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 Univ of Maryland - College Park
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML16061A003 (60)

Text

UNIVERSIYO GLENN L. MARTIN INSTITUTE OF TECHNOLOGY A. JAMES CLARK SCHOOL OF ENGINEERING Department of Materials Science & Engineering Nuclear Reactor & Radiation Facilities Timothy W. Koeth, Director Building 090 College Park, Maryland 20742-2115 301.405.4952 TEL 301.405.6327 FAX 609.577.8790 CELL koethi@umd.edu February 29, 2016 Document Control Desk United States Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUBJECT:

UNIVERSITY OF MARYLAND - REQUEST FOR ADDITIONAL INFORMATION RE: FOR THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-70 THE MARYLAND UNIVERSITY TRAINING REACTOR DOCKET NO. 50-166 Enclosed please find the response to the RAI dated August 24, 2015 for the University of Maryland 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 Director University of Maryland Training Reactor & Radiation Facilities 10/p

Response To:

OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION FOR THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-70 THE MARYLAND UNIVERSITY TRAINING REACTOR DOCKET NO. 50-166

1. MUTR SAR, Section 4.5.2, "Reactor Core Physics Parameters," (Ref. 1) lists three reactivity coefficients and 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-Power Reactors Standard Review Plan and Acceptance Criteria," Section 4.5.2 provides guidance that an analysis should show that reactivity coefficients are sufficiently negative to prevent or mitigate damaging reactor transients. Describe what constitutes a power coefficient and show how overall reactivity coefficients are negative; or justify why the current method is acceptable.

Fuel Temperature Coefficient:

-1 1.2¢/ 0C Moderator Temperature Coefficient:

+3.0 ¢/°C Reactor Power Coefficient:

-0.53 ¢!kW Listed above are the reactivity coefficients of MUIR. At first glance it may seem as though the sum coefficient has a positive value. Howvever, the sole positive contribution to the reactivity is from the moderator temperature. The moderator temperature increases very slowly in comparison to the fuel temperature 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 already significant negative reactivity added before any positive reactivity results from an increase in moderator temperature.

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 for Preparing 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 a minimum value of 2. Provide a DNBR analysis that indicates a minimum value of at least 2, or justify why one is not needed.

A conservative calculation of the DNBR for MUTR has returned a value of 2.96 while operating at 600kW and inlet temperature of 92 degrees Celsius. Support documentation for this DNBR is attached in document 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 will have no negative impact on the environment. The MUTR environmental monitoring program results were provided in response to RAIs No. 47 and No. 72 (Refs. 6 and 2, respectively). However, the results are from 2004, and therefore, are out of date. NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content, Section 11.1.7 provides guidance that an appropriate monitoring program should contain probable pathways to people, and trends of recorded results. Provide updated information on the environmental monitoring program or justify why it is not needed.

Fixed environmental area monitors are located at the MUTR restricted area boundary and at locations on the university campus, to record and tract the potential radiological impact the MUTR operations have on the surrounding environment.

Monitors are exchanged and analyzed at frequency not to exceed once per calendar quarter. Records are maintained in accordance with 10 CFR 20.2103. Historically, and over the past 5 years, dose determinations for members of the public based upon this program, indicate doses to the public are in compliance with the limits of 20 CFR 20.1301.

In addition to fixed environmental area monitors, exposure rate Geiger Muller measurements are taken monthly in unrestricted areas outside of the MUTR boundary to monitor potential exposure to the public 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, and determination 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 the course of events and mitigate consequences (including automatic functions and operator actions) until a stabilized condition has been reached should be described. The accident analysis appears to be limited to uniform mixing of fission products in the reactor room and subsequent elevated 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 separate scenarios are analyzed; (iv) what the sequence for the analyzed exposure times (question 4(e)ii of this document); or (v) when a stable condition would be reached. Provide an updated analysis describing the sequence of events including initiation of engineered safety features to mitigate an 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 and assumptions developed for the "Radiation Protection Program and Waste Management,"

chapter of the SAR should be adapted as appropriate for the analysis. Submitted information should allow the results to be independently verified. The following parameters require further clarification:

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 per second (page 1, Ref. 3), and room leakage parameter of 0.00236 meters cubed per second (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 (pages 4, 8, and 12, Ref. 3) is inconsistent with the breathing rate of 4.27x20-04 meters cubed per 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 are provided as input parameters for "HOTSPOT" (page 16, Ref. 3). However, dispersion values for various distances and atmospheric stability classes (page 3, Ref. 3) cannot be verified using these input parameters. Provide an updated analysis clearly stating confinement leakage, breathing rates, release heights, and wind speed parameters as necessary before each series of computations, or justify why the current method is acceptable.

See Attachment Titled "Accident Analysis MHA" c) Guidance in NUREG-1537, Section 13.2, item (6) provides for defining the source term quantity of radionuclides. The fission product inventory is 25 percent equivalent of those described in NUREG/CR-2387 (page 2, Ref. 3). It appears the activities of Cesium and Strontium are less than 25 percent of those values listed in NUREG/CR-2387. Provide an updated analysis using consistent methodology for determining the source term, or justify why the current method is acceptable.

See Attachment Titled "Accident Analysis MHA" d) Guidance in NUREG-1537, Section 13.2, item (6) provides for describing a source term that could 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 to members 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 not considered. NUREG-1537, Section 13.2, item (7) provides guidance for evaluating exposure of a member of the public until the situation is terminated or the person is moved. Provide an updated analysis to include direct or scattered radiation exposure to members of the public specific 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 of the public specific to the facility until the situation are stabilized. The following locations for members of the public and times of exposure require further clarification:

i. Potential radiological consequences to members of the public in unrestricted areas are evaluated 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-campus residence hall and nearest off campus public residence from the reactor building at approximately 230 and 370 meters, respectively. A maximum exposed members of the public 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 for other locations of interest that may be applicable to the MUTR facility is provided in NUREG-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 a receptor 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 clear how to chronologically view the events, or if exposure times are consistent with one another. Provide an updated analysis clearly indicating exposure times and subsequent dose 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 the current method is acceptable.

See Attachment Titled "Accident Analysis MHA"

5. MUTR proposed TS 3.1, "Reactor Core Parameters," Specification (5) describes reactivity coefficients at the MUTR (Ref. 7). NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content, Section 14, Appendix 14.1, Section 4 provides guidance that certain limiting conditions for operations have accompanying surveillance requirements to include test, method, frequency, and acceptability. It appears the reactivity coefficients do not have a surveillance requirement. Provide a surveillance specification for TS 3.1 Specification (5), or justify why one is not necessary.

See updated TS 3.1

6. MUTR proposed TS 3.7, "Limitations On Experiments," Specification (4) describes limits on experiments (Ref. 7). Specification (4) describes explosive materials in quantities greater than 25 milligrams and less than 25 milligrams, but does not include quantities equal to 25 milligrams. Provide a revised TS 3.7 Specification (4) to provide for explosive material quantities equal to 25 milligrams, or justify why no change is necessary.

See updated TS 3.7

7. MUTR proposed TS 4.1, "Reactor Core Parameters," Specification (5) describes annual inspections of fuel 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 in the 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 Format and Content, Section 14, Appendix 14.1, Section 1.2.2 provides guidance that the proposed TS basis should be reference to the facility's analysis. Provide a revision to proposed TS 4.4 to include a qualitative 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 is necessary.

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 and Reviewing 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's analysis. 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 is necessary.

See updated 1S 5.2

10. 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 for Preparing 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 for restricting certain fuel bundles from core positions so that assumptions used in the development safety limits are met. NUREG-1537, Section 14, Appendix 14.1, Section 1.2.2 provides guidance that a TS should include a basis for each specification. Propose a TS including a basis that incorporates acceptable IFE locations, or justify why no change is necessary.

See updated TS 2.2

11. 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, "MUIR Organizational Structure," show solid-line and dashed-line connections, but appear to be missing 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 a description of the connection lines in Figures 6.1 and 6.2.

See updated TS Figures 6.1 & 6.2 b) Figure 6.2, "MUIR Organizational Structure," shows members of the MUIR organization including staff and management. However, the TSs do not appear to correlate the MUIR members 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 provided in ANSl/ANS-25.4-2007, Section 3. Clarify the level of assignment in the TSs for the members shown in Figure 6.2.

See updated TS Figure 6.2 c) ANSl/ANS-15.1-2007, Section 6.1.2 provides guidance that management not only be responsible for policies and operation, but shall also adhere to all requirements of the operating license and TSs. MUIR proposed TS 6.1.2, "Responsibility," describes specific responsibilities for the 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 MUTR member shown in Figure 6.2.

TS 6.1.2 has been rewritten to include responsibilities of all MUTR members shown in Figure 6.2

12. MUTR proposed TS 6.1.3, "Facility Staff Requirements," Specification (1) describes facility staffing requirements when the "reactor is operating" (Ref. 7). However, ANSI/ANS-15.1-2007, Section 6.1.3 provides 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 of radioactivity," (Ref. 7). However, "new experiment," is not defined, nor is the terminology consistent with the MUTR proposed TS Definition 1.7 or Specification 6.5. It is not clear which category of experiments are applicable in proposed TS 6.2.1.2. Provide a revised TS 6.2.1.2 to delineate which experiments 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 of an 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 are considered new.

14. MUTR proposed TS 6.5, "Experiment Review And Approval," Specification (3) uses the term "desired alternate," 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 a revision to the proposed TS 6.5 or justify why no change is necessary.

See updated TS 6.5

15. 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." It appears Definition 1.27 is erroneously used in proposed TS 6.7.2. Provide a revision to proposed TS 6.7.2 or justify why no change is necessary.

See updated TS 6.7.2

16. The following typographical errors were noticed. Consider reviewing the proposed TSs for other typographical or formatting errors and propose corrections as necessary.

a)

MUTR proposed TS Definition 1.37 may contain a grammatical error, See updated TS 1.37 b) MUTR proposed TS Definition 1.41 is numbered as 1.401, See updated TS 1.41 c) MUTR proposed TS 4.4 Specification may contain a grammatical error,

See updated TS 4.4 d) MUTR proposed TS 5.2.1 Specification (1) appears to erroneously use "connective,"

See updated TS 5.2 e) MUTR proposed TS 5.3.1 Specification (4) appears to be missing, See updated TS 5.3.1 f) MUTR proposed TS 5.3.2 Specification (1) states the control rods will contain borated graphite BvC, and See updated TS 5.3.2 g) MUTR proposed TS 5.4 Specification (3) appears to be missing.

See updated TS 5.4 OTHER 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 SPECIFICATIONS FOR THE MARYLAND UNIVERSITY TRAINING REACTOR License Number R-70 Docket Number 50-166 Submitted to United States Nuclear Regulatory Commission 29 February 2016 (Superseding 27 September 2011 Submission)

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16Rev02292016.docx Last edit February 29, 2016 TABLE OF CONTENTS TABLE OF CONTENTS

.i LIST OF TABLES

.iv LIST OF FIGURES.

.v 1.0 DEFINITIONS

.1 1.1 ALARA

.1 1.2 Channel

.1 1.3 Confinement

.1 1.4 Control Rod Guide Tube

.1 1.5 Core Configuration

.1 1.6 Excess Reactivity

.1 1.7 Experiment

.1 1.8 Experimental Facilities

.2 1.9 Experiment Safety Systems

.2 1.10 Four Element Fuel Bundle

.2 1.11 Fuel Element

.2 1.12 Fueled Device.

.2 1.13 Full Power

.2 1.14 Instrumented Element.

.2 1.15 Isolation

.2 1.16 Limiting Conditions for Operation

.2 1.17 Limiting Safety System Setting

.2 1.18 Measuring Channel

.2 1.19 Measured Value

.2 1.20 Moveable Experiment.

.2 1.21 On Call

.3 1.22 Operable

.3 1.23 Operating

.3 1.24 Reactivity Worth of an Experiment

.3 1.25 Reactor Console Secured

.3 1.26 Reactor Operating

.3 1.27 Reactor Operator

.3 1.28 Reactor Safety Systems

.3 1.29 Reactor Secured

.3 1.30 Reactor Shutdown

.3 1.31 Reference Core Condition

.4 1.32 Reportable Occurrence

.4 1.33 Rod-Control

.4 1.34 Safety Channel

.4 1.35 Safety Limit

.4

O:\\M UTR\\2016M UTRLivingDocs\\WorlingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 1.36 Scram Time

.4 1.37 Secured Experiment

.4 1.38 Secured Shutdown

.5 1.39 Senior Reactor Operator

.5 1.40 Shall, Should, May

.5 1.41 Shutdown Margin

.5 1.42 Shutdown Reactivity.

5 1.43 Standard Core.

.5 1.44 Steady State Mode

.5 1.45 Three Element Fuel Bundle

.5 1.46 True Value

.5 1.47 Unscheduled Shutdown 5

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 6

2.1 Safety Limit

.6 2.2 Limiting Safety System Settings.

.6 3.0 LIMITING CONDITIONS FOR OPERATION.

8 3.1 Reactor Core Parameters.

.8 3.2 Reactor Control and Safety Systems.

.9 3.3 Primary Coolant System.

.13 3.4 Confinement

.14 3.5 Ventilation Systems

.14 3.6 Radiation Monitoring System and Effluents 15 3.6.1 Radiation Monitoring System.

15 3.6.2 Effluents.

.16 3.7 Limitations on Experiments 17 4.0 SURVEILLANCE REQUIREMENTS 19 4.1 Reactor Core Parameters 19 4.2 Reactor Control and Safety Systems.

20 4.3 Primary Coolant System 21 4.4 Confinement 22 4.5 Ventilation System 22 4.6 Radiation Monitoring System and Effluents 23 4.6.1 Radiation Monitoring System.

23 4.6.2 Effluents

.23 4.7 Experiments

.24 5.0 DESIGN FEATURES 25 5.1 Site Characteristics 25 5.2 Reactor Coolant System 25 ii

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 5.3 Reactor Core and Fuel 26 5.3.1 Reactor Fuel

.27 5.3.2 Control Rods.

27 5.4 Fissionable Material Storage 28 6.0 ADMINISTRATION

.29 6.1 Organization

.29 6.1.1 Structure 29 6.1.2 Responsibility.

29 6.1.3 Facility Staff Requirements 29 6.1.4 Selection and Training of Personnel 33 6.2 Review and Audit 33 6.2.1 Reactor Safety Committee 33 6.2.1.1 Reactor Safety Committee Charter and Rules 33 6.2.1.2 Reactor Safety Committee Review Function 34 6.2.1.3 Reactor Safety Committee Audit Function 34 6.2.2 Audit of ALARA Program 35 6.3 Radiation Safety

.35 6.4 Operating Procedures.

35 6.5 Experiment Review and Approval 36 6.6 Required Actions 36 6.6.1 Actions to be Taken in Case of Safety Limit Violation 36 6.6.2 Actions to be Taken in the Event of a Reportable Occurrence 37 6.7 Reports.

.37 6.7.1 Annual Operating Report 37 6.7.2 Special Reports

.38 6.7.3 Unusual Event Report 39 6.8 Records

.39 3

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 LIST OF TABLES Table 3.1 Reactor Safety Channels: Scram Channels 11 Table 3.2 Reactor Safety Channels: Interlocks 11 Table 3.3 Reactor Safety Channels: Scram Channel Bases.

12 Table 3.4 Reactor Safety Channels: Interlock Bases 12 Table 3.5 Minimum Radiation Monitoring Channels 16 4

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 LIST OF FIGURES Figure 6.1 MUTR Position in University of Maryland Structure 30 Figure 6.1 MUTR Organizational Structure 31 5

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Included in this document are the Technical Specifications and the "Bases" for the Technical Specifications. These bases, which provide the technical support for the individual technical specifications, are included for information purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

1.0 DEFINITIONS 1.1 ALARA.(acronym for "as low as is reasonably achievable") means making every reasonable effort to maintain exposures to radiation as far below the dose limits in 10 CFR Part 20 as is practical consistent with the purpose for which the licensed activity is undertaken, taking into account 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 other societal and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed materials in the public interest.

1.2 CHANNEL - A channel is the combination of sensors, lines, amplifiers, and output devices which are 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 its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, 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 by observation of channel behavior, or by comparison of the channel with other independent channels 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 is operable.

1.3 CONFINEMENT - Confinement means a closure on the overall facility that controls the movement 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 fuel elements, arranged in a rectangular array with one bundle displaced for the pneumatic experimental system; three control rods; and two graphite reflectors.

1.6 EXCESS REACTIVITY - Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive condition from the point where the reactor is exactly 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 for irradiation within the pool, on or in a beamport or irradiation facility, and that is not rigidly secured 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 performed in the course of the reactor program.

1

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2. Modified Routine Experiments - Modified routine experiments are those which have not been performed previously but are similar to routine experiments in that the hazards are neither greater nor significantly different than those for the corresponding routine experiments.

3.

Special experiments - Special experiments are those which are not routine or modified routine experiments.

1.8 EXPERIMENTAL FACILITIES - Experimental facilities are facilities used to perform experiments 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, including their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated.

1.10 FOUR ELEMENT FUEL BUNDLE - The 4-element fuel bundle consists of an aluminum bottom, 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 a sheathed chromel-alumel or equivalent thermocouple is embedded in the fuel.

1.15 ISOLATION - Isolation is the establishment of confinement, closing of the doors leading from the reactor bay area leading into the balcony area on the top floor, the door to the reception area on the ground floor, and the building exterior doors.

1.16 LIMITING CONDITIONS FOR OPERATION - Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

1.17 LIMITING SAFETY SYSTEM SETTING-Limiting safety system settings (LSSS) for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.

1.18 MEASURING CHANNEL - A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device, which are connected for the purpose of measuring the value of a variable.

1.19 MEASURED VALUE - The measured value is the value of a parameter as it appears on the output of a channel.

1.20 MOVEABLE EXPERIMENT - A movable experiment is one where it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.

2

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTech nicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 1.21 ON CALL - A senior operator is available "on call" if the senior operator is either on the College Park campus or within 10 miles from the facility and can reach the facility within one half hour following a request.

1.22 OPERABLE - Operable means a component or system is capable of performing its intended function.

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 the Value of the reactivity change that results from the experiment being inserted into or removed from 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 the console.

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 their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action. Manual protective action is considered 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 is insufficient fissile material present in the reactor to attain criticality under optimum available conditions of moderator and reflection, or

2. 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, or control rod drives unless they are physically decoupled from the control rods; and

d. No experiments in or near the reactor are being moved or serviced that have, on movement, the smaller of: a reactivity worth exceeding the maximum value allowed for 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 the reference core condition with the reactivity worth of all installed experiments included and the following conditions exist:

a. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods; 3

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b. No experiments are being moved or serviced that have, on movement, a reactivity worth 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 of the core when it is at 20 °C and the reactivity worth of xenon is zero (i.e., cold, clean, and critical).

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 the Limiting Safety-System Settings specified in technical specifications 2.2.

2.

Operation in violation of the Limiting Conditions for Operation established in the technical specifications.

3.

A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests. (Note: Where components or systems are provided in addition to those required by the Technical Specifications, the failure of the extra components or systems is not considered reportable provided that the minimum number of components 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 such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.

1.33 ROD-CONTROL - A control rod is a device fabricated from neutron absorbing material which is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod may be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

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 be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity.

1.36 SCRAM TIME - Scram time is the elapsed time between the initiation of a scram signal by either automated or operator initiated action and the time required for the control rods to reach a fully inserted position into the core.

1.37 SECURED EXPERIMENT - A secured experiment is any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces that can arise as a result of 4

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1.38 SECURED SHUTDOWN - Secured shutdown is achieved when the reactor meets the requirements of the definition of "reactor secured" and the facility administrative requirements for leaving the facility with no licensed reactor operators present.

1.39 SENIOR REACTOR OPERATOR - A senior reactor operator (SRO) is an individual who is licensed 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,,shouldee is used to denote a recommendation; and the word,,may" is used to denote permission, neither a requirement nor a recommendation.

1.41 SHUTDOWN MARGIN - Shutdown margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems starting from any permissible operation condition and with the most reactive rod in its most 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 reactor with all control rods in their least reactive position (e.g., inserted). The value of shutdown reactivity includes the reactivity value of all installed experiments and is determined with the reactor at ambient conditions.

1.43 STANDARD CORE - A standard core is an arrangement of standard TRIGA fuel in the reactor grid plate.

1.44 STEADY STATE MODE - Steady state mode operation shall mean operation of the reactor with the mode selector switch in the STEADY STATE position.

1.45 THREE ELEMENT FUEL BUNDLE - The 3-element fuel bundle consists of an aluminum bottom, 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 unplanned shutdown 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 adversely affect safe operation, not to include shutdowns which occur during testing or check-out operations.

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O :\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 2.0 SAFETY LIMITS AN]) LIMITING SAFETY SYSTEM SETTING 2.1 SAFETY LIMIT Applicability This specification applies to the temperature of the reactor fuel.

Objective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding will result.

Specification The temperature in a standard TRIGA fuel element shall not exceed 1000 °C under any conditions of operation, with the fuel fully immersed in water.

Basis The important parameter for TRIGA reactor is the UZrH fuel element temperature. This parameter is well suited as a single specification especially since it can be measured. A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the fuel-moderator. The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium. The data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of ZrHx will remain below the ultimate stress if the temperature in the fuel does not exceed 1000 °C and the fuel cladding is water-cooled.

It has been shown by experience that operation of TRIGA reactors at a power level of 1000 kW will not result in damage to the fuel. Several reactors of this type have operated successfully for several years at power levels up to 1500 kW. Analysis and measurements on other TRIGA reactors have shown that a power level of 1000 kW corresponds to a peak fuel temperature of approximately 400 °C.

2.2 LIMITIN4G SAFETY SYSTEM SETTINGS Applicability This specification applies to the reactor scram setting that prevents the reactor fuel temperature from reaching the safety limit.

Objective The objective is to provide a reactor scram to prevent the safety limit (fuel element temperature of 1000 °C) from being reached.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Specification The 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 is placed in any position other than D8 in the grid plate, an analysis must be performed. The analysis shall indicate that the proposed location shall have a peaking factor not less than 50% of the highest fuel element in the core.

Basis A Limiting Safety Setting of 175 °C provides a safety margin of 650 °C. A part of the safety margin is used to account for the difference between the temperature at the hot spot in the fuel and the measured temperature resulting from the actual location of the thermocouple. If the thermocouple element is located in the hottest position in the core, the difference between the true and measured temperatures will be only a few degrees since the thermocouple junction is at the mid-plane of the element and close to the anticipated hot spot. If the thermocouple element is located in a region of lower temperature, such as on the periphery of the core, the measured temperature 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 the core, the true temperature at the hottest location in the core will differ from the measured temperature by no more than a factor of two. Thus, with the WFE positioned in the location specified by the license, when the temperature in the thermocouple element reaches the setting of 175 °C, the true temperature at the hottest location would be no greater than 350 °C, providing a margin to the safety limit of at least 650 °C. This margin is ample to account for the remaining uncertainty in the accuracy of the fuel temperature, measurement channel, and any overshoot in reactor power resulting from a reactor transient during steady state mode operation.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR CORE PARAMETERS Applicability These specifications shall apply to the reactor at all times it is operating.

Objective The objectives are to ensure that the reactor can be controlled and shut down at all times and that the safety limits will not be exceeded.

Specifications

1. The excess reactivity relative to the cold critical conditions, with or without experiments in 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; and

b.

Total worth of all in-core experiments in their most reactive state; and

c.

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 by more than the worth of the most reactive fuel element.

c.

No control rods shall be removed from the core unless a minimum of four fuel bundles 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 product release into the reactor coolant) except to locate such fuel.

5. The reactivity coefficients for the reactor are:

Fuel:

-1.2 ¢/°0C Moderator:

+3.0 ¢/°0C Power:

-0.53 ¢/kW The Fuel Temperature Coefficient, and the Moderator Temperature Coefficient, shall be verified any time the standard core is modified either by the rotation of a fuel bundle, a change in fuel bundle or reflector location, or the replacement of any fuel or reflector. The Power Temperature Coefficient shall be recalculated if either the Fuel Temperature or Moderator Temperature Coefficient are measured to be more than 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.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Bases

1. While specification 3.1.1, in conjunction with specification 3.1.2, tends to over constrain the excess reactivity, it helps ensure that the operable core is similar to the core analyzed in the FSAR.

2.

The value of the shutdown margin as required by specification 3.1.2 assures that the reactor can be shutdown from any operating condition even if the highest worth control rod should remain in the fully withdrawn position.

3. Specification 3.1.3 ensures that the operable core is similar to the core analyzed in the FSAR. It also ensures that accidental criticality will not occur during fuel or control rod manipulations.

4. Specification 3.1.4 limits the fission product release that might accompany operation with a damaged fuel element. Fuel will be considered potentially "Damaged" if said fuel is found to be leaking under the air and/or water sampling or under such case that the fuel has been exposed to temperature above 175 °C as measured on the instrumented fuel element. The criteria of the water and air sampling to determine a leaking fuel element is considered 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 is negative.

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 SYSTEMS Applicability These specifications apply to reactor control and safety systems and safety-related instrumentation that must be operable when the reactor is in operation.

Objective The objective of these specifications is to specify the lowest acceptable level of performance or the minimum number of operable components for the reactor control and safety systems.

Specifications

1. The drop time of each of the three standard control rods from the fully withdrawn position to the fully inserted position shall not exceed one second.

2. Maximum positive reactivity insertion rate by control rod motion shall not exceed $0.30 per second.

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3. The reactor safety channels shall be operable in accordance with Table 3.1l, including the minimum number of channels and the indicated maximum or minimum set points for the scram channels.

4. The safety interlocks shall be operable in accordance with Table 3.2, including the minimum number of interlocks.

5. The Beam Port and Through Tube interlocks may be bypassed during a reactor operation with the permission of the Reactor Director.

6.

A minimum of one reactor power channel, calibrated for reactor thermal power, must be attached to a recording device sufficient for auditing of reactor operation history.

Bases

1. Specification 3.2.1 assures that the reactor will be shutdown promptly when a scram signal is initiated. Experiments and analysis have indicated that for the range of transients anticipated for the MUTR TRIGA reactor, the specified control rod drop time is 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 the normally available reactivity and insertion, rate cannot generate operating conditions that exceed the Safety Limit. (See FSAR)

3.

Specification 3.2.3 provides protection against the reactor operating outside of the safety limits. 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 safety limits. 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 intended purpose. This purpose is to assure that the operator is aware of the status of both the beam ports and the through tube.

6.

Specification 3.2.6 provides for a means to monitor reactor operations and verify that the reactor is not operated outside of its license condition.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Table 3.1: Reactor Safety Channels: Scram Channels Scram Channel Minimum Required Operable 2

Scram Setpoint Reactor Power Level Not to exceed 120 %

Fuel Element Temperature Reactor Power Channel Detector Power Supply 1

2 Manual Scram Console Electrical Supply Rate of power change - Period Radiation Area Monitors 1

1 1

1 Not to exceed 175 °C Loss of power supply voltage to chamber N/A Loss of electrical power to the control console Not less than 5 seconds 50 mr/hr (bridge monitor) 10 mr/hr (exhaust monitor)

Table 3.2: Reactor Safety Channels: Interlocks Interlock/Channel Log Power Level Startup Count rate Safety 1 Trip Test Plug Electrical Connection Rod Drive Control Function Provide signal to period rate and minimum source channels. Prevent control rod withdrawal when neutron count rate is less than l cps.

Prevent control rod withdrawal when neutron count rate is less than 1 cps.

Prevent control rod withdrawal when Safety 1 Trip Test switch is activated.

Disable magnet power when Beam Port or Through Tube plug is removed unless bypass has been activated.

Prevent simultaneous manual withdrawal of two or more control rods in the steady state mode of operation.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Table 3.3: Reactor Safety Channels: Scram Channel Bases Scram Channel Bases Reactor Power Level Fuel Element Temperature Reactor Power Channel Detector Power Supply Manual Scram Console Electric Supply Rate of power change -

Period Radiation Area Monitors Provides protection to assure that the reactor can be shut down before the safety limit on the fuel element temperature will be exceeded.

Provides protection to assure that the reactor cannot be operated unless the neutron detectors that input to each of the linear power channels are operable.

Allows the operator to shut down the reactor if an unsafe or abnormal condition occurs.

Assures that the reactor cannot be operated without a secure electric supply.

Assures that the reactor is operated in a manner that allows the operator time to shut down the reactor before the licensed power restriction is exceeded.

Assures that the reactor automatically scrams if a high airborne radiation level is detected.

Table 3.4: Reactor Safety Channels: Interlock Bases Interlock/Channel Log Power Level Startup Count rate Safety 1 Trip Test Plug Electrical Connection Rod Drive Control Bases This channel is required to provide a neutron detector input signal to the startup count rate channel.

Assures sufficient amount of startup neutrons are available to achieve a controlled approach to criticality.

Assures that the 1 cps interlock cannot be bypassed by creating an artificial 1 cps signal with the Safety 1 trip test switch Assures that the reactor cannot be operated with Beamport or Through Tube plugs removed without further precautions.

Limits the maximum positive reactivity insertion rate available for steady state operation.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 3.3 PRIMARY COOLANT SYSTEM Applicability_

This specification applies to the quality and quantity of the primary coolant in contact with the fuel cladding at the time of reactor startup.

Objectives

1. 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.

Specifications

1. 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 shall be 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 located in the pool.

Bases

1. Specification 3.3.1 ensures that both sufficient cooling capability and sufficient biological shielding are available for safe reactor operation.

2. A small rate of corrosion continuously occurs in a water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limit provides acceptable control. In addition, by limiting the concentration of dissolved materials in the water, the radioactivity of neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel exposures during maintenance and operation.

3. Specification 3.3.3 ensures that a fuel failure with release of radioactive materials into the pool will be determined.

4. Specification 3.3.4 ensures a DNBR value greater than 2.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 3.4 CONFINEMENT

_Applicability This specification applies to that part of the facility that contains the reactor, its controls and shielding.

Objective The objective of these specifications is to ensure that sufficient confinement volume is available for the dilution of radioactive releases and to limit the rate of release of radioactive material to the outside environment.

_Specifications

1. Confinement shall be considered established when the doors leading from the reactor bay area leading into the balcony area on the top floor, and the reception area as well as the building exterior are secured.

2.

Confinement shall be established whenever the reactor is in an unsecured mode with the exception of the time that persons are physically entering or leaving the confinement area.

Bases

1. This specification provides the necessary requirements for confinement, which ensures releases to the outside environment are within 10 CFR Part 20 requirements.

2. This specification provides the reactor status condition for confimement, as well as allows personnel to enter and leave the reactor building, as required, when the reactor is unsecured.

3.5 VENTILATJON SYSTEMS Applicability These specifications apply to the ventilation systems for the reactor building.

_Objective The objective of these specifications is to ensure that air exchanges between the reactor confinement building and the environment do not impact negatively on the general public.

_Specifications

1. Air within the reactor building shall not be exchanged with other occupied spaces in the building.

2. All locations where ventilation systems exchange air with the environment shall have failsafe closure mechanisms.

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3. Forced air ventilation to the outside shall automatically secure without operator intervention in such case that the radiation levels exceed a preset level as defined in facility procedures. The setpoints are: 50 mR/hr (bridge monitor), 10 mR/hr (exhaust monitor).

Bases

1. This specification ensures that radioactive releases inside the reactor building will not be transported to the remainder of the building.

2.

This specification ensures that the reactor building can always be isolated from the environment.

3. This specification ensures that radioactive release will be minimized by stopping forced flow to the outside environment.

3.6 _

RADIATION MONITORING SYSTEM ANT) EFFLUENTS 3.6.1 Radiation Monitoring System Applicability This specification applies to the radiation monitoring information that must be available to the reactor operator during reactor operation.

Objective The objective is to assure that sufficient radiation monitoring information is available to the operator to assure safe operation of the reactor.

Specifications

1. The reactor shall not be operated unless a minimum of one of the two radiation area monitor 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 radiation monitor channels, the intent of specification 3.6.1 will be satisfied if they are replaced with portable gamma sensitive instruments having their own alarms or which shall be observable by the reactor operator.

3. The alarm set points shall be stated in a facility operating procedure. The alarm setpoints for the bridge monitor are:.37 mR/hr (alert), 50 mR/hr (scram). The setpoints for the exhaust monitor are: 8 mR/hr (alert), 10 mR/hr (scram).

4.

The campus radiation safety organization shall maintain an environmental monitor at the MUJTR site boundary.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Table 3.5: Minimum Radiation Monitoring Channels Radiation Area Monitors Exhaust Radiation Monitor Bridge Radiation Monitor Function Minimum Number Operable Monitor radiation levels in Reactor Bay area at an Exhaust Fan location Monitor radiation levels in Reactor Bay area at the Reactor Bridge location A minimum of 1 of the 2 monitors shall be operable Bases

1. Specification 3.6.1.1 ensures that a significant fuel failure with release of radioactive materials will be determined and that any large releases will be mitigated by the specified protective actions.

2.

Specification 3.6.1.2 allows for continued reactor operation if maintenance and/or calibration of the radiation area monitors is required.

3. The alarm and scram set points shall be designed to ensure that dose rates delivered to areas accessible to members of the general public do not exceed the levels defined in 10 CFR Part 20. Additionally, the radiation area monitors provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of 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 a dose to the general public greater than that allowed by 10 CFR Part 20.

3.6.2 Effluents Applicability.

This specification applies to limits on effluent release.

Objective The objective is to ensure that the release of radioactive materials from the reactor facility to unrestricted areas do not exceed federal regulations.

Specification All effluents from the MUTR shall conform to the standards set forth in 10 CFR Part 20.

Basis The intent of 3.6.2 is to ensure that, in the event that radioactive effluents are released, the dose to the general public will be less than that allowed by 10 CFR Part 20.

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O :\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 3.7 LIMITATIONS ON EXPERIMENTS Applicability The specification applies to experiments installed in the reactor and its experimental facilities.

Objective The objective is to prevent damage to the reactor or excessive release of radioactive material in the event of an experiment failure.

Specifications The 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 and experiment flooding or voiding.

3. Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials, and liquid fissionable materials shall be doubly encapsulated.

4. Explosive materials in quantities greater than 25 mg TNT or its equivalent shall not be irradiated in the reactor or experimental facilities. Explosive materials in quantities equal to or less than 25 mg may be irradiated provided the pressture produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the failure pressure of the container. The failure pressure of the container is one half of the design pressure. Total explosive material inventory in the reactor facility may not exceed 100 mg TNT or its equivalent.

5. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor or (3) possible accident conditions in the experiment shall be limited in type and quantity such that if 100 % of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne radioactivity in the reactor room or outside environment will not result in exceeding 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, which closes automatically on high radiation level, at least 10 % of the gaseous activity or aerosols produced will escape.

b. If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99 % efficiency for 0.3 *na particles, at least 10 % of these particles can escape.

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c.

If an experiment fails and releases radioactive gases or aerosols to the reactor bay or atmosphere, 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 (540 C), the vapors of at least 10 percent of the materials escape through an undisturbed column of water above the core.

6.

Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 5 mCi.

Bases

1. This specification is intended to provide assurance that the worth of a single unsecured experiment will be limited to a value such that the safety limit will not be exceeded if the positive 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 cold critical reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its inadvertent removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

The maximum worth of all experiments is also limited to a reactivity value such that the cold reactor will not achieve a power level high enough to exceed the core temperature safety limit if the experiments were removed inadvertently.

3.

This specification is intended to prevent damage to reactor components resulting from experiment failure. If an experiment fails, inspection of reactor structures and components 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 from failure of an experiment involving explosive materials, especially the accidental detonation of the explosive. If an experiment fails, inspection of reactor structures and components 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 of the limits of Table 2 of Appendix B of 10 CFR Part 20 will be released to the atmosphere outside the facility boundary.

6.

The 5 mCi limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less than that allowed by 10 CFR Part 20 for an unrestricted area. (See SAR) 18

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 4.0 SURVEILLANCE REQUIREMENTS INTRODUCTION Surveillances shall be performed on a timely basis as defined in the individual procedures governing the performance of the surveillance. In the event that the reactor is not in an operable condition, such as during periods of refueling, or replacement or repair of safety equipment, surveillances may be postponed until such time that the reactor is operable. In such case that any surveillance must be postponed, a written directive signed by the Facility Director, shall be placed in the records indicating the reason why and the expected completion date of the required surveillance. This directive shall be written before the date that the surveillance is due. Under no circumstance shall the reactor perform routine operations until such 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/component continues to meet performance requirements.

4.1 REACTOR CORE PARAMETERS Applicability These specifications apply to the surveillance requirements for the reactor core.

O~bjective The objective of these specifications is to ensure that the specifications of Section 3.1 are satisfied.

Specifications

1. 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 any removal 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 any removal or replacement of control rods

3. 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 to exceed 15 months, such that the entire core is inspected over a five year period.

6. Burnup shall be verified in the Annual Report.

Bases Experience has shown that the identified frequencies ensure performance and operability for each of these systems or components. For excess reactivity and shutdown margin, long-term changes are slow to develop. For fuel inspection, visually inspecting 20% of the bundles annually will identify any developing fuel integrity issues throughout the core.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 4.2 REACTOR CONTROL AND SAFETY SYSTEMS App~licability These specifications apply to the surveillance requirements for reactor control and safety systems.

Objective The objective of these specifications is to ensure that the specifications of Section 3.2 are satisfied.

Specifications

1. The reactivity worth of each standard control rod shall be determined annually, intervals not to exceed 15 months, and after each time the core fuel configuration is changed or a control rod is changed.

2.

The control rod withdrawal and insertion speeds shall be determined annually, intervals not to exceed 15 months, or whenever maintenance or repairs are made that could affect rod travel times.

3.

Control rod drop times shall be measured annually; intervals not to exceed 15 months, or whenever maintenance or repairs are made that could affect their drop time.

4. All scram channels and power measuring channels shall have a channel test, including trip actions with safety rod release and specified interlocks performed after each secured shutdown, before the first operation of the day, or prior to any operation scheduled to last more 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 channels shall be calibrated annually, intervals not to exceed 15 months.

5. Operability tests shall be performed on all affected safety and control systems after any maintenance is performed.

6. A channel calibration shall be made of the linear power level monitoring channels annually, intervals not to exceed 15 months.

7. A visual inspection of the control rod poison sections shall be made biennially, intervals not to exceed 28 months.

8. A visual inspection of the control rod drive and scram mechanisms shall be made annually, intervals not to exceed 15 months.

Bases

1. The reactivity worth of the control rods, specification 4.2.1, is measured to assure that the required shutdown margin is available and to provide a means to measure the reactivity worth of experiments. Long term effects of TRIGA reactor operation are such that measurements of the reactivity worths on an annual basis are adequate to insure that no significant changes in shutdown margin have occurred.

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2.

The control rod withdrawal and insertion rates, specification 4.2.2, are measured to insure that 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 can perform their safety function properly.

4.

The surveillance requirement specified in specification 4.2.4 for the reactor safety scram channels ensures that the overall functional capability is maintained.

5. The surveillance test performed after maintenance or repairs to the reactor safety system as required by specification 4.2.5 ensures that the affected channel will perform as intended.

6. The linear power level channel calibration specified in specification 4.2.6 will assure that the 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 made to evaluate corrosion and wear characteristics and any damage caused by operation in the reactor.

8. Specification 4.2.8 assures that a visual inspection of control drive mechanisms is made to evaluate corrosion and wear characteristics and any damage caused by operation in the reactor.

4.3 PRIMARY COOLANT SYSTEM Applicability These specifications apply to the surveillance requirements of the reactor primary coolant system.

Objective The objective of these specifications is to ensure the operability of the reactor primary coolant system as described in Section 3.3.

Specifications

1. The primary coolant level shall be verified before each reactor startup or daily during operations 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 pool water 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 exceed six weeks. If gross gamma activity is high (greater than twice historical data), gamma spectroscopy shall be performed.

4. Pool water temperature shall be measured prior to the reactor startup and shall be monitored during reactor operation.

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1. Specification 4.3.1 ensures that sufficient water exists above the core to provide both sufficient cooling capacity and an adequate biological shield.

2.

Specification 4.3.2 ensures that poor pool water quality could not exist for long without being detected. Years of experience at the MIUTR have shown that pool water analysis on a monthly basis is adequate to detect degraded conditions of the pooi water in a timely manner.

3.

Gross gamma activity measurements are conducted to detect fission product releases from damaged fuel element cladding.

4.

Specification 4.3.4 ensures that the maximum allowable pool water temperature is not exceeded.

4.4 CONFINEMENT Applicability This specification applies to that part of the facility which contains the reactor, its controls and shielding.

Objective The objective of this specification is to ensure that radioactive releases from the confinement can be limited.

_Specification Prior to putting the reactor in an unsecured mode, the isolation of the confinement building shall be visually verified.

Bases This specification ensures that the minimal leakage rate assumed in the SAR is actually present during reactor operations in order to limit the release of radioactive material to the environs.

4.5 VENTTLATION SYSTEM Applicability This specification applies to the reactor ventilation system.

_Objective The objective is to assure that provisions are made to restrict the amount of radioactivity released to the environment.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Specification The ability to secure the ventilation system shall be verified before the first reactor operation of the day.

Bases The facility is designed such that in the event that excessive airborne radioactivity is detected the ventilation system shall be shutdown to minimize transport of airborne materials. Analysis indicates that in the event of a major fuel element failure personnel would have sufficient time to evacuate the facility before the maximum permissible dose (10 CFR Part 20) is exceeded.

4.6 RADIATION MONITORING SYSTEM ANT) EFFLUENTS 4.6.1 Radiation Monitoring System Applicability This specification applies to the surveillance requirements for the Radiation Area Monitoring System (RAMS).

Objective The objective of these specifications is to ensure the operability of each radiation area monitoring channel as required by Section 3.6 and to ensure that releases to the environment are kept below allowable limits.

Specifications

1. A channel calibration shall be made for each channel listed in Table 3.5 annually but at intervals not to exceed 15 months or whenever maintenance or repairs are made that could affect their calibration.

2. A channel test shall be made for each channel listed in Table 3.5 prior to starting up the reactor to ensure reactor scram, fan shutdown, and louver closing.

Bases Specifications 4.6.1.1 and 4.6.1.2 ensure that the various radiation area monitors are checked and calibrated on a routine basis, in order to assure compliance with 10 CFR Part 20.

4.6.2 Effluents Applicability This specification applies to the surveillance requirements for air and water effluents.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Objective The objective of these specifications is to that releases to the environment are kept below allowable limits.

Specifications

1. Reactor building air samples shall be counted for gross gamma activity monthly, intervals not to exceed 6 weeks.

2.

A sample of any water discharged from the reactor building sump shall be counted for gross gamma activity before its release to the environs.

Bases Specifications 4.6.2.1 and 4.6.2.2 ensure that the facility effluents comply with 10 CFR Part 20.

4.7 EXPERIMENTS Applicability This specification applies to the surveillance requirements for experiments installed in the reactor and its irradiation facilities.

Objective The objective of this specification is to prevent the conduct of experiments which may damage the reactor or release excessive amounts of radioactive materials as a result of experiment failure.

Specifications

1. The reactivity worth of an experiment shall be estimated or measured, as appropriate, before reactor operation with said experiment

2.

An experiment shall not be installed in the reactor or its irradiation facilities unless a safety analysis has been performed and reviewed for compliance with Section 3.7 by the Reactor Safety Committee (new experiment) or Facility Director (modified routine experiment), in full accord with Sections 6.1.2 and 6.2.1 of these Technical Specifications and the procedures which are established for this purpose.

Basis Experience has shown that experiments reviewed and approved by the Reactor Safety Committee or Facility Director can be conducted without endangering the safety of the reactor, personnel, or exceeding Technical Specification limits.

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O:\\M UTR\\2Oi6M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 5.0 DESIGN FEATURES 5.1 SITE CHARACTERISTICS Applicability This specification applies to the reactor facility and its site boundary.

Objective The objective is to assure that appropriate physical security is maintained for the reactor facility and the radioactive materials contained within it.

Specifications

1. The reactor shall be housed in a closed room designed to restrict leakage. The closed room does not include the West balcony area.

2.

The reactor site boundary shall consist of the outer walls of the reactor building and the area enclosed by the loading dock fence.

3. The restricted area shall consist of all areas interior to the reactor building including the west balcony and lower entryway.

4.

The controlled area shall consist of all areas interior to the reactor building including the west balcony and lower entryway.

Bases These specifications assure that appropriate control is maintained over access to the facility by members of the general public.

5.2 REACTOR PRIMARY COOLANT SYSTEM Applicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.

Objective The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding.

Specifications

1. 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.

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3.

The pooi water return (outlet) pipe shall not extend more than 50.8 cm (20 in) below the overflow outlet pipe when fuel is in the core.

Bases

1. Specification 5.2.1 is based on thermal and hydraulic calculations and operation of the MUTR that show that the core can operate in a safe manner at power levels up to 300 kW with natural convection flow of the coolant.

2.

Specifications 5.2.2 and 5.2.3 ensures that the pool water level can normally decrease only by 50.8 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 FUEL Applicability This specification applies to the configuration of the core and in-core experiments.

Objective The objective is to ensure that the core configuration is as specified in the license.

Specifications

1. The core shall consist of 93 TRIGA fuel elements assembled into 24 fuel bundles - 21 bundles shall contain four fuel elements and 3 bundles shall contain three fuel elements and a control rod guide tube.

2. The fuel bundles shall be arranged in a rectangular 4 x 6 configuration, with one bundle displaced 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 water Basis

1. 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 limits for the maximum temperature in the fuel.

4. The reflector reduces the neutron leakage from the reactor core.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 5.3.1 Reactor Fuel Applicability This specification applies to the fuel elements used in the reactor core.

Objective The objective is to assure that the fuel elements are of such design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics, and that the fuel used in the reactor has characteristics consistent with the fuel assumed in the SAR and the license.

Specifications The individual unirradiated standard TRIGA fuel elements shall have the following characteristics:

1. Uranium content: a maximum of 9.0 weight % uranium enriched to less than 20 % 235U

2.

Zirconium hydride atom ratio: nominal 1.5 - 1.8 hydrogen-to-zirconium, ZrHx

3. 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 15 inches.

Basis The design basis of the standard TRIGA fuel element demonstrates that 250 kW steady state operation presents a conservative limitation with respect to safety limits for the maximum temperature generated in the fuel.

5.3.2 Control Rods Applicability This specification applies to the control rods used in the reactor core.

Objective The objective is to assure that the control rods are of such a design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications

1. The three control rods shall have scram capability, shall be used for reactivity control, and shall contain borated graphite, B4C, in powder form.

2.

The control rod cladding shall be aluminum with nominal thickness 0.71 mm (0.028") and length 17".

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTech nicai Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Basis The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, B4C, powder. These materials must be contained in a suitable clad material such as aluminum to ensure mechanical stability during movement and to isolate the poison from the tank water environments. Scram capabilities are provided for rapid insertion of the control rods, which is the primary safety feature of the reactor.

5.4 FISSIONABLE MATERIAL STORAGE Applicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Objective The objective is to assure that fuel that is being stored will not become critical and will not reach an unsafe temperature.

Specifications

1. All fuel elements shall be stored either in a geometrical array where the k-effective is less than 0.8 for 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 permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.

3. When fuel is in storage in any area other than the grid plate, that area must be equipped with monitoring devices that both measure and record the radiation levels and temperature of the region surrounding the fuel.

Basis The limits imposed by Specifications 5.4.1 and 5.4.2 are conservative and assure safe storage.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 6.0 ADMINISTRATION 6.1 ORGANIZATION The Maryland University Training Reactor (MUTR) is owned and operated by the University of Maryland, College Park. Its position in the university's structure is shown in Figure 6.1 The university shall provide whatever resources are required to maintain the facility in a condition that poses no hazard to the general public or to the environment.

6.1.1 Structure Figure 6.2 shows the MUTR organizational structure.

6.1.2 Responsibility The Dean College of Engineering is responsible for the oversight and operation of the school of engineering.

The Chair of the Department of Materials Science and Engineering is responsible for the oversight and operation of the Department of Materials Science and Engineering.

The Director of MUTR: Responsibility for the safe operation of the reactor facility and radiological safety shall rest with the Facility Director. The members of the organization chart shown in Figure 6.2 shall be responsible for safeguarding the public and facility personnel from undue 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 the activities of reactor operators.

The Reactor Operators are individuals who are licensed by the U.S. Nuclear Regulatory Commission (NRC) to manipulate the controls of the reactor.

6.1.3 Facility Staff Requirements

1. 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 be present in the control room.

b. A minimum of two persons shall be present in the facility or in the Chemical and Nuclear Engineering Building while the reactor is not secured: the operator in the control room and a second person who can be reached from the control room who is able to carry out prescribed written instructions which may involve activating elements of the Emergency Plan, including evacuation and initial notification procedures.

c.

A licensed SRO shall be present or readily available on call. "Readily Available on Call" means an individual who (1) has been specifically designated and the designation known to the operator on duty, (2) keeps the operator on duty informed of where he/she may be rapidly contacted and the method of contact, and (3) is 29

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 capable of arriving at the reactor facility within a reasonable amount of time under normal conditions. At no time while the reactor is not secured shall the designated SRO 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 in the control room for use by the operator. The list shall include:

a.

Management personnel

b.

Radiation safety personnel

c.

Licensed operators 30

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 President University of Maryland Provost & Senior Vice President Dean Clark School of Engineering Vice President Administrative Affairs Director Department of Environmental Safety Radiation Safety Committee Chair Department of Materials Science and Engineering Radiation Safety Officer 4L Director Nuclear Reactor Facility Reviews* * -

- Audit Reactor

.iembr Safety

."Mme Committee Radiation Safety Office Staff Services i

Reactor Operations Staff

-- Normal Administrative Reporting Channel Communication Lines Figure 6.1: MUTR Position in University of Maryland Structure 31

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Dean Clark School of Engineering (Level 1)

Chair Department of Materials Science and Engineering (Level 1)

Director Nuclear Reactor Facility (Level 2)

,I!

Senior Reactor Operator (Level 3)

Reactor Operator (Level 4)

Reactor Safety Committee 4---------

-- Normal Administrative Reporting Channel Communication Lines Figure 6.2: MUTR Organizational Structure 32

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3. The following operations shall be supervised by a senior reactor operator:

a.

Initial startup and approach to power following new fuel loading or fuel rearrangement

b.

When experiments are being manipulated in the core that have an estimated worth greater than $0.80

c. Removal of control rods or fuel manipulations in the core

d. Resumption of operation following an unplanned or unscheduled shutdown or any unplanned or unexpected significant reduction in power.

6.1.4 Selection and Training of Personnel The 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 the training 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 and 15.4.

6.2 REVIEW AND AUDIT 6.2.1 Reactor Safety Committee A Reactor Safety Committee (RSC) shall exist for the purpose of reviewing matters relating to the health and safety of the public and facility staff and the safe operation of the facility. It is appointed by and reports to the Chairperson of the Department of Materials Science and Engineering. The RSC shall consist of a minimum of five persons with expertise in the physical sciences and preferably some nuclear experience. Permanent members of the committee are the Facility 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 the committee. Alternates may be appointed by the Chairperson of the RSC to serve on a temporary basis. At least one committee member must be from outside the Department of Materials Science and Engineering.

6.2.1.1l Reactor Safety Committee Charter And Rules

1. 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 whom shall be the Campus Radiation Safety Officer (or designated alternate). No more than two alternates shall be used to make a quorum. MUTR staff members shall not constitute the majority of a voting quorum.

3. Minutes of all meetings will be retained in a file and distributed to all RSC members.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 6.2.1.2 Reactor Safety Committee Review Function The RSC shall review the following:

1. Determinations that proposed changes in equipment, systems, test, experiments, or procedures are allowed without prior authorization by the responsible authority, e.g. 10 CFR 50.59;

2. All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment, or systems having safety significance;

3.

All new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity;

4.

Proposed changes in technical specifications, or license;

5. Violations of technical specifications or license. Violations of internal procedures or instructions 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 1 management, the Facility Director, and the RSC members in a timely manner after the review has been completed.

6.2.1.3 Reactor Safety Committee Audit Function

1. An annual audit and review of the reactor operations will be performed by an outside individual or group familiar with research reactor operations. They shall submit a report to the 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 reactor facility equipment, systems, structures, or methods of operation that affect reactor safety; and 34

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g.

Reactor facility emergency plan and implementing procedures.

Deficiencies uncovered that affect reactor safety shall immediately be reported to Level 1 management and the Facility Director. A written report of the findings of the audit shall be submitted to Level 1 management, the Facility Director, and the RSC members within 3 months after the audit has been completed.

6.2.2 Audit of ALARA Program The Facility Director or his designated alternate shall conduct an audit of the reactor facility ALARA Program at least once per calendar year (not to exceed fifteen months). The results of the audit shall be presented to the RSC at the next scheduled meeting. This audit may occur as part of a review of the overall campus ALARA program.

6.3 RADIATION SAFETY A radiation safety program following the requirements established in 10 CFR Part 20 will be undertaken by the Radiation Safety Office. The facility director will ensure that ALARA principles are followed during all facility activities.

6.4 OPERATING PROCEDURES Written procedures, reviewed and approved by the Reactor Safety Committee, shall be in effect and followed for the following items prior to performance of the activity. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use of independent judgment and action should the situation require such.

1. Start-up, operation, and shutdown of the reactor

2.

Installation or removal of fuel elements, control rods, experiments, and experimental facilities

3. Maintenance procedures that could have an effect on reactor safety

4.

Periodic surveillance checks, calibrations, and inspections required by the Technical Specifications or those that may have an effect on reactor safety

5. Administrative controls for operations and maintenance and for the conduct of irradiations and experiments that could affect reactor safety or core reactivity

6. For any activity pertaining to shipping, possession, and transfer of radioactive material, these procedures shall be written in conjunction with the Radiation Safety Office and the Radiation Safety Officer who shall inform the Reactor Director of any changes in regulations or laws that may require modification of these procedures. All shipping and receiving of radioactive material shall be performed in conjunction w i t h, and with the approval of the Radiation Safety Office.

7.

Implementation, maintenance, and modification to the Emergency Plan

8. Implementation, maintenance, and modification to the Security Plan 35

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9. Implementation, maintenance, and modification to the Radiation Protection Plan.

The Radiation Protection Plan shall include an ALARA plan as defined in ANSI/ANS-15.l11

10. Use, receipt, and transfer of byproduct material Substantive changes to the above procedures shall be made with the approval of the Facility Director 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 only exception to this clause is in such a case where the delay in implementation would cause a credible risk to the public or the facility. If such a case exists as determined by the Facility Director, temporary approval may be granted by the Director but must be approved by the Reactor Safety Committee within thirty days. Temporary or minor changes to procedures shall be documented and subsequently reviewed by the Reactor Safety Committee at the next scheduled meeting. The Reactor Director shall have the power to approve minor changes such as phone number changes, typographical error correction or any other change that does not change the effectiveness or the intent of the procedure. It shall be considered sufficient approval and documentation when the Director forwards by electronic means to both the Radiation Safety Officer and the Chair of the Reactor Safety Committee. A copy of the transmission shall be filed with the appropriate procedure.

6.5 EXPERIMENT REVTEW AND APPROVAL

1. Routine experiments may be performed at the discretion of the duty senior reactor operator without the necessity of further review or approval.

2. Modified routine experiments shall be reviewed and approved in writing by the Facility Director, or designated alternate.

3.

Special experiments shall be reviewed by the RSC and approved by the RSC and the Facility Director 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 its effect 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 ACTIONS 6.6.1 Actions To Be Taken In Case Of Safety Limit Violation In the event a safety limit is exceeded:

1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

2. The event shall be reported to the Reactor Director who will report to the NRC as required in section 6.7.2.

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3. An immediate report of the occurrence shall be made to the Chairman, Reactor Safety Committee, and reports shall be made to the NRC in accordance with Section 6.7.2 of these specifications, and

4. A report, and any follow-up report, shall be prepared. The report shall describe the following:

a.

Applicable circumstances leading to the violation, including when known, the cause, and contributing factors;

b. Effect of the violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public; and

c. Corrective action to be taken to prevent recurrence.

The report shall be reviewed by the Reactor Safety Committee and submitted to the NRC when authorization is sought to resume operation of the reactor.

6.6.2 Actions to Be Taken In The Event Of a Reportable Occurrence In the event of a reportable occurrence, as defined in section 1.32 of these Technical Specifications, the following actions will be taken:

1. Immediate action shall be taken to correct the situation and to mitigate the consequences of the occurrence.

2.

The reactor shall be shut down and reactor operation shall not be resumed until authorized by the Facility Director.

3. The event shall be reported to the Facility Director who will report to the NRC as required in section 6.7.2.

4.

The Reactor Safety Committee shall investigate the causes of the occurrence at its next meeting. 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 the occurrence, the effectiveness of corrective actions taken, and recommendations of measures to prevent or reduce the probability or consequences of recurrence.

6.7 REPORTS 6.7.1 Annual Operating Report A report summarizing facility operations shall be prepared annually for the reporting period ending June 30. This report shall be submitted by December 30 of each year to the NRC Document Control Desk. The report shall include the following:

1. A brief narrative summary of results of reactor operations and surveillance tests and inspections required in section 4.0 of these Technical Specifications

2.

A tabulation showing the energy generated in MW hr-' for the year 37

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3. A list of unscheduled shutdowns including the reasons therefore and corrective action taken, if any

4. A tabulation of the major maintenance operations performed during the period, including the effects, if any, on safe operation of the reactor, and the reason for any corrective maintenance required

5.

A brief description of

a.

Each change to the facility to the extent that it changes a description of the facility in the Final Safety Analysis Report

b. 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 the environment

7.

A description of any environmental surveys performed outside of the facility

8. A summary of exposure received by facility personnel and visitors where such exposures are greater than 25 percent of limits allowed by 10 CFR Part 20

9.

Changes in facility organization 6.7.2 Special Reports Notification 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, followed by 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 document

2. Release of radioactivity from the site above allowed limits

3. Exceeding the Safety Limit The written report shall be sent to the NRC document control desk. The written report and, to the extent possible, the preliminary telephone or facsimile notification shall:

1. Describe, analyze, and evaluate safety implications

2.

Outline the measures taken to ensure that the cause of the condition is determine

3.

Indicate the corrective action taken to prevent repetition of the occurrence including chances to procedures

4. Evaluate the safety implications of the incident in light of the cumulative experience obtained from the report of previous failure and malfunction of similar systems and components 38

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 6.7.3 Unusual Event Report A written report shall be forwarded within 3 0 days to the NRC Document Control Desk, with a copy 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 methods used for such analysis as described in the Safety Analysis Report or in the bases for the Technical Specifications

2.

Discovery of any substantial variance from performance specifications contained in the Technical Specifications or Safety Analysis Report

3. Discovery of any condition involving a possible single failure which, for a system designed against assumed failure, could result in a loss of the capability of the system to perform its safety function

4. A permanent change in the position of Department Chair or Facility Director 6.8 RECORDS

1. The following records shall be retained for a period of at least five years:

a. Normal reactor facility operation and maintenance

b.

Reportable occurrences

c. Surveillance activities required by Technical Specifications

d. Facility radiation and contamination surveys

e. Experiments performed with the reactor

f. Reactor fuel inventories, receipts, and shipments

g.

Approved changes in procedures required by these Technical Specifications

h. Minutes of the Reactor Safety Committee meetings

i. Results of External Audits

2.

Retraining and requalification records of current licensed operators shall be maintained at all times 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 environs

b.

Gaseous radioactive effluents released to the environs

c.

Radiation exposure for all facility personnel 39

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016

d. Radiation exposures monitored at site boundary

e. As-built facility drawing

f.

Violation of the Safety Limit

g. 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 as records in subsection 6.8.3 above.

40

Accident Analysis MHA The NRC licenses research and test reactors consistent with the NRC mission to ensure adequate protection of the public health and safety and to promote and protect the environment. NUREG 1537 Part 2 Chapter 13 Accident Analysis provides guidance and acceptable format and content for licensees to present regarding a Maximum Hypothetical Accident, MHA.

Utilizing guidance from both NUJREG 1537 and NUJREG/CR-23 87, Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors, the bounding and limiting credible accident scenario is a fuel element failure which 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 of the reactor building outside of the biological shield. Fission products are released in air from the gap and the cladding and instantaneously and uniformly mix in the volume of the reactor building. The reactor facility exhaust fans are not running and are closed during this event. No immediate protective functions are activated by radiation 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 the unrestricted 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 MHA Doses are calculated for air leakage out of the south side of the reactor. Internal, external, and shine doses are 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 time air leakage occurs out of the reactor on the north side entrances, predominantly through the roll up door and significantly less through a single entrance door. The remaining 10% of the time analysis has shown air leakage sites are located, on the eastern, southern and western sides of the reactor building, the most prominent site being on the southern side and contributing 38% of the total air leakage to a single location. A highly conservative approach assumes the MHA's airborne radioactivity escapes continuously during the event from these predominant locations from the onset of the event until the end of the leakage time. This is the source of the Maximally Exposed Individual (MET) member of the public outside of the confinement space, and at locations downwind of the MUTR. Occupational personnel located within the MUTR are exposed to internal and external radiation from the release 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 space will 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, members of the public remain in place for the duration of the event, and doses are determined over this time interval. In reality 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 work together to secure and maintain acceptable perimeters relative to exposure and dose measurements surrounding the MULTR facility.

Per 10 CFR 20.1301 the NRC regulations limit the internal and external dose to 100 mrem for members of the public; this includes doses incurred during an incident such as the MHA.

1

Accident Analysis MHA Internal and external doses are accrued by occupational personnel present in the reactor at the start of the event until they have evacuated the confinement space. Evacuation time is overly conservative and set at 5 minutes.

NUREG 2387 provides guidance and analysis for a 1MW TRIGA after 1 year of continuous operation at full power, or 365MWd. The Maryland University Training Reactor (MUTR) is licensed as a 250 kW TRIGA reactor and therefore not capable of this level of operation. The analysis, inventories and released activities from the damaged fuel elements are thus scaled back to 25% in order to determine doses to occupational personnel and members of the general public.

NUREG analysis assumes 50 elements were present in the referenced core and the central element experiences a greater than average burn up. At 1/50 or 2% of the total, the central element would contain 4% of the 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 NUREG example) for 365 days is 91.25 MWd and the corresponding activities would be 957.2 Ci of Krypton, 2357.8 Ci of Iodine and 983.3 Ci of Xenon. The MIUTR contains 93 fueled elements whereas the NUREG/General Atomics example uses 50 elements in its assessment. Therefore, the activity per MUTR element would be decreased by a factor 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 to retain the fission products. According to NUREG 2387 the gap activity fraction is approximately 1.5x10-5.

Isotope Released Activities (mCi)

Isotope Released Activities (mCi)

Kr-83 m 0.2492 1-134 5.1290 Kr-85m 0.5782 1-135 4.4621 Kr-85 0.0097 Sr-89 2.0161 Kr-87 1.1129 Sr-90 0.0625 Kr-88 1.5911 Sr-91 2.6048 Xe-133m 0.0782 Sr-92 2.9516 Xe-133 4.5702 Cs-134 0.0060 Xe-135m 1.2056 Cs-134m 0.00363 Xe-135 2.0669 Cs-136 0.0524 1-131 2.1774 Cs-137 1.0000 1-132 3.3492 Cs-138 4.153 1-133 3.8976 In addition to using MUITR release activities, allowance is taken for air leakage, radionuclide decay and shielding over the course of the event. Air leakage rates were determined using engineering analysis of the M\\UTR. Decay rates were calculated from the Chart of the Nuclides as well as the Health Physics and Radiological Health Handbook (HTPRRH). The reduction in shine dose due to shielding was determined from 2

Accident Analysis MHA figure 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 to occupational personnel and members of the public. Doses to the public are from ground level release due to air leakage from the south side of the reactor building. Horizontal and vertical diffusion coefficients, where applicable, 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 are extrapolated. A Pasquill category F, moderately stable condition, was chosen for all releases as a conservative category.

3

Accident Analysis MHA Methodology for Occupational Dose Calculations The following are the formulae used to calculate the occupational doses:

Committed Dose Equivalent (CDE) to the thyroid and CEDE for reactor occupational personnel CEDE = E

[.BR

• DCF1 nt

• Ai[1-exp(-aef, leak tst)))

Deep Dose Equivalent (DDE) to reactor occupational personnel Terms used in the above Dose Equations BR 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]

9*eff 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 [in 3]

tst Reactor personnel stay time (evacuation time) [s]

4

Accident Analysis MHA Methodology for Public Dose Calculations For the 10% of the time the air in the MUTR predominantly flows out of the southern side of the reactor due to atmospheric conditions. This scenario describes the MEl since during the other 90% of the time doses to any member 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

• Tstay Parameters in CEDE Equation BR - Breathing rate per NRC Guidance [in 3 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

• Tsa Parameters in DDE Equation DCFext - 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 Equation F-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 MHA Summary of Doses CEDE Occupational 10.2 mrem DDE Occupational 1.62 mrem TEDE Occupational 11.82 mrem CEDE public At MEl1-88.5 mrem DDE public At MEl1-2.78 mrem Shine Dose Public At ME1-7.325 mrem TEDE Public MEI - 98.605 mrem 6

UNIVERSIYO GLENN L. MARTIN INSTITUTE OF TECHNOLOGY A. JAMES CLARK SCHOOL OF ENGINEERING Department of Materials Science & Engineering Nuclear Reactor & Radiation Facilities Timothy W. Koeth, Director Building 090 College Park, Maryland 20742-2115 301.405.4952 TEL 301.405.6327 FAX 609.577.8790 CELL koethi@umd.edu February 29, 2016 Document Control Desk United States Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUBJECT:

UNIVERSITY OF MARYLAND - REQUEST FOR ADDITIONAL INFORMATION RE: FOR THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-70 THE MARYLAND UNIVERSITY TRAINING REACTOR DOCKET NO. 50-166 Enclosed please find the response to the RAI dated August 24, 2015 for the University of Maryland 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 Director University of Maryland Training Reactor & Radiation Facilities 10/p

Response To:

OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION FOR THE RENEWAL OF FACILITY OPERATING LICENSE NO. R-70 THE MARYLAND UNIVERSITY TRAINING REACTOR DOCKET NO. 50-166

1. MUTR SAR, Section 4.5.2, "Reactor Core Physics Parameters," (Ref. 1) lists three reactivity coefficients and 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-Power Reactors Standard Review Plan and Acceptance Criteria," Section 4.5.2 provides guidance that an analysis should show that reactivity coefficients are sufficiently negative to prevent or mitigate damaging reactor transients. Describe what constitutes a power coefficient and show how overall reactivity coefficients are negative; or justify why the current method is acceptable.

Fuel Temperature Coefficient:

-1 1.2¢/ 0C Moderator Temperature Coefficient:

+3.0 ¢/°C Reactor Power Coefficient:

-0.53 ¢!kW Listed above are the reactivity coefficients of MUIR. At first glance it may seem as though the sum coefficient has a positive value. Howvever, the sole positive contribution to the reactivity is from the moderator temperature. The moderator temperature increases very slowly in comparison to the fuel temperature 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 already significant negative reactivity added before any positive reactivity results from an increase in moderator temperature.

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 for Preparing 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 a minimum value of 2. Provide a DNBR analysis that indicates a minimum value of at least 2, or justify why one is not needed.

A conservative calculation of the DNBR for MUTR has returned a value of 2.96 while operating at 600kW and inlet temperature of 92 degrees Celsius. Support documentation for this DNBR is attached in document 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 will have no negative impact on the environment. The MUTR environmental monitoring program results were provided in response to RAIs No. 47 and No. 72 (Refs. 6 and 2, respectively). However, the results are from 2004, and therefore, are out of date. NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content, Section 11.1.7 provides guidance that an appropriate monitoring program should contain probable pathways to people, and trends of recorded results. Provide updated information on the environmental monitoring program or justify why it is not needed.

Fixed environmental area monitors are located at the MUTR restricted area boundary and at locations on the university campus, to record and tract the potential radiological impact the MUTR operations have on the surrounding environment.

Monitors are exchanged and analyzed at frequency not to exceed once per calendar quarter. Records are maintained in accordance with 10 CFR 20.2103. Historically, and over the past 5 years, dose determinations for members of the public based upon this program, indicate doses to the public are in compliance with the limits of 20 CFR 20.1301.

In addition to fixed environmental area monitors, exposure rate Geiger Muller measurements are taken monthly in unrestricted areas outside of the MUTR boundary to monitor potential exposure to the public 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, and determination 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 the course of events and mitigate consequences (including automatic functions and operator actions) until a stabilized condition has been reached should be described. The accident analysis appears to be limited to uniform mixing of fission products in the reactor room and subsequent elevated 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 separate scenarios are analyzed; (iv) what the sequence for the analyzed exposure times (question 4(e)ii of this document); or (v) when a stable condition would be reached. Provide an updated analysis describing the sequence of events including initiation of engineered safety features to mitigate an 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 and assumptions developed for the "Radiation Protection Program and Waste Management,"

chapter of the SAR should be adapted as appropriate for the analysis. Submitted information should allow the results to be independently verified. The following parameters require further clarification:

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 per second (page 1, Ref. 3), and room leakage parameter of 0.00236 meters cubed per second (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 (pages 4, 8, and 12, Ref. 3) is inconsistent with the breathing rate of 4.27x20-04 meters cubed per 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 are provided as input parameters for "HOTSPOT" (page 16, Ref. 3). However, dispersion values for various distances and atmospheric stability classes (page 3, Ref. 3) cannot be verified using these input parameters. Provide an updated analysis clearly stating confinement leakage, breathing rates, release heights, and wind speed parameters as necessary before each series of computations, or justify why the current method is acceptable.

See Attachment Titled "Accident Analysis MHA" c) Guidance in NUREG-1537, Section 13.2, item (6) provides for defining the source term quantity of radionuclides. The fission product inventory is 25 percent equivalent of those described in NUREG/CR-2387 (page 2, Ref. 3). It appears the activities of Cesium and Strontium are less than 25 percent of those values listed in NUREG/CR-2387. Provide an updated analysis using consistent methodology for determining the source term, or justify why the current method is acceptable.

See Attachment Titled "Accident Analysis MHA" d) Guidance in NUREG-1537, Section 13.2, item (6) provides for describing a source term that could 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 to members 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 not considered. NUREG-1537, Section 13.2, item (7) provides guidance for evaluating exposure of a member of the public until the situation is terminated or the person is moved. Provide an updated analysis to include direct or scattered radiation exposure to members of the public specific 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 of the public specific to the facility until the situation are stabilized. The following locations for members of the public and times of exposure require further clarification:

i. Potential radiological consequences to members of the public in unrestricted areas are evaluated 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-campus residence hall and nearest off campus public residence from the reactor building at approximately 230 and 370 meters, respectively. A maximum exposed members of the public 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 for other locations of interest that may be applicable to the MUTR facility is provided in NUREG-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 a receptor 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 clear how to chronologically view the events, or if exposure times are consistent with one another. Provide an updated analysis clearly indicating exposure times and subsequent dose 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 the current method is acceptable.

See Attachment Titled "Accident Analysis MHA"

5. MUTR proposed TS 3.1, "Reactor Core Parameters," Specification (5) describes reactivity coefficients at the MUTR (Ref. 7). NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors Format and Content, Section 14, Appendix 14.1, Section 4 provides guidance that certain limiting conditions for operations have accompanying surveillance requirements to include test, method, frequency, and acceptability. It appears the reactivity coefficients do not have a surveillance requirement. Provide a surveillance specification for TS 3.1 Specification (5), or justify why one is not necessary.

See updated TS 3.1

6. MUTR proposed TS 3.7, "Limitations On Experiments," Specification (4) describes limits on experiments (Ref. 7). Specification (4) describes explosive materials in quantities greater than 25 milligrams and less than 25 milligrams, but does not include quantities equal to 25 milligrams. Provide a revised TS 3.7 Specification (4) to provide for explosive material quantities equal to 25 milligrams, or justify why no change is necessary.

See updated TS 3.7

7. MUTR proposed TS 4.1, "Reactor Core Parameters," Specification (5) describes annual inspections of fuel 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 in the 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 Format and Content, Section 14, Appendix 14.1, Section 1.2.2 provides guidance that the proposed TS basis should be reference to the facility's analysis. Provide a revision to proposed TS 4.4 to include a qualitative 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 is necessary.

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 and Reviewing 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's analysis. 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 is necessary.

See updated 1S 5.2

10. 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 for Preparing 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 for restricting certain fuel bundles from core positions so that assumptions used in the development safety limits are met. NUREG-1537, Section 14, Appendix 14.1, Section 1.2.2 provides guidance that a TS should include a basis for each specification. Propose a TS including a basis that incorporates acceptable IFE locations, or justify why no change is necessary.

See updated TS 2.2

11. 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, "MUIR Organizational Structure," show solid-line and dashed-line connections, but appear to be missing 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 a description of the connection lines in Figures 6.1 and 6.2.

See updated TS Figures 6.1 & 6.2 b) Figure 6.2, "MUIR Organizational Structure," shows members of the MUIR organization including staff and management. However, the TSs do not appear to correlate the MUIR members 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 provided in ANSl/ANS-25.4-2007, Section 3. Clarify the level of assignment in the TSs for the members shown in Figure 6.2.

See updated TS Figure 6.2 c) ANSl/ANS-15.1-2007, Section 6.1.2 provides guidance that management not only be responsible for policies and operation, but shall also adhere to all requirements of the operating license and TSs. MUIR proposed TS 6.1.2, "Responsibility," describes specific responsibilities for the 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 MUTR member shown in Figure 6.2.

TS 6.1.2 has been rewritten to include responsibilities of all MUTR members shown in Figure 6.2

12. MUTR proposed TS 6.1.3, "Facility Staff Requirements," Specification (1) describes facility staffing requirements when the "reactor is operating" (Ref. 7). However, ANSI/ANS-15.1-2007, Section 6.1.3 provides 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 of radioactivity," (Ref. 7). However, "new experiment," is not defined, nor is the terminology consistent with the MUTR proposed TS Definition 1.7 or Specification 6.5. It is not clear which category of experiments are applicable in proposed TS 6.2.1.2. Provide a revised TS 6.2.1.2 to delineate which experiments 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 of an 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 are considered new.

14. MUTR proposed TS 6.5, "Experiment Review And Approval," Specification (3) uses the term "desired alternate," 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 a revision to the proposed TS 6.5 or justify why no change is necessary.

See updated TS 6.5

15. 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." It appears Definition 1.27 is erroneously used in proposed TS 6.7.2. Provide a revision to proposed TS 6.7.2 or justify why no change is necessary.

See updated TS 6.7.2

16. The following typographical errors were noticed. Consider reviewing the proposed TSs for other typographical or formatting errors and propose corrections as necessary.

a)

MUTR proposed TS Definition 1.37 may contain a grammatical error, See updated TS 1.37 b) MUTR proposed TS Definition 1.41 is numbered as 1.401, See updated TS 1.41 c) MUTR proposed TS 4.4 Specification may contain a grammatical error,

See updated TS 4.4 d) MUTR proposed TS 5.2.1 Specification (1) appears to erroneously use "connective,"

See updated TS 5.2 e) MUTR proposed TS 5.3.1 Specification (4) appears to be missing, See updated TS 5.3.1 f) MUTR proposed TS 5.3.2 Specification (1) states the control rods will contain borated graphite BvC, and See updated TS 5.3.2 g) MUTR proposed TS 5.4 Specification (3) appears to be missing.

See updated TS 5.4 OTHER 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 SPECIFICATIONS FOR THE MARYLAND UNIVERSITY TRAINING REACTOR License Number R-70 Docket Number 50-166 Submitted to United States Nuclear Regulatory Commission 29 February 2016 (Superseding 27 September 2011 Submission)

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16Rev02292016.docx Last edit February 29, 2016 TABLE OF CONTENTS TABLE OF CONTENTS

.i LIST OF TABLES

.iv LIST OF FIGURES.

.v 1.0 DEFINITIONS

.1 1.1 ALARA

.1 1.2 Channel

.1 1.3 Confinement

.1 1.4 Control Rod Guide Tube

.1 1.5 Core Configuration

.1 1.6 Excess Reactivity

.1 1.7 Experiment

.1 1.8 Experimental Facilities

.2 1.9 Experiment Safety Systems

.2 1.10 Four Element Fuel Bundle

.2 1.11 Fuel Element

.2 1.12 Fueled Device.

.2 1.13 Full Power

.2 1.14 Instrumented Element.

.2 1.15 Isolation

.2 1.16 Limiting Conditions for Operation

.2 1.17 Limiting Safety System Setting

.2 1.18 Measuring Channel

.2 1.19 Measured Value

.2 1.20 Moveable Experiment.

.2 1.21 On Call

.3 1.22 Operable

.3 1.23 Operating

.3 1.24 Reactivity Worth of an Experiment

.3 1.25 Reactor Console Secured

.3 1.26 Reactor Operating

.3 1.27 Reactor Operator

.3 1.28 Reactor Safety Systems

.3 1.29 Reactor Secured

.3 1.30 Reactor Shutdown

.3 1.31 Reference Core Condition

.4 1.32 Reportable Occurrence

.4 1.33 Rod-Control

.4 1.34 Safety Channel

.4 1.35 Safety Limit

.4

O:\\M UTR\\2016M UTRLivingDocs\\WorlingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 1.36 Scram Time

.4 1.37 Secured Experiment

.4 1.38 Secured Shutdown

.5 1.39 Senior Reactor Operator

.5 1.40 Shall, Should, May

.5 1.41 Shutdown Margin

.5 1.42 Shutdown Reactivity.

5 1.43 Standard Core.

.5 1.44 Steady State Mode

.5 1.45 Three Element Fuel Bundle

.5 1.46 True Value

.5 1.47 Unscheduled Shutdown 5

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 6

2.1 Safety Limit

.6 2.2 Limiting Safety System Settings.

.6 3.0 LIMITING CONDITIONS FOR OPERATION.

8 3.1 Reactor Core Parameters.

.8 3.2 Reactor Control and Safety Systems.

.9 3.3 Primary Coolant System.

.13 3.4 Confinement

.14 3.5 Ventilation Systems

.14 3.6 Radiation Monitoring System and Effluents 15 3.6.1 Radiation Monitoring System.

15 3.6.2 Effluents.

.16 3.7 Limitations on Experiments 17 4.0 SURVEILLANCE REQUIREMENTS 19 4.1 Reactor Core Parameters 19 4.2 Reactor Control and Safety Systems.

20 4.3 Primary Coolant System 21 4.4 Confinement 22 4.5 Ventilation System 22 4.6 Radiation Monitoring System and Effluents 23 4.6.1 Radiation Monitoring System.

23 4.6.2 Effluents

.23 4.7 Experiments

.24 5.0 DESIGN FEATURES 25 5.1 Site Characteristics 25 5.2 Reactor Coolant System 25 ii

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 5.3 Reactor Core and Fuel 26 5.3.1 Reactor Fuel

.27 5.3.2 Control Rods.

27 5.4 Fissionable Material Storage 28 6.0 ADMINISTRATION

.29 6.1 Organization

.29 6.1.1 Structure 29 6.1.2 Responsibility.

29 6.1.3 Facility Staff Requirements 29 6.1.4 Selection and Training of Personnel 33 6.2 Review and Audit 33 6.2.1 Reactor Safety Committee 33 6.2.1.1 Reactor Safety Committee Charter and Rules 33 6.2.1.2 Reactor Safety Committee Review Function 34 6.2.1.3 Reactor Safety Committee Audit Function 34 6.2.2 Audit of ALARA Program 35 6.3 Radiation Safety

.35 6.4 Operating Procedures.

35 6.5 Experiment Review and Approval 36 6.6 Required Actions 36 6.6.1 Actions to be Taken in Case of Safety Limit Violation 36 6.6.2 Actions to be Taken in the Event of a Reportable Occurrence 37 6.7 Reports.

.37 6.7.1 Annual Operating Report 37 6.7.2 Special Reports

.38 6.7.3 Unusual Event Report 39 6.8 Records

.39 3

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 LIST OF TABLES Table 3.1 Reactor Safety Channels: Scram Channels 11 Table 3.2 Reactor Safety Channels: Interlocks 11 Table 3.3 Reactor Safety Channels: Scram Channel Bases.

12 Table 3.4 Reactor Safety Channels: Interlock Bases 12 Table 3.5 Minimum Radiation Monitoring Channels 16 4

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 LIST OF FIGURES Figure 6.1 MUTR Position in University of Maryland Structure 30 Figure 6.1 MUTR Organizational Structure 31 5

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Included in this document are the Technical Specifications and the "Bases" for the Technical Specifications. These bases, which provide the technical support for the individual technical specifications, are included for information purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

1.0 DEFINITIONS 1.1 ALARA.(acronym for "as low as is reasonably achievable") means making every reasonable effort to maintain exposures to radiation as far below the dose limits in 10 CFR Part 20 as is practical consistent with the purpose for which the licensed activity is undertaken, taking into account 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 other societal and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed materials in the public interest.

1.2 CHANNEL - A channel is the combination of sensors, lines, amplifiers, and output devices which are 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 its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, 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 by observation of channel behavior, or by comparison of the channel with other independent channels 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 is operable.

1.3 CONFINEMENT - Confinement means a closure on the overall facility that controls the movement 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 fuel elements, arranged in a rectangular array with one bundle displaced for the pneumatic experimental system; three control rods; and two graphite reflectors.

1.6 EXCESS REACTIVITY - Excess reactivity is that amount of reactivity that would exist if all control rods were moved to the maximum reactive condition from the point where the reactor is exactly 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 for irradiation within the pool, on or in a beamport or irradiation facility, and that is not rigidly secured 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 performed in the course of the reactor program.

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2. Modified Routine Experiments - Modified routine experiments are those which have not been performed previously but are similar to routine experiments in that the hazards are neither greater nor significantly different than those for the corresponding routine experiments.

3.

Special experiments - Special experiments are those which are not routine or modified routine experiments.

1.8 EXPERIMENTAL FACILITIES - Experimental facilities are facilities used to perform experiments 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, including their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated.

1.10 FOUR ELEMENT FUEL BUNDLE - The 4-element fuel bundle consists of an aluminum bottom, 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 a sheathed chromel-alumel or equivalent thermocouple is embedded in the fuel.

1.15 ISOLATION - Isolation is the establishment of confinement, closing of the doors leading from the reactor bay area leading into the balcony area on the top floor, the door to the reception area on the ground floor, and the building exterior doors.

1.16 LIMITING CONDITIONS FOR OPERATION - Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

1.17 LIMITING SAFETY SYSTEM SETTING-Limiting safety system settings (LSSS) for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.

1.18 MEASURING CHANNEL - A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device, which are connected for the purpose of measuring the value of a variable.

1.19 MEASURED VALUE - The measured value is the value of a parameter as it appears on the output of a channel.

1.20 MOVEABLE EXPERIMENT - A movable experiment is one where it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.

2

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTech nicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 1.21 ON CALL - A senior operator is available "on call" if the senior operator is either on the College Park campus or within 10 miles from the facility and can reach the facility within one half hour following a request.

1.22 OPERABLE - Operable means a component or system is capable of performing its intended function.

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 the Value of the reactivity change that results from the experiment being inserted into or removed from 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 the console.

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 their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action. Manual protective action is considered 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 is insufficient fissile material present in the reactor to attain criticality under optimum available conditions of moderator and reflection, or

2. 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, or control rod drives unless they are physically decoupled from the control rods; and

d. No experiments in or near the reactor are being moved or serviced that have, on movement, the smaller of: a reactivity worth exceeding the maximum value allowed for 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 the reference core condition with the reactivity worth of all installed experiments included and the following conditions exist:

a. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control rods; 3

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b. No experiments are being moved or serviced that have, on movement, a reactivity worth 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 of the core when it is at 20 °C and the reactivity worth of xenon is zero (i.e., cold, clean, and critical).

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 the Limiting Safety-System Settings specified in technical specifications 2.2.

2.

Operation in violation of the Limiting Conditions for Operation established in the technical specifications.

3.

A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests. (Note: Where components or systems are provided in addition to those required by the Technical Specifications, the failure of the extra components or systems is not considered reportable provided that the minimum number of components 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 such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.

1.33 ROD-CONTROL - A control rod is a device fabricated from neutron absorbing material which is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod may be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

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 be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity.

1.36 SCRAM TIME - Scram time is the elapsed time between the initiation of a scram signal by either automated or operator initiated action and the time required for the control rods to reach a fully inserted position into the core.

1.37 SECURED EXPERIMENT - A secured experiment is any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces that can arise as a result of 4

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 credible malfunctions.

1.38 SECURED SHUTDOWN - Secured shutdown is achieved when the reactor meets the requirements of the definition of "reactor secured" and the facility administrative requirements for leaving the facility with no licensed reactor operators present.

1.39 SENIOR REACTOR OPERATOR - A senior reactor operator (SRO) is an individual who is licensed 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,,shouldee is used to denote a recommendation; and the word,,may" is used to denote permission, neither a requirement nor a recommendation.

1.41 SHUTDOWN MARGIN - Shutdown margin is the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems starting from any permissible operation condition and with the most reactive rod in its most 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 reactor with all control rods in their least reactive position (e.g., inserted). The value of shutdown reactivity includes the reactivity value of all installed experiments and is determined with the reactor at ambient conditions.

1.43 STANDARD CORE - A standard core is an arrangement of standard TRIGA fuel in the reactor grid plate.

1.44 STEADY STATE MODE - Steady state mode operation shall mean operation of the reactor with the mode selector switch in the STEADY STATE position.

1.45 THREE ELEMENT FUEL BUNDLE - The 3-element fuel bundle consists of an aluminum bottom, 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 unplanned shutdown 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 adversely affect safe operation, not to include shutdowns which occur during testing or check-out operations.

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O :\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 2.0 SAFETY LIMITS AN]) LIMITING SAFETY SYSTEM SETTING 2.1 SAFETY LIMIT Applicability This specification applies to the temperature of the reactor fuel.

Objective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding will result.

Specification The temperature in a standard TRIGA fuel element shall not exceed 1000 °C under any conditions of operation, with the fuel fully immersed in water.

Basis The important parameter for TRIGA reactor is the UZrH fuel element temperature. This parameter is well suited as a single specification especially since it can be measured. A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the fuel-moderator. The magnitude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium. The data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of ZrHx will remain below the ultimate stress if the temperature in the fuel does not exceed 1000 °C and the fuel cladding is water-cooled.

It has been shown by experience that operation of TRIGA reactors at a power level of 1000 kW will not result in damage to the fuel. Several reactors of this type have operated successfully for several years at power levels up to 1500 kW. Analysis and measurements on other TRIGA reactors have shown that a power level of 1000 kW corresponds to a peak fuel temperature of approximately 400 °C.

2.2 LIMITIN4G SAFETY SYSTEM SETTINGS Applicability This specification applies to the reactor scram setting that prevents the reactor fuel temperature from reaching the safety limit.

Objective The objective is to provide a reactor scram to prevent the safety limit (fuel element temperature of 1000 °C) from being reached.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Specification The 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 is placed in any position other than D8 in the grid plate, an analysis must be performed. The analysis shall indicate that the proposed location shall have a peaking factor not less than 50% of the highest fuel element in the core.

Basis A Limiting Safety Setting of 175 °C provides a safety margin of 650 °C. A part of the safety margin is used to account for the difference between the temperature at the hot spot in the fuel and the measured temperature resulting from the actual location of the thermocouple. If the thermocouple element is located in the hottest position in the core, the difference between the true and measured temperatures will be only a few degrees since the thermocouple junction is at the mid-plane of the element and close to the anticipated hot spot. If the thermocouple element is located in a region of lower temperature, such as on the periphery of the core, the measured temperature 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 the core, the true temperature at the hottest location in the core will differ from the measured temperature by no more than a factor of two. Thus, with the WFE positioned in the location specified by the license, when the temperature in the thermocouple element reaches the setting of 175 °C, the true temperature at the hottest location would be no greater than 350 °C, providing a margin to the safety limit of at least 650 °C. This margin is ample to account for the remaining uncertainty in the accuracy of the fuel temperature, measurement channel, and any overshoot in reactor power resulting from a reactor transient during steady state mode operation.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR CORE PARAMETERS Applicability These specifications shall apply to the reactor at all times it is operating.

Objective The objectives are to ensure that the reactor can be controlled and shut down at all times and that the safety limits will not be exceeded.

Specifications

1. The excess reactivity relative to the cold critical conditions, with or without experiments in 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; and

b.

Total worth of all in-core experiments in their most reactive state; and

c.

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 by more than the worth of the most reactive fuel element.

c.

No control rods shall be removed from the core unless a minimum of four fuel bundles 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 product release into the reactor coolant) except to locate such fuel.

5. The reactivity coefficients for the reactor are:

Fuel:

-1.2 ¢/°0C Moderator:

+3.0 ¢/°0C Power:

-0.53 ¢/kW The Fuel Temperature Coefficient, and the Moderator Temperature Coefficient, shall be verified any time the standard core is modified either by the rotation of a fuel bundle, a change in fuel bundle or reflector location, or the replacement of any fuel or reflector. The Power Temperature Coefficient shall be recalculated if either the Fuel Temperature or Moderator Temperature Coefficient are measured to be more than 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.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Bases

1. While specification 3.1.1, in conjunction with specification 3.1.2, tends to over constrain the excess reactivity, it helps ensure that the operable core is similar to the core analyzed in the FSAR.

2.

The value of the shutdown margin as required by specification 3.1.2 assures that the reactor can be shutdown from any operating condition even if the highest worth control rod should remain in the fully withdrawn position.

3. Specification 3.1.3 ensures that the operable core is similar to the core analyzed in the FSAR. It also ensures that accidental criticality will not occur during fuel or control rod manipulations.

4. Specification 3.1.4 limits the fission product release that might accompany operation with a damaged fuel element. Fuel will be considered potentially "Damaged" if said fuel is found to be leaking under the air and/or water sampling or under such case that the fuel has been exposed to temperature above 175 °C as measured on the instrumented fuel element. The criteria of the water and air sampling to determine a leaking fuel element is considered 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 is negative.

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 SYSTEMS Applicability These specifications apply to reactor control and safety systems and safety-related instrumentation that must be operable when the reactor is in operation.

Objective The objective of these specifications is to specify the lowest acceptable level of performance or the minimum number of operable components for the reactor control and safety systems.

Specifications

1. The drop time of each of the three standard control rods from the fully withdrawn position to the fully inserted position shall not exceed one second.

2. Maximum positive reactivity insertion rate by control rod motion shall not exceed $0.30 per second.

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3. The reactor safety channels shall be operable in accordance with Table 3.1l, including the minimum number of channels and the indicated maximum or minimum set points for the scram channels.

4. The safety interlocks shall be operable in accordance with Table 3.2, including the minimum number of interlocks.

5. The Beam Port and Through Tube interlocks may be bypassed during a reactor operation with the permission of the Reactor Director.

6.

A minimum of one reactor power channel, calibrated for reactor thermal power, must be attached to a recording device sufficient for auditing of reactor operation history.

Bases

1. Specification 3.2.1 assures that the reactor will be shutdown promptly when a scram signal is initiated. Experiments and analysis have indicated that for the range of transients anticipated for the MUTR TRIGA reactor, the specified control rod drop time is 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 the normally available reactivity and insertion, rate cannot generate operating conditions that exceed the Safety Limit. (See FSAR)

3.

Specification 3.2.3 provides protection against the reactor operating outside of the safety limits. 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 safety limits. 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 intended purpose. This purpose is to assure that the operator is aware of the status of both the beam ports and the through tube.

6.

Specification 3.2.6 provides for a means to monitor reactor operations and verify that the reactor is not operated outside of its license condition.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Table 3.1: Reactor Safety Channels: Scram Channels Scram Channel Minimum Required Operable 2

Scram Setpoint Reactor Power Level Not to exceed 120 %

Fuel Element Temperature Reactor Power Channel Detector Power Supply 1

2 Manual Scram Console Electrical Supply Rate of power change - Period Radiation Area Monitors 1

1 1

1 Not to exceed 175 °C Loss of power supply voltage to chamber N/A Loss of electrical power to the control console Not less than 5 seconds 50 mr/hr (bridge monitor) 10 mr/hr (exhaust monitor)

Table 3.2: Reactor Safety Channels: Interlocks Interlock/Channel Log Power Level Startup Count rate Safety 1 Trip Test Plug Electrical Connection Rod Drive Control Function Provide signal to period rate and minimum source channels. Prevent control rod withdrawal when neutron count rate is less than l cps.

Prevent control rod withdrawal when neutron count rate is less than 1 cps.

Prevent control rod withdrawal when Safety 1 Trip Test switch is activated.

Disable magnet power when Beam Port or Through Tube plug is removed unless bypass has been activated.

Prevent simultaneous manual withdrawal of two or more control rods in the steady state mode of operation.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Table 3.3: Reactor Safety Channels: Scram Channel Bases Scram Channel Bases Reactor Power Level Fuel Element Temperature Reactor Power Channel Detector Power Supply Manual Scram Console Electric Supply Rate of power change -

Period Radiation Area Monitors Provides protection to assure that the reactor can be shut down before the safety limit on the fuel element temperature will be exceeded.

Provides protection to assure that the reactor cannot be operated unless the neutron detectors that input to each of the linear power channels are operable.

Allows the operator to shut down the reactor if an unsafe or abnormal condition occurs.

Assures that the reactor cannot be operated without a secure electric supply.

Assures that the reactor is operated in a manner that allows the operator time to shut down the reactor before the licensed power restriction is exceeded.

Assures that the reactor automatically scrams if a high airborne radiation level is detected.

Table 3.4: Reactor Safety Channels: Interlock Bases Interlock/Channel Log Power Level Startup Count rate Safety 1 Trip Test Plug Electrical Connection Rod Drive Control Bases This channel is required to provide a neutron detector input signal to the startup count rate channel.

Assures sufficient amount of startup neutrons are available to achieve a controlled approach to criticality.

Assures that the 1 cps interlock cannot be bypassed by creating an artificial 1 cps signal with the Safety 1 trip test switch Assures that the reactor cannot be operated with Beamport or Through Tube plugs removed without further precautions.

Limits the maximum positive reactivity insertion rate available for steady state operation.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 3.3 PRIMARY COOLANT SYSTEM Applicability_

This specification applies to the quality and quantity of the primary coolant in contact with the fuel cladding at the time of reactor startup.

Objectives

1. 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.

Specifications

1. 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 shall be 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 located in the pool.

Bases

1. Specification 3.3.1 ensures that both sufficient cooling capability and sufficient biological shielding are available for safe reactor operation.

2. A small rate of corrosion continuously occurs in a water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limit provides acceptable control. In addition, by limiting the concentration of dissolved materials in the water, the radioactivity of neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel exposures during maintenance and operation.

3. Specification 3.3.3 ensures that a fuel failure with release of radioactive materials into the pool will be determined.

4. Specification 3.3.4 ensures a DNBR value greater than 2.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 3.4 CONFINEMENT

_Applicability This specification applies to that part of the facility that contains the reactor, its controls and shielding.

Objective The objective of these specifications is to ensure that sufficient confinement volume is available for the dilution of radioactive releases and to limit the rate of release of radioactive material to the outside environment.

_Specifications

1. Confinement shall be considered established when the doors leading from the reactor bay area leading into the balcony area on the top floor, and the reception area as well as the building exterior are secured.

2.

Confinement shall be established whenever the reactor is in an unsecured mode with the exception of the time that persons are physically entering or leaving the confinement area.

Bases

1. This specification provides the necessary requirements for confinement, which ensures releases to the outside environment are within 10 CFR Part 20 requirements.

2. This specification provides the reactor status condition for confimement, as well as allows personnel to enter and leave the reactor building, as required, when the reactor is unsecured.

3.5 VENTILATJON SYSTEMS Applicability These specifications apply to the ventilation systems for the reactor building.

_Objective The objective of these specifications is to ensure that air exchanges between the reactor confinement building and the environment do not impact negatively on the general public.

_Specifications

1. Air within the reactor building shall not be exchanged with other occupied spaces in the building.

2. All locations where ventilation systems exchange air with the environment shall have failsafe closure mechanisms.

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3. Forced air ventilation to the outside shall automatically secure without operator intervention in such case that the radiation levels exceed a preset level as defined in facility procedures. The setpoints are: 50 mR/hr (bridge monitor), 10 mR/hr (exhaust monitor).

Bases

1. This specification ensures that radioactive releases inside the reactor building will not be transported to the remainder of the building.

2.

This specification ensures that the reactor building can always be isolated from the environment.

3. This specification ensures that radioactive release will be minimized by stopping forced flow to the outside environment.

3.6 _

RADIATION MONITORING SYSTEM ANT) EFFLUENTS 3.6.1 Radiation Monitoring System Applicability This specification applies to the radiation monitoring information that must be available to the reactor operator during reactor operation.

Objective The objective is to assure that sufficient radiation monitoring information is available to the operator to assure safe operation of the reactor.

Specifications

1. The reactor shall not be operated unless a minimum of one of the two radiation area monitor 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 radiation monitor channels, the intent of specification 3.6.1 will be satisfied if they are replaced with portable gamma sensitive instruments having their own alarms or which shall be observable by the reactor operator.

3. The alarm set points shall be stated in a facility operating procedure. The alarm setpoints for the bridge monitor are:.37 mR/hr (alert), 50 mR/hr (scram). The setpoints for the exhaust monitor are: 8 mR/hr (alert), 10 mR/hr (scram).

4.

The campus radiation safety organization shall maintain an environmental monitor at the MUJTR site boundary.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Table 3.5: Minimum Radiation Monitoring Channels Radiation Area Monitors Exhaust Radiation Monitor Bridge Radiation Monitor Function Minimum Number Operable Monitor radiation levels in Reactor Bay area at an Exhaust Fan location Monitor radiation levels in Reactor Bay area at the Reactor Bridge location A minimum of 1 of the 2 monitors shall be operable Bases

1. Specification 3.6.1.1 ensures that a significant fuel failure with release of radioactive materials will be determined and that any large releases will be mitigated by the specified protective actions.

2.

Specification 3.6.1.2 allows for continued reactor operation if maintenance and/or calibration of the radiation area monitors is required.

3. The alarm and scram set points shall be designed to ensure that dose rates delivered to areas accessible to members of the general public do not exceed the levels defined in 10 CFR Part 20. Additionally, the radiation area monitors provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of 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 a dose to the general public greater than that allowed by 10 CFR Part 20.

3.6.2 Effluents Applicability.

This specification applies to limits on effluent release.

Objective The objective is to ensure that the release of radioactive materials from the reactor facility to unrestricted areas do not exceed federal regulations.

Specification All effluents from the MUTR shall conform to the standards set forth in 10 CFR Part 20.

Basis The intent of 3.6.2 is to ensure that, in the event that radioactive effluents are released, the dose to the general public will be less than that allowed by 10 CFR Part 20.

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O :\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 3.7 LIMITATIONS ON EXPERIMENTS Applicability The specification applies to experiments installed in the reactor and its experimental facilities.

Objective The objective is to prevent damage to the reactor or excessive release of radioactive material in the event of an experiment failure.

Specifications The 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 and experiment flooding or voiding.

3. Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials, and liquid fissionable materials shall be doubly encapsulated.

4. Explosive materials in quantities greater than 25 mg TNT or its equivalent shall not be irradiated in the reactor or experimental facilities. Explosive materials in quantities equal to or less than 25 mg may be irradiated provided the pressture produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the failure pressure of the container. The failure pressure of the container is one half of the design pressure. Total explosive material inventory in the reactor facility may not exceed 100 mg TNT or its equivalent.

5. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor or (3) possible accident conditions in the experiment shall be limited in type and quantity such that if 100 % of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne radioactivity in the reactor room or outside environment will not result in exceeding 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, which closes automatically on high radiation level, at least 10 % of the gaseous activity or aerosols produced will escape.

b. If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99 % efficiency for 0.3 *na particles, at least 10 % of these particles can escape.

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c.

If an experiment fails and releases radioactive gases or aerosols to the reactor bay or atmosphere, 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 (540 C), the vapors of at least 10 percent of the materials escape through an undisturbed column of water above the core.

6.

Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 5 mCi.

Bases

1. This specification is intended to provide assurance that the worth of a single unsecured experiment will be limited to a value such that the safety limit will not be exceeded if the positive 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 cold critical reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its inadvertent removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

The maximum worth of all experiments is also limited to a reactivity value such that the cold reactor will not achieve a power level high enough to exceed the core temperature safety limit if the experiments were removed inadvertently.

3.

This specification is intended to prevent damage to reactor components resulting from experiment failure. If an experiment fails, inspection of reactor structures and components 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 from failure of an experiment involving explosive materials, especially the accidental detonation of the explosive. If an experiment fails, inspection of reactor structures and components 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 of the limits of Table 2 of Appendix B of 10 CFR Part 20 will be released to the atmosphere outside the facility boundary.

6.

The 5 mCi limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less than that allowed by 10 CFR Part 20 for an unrestricted area. (See SAR) 18

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 4.0 SURVEILLANCE REQUIREMENTS INTRODUCTION Surveillances shall be performed on a timely basis as defined in the individual procedures governing the performance of the surveillance. In the event that the reactor is not in an operable condition, such as during periods of refueling, or replacement or repair of safety equipment, surveillances may be postponed until such time that the reactor is operable. In such case that any surveillance must be postponed, a written directive signed by the Facility Director, shall be placed in the records indicating the reason why and the expected completion date of the required surveillance. This directive shall be written before the date that the surveillance is due. Under no circumstance shall the reactor perform routine operations until such 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/component continues to meet performance requirements.

4.1 REACTOR CORE PARAMETERS Applicability These specifications apply to the surveillance requirements for the reactor core.

O~bjective The objective of these specifications is to ensure that the specifications of Section 3.1 are satisfied.

Specifications

1. 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 any removal 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 any removal or replacement of control rods

3. 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 to exceed 15 months, such that the entire core is inspected over a five year period.

6. Burnup shall be verified in the Annual Report.

Bases Experience has shown that the identified frequencies ensure performance and operability for each of these systems or components. For excess reactivity and shutdown margin, long-term changes are slow to develop. For fuel inspection, visually inspecting 20% of the bundles annually will identify any developing fuel integrity issues throughout the core.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 4.2 REACTOR CONTROL AND SAFETY SYSTEMS App~licability These specifications apply to the surveillance requirements for reactor control and safety systems.

Objective The objective of these specifications is to ensure that the specifications of Section 3.2 are satisfied.

Specifications

1. The reactivity worth of each standard control rod shall be determined annually, intervals not to exceed 15 months, and after each time the core fuel configuration is changed or a control rod is changed.

2.

The control rod withdrawal and insertion speeds shall be determined annually, intervals not to exceed 15 months, or whenever maintenance or repairs are made that could affect rod travel times.

3.

Control rod drop times shall be measured annually; intervals not to exceed 15 months, or whenever maintenance or repairs are made that could affect their drop time.

4. All scram channels and power measuring channels shall have a channel test, including trip actions with safety rod release and specified interlocks performed after each secured shutdown, before the first operation of the day, or prior to any operation scheduled to last more 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 channels shall be calibrated annually, intervals not to exceed 15 months.

5. Operability tests shall be performed on all affected safety and control systems after any maintenance is performed.

6. A channel calibration shall be made of the linear power level monitoring channels annually, intervals not to exceed 15 months.

7. A visual inspection of the control rod poison sections shall be made biennially, intervals not to exceed 28 months.

8. A visual inspection of the control rod drive and scram mechanisms shall be made annually, intervals not to exceed 15 months.

Bases

1. The reactivity worth of the control rods, specification 4.2.1, is measured to assure that the required shutdown margin is available and to provide a means to measure the reactivity worth of experiments. Long term effects of TRIGA reactor operation are such that measurements of the reactivity worths on an annual basis are adequate to insure that no significant changes in shutdown margin have occurred.

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2.

The control rod withdrawal and insertion rates, specification 4.2.2, are measured to insure that 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 can perform their safety function properly.

4.

The surveillance requirement specified in specification 4.2.4 for the reactor safety scram channels ensures that the overall functional capability is maintained.

5. The surveillance test performed after maintenance or repairs to the reactor safety system as required by specification 4.2.5 ensures that the affected channel will perform as intended.

6. The linear power level channel calibration specified in specification 4.2.6 will assure that the 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 made to evaluate corrosion and wear characteristics and any damage caused by operation in the reactor.

8. Specification 4.2.8 assures that a visual inspection of control drive mechanisms is made to evaluate corrosion and wear characteristics and any damage caused by operation in the reactor.

4.3 PRIMARY COOLANT SYSTEM Applicability These specifications apply to the surveillance requirements of the reactor primary coolant system.

Objective The objective of these specifications is to ensure the operability of the reactor primary coolant system as described in Section 3.3.

Specifications

1. The primary coolant level shall be verified before each reactor startup or daily during operations 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 pool water 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 exceed six weeks. If gross gamma activity is high (greater than twice historical data), gamma spectroscopy shall be performed.

4. Pool water temperature shall be measured prior to the reactor startup and shall be monitored during reactor operation.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Bases

1. Specification 4.3.1 ensures that sufficient water exists above the core to provide both sufficient cooling capacity and an adequate biological shield.

2.

Specification 4.3.2 ensures that poor pool water quality could not exist for long without being detected. Years of experience at the MIUTR have shown that pool water analysis on a monthly basis is adequate to detect degraded conditions of the pooi water in a timely manner.

3.

Gross gamma activity measurements are conducted to detect fission product releases from damaged fuel element cladding.

4.

Specification 4.3.4 ensures that the maximum allowable pool water temperature is not exceeded.

4.4 CONFINEMENT Applicability This specification applies to that part of the facility which contains the reactor, its controls and shielding.

Objective The objective of this specification is to ensure that radioactive releases from the confinement can be limited.

_Specification Prior to putting the reactor in an unsecured mode, the isolation of the confinement building shall be visually verified.

Bases This specification ensures that the minimal leakage rate assumed in the SAR is actually present during reactor operations in order to limit the release of radioactive material to the environs.

4.5 VENTTLATION SYSTEM Applicability This specification applies to the reactor ventilation system.

_Objective The objective is to assure that provisions are made to restrict the amount of radioactivity released to the environment.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Specification The ability to secure the ventilation system shall be verified before the first reactor operation of the day.

Bases The facility is designed such that in the event that excessive airborne radioactivity is detected the ventilation system shall be shutdown to minimize transport of airborne materials. Analysis indicates that in the event of a major fuel element failure personnel would have sufficient time to evacuate the facility before the maximum permissible dose (10 CFR Part 20) is exceeded.

4.6 RADIATION MONITORING SYSTEM ANT) EFFLUENTS 4.6.1 Radiation Monitoring System Applicability This specification applies to the surveillance requirements for the Radiation Area Monitoring System (RAMS).

Objective The objective of these specifications is to ensure the operability of each radiation area monitoring channel as required by Section 3.6 and to ensure that releases to the environment are kept below allowable limits.

Specifications

1. A channel calibration shall be made for each channel listed in Table 3.5 annually but at intervals not to exceed 15 months or whenever maintenance or repairs are made that could affect their calibration.

2. A channel test shall be made for each channel listed in Table 3.5 prior to starting up the reactor to ensure reactor scram, fan shutdown, and louver closing.

Bases Specifications 4.6.1.1 and 4.6.1.2 ensure that the various radiation area monitors are checked and calibrated on a routine basis, in order to assure compliance with 10 CFR Part 20.

4.6.2 Effluents Applicability This specification applies to the surveillance requirements for air and water effluents.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Objective The objective of these specifications is to that releases to the environment are kept below allowable limits.

Specifications

1. Reactor building air samples shall be counted for gross gamma activity monthly, intervals not to exceed 6 weeks.

2.

A sample of any water discharged from the reactor building sump shall be counted for gross gamma activity before its release to the environs.

Bases Specifications 4.6.2.1 and 4.6.2.2 ensure that the facility effluents comply with 10 CFR Part 20.

4.7 EXPERIMENTS Applicability This specification applies to the surveillance requirements for experiments installed in the reactor and its irradiation facilities.

Objective The objective of this specification is to prevent the conduct of experiments which may damage the reactor or release excessive amounts of radioactive materials as a result of experiment failure.

Specifications

1. The reactivity worth of an experiment shall be estimated or measured, as appropriate, before reactor operation with said experiment

2.

An experiment shall not be installed in the reactor or its irradiation facilities unless a safety analysis has been performed and reviewed for compliance with Section 3.7 by the Reactor Safety Committee (new experiment) or Facility Director (modified routine experiment), in full accord with Sections 6.1.2 and 6.2.1 of these Technical Specifications and the procedures which are established for this purpose.

Basis Experience has shown that experiments reviewed and approved by the Reactor Safety Committee or Facility Director can be conducted without endangering the safety of the reactor, personnel, or exceeding Technical Specification limits.

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O:\\M UTR\\2Oi6M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 5.0 DESIGN FEATURES 5.1 SITE CHARACTERISTICS Applicability This specification applies to the reactor facility and its site boundary.

Objective The objective is to assure that appropriate physical security is maintained for the reactor facility and the radioactive materials contained within it.

Specifications

1. The reactor shall be housed in a closed room designed to restrict leakage. The closed room does not include the West balcony area.

2.

The reactor site boundary shall consist of the outer walls of the reactor building and the area enclosed by the loading dock fence.

3. The restricted area shall consist of all areas interior to the reactor building including the west balcony and lower entryway.

4.

The controlled area shall consist of all areas interior to the reactor building including the west balcony and lower entryway.

Bases These specifications assure that appropriate control is maintained over access to the facility by members of the general public.

5.2 REACTOR PRIMARY COOLANT SYSTEM Applicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.

Objective The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding.

Specifications

1. 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.

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3.

The pooi water return (outlet) pipe shall not extend more than 50.8 cm (20 in) below the overflow outlet pipe when fuel is in the core.

Bases

1. Specification 5.2.1 is based on thermal and hydraulic calculations and operation of the MUTR that show that the core can operate in a safe manner at power levels up to 300 kW with natural convection flow of the coolant.

2.

Specifications 5.2.2 and 5.2.3 ensures that the pool water level can normally decrease only by 50.8 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 FUEL Applicability This specification applies to the configuration of the core and in-core experiments.

Objective The objective is to ensure that the core configuration is as specified in the license.

Specifications

1. The core shall consist of 93 TRIGA fuel elements assembled into 24 fuel bundles - 21 bundles shall contain four fuel elements and 3 bundles shall contain three fuel elements and a control rod guide tube.

2. The fuel bundles shall be arranged in a rectangular 4 x 6 configuration, with one bundle displaced 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 water Basis

1. 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 limits for the maximum temperature in the fuel.

4. The reflector reduces the neutron leakage from the reactor core.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 5.3.1 Reactor Fuel Applicability This specification applies to the fuel elements used in the reactor core.

Objective The objective is to assure that the fuel elements are of such design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics, and that the fuel used in the reactor has characteristics consistent with the fuel assumed in the SAR and the license.

Specifications The individual unirradiated standard TRIGA fuel elements shall have the following characteristics:

1. Uranium content: a maximum of 9.0 weight % uranium enriched to less than 20 % 235U

2.

Zirconium hydride atom ratio: nominal 1.5 - 1.8 hydrogen-to-zirconium, ZrHx

3. 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 15 inches.

Basis The design basis of the standard TRIGA fuel element demonstrates that 250 kW steady state operation presents a conservative limitation with respect to safety limits for the maximum temperature generated in the fuel.

5.3.2 Control Rods Applicability This specification applies to the control rods used in the reactor core.

Objective The objective is to assure that the control rods are of such a design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.

Specifications

1. The three control rods shall have scram capability, shall be used for reactivity control, and shall contain borated graphite, B4C, in powder form.

2.

The control rod cladding shall be aluminum with nominal thickness 0.71 mm (0.028") and length 17".

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTech nicai Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Basis The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, B4C, powder. These materials must be contained in a suitable clad material such as aluminum to ensure mechanical stability during movement and to isolate the poison from the tank water environments. Scram capabilities are provided for rapid insertion of the control rods, which is the primary safety feature of the reactor.

5.4 FISSIONABLE MATERIAL STORAGE Applicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.

Objective The objective is to assure that fuel that is being stored will not become critical and will not reach an unsafe temperature.

Specifications

1. All fuel elements shall be stored either in a geometrical array where the k-effective is less than 0.8 for 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 permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.

3. When fuel is in storage in any area other than the grid plate, that area must be equipped with monitoring devices that both measure and record the radiation levels and temperature of the region surrounding the fuel.

Basis The limits imposed by Specifications 5.4.1 and 5.4.2 are conservative and assure safe storage.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 6.0 ADMINISTRATION 6.1 ORGANIZATION The Maryland University Training Reactor (MUTR) is owned and operated by the University of Maryland, College Park. Its position in the university's structure is shown in Figure 6.1 The university shall provide whatever resources are required to maintain the facility in a condition that poses no hazard to the general public or to the environment.

6.1.1 Structure Figure 6.2 shows the MUTR organizational structure.

6.1.2 Responsibility The Dean College of Engineering is responsible for the oversight and operation of the school of engineering.

The Chair of the Department of Materials Science and Engineering is responsible for the oversight and operation of the Department of Materials Science and Engineering.

The Director of MUTR: Responsibility for the safe operation of the reactor facility and radiological safety shall rest with the Facility Director. The members of the organization chart shown in Figure 6.2 shall be responsible for safeguarding the public and facility personnel from undue 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 the activities of reactor operators.

The Reactor Operators are individuals who are licensed by the U.S. Nuclear Regulatory Commission (NRC) to manipulate the controls of the reactor.

6.1.3 Facility Staff Requirements

1. 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 be present in the control room.

b. A minimum of two persons shall be present in the facility or in the Chemical and Nuclear Engineering Building while the reactor is not secured: the operator in the control room and a second person who can be reached from the control room who is able to carry out prescribed written instructions which may involve activating elements of the Emergency Plan, including evacuation and initial notification procedures.

c.

A licensed SRO shall be present or readily available on call. "Readily Available on Call" means an individual who (1) has been specifically designated and the designation known to the operator on duty, (2) keeps the operator on duty informed of where he/she may be rapidly contacted and the method of contact, and (3) is 29

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 capable of arriving at the reactor facility within a reasonable amount of time under normal conditions. At no time while the reactor is not secured shall the designated SRO 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 in the control room for use by the operator. The list shall include:

a.

Management personnel

b.

Radiation safety personnel

c.

Licensed operators 30

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnicalI Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 President University of Maryland Provost & Senior Vice President Dean Clark School of Engineering Vice President Administrative Affairs Director Department of Environmental Safety Radiation Safety Committee Chair Department of Materials Science and Engineering Radiation Safety Officer 4L Director Nuclear Reactor Facility Reviews* * -

- Audit Reactor

.iembr Safety

."Mme Committee Radiation Safety Office Staff Services i

Reactor Operations Staff

-- Normal Administrative Reporting Channel Communication Lines Figure 6.1: MUTR Position in University of Maryland Structure 31

O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 Dean Clark School of Engineering (Level 1)

Chair Department of Materials Science and Engineering (Level 1)

Director Nuclear Reactor Facility (Level 2)

,I!

Senior Reactor Operator (Level 3)

Reactor Operator (Level 4)

Reactor Safety Committee 4---------

-- Normal Administrative Reporting Channel Communication Lines Figure 6.2: MUTR Organizational Structure 32

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3. The following operations shall be supervised by a senior reactor operator:

a.

Initial startup and approach to power following new fuel loading or fuel rearrangement

b.

When experiments are being manipulated in the core that have an estimated worth greater than $0.80

c. Removal of control rods or fuel manipulations in the core

d. Resumption of operation following an unplanned or unscheduled shutdown or any unplanned or unexpected significant reduction in power.

6.1.4 Selection and Training of Personnel The 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 the training 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 and 15.4.

6.2 REVIEW AND AUDIT 6.2.1 Reactor Safety Committee A Reactor Safety Committee (RSC) shall exist for the purpose of reviewing matters relating to the health and safety of the public and facility staff and the safe operation of the facility. It is appointed by and reports to the Chairperson of the Department of Materials Science and Engineering. The RSC shall consist of a minimum of five persons with expertise in the physical sciences and preferably some nuclear experience. Permanent members of the committee are the Facility 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 the committee. Alternates may be appointed by the Chairperson of the RSC to serve on a temporary basis. At least one committee member must be from outside the Department of Materials Science and Engineering.

6.2.1.1l Reactor Safety Committee Charter And Rules

1. 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 whom shall be the Campus Radiation Safety Officer (or designated alternate). No more than two alternates shall be used to make a quorum. MUTR staff members shall not constitute the majority of a voting quorum.

3. Minutes of all meetings will be retained in a file and distributed to all RSC members.

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O:\\M UTR\\2016M UTRLivingDocs\\WorkingCopyTechnical Specifications 29 feb 16RevO2292016.docx Last edit February 29, 2016 6.2.1.2 Reactor Safety Committee Review Function The RSC shall review the following:

1. Determinations that proposed changes in equipment, systems, test, experiments, or procedures are allowed without prior authorization by the responsible authority, e.g. 10 CFR 50.59;

2. All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment, or systems having safety significance;

3.

All new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity;

4.

Proposed changes in technical specifications, or license;

5. Violations of technical specifications or license. Violations of internal procedures or instructions 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 1 management, the Facility Director, and the RSC members in a timely manner after the review has been completed.

6.2.1.3 Reactor Safety Committee Audit Function

1. An annual audit and review of the reactor operations will be performed by an outside individual or group familiar with research reactor operations. They shall submit a report to the 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 reactor facility equipment, systems, structures, or methods of operation that affect reactor safety; and 34

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g.

Reactor facility emergency plan and implementing procedures.

Deficiencies uncovered that affect reactor safety shall immediately be reported to Level 1 management and the Facility Director. A written report of the findings of the audit shall be submitted to Level 1 management, the Facility Director, and the RSC members within 3 months after the audit has been completed.

6.2.2 Audit of ALARA Program The Facility Director or his designated alternate shall conduct an audit of the reactor facility ALARA Program at least once per calendar year (not to exceed fifteen months). The results of the audit shall be presented to the RSC at the next scheduled meeting. This audit may occur as part of a review of the overall campus ALARA program.

6.3 RADIATION SAFETY A radiation safety program following the requirements established in 10 CFR Part 20 will be undertaken by the Radiation Safety Office. The facility director will ensure that ALARA principles are followed during all facility activities.

6.4 OPERATING PROCEDURES Written procedures, reviewed and approved by the Reactor Safety Committee, shall be in effect and followed for the following items prior to performance of the activity. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use of independent judgment and action should the situation require such.

1. Start-up, operation, and shutdown of the reactor

2.

Installation or removal of fuel elements, control rods, experiments, and experimental facilities

3. Maintenance procedures that could have an effect on reactor safety

4.

Periodic surveillance checks, calibrations, and inspections required by the Technical Specifications or those that may have an effect on reactor safety

5. Administrative controls for operations and maintenance and for the conduct of irradiations and experiments that could affect reactor safety or core reactivity

6. For any activity pertaining to shipping, possession, and transfer of radioactive material, these procedures shall be written in conjunction with the Radiation Safety Office and the Radiation Safety Officer who shall inform the Reactor Director of any changes in regulations or laws that may require modification of these procedures. All shipping and receiving of radioactive material shall be performed in conjunction w i t h, and with the approval of the Radiation Safety Office.

7.

Implementation, maintenance, and modification to the Emergency Plan

8. Implementation, maintenance, and modification to the Security Plan 35

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9. Implementation, maintenance, and modification to the Radiation Protection Plan.

The Radiation Protection Plan shall include an ALARA plan as defined in ANSI/ANS-15.l11

10. Use, receipt, and transfer of byproduct material Substantive changes to the above procedures shall be made with the approval of the Facility Director 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 only exception to this clause is in such a case where the delay in implementation would cause a credible risk to the public or the facility. If such a case exists as determined by the Facility Director, temporary approval may be granted by the Director but must be approved by the Reactor Safety Committee within thirty days. Temporary or minor changes to procedures shall be documented and subsequently reviewed by the Reactor Safety Committee at the next scheduled meeting. The Reactor Director shall have the power to approve minor changes such as phone number changes, typographical error correction or any other change that does not change the effectiveness or the intent of the procedure. It shall be considered sufficient approval and documentation when the Director forwards by electronic means to both the Radiation Safety Officer and the Chair of the Reactor Safety Committee. A copy of the transmission shall be filed with the appropriate procedure.

6.5 EXPERIMENT REVTEW AND APPROVAL

1. Routine experiments may be performed at the discretion of the duty senior reactor operator without the necessity of further review or approval.

2. Modified routine experiments shall be reviewed and approved in writing by the Facility Director, or designated alternate.

3.

Special experiments shall be reviewed by the RSC and approved by the RSC and the Facility Director 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 its effect 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 ACTIONS 6.6.1 Actions To Be Taken In Case Of Safety Limit Violation In the event a safety limit is exceeded:

1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

2. The event shall be reported to the Reactor Director who will report to the NRC as required in section 6.7.2.

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3. An immediate report of the occurrence shall be made to the Chairman, Reactor Safety Committee, and reports shall be made to the NRC in accordance with Section 6.7.2 of these specifications, and

4. A report, and any follow-up report, shall be prepared. The report shall describe the following:

a.

Applicable circumstances leading to the violation, including when known, the cause, and contributing factors;

b. Effect of the violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public; and

c. Corrective action to be taken to prevent recurrence.

The report shall be reviewed by the Reactor Safety Committee and submitted to the NRC when authorization is sought to resume operation of the reactor.

6.6.2 Actions to Be Taken In The Event Of a Reportable Occurrence In the event of a reportable occurrence, as defined in section 1.32 of these Technical Specifications, the following actions will be taken:

1. Immediate action shall be taken to correct the situation and to mitigate the consequences of the occurrence.

2.

The reactor shall be shut down and reactor operation shall not be resumed until authorized by the Facility Director.

3. The event shall be reported to the Facility Director who will report to the NRC as required in section 6.7.2.

4.

The Reactor Safety Committee shall investigate the causes of the occurrence at its next meeting. 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 the occurrence, the effectiveness of corrective actions taken, and recommendations of measures to prevent or reduce the probability or consequences of recurrence.

6.7 REPORTS 6.7.1 Annual Operating Report A report summarizing facility operations shall be prepared annually for the reporting period ending June 30. This report shall be submitted by December 30 of each year to the NRC Document Control Desk. The report shall include the following:

1. A brief narrative summary of results of reactor operations and surveillance tests and inspections required in section 4.0 of these Technical Specifications

2.

A tabulation showing the energy generated in MW hr-' for the year 37

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3. A list of unscheduled shutdowns including the reasons therefore and corrective action taken, if any

4. A tabulation of the major maintenance operations performed during the period, including the effects, if any, on safe operation of the reactor, and the reason for any corrective maintenance required

5.

A brief description of

a.

Each change to the facility to the extent that it changes a description of the facility in the Final Safety Analysis Report

b. 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 the environment

7.

A description of any environmental surveys performed outside of the facility

8. A summary of exposure received by facility personnel and visitors where such exposures are greater than 25 percent of limits allowed by 10 CFR Part 20

9.

Changes in facility organization 6.7.2 Special Reports Notification 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, followed by 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 document

2. Release of radioactivity from the site above allowed limits

3. Exceeding the Safety Limit The written report shall be sent to the NRC document control desk. The written report and, to the extent possible, the preliminary telephone or facsimile notification shall:

1. Describe, analyze, and evaluate safety implications

2.

Outline the measures taken to ensure that the cause of the condition is determine

3.

Indicate the corrective action taken to prevent repetition of the occurrence including chances to procedures

4. Evaluate the safety implications of the incident in light of the cumulative experience obtained from the report of previous failure and malfunction of similar systems and components 38

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1. Discovery of any substantial errors in the transient or accident analysis or in the methods used for such analysis as described in the Safety Analysis Report or in the bases for the Technical Specifications

2.

Discovery of any substantial variance from performance specifications contained in the Technical Specifications or Safety Analysis Report

3. Discovery of any condition involving a possible single failure which, for a system designed against assumed failure, could result in a loss of the capability of the system to perform its safety function

4. A permanent change in the position of Department Chair or Facility Director 6.8 RECORDS

1. The following records shall be retained for a period of at least five years:

a. Normal reactor facility operation and maintenance

b.

Reportable occurrences

c. Surveillance activities required by Technical Specifications

d. Facility radiation and contamination surveys

e. Experiments performed with the reactor

f. Reactor fuel inventories, receipts, and shipments

g.

Approved changes in procedures required by these Technical Specifications

h. Minutes of the Reactor Safety Committee meetings

i. Results of External Audits

2.

Retraining and requalification records of current licensed operators shall be maintained at all times 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 environs

b.

Gaseous radioactive effluents released to the environs

c.

Radiation exposure for all facility personnel 39

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d. Radiation exposures monitored at site boundary

e. As-built facility drawing

f.

Violation of the Safety Limit

g. 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 as records in subsection 6.8.3 above.

40

Accident Analysis MHA The NRC licenses research and test reactors consistent with the NRC mission to ensure adequate protection of the public health and safety and to promote and protect the environment. NUREG 1537 Part 2 Chapter 13 Accident Analysis provides guidance and acceptable format and content for licensees to present regarding a Maximum Hypothetical Accident, MHA.

Utilizing guidance from both NUJREG 1537 and NUJREG/CR-23 87, Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors, the bounding and limiting credible accident scenario is a fuel element failure which 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 of the reactor building outside of the biological shield. Fission products are released in air from the gap and the cladding and instantaneously and uniformly mix in the volume of the reactor building. The reactor facility exhaust fans are not running and are closed during this event. No immediate protective functions are activated by radiation 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 the unrestricted 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 MHA Doses are calculated for air leakage out of the south side of the reactor. Internal, external, and shine doses are 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 time air leakage occurs out of the reactor on the north side entrances, predominantly through the roll up door and significantly less through a single entrance door. The remaining 10% of the time analysis has shown air leakage sites are located, on the eastern, southern and western sides of the reactor building, the most prominent site being on the southern side and contributing 38% of the total air leakage to a single location. A highly conservative approach assumes the MHA's airborne radioactivity escapes continuously during the event from these predominant locations from the onset of the event until the end of the leakage time. This is the source of the Maximally Exposed Individual (MET) member of the public outside of the confinement space, and at locations downwind of the MUTR. Occupational personnel located within the MUTR are exposed to internal and external radiation from the release 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 space will 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, members of the public remain in place for the duration of the event, and doses are determined over this time interval. In reality 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 work together to secure and maintain acceptable perimeters relative to exposure and dose measurements surrounding the MULTR facility.

Per 10 CFR 20.1301 the NRC regulations limit the internal and external dose to 100 mrem for members of the public; this includes doses incurred during an incident such as the MHA.

1

Accident Analysis MHA Internal and external doses are accrued by occupational personnel present in the reactor at the start of the event until they have evacuated the confinement space. Evacuation time is overly conservative and set at 5 minutes.

NUREG 2387 provides guidance and analysis for a 1MW TRIGA after 1 year of continuous operation at full power, or 365MWd. The Maryland University Training Reactor (MUTR) is licensed as a 250 kW TRIGA reactor and therefore not capable of this level of operation. The analysis, inventories and released activities from the damaged fuel elements are thus scaled back to 25% in order to determine doses to occupational personnel and members of the general public.

NUREG analysis assumes 50 elements were present in the referenced core and the central element experiences a greater than average burn up. At 1/50 or 2% of the total, the central element would contain 4% of the 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 NUREG example) for 365 days is 91.25 MWd and the corresponding activities would be 957.2 Ci of Krypton, 2357.8 Ci of Iodine and 983.3 Ci of Xenon. The MIUTR contains 93 fueled elements whereas the NUREG/General Atomics example uses 50 elements in its assessment. Therefore, the activity per MUTR element would be decreased by a factor 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 to retain the fission products. According to NUREG 2387 the gap activity fraction is approximately 1.5x10-5.

Isotope Released Activities (mCi)

Isotope Released Activities (mCi)

Kr-83 m 0.2492 1-134 5.1290 Kr-85m 0.5782 1-135 4.4621 Kr-85 0.0097 Sr-89 2.0161 Kr-87 1.1129 Sr-90 0.0625 Kr-88 1.5911 Sr-91 2.6048 Xe-133m 0.0782 Sr-92 2.9516 Xe-133 4.5702 Cs-134 0.0060 Xe-135m 1.2056 Cs-134m 0.00363 Xe-135 2.0669 Cs-136 0.0524 1-131 2.1774 Cs-137 1.0000 1-132 3.3492 Cs-138 4.153 1-133 3.8976 In addition to using MUITR release activities, allowance is taken for air leakage, radionuclide decay and shielding over the course of the event. Air leakage rates were determined using engineering analysis of the M\\UTR. Decay rates were calculated from the Chart of the Nuclides as well as the Health Physics and Radiological Health Handbook (HTPRRH). The reduction in shine dose due to shielding was determined from 2

Accident Analysis MHA figure 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 to occupational personnel and members of the public. Doses to the public are from ground level release due to air leakage from the south side of the reactor building. Horizontal and vertical diffusion coefficients, where applicable, 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 are extrapolated. A Pasquill category F, moderately stable condition, was chosen for all releases as a conservative category.

3

Accident Analysis MHA Methodology for Occupational Dose Calculations The following are the formulae used to calculate the occupational doses:

Committed Dose Equivalent (CDE) to the thyroid and CEDE for reactor occupational personnel CEDE = E

[.BR

• DCF1 nt

• Ai[1-exp(-aef, leak tst)))

Deep Dose Equivalent (DDE) to reactor occupational personnel Terms used in the above Dose Equations BR 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]

9*eff 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 [in 3]

tst Reactor personnel stay time (evacuation time) [s]

4

Accident Analysis MHA Methodology for Public Dose Calculations For the 10% of the time the air in the MUTR predominantly flows out of the southern side of the reactor due to atmospheric conditions. This scenario describes the MEl since during the other 90% of the time doses to any member 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

• Tstay Parameters in CEDE Equation BR - Breathing rate per NRC Guidance [in 3 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

• Tsa Parameters in DDE Equation DCFext - 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 Equation F-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 MHA Summary of Doses CEDE Occupational 10.2 mrem DDE Occupational 1.62 mrem TEDE Occupational 11.82 mrem CEDE public At MEl1-88.5 mrem DDE public At MEl1-2.78 mrem Shine Dose Public At ME1-7.325 mrem TEDE Public MEI - 98.605 mrem 6