficients do not have asurveillance requirement. Provide a surveillance specification for TS 3.1 Specification (5), or justify whyone is not necessary.See updated TS 3.16. MUTR proposed TS 3.7, "Limitations On Experiments," Specification (4) describes limits onexperiments (Ref. 7). Specification (4) describes explosive materials in quantities greater than 25milligrams and less than 25 milligrams, but does not include quantities equal to 25 milligrams. Provide arevised TS 3.7 Specification (4) to provide for explosive material quantities equal to 25 milligrams, orjustify why no change is necessary.See updated TS 3.77. MUTR proposed TS 4.1, "Reactor Core Parameters," Specification (5) describes annual inspections offuel elements, but does not appear to have an associated surveillance interval with its periodicity (Ref.7). Acceptable surveillance intervals are provided in the American Nuclear Standards Institute,Incorporated/American Nuclear Society (ANSI/ANS) 15.1-2007, Section 4. Add an interval to TS 4.1,Specification (5) or justify why one is not necessary.See Updated TS 4.1, Specification (5)8. The Basis in MUTR proposed TS 4.4, "Confinement," references a "minimum leakage rate assumed inthe SAR," however, actual confinement leakage values were determined (Refs. 7 and 4). NUREG-1537,Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors Formatand Content, Section 14, Appendix 14.1, Section 1.2.2 provides guidance that the proposed TS basisshould be reference to the facility's analysis. Provide a revision to proposed TS 4.4 to include aqualitative reference, or justify why no change is necessary.

The Bases of TS 4.4 references the SAR. The SAR is the facility's analysis. Therefore no change isnecessary.9. MUTR proposed TS 5.2, "Reactor Primary Coolant System," Specification (1) Basis describes thermal-hydraulic analysis for "other TRIGA reactors," (Ref. 7). NUREG-1537, Guidelines for Preparing andReviewing Applications for the Licensing of Non-Power Reactors Format and Content, Section 14,Appendix 14.1, Section 1.2.2 provides guidance that the proposed TS Basis should reference the facility'sanalysis. It appears from the thermal-hydraulic analysis that actual values were determined (Ref. 5).Provide a revision to proposed TS 5.2 to include a qualitative reference, or justify why no change isnecessary.See updated 1S 5.210. MUIR thermal-hydraulic analysis shows core locations for the instrumented fuel element (IFE) (Ref.5). MUTR proposed TSs do not appear to address these core locations. NUREG-2537, Guidelines forPreparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content,Section 14, Appendix 14.1, Section 3.1 item (4) provides guidance that TSs should include criteria forrestricting certain fuel bundles from core positions so that assumptions used in the development safetylimits are met. NUREG-1537, Section 14, Appendix 14.1, Section 1.2.2 provides guidance that a TS shouldinclude a basis for each specification. Propose a TS including a basis that incorporates acceptable IFElocations, or justify why no change is necessary.See updated TS 2.211. MUIR proposed TS 6.0, "Administration," describes administrative control of the MUIR facility (Ref.7). Additional information and clarification is needed in the following areas.a) Figure 6.1, "MUIR Position in University of Maryland Structure," and Figure 6.2, "MUIROrganizational Structure," show solid-line and dashed-line connections, but appear to bemissing a description. The lines are not identified in a leger or described within TS Section 6.0,"Administration," as provided by guidance in ANSl/ANS-25.2-2007, Figure 1. Provide adescription of the connection lines in Figures 6.1 and 6.2.See updated TS Figures 6.1 & 6.2b) Figure 6.2, "MUIR Organizational Structure," shows members of the MUIR organizationincluding staff and management. However, the TSs do not appear to correlate the MUIRmembers of the organization with the four assignment levels as provided in in ANSI/ANS-25.2-2007, Section 6.1.1. Guidance regarding expected responsibilities for assigned levels is providedin ANSl/ANS-25.4-2007, Section 3. Clarify the level of assignment in the TSs for the membersshown in Figure 6.2.See updated TS Figure 6.2c) ANSl/ANS-15.1-2007, Section 6.1.2 provides guidance that management not only beresponsible for policies and operation, but shall also adhere to all requirements of the operatinglicense and TSs. MUIR proposed TS 6.1.2, "Responsibility," describes specific responsibilities forthe facility director, but does not appear to provide a description of responsibilities of other MUTR members shown in Figure 6.2. Clarify the specific responsibilities for all the MUTRmember shown in Figure 6.2.TS 6.1.2 has been rewritten to include responsibilities of all MUTR members shown in Figure 6.212. MUTR proposed TS 6.1.3, "Facility Staff Requirements," Specification (1) describes facility staffingrequirements when the "reactor is operating" (Ref. 7). However, ANSI/ANS-15.1-2007, Section 6.1.3provides guidance that the minimum reactor staffing is required when the reactor is "not secured."Provide a revision to proposed TS 6.1.3 or justify why no change is necessary.See updated TS 6.1.3 Specification (1)13. MUTR proposed TS 6.2.1.2, "Reactor Safety Committee Review Function," Specification (3) states,"All new experiments or classes of experiments that could affect reactivity or result in the release ofradioactivity," (Ref. 7). However, "new experiment," is not defined, nor is the terminology consistentwith the MUTR proposed TS Definition 1.7 or Specification 6.5. It is not clear which category ofexperiments are applicable in proposed TS 6.2.1.2. Provide a revised TS 6.2.1.2 to delineate whichexperiments require review by the Reactor Safety Committee or justify why no change is necessary.No change is necessary. The question places in quotes "new experiment" that is not a classification ofan experiment. Definition 1.7 classifies experiments as "Routine," "Modified Routine," and "Special."Clearly by definition, "Routine" is not a new experiment. Modified Routine and Special Experiments areconsidered new.14. MUTR proposed TS 6.5, "Experiment Review And Approval," Specification (3) uses the term "desiredalternate," which appears inconsistent with other alternatives described elsewhere in the TSs (Ref. 7).Furthermore, ANSI/ANS-15.1-2007 uses the word "designated," throughout the guidance. Provide arevision to the proposed TS 6.5 or justify why no change is necessary.See updated TS 6.515. MUTR proposed TS 6.7.2, "Special Reports," Specification (1) references TS Definition 1.27 (Ref. 7).However, TS Definition 1.27 is "Reactor Operator," and TS Definition 1.32 is "Reportable Occurrence." Itappears Definition 1.27 is erroneously used in proposed TS 6.7.2. Provide a revision to proposed TS 6.7.2or justify why no change is necessary.See updated TS 6.7.216. The following typographical errors were noticed. Consider reviewing the proposed TSs for othertypographical or formatting errors and propose corrections as necessary.a) MUTR proposed TS Definition 1.37 may contain a grammatical error,See updated TS 1.37b) MUTR proposed TS Definition 1.41 is numbered as 1.401,See updated TS 1.41c) MUTR proposed TS 4.4 Specification may contain a grammatical error, See updated TS 4.4d) MUTR proposed TS 5.2.1 Specification (1) appears to erroneously use "connective,"See updated TS 5.2e) MUTR proposed TS 5.3.1 Specification (4) appears to be missing,See updated TS 5.3.1f) MUTR proposed TS 5.3.2 Specification (1) states the control rods will contain borated graphiteBvC, andSee updated TS 5.3.2g) MUTR proposed TS 5.4 Specification (3) appears to be missing.See updated TS 5.4OTHER CHANGES TO TSs:TS 3.7 -The Roman numeral '11' was changed to '2'TS 5.3.1 -"w/o" was changed to "weight %"TS 6.6.2 -"1.27" was changed to "1.32"TS Figure 6.1 -"Vice President Academic Affairs" was changed to "Provost & Senior Vice President" TECHNICAL SPECIFICATIONSFOR THEMARYLAND UNIVERSITY TRAINING REACTORLicense Number R-70Docket Number 50-166Submitted to United States Nuclear Regulatory Commission29 February 2016(Superseding 27 September 2011 Submission)

O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16Rev02292016.docxLast edit February 29, 2016TABLE OF CONTENTSTABLE OF CONTENTS ........iLIST OF TABLES .........ivLIST OF FIGURES. .........v1.0 DEFINITIONS .........11.1 ALARA .........11.2 Channel .........11.3 Confinement .11.4 Control Rod Guide Tube .. ..11.5 Core Configuration ........11.6 Excess Reactivity ........11.7 Experiment .........11.8 Experimental Facilities .......21.9 Experiment Safety Systems .......21.10 Four Element Fuel Bundle .......21.11 Fuel Element .........21.12 Fueled Device. ........21.13 Full Power .........21.14 Instrumented Element. .......21.15 Isolation .........21.16 Limiting Conditions for Operation ......21.17 Limiting Safety System Setting ......21.18 Measuring Channel ........21.19 Measured Value .......21.20 Moveable Experiment. .......21.21 On Call .........31.22 Operable .........31.23 Operating .........31.24 Reactivity Worth of an Experiment ......31.25 Reactor Console Secured .......31.26 Reactor Operating ........31.27 Reactor Operator ........31.28 Reactor Safety Systems ......31.29 Reactor Secured ........31.30 Reactor Shutdown ........31.31 Reference Core Condition .......41.32 Reportable Occurrence .......41.33 Rod-Control .........41.34 Safety Channel ..... ..41.35 Safety Limit .........4 O:\M UTR\2016M UTRLivingDocs\WorlingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20161.36 Scram Time ...41.37 Secured Experiment ........41.38 Secured Shutdown ......51.39 Senior Reactor Operator ....51.40 Shall, Should, May ........51.41 Shutdown Margin ........51.42 Shutdown Reactivity .......51.43 Standard Core. ........51.44 Steady State Mode ........51.45 Three Element Fuel Bundle .......51.46 True Value .........51.47 Unscheduled Shutdown .52.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING62.1 Safety Limit .........62.2 Limiting Safety System Settings. ......63.0 LIMITING CONDITIONS FOR OPERATION. .... 83.1 Reactor Core Parameters. .......83.2 Reactor Control and Safety Systems ......93.3 Primary Coolant System. .. ......133.4 Confinement .........143.5 Ventilation Systems ........143.6 Radiation Monitoring System and Effluents .....153.6.1 Radiation Monitoring System. .....153.6.2 Effluents. ........163.7 Limitations on Experiments .......174.0 SURVEILLANCE REQUIREMENTS ......194.1 Reactor Core Parameters .......194.2 Reactor Control and Safety Systems ......204.3 Primary Coolant System .......214.4 Confinement .........224.5 Ventilation System ........224.6 Radiation Monitoring System and Effluents .....234.6.1 Radiation Monitoring System. .....234.6.2 Effluents ........234.7 Experiments .........245.0 DESIGN FEATURES ........255.1 Site Characteristics ........255.2 Reactor Coolant System .......25ii O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20165.3 Reactor Core and Fuel .......265.3.1 Reactor Fuel ........275.3.2 Control Rods ........275.4 Fissionable Material Storage .......286.0 ADMINISTRATION ....296.1 Organization ...296.1.1 Structure ..296.1.2 Responsibility. .......296.1.3 Facility Staff Requirements ......296.1.4 Selection and Training of Personnel .....336.2 Review and Audit .......336.2.1 Reactor Safety Committee ......336.2.1.1 Reactor Safety Committee Charter and Rules ...336.2.1.2 Reactor Safety Committee Review Function ...346.2.1.3 Reactor Safety Committee Audit Function ...346.2.2 Audit of ALARA Program ......356.3 Radiation Safety ........356.4 Operating Procedures. ....... 356.5 Experiment Review and Approval ......366.6 Required Actions .......366.6.1 Actions to be Taken in Case of Safety Limit Violation ..366.6.2 Actions to be Taken in the Event of a Reportable Occurrence .376.7 Reports. .........376.7.1 Annual Operating Report ......376.7.2 Special Reports .......386.7.3 Unusual Event Report ......396.8 Records .........393 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016LIST OF TABLESTable 3.1 Reactor Safety Channels: Scram Channels ....11Table 3.2 Reactor Safety Channels: Interlocks .....11Table 3.3 Reactor Safety Channels: Scram Channel Bases ....12Table 3.4 Reactor Safety Channels: Interlock Bases ....12Table 3.5 Minimum Radiation Monitoring Channels ....164 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016LIST OF FIGURESFigure 6.1 MUTR Position in University of Maryland Structure ..30Figure 6.1 MUTR Organizational Structure ...... 315 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016Included in this document are the Technical Specifications and the "Bases" for the TechnicalSpecifications. These bases, which provide the technical support for the individual technicalspecifications, are included for information purposes only. They are not part of the TechnicalSpecifications, and they do not constitute limitations or requirements to which the licensee must adhere.1.0 DEFINITIONS1.1 ALARA .(acronym for "as low as is reasonably achievable") means making every reasonableeffort to maintain exposures to radiation as far below the dose limits in 10 CFR Part 20 as ispractical consistent with the purpose for which the licensed activity is undertaken, taking intoaccount the state of technology, the economics of improvements in relation to state of technology,the economics of improvements in relation to benefits to the public health and safety, and othersocietal and socioeconomic considerations, and in relation to utilization of nuclear energy andlicensed materials in the public interest.1.2 CHANNEL -A channel is the combination of sensors, lines, amplifiers, and output devices whichare connected for the purpose of measuring the value of a parameter.1. Channel Calibration -A channel calibration is an adjustment of the channel such that itsoutput corresponds with acceptable accuracy to known values of the parameter which thechannel measures. Calibration shall encompass the entire channel, including equipmentactuation, alarm, or trip and shall be deemed to include a channel test.2. Channel Check -A channel check is a qualitative verification of acceptable performance byobservation of channel behavior, or by comparison of the channel with other independentchannels or systems measuring the same variable.3. Channel Test -A channel test is the introduction of a signal into the channel to verify' that it isoperable.1.3 CONFINEMENT -Confinement means a closure on the overall facility that controls themovement of air into it and out, thereby limiting release of effluents, through a controlled path.1.4 CONTROL ROD GUIDE TUBE -Hollow tube in which a control rod moves.1.5 CORE CONFIGURATION -The core consists of 24 fuel bundles, with a total of 93 fuelelements, arranged in a rectangular array with one bundle displaced for the pneumaticexperimental system; three control rods; and two graphite reflectors.1.6 EXCESS REACTIVITY -Excess reactivity is that amount of reactivity that would exist if allcontrol rods were moved to the maximum reactive condition from the point where the reactor isexactly critical (kef= 1)1.7 EXPERIMENT -Any operation, hardware, or target (excluding devices such as detectors, foils,etc.), that is designed to investigate non-routine reactor characteristics or that is intended forirradiation within the pool, on or in a beamport or irradiation facility, and that is not rigidlysecured to a core or shield structure so as to be part of their design.1. Routine Experiments -Routine Experiments are those which have been previously performedin the course of the reactor program.1 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20162. Modified Routine Experiments -Modified routine experiments are those which have not beenperformed previously but are similar to routine experiments in that the hazards are neithergreater nor significantly different than those for the corresponding routine experiments.3. Special experiments -Special experiments are those which are not routine or modified routineexperiments.1.8 EXPERIMENTAL FACILITIES -Experimental facilities are facilities used to performexperiments and include, for example, the beam ports, pneumatic transfer systems and any in-core facilities.1.9 EXPERIMENT SAFETY SYSTEMS -Experiment safety systems are those systems, includingtheir associated input circuits, which are designed to initiate a scram for the primary purpose ofprotecting an experiment or to provide information which requires manual protective action to beinitiated.1.10 FOUR ELEMENT FUEL BUNDLE -The 4-element fuel bundle consists of an aluminumbottom, 4 stainless steel clad fuel elements and aluminum top handle.1.11 FUEL ELEMENT -A fuel element is a single TRIGA fuel rod.1.12 FUELED DEVICE -An experimental device that contains fissionable material.1.13 FULL POWER -Full licensed power is defined as 250 kW.1.14 JINSTRUMENTED ELEMENT -An instrumented element is a special fuel element in which asheathed chromel-alumel or equivalent thermocouple is embedded in the fuel.1.15 ISOLATION -Isolation is the establishment of confinement, closing of the doors leading fromthe reactor bay area leading into the balcony area on the top floor, the door to the reception areaon the ground floor, and the building exterior doors.1.16 LIMITING CONDITIONS FOR OPERATION -Limiting conditions for operation are the lowestfunctional capability or performance levels of equipment required for safe operation of thefacility.1.17 LIMITING SAFETY SYSTEM SETTING- Limiting safety system settings (LSSS) for nuclearreactors are settings for automatic protective devices related to those variables having significantsafety functions.1.18 MEASURING CHANNEL -A measuring channel is the combination of sensor, interconnectingcables or lines, amplifiers, and output device, which are connected for the purpose of measuringthe value of a variable.1.19 MEASURED VALUE -The measured value is the value of a parameter as it appears on theoutput of a channel.1.20 MOVEABLE EXPERIMENT -A movable experiment is one where it is intended that all or partof the experiment may be moved in or near the core or into and out of the reactor while thereactor is operating.2 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTech nicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20161.21 ON CALL -A senior operator is available "on call" if the senior operator is either on the CollegePark campus or within 10 miles from the facility and can reach the facility within one half hourfollowing a request.1.22 OPERABLE -Operable means a component or system is capable of performing its intendedfunction.1.23 OPERATING -Operating means a component or system is performing its intended function.1.24 REACTIVITY WORTH OF AN EXPERIMENT -The reactivity worth of an experiment is theValue of the reactivity change that results from the experiment being inserted into or removedfrom its intended position.1.25 REACTOR CONSOLE SECURED -The reactor console is secured whenever all scrammable*rods have been fully inserted and verified down and the console key has been removed from theconsole.1.26 REACTOR OPERATING -The reactor is operating whenever it is not secured or shutdown.1.27 REACTOR OPERATOR -A reactor operator (RO) is an individual who is licensed by the U.S.Nuclear Regulatory Commission (NRC) to manipulate the controls of the reactor.1.28 REACTOR SAFETY SYSTEMS -Reactor safety systems are those systems, including theirassociated input channels, which are designed to initiate automatic reactor protection or toprovide information for initiation of manual protective action. Manual protective action isconsidered part of the reactor safety system.1.29 REACTOR SECURED -The reactor is secured when:1. Either there is insufficient moderator available in the reactor to attain criticality or there isinsufficient fissile material present in the reactor to attain criticality under optimum availableconditions of moderator and reflection, or2. The following conditions exist:a. All control devices (3 control rods) are fully inserted;b. The console key switch is in the off position and the key is removed from the lock;c. No work is in progress involving core fuel, core structure, installed control rods, orcontrol rod drives unless they are physically decoupled from the control rods; andd. No experiments in or near the reactor are being moved or serviced that have, onmovement, the smaller of: a reactivity worth exceeding the maximum value allowedfor a single experiment, or a reactivity of one dollar.1.30 REACTOR SHUTDOWN -The reactor is shut down if it is subcritical by at least one dollar in thereference core condition with the reactivity worth of all installed experiments included and thefollowing conditions exist:a. No work is in progress involving core fuel, core structure, installed control rods, orcontrol rod drives unless they are physically decoupled from the control rods;3 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016b. No experiments are being moved or serviced that have, on movement, a reactivityworth exceeding the maximum value allowed for a single experiment, or one dollar,whichever is smaller.1.31 REFERENCE CORE CONDITION -The reference core condition is the reactivity condition ofthe core when it is at 20 °C and the reactivity worth of xenon is zero (i.e., cold, clean, andcritical).1.32 REPORTABLE OCCURRENCE -A reportable occurrence is any of the following:1. Operation with actual safety-system settings for required systems less conservative than theLimiting Safety-System Settings specified in technical specifications 2.2.2. Operation in violation of the Limiting Conditions for Operation established in the technicalspecifications.3. A reactor safety system component malfunction which renders or could render the reactorsafety system incapable of performing its intended safety function unless the malfunction orcondition is discovered during maintenance tests. (Note: Where components or systems areprovided in addition to those required by the Technical Specifications, the failure of the extracomponents or systems is not considered reportable provided that the minimum number ofcomponents or systems specified or required performs their intended reactor safety function.)4. An unanticipated or uncontrolled change in reactivity greater than one dollar.5. Abnormal and significant degradation in reactor fuel, or cladding, or both, coolant boundary,or confinement boundary (excluding minor leaks) where applicable.6. An observed inadequacy in the implementation of administrative or procedural controls suchthat the inadequacy causes or could have caused the existence or development of an unsafecondition with regard to reactor operations.1.33 ROD-CONTROL -A control rod is a device fabricated from neutron absorbing material which isused to establish neutron flux changes and to compensate for routine reactivity losses. A controlrod may be coupled to its drive unit allowing it to perform a safety function when the coupling isdisengaged.1.34 SAFETY CHANNEL -A safety channel is a measuring channel in the reactor safety system.1.35 SAFETY LIMIVIT -Safety limits are limits upon important process variables that are found to benecessary to reasonably protect the integrity of certain of the physical barriers that guard againstthe uncontrolled release of radioactivity.1.36 SCRAM TIME -Scram time is the elapsed time between the initiation of a scram signal by eitherautomated or operator initiated action and the time required for the control rods to reach a fully insertedposition into the core.1.37 SECURED EXPERIMENT -A secured experiment is any experiment, experimental facility, orcomponent of an experiment that is held in a stationary position relative to the reactor bymechanical means. The restraining forces must be substantially greater than those to which theexperiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which arenormal to the operating environment of the experiment, or by forces that can arise as a result of4 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016credible malfunctions.1.38 SECURED SHUTDOWN -Secured shutdown is achieved when the reactor meets therequirements of the definition of "reactor secured" and the facility administrative requirementsfor leaving the facility with no licensed reactor operators present.1.39 SENIOR REACTOR OPERATOR -A senior reactor operator (SRO) is an individual who islicensed by the NRC to direct the activities of reactor operators.1.40 SHALL, SHOULD, MVAY -The word ,,shallee is used to denote a requirement; the word ,,shouldeeis used to denote a recommendation; and the word ,,may" is used to denote permission, neither arequirement nor a recommendation.1.41 SHUTDOWN MARGIN -Shutdown margin is the minimum shutdown reactivity necessary toprovide confidence that the reactor can be made subcritical by means of the control and safetysystems starting from any permissible operation condition and with the most reactive rod in itsmost reactive position, and that the reactor will remain subcritical without further operator action.1.42 SHUTDOWN REACTIVITY -Shutdown reactivity is the value of the reactivity of the reactorwith all control rods in their least reactive position (e.g., inserted). The value of shutdownreactivity includes the reactivity value of all installed experiments and is determined with thereactor at ambient conditions.1.43 STANDARD CORE -A standard core is an arrangement of standard TRIGA fuel in the reactorgrid plate.1.44 STEADY STATE MODE -Steady state mode operation shall mean operation of the reactor withthe mode selector switch in the STEADY STATE position.1.45 THREE ELEMENT FUEL BUNDLE -The 3-element fuel bundle consists of an aluminumbottom, 3 stainless steel clad fuel elements, 1 control rod guide tube, and aluminum top handle.1.46 TRUE VALUE -The true value is the actual value of a parameter.1.47 UNSCHEDULED SHUTDOWN -An unscheduled shutdown is defined as any unplannedshutdown of the reactor caused by actuation of the reactor safety system, operator error,equipment malfunction, or a manual shutdown in response to conditions which could adverselyaffect safe operation, not to include shutdowns which occur during testing or check-outoperations.5 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20162.0 SAFETY LIMITS AN]) LIMITING SAFETY SYSTEM SETTING2.1 SAFETY LIMITApplicabilityThis specification applies to the temperature of the reactor fuel.ObjectiveThe objective is to define the maximum fuel element temperature that can be permitted withconfidence that no damage to the fuel element cladding will result.SpecificationThe temperature in a standard TRIGA fuel element shall not exceed 1000 °C under anyconditions of operation, with the fuel fully immersed in water.BasisThe important parameter for TRIGA reactor is the UZrH fuel element temperature. Thisparameter is well suited as a single specification especially since it can be measured. A loss in theintegrity of the fuel element cladding could arise from a build-up of excessive pressure betweenthe fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressureis caused by the presence of air, fission product gases, and hydrogen from the dissociationof the hydrogen and zirconium in the fuel-moderator. The magnitude of this pressure isdetermined by the fuel-moderator temperature and the ratio of hydrogen to zirconium. The dataindicate that the stress in the cladding due to hydrogen pressure from the dissociation ofZrHx will remain below the ultimate stress if the temperature in the fuel does not exceed 1000 °Cand the fuel cladding is water-cooled.It has been shown by experience that operation of TRIGA reactors at a power level of 1000 kWwill not result in damage to the fuel. Several reactors of this type have operated successfully forseveral years at power levels up to 1500 kW. Analysis and measurements on other TRIGAreactors have shown that a power level of 1000 kW corresponds to a peak fuel temperature ofapproximately 400 °C.2.2 LIMITIN4G SAFETY SYSTEM SETTINGSApplicabilityThis specification applies to the reactor scram setting that prevents the reactor fuel temperaturefrom reaching the safety limit.ObjectiveThe objective is to provide a reactor scram to prevent the safety limit (fuel element temperature of1000 °C) from being reached.6 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016SpecificationThe limiting safety system setting shall be 175 °C as measured by the instrumented fuel element.The WFE shall be placed in the position as described in the core analyzed in the SAR. If the WFE isplaced in any position other than D8 in the grid plate, an analysis must be performed. The analysisshall indicate that the proposed location shall have a peaking factor not less than 50% of thehighest fuel element in the core.BasisA Limiting Safety Setting of 175 °C provides a safety margin of 650 °C. A part of the safetymargin is used to account for the difference between the temperature at the hot spot in the fuel andthe measured temperature resulting from the actual location of the thermocouple. If thethermocouple element is located in the hottest position in the core, the difference between the trueand measured temperatures will be only a few degrees since the thermocouple junction is at themid-plane of the element and close to the anticipated hot spot. If the thermocouple element islocated in a region of lower temperature, such as on the periphery of the core, the measuredtemperature will differ by a greater amount from that actually occurring at the core hot spot.Calculations have shown that if the thermocouple element were located on the periphery of thecore, the true temperature at the hottest location in the core will differ from the measuredtemperature by no more than a factor of two. Thus, with the WFE positioned in the locationspecified by the license, when the temperature in the thermocouple element reaches the setting of175 °C, the true temperature at the hottest location would be no greater than 350 °C, providing amargin to the safety limit of at least 650 °C. This margin is ample to account for the remaininguncertainty in the accuracy of the fuel temperature, measurement channel, and any overshoot inreactor power resulting from a reactor transient during steady state mode operation.7 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163.0 LIMITING CONDITIONS FOR OPERATION3.1 REACTOR CORE PARAMETERSApplicabilityThese specifications shall apply to the reactor at all times it is operating.ObjectiveThe objectives are to ensure that the reactor can be controlled and shut down at all times and thatthe safety limits will not be exceeded.Specifications1. The excess reactivity relative to the cold critical conditions, with or without experimentsin place shall not be greater than $3.50.2. The shutdown margin shall not be less than $0.50 with:a. The reactor in the reference core condition; andb. Total worth of all in-core experiments in their most reactive state; andc. Most reactive control rod fully withdrawn.3. Core configurations:a. The reactor shall only be operated with a standard core.b. No fuel shall be inserted or removed from the core unless the reactor is subcritical bymore than the worth of the most reactive fuel element.c. No control rods shall be removed from the core unless a minimum of four fuelbundles are removed from the core.d. The reactor shall be operated only with three operable control rods.4. No operation with damaged fuel (defined as a clad defect that results in fission productrelease into the reactor coolant) except to locate such fuel.5. The reactivity coefficients for the reactor are:Fuel: -1.2 ¢/°0CModerator: +3.0 ¢/°0CPower: -0.53 ¢/kWThe Fuel Temperature Coefficient, and the Moderator Temperature Coefficient, shall be verified any timethe standard core is modified either by the rotation of a fuel bundle, a change in fuel bundle or reflectorlocation, or the replacement of any fuel or reflector. The Power Temperature Coefficient shall berecalculated if either the Fuel Temperature or Moderator Temperature Coefficient are measured to be morethan 4-5% from the established values. Records of these tests shall be retained for a minimum of five years.6. The burnup ofU-235 in the UZrH fuel matrix shall not exceed 50 % of the initial concentration.8 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016Bases1. While specification 3.1.1, in conjunction with specification 3.1.2, tends to over constrainthe excess reactivity, it helps ensure that the operable core is similar to the core analyzedin the FSAR.2. The value of the shutdown margin as required by specification 3.1.2 assures that thereactor can be shutdown from any operating condition even if the highest worth controlrod should remain in the fully withdrawn position.3. Specification 3.1.3 ensures that the operable core is similar to the core analyzed in theFSAR. It also ensures that accidental criticality will not occur during fuel or control rodmanipulations.4. Specification 3.1.4 limits the fission product release that might accompany operation witha damaged fuel element. Fuel will be considered potentially "Damaged" if said fuel isfound to be leaking under the air and/or water sampling or under such case that the fuelhas been exposed to temperature above 175 °C as measured on the instrumented fuelelement. The criteria of the water and air sampling to determine a leaking fuel element isconsidered positive if either sample is found to contain I- 129 through I- 135, Xe- 135, Kr-85, 87 and Kr-88, Cs-135 and Cs-137, or Sr-89 through Sr-92.5. The reactivity coefficients in Specification 3.1.5 ensure that the net reactivity feedback isnegative.6. General Atomic tests of TRIGA fuel indicate that keeping fuel element burnup below 50% of the original 235U loading will avoid damage to the fuel from fission product buildup.3.2 REACTOR CONTROL AND SAFETY SYSTEMSApplicabilityThese specifications apply to reactor control and safety systems and safety-relatedinstrumentation that must be operable when the reactor is in operation.ObjectiveThe objective of these specifications is to specify the lowest acceptable level of performance orthe minimum number of operable components for the reactor control and safety systems.Specifications1. The drop time of each of the three standard control rods from the fully withdrawnposition to the fully inserted position shall not exceed one second.2. Maximum positive reactivity insertion rate by control rod motion shall not exceed $0.30per second.9 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. The reactor safety channels shall be operable in accordance with Table 3.1l, including theminimum number of channels and the indicated maximum or minimum set points for thescram channels.4. The safety interlocks shall be operable in accordance with Table 3.2, including theminimum number of interlocks.5. The Beam Port and Through Tube interlocks may be bypassed during a reactor operationwith the permission of the Reactor Director.6. A minimum of one reactor power channel, calibrated for reactor thermal power, must beattached to a recording device sufficient for auditing of reactor operation history.Bases1. Specification 3.2.1 assures that the reactor will be shutdown promptly when a scramsignal is initiated. Experiments and analysis have indicated that for the range oftransients anticipated for the MUTR TRIGA reactor, the specified control rod drop timeis adequate to assure the safety of the reactor.2. Specification 3.2.2 establishes a limit on the rate of change of power to ensure that thenormally available reactivity and insertion, rate cannot generate operating conditions thatexceed the Safety Limit. (See FSAR)3. Specification 3.2.3 provides protection against the reactor operating outside of the safetylimits. Table 3.3 describes the basis for each of the reactor safety channels.4. Specification 3.2.4 provides protection against the reactor operating outside of the safetylimits. Table 3.4 describes the basis for each of the reactor safety interlocks.5. Specification 3.2.5 ensures that reactor interlocks will always serve their intendedpurpose. This purpose is to assure that the operator is aware of the status of both thebeam ports and the through tube.6. Specification 3.2.6 provides for a means to monitor reactor operations and verify that thereactor is not operated outside of its license condition.10 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb16RevO2292016.docxLast edit February 29, 2016Table 3.1: Reactor Safety Channels: Scram ChannelsScram ChannelMinimum RequiredOperable2Scram SetpointReactor Power LevelNot to exceed 120 %Fuel Element TemperatureReactor Power ChannelDetector Power Supply12Manual ScramConsole Electrical SupplyRate of power change -PeriodRadiation Area Monitors1111Not to exceed 175 °CLoss of power supply voltage tochamberN/ALoss of electrical power to thecontrol consoleNot less than 5 seconds50 mr/hr (bridge monitor)10 mr/hr (exhaust monitor)Table 3.2: Reactor Safety Channels: InterlocksInterlock/ChannelLog Power LevelStartup Count rateSafety 1 Trip TestPlug ElectricalConnectionRod Drive ControlFunctionProvide signal to period rate and minimum sourcechannels. Prevent control rod withdrawal when neutroncount rate is less than l cps.Prevent control rod withdrawal when neutron count rate isless than 1 cps.Prevent control rod withdrawal when Safety 1 Trip Testswitch is activated.Disable magnet power when Beam Port or Through Tubeplug is removed unless bypass has been activated.Prevent simultaneous manual withdrawal of two or morecontrol rods in the steady state mode of operation.11 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb16RevO2292016.docxLast edit February 29, 2016Table 3.3: Reactor Safety Channels: Scram Channel BasesScram ChannelBasesReactor Power LevelFuel Element TemperatureReactor Power ChannelDetector Power SupplyManual ScramConsole Electric SupplyRate of power change -PeriodRadiation Area MonitorsProvides protection to assure that the reactor can beshut down before the safety limit on the fuelelement temperature will be exceeded.Provides protection to assure that the reactor cannotbe operated unless the neutron detectors that inputto each of the linear power channels are operable.Allows the operator to shut down the reactor if anunsafe or abnormal condition occurs.Assures that the reactor cannot be operated withouta secure electric supply.Assures that the reactor is operated in a manner thatallows the operator time to shut down the reactorbefore the licensed power restriction is exceeded.Assures that the reactor automatically scrams if ahigh airborne radiation level is detected.Table 3.4: Reactor Safety Channels: Interlock BasesInterlock/ChannelLog Power LevelStartup Count rateSafety 1 Trip TestPlug Electrical ConnectionRod Drive ControlBasesThis channel is required to provide a neutrondetector input signal to the startup count ratechannel.Assures sufficient amount of startup neutrons areavailable to achieve a controlled approach tocriticality.Assures that the 1 cps interlock cannot be bypassedby creating an artificial 1 cps signal with the Safety1 trip test switchAssures that the reactor cannot be operated withBeamport or Through Tube plugs removed withoutfurther precautions.Limits the maximum positive reactivity insertionrate available for steady state operation.12 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163.3 PRIMARY COOLANT SYSTEMApplicability_This specification applies to the quality and quantity of the primary coolant in contact with thefuel cladding at the time of reactor startup.Objectives1. To minimize the possibility for corrosion of the cladding on the fuel elements.2. To minimize neutron activation of dissolved materials.3. To ensure sufficient biological shielding during reactor operations.4. To maintain water clarity.Specifications1. A minimum of 15 ft. of coolant shall be above the core.2. Conductivity of the pool water shall be no higher than 5x10-6 mhos/cm and the pH shallbe between 5.0 and 7.5. Conductivity shall be measured before each reactor operation.pH shall be measured monthly, interval not to exceed six weeks.3. Gross gamma measurement shall be less than two times historical data measurements.Gross gamma activity shall be measured monthly, interval not to exceed six weeks.4. The pool water temperature shall not exceed 90 C, as measured by thermocouples locatedin the pool.Bases1. Specification 3.3.1 ensures that both sufficient cooling capability and sufficient biologicalshielding are available for safe reactor operation.2. A small rate of corrosion continuously occurs in a water-metal system. In order to limitthis rate, and thereby extend the longevity and integrity of the fuel cladding, a watercleanup system is required. Experience with water quality control at many reactorfacilities has shown that maintenance within the specified limit provides acceptablecontrol. In addition, by limiting the concentration of dissolved materials in the water, theradioactivity of neutron activation products is limited. This is consistent with theALARA principle, and tends to decrease the inventory of radionuclides in the entirecoolant system, which will decrease personnel exposures during maintenance andoperation.3. Specification 3.3.3 ensures that a fuel failure with release of radioactive materials into thepool will be determined.4. Specification 3.3.4 ensures a DNBR value greater than 2.13 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163.4 CONFINEMENT_ApplicabilityThis specification applies to that part of the facility that contains the reactor, its controls andshielding.ObjectiveThe objective of these specifications is to ensure that sufficient confinement volume is availablefor the dilution of radioactive releases and to limit the rate of release of radioactive material to theoutside environment._Specifications1. Confinement shall be considered established when the doors leading from the reactor bayarea leading into the balcony area on the top floor, and the reception area as well as thebuilding exterior are secured.2. Confinement shall be established whenever the reactor is in an unsecured mode with theexception of the time that persons are physically entering or leaving the confinementarea.Bases1. This specification provides the necessary requirements for confinement, which ensuresreleases to the outside environment are within 10 CFR Part 20 requirements.2. This specification provides the reactor status condition for confimement, as well as allowspersonnel to enter and leave the reactor building, as required, when the reactor isunsecured.3.5 VENTILATJON SYSTEMSApplicabilityThese specifications apply to the ventilation systems for the reactor building._ObjectiveThe objective of these specifications is to ensure that air exchanges between the reactorconfinement building and the environment do not impact negatively on the general public._Specifications1. Air within the reactor building shall not be exchanged with other occupied spaces in thebuilding.2. All locations where ventilation systems exchange air with the environment shall havefailsafe closure mechanisms.14 O:\M UTR\2O16M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. Forced air ventilation to the outside shall automatically secure without operatorintervention in such case that the radiation levels exceed a preset level as defined infacility procedures. The setpoints are: 50 mR/hr (bridge monitor), 10 mR/hr (exhaustmonitor).Bases1. This specification ensures that radioactive releases inside the reactor building will not betransported to the remainder of the building.2. This specification ensures that the reactor building can always be isolated from theenvironment.3. This specification ensures that radioactive release will be minimized by stopping forcedflow to the outside environment.3.6 _ RADIATION MONITORING SYSTEM ANT) EFFLUENTS3.6.1 Radiation Monitoring SystemApplicabilityThis specification applies to the radiation monitoring information that must be available to thereactor operator during reactor operation.ObjectiveThe objective is to assure that sufficient radiation monitoring information is available to theoperator to assure safe operation of the reactor.Specifications1. The reactor shall not be operated unless a minimum of one of the two radiation areamonitor channels listed in Table 3.5 are operable.2. For a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for maintenance or calibration to the radiationmonitor channels, the intent of specification 3.6.1 will be satisfied if they are replacedwith portable gamma sensitive instruments having their own alarms or which shall beobservable by the reactor operator.3. The alarm set points shall be stated in a facility operating procedure. The alarm setpointsfor the bridge monitor are: .37 mR/hr (alert), 50 mR/hr (scram). The setpoints for theexhaust monitor are: 8 mR/hr (alert), 10 mR/hr (scram).4. The campus radiation safety organization shall maintain an environmental monitor at theMUJTR site boundary.15 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016Table 3.5: Minimum Radiation Monitoring ChannelsRadiation Area MonitorsExhaust Radiation MonitorBridge Radiation MonitorFunctionMinimum Number OperableMonitor radiation levels in ReactorBay area at an Exhaust Fan locationMonitor radiation levels in ReactorBay area at the Reactor BridgelocationA minimum of 1 of the 2monitors shall be operableBases1. Specification 3.6.1.1 ensures that a significant fuel failure with release of radioactivematerials will be determined and that any large releases will be mitigated by the specifiedprotective actions.2. Specification 3.6.1.2 allows for continued reactor operation if maintenance and/orcalibration of the radiation area monitors is required.3. The alarm and scram set points shall be designed to ensure that dose rates delivered toareas accessible to members of the general public do not exceed the levels defined in 10CFR Part 20. Additionally, the radiation area monitors provide information to operatingpersonnel of any impending or existing danger from radiation so that there will besufficient time to evacuate the facility and take the necessary steps to prevent the spreadof radioactivity to the surroundings.4. The intent of Specifications 3.6.1.3 and 3.6.1.4 is to ensure that facility does not lead to adose to the general public greater than that allowed by 10 CFR Part 20.3.6.2 EffluentsApplicability.This specification applies to limits on effluent release.ObjectiveThe objective is to ensure that the release of radioactive materials from the reactor facility tounrestricted areas do not exceed federal regulations.SpecificationAll effluents from the MUTR shall conform to the standards set forth in 10 CFR Part 20.BasisThe intent of 3.6.2 is to ensure that, in the event that radioactive effluents are released, the dose tothe general public will be less than that allowed by 10 CFR Part 20.16 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163.7 LIMITATIONS ON EXPERIMENTSApplicabilityThe specification applies to experiments installed in the reactor and its experimental facilities.ObjectiveThe objective is to prevent damage to the reactor or excessive release of radioactive material inthe event of an experiment failure.SpecificationsThe reactor shall not be operated unless the following conditions governing experiments exist.1. The reactivity worth of any single experiment shall be less than $1.00.2. The total absolute reactivity worth of in-core experiments shall not exceed $3.00,including, the potential reactivity which might result from experimental malfunction andexperiment flooding or voiding.3. Experiments containing materials corrosive to reactor components, compounds highlyreactive with water, potentially explosive materials, and liquid fissionable materials shallbe doubly encapsulated.4. Explosive materials in quantities greater than 25 mg TNT or its equivalent shall not beirradiated in the reactor or experimental facilities. Explosive materials in quantities equalto or less than 25 mg may be irradiated provided the pressture produced upon detonationof the explosive has been calculated and/or experimentally demonstrated to be less thanthe failure pressure of the container. The failure pressure of the container is one half ofthe design pressure. Total explosive material inventory in the reactor facility may notexceed 100 mg TNT or its equivalent.5. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, orproduce aerosols under (1) normal operating conditions of the experiment or reactor, (2)credible accident conditions in the reactor or (3) possible accident conditions in theexperiment shall be limited in type and quantity such that if 100 % of the gaseous activityor radioactive aerosols produced escaped to the reactor room or the atmosphere, theairborne radioactivity in the reactor room or outside environment will not result inexceeding the applicable dose limits set forth in 10 CFR Part 20.In calculations pursuant to 3.7.5 above, the following assumptions shall be used:a. If the effluent from an experimental facility exhausts through a holdup tank, whichcloses automatically on high radiation level, at least 10 % of the gaseous activity oraerosols produced will escape.b. If the effluent from an experimental facility exhausts through a filter installationdesigned for greater than 99 % efficiency for 0.3 particles, at least 10 % of theseparticles can escape.17 O:\M UTR\2016M UTRLivingoocs\WorkingCopyTechnical Specifications 29 feb 16Rev02292016.docxLast edit February 29, 2016c. If an experiment fails and releases radioactive gases or aerosols to the reactor bay oratmosphere, 100 per cent of the radioactive gases or aerosols escape.d. If an experiment fails that contains materials with a boiling point above 1300 F (540C), the vapors of at least 10 percent of the materials escape through an undisturbedcolumn of water above the core.6. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes131 through 135 in the experiment is no greater than 5 mCi.Bases1. This specification is intended to provide assurance that the worth of a single unsecuredexperiment will be limited to a value such that the safety limit will not be exceeded if thepositive worth of the experiment were to be inserted suddenly.2. The maximum worth of a single experiment is limited so that its removal from the coldcritical reactor will not result in the reactor achieving a power level high enough toexceed the core temperature safety limit. Since experiments of such worth must befastened in place, its inadvertent removal from the reactor operating at full power wouldresult in a relatively slow power increase such that the reactor protective systems wouldact to prevent high power levels from being attained.The maximum worth of all experiments is also limited to a reactivity value such that thecold reactor will not achieve a power level high enough to exceed the core temperaturesafety limit if the experiments were removed inadvertently.3. This specification is intended to prevent damage to reactor components resulting fromexperiment failure. If an experiment fails, inspection of reactor structures andcomponents shall be performed in order to verifyr that the failure did not cause damage.If damage is found, appropriate corrective actions shall be taken.4. This specification is intended to prevent damage to reactor components resulting fromfailure of an experiment involving explosive materials, especially the accidentaldetonation of the explosive. If an experiment fails, inspection of reactor structures andcomponents shall be performed in order to verify' that the failure did not cause damage.If damage is found, appropriate corrective actions shall be taken.5. This specification is intended to reduce the likelihood that airborne activities in excess ofthe limits of Table 2 of Appendix B of 10 CFR Part 20 will be released to the atmosphereoutside the facility boundary.6. The 5 mCi limitation on iodine 131 through 135 assures that in the event of failure of afueled experiment leading to total release of the iodine, the exposure dose at theexclusion area boundary will be less than that allowed by 10 CFR Part 20 for anunrestricted area. (See SAR)18 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20164.0 SURVEILLANCE REQUIREMENTSINTRODUCTIONSurveillances shall be performed on a timely basis as defined in the individual procedures governing theperformance of the surveillance. In the event that the reactor is not in an operable condition, such asduring periods of refueling, or replacement or repair of safety equipment, surveillances may be postponeduntil such time that the reactor is operable. In such case that any surveillance must be postponed, awritten directive signed by the Facility Director, shall be placed in the records indicating the reason whyand the expected completion date of the required surveillance. This directive shall be written before thedate that the surveillance is due. Under no circumstance shall the reactor perform routine operations untilsuch time that all surveillances are current and up to date. Any system or component that is modified,replaced, or had maintenance performed will undergo testing to ensure that the system/componentcontinues to meet performance requirements.4.1 REACTOR CORE PARAMETERSApplicabilityThese specifications apply to the surveillance requirements for the reactor core.O~bjectiveThe objective of these specifications is to ensure that the specifications of Section 3.1 aresatisfied.Specifications1. The excess reactivity shall be determined annually, at intervals not to exceed 15 months,and after each time the core fuel configuration is changed, these changes include anyremoval or replacement of control rods.2. The shutdown margin shall be determined annually, at intervals not to exceed 15 months,and after each time the core fuel configuration is changed, these changes include anyremoval or replacement of control rods3. Core configuration shall be verified prior to the first startup of the day.4. Gross gamma measurements shall be taken monthly, at intervals not to exceed six weeks.5. Twenty percent of the fuel elements shall be visually inspected annually, not toexceed 15 months, such that the entire core is inspected over a five year period.6. Burnup shall be verified in the Annual Report.BasesExperience has shown that the identified frequencies ensure performance and operability for eachof these systems or components. For excess reactivity and shutdown margin, long-term changesare slow to develop. For fuel inspection, visually inspecting 20% of the bundles annually willidentify any developing fuel integrity issues throughout the core.19 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20164.2 REACTOR CONTROL AND SAFETY SYSTEMSApp~licabilityThese specifications apply to the surveillance requirements for reactor control and safety systems.ObjectiveThe objective of these specifications is to ensure that the specifications of Section 3.2 aresatisfied.Specifications1. The reactivity worth of each standard control rod shall be determined annually, intervalsnot to exceed 15 months, and after each time the core fuel configuration is changed or acontrol rod is changed.2. The control rod withdrawal and insertion speeds shall be determined annually, intervalsnot to exceed 15 months, or whenever maintenance or repairs are made that could affectrod travel times.3. Control rod drop times shall be measured annually; intervals not to exceed 15 months, orwhenever maintenance or repairs are made that could affect their drop time.4. All scram channels and power measuring channels shall have a channel test, includingtrip actions with safety rod release and specified interlocks performed after each securedshutdown, before the first operation of the day, or prior to any operation scheduled to lastmore than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or quarterly, with intervals not to exceed 4 months. Scram channelsshall be calibrated annually, intervals not to exceed 15 months.5. Operability tests shall be performed on all affected safety and control systems after anymaintenance is performed.6. A channel calibration shall be made of the linear power level monitoring channelsannually, intervals not to exceed 15 months.7. A visual inspection of the control rod poison sections shall be made biennially, intervalsnot to exceed 28 months.8. A visual inspection of the control rod drive and scram mechanisms shall be madeannually, intervals not to exceed 15 months.Bases1. The reactivity worth of the control rods, specification 4.2.1, is measured to assure that therequired shutdown margin is available and to provide a means to measure the reactivityworth of experiments. Long term effects of TRIGA reactor operation are such thatmeasurements of the reactivity worths on an annual basis are adequate to insure that nosignificant changes in shutdown margin have occurred.20 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20162. The control rod withdrawal and insertion rates, specification 4.2.2, are measured to insurethat the limits on maximum reactivity insertion rates are not exceeded.3. Measurement of the control rod drop time, specification 4.2.3, ensures that the rods canperform their safety function properly.4. The surveillance requirement specified in specification 4.2.4 for the reactor safety scramchannels ensures that the overall functional capability is maintained.5. The surveillance test performed after maintenance or repairs to the reactor safety systemas required by specification 4.2.5 ensures that the affected channel will perform asintended.6. The linear power level channel calibration specified in specification 4.2.6 will assure thatthe reactor will be operated at the licensed power levels.7. Specification 4.2.7 assures that a visual inspection of control rod poison sections is madeto evaluate corrosion and wear characteristics and any damage caused by operation in thereactor.8. Specification 4.2.8 assures that a visual inspection of control drive mechanisms is madeto evaluate corrosion and wear characteristics and any damage caused by operation in thereactor.4.3 PRIMARY COOLANT SYSTEMApplicabilityThese specifications apply to the surveillance requirements of the reactor primary coolant system.ObjectiveThe objective of these specifications is to ensure the operability of the reactor primary coolantsystem as described in Section 3.3.Specifications1. The primary coolant level shall be verified before each reactor startup or daily duringoperations exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.2. Pool water conductivity shall be determined prior to the first startup of the day, and poolwater pH shall be determined monthly at intervals not to exceed six weeks.3. Pool water gross gamma activity shall be determined monthly, at intervals not to exceedsix weeks. If gross gamma activity is high (greater than twice historical data), gammaspectroscopy shall be performed.4. Pool water temperature shall be measured prior to the reactor startup and shall bemonitored during reactor operation.21 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016Bases1. Specification 4.3.1 ensures that sufficient water exists above the core to provide bothsufficient cooling capacity and an adequate biological shield.2. Specification 4.3.2 ensures that poor pool water quality could not exist for long withoutbeing detected. Years of experience at the MIUTR have shown that pool water analysison a monthly basis is adequate to detect degraded conditions of the pooi water in a timelymanner.3. Gross gamma activity measurements are conducted to detect fission product releasesfrom damaged fuel element cladding.4. Specification 4.3.4 ensures that the maximum allowable pool water temperature is notexceeded.4.4 CONFINEMENTApplicabilityThis specification applies to that part of the facility which contains the reactor, its controls andshielding.ObjectiveThe objective of this specification is to ensure that radioactive releases from the confinementcan be limited._SpecificationPrior to putting the reactor in an unsecured mode, the isolation of the confinement building shallbe visually verified.BasesThis specification ensures that the minimal leakage rate assumed in the SAR is actually presentduring reactor operations in order to limit the release of radioactive material to the environs.4.5 VENTTLATION SYSTEMApplicabilityThis specification applies to the reactor ventilation system._ObjectiveThe objective is to assure that provisions are made to restrict the amount of radioactivity releasedto the environment.22 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016SpecificationThe ability to secure the ventilation system shall be verified before the first reactor operation of theday.BasesThe facility is designed such that in the event that excessive airborne radioactivity is detected theventilation system shall be shutdown to minimize transport of airborne materials. Analysisindicates that in the event of a major fuel element failure personnel would have sufficient time toevacuate the facility before the maximum permissible dose (10 CFR Part 20) is exceeded.4.6 RADIATION MONITORING SYSTEM ANT) EFFLUENTS4.6.1 Radiation Monitoring SystemApplicabilityThis specification applies to the surveillance requirements for the Radiation Area MonitoringSystem (RAMS).ObjectiveThe objective of these specifications is to ensure the operability of each radiation area monitoringchannel as required by Section 3.6 and to ensure that releases to the environment are kept belowallowable limits.Specifications1. A channel calibration shall be made for each channel listed in Table 3.5 annually but atintervals not to exceed 15 months or whenever maintenance or repairs are made thatcould affect their calibration.2. A channel test shall be made for each channel listed in Table 3.5 prior to starting up thereactor to ensure reactor scram, fan shutdown, and louver closing.BasesSpecifications 4.6.1.1 and 4.6.1.2 ensure that the various radiation area monitors are checked andcalibrated on a routine basis, in order to assure compliance with 10 CFR Part 20.4.6.2 EffluentsApplicabilityThis specification applies to the surveillance requirements for air and water effluents.23 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016ObjectiveThe objective of these specifications is to that releases to the environment are kept belowallowable limits.Specifications1. Reactor building air samples shall be counted for gross gamma activity monthly, intervalsnot to exceed 6 weeks.2. A sample of any water discharged from the reactor building sump shall be counted forgross gamma activity before its release to the environs.BasesSpecifications 4.6.2.1 and 4.6.2.2 ensure that the facility effluents comply with 10 CFR Part 20.4.7 EXPERIMENTSApplicabilityThis specification applies to the surveillance requirements for experiments installed in the reactorand its irradiation facilities.ObjectiveThe objective of this specification is to prevent the conduct of experiments which may damagethe reactor or release excessive amounts of radioactive materials as a result of experiment failure.Specifications1. The reactivity worth of an experiment shall be estimated or measured, as appropriate, beforereactor operation with said experiment2. An experiment shall not be installed in the reactor or its irradiation facilities unless a safetyanalysis has been performed and reviewed for compliance with Section 3.7 by the ReactorSafety Committee (new experiment) or Facility Director (modified routine experiment), infull accord with Sections 6.1.2 and 6.2.1 of these Technical Specifications and the procedureswhich are established for this purpose.BasisExperience has shown that experiments reviewed and approved by the Reactor Safety Committee orFacility Director can be conducted without endangering the safety of the reactor, personnel, orexceeding Technical Specification limits.24 O:\M UTR\2Oi6M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20165.0 DESIGN FEATURES5.1 SITE CHARACTERISTICSApplicabilityThis specification applies to the reactor facility and its site boundary.ObjectiveThe objective is to assure that appropriate physical security is maintained for the reactor facilityand the radioactive materials contained within it.Specifications1. The reactor shall be housed in a closed room designed to restrict leakage. The closedroom does not include the West balcony area.2. The reactor site boundary shall consist of the outer walls of the reactor building and thearea enclosed by the loading dock fence.3. The restricted area shall consist of all areas interior to the reactor building including thewest balcony and lower entryway.4. The controlled area shall consist of all areas interior to the reactor building includingthe west balcony and lower entryway.BasesThese specifications assure that appropriate control is maintained over access to the facility bymembers of the general public.5.2 REACTOR PRIMARY COOLANT SYSTEMApplicabilityThis specification applies to the pool containing the reactor and to the cooling of the core by thepool water.ObjectiveThe objective is to assure that coolant water shall be available to provide adequate cooling of thereactor core and adequate radiation shielding.Specifications1. The reactor core shall be cooled by natural convective water flow.2. The pool water inlet pipe is equipped with a siphon break at the surface of the pool.25 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. The pooi water return (outlet) pipe shall not extend more than 50.8 cm (20 in) below theoverflow outlet pipe when fuel is in the core.Bases1. Specification 5.2.1 is based on thermal and hydraulic calculations and operation of the MUTRthat show that the core can operate in a safe manner at power levels up to 300 kW withnatural convection flow of the coolant.2. Specifications 5.2.2 and 5.2.3 ensures that the pool water level can normally decrease only by50.8 cm (20 in) if the coolant piping were to rupture and siphon water from the reactor tank.Thus, the core will be covered by at least 4.57 m (15 ft.) of water.5.3 REACTOR CORE AND FUELApplicabilityThis specification applies to the configuration of the core and in-core experiments.ObjectiveThe objective is to ensure that the core configuration is as specified in the license.Specifications1. The core shall consist of 93 TRIGA fuel elements assembled into 24 fuel bundles -21bundles shall contain four fuel elements and 3 bundles shall contain three fuel elements and acontrol rod guide tube.2. The fuel bundles shall be arranged in a rectangular 4 x 6 configuration, with one bundledisplaced for the in-core pneumatic experimental system.3. The reactor shall not be operated at power levels exceeding 250 kW.4. The reflector shall be a combination of two graphite reflector elements and waterBasis1. Only TRIGA fuel elements shall be used in the fuel bundles.2. The experimental system allows insertion of small samples directly into the reactor core.3. The maximum power level presents a conservative limitation with respect to the safety limitsfor the maximum temperature in the fuel.4. The reflector reduces the neutron leakage from the reactor core.26 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20165.3.1 Reactor FuelApplicabilityThis specification applies to the fuel elements used in the reactor core.ObjectiveThe objective is to assure that the fuel elements are of such design and fabricated in such a manner asto permit their use with a high degree of reliability with respect to their physical and nuclearcharacteristics, and that the fuel used in the reactor has characteristics consistent with the fuelassumed in the SAR and the license.SpecificationsThe individual unirradiated standard TRIGA fuel elements shall have the followingcharacteristics:1. Uranium content: a maximum of 9.0 weight % uranium enriched to less than 20 % 235U2. Zirconium hydride atom ratio: nominal 1.5 -1.8 hydrogen-to-zirconium, ZrHx3. Cladding: 304 stainless steel, nominal thickness of 0.508 mm (.020 in)4. The overall length of a fuel element shall be 30 inches, and the fueled length shall be 15inches.BasisThe design basis of the standard TRIGA fuel element demonstrates that 250 kW steady stateoperation presents a conservative limitation with respect to safety limits for the maximumtemperature generated in the fuel.5.3.2 Control RodsApplicabilityThis specification applies to the control rods used in the reactor core.ObjectiveThe objective is to assure that the control rods are of such a design as to permit their use with a highdegree of reliability with respect to their physical and nuclear characteristics.Specifications1. The three control rods shall have scram capability, shall be used for reactivity control, andshall contain borated graphite, B4C, in powder form.2. The control rod cladding shall be aluminum with nominal thickness 0.71 mm (0.028") andlength 17".27 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTech nicai Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016BasisThe poison requirements for the control rods are satisfied by using neutron absorbing boratedgraphite, B4C, powder. These materials must be contained in a suitable clad material such asaluminum to ensure mechanical stability during movement and to isolate the poison from the tankwater environments. Scram capabilities are provided for rapid insertion of the control rods, which isthe primary safety feature of the reactor.5.4 FISSIONABLE MATERIAL STORAGEApplicabilityThis specification applies to the storage of reactor fuel at times when it is not in the reactor core.ObjectiveThe objective is to assure that fuel that is being stored will not become critical and will not reachan unsafe temperature.Specifications1. All fuel elements shall be stored either in a geometrical array where the k-effective is less than 0.8for all conditions of moderation and reflection or stored in an approved fuel shipping container.2. Irradiated fuel elements and fueled devices shall be stored in an array which will permitsufficient natural convection cooling by water or air such that the fuel element or fueled devicetemperature will not exceed design values.3. When fuel is in storage in any area other than the grid plate, that area must be equipped withmonitoring devices that both measure and record the radiation levels and temperature of theregion surrounding the fuel.BasisThe limits imposed by Specifications 5.4.1 and 5.4.2 are conservative and assure safe storage.28 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20166.0 ADMINISTRATION6.1 ORGANIZATIONThe Maryland University Training Reactor (MUTR) is owned and operated by the University ofMaryland, College Park. Its position in the university's structure is shown in Figure 6.1The university shall provide whatever resources are required to maintain the facility in acondition that poses no hazard to the general public or to the environment.6.1.1 StructureFigure 6.2 shows the MUTR organizational structure.6.1.2 ResponsibilityThe Dean College of Engineering is responsible for the oversight and operation of the school ofengineering.The Chair of the Department of Materials Science and Engineering is responsible for the oversightand operation of the Department of Materials Science and Engineering.The Director of MUTR: Responsibility for the safe operation of the reactor facility andradiological safety shall rest with the Facility Director. The members of the organization chartshown in Figure 6.2 shall be responsible for safeguarding the public and facility personnel fromundue radiation exposure and for adhering to all requirements of the operating license.The Senior Reactor Operators (SRO) are individuals who are licensed by the NRC to direct theactivities of reactor operators.The Reactor Operators are individuals who are licensed by the U.S. Nuclear RegulatoryCommission (NRC) to manipulate the controls of the reactor.6.1.3 Facility Staff Requirements1. The minimum staffing while the reactor is not secured shall be:a. A licensed reactor operator (RO) or a licensed senior reactor operator (SRO) shall bepresent in the control room.b. A minimum of two persons shall be present in the facility or in the Chemical andNuclear Engineering Building while the reactor is not secured: the operator in thecontrol room and a second person who can be reached from the control room who isable to carry out prescribed written instructions which may involve activatingelements of the Emergency Plan, including evacuation and initial notificationprocedures.c. A licensed SRO shall be present or readily available on call. "Readily Available onCall" means an individual who (1) has been specifically designated and thedesignation known to the operator on duty, (2) keeps the operator on duty informedof where he/she may be rapidly contacted and the method of contact, and (3) is29 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016capable of arriving at the reactor facility within a reasonable amount of time undernormal conditions. At no time while the reactor is not secured shall the designatedSRO be more than thirty minutes or ten miles from the facility.2. A list of reactor facility personnel by name and telephone number shall be readily available inthe control room for use by the operator. The list shall include:a. Management personnelb. Radiation safety personnelc. Licensed operators30 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016PresidentUniversity of MarylandProvost & SeniorVice PresidentDeanClark School ofEngineeringVice PresidentAdministrative AffairsDirectorDepartment ofEnvironmental SafetyRadiationSafety .Committee %%ChairDepartment ofMaterials Science andEngineeringRadiation SafetyOfficer4LDirectorNuclear ReactorFacilityReviews* * --AuditReactor .iembr-Safety ."MmeCommitteeRadiation Safety OfficeStaffServicesiReactor OperationsStaff-- Normal Administrative Reporting Channel-...Communication LinesFigure 6.1: MUTR Position in University of Maryland Structure31 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016DeanClark School of Engineering(Level 1)ChairDepartment of MaterialsScience and Engineering(Level 1)DirectorNuclear Reactor Facility(Level 2),I!Senior Reactor Operator(Level 3)Reactor Operator(Level 4)Reactor SafetyCommittee4----------- Normal Administrative Reporting Channel-------------------Communication LinesFigure 6.2: MUTR Organizational Structure32 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTech nicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. The following operations shall be supervised by a senior reactor operator:a. Initial startup and approach to power following new fuel loading or fuelrearrangementb. When experiments are being manipulated in the core that have an estimated worthgreater than $0.80c. Removal of control rods or fuel manipulations in the cored. Resumption of operation following an unplanned or unscheduled shutdown or anyunplanned or unexpected significant reduction in power.6.1.4 Selection and Training of PersonnelThe selection and training of operations personnel should be in accordance with the following:1. Responsibility -The Facility Director or his designated alternate is responsible for thetraining and requalification of the facility reactor operators and senior reactor operators.This selection shall be in conjunction with the guidelines set forth in ANSJIANS 15.1 and15.4.6.2 REVIEW AND AUDIT6.2.1 Reactor Safety CommitteeA Reactor Safety Committee (RSC) shall exist for the purpose of reviewing matters relating to thehealth and safety of the public and facility staff and the safe operation of the facility. It isappointed by and reports to the Chairperson of the Department of Materials Science andEngineering. The RSC shall consist of a minimum of five persons with expertise in the physicalsciences and preferably some nuclear experience. Permanent members of the committee are theFacility Director and the Campus Radiation Safety Officer or that office's designated alternate,neither may serve as the committee's chairperson. Qualified alternates may serve on thecommittee. Alternates may be appointed by the Chairperson of the RSC to serve on a temporarybasis. At least one committee member must be from outside the Department of Materials Scienceand Engineering.6.2.1.1l Reactor Safety Committee Charter And Rules1. The RSC shall meet at least twice per year, and more often as required.2. A quorum of the RSC shall be not less than half of the committee members, one of whomshall be the Campus Radiation Safety Officer (or designated alternate). No more than twoalternates shall be used to make a quorum. MUTR staff members shall not constitute themajority of a voting quorum.3. Minutes of all meetings will be retained in a file and distributed to all RSC members.33 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20166.2.1.2 Reactor Safety Committee Review FunctionThe RSC shall review the following:1. Determinations that proposed changes in equipment, systems, test, experiments, orprocedures are allowed without prior authorization by the responsible authority, e.g. 10 CFR50.59;2. All new procedures and major revisions thereto having safety significance, proposed changesin reactor facility equipment, or systems having safety significance;3. All new experiments or classes of experiments that could affect reactivity or result in therelease of radioactivity;4. Proposed changes in technical specifications, or license;5. Violations of technical specifications or license. Violations of internal procedures orinstructions having safety significance;6. Operating abnormalities having safety significance;7. Reportable occurrences listed in Section 6.7.2;8. Audit reports.A written report of the findings and recommendations of the RSC shall be submitted to Level 1management, the Facility Director, and the RSC members in a timely manner after the review hasbeen completed.6.2.1.3 Reactor Safety Committee Audit Function1. An annual audit and review of the reactor operations will be performed by an outsideindividual or group familiar with research reactor operations. They shall submit a report tothe Facility Director and the Reactor Safety Committee.2. The following shall be reviewed:a. Reactor operators and operational records for compliance with internal rules,procedures, and regulations, and with license provisions;b. Existing operating procedures for adequacy and accuracy;c. Plant equipment performance and its surveillance requirements;d. Records of releases of radioactive effluents to the environment;e. Operator training and requalification;f. Results of actions taken to correct those deficiencies that may occur in the reactorfacility equipment, systems, structures, or methods of operation that affect reactorsafety; and34 O :\M UTR\2016M UTRLivingDocs\WorlingCopyTech nicalI Specifications 29 feb 16RevO22920l6.docxLast edit February 29, 2016g. Reactor facility emergency plan and implementing procedures.Deficiencies uncovered that affect reactor safety shall immediately be reported to Level 1management and the Facility Director. A written report of the findings of the audit shall besubmitted to Level 1 management, the Facility Director, and the RSC members within 3 monthsafter the audit has been completed.6.2.2 Audit of ALARA ProgramThe Facility Director or his designated alternate shall conduct an audit of the reactor facilityALARA Program at least once per calendar year (not to exceed fifteen months). The results ofthe audit shall be presented to the RSC at the next scheduled meeting. This audit may occur aspart of a review of the overall campus ALARA program.6.3 RADIATION SAFETYA radiation safety program following the requirements established in 10 CFR Part 20 will beundertaken by the Radiation Safety Office. The facility director will ensure that ALARAprinciples are followed during all facility activities.6.4 OPERATING PROCEDURESWritten procedures, reviewed and approved by the Reactor Safety Committee, shall be in effectand followed for the following items prior to performance of the activity. The procedures shallbe adequate to assure the safety of the reactor, but should not preclude the use of independentjudgment and action should the situation require such.1. Start-up, operation, and shutdown of the reactor2. Installation or removal of fuel elements, control rods, experiments, and experimentalfacilities3. Maintenance procedures that could have an effect on reactor safety4. Periodic surveillance checks, calibrations, and inspections required by the TechnicalSpecifications or those that may have an effect on reactor safety5. Administrative controls for operations and maintenance and for the conduct ofirradiations and experiments that could affect reactor safety or core reactivity6. For any activity pertaining to shipping, possession, and transfer of radioactivematerial, these procedures shall be written in conjunction with the Radiation SafetyOffice and the Radiation Safety Officer who shall inform the Reactor Director of anychanges in regulations or laws that may require modification of these procedures. Allshipping and receiving of radioactive material shall be performed in conjunctionw i t h , and with the approval of the Radiation Safety Office.7. Implementation, maintenance, and modification to the Emergency Plan8. Implementation, maintenance, and modification to the Security Plan35 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20169. Implementation, maintenance, and modification to the Radiation Protection Plan.The Radiation Protection Plan shall include an ALARA plan as defined inANSI/ANS- 15.l1110. Use, receipt, and transfer of byproduct materialSubstantive changes to the above procedures shall be made with the approval of the FacilityDirector and the Reactor Safety Committee and shall be made in accordance with 10 CFR 50.59.This approval shall be granted before the changes may be considered in effect. The onlyexception to this clause is in such a case where the delay in implementation would cause acredible risk to the public or the facility. If such a case exists as determined by the FacilityDirector, temporary approval may be granted by the Director but must be approved by theReactor Safety Committee within thirty days. Temporary or minor changes to procedures shallbe documented and subsequently reviewed by the Reactor Safety Committee at the nextscheduled meeting. The Reactor Director shall have the power to approve minor changes such asphone number changes, typographical error correction or any other change that does not changethe effectiveness or the intent of the procedure. It shall be considered sufficient approval anddocumentation when the Director forwards by electronic means to both the Radiation SafetyOfficer and the Chair of the Reactor Safety Committee. A copy of the transmission shall be filedwith the appropriate procedure.6.5 EXPERIMENT REVTEW AND APPROVAL1. Routine experiments may be performed at the discretion of the duty senior reactor operatorwithout the necessity of further review or approval.2. Modified routine experiments shall be reviewed and approved in writing by the FacilityDirector, or designated alternate.3. Special experiments shall be reviewed by the RSC and approved by the RSC and the FacilityDirector or designated alternate prior to initiation.4. The review of an experiment listed in subsections 6.5.2 and 6.5.3 above, shall consider itseffect on reactor operation and the possibility and consequences of its failure, including,where significant, chemical reactions, physical integrity, design life, proper cooling,interaction with core components, and any reactivity effects.6.6 REQUIRED ACTIONS6.6.1 Actions To Be Taken In Case Of Safety Limit ViolationIn the event a safety limit is exceeded:1. The reactor shall be shut down and reactor operation shall not be resumed untilauthorized by the NRC.2. The event shall be reported to the Reactor Director who will report to the NRC asrequired in section 6.7.2.36 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. An immediate report of the occurrence shall be made to the Chairman, Reactor SafetyCommittee, and reports shall be made to the NRC in accordance with Section 6.7.2 ofthese specifications, and4. A report, and any follow-up report, shall be prepared. The report shall describe thefollowing:a. Applicable circumstances leading to the violation, including when known, thecause, and contributing factors;b. Effect of the violation upon reactor facility components, systems, or structuresand on the health and safety of personnel and the public; andc. Corrective action to be taken to prevent recurrence.The report shall be reviewed by the Reactor Safety Committee and submitted to the NRCwhen authorization is sought to resume operation of the reactor.6.6.2 Actions to Be Taken In The Event Of a Reportable OccurrenceIn the event of a reportable occurrence, as defined in section 1.32 of these TechnicalSpecifications, the following actions will be taken:1. Immediate action shall be taken to correct the situation and to mitigate the consequencesof the occurrence.2. The reactor shall be shut down and reactor operation shall not be resumed untilauthorized by the Facility Director.3. The event shall be reported to the Facility Director who will report to the NRC asrequired in section 6.7.2.4. The Reactor Safety Committee shall investigate the causes of the occurrence at its nextmeeting. The Reactor Safety Committee shall report its findings to the NRC and Dean,School of Engineering. The report shall include an analysis of the causes of theoccurrence, the effectiveness of corrective actions taken, and recommendations ofmeasures to prevent or reduce the probability or consequences of recurrence.6.7 REPORTS6.7.1 Annual Operating ReportA report summarizing facility operations shall be prepared annually for the reporting periodending June 30. This report shall be submitted by December 30 of each year to the NRCDocument Control Desk. The report shall include the following:1. A brief narrative summary of results of reactor operations and surveillance tests andinspections required in section 4.0 of these Technical Specifications2. A tabulation showing the energy generated in MW hr-' for the year37 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. A list of unscheduled shutdowns including the reasons therefore and corrective actiontaken, if any4. A tabulation of the major maintenance operations performed during the period, includingthe effects, if any, on safe operation of the reactor, and the reason for any correctivemaintenance required5. A brief description ofa. Each change to the facility to the extent that it changes a description of thefacility in the Final Safety Analysis Reportb. Review of changes, tests, and experiments made pursuant to 10 CFR Part 50.59.6. A summary of the nature and amount of radioactive effluents released or discharged to theenvironment7. A description of any environmental surveys performed outside of the facility8. A summary of exposure received by facility personnel and visitors where such exposuresare greater than 25 percent of limits allowed by 10 CFR Part 209. Changes in facility organization6.7.2 Special ReportsNotification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone to the NRC Operations Center, followedby a written report faxed within 14 days in the event of the following:1. A reportable occurrence, as defined in Section 1.32 of this document2. Release of radioactivity from the site above allowed limits3. Exceeding the Safety LimitThe written report shall be sent to the NRC document control desk. The written report and, to theextent possible, the preliminary telephone or facsimile notification shall:1. Describe, analyze, and evaluate safety implications2. Outline the measures taken to ensure that the cause of the condition is determine3. Indicate the corrective action taken to prevent repetition of the occurrence includingchances to procedures4. Evaluate the safety implications of the incident in light of the cumulative experienceobtained from the report of previous failure and malfunction of similar systems andcomponents38 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20166.7.3 Unusual Event ReportA written report shall be forwarded within 3 0 days to the NRC Document Control Desk, with acopy to the Regional Administrator, Region I, NRC, in the event of:1. Discovery of any substantial errors in the transient or accident analysis or in the methodsused for such analysis as described in the Safety Analysis Report or in the bases for theTechnical Specifications2. Discovery of any substantial variance from performance specifications contained in theTechnical Specifications or Safety Analysis Report3. Discovery of any condition involving a possible single failure which, for a systemdesigned against assumed failure, could result in a loss of the capability of the system toperform its safety function4. A permanent change in the position of Department Chair or Facility Director6.8 RECORDS1. The following records shall be retained for a period of at least five years:a. Normal reactor facility operation and maintenanceb. Reportable occurrencesc. Surveillance activities required by Technical Specificationsd. Facility radiation and contamination surveyse. Experiments performed with the reactorf. Reactor fuel inventories, receipts, and shipmentsg. Approved changes in procedures required by these Technical Specificationsh. Minutes of the Reactor Safety Committee meetingsi. Results of External Audits2. Retraining and requalification records of current licensed operators shall be maintained at alltimes that an operator is employed or until the operator's license is renewed.3. The following records shall be retained for the lifetime of the facility:a. Liquid radioactive effluents released to the environsb. Gaseous radioactive effluents released to the environsc. Radiation exposure for all facility personnel39 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb16RevO2292016.docxLast edit February 29, 2016d. Radiation exposures monitored at site boundarye. As-built facility drawingf. Violation of the Safety Limitg. Violation of any Limited Safety System Setting (LSSS)h. Violation of any Limiting Condition of Operation (LCO)4. Requirement 6.8.1 (a) above does not include supporting documents such as checklists,logsheets and recorder charts, which shall be maintained for a period of at least one year.5. Applicable annual reports, if they contain any of the required information may be used asrecords in subsection 6.8.3 above.40 Accident Analysis MHAThe NRC licenses research and test reactors consistent with the NRC mission to ensure adequate protectionof the public health and safety and to promote and protect the environment. NUREG 1537 Part 2 Chapter 13Accident Analysis provides guidance and acceptable format and content for licensees to present regarding aMaximum Hypothetical Accident, MHA.Utilizing guidance from both NUJREG 1537 and NUJREG/CR-23 87, Credible Accident Analyses forTRIGA and TRIGA-Fueled Reactors, the bounding and limiting credible accident scenario is a fuel element failurewhich can occur at any time during normal operations or when the reactor is at rest and shutdown.In this worst case scenario a single element has been removed from the reactor and dropped to the floor ofthe reactor building outside of the biological shield. Fission products are released in air from the gap and thecladding and instantaneously and uniformly mix in the volume of the reactor building. The reactor facilityexhaust fans are not running and are closed during this event. No immediate protective functions are activated byradiation detectors or personnel present at the start of the event.In general, the escape of fission products from fuel or fueled experiments and their release to theunrestricted environment would be the most hazardous radiological accident conceivable at a non-power reactor.However, non-power reactors are designed and operated so that a fission product release is not credible for most.Therefore, this release under accident conditions can reasonably be selected as the MHADoses are calculated for air leakage out of the south side of the reactor. Internal, external, and shine dosesare determined for members of the public. Internal and external doses are determined for reactor personnel.Engineering analysis at the Maryland University Training Reactor (MUTR) has shown that 90% of the timeair leakage occurs out of the reactor on the north side entrances, predominantly through the roll up door andsignificantly less through a single entrance door. The remaining 10% of the time analysis has shown air leakagesites are located, on the eastern, southern and western sides of the reactor building, the most prominent site beingon the southern side and contributing 38% of the total air leakage to a single location. A highly conservativeapproach assumes the MHA's airborne radioactivity escapes continuously during the event from these predominantlocations from the onset of the event until the end of the leakage time. This is the source of the Maximally ExposedIndividual (MET) member of the public outside of the confinement space, and at locations downwind of theMUTR. Occupational personnel located within the MUTR are exposed to internal and external radiation from therelease during the time it takes to evacuate the reactor.On the south side of the reactor, 10% of the time, a steady nonstop air leakage rate from the reactor spacewill empty the air in 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> at a rate of 6.25E-3 m3 s-1. To maintain an overly conservative assumption, membersof the public remain in place for the duration of the event, and doses are determined over this time interval. Inreality individuals would not remain in these locations directly outside of confinement for the duration of an MHA.Reactor staff, Radiation Safety Office support personnel, and local response agency personnel would worktogether to secure and maintain acceptable perimeters relative to exposure and dose measurements surrounding theMULTR facility.Per 10 CFR 20.1301 the NRC regulations limit the internal and external dose to 100 mrem for members ofthe public; this includes doses incurred during an incident such as the MHA.1 Accident Analysis MHAInternal and external doses are accrued by occupational personnel present in the reactor at the start of theevent until they have evacuated the confinement space. Evacuation time is overly conservative and set at 5minutes.NUREG 2387 provides guidance and analysis for a 1MW TRIGA after 1 year of continuous operation atfull power, or 365MWd. The Maryland University Training Reactor (MUTR) is licensed as a 250 kW TRIGAreactor and therefore not capable of this level of operation. The analysis, inventories and released activities fromthe damaged fuel elements are thus scaled back to 25% in order to determine doses to occupational personnel andmembers of the general public.NUREG analysis assumes 50 elements were present in the referenced core and the central elementexperiences a greater than average burn up. At 1/50 or 2% of the total, the central element would contain 4% ofthe total activity in the core. The noble gas and radioiodine activities in this element are 3828.8 Ci of Krypton,9431 Ci of Iodine, and 3933 Ci of Xenon. A one year operation of the MUTR at 250 kW (25% of the NUREGexample) for 365 days is 91.25 MWd and the corresponding activities would be 957.2 Ci of Krypton, 2357.8 Ci ofIodine and 983.3 Ci of Xenon. The MIUTR contains 93 fueled elements whereas the NUREG/General Atomicsexample uses 50 elements in its assessment. Therefore, the activity per MUTR element would be decreased by afactor of 93/5 0, or 1.86. The scaled release activity table for the MUTR is shown below.Not all of the fission product activity would be released from the element as the fuel matrix acts strongly toretain the fission products. According to NUREG 2387 the gap activity fraction is approximately 1 .5x10-5.IsotopeReleased Activities(mCi)IsotopeReleasedActivities (mCi)Kr-83 m 0.2492 1-134 5.1290Kr-85m 0.5782 1-135 4.4621Kr-85 0.0097 Sr-89 2.0161Kr-87 1.1129 Sr-90 0.0625Kr-88 1.5911 Sr-91 2.6048Xe-133m 0.0782 Sr-92 2.9516Xe-133 4.5702 Cs-134 0.0060Xe-135m 1.2056 Cs-134m 0.00363Xe-135 2.0669 Cs-136 0.05241-131 2.1774 Cs-137 1.00001-132 3.3492 Cs-138 4.1531-133 3.8976In addition to using MUITR release activities, allowance is taken for air leakage, radionuclide decay andshielding over the course of the event. Air leakage rates were determined using engineering analysis of theM\UTR. Decay rates were calculated from the Chart of the Nuclides as well as the Health Physics andRadiological Health Handbook (HTPRRH). The reduction in shine dose due to shielding was determined from2 Accident Analysis MHAfigure 6.11 of the HPRRH, Average Half-Value and Tenth Value Layers of Shielding Materials (Broad Beams),obtained from the NBS Handbook 138 1982 and Wachsman and Drexier 1975.Dose conversion factors in Federal Guidance Reports 11 and 12 are utilized in calculating doses tooccupational personnel and members of the public. Doses to the public are from ground level release due to airleakage from the south side of the reactor building. Horizontal and vertical diffusion coefficients, whereapplicable, were taken from Cember, Introduction to Health Physics third edition as referenced from D.H. Slade,Meteorology and Atomic Energy Tech Inform, 1968. Diffusion coefficients for distances less than 100 meters areextrapolated. A Pasquill category F, moderately stable condition, was chosen for all releases as a conservativecategory.3 Accident Analysis MHAMethodology for Occupational Dose CalculationsThe following are the formulae used to calculate the occupational doses:Committed Dose Equivalent (CDE) to the thyroid and CEDE for reactor occupational personnelCEDE = E [.BR

• DCF1nt

• Ai[1- exp(-aef, leak tst)]]Deep Dose Equivalent (DDE) to reactor occupational personnelTerms used in the above Dose EquationsBR Breathing Rate, per NRC Guidance [in3 S1l]DCFint Internal Dose Conversion Factor per FGR 11 [mrem uCi"1]DCFext External Dose Conversion Factor per FGR 12 [mrem m3 uCi-' s-a]Ai MUTR released activity per nuclide [uCi]leak Effective removal rate or leak constant, (24i +t-2) [S'l]24i Decay constant per nuclide [s-1])Xl Leakage constant per nuclide [s"1]V MUTR volume [in3]tst Reactor personnel stay time (evacuation time) [s]4 Accident Analysis MHAMethodology for Public Dose CalculationsFor the 10% of the time the air in the MUTR predominantly flows out of the southern side of the reactor due toatmospheric conditions. This scenario describes the MEl since during the other 90% of the time doses to anymember of the public are drastically lower.The CEDE, DDE, and Shine doses to the MEl member of the public are calculated as follows:CEDE = BR

• DCF1nt

• Cavg

• TstayParameters in CEDE EquationBR -Breathing rate per NRC Guidance [in3 s']DCFint -Internal Dose Conversion Factor per FRG 11 [inrem uCi-']Cavg -Average concentration in room 1398 [uCi m3]Tstay -Stay time [s]DDE = DCFext

• Ca

• TsaParameters in DDE EquationDCFext -External Dose Conversion Factor per FRG 12 [toremo m3 uCi' s1]Cavg -Average concentration in room 1398 [uCi m-3]Tstay -Stay time [s]Shine Dose = w* F**-(1 -e-.a), *In(2~Parameters in Shine Dose EquationF- Gamma constant for nuclide [remn hrI Ci-1 in2]C -Average concentration in the cloud in the MUTR [Ci m-3]r -Radius of the cloud in the MUJTR [in]h -Dose location distance from the surface of the cloud [in]S- lien, the linear energy absorption coefficient [m1]d -Diameter of the cloud in the MUTR [in]5 Accident Analysis MHASummary of DosesCEDE Occupational10.2 mremDDE Occupational1.62 mremTEDE Occupational11.82 mremCEDE publicAt MEl1-88.5 mremDDE publicAt MEl1-2.78 mremShine Dose PublicAt ME1-7.325 mremTEDE PublicMEI -98.605 mrem6 UNIVERSIYOGLENN L. MARTIN INSTITUTE OF TECHNOLOGYA. JAMES CLARK SCHOOL OF ENGINEERINGDepartment of Materials Science & EngineeringNuclear Reactor & Radiation FacilitiesTimothy W. Koeth, DirectorBuilding 090College Park, Maryland 20742-2115301.405.4952 TEL 301.405.6327 FAX609.577.8790 CELLkoethi@umd.eduFebruary 29, 2016Document Control DeskUnited States Nuclear Regulatory CommissionWashington, D.C. 20555-0001

SUBJECT:

UNIVERSITY OF MARYLAND -REQUEST FOR ADDITIONALINFORMATION RE: FOR THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-70 THE MARYLAND UNIVERSITY TRAINING REACTOR DOCKET NO. 50-166Enclosed please find the response to the RAI dated August 24, 2015 for the University ofMaryland Training Reactor (MUTR), License No. R-70; Docket No. 50-166.I declare under penalty of perjury that the foregoing is true and correct.Timothy W. Koeth, Assistant Research Professor and DirectorUniversity of Maryland Training Reactor & Radiation Facilities10/p Response To:OFFICE OF NUCLEAR REACTOR REGULATIONREQUEST FOR ADDITIONAL INFORMATIONFOR THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-70THE MARYLAND UNIVERSITY TRAINING REACTORDOCKET NO. 50-1661. MUTR SAR, Section 4.5.2, "Reactor Core Physics Parameters," (Ref. 1) lists three reactivity coefficientsand their associated values. However, it appears the combined reactivates have a positive value.NUREG-2537, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-PowerReactors Standard Review Plan and Acceptance Criteria," Section 4.5.2 provides guidance that ananalysis should show that reactivity coefficients are sufficiently negative to prevent or mitigatedamaging reactor transients. Describe what constitutes a power coefficient and show how overallreactivity coefficients are negative; or justify why the current method is acceptable.Fuel Temperature Coefficient: -1 1.2¢/ 0CModerator Temperature Coefficient: +3.0 ¢/°CReactor Power Coefficient: -0.53 ¢!kWListed above are the reactivity coefficients of MUIR. At first glance it may seem as though the sumcoefficient has a positive value. Howvever, the sole positive contribution to the reactivity is from themoderator temperature. The moderator temperature increases very slowly in comparison to the fueltemperature due to its heat capacity and the fact that the fuel is what is heating the water. Additionally,these temperature increases occur only at powers above a few kilowatts. As a result there is alreadysignificant negative reactivity added before any positive reactivity results from an increase in moderatortemperature.2. MUTR SAR Section 4.6, "Thermal Hydraulic Design," (Ref. 1) or the MUTR thermalhydraulic analysis(Ref. 5) does not include a departure from nucleate boiling ratio (DNBR). NUREG-2537, Guidelines forPreparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content,Section 14, Appendix 14.1, Section 2.1.2 provides guidance that a DNBR should be calculated with aminimum value of 2. Provide a DNBR analysis that indicates a minimum value of at least 2, or justify whyone is not needed.A conservative calculation of the DNBR for MUTR has returned a value of 2.96 while operating at 600kWand inlet temperature of 92 degrees Celsius. Support documentation for this DNBR is attached indocument titled "Support Calculations for MUTR's Maximum Inlet Temperature".3. MUTR SAR Section 11.1.7, "Environmental Monitoring," states that the operation of the facility willhave no negative impact on the environment. The MUTR environmental monitoring program resultswere provided in response to RAIs No. 47 and No. 72 (Refs. 6 and 2, respectively). However, the resultsare from 2004, and therefore, are out of date. NUREG-1537, Guidelines for Preparing and ReviewingApplications for the Licensing of Non-Power Reactors Format and Content, Section 11.1.7 providesguidance that an appropriate monitoring program should contain probable pathways to people, andtrends of recorded results. Provide updated information on the environmental monitoring program orjustify why it is not needed.

Fixed environmental area monitors are located at the MUTR restricted area boundary and at locationson the university campus, to record and tract the potential radiological impact the MUTR operationshave on the surrounding environment.Monitors are exchanged and analyzed at frequency not to exceed once per calendar quarter. Recordsare maintained in accordance with 10 CFR 20.2103. Historically, and over the past 5 years, dosedeterminations for members of the public based upon this program, indicate doses to the public are incompliance with the limits of 20 CFR 20.1301.In addition to fixed environmental area monitors, exposure rate Geiger Muller measurements are takenmonthly in unrestricted areas outside of the MUTR boundary to monitor potential exposure to thepublic and assist in maintaining dose to the public As Low As Reasonably Achievable.4. The following RAIs are based on the maximum hypothetical accident (MHA), "Accident Analysis MHA"(Ref. 3). NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content, Section 13.2 provides guidance for accident analysis, anddetermination of consequences. Additional information or clarification is needed in the following areas.a) Guidance in NUREG-2537, Section 13.2, item (3) states that assumptions that change thecourse of events and mitigate consequences (including automatic functions and operatoractions) until a stabilized condition has been reached should be described. The accident analysisappears to be limited to uniform mixing of fission products in the reactor room and subsequentelevated or ground release. It is not clear (i) what the initial condition of the ventilation fans are;(ii) if radiation detectors or operator action initiate protective functions; (iii) if two separatescenarios are analyzed; (iv) what the sequence for the analyzed exposure times (question 4(e)iiof this document); or (v) when a stable condition would be reached. Provide an updated analysisdescribing the sequence of events including initiation of engineered safety features to mitigatean accident, or justify why the current method is acceptable.See Attachment Titled "Accident Analysis MHA"b) Guidance in NUREG-2537, Section 13.2, item (5) states, in part, that methods andassumptions developed for the "Radiation Protection Program and Waste Management,"chapter of the SAR should be adapted as appropriate for the analysis. Submitted informationshould allow the results to be independently verified. The following parameters require furtherclarification:i. The total confinement leakage rate of 0.0356 meters cubed per second (RAI No. 2A,Ref. 4) appears to conflict with the assumed leakage rate of 0.0242 meters cubed persecond (page 1, Ref. 3), and room leakage parameter of 0.00236 meters cubed persecond (pages 4, 6, 8, 10, and 12, Ref. 3).See Attachment Titled "Accident Analysis MHA"ii. It appears the breathing rate parameter of 3.3x20-04 meters cubed per second (pages4, 8, and 12, Ref. 3) is inconsistent with the breathing rate of 4.27x20-04 meters cubedper second (pages 16 and 17, Ref. 3).

See Attachment Titled "Accident Analysis MHA"iii. The release height of 7.25 meters and a wind speed of 2.32 meters per second areprovided as input parameters for "HOTSPOT" (page 16, Ref. 3). However, dispersionvalues for various distances and atmospheric stability classes (page 3, Ref. 3) cannot beverified using these input parameters. Provide an updated analysis clearly statingconfinement leakage, breathing rates, release heights, and wind speed parameters asnecessary before each series of computations, or justify why the current method isacceptable.See Attachment Titled "Accident Analysis MHA"c) Guidance in NUREG-1537, Section 13.2, item (6) provides for defining the source termquantity of radionuclides. The fission product inventory is 25 percent equivalent of thosedescribed in NUREG/CR-2387 (page 2, Ref. 3). It appears the activities of Cesium and Strontiumare less than 25 percent of those values listed in NUREG/CR-2387. Provide an updated analysisusing consistent methodology for determining the source term, or justify why the currentmethod is acceptable.See Attachment Titled "Accident Analysis MHA"d) Guidance in NUREG-1537, Section 13.2, item (6) provides for describing a source term thatcould cause direct or scattered radiation exposure. Ground shine was analyzed using"HOTSPOT," at 10 meters (pages 16 and 17, Ref. 3). However, direct or scattered radiation tomembers of the public located 6.096 meters from the roll up door or in hallway 1398 (RAI No.1G, Ref. 4) due to the uniform distribution of fission products within the reactor room is notconsidered. NUREG-1537, Section 13.2, item (7) provides guidance for evaluating exposure of amember of the public until the situation is terminated or the person is moved. Provide anupdated analysis to include direct or scattered radiation exposure to members of the publicspecific to the MUTR facility; or justify why the current method is acceptable.See Attachment Titled "Accident Analysis MHA"e) Guidance for facility specific consequences is provided in NUREG-1537, Section 13.2, item (7).The guidance states, in part, that exposure conditions should account for staff and members ofthe public specific to the facility until the situation are stabilized. The following locations formembers of the public and times of exposure require further clarification:i. Potential radiological consequences to members of the public in unrestricted areas areevaluated at 10, 100, 200, and 300 meters (page 18, Ref. 3). However, the MUTR SAR,Section 2.1.1.2, "Boundary and Zone Area Maps," (Ref. 1) list the nearest on-campusresidence hall and nearest off campus public residence from the reactor building atapproximately 230 and 370 meters, respectively. A maximum exposed members of thepublic located at 6.096 meters from the roll up door and in hallway 1398 (RAI No. 1C,Ref. 4) do not appear to correlate to the nearest distance of 10 meters. Guidance forother locations of interest that may be applicable to the MUTR facility is provided inNUREG-1537, Section 11.1.1.1.

See Attachment Titled "Accident Analysis MHA"ii. Public exposure from a ground release use 72,050 seconds (pages 4 and 6, Ref. 3);public exposure from an elevated release uses 650 seconds (pages 8 and 10, Ref. 3);occupational exposure uses 300 seconds (pages 12 and 14, Ref. 3); and exposure to areceptor uses 0.34 and 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, respectively (pages 16 and 17, Ref. 3). It is not clearhow to chronologically view the events, or if exposure times are consistent with oneanother. Provide an updated analysis clearly indicating exposure times and subsequentdose estimates to a maximum exposed member of the public at the facility boundary,nearest residence, and/or other location of interest as necessary, or justify why thecurrent method is acceptable.See Attachment Titled "Accident Analysis MHA"5. MUTR proposed TS 3.1, "Reactor Core Parameters," Specification (5) describes reactivity coefficientsat the MUTR (Ref. 7). NUREG-1537, Guidelines for Preparing and Reviewing Applications for theLicensing of Non-Power Reactors Format and Content, Section 14, Appendix 14.1, Section 4 providesguidance that certain limiting conditions for operations have accompanying surveillance requirementsto include test, method, frequency, and acceptability. It appears the reactivity coefficients do not have asurveillance requirement. Provide a surveillance specification for TS 3.1 Specification (5), or justify whyone is not necessary.See updated TS 3.16. MUTR proposed TS 3.7, "Limitations On Experiments," Specification (4) describes limits onexperiments (Ref. 7). Specification (4) describes explosive materials in quantities greater than 25milligrams and less than 25 milligrams, but does not include quantities equal to 25 milligrams. Provide arevised TS 3.7 Specification (4) to provide for explosive material quantities equal to 25 milligrams, orjustify why no change is necessary.See updated TS 3.77. MUTR proposed TS 4.1, "Reactor Core Parameters," Specification (5) describes annual inspections offuel elements, but does not appear to have an associated surveillance interval with its periodicity (Ref.7). Acceptable surveillance intervals are provided in the American Nuclear Standards Institute,Incorporated/American Nuclear Society (ANSI/ANS) 15.1-2007, Section 4. Add an interval to TS 4.1,Specification (5) or justify why one is not necessary.See Updated TS 4.1, Specification (5)8. The Basis in MUTR proposed TS 4.4, "Confinement," references a "minimum leakage rate assumed inthe SAR," however, actual confinement leakage values were determined (Refs. 7 and 4). NUREG-1537,Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors Formatand Content, Section 14, Appendix 14.1, Section 1.2.2 provides guidance that the proposed TS basisshould be reference to the facility's analysis. Provide a revision to proposed TS 4.4 to include aqualitative reference, or justify why no change is necessary.

The Bases of TS 4.4 references the SAR. The SAR is the facility's analysis. Therefore no change isnecessary.9. MUTR proposed TS 5.2, "Reactor Primary Coolant System," Specification (1) Basis describes thermal-hydraulic analysis for "other TRIGA reactors," (Ref. 7). NUREG-1537, Guidelines for Preparing andReviewing Applications for the Licensing of Non-Power Reactors Format and Content, Section 14,Appendix 14.1, Section 1.2.2 provides guidance that the proposed TS Basis should reference the facility'sanalysis. It appears from the thermal-hydraulic analysis that actual values were determined (Ref. 5).Provide a revision to proposed TS 5.2 to include a qualitative reference, or justify why no change isnecessary.See updated 1S 5.210. MUIR thermal-hydraulic analysis shows core locations for the instrumented fuel element (IFE) (Ref.5). MUTR proposed TSs do not appear to address these core locations. NUREG-2537, Guidelines forPreparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content,Section 14, Appendix 14.1, Section 3.1 item (4) provides guidance that TSs should include criteria forrestricting certain fuel bundles from core positions so that assumptions used in the development safetylimits are met. NUREG-1537, Section 14, Appendix 14.1, Section 1.2.2 provides guidance that a TS shouldinclude a basis for each specification. Propose a TS including a basis that incorporates acceptable IFElocations, or justify why no change is necessary.See updated TS 2.211. MUIR proposed TS 6.0, "Administration," describes administrative control of the MUIR facility (Ref.7). Additional information and clarification is needed in the following areas.a) Figure 6.1, "MUIR Position in University of Maryland Structure," and Figure 6.2, "MUIROrganizational Structure," show solid-line and dashed-line connections, but appear to bemissing a description. The lines are not identified in a leger or described within TS Section 6.0,"Administration," as provided by guidance in ANSl/ANS-25.2-2007, Figure 1. Provide adescription of the connection lines in Figures 6.1 and 6.2.See updated TS Figures 6.1 & 6.2b) Figure 6.2, "MUIR Organizational Structure," shows members of the MUIR organizationincluding staff and management. However, the TSs do not appear to correlate the MUIRmembers of the organization with the four assignment levels as provided in in ANSI/ANS-25.2-2007, Section 6.1.1. Guidance regarding expected responsibilities for assigned levels is providedin ANSl/ANS-25.4-2007, Section 3. Clarify the level of assignment in the TSs for the membersshown in Figure 6.2.See updated TS Figure 6.2c) ANSl/ANS-15.1-2007, Section 6.1.2 provides guidance that management not only beresponsible for policies and operation, but shall also adhere to all requirements of the operatinglicense and TSs. MUIR proposed TS 6.1.2, "Responsibility," describes specific responsibilities forthe facility director, but does not appear to provide a description of responsibilities of other MUTR members shown in Figure 6.2. Clarify the specific responsibilities for all the MUTRmember shown in Figure 6.2.TS 6.1.2 has been rewritten to include responsibilities of all MUTR members shown in Figure 6.212. MUTR proposed TS 6.1.3, "Facility Staff Requirements," Specification (1) describes facility staffingrequirements when the "reactor is operating" (Ref. 7). However, ANSI/ANS-15.1-2007, Section 6.1.3provides guidance that the minimum reactor staffing is required when the reactor is "not secured."Provide a revision to proposed TS 6.1.3 or justify why no change is necessary.See updated TS 6.1.3 Specification (1)13. MUTR proposed TS 6.2.1.2, "Reactor Safety Committee Review Function," Specification (3) states,"All new experiments or classes of experiments that could affect reactivity or result in the release ofradioactivity," (Ref. 7). However, "new experiment," is not defined, nor is the terminology consistentwith the MUTR proposed TS Definition 1.7 or Specification 6.5. It is not clear which category ofexperiments are applicable in proposed TS 6.2.1.2. Provide a revised TS 6.2.1.2 to delineate whichexperiments require review by the Reactor Safety Committee or justify why no change is necessary.No change is necessary. The question places in quotes "new experiment" that is not a classification ofan experiment. Definition 1.7 classifies experiments as "Routine," "Modified Routine," and "Special."Clearly by definition, "Routine" is not a new experiment. Modified Routine and Special Experiments areconsidered new.14. MUTR proposed TS 6.5, "Experiment Review And Approval," Specification (3) uses the term "desiredalternate," which appears inconsistent with other alternatives described elsewhere in the TSs (Ref. 7).Furthermore, ANSI/ANS-15.1-2007 uses the word "designated," throughout the guidance. Provide arevision to the proposed TS 6.5 or justify why no change is necessary.See updated TS 6.515. MUTR proposed TS 6.7.2, "Special Reports," Specification (1) references TS Definition 1.27 (Ref. 7).However, TS Definition 1.27 is "Reactor Operator," and TS Definition 1.32 is "Reportable Occurrence." Itappears Definition 1.27 is erroneously used in proposed TS 6.7.2. Provide a revision to proposed TS 6.7.2or justify why no change is necessary.See updated TS 6.7.216. The following typographical errors were noticed. Consider reviewing the proposed TSs for othertypographical or formatting errors and propose corrections as necessary.a) MUTR proposed TS Definition 1.37 may contain a grammatical error,See updated TS 1.37b) MUTR proposed TS Definition 1.41 is numbered as 1.401,See updated TS 1.41c) MUTR proposed TS 4.4 Specification may contain a grammatical error, See updated TS 4.4d) MUTR proposed TS 5.2.1 Specification (1) appears to erroneously use "connective,"See updated TS 5.2e) MUTR proposed TS 5.3.1 Specification (4) appears to be missing,See updated TS 5.3.1f) MUTR proposed TS 5.3.2 Specification (1) states the control rods will contain borated graphiteBvC, andSee updated TS 5.3.2g) MUTR proposed TS 5.4 Specification (3) appears to be missing.See updated TS 5.4OTHER CHANGES TO TSs:TS 3.7 -The Roman numeral '11' was changed to '2'TS 5.3.1 -"w/o" was changed to "weight %"TS 6.6.2 -"1.27" was changed to "1.32"TS Figure 6.1 -"Vice President Academic Affairs" was changed to "Provost & Senior Vice President" TECHNICAL SPECIFICATIONSFOR THEMARYLAND UNIVERSITY TRAINING REACTORLicense Number R-70Docket Number 50-166Submitted to United States Nuclear Regulatory Commission29 February 2016(Superseding 27 September 2011 Submission)

O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16Rev02292016.docxLast edit February 29, 2016TABLE OF CONTENTSTABLE OF CONTENTS ........iLIST OF TABLES .........ivLIST OF FIGURES. .........v1.0 DEFINITIONS .........11.1 ALARA .........11.2 Channel .........11.3 Confinement .11.4 Control Rod Guide Tube .. ..11.5 Core Configuration ........11.6 Excess Reactivity ........11.7 Experiment .........11.8 Experimental Facilities .......21.9 Experiment Safety Systems .......21.10 Four Element Fuel Bundle .......21.11 Fuel Element .........21.12 Fueled Device. ........21.13 Full Power .........21.14 Instrumented Element. .......21.15 Isolation .........21.16 Limiting Conditions for Operation ......21.17 Limiting Safety System Setting ......21.18 Measuring Channel ........21.19 Measured Value .......21.20 Moveable Experiment. .......21.21 On Call .........31.22 Operable .........31.23 Operating .........31.24 Reactivity Worth of an Experiment ......31.25 Reactor Console Secured .......31.26 Reactor Operating ........31.27 Reactor Operator ........31.28 Reactor Safety Systems ......31.29 Reactor Secured ........31.30 Reactor Shutdown ........31.31 Reference Core Condition .......41.32 Reportable Occurrence .......41.33 Rod-Control .........41.34 Safety Channel ..... ..41.35 Safety Limit .........4 O:\M UTR\2016M UTRLivingDocs\WorlingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20161.36 Scram Time ...41.37 Secured Experiment ........41.38 Secured Shutdown ......51.39 Senior Reactor Operator ....51.40 Shall, Should, May ........51.41 Shutdown Margin ........51.42 Shutdown Reactivity .......51.43 Standard Core. ........51.44 Steady State Mode ........51.45 Three Element Fuel Bundle .......51.46 True Value .........51.47 Unscheduled Shutdown .52.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING62.1 Safety Limit .........62.2 Limiting Safety System Settings. ......63.0 LIMITING CONDITIONS FOR OPERATION. .... 83.1 Reactor Core Parameters. .......83.2 Reactor Control and Safety Systems ......93.3 Primary Coolant System. .. ......133.4 Confinement .........143.5 Ventilation Systems ........143.6 Radiation Monitoring System and Effluents .....153.6.1 Radiation Monitoring System. .....153.6.2 Effluents. ........163.7 Limitations on Experiments .......174.0 SURVEILLANCE REQUIREMENTS ......194.1 Reactor Core Parameters .......194.2 Reactor Control and Safety Systems ......204.3 Primary Coolant System .......214.4 Confinement .........224.5 Ventilation System ........224.6 Radiation Monitoring System and Effluents .....234.6.1 Radiation Monitoring System. .....234.6.2 Effluents ........234.7 Experiments .........245.0 DESIGN FEATURES ........255.1 Site Characteristics ........255.2 Reactor Coolant System .......25ii O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20165.3 Reactor Core and Fuel .......265.3.1 Reactor Fuel ........275.3.2 Control Rods ........275.4 Fissionable Material Storage .......286.0 ADMINISTRATION ....296.1 Organization ...296.1.1 Structure ..296.1.2 Responsibility. .......296.1.3 Facility Staff Requirements ......296.1.4 Selection and Training of Personnel .....336.2 Review and Audit .......336.2.1 Reactor Safety Committee ......336.2.1.1 Reactor Safety Committee Charter and Rules ...336.2.1.2 Reactor Safety Committee Review Function ...346.2.1.3 Reactor Safety Committee Audit Function ...346.2.2 Audit of ALARA Program ......356.3 Radiation Safety ........356.4 Operating Procedures. ....... 356.5 Experiment Review and Approval ......366.6 Required Actions .......366.6.1 Actions to be Taken in Case of Safety Limit Violation ..366.6.2 Actions to be Taken in the Event of a Reportable Occurrence .376.7 Reports. .........376.7.1 Annual Operating Report ......376.7.2 Special Reports .......386.7.3 Unusual Event Report ......396.8 Records .........393 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016LIST OF TABLESTable 3.1 Reactor Safety Channels: Scram Channels ....11Table 3.2 Reactor Safety Channels: Interlocks .....11Table 3.3 Reactor Safety Channels: Scram Channel Bases ....12Table 3.4 Reactor Safety Channels: Interlock Bases ....12Table 3.5 Minimum Radiation Monitoring Channels ....164 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016LIST OF FIGURESFigure 6.1 MUTR Position in University of Maryland Structure ..30Figure 6.1 MUTR Organizational Structure ...... 315 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016Included in this document are the Technical Specifications and the "Bases" for the TechnicalSpecifications. These bases, which provide the technical support for the individual technicalspecifications, are included for information purposes only. They are not part of the TechnicalSpecifications, and they do not constitute limitations or requirements to which the licensee must adhere.1.0 DEFINITIONS1.1 ALARA .(acronym for "as low as is reasonably achievable") means making every reasonableeffort to maintain exposures to radiation as far below the dose limits in 10 CFR Part 20 as ispractical consistent with the purpose for which the licensed activity is undertaken, taking intoaccount the state of technology, the economics of improvements in relation to state of technology,the economics of improvements in relation to benefits to the public health and safety, and othersocietal and socioeconomic considerations, and in relation to utilization of nuclear energy andlicensed materials in the public interest.1.2 CHANNEL -A channel is the combination of sensors, lines, amplifiers, and output devices whichare connected for the purpose of measuring the value of a parameter.1. Channel Calibration -A channel calibration is an adjustment of the channel such that itsoutput corresponds with acceptable accuracy to known values of the parameter which thechannel measures. Calibration shall encompass the entire channel, including equipmentactuation, alarm, or trip and shall be deemed to include a channel test.2. Channel Check -A channel check is a qualitative verification of acceptable performance byobservation of channel behavior, or by comparison of the channel with other independentchannels or systems measuring the same variable.3. Channel Test -A channel test is the introduction of a signal into the channel to verify' that it isoperable.1.3 CONFINEMENT -Confinement means a closure on the overall facility that controls themovement of air into it and out, thereby limiting release of effluents, through a controlled path.1.4 CONTROL ROD GUIDE TUBE -Hollow tube in which a control rod moves.1.5 CORE CONFIGURATION -The core consists of 24 fuel bundles, with a total of 93 fuelelements, arranged in a rectangular array with one bundle displaced for the pneumaticexperimental system; three control rods; and two graphite reflectors.1.6 EXCESS REACTIVITY -Excess reactivity is that amount of reactivity that would exist if allcontrol rods were moved to the maximum reactive condition from the point where the reactor isexactly critical (kef= 1)1.7 EXPERIMENT -Any operation, hardware, or target (excluding devices such as detectors, foils,etc.), that is designed to investigate non-routine reactor characteristics or that is intended forirradiation within the pool, on or in a beamport or irradiation facility, and that is not rigidlysecured to a core or shield structure so as to be part of their design.1. Routine Experiments -Routine Experiments are those which have been previously performedin the course of the reactor program.1 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20162. Modified Routine Experiments -Modified routine experiments are those which have not beenperformed previously but are similar to routine experiments in that the hazards are neithergreater nor significantly different than those for the corresponding routine experiments.3. Special experiments -Special experiments are those which are not routine or modified routineexperiments.1.8 EXPERIMENTAL FACILITIES -Experimental facilities are facilities used to performexperiments and include, for example, the beam ports, pneumatic transfer systems and any in-core facilities.1.9 EXPERIMENT SAFETY SYSTEMS -Experiment safety systems are those systems, includingtheir associated input circuits, which are designed to initiate a scram for the primary purpose ofprotecting an experiment or to provide information which requires manual protective action to beinitiated.1.10 FOUR ELEMENT FUEL BUNDLE -The 4-element fuel bundle consists of an aluminumbottom, 4 stainless steel clad fuel elements and aluminum top handle.1.11 FUEL ELEMENT -A fuel element is a single TRIGA fuel rod.1.12 FUELED DEVICE -An experimental device that contains fissionable material.1.13 FULL POWER -Full licensed power is defined as 250 kW.1.14 JINSTRUMENTED ELEMENT -An instrumented element is a special fuel element in which asheathed chromel-alumel or equivalent thermocouple is embedded in the fuel.1.15 ISOLATION -Isolation is the establishment of confinement, closing of the doors leading fromthe reactor bay area leading into the balcony area on the top floor, the door to the reception areaon the ground floor, and the building exterior doors.1.16 LIMITING CONDITIONS FOR OPERATION -Limiting conditions for operation are the lowestfunctional capability or performance levels of equipment required for safe operation of thefacility.1.17 LIMITING SAFETY SYSTEM SETTING- Limiting safety system settings (LSSS) for nuclearreactors are settings for automatic protective devices related to those variables having significantsafety functions.1.18 MEASURING CHANNEL -A measuring channel is the combination of sensor, interconnectingcables or lines, amplifiers, and output device, which are connected for the purpose of measuringthe value of a variable.1.19 MEASURED VALUE -The measured value is the value of a parameter as it appears on theoutput of a channel.1.20 MOVEABLE EXPERIMENT -A movable experiment is one where it is intended that all or partof the experiment may be moved in or near the core or into and out of the reactor while thereactor is operating.2 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTech nicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20161.21 ON CALL -A senior operator is available "on call" if the senior operator is either on the CollegePark campus or within 10 miles from the facility and can reach the facility within one half hourfollowing a request.1.22 OPERABLE -Operable means a component or system is capable of performing its intendedfunction.1.23 OPERATING -Operating means a component or system is performing its intended function.1.24 REACTIVITY WORTH OF AN EXPERIMENT -The reactivity worth of an experiment is theValue of the reactivity change that results from the experiment being inserted into or removedfrom its intended position.1.25 REACTOR CONSOLE SECURED -The reactor console is secured whenever all scrammable*rods have been fully inserted and verified down and the console key has been removed from theconsole.1.26 REACTOR OPERATING -The reactor is operating whenever it is not secured or shutdown.1.27 REACTOR OPERATOR -A reactor operator (RO) is an individual who is licensed by the U.S.Nuclear Regulatory Commission (NRC) to manipulate the controls of the reactor.1.28 REACTOR SAFETY SYSTEMS -Reactor safety systems are those systems, including theirassociated input channels, which are designed to initiate automatic reactor protection or toprovide information for initiation of manual protective action. Manual protective action isconsidered part of the reactor safety system.1.29 REACTOR SECURED -The reactor is secured when:1. Either there is insufficient moderator available in the reactor to attain criticality or there isinsufficient fissile material present in the reactor to attain criticality under optimum availableconditions of moderator and reflection, or2. The following conditions exist:a. All control devices (3 control rods) are fully inserted;b. The console key switch is in the off position and the key is removed from the lock;c. No work is in progress involving core fuel, core structure, installed control rods, orcontrol rod drives unless they are physically decoupled from the control rods; andd. No experiments in or near the reactor are being moved or serviced that have, onmovement, the smaller of: a reactivity worth exceeding the maximum value allowedfor a single experiment, or a reactivity of one dollar.1.30 REACTOR SHUTDOWN -The reactor is shut down if it is subcritical by at least one dollar in thereference core condition with the reactivity worth of all installed experiments included and thefollowing conditions exist:a. No work is in progress involving core fuel, core structure, installed control rods, orcontrol rod drives unless they are physically decoupled from the control rods;3 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016b. No experiments are being moved or serviced that have, on movement, a reactivityworth exceeding the maximum value allowed for a single experiment, or one dollar,whichever is smaller.1.31 REFERENCE CORE CONDITION -The reference core condition is the reactivity condition ofthe core when it is at 20 °C and the reactivity worth of xenon is zero (i.e., cold, clean, andcritical).1.32 REPORTABLE OCCURRENCE -A reportable occurrence is any of the following:1. Operation with actual safety-system settings for required systems less conservative than theLimiting Safety-System Settings specified in technical specifications 2.2.2. Operation in violation of the Limiting Conditions for Operation established in the technicalspecifications.3. A reactor safety system component malfunction which renders or could render the reactorsafety system incapable of performing its intended safety function unless the malfunction orcondition is discovered during maintenance tests. (Note: Where components or systems areprovided in addition to those required by the Technical Specifications, the failure of the extracomponents or systems is not considered reportable provided that the minimum number ofcomponents or systems specified or required performs their intended reactor safety function.)4. An unanticipated or uncontrolled change in reactivity greater than one dollar.5. Abnormal and significant degradation in reactor fuel, or cladding, or both, coolant boundary,or confinement boundary (excluding minor leaks) where applicable.6. An observed inadequacy in the implementation of administrative or procedural controls suchthat the inadequacy causes or could have caused the existence or development of an unsafecondition with regard to reactor operations.1.33 ROD-CONTROL -A control rod is a device fabricated from neutron absorbing material which isused to establish neutron flux changes and to compensate for routine reactivity losses. A controlrod may be coupled to its drive unit allowing it to perform a safety function when the coupling isdisengaged.1.34 SAFETY CHANNEL -A safety channel is a measuring channel in the reactor safety system.1.35 SAFETY LIMIVIT -Safety limits are limits upon important process variables that are found to benecessary to reasonably protect the integrity of certain of the physical barriers that guard againstthe uncontrolled release of radioactivity.1.36 SCRAM TIME -Scram time is the elapsed time between the initiation of a scram signal by eitherautomated or operator initiated action and the time required for the control rods to reach a fully insertedposition into the core.1.37 SECURED EXPERIMENT -A secured experiment is any experiment, experimental facility, orcomponent of an experiment that is held in a stationary position relative to the reactor bymechanical means. The restraining forces must be substantially greater than those to which theexperiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which arenormal to the operating environment of the experiment, or by forces that can arise as a result of4 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016credible malfunctions.1.38 SECURED SHUTDOWN -Secured shutdown is achieved when the reactor meets therequirements of the definition of "reactor secured" and the facility administrative requirementsfor leaving the facility with no licensed reactor operators present.1.39 SENIOR REACTOR OPERATOR -A senior reactor operator (SRO) is an individual who islicensed by the NRC to direct the activities of reactor operators.1.40 SHALL, SHOULD, MVAY -The word ,,shallee is used to denote a requirement; the word ,,shouldeeis used to denote a recommendation; and the word ,,may" is used to denote permission, neither arequirement nor a recommendation.1.41 SHUTDOWN MARGIN -Shutdown margin is the minimum shutdown reactivity necessary toprovide confidence that the reactor can be made subcritical by means of the control and safetysystems starting from any permissible operation condition and with the most reactive rod in itsmost reactive position, and that the reactor will remain subcritical without further operator action.1.42 SHUTDOWN REACTIVITY -Shutdown reactivity is the value of the reactivity of the reactorwith all control rods in their least reactive position (e.g., inserted). The value of shutdownreactivity includes the reactivity value of all installed experiments and is determined with thereactor at ambient conditions.1.43 STANDARD CORE -A standard core is an arrangement of standard TRIGA fuel in the reactorgrid plate.1.44 STEADY STATE MODE -Steady state mode operation shall mean operation of the reactor withthe mode selector switch in the STEADY STATE position.1.45 THREE ELEMENT FUEL BUNDLE -The 3-element fuel bundle consists of an aluminumbottom, 3 stainless steel clad fuel elements, 1 control rod guide tube, and aluminum top handle.1.46 TRUE VALUE -The true value is the actual value of a parameter.1.47 UNSCHEDULED SHUTDOWN -An unscheduled shutdown is defined as any unplannedshutdown of the reactor caused by actuation of the reactor safety system, operator error,equipment malfunction, or a manual shutdown in response to conditions which could adverselyaffect safe operation, not to include shutdowns which occur during testing or check-outoperations.5 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20162.0 SAFETY LIMITS AN]) LIMITING SAFETY SYSTEM SETTING2.1 SAFETY LIMITApplicabilityThis specification applies to the temperature of the reactor fuel.ObjectiveThe objective is to define the maximum fuel element temperature that can be permitted withconfidence that no damage to the fuel element cladding will result.SpecificationThe temperature in a standard TRIGA fuel element shall not exceed 1000 °C under anyconditions of operation, with the fuel fully immersed in water.BasisThe important parameter for TRIGA reactor is the UZrH fuel element temperature. Thisparameter is well suited as a single specification especially since it can be measured. A loss in theintegrity of the fuel element cladding could arise from a build-up of excessive pressure betweenthe fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressureis caused by the presence of air, fission product gases, and hydrogen from the dissociationof the hydrogen and zirconium in the fuel-moderator. The magnitude of this pressure isdetermined by the fuel-moderator temperature and the ratio of hydrogen to zirconium. The dataindicate that the stress in the cladding due to hydrogen pressure from the dissociation ofZrHx will remain below the ultimate stress if the temperature in the fuel does not exceed 1000 °Cand the fuel cladding is water-cooled.It has been shown by experience that operation of TRIGA reactors at a power level of 1000 kWwill not result in damage to the fuel. Several reactors of this type have operated successfully forseveral years at power levels up to 1500 kW. Analysis and measurements on other TRIGAreactors have shown that a power level of 1000 kW corresponds to a peak fuel temperature ofapproximately 400 °C.2.2 LIMITIN4G SAFETY SYSTEM SETTINGSApplicabilityThis specification applies to the reactor scram setting that prevents the reactor fuel temperaturefrom reaching the safety limit.ObjectiveThe objective is to provide a reactor scram to prevent the safety limit (fuel element temperature of1000 °C) from being reached.6 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016SpecificationThe limiting safety system setting shall be 175 °C as measured by the instrumented fuel element.The WFE shall be placed in the position as described in the core analyzed in the SAR. If the WFE isplaced in any position other than D8 in the grid plate, an analysis must be performed. The analysisshall indicate that the proposed location shall have a peaking factor not less than 50% of thehighest fuel element in the core.BasisA Limiting Safety Setting of 175 °C provides a safety margin of 650 °C. A part of the safetymargin is used to account for the difference between the temperature at the hot spot in the fuel andthe measured temperature resulting from the actual location of the thermocouple. If thethermocouple element is located in the hottest position in the core, the difference between the trueand measured temperatures will be only a few degrees since the thermocouple junction is at themid-plane of the element and close to the anticipated hot spot. If the thermocouple element islocated in a region of lower temperature, such as on the periphery of the core, the measuredtemperature will differ by a greater amount from that actually occurring at the core hot spot.Calculations have shown that if the thermocouple element were located on the periphery of thecore, the true temperature at the hottest location in the core will differ from the measuredtemperature by no more than a factor of two. Thus, with the WFE positioned in the locationspecified by the license, when the temperature in the thermocouple element reaches the setting of175 °C, the true temperature at the hottest location would be no greater than 350 °C, providing amargin to the safety limit of at least 650 °C. This margin is ample to account for the remaininguncertainty in the accuracy of the fuel temperature, measurement channel, and any overshoot inreactor power resulting from a reactor transient during steady state mode operation.7 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163.0 LIMITING CONDITIONS FOR OPERATION3.1 REACTOR CORE PARAMETERSApplicabilityThese specifications shall apply to the reactor at all times it is operating.ObjectiveThe objectives are to ensure that the reactor can be controlled and shut down at all times and thatthe safety limits will not be exceeded.Specifications1. The excess reactivity relative to the cold critical conditions, with or without experimentsin place shall not be greater than $3.50.2. The shutdown margin shall not be less than $0.50 with:a. The reactor in the reference core condition; andb. Total worth of all in-core experiments in their most reactive state; andc. Most reactive control rod fully withdrawn.3. Core configurations:a. The reactor shall only be operated with a standard core.b. No fuel shall be inserted or removed from the core unless the reactor is subcritical bymore than the worth of the most reactive fuel element.c. No control rods shall be removed from the core unless a minimum of four fuelbundles are removed from the core.d. The reactor shall be operated only with three operable control rods.4. No operation with damaged fuel (defined as a clad defect that results in fission productrelease into the reactor coolant) except to locate such fuel.5. The reactivity coefficients for the reactor are:Fuel: -1.2 ¢/°0CModerator: +3.0 ¢/°0CPower: -0.53 ¢/kWThe Fuel Temperature Coefficient, and the Moderator Temperature Coefficient, shall be verified any timethe standard core is modified either by the rotation of a fuel bundle, a change in fuel bundle or reflectorlocation, or the replacement of any fuel or reflector. The Power Temperature Coefficient shall berecalculated if either the Fuel Temperature or Moderator Temperature Coefficient are measured to be morethan 4-5% from the established values. Records of these tests shall be retained for a minimum of five years.6. The burnup ofU-235 in the UZrH fuel matrix shall not exceed 50 % of the initial concentration.8 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016Bases1. While specification 3.1.1, in conjunction with specification 3.1.2, tends to over constrainthe excess reactivity, it helps ensure that the operable core is similar to the core analyzedin the FSAR.2. The value of the shutdown margin as required by specification 3.1.2 assures that thereactor can be shutdown from any operating condition even if the highest worth controlrod should remain in the fully withdrawn position.3. Specification 3.1.3 ensures that the operable core is similar to the core analyzed in theFSAR. It also ensures that accidental criticality will not occur during fuel or control rodmanipulations.4. Specification 3.1.4 limits the fission product release that might accompany operation witha damaged fuel element. Fuel will be considered potentially "Damaged" if said fuel isfound to be leaking under the air and/or water sampling or under such case that the fuelhas been exposed to temperature above 175 °C as measured on the instrumented fuelelement. The criteria of the water and air sampling to determine a leaking fuel element isconsidered positive if either sample is found to contain I- 129 through I- 135, Xe- 135, Kr-85, 87 and Kr-88, Cs-135 and Cs-137, or Sr-89 through Sr-92.5. The reactivity coefficients in Specification 3.1.5 ensure that the net reactivity feedback isnegative.6. General Atomic tests of TRIGA fuel indicate that keeping fuel element burnup below 50% of the original 235U loading will avoid damage to the fuel from fission product buildup.3.2 REACTOR CONTROL AND SAFETY SYSTEMSApplicabilityThese specifications apply to reactor control and safety systems and safety-relatedinstrumentation that must be operable when the reactor is in operation.ObjectiveThe objective of these specifications is to specify the lowest acceptable level of performance orthe minimum number of operable components for the reactor control and safety systems.Specifications1. The drop time of each of the three standard control rods from the fully withdrawnposition to the fully inserted position shall not exceed one second.2. Maximum positive reactivity insertion rate by control rod motion shall not exceed $0.30per second.9 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. The reactor safety channels shall be operable in accordance with Table 3.1l, including theminimum number of channels and the indicated maximum or minimum set points for thescram channels.4. The safety interlocks shall be operable in accordance with Table 3.2, including theminimum number of interlocks.5. The Beam Port and Through Tube interlocks may be bypassed during a reactor operationwith the permission of the Reactor Director.6. A minimum of one reactor power channel, calibrated for reactor thermal power, must beattached to a recording device sufficient for auditing of reactor operation history.Bases1. Specification 3.2.1 assures that the reactor will be shutdown promptly when a scramsignal is initiated. Experiments and analysis have indicated that for the range oftransients anticipated for the MUTR TRIGA reactor, the specified control rod drop timeis adequate to assure the safety of the reactor.2. Specification 3.2.2 establishes a limit on the rate of change of power to ensure that thenormally available reactivity and insertion, rate cannot generate operating conditions thatexceed the Safety Limit. (See FSAR)3. Specification 3.2.3 provides protection against the reactor operating outside of the safetylimits. Table 3.3 describes the basis for each of the reactor safety channels.4. Specification 3.2.4 provides protection against the reactor operating outside of the safetylimits. Table 3.4 describes the basis for each of the reactor safety interlocks.5. Specification 3.2.5 ensures that reactor interlocks will always serve their intendedpurpose. This purpose is to assure that the operator is aware of the status of both thebeam ports and the through tube.6. Specification 3.2.6 provides for a means to monitor reactor operations and verify that thereactor is not operated outside of its license condition.10 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb16RevO2292016.docxLast edit February 29, 2016Table 3.1: Reactor Safety Channels: Scram ChannelsScram ChannelMinimum RequiredOperable2Scram SetpointReactor Power LevelNot to exceed 120 %Fuel Element TemperatureReactor Power ChannelDetector Power Supply12Manual ScramConsole Electrical SupplyRate of power change -PeriodRadiation Area Monitors1111Not to exceed 175 °CLoss of power supply voltage tochamberN/ALoss of electrical power to thecontrol consoleNot less than 5 seconds50 mr/hr (bridge monitor)10 mr/hr (exhaust monitor)Table 3.2: Reactor Safety Channels: InterlocksInterlock/ChannelLog Power LevelStartup Count rateSafety 1 Trip TestPlug ElectricalConnectionRod Drive ControlFunctionProvide signal to period rate and minimum sourcechannels. Prevent control rod withdrawal when neutroncount rate is less than l cps.Prevent control rod withdrawal when neutron count rate isless than 1 cps.Prevent control rod withdrawal when Safety 1 Trip Testswitch is activated.Disable magnet power when Beam Port or Through Tubeplug is removed unless bypass has been activated.Prevent simultaneous manual withdrawal of two or morecontrol rods in the steady state mode of operation.11 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb16RevO2292016.docxLast edit February 29, 2016Table 3.3: Reactor Safety Channels: Scram Channel BasesScram ChannelBasesReactor Power LevelFuel Element TemperatureReactor Power ChannelDetector Power SupplyManual ScramConsole Electric SupplyRate of power change -PeriodRadiation Area MonitorsProvides protection to assure that the reactor can beshut down before the safety limit on the fuelelement temperature will be exceeded.Provides protection to assure that the reactor cannotbe operated unless the neutron detectors that inputto each of the linear power channels are operable.Allows the operator to shut down the reactor if anunsafe or abnormal condition occurs.Assures that the reactor cannot be operated withouta secure electric supply.Assures that the reactor is operated in a manner thatallows the operator time to shut down the reactorbefore the licensed power restriction is exceeded.Assures that the reactor automatically scrams if ahigh airborne radiation level is detected.Table 3.4: Reactor Safety Channels: Interlock BasesInterlock/ChannelLog Power LevelStartup Count rateSafety 1 Trip TestPlug Electrical ConnectionRod Drive ControlBasesThis channel is required to provide a neutrondetector input signal to the startup count ratechannel.Assures sufficient amount of startup neutrons areavailable to achieve a controlled approach tocriticality.Assures that the 1 cps interlock cannot be bypassedby creating an artificial 1 cps signal with the Safety1 trip test switchAssures that the reactor cannot be operated withBeamport or Through Tube plugs removed withoutfurther precautions.Limits the maximum positive reactivity insertionrate available for steady state operation.12 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163.3 PRIMARY COOLANT SYSTEMApplicability_This specification applies to the quality and quantity of the primary coolant in contact with thefuel cladding at the time of reactor startup.Objectives1. To minimize the possibility for corrosion of the cladding on the fuel elements.2. To minimize neutron activation of dissolved materials.3. To ensure sufficient biological shielding during reactor operations.4. To maintain water clarity.Specifications1. A minimum of 15 ft. of coolant shall be above the core.2. Conductivity of the pool water shall be no higher than 5x10-6 mhos/cm and the pH shallbe between 5.0 and 7.5. Conductivity shall be measured before each reactor operation.pH shall be measured monthly, interval not to exceed six weeks.3. Gross gamma measurement shall be less than two times historical data measurements.Gross gamma activity shall be measured monthly, interval not to exceed six weeks.4. The pool water temperature shall not exceed 90 C, as measured by thermocouples locatedin the pool.Bases1. Specification 3.3.1 ensures that both sufficient cooling capability and sufficient biologicalshielding are available for safe reactor operation.2. A small rate of corrosion continuously occurs in a water-metal system. In order to limitthis rate, and thereby extend the longevity and integrity of the fuel cladding, a watercleanup system is required. Experience with water quality control at many reactorfacilities has shown that maintenance within the specified limit provides acceptablecontrol. In addition, by limiting the concentration of dissolved materials in the water, theradioactivity of neutron activation products is limited. This is consistent with theALARA principle, and tends to decrease the inventory of radionuclides in the entirecoolant system, which will decrease personnel exposures during maintenance andoperation.3. Specification 3.3.3 ensures that a fuel failure with release of radioactive materials into thepool will be determined.4. Specification 3.3.4 ensures a DNBR value greater than 2.13 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163.4 CONFINEMENT_ApplicabilityThis specification applies to that part of the facility that contains the reactor, its controls andshielding.ObjectiveThe objective of these specifications is to ensure that sufficient confinement volume is availablefor the dilution of radioactive releases and to limit the rate of release of radioactive material to theoutside environment._Specifications1. Confinement shall be considered established when the doors leading from the reactor bayarea leading into the balcony area on the top floor, and the reception area as well as thebuilding exterior are secured.2. Confinement shall be established whenever the reactor is in an unsecured mode with theexception of the time that persons are physically entering or leaving the confinementarea.Bases1. This specification provides the necessary requirements for confinement, which ensuresreleases to the outside environment are within 10 CFR Part 20 requirements.2. This specification provides the reactor status condition for confimement, as well as allowspersonnel to enter and leave the reactor building, as required, when the reactor isunsecured.3.5 VENTILATJON SYSTEMSApplicabilityThese specifications apply to the ventilation systems for the reactor building._ObjectiveThe objective of these specifications is to ensure that air exchanges between the reactorconfinement building and the environment do not impact negatively on the general public._Specifications1. Air within the reactor building shall not be exchanged with other occupied spaces in thebuilding.2. All locations where ventilation systems exchange air with the environment shall havefailsafe closure mechanisms.14 O:\M UTR\2O16M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. Forced air ventilation to the outside shall automatically secure without operatorintervention in such case that the radiation levels exceed a preset level as defined infacility procedures. The setpoints are: 50 mR/hr (bridge monitor), 10 mR/hr (exhaustmonitor).Bases1. This specification ensures that radioactive releases inside the reactor building will not betransported to the remainder of the building.2. This specification ensures that the reactor building can always be isolated from theenvironment.3. This specification ensures that radioactive release will be minimized by stopping forcedflow to the outside environment.3.6 _ RADIATION MONITORING SYSTEM ANT) EFFLUENTS3.6.1 Radiation Monitoring SystemApplicabilityThis specification applies to the radiation monitoring information that must be available to thereactor operator during reactor operation.ObjectiveThe objective is to assure that sufficient radiation monitoring information is available to theoperator to assure safe operation of the reactor.Specifications1. The reactor shall not be operated unless a minimum of one of the two radiation areamonitor channels listed in Table 3.5 are operable.2. For a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for maintenance or calibration to the radiationmonitor channels, the intent of specification 3.6.1 will be satisfied if they are replacedwith portable gamma sensitive instruments having their own alarms or which shall beobservable by the reactor operator.3. The alarm set points shall be stated in a facility operating procedure. The alarm setpointsfor the bridge monitor are: .37 mR/hr (alert), 50 mR/hr (scram). The setpoints for theexhaust monitor are: 8 mR/hr (alert), 10 mR/hr (scram).4. The campus radiation safety organization shall maintain an environmental monitor at theMUJTR site boundary.15 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016Table 3.5: Minimum Radiation Monitoring ChannelsRadiation Area MonitorsExhaust Radiation MonitorBridge Radiation MonitorFunctionMinimum Number OperableMonitor radiation levels in ReactorBay area at an Exhaust Fan locationMonitor radiation levels in ReactorBay area at the Reactor BridgelocationA minimum of 1 of the 2monitors shall be operableBases1. Specification 3.6.1.1 ensures that a significant fuel failure with release of radioactivematerials will be determined and that any large releases will be mitigated by the specifiedprotective actions.2. Specification 3.6.1.2 allows for continued reactor operation if maintenance and/orcalibration of the radiation area monitors is required.3. The alarm and scram set points shall be designed to ensure that dose rates delivered toareas accessible to members of the general public do not exceed the levels defined in 10CFR Part 20. Additionally, the radiation area monitors provide information to operatingpersonnel of any impending or existing danger from radiation so that there will besufficient time to evacuate the facility and take the necessary steps to prevent the spreadof radioactivity to the surroundings.4. The intent of Specifications 3.6.1.3 and 3.6.1.4 is to ensure that facility does not lead to adose to the general public greater than that allowed by 10 CFR Part 20.3.6.2 EffluentsApplicability.This specification applies to limits on effluent release.ObjectiveThe objective is to ensure that the release of radioactive materials from the reactor facility tounrestricted areas do not exceed federal regulations.SpecificationAll effluents from the MUTR shall conform to the standards set forth in 10 CFR Part 20.BasisThe intent of 3.6.2 is to ensure that, in the event that radioactive effluents are released, the dose tothe general public will be less than that allowed by 10 CFR Part 20.16 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163.7 LIMITATIONS ON EXPERIMENTSApplicabilityThe specification applies to experiments installed in the reactor and its experimental facilities.ObjectiveThe objective is to prevent damage to the reactor or excessive release of radioactive material inthe event of an experiment failure.SpecificationsThe reactor shall not be operated unless the following conditions governing experiments exist.1. The reactivity worth of any single experiment shall be less than $1.00.2. The total absolute reactivity worth of in-core experiments shall not exceed $3.00,including, the potential reactivity which might result from experimental malfunction andexperiment flooding or voiding.3. Experiments containing materials corrosive to reactor components, compounds highlyreactive with water, potentially explosive materials, and liquid fissionable materials shallbe doubly encapsulated.4. Explosive materials in quantities greater than 25 mg TNT or its equivalent shall not beirradiated in the reactor or experimental facilities. Explosive materials in quantities equalto or less than 25 mg may be irradiated provided the pressture produced upon detonationof the explosive has been calculated and/or experimentally demonstrated to be less thanthe failure pressure of the container. The failure pressure of the container is one half ofthe design pressure. Total explosive material inventory in the reactor facility may notexceed 100 mg TNT or its equivalent.5. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, orproduce aerosols under (1) normal operating conditions of the experiment or reactor, (2)credible accident conditions in the reactor or (3) possible accident conditions in theexperiment shall be limited in type and quantity such that if 100 % of the gaseous activityor radioactive aerosols produced escaped to the reactor room or the atmosphere, theairborne radioactivity in the reactor room or outside environment will not result inexceeding the applicable dose limits set forth in 10 CFR Part 20.In calculations pursuant to 3.7.5 above, the following assumptions shall be used:a. If the effluent from an experimental facility exhausts through a holdup tank, whichcloses automatically on high radiation level, at least 10 % of the gaseous activity oraerosols produced will escape.b. If the effluent from an experimental facility exhausts through a filter installationdesigned for greater than 99 % efficiency for 0.3 particles, at least 10 % of theseparticles can escape.17 O:\M UTR\2016M UTRLivingoocs\WorkingCopyTechnical Specifications 29 feb 16Rev02292016.docxLast edit February 29, 2016c. If an experiment fails and releases radioactive gases or aerosols to the reactor bay oratmosphere, 100 per cent of the radioactive gases or aerosols escape.d. If an experiment fails that contains materials with a boiling point above 1300 F (540C), the vapors of at least 10 percent of the materials escape through an undisturbedcolumn of water above the core.6. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes131 through 135 in the experiment is no greater than 5 mCi.Bases1. This specification is intended to provide assurance that the worth of a single unsecuredexperiment will be limited to a value such that the safety limit will not be exceeded if thepositive worth of the experiment were to be inserted suddenly.2. The maximum worth of a single experiment is limited so that its removal from the coldcritical reactor will not result in the reactor achieving a power level high enough toexceed the core temperature safety limit. Since experiments of such worth must befastened in place, its inadvertent removal from the reactor operating at full power wouldresult in a relatively slow power increase such that the reactor protective systems wouldact to prevent high power levels from being attained.The maximum worth of all experiments is also limited to a reactivity value such that thecold reactor will not achieve a power level high enough to exceed the core temperaturesafety limit if the experiments were removed inadvertently.3. This specification is intended to prevent damage to reactor components resulting fromexperiment failure. If an experiment fails, inspection of reactor structures andcomponents shall be performed in order to verifyr that the failure did not cause damage.If damage is found, appropriate corrective actions shall be taken.4. This specification is intended to prevent damage to reactor components resulting fromfailure of an experiment involving explosive materials, especially the accidentaldetonation of the explosive. If an experiment fails, inspection of reactor structures andcomponents shall be performed in order to verify' that the failure did not cause damage.If damage is found, appropriate corrective actions shall be taken.5. This specification is intended to reduce the likelihood that airborne activities in excess ofthe limits of Table 2 of Appendix B of 10 CFR Part 20 will be released to the atmosphereoutside the facility boundary.6. The 5 mCi limitation on iodine 131 through 135 assures that in the event of failure of afueled experiment leading to total release of the iodine, the exposure dose at theexclusion area boundary will be less than that allowed by 10 CFR Part 20 for anunrestricted area. (See SAR)18 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20164.0 SURVEILLANCE REQUIREMENTSINTRODUCTIONSurveillances shall be performed on a timely basis as defined in the individual procedures governing theperformance of the surveillance. In the event that the reactor is not in an operable condition, such asduring periods of refueling, or replacement or repair of safety equipment, surveillances may be postponeduntil such time that the reactor is operable. In such case that any surveillance must be postponed, awritten directive signed by the Facility Director, shall be placed in the records indicating the reason whyand the expected completion date of the required surveillance. This directive shall be written before thedate that the surveillance is due. Under no circumstance shall the reactor perform routine operations untilsuch time that all surveillances are current and up to date. Any system or component that is modified,replaced, or had maintenance performed will undergo testing to ensure that the system/componentcontinues to meet performance requirements.4.1 REACTOR CORE PARAMETERSApplicabilityThese specifications apply to the surveillance requirements for the reactor core.O~bjectiveThe objective of these specifications is to ensure that the specifications of Section 3.1 aresatisfied.Specifications1. The excess reactivity shall be determined annually, at intervals not to exceed 15 months,and after each time the core fuel configuration is changed, these changes include anyremoval or replacement of control rods.2. The shutdown margin shall be determined annually, at intervals not to exceed 15 months,and after each time the core fuel configuration is changed, these changes include anyremoval or replacement of control rods3. Core configuration shall be verified prior to the first startup of the day.4. Gross gamma measurements shall be taken monthly, at intervals not to exceed six weeks.5. Twenty percent of the fuel elements shall be visually inspected annually, not toexceed 15 months, such that the entire core is inspected over a five year period.6. Burnup shall be verified in the Annual Report.BasesExperience has shown that the identified frequencies ensure performance and operability for eachof these systems or components. For excess reactivity and shutdown margin, long-term changesare slow to develop. For fuel inspection, visually inspecting 20% of the bundles annually willidentify any developing fuel integrity issues throughout the core.19 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20164.2 REACTOR CONTROL AND SAFETY SYSTEMSApp~licabilityThese specifications apply to the surveillance requirements for reactor control and safety systems.ObjectiveThe objective of these specifications is to ensure that the specifications of Section 3.2 aresatisfied.Specifications1. The reactivity worth of each standard control rod shall be determined annually, intervalsnot to exceed 15 months, and after each time the core fuel configuration is changed or acontrol rod is changed.2. The control rod withdrawal and insertion speeds shall be determined annually, intervalsnot to exceed 15 months, or whenever maintenance or repairs are made that could affectrod travel times.3. Control rod drop times shall be measured annually; intervals not to exceed 15 months, orwhenever maintenance or repairs are made that could affect their drop time.4. All scram channels and power measuring channels shall have a channel test, includingtrip actions with safety rod release and specified interlocks performed after each securedshutdown, before the first operation of the day, or prior to any operation scheduled to lastmore than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or quarterly, with intervals not to exceed 4 months. Scram channelsshall be calibrated annually, intervals not to exceed 15 months.5. Operability tests shall be performed on all affected safety and control systems after anymaintenance is performed.6. A channel calibration shall be made of the linear power level monitoring channelsannually, intervals not to exceed 15 months.7. A visual inspection of the control rod poison sections shall be made biennially, intervalsnot to exceed 28 months.8. A visual inspection of the control rod drive and scram mechanisms shall be madeannually, intervals not to exceed 15 months.Bases1. The reactivity worth of the control rods, specification 4.2.1, is measured to assure that therequired shutdown margin is available and to provide a means to measure the reactivityworth of experiments. Long term effects of TRIGA reactor operation are such thatmeasurements of the reactivity worths on an annual basis are adequate to insure that nosignificant changes in shutdown margin have occurred.20 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20162. The control rod withdrawal and insertion rates, specification 4.2.2, are measured to insurethat the limits on maximum reactivity insertion rates are not exceeded.3. Measurement of the control rod drop time, specification 4.2.3, ensures that the rods canperform their safety function properly.4. The surveillance requirement specified in specification 4.2.4 for the reactor safety scramchannels ensures that the overall functional capability is maintained.5. The surveillance test performed after maintenance or repairs to the reactor safety systemas required by specification 4.2.5 ensures that the affected channel will perform asintended.6. The linear power level channel calibration specified in specification 4.2.6 will assure thatthe reactor will be operated at the licensed power levels.7. Specification 4.2.7 assures that a visual inspection of control rod poison sections is madeto evaluate corrosion and wear characteristics and any damage caused by operation in thereactor.8. Specification 4.2.8 assures that a visual inspection of control drive mechanisms is madeto evaluate corrosion and wear characteristics and any damage caused by operation in thereactor.4.3 PRIMARY COOLANT SYSTEMApplicabilityThese specifications apply to the surveillance requirements of the reactor primary coolant system.ObjectiveThe objective of these specifications is to ensure the operability of the reactor primary coolantsystem as described in Section 3.3.Specifications1. The primary coolant level shall be verified before each reactor startup or daily duringoperations exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.2. Pool water conductivity shall be determined prior to the first startup of the day, and poolwater pH shall be determined monthly at intervals not to exceed six weeks.3. Pool water gross gamma activity shall be determined monthly, at intervals not to exceedsix weeks. If gross gamma activity is high (greater than twice historical data), gammaspectroscopy shall be performed.4. Pool water temperature shall be measured prior to the reactor startup and shall bemonitored during reactor operation.21 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016Bases1. Specification 4.3.1 ensures that sufficient water exists above the core to provide bothsufficient cooling capacity and an adequate biological shield.2. Specification 4.3.2 ensures that poor pool water quality could not exist for long withoutbeing detected. Years of experience at the MIUTR have shown that pool water analysison a monthly basis is adequate to detect degraded conditions of the pooi water in a timelymanner.3. Gross gamma activity measurements are conducted to detect fission product releasesfrom damaged fuel element cladding.4. Specification 4.3.4 ensures that the maximum allowable pool water temperature is notexceeded.4.4 CONFINEMENTApplicabilityThis specification applies to that part of the facility which contains the reactor, its controls andshielding.ObjectiveThe objective of this specification is to ensure that radioactive releases from the confinementcan be limited._SpecificationPrior to putting the reactor in an unsecured mode, the isolation of the confinement building shallbe visually verified.BasesThis specification ensures that the minimal leakage rate assumed in the SAR is actually presentduring reactor operations in order to limit the release of radioactive material to the environs.4.5 VENTTLATION SYSTEMApplicabilityThis specification applies to the reactor ventilation system._ObjectiveThe objective is to assure that provisions are made to restrict the amount of radioactivity releasedto the environment.22 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016SpecificationThe ability to secure the ventilation system shall be verified before the first reactor operation of theday.BasesThe facility is designed such that in the event that excessive airborne radioactivity is detected theventilation system shall be shutdown to minimize transport of airborne materials. Analysisindicates that in the event of a major fuel element failure personnel would have sufficient time toevacuate the facility before the maximum permissible dose (10 CFR Part 20) is exceeded.4.6 RADIATION MONITORING SYSTEM ANT) EFFLUENTS4.6.1 Radiation Monitoring SystemApplicabilityThis specification applies to the surveillance requirements for the Radiation Area MonitoringSystem (RAMS).ObjectiveThe objective of these specifications is to ensure the operability of each radiation area monitoringchannel as required by Section 3.6 and to ensure that releases to the environment are kept belowallowable limits.Specifications1. A channel calibration shall be made for each channel listed in Table 3.5 annually but atintervals not to exceed 15 months or whenever maintenance or repairs are made thatcould affect their calibration.2. A channel test shall be made for each channel listed in Table 3.5 prior to starting up thereactor to ensure reactor scram, fan shutdown, and louver closing.BasesSpecifications 4.6.1.1 and 4.6.1.2 ensure that the various radiation area monitors are checked andcalibrated on a routine basis, in order to assure compliance with 10 CFR Part 20.4.6.2 EffluentsApplicabilityThis specification applies to the surveillance requirements for air and water effluents.23 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016ObjectiveThe objective of these specifications is to that releases to the environment are kept belowallowable limits.Specifications1. Reactor building air samples shall be counted for gross gamma activity monthly, intervalsnot to exceed 6 weeks.2. A sample of any water discharged from the reactor building sump shall be counted forgross gamma activity before its release to the environs.BasesSpecifications 4.6.2.1 and 4.6.2.2 ensure that the facility effluents comply with 10 CFR Part 20.4.7 EXPERIMENTSApplicabilityThis specification applies to the surveillance requirements for experiments installed in the reactorand its irradiation facilities.ObjectiveThe objective of this specification is to prevent the conduct of experiments which may damagethe reactor or release excessive amounts of radioactive materials as a result of experiment failure.Specifications1. The reactivity worth of an experiment shall be estimated or measured, as appropriate, beforereactor operation with said experiment2. An experiment shall not be installed in the reactor or its irradiation facilities unless a safetyanalysis has been performed and reviewed for compliance with Section 3.7 by the ReactorSafety Committee (new experiment) or Facility Director (modified routine experiment), infull accord with Sections 6.1.2 and 6.2.1 of these Technical Specifications and the procedureswhich are established for this purpose.BasisExperience has shown that experiments reviewed and approved by the Reactor Safety Committee orFacility Director can be conducted without endangering the safety of the reactor, personnel, orexceeding Technical Specification limits.24 O:\M UTR\2Oi6M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20165.0 DESIGN FEATURES5.1 SITE CHARACTERISTICSApplicabilityThis specification applies to the reactor facility and its site boundary.ObjectiveThe objective is to assure that appropriate physical security is maintained for the reactor facilityand the radioactive materials contained within it.Specifications1. The reactor shall be housed in a closed room designed to restrict leakage. The closedroom does not include the West balcony area.2. The reactor site boundary shall consist of the outer walls of the reactor building and thearea enclosed by the loading dock fence.3. The restricted area shall consist of all areas interior to the reactor building including thewest balcony and lower entryway.4. The controlled area shall consist of all areas interior to the reactor building includingthe west balcony and lower entryway.BasesThese specifications assure that appropriate control is maintained over access to the facility bymembers of the general public.5.2 REACTOR PRIMARY COOLANT SYSTEMApplicabilityThis specification applies to the pool containing the reactor and to the cooling of the core by thepool water.ObjectiveThe objective is to assure that coolant water shall be available to provide adequate cooling of thereactor core and adequate radiation shielding.Specifications1. The reactor core shall be cooled by natural convective water flow.2. The pool water inlet pipe is equipped with a siphon break at the surface of the pool.25 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. The pooi water return (outlet) pipe shall not extend more than 50.8 cm (20 in) below theoverflow outlet pipe when fuel is in the core.Bases1. Specification 5.2.1 is based on thermal and hydraulic calculations and operation of the MUTRthat show that the core can operate in a safe manner at power levels up to 300 kW withnatural convection flow of the coolant.2. Specifications 5.2.2 and 5.2.3 ensures that the pool water level can normally decrease only by50.8 cm (20 in) if the coolant piping were to rupture and siphon water from the reactor tank.Thus, the core will be covered by at least 4.57 m (15 ft.) of water.5.3 REACTOR CORE AND FUELApplicabilityThis specification applies to the configuration of the core and in-core experiments.ObjectiveThe objective is to ensure that the core configuration is as specified in the license.Specifications1. The core shall consist of 93 TRIGA fuel elements assembled into 24 fuel bundles -21bundles shall contain four fuel elements and 3 bundles shall contain three fuel elements and acontrol rod guide tube.2. The fuel bundles shall be arranged in a rectangular 4 x 6 configuration, with one bundledisplaced for the in-core pneumatic experimental system.3. The reactor shall not be operated at power levels exceeding 250 kW.4. The reflector shall be a combination of two graphite reflector elements and waterBasis1. Only TRIGA fuel elements shall be used in the fuel bundles.2. The experimental system allows insertion of small samples directly into the reactor core.3. The maximum power level presents a conservative limitation with respect to the safety limitsfor the maximum temperature in the fuel.4. The reflector reduces the neutron leakage from the reactor core.26 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20165.3.1 Reactor FuelApplicabilityThis specification applies to the fuel elements used in the reactor core.ObjectiveThe objective is to assure that the fuel elements are of such design and fabricated in such a manner asto permit their use with a high degree of reliability with respect to their physical and nuclearcharacteristics, and that the fuel used in the reactor has characteristics consistent with the fuelassumed in the SAR and the license.SpecificationsThe individual unirradiated standard TRIGA fuel elements shall have the followingcharacteristics:1. Uranium content: a maximum of 9.0 weight % uranium enriched to less than 20 % 235U2. Zirconium hydride atom ratio: nominal 1.5 -1.8 hydrogen-to-zirconium, ZrHx3. Cladding: 304 stainless steel, nominal thickness of 0.508 mm (.020 in)4. The overall length of a fuel element shall be 30 inches, and the fueled length shall be 15inches.BasisThe design basis of the standard TRIGA fuel element demonstrates that 250 kW steady stateoperation presents a conservative limitation with respect to safety limits for the maximumtemperature generated in the fuel.5.3.2 Control RodsApplicabilityThis specification applies to the control rods used in the reactor core.ObjectiveThe objective is to assure that the control rods are of such a design as to permit their use with a highdegree of reliability with respect to their physical and nuclear characteristics.Specifications1. The three control rods shall have scram capability, shall be used for reactivity control, andshall contain borated graphite, B4C, in powder form.2. The control rod cladding shall be aluminum with nominal thickness 0.71 mm (0.028") andlength 17".27 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTech nicai Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016BasisThe poison requirements for the control rods are satisfied by using neutron absorbing boratedgraphite, B4C, powder. These materials must be contained in a suitable clad material such asaluminum to ensure mechanical stability during movement and to isolate the poison from the tankwater environments. Scram capabilities are provided for rapid insertion of the control rods, which isthe primary safety feature of the reactor.5.4 FISSIONABLE MATERIAL STORAGEApplicabilityThis specification applies to the storage of reactor fuel at times when it is not in the reactor core.ObjectiveThe objective is to assure that fuel that is being stored will not become critical and will not reachan unsafe temperature.Specifications1. All fuel elements shall be stored either in a geometrical array where the k-effective is less than 0.8for all conditions of moderation and reflection or stored in an approved fuel shipping container.2. Irradiated fuel elements and fueled devices shall be stored in an array which will permitsufficient natural convection cooling by water or air such that the fuel element or fueled devicetemperature will not exceed design values.3. When fuel is in storage in any area other than the grid plate, that area must be equipped withmonitoring devices that both measure and record the radiation levels and temperature of theregion surrounding the fuel.BasisThe limits imposed by Specifications 5.4.1 and 5.4.2 are conservative and assure safe storage.28 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20166.0 ADMINISTRATION6.1 ORGANIZATIONThe Maryland University Training Reactor (MUTR) is owned and operated by the University ofMaryland, College Park. Its position in the university's structure is shown in Figure 6.1The university shall provide whatever resources are required to maintain the facility in acondition that poses no hazard to the general public or to the environment.6.1.1 StructureFigure 6.2 shows the MUTR organizational structure.6.1.2 ResponsibilityThe Dean College of Engineering is responsible for the oversight and operation of the school ofengineering.The Chair of the Department of Materials Science and Engineering is responsible for the oversightand operation of the Department of Materials Science and Engineering.The Director of MUTR: Responsibility for the safe operation of the reactor facility andradiological safety shall rest with the Facility Director. The members of the organization chartshown in Figure 6.2 shall be responsible for safeguarding the public and facility personnel fromundue radiation exposure and for adhering to all requirements of the operating license.The Senior Reactor Operators (SRO) are individuals who are licensed by the NRC to direct theactivities of reactor operators.The Reactor Operators are individuals who are licensed by the U.S. Nuclear RegulatoryCommission (NRC) to manipulate the controls of the reactor.6.1.3 Facility Staff Requirements1. The minimum staffing while the reactor is not secured shall be:a. A licensed reactor operator (RO) or a licensed senior reactor operator (SRO) shall bepresent in the control room.b. A minimum of two persons shall be present in the facility or in the Chemical andNuclear Engineering Building while the reactor is not secured: the operator in thecontrol room and a second person who can be reached from the control room who isable to carry out prescribed written instructions which may involve activatingelements of the Emergency Plan, including evacuation and initial notificationprocedures.c. A licensed SRO shall be present or readily available on call. "Readily Available onCall" means an individual who (1) has been specifically designated and thedesignation known to the operator on duty, (2) keeps the operator on duty informedof where he/she may be rapidly contacted and the method of contact, and (3) is29 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016capable of arriving at the reactor facility within a reasonable amount of time undernormal conditions. At no time while the reactor is not secured shall the designatedSRO be more than thirty minutes or ten miles from the facility.2. A list of reactor facility personnel by name and telephone number shall be readily available inthe control room for use by the operator. The list shall include:a. Management personnelb. Radiation safety personnelc. Licensed operators30 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016PresidentUniversity of MarylandProvost & SeniorVice PresidentDeanClark School ofEngineeringVice PresidentAdministrative AffairsDirectorDepartment ofEnvironmental SafetyRadiationSafety .Committee %%ChairDepartment ofMaterials Science andEngineeringRadiation SafetyOfficer4LDirectorNuclear ReactorFacilityReviews* * --AuditReactor .iembr-Safety ."MmeCommitteeRadiation Safety OfficeStaffServicesiReactor OperationsStaff-- Normal Administrative Reporting Channel-...Communication LinesFigure 6.1: MUTR Position in University of Maryland Structure31 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 2016DeanClark School of Engineering(Level 1)ChairDepartment of MaterialsScience and Engineering(Level 1)DirectorNuclear Reactor Facility(Level 2),I!Senior Reactor Operator(Level 3)Reactor Operator(Level 4)Reactor SafetyCommittee4----------- Normal Administrative Reporting Channel-------------------Communication LinesFigure 6.2: MUTR Organizational Structure32 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTech nicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. The following operations shall be supervised by a senior reactor operator:a. Initial startup and approach to power following new fuel loading or fuelrearrangementb. When experiments are being manipulated in the core that have an estimated worthgreater than $0.80c. Removal of control rods or fuel manipulations in the cored. Resumption of operation following an unplanned or unscheduled shutdown or anyunplanned or unexpected significant reduction in power.6.1.4 Selection and Training of PersonnelThe selection and training of operations personnel should be in accordance with the following:1. Responsibility -The Facility Director or his designated alternate is responsible for thetraining and requalification of the facility reactor operators and senior reactor operators.This selection shall be in conjunction with the guidelines set forth in ANSJIANS 15.1 and15.4.6.2 REVIEW AND AUDIT6.2.1 Reactor Safety CommitteeA Reactor Safety Committee (RSC) shall exist for the purpose of reviewing matters relating to thehealth and safety of the public and facility staff and the safe operation of the facility. It isappointed by and reports to the Chairperson of the Department of Materials Science andEngineering. The RSC shall consist of a minimum of five persons with expertise in the physicalsciences and preferably some nuclear experience. Permanent members of the committee are theFacility Director and the Campus Radiation Safety Officer or that office's designated alternate,neither may serve as the committee's chairperson. Qualified alternates may serve on thecommittee. Alternates may be appointed by the Chairperson of the RSC to serve on a temporarybasis. At least one committee member must be from outside the Department of Materials Scienceand Engineering.6.2.1.1l Reactor Safety Committee Charter And Rules1. The RSC shall meet at least twice per year, and more often as required.2. A quorum of the RSC shall be not less than half of the committee members, one of whomshall be the Campus Radiation Safety Officer (or designated alternate). No more than twoalternates shall be used to make a quorum. MUTR staff members shall not constitute themajority of a voting quorum.3. Minutes of all meetings will be retained in a file and distributed to all RSC members.33 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20166.2.1.2 Reactor Safety Committee Review FunctionThe RSC shall review the following:1. Determinations that proposed changes in equipment, systems, test, experiments, orprocedures are allowed without prior authorization by the responsible authority, e.g. 10 CFR50.59;2. All new procedures and major revisions thereto having safety significance, proposed changesin reactor facility equipment, or systems having safety significance;3. All new experiments or classes of experiments that could affect reactivity or result in therelease of radioactivity;4. Proposed changes in technical specifications, or license;5. Violations of technical specifications or license. Violations of internal procedures orinstructions having safety significance;6. Operating abnormalities having safety significance;7. Reportable occurrences listed in Section 6.7.2;8. Audit reports.A written report of the findings and recommendations of the RSC shall be submitted to Level 1management, the Facility Director, and the RSC members in a timely manner after the review hasbeen completed.6.2.1.3 Reactor Safety Committee Audit Function1. An annual audit and review of the reactor operations will be performed by an outsideindividual or group familiar with research reactor operations. They shall submit a report tothe Facility Director and the Reactor Safety Committee.2. The following shall be reviewed:a. Reactor operators and operational records for compliance with internal rules,procedures, and regulations, and with license provisions;b. Existing operating procedures for adequacy and accuracy;c. Plant equipment performance and its surveillance requirements;d. Records of releases of radioactive effluents to the environment;e. Operator training and requalification;f. Results of actions taken to correct those deficiencies that may occur in the reactorfacility equipment, systems, structures, or methods of operation that affect reactorsafety; and34 O :\M UTR\2016M UTRLivingDocs\WorlingCopyTech nicalI Specifications 29 feb 16RevO22920l6.docxLast edit February 29, 2016g. Reactor facility emergency plan and implementing procedures.Deficiencies uncovered that affect reactor safety shall immediately be reported to Level 1management and the Facility Director. A written report of the findings of the audit shall besubmitted to Level 1 management, the Facility Director, and the RSC members within 3 monthsafter the audit has been completed.6.2.2 Audit of ALARA ProgramThe Facility Director or his designated alternate shall conduct an audit of the reactor facilityALARA Program at least once per calendar year (not to exceed fifteen months). The results ofthe audit shall be presented to the RSC at the next scheduled meeting. This audit may occur aspart of a review of the overall campus ALARA program.6.3 RADIATION SAFETYA radiation safety program following the requirements established in 10 CFR Part 20 will beundertaken by the Radiation Safety Office. The facility director will ensure that ALARAprinciples are followed during all facility activities.6.4 OPERATING PROCEDURESWritten procedures, reviewed and approved by the Reactor Safety Committee, shall be in effectand followed for the following items prior to performance of the activity. The procedures shallbe adequate to assure the safety of the reactor, but should not preclude the use of independentjudgment and action should the situation require such.1. Start-up, operation, and shutdown of the reactor2. Installation or removal of fuel elements, control rods, experiments, and experimentalfacilities3. Maintenance procedures that could have an effect on reactor safety4. Periodic surveillance checks, calibrations, and inspections required by the TechnicalSpecifications or those that may have an effect on reactor safety5. Administrative controls for operations and maintenance and for the conduct ofirradiations and experiments that could affect reactor safety or core reactivity6. For any activity pertaining to shipping, possession, and transfer of radioactivematerial, these procedures shall be written in conjunction with the Radiation SafetyOffice and the Radiation Safety Officer who shall inform the Reactor Director of anychanges in regulations or laws that may require modification of these procedures. Allshipping and receiving of radioactive material shall be performed in conjunctionw i t h , and with the approval of the Radiation Safety Office.7. Implementation, maintenance, and modification to the Emergency Plan8. Implementation, maintenance, and modification to the Security Plan35 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20169. Implementation, maintenance, and modification to the Radiation Protection Plan.The Radiation Protection Plan shall include an ALARA plan as defined inANSI/ANS- 15.l1110. Use, receipt, and transfer of byproduct materialSubstantive changes to the above procedures shall be made with the approval of the FacilityDirector and the Reactor Safety Committee and shall be made in accordance with 10 CFR 50.59.This approval shall be granted before the changes may be considered in effect. The onlyexception to this clause is in such a case where the delay in implementation would cause acredible risk to the public or the facility. If such a case exists as determined by the FacilityDirector, temporary approval may be granted by the Director but must be approved by theReactor Safety Committee within thirty days. Temporary or minor changes to procedures shallbe documented and subsequently reviewed by the Reactor Safety Committee at the nextscheduled meeting. The Reactor Director shall have the power to approve minor changes such asphone number changes, typographical error correction or any other change that does not changethe effectiveness or the intent of the procedure. It shall be considered sufficient approval anddocumentation when the Director forwards by electronic means to both the Radiation SafetyOfficer and the Chair of the Reactor Safety Committee. A copy of the transmission shall be filedwith the appropriate procedure.6.5 EXPERIMENT REVTEW AND APPROVAL1. Routine experiments may be performed at the discretion of the duty senior reactor operatorwithout the necessity of further review or approval.2. Modified routine experiments shall be reviewed and approved in writing by the FacilityDirector, or designated alternate.3. Special experiments shall be reviewed by the RSC and approved by the RSC and the FacilityDirector or designated alternate prior to initiation.4. The review of an experiment listed in subsections 6.5.2 and 6.5.3 above, shall consider itseffect on reactor operation and the possibility and consequences of its failure, including,where significant, chemical reactions, physical integrity, design life, proper cooling,interaction with core components, and any reactivity effects.6.6 REQUIRED ACTIONS6.6.1 Actions To Be Taken In Case Of Safety Limit ViolationIn the event a safety limit is exceeded:1. The reactor shall be shut down and reactor operation shall not be resumed untilauthorized by the NRC.2. The event shall be reported to the Reactor Director who will report to the NRC asrequired in section 6.7.2.36 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. An immediate report of the occurrence shall be made to the Chairman, Reactor SafetyCommittee, and reports shall be made to the NRC in accordance with Section 6.7.2 ofthese specifications, and4. A report, and any follow-up report, shall be prepared. The report shall describe thefollowing:a. Applicable circumstances leading to the violation, including when known, thecause, and contributing factors;b. Effect of the violation upon reactor facility components, systems, or structuresand on the health and safety of personnel and the public; andc. Corrective action to be taken to prevent recurrence.The report shall be reviewed by the Reactor Safety Committee and submitted to the NRCwhen authorization is sought to resume operation of the reactor.6.6.2 Actions to Be Taken In The Event Of a Reportable OccurrenceIn the event of a reportable occurrence, as defined in section 1.32 of these TechnicalSpecifications, the following actions will be taken:1. Immediate action shall be taken to correct the situation and to mitigate the consequencesof the occurrence.2. The reactor shall be shut down and reactor operation shall not be resumed untilauthorized by the Facility Director.3. The event shall be reported to the Facility Director who will report to the NRC asrequired in section 6.7.2.4. The Reactor Safety Committee shall investigate the causes of the occurrence at its nextmeeting. The Reactor Safety Committee shall report its findings to the NRC and Dean,School of Engineering. The report shall include an analysis of the causes of theoccurrence, the effectiveness of corrective actions taken, and recommendations ofmeasures to prevent or reduce the probability or consequences of recurrence.6.7 REPORTS6.7.1 Annual Operating ReportA report summarizing facility operations shall be prepared annually for the reporting periodending June 30. This report shall be submitted by December 30 of each year to the NRCDocument Control Desk. The report shall include the following:1. A brief narrative summary of results of reactor operations and surveillance tests andinspections required in section 4.0 of these Technical Specifications2. A tabulation showing the energy generated in MW hr-' for the year37 O :\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20163. A list of unscheduled shutdowns including the reasons therefore and corrective actiontaken, if any4. A tabulation of the major maintenance operations performed during the period, includingthe effects, if any, on safe operation of the reactor, and the reason for any correctivemaintenance required5. A brief description ofa. Each change to the facility to the extent that it changes a description of thefacility in the Final Safety Analysis Reportb. Review of changes, tests, and experiments made pursuant to 10 CFR Part 50.59.6. A summary of the nature and amount of radioactive effluents released or discharged to theenvironment7. A description of any environmental surveys performed outside of the facility8. A summary of exposure received by facility personnel and visitors where such exposuresare greater than 25 percent of limits allowed by 10 CFR Part 209. Changes in facility organization6.7.2 Special ReportsNotification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone to the NRC Operations Center, followedby a written report faxed within 14 days in the event of the following:1. A reportable occurrence, as defined in Section 1.32 of this document2. Release of radioactivity from the site above allowed limits3. Exceeding the Safety LimitThe written report shall be sent to the NRC document control desk. The written report and, to theextent possible, the preliminary telephone or facsimile notification shall:1. Describe, analyze, and evaluate safety implications2. Outline the measures taken to ensure that the cause of the condition is determine3. Indicate the corrective action taken to prevent repetition of the occurrence includingchances to procedures4. Evaluate the safety implications of the incident in light of the cumulative experienceobtained from the report of previous failure and malfunction of similar systems andcomponents38 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docxLast edit February 29, 20166.7.3 Unusual Event ReportA written report shall be forwarded within 3 0 days to the NRC Document Control Desk, with acopy to the Regional Administrator, Region I, NRC, in the event of:1. Discovery of any substantial errors in the transient or accident analysis or in the methodsused for such analysis as described in the Safety Analysis Report or in the bases for theTechnical Specifications2. Discovery of any substantial variance from performance specifications contained in theTechnical Specifications or Safety Analysis Report3. Discovery of any condition involving a possible single failure which, for a systemdesigned against assumed failure, could result in a loss of the capability of the system toperform its safety function4. A permanent change in the position of Department Chair or Facility Director6.8 RECORDS1. The following records shall be retained for a period of at least five years:a. Normal reactor facility operation and maintenanceb. Reportable occurrencesc. Surveillance activities required by Technical Specificationsd. Facility radiation and contamination surveyse. Experiments performed with the reactorf. Reactor fuel inventories, receipts, and shipmentsg. Approved changes in procedures required by these Technical Specificationsh. Minutes of the Reactor Safety Committee meetingsi. Results of External Audits2. Retraining and requalification records of current licensed operators shall be maintained at alltimes that an operator is employed or until the operator's license is renewed.3. The following records shall be retained for the lifetime of the facility:a. Liquid radioactive effluents released to the environsb. Gaseous radioactive effluents released to the environsc. Radiation exposure for all facility personnel39 O:\M UTR\2016M UTRLivingDocs\WorkingCopyTechnical Specifications 29 feb16RevO2292016.docxLast edit February 29, 2016d. Radiation exposures monitored at site boundarye. As-built facility drawingf. Violation of the Safety Limitg. Violation of any Limited Safety System Setting (LSSS)h. Violation of any Limiting Condition of Operation (LCO)4. Requirement 6.8.1 (a) above does not include supporting documents such as checklists,logsheets and recorder charts, which shall be maintained for a period of at least one year.5. Applicable annual reports, if they contain any of the required information may be used asrecords in subsection 6.8.3 above.40 Accident Analysis MHAThe NRC licenses research and test reactors consistent with the NRC mission to ensure adequate protectionof the public health and safety and to promote and protect the environment. NUREG 1537 Part 2 Chapter 13Accident Analysis provides guidance and acceptable format and content for licensees to present regarding aMaximum Hypothetical Accident, MHA.Utilizing guidance from both NUJREG 1537 and NUJREG/CR-23 87, Credible Accident Analyses forTRIGA and TRIGA-Fueled Reactors, the bounding and limiting credible accident scenario is a fuel element failurewhich can occur at any time during normal operations or when the reactor is at rest and shutdown.In this worst case scenario a single element has been removed from the reactor and dropped to the floor ofthe reactor building outside of the biological shield. Fission products are released in air from the gap and thecladding and instantaneously and uniformly mix in the volume of the reactor building. The reactor facilityexhaust fans are not running and are closed during this event. No immediate protective functions are activated byradiation detectors or personnel present at the start of the event.In general, the escape of fission products from fuel or fueled experiments and their release to theunrestricted environment would be the most hazardous radiological accident conceivable at a non-power reactor.However, non-power reactors are designed and operated so that a fission product release is not credible for most.Therefore, this release under accident conditions can reasonably be selected as the MHADoses are calculated for air leakage out of the south side of the reactor. Internal, external, and shine dosesare determined for members of the public. Internal and external doses are determined for reactor personnel.Engineering analysis at the Maryland University Training Reactor (MUTR) has shown that 90% of the timeair leakage occurs out of the reactor on the north side entrances, predominantly through the roll up door andsignificantly less through a single entrance door. The remaining 10% of the time analysis has shown air leakagesites are located, on the eastern, southern and western sides of the reactor building, the most prominent site beingon the southern side and contributing 38% of the total air leakage to a single location. A highly conservativeapproach assumes the MHA's airborne radioactivity escapes continuously during the event from these predominantlocations from the onset of the event until the end of the leakage time. This is the source of the Maximally ExposedIndividual (MET) member of the public outside of the confinement space, and at locations downwind of theMUTR. Occupational personnel located within the MUTR are exposed to internal and external radiation from therelease during the time it takes to evacuate the reactor.On the south side of the reactor, 10% of the time, a steady nonstop air leakage rate from the reactor spacewill empty the air in 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> at a rate of 6.25E-3 m3 s-1. To maintain an overly conservative assumption, membersof the public remain in place for the duration of the event, and doses are determined over this time interval. Inreality individuals would not remain in these locations directly outside of confinement for the duration of an MHA.Reactor staff, Radiation Safety Office support personnel, and local response agency personnel would worktogether to secure and maintain acceptable perimeters relative to exposure and dose measurements surrounding theMULTR facility.Per 10 CFR 20.1301 the NRC regulations limit the internal and external dose to 100 mrem for members ofthe public; this includes doses incurred during an incident such as the MHA.1 Accident Analysis MHAInternal and external doses are accrued by occupational personnel present in the reactor at the start of theevent until they have evacuated the confinement space. Evacuation time is overly conservative and set at 5minutes.NUREG 2387 provides guidance and analysis for a 1MW TRIGA after 1 year of continuous operation atfull power, or 365MWd. The Maryland University Training Reactor (MUTR) is licensed as a 250 kW TRIGAreactor and therefore not capable of this level of operation. The analysis, inventories and released activities fromthe damaged fuel elements are thus scaled back to 25% in order to determine doses to occupational personnel andmembers of the general public.NUREG analysis assumes 50 elements were present in the referenced core and the central elementexperiences a greater than average burn up. At 1/50 or 2% of the total, the central element would contain 4% ofthe total activity in the core. The noble gas and radioiodine activities in this element are 3828.8 Ci of Krypton,9431 Ci of Iodine, and 3933 Ci of Xenon. A one year operation of the MUTR at 250 kW (25% of the NUREGexample) for 365 days is 91.25 MWd and the corresponding activities would be 957.2 Ci of Krypton, 2357.8 Ci ofIodine and 983.3 Ci of Xenon. The MIUTR contains 93 fueled elements whereas the NUREG/General Atomicsexample uses 50 elements in its assessment. Therefore, the activity per MUTR element would be decreased by afactor of 93/5 0, or 1.86. The scaled release activity table for the MUTR is shown below.Not all of the fission product activity would be released from the element as the fuel matrix acts strongly toretain the fission products. According to NUREG 2387 the gap activity fraction is approximately 1 .5x10-5.IsotopeReleased Activities(mCi)IsotopeReleasedActivities (mCi)Kr-83 m 0.2492 1-134 5.1290Kr-85m 0.5782 1-135 4.4621Kr-85 0.0097 Sr-89 2.0161Kr-87 1.1129 Sr-90 0.0625Kr-88 1.5911 Sr-91 2.6048Xe-133m 0.0782 Sr-92 2.9516Xe-133 4.5702 Cs-134 0.0060Xe-135m 1.2056 Cs-134m 0.00363Xe-135 2.0669 Cs-136 0.05241-131 2.1774 Cs-137 1.00001-132 3.3492 Cs-138 4.1531-133 3.8976In addition to using MUITR release activities, allowance is taken for air leakage, radionuclide decay andshielding over the course of the event. Air leakage rates were determined using engineering analysis of theM\UTR. Decay rates were calculated from the Chart of the Nuclides as well as the Health Physics andRadiological Health Handbook (HTPRRH). The reduction in shine dose due to shielding was determined from2 Accident Analysis MHAfigure 6.11 of the HPRRH, Average Half-Value and Tenth Value Layers of Shielding Materials (Broad Beams),obtained from the NBS Handbook 138 1982 and Wachsman and Drexier 1975.Dose conversion factors in Federal Guidance Reports 11 and 12 are utilized in calculating doses tooccupational personnel and members of the public. Doses to the public are from ground level release due to airleakage from the south side of the reactor building. Horizontal and vertical diffusion coefficients, whereapplicable, were taken from Cember, Introduction to Health Physics third edition as referenced from D.H. Slade,Meteorology and Atomic Energy Tech Inform, 1968. Diffusion coefficients for distances less than 100 meters areextrapolated. A Pasquill category F, moderately stable condition, was chosen for all releases as a conservativecategory.3 Accident Analysis MHAMethodology for Occupational Dose CalculationsThe following are the formulae used to calculate the occupational doses:Committed Dose Equivalent (CDE) to the thyroid and CEDE for reactor occupational personnelCEDE = E [.BR

• DCF1nt

• Ai[1- exp(-aef, leak tst)]]Deep Dose Equivalent (DDE) to reactor occupational personnelTerms used in the above Dose EquationsBR Breathing Rate, per NRC Guidance [in3 S1l]DCFint Internal Dose Conversion Factor per FGR 11 [mrem uCi"1]DCFext External Dose Conversion Factor per FGR 12 [mrem m3 uCi-' s-a]Ai MUTR released activity per nuclide [uCi]leak Effective removal rate or leak constant, (24i +t-2) [S'l]24i Decay constant per nuclide [s-1])Xl Leakage constant per nuclide [s"1]V MUTR volume [in3]tst Reactor personnel stay time (evacuation time) [s]4 Accident Analysis MHAMethodology for Public Dose CalculationsFor the 10% of the time the air in the MUTR predominantly flows out of the southern side of the reactor due toatmospheric conditions. This scenario describes the MEl since during the other 90% of the time doses to anymember of the public are drastically lower.The CEDE, DDE, and Shine doses to the MEl member of the public are calculated as follows:CEDE = BR

• DCF1nt

• Cavg

• TstayParameters in CEDE EquationBR -Breathing rate per NRC Guidance [in3 s']DCFint -Internal Dose Conversion Factor per FRG 11 [inrem uCi-']Cavg -Average concentration in room 1398 [uCi m3]Tstay -Stay time [s]DDE = DCFext

• Ca

• TsaParameters in DDE EquationDCFext -External Dose Conversion Factor per FRG 12 [toremo m3 uCi' s1]Cavg -Average concentration in room 1398 [uCi m-3]Tstay -Stay time [s]Shine Dose = w* F**-(1 -e-.a), *In(2~Parameters in Shine Dose EquationF- Gamma constant for nuclide [remn hrI Ci-1 in2]C -Average concentration in the cloud in the MUTR [Ci m-3]r -Radius of the cloud in the MUJTR [in]h -Dose location distance from the surface of the cloud [in]S- lien, the linear energy absorption coefficient [m1]d -Diameter of the cloud in the MUTR [in]5 Accident Analysis MHASummary of DosesCEDE Occupational10.2 mremDDE Occupational1.62 mremTEDE Occupational11.82 mremCEDE publicAt MEl1-88.5 mremDDE publicAt MEl1-2.78 mremShine Dose PublicAt ME1-7.325 mremTEDE PublicMEI -98.605 mrem6