ML13246A014

Request for Renewal of Facility Operating License R-129

ML13246A014
Person / Time
Site:	University of Texas at Austin
Issue date:	08/21/2013
From:	Whaley P University of Texas at Austin
To:	Patrick Boyle Document Control Desk, Division of Policy and Rulemaking
References
TAC ME7694
Download: ML13246A014 (32)

Text

Department oCMechanical Engineering V THE UNIVERSITY OF TEXAS A" AUSTIN

-' )..*,., Laboratory-Austin, Texas 78758 NuclearEnginieering "lhaching 512-232-5370 -.,FAX512-4714-4589 h- p.'llwww.rne.utexas.eda/-netl/

08/21/2013 ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 Patrick G. Boyle Nuclear Engineer NRR/DPR/PRLB

SUBJECT:

Docket No. 50-602, Request for Renewal of Facility Operating License R-129 REF: UNIVERSITY OF TEXAS AT AUSTIN - REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE RENEWAL REQUEST FOR THE NUCLEAR ENGINEERING TEACHING LABORATORY TRIGA MARK II NUCLEAR RESEARCH REACTOR (TAC NO. ME7694)

Sir:

Attached is a partial response to the Request for Additional Information referenced above, including:

• RAI 4

• RAI 5

• RAI 16.2

• RAI 17 (in 2 parts)

• RAI 25

• RAI 30

" RAI 32.1

" RAI 32.2

• RAI 35.7

" RAI 35.8

• RAI 36.2

• RAI 36.4

• RAI 40.1

• RAI 40.3

• RAI 40.5

• RAI 40.6

• RAI 40.7 kf~

If there are any questions, please feel free to contact P. M. Whaley at 512 232 5373 or whaley@mail.utexas.edu.

Your attention in this matter is greatly appreciated, P. M. Whaley /

I declare under penalty of perjury that the foregoing is true and correct.

S. Biegalski ATT: RAI RESPONSES

RAI 16.2 The coolant flow rates cited in UT SAR Table 5-1 for the tubes and shell side of the primary coolant heat exchanger appear to be in error. Please confirm and revise accordingly.

RESPONSE

The shell and tube flow rate values for Table 5-1 appear to have been inadvertently transposed, and will be changed to indicate:

Flow Rate (shell) 400 gpm (25.2 Ips)

Flow Rate (tubes) 250 gpm (15.8 Ips)

RAI 17 (a)

The guidance in NUREG-1537 Section 5.2, "Primary Coolant System," requests that the licensee provide information regarding the coolant system control and safety instrumentation, including the location and functions of sensors and instruments; the SCRAM or interlock functions that prevent exceeding the safety limits should be shown and discussed. UT SAR Section 5.3.4 provides a summary of the control system, but does not include information on the pool low-level scram setpoint referred to in UT SAR, Section 13.6.2 and the UT TRIGA TS 3.4, Table 2....

RESPONSE

Section 5.3.4 Control System will be revised to add information and structure as indicated:

a. Level gage A metric scale is used to provide local indication of pool level. With the 23.5 cm mark secured to the tank equipment monitoring ring, 8 cm corresponds to 8.10 meters above the floor of the pool. Pool level is normally maintained at 8.10 + 0.05 m, corresponding to an indicated 8 +/- 5 cm.

b. Level Control Functions Pool level is monitored by 5 float switches using sealed relays on two separate assemblies. Two level switches are adjusted to alarm if water level is 5 cm above or 10 cm below the nominal value (8.15 or 8.05 m above the pool floor, 8.8 or 7.5 cm gage level). Two level switches are adjusted to provide a reactor trip initiation if water level falls to 7.79 m above the pool floor, 30 cm below normal pool level. One level switch is adjusted below the reactor trip switches to alarm at a remote 24-hour response station.

c. Temperature Control (Followed by the information currently in the draft Safety Analysis Report.)

RAI 17, Part 2: The guidance in NUREG-1537 Section 5.2, "Primary Coolant System," requests that the licensee provide information regarding the coolant system control and safety instrumentation, including the location and functions of sensors and instruments; the SCRAM or interlock functions that prevent exceeding the safety limits should be shown and discussed...In addition, it does not describe actions that are needed, or should be taken, should pressure at the chilled water outlet rise above the pressure at the pool inlet to the heat exchanger. Please confirm and revise accordingly.

RESPONSE

Section 3.8.3 specifies:

D The pressure difference between chilled water outlet from the pool heat exchanger and pool water inlet SHALL NOT be less than 7 kPa (1 psig)

Section 3.8.4 specifies:

D.1 ENSURE the reactor is D.1 IMMEDIATE SHUTDOWN OR D. The pressure difference D.2 Verify pressure differential is D.2 IMMEDIATE between chilled water outlet greater than 7 kPa (1 psig) from the pool heat exchanger and pool water OR inlet is less than 7 kPa (1 D.3 IMMEDIATE psig) D.3 RESTORE pressure difference to greater than 7 kPa (1 psig)

OR D.4 Isolate chill water D.4 IMMEDIATE "Actions that are needed, or should be taken, should pressure at the chilled water outlet rise above the pressure at the pool inlet to the heat exchanger" are to (1) ensure the reactor is shutdown or (2) one of the following (a) verify differential pressure is greater than 7 kPa (1 psig), (b) restore the pressure difference to greater than 7 kPa (1 psig),

or (c) isolate the chill water.

The basis states, "The specified pressure difference assures that any postulated heat exchanger leakage will not release potentially contaminated water to the chill water system." If a limiting condition for operation cannot be met, the reactor should be shutdown. It is possible that the differential pressure channel is not operating properly, and actual differential pressure is adequate to me the specification. If it can be verified that actual pressure is adequate, operations may continue. If differential pressure does not meet the specification but can be restored, operation contingent on restoration is acceptable. If the pressure differential does not meet the specification and cannot be restored, isolation of the chill water system from the heat exchanger will prevent flow from the pool water system into the chill water system.

RAI 25

The guidance in NUREG-1537 Section 11.1.5, "Radiation Exposure Control and Dosimetry," requests that the licensee consider of all groups (including embryos and fetuses, declared pregnant women, minors, and students) when establishing dose limits. The UT SAR does not mention these groups, or whether they were considered in establishing dose limits. Please confirm that these considerations are included in your radiation protection procedures or elsewhere.

RESPONSE

While ALARA has always been applied to doses to all groups, an exposure limit for special categories of radiation workers (declared pregnant workers, minors, etc.) was not explicitly stated in the NETL ALARA program. A revised version of the NETL ALARA program has been drafted and will be submitted to the Reactor Oversight Committee for approval stating that for special categories of radiation workers (declared pregnant workers, minors, etc.), the dose limit will be established on a case-by-case basis by the NETL Director with concurrence from the HP.

RAI 30

The guidance in NUREG-1537 Section 13.1.6, "Experiment Malfunction," requests that the licensee provide analysis of an experiment malfunction event. UT SAR, Section 13.8.2.C includes an analysis of fueled experiment fission product inventory to estimate the allowable limit on production of iodine and strontium in an experiment based on a 2-hour DAC limit; the analysis uses fission yield data from uranium-235. Please clarify the source of the activation materials for producing iodine, and identify actions needed if it is not 231U. In addition, please show that the limits for the iodine and strontium are less than the values that could be released in an MHA and the doses to the public from such releases are within the 10 CFR Part 20 limits.

RESPONSE

This RAI is addressed in 2 parts. Fission product isotopes of interest are first considered, then neutron activated materials. In both cases, the derived air concentrations of 10CFR20 appendix B is identified for each element of interest, then the activity that (if released into the reactor bay, volume of 4.12E9 ml) would result in 1000 DAC is calculated. Finally, the maximum potential exposure associated with a postulated release is calculated using CAP-88.

FISSION PRODUCT ISOTOPES OF INTEREST Fission products of interest include radioactive strontium and iodine. The results of calculations for fueled experiments are provided in Table 1.

Table 1, Maximum Potential Dose Outside Reactor Bay from Fueled Experiments Fueled Exp. Bay Activity Max. Dose Isotope Limit Max. Dose DAC (1000 DAC) (1000 DAC) tCi mrem/yr ICi/ml PCi mrem/yr Sr-90 (D) 93500 2.65 8E-09 32960 0.93416 Sr-90 (Y) 93500 2.65 2E-09 8240 0.23354 1-131 932000 0.414 2E-08 82400 0.036603 1-132 932000 0.00509 0.000003 12360000 0.067503 1-132m 932000 0.00142 0.000004 16480000 0.025109 1-133 932000 0.0452 1E-07 412000 0.019981 1-134 932000 0.00461 0.00002 82400000 0.407579 1-135 932000 0.0137 7E-07 2884000 0.042394 NEUTRON ACTIVATED MATERIALS Neutron activation products with the potential to exist in a vapor form (e.g., noble gases, low vapor pressure, and sublimation potential at near atmospheric pressure) were identified. The maximum potential dose associated with potentially gaseous activation products is provided in Table 2, with the exception that materials with a half-life too short to be included in CAP-88 calculations are not included.

Table 2: Maximum Potential Dose from Release of Volatile Material (at 1000 DAC in the Reactor Bay)

Isotope DAC Bay Activity Max. Dose jiCi/ml PCi mrem/yr H-3 0.00002 82400000 0.0498 F-18 1E-07 412000 0.000984 CI-36 0.000001 4120000 323 CI-38 0.00002 82400000 0.259 Ar-37 1 4.12E+12 0 Ar-39 0.0002 8.24E+08 0.00399 Ar-41 0.000003 12360000 3.04E-05 Ga-70 0.00008 3.3E+08 0.0513 Ga-72 0.000001 4120000 0.143 As-74 3E-07 1236000 0.732 As-76 6E-07 2472000 0.0289 As-77 0.000002 8240000 0.036 Se-75 3E-07 1236000 2.58 Se-77m 1E-07 412000 0 Se-81 0.0001 4.12E+08 0.037 Se-81m 0.00003 1.24E+08 0.0261 Se-83 0.00005 2.06E+08 0.799 Br-80 0.00009 3.71E+08 0.0701 Br-80m 0.000007 28840000 0.0279 Br-82 0.000002 8240000 0.00101 Kr-85m 0.00002 82400000 0.0234 Kr-87 0.000005 20600000 0.032 Kr-88 0.000002 8240000 0.0365 Rb-86 3E-07 1236000 0.341 Rb-88 0.00003 1.24E+08 0.151 Xe-125 0.00002 82400000 2.81 Xe-127 0.00001 41200000 0.0196 Xe-129m 0.0002 8.24E+08 0.0318 Xe-131m 0.0004 1.65E+09 0.0243 Xe-133 0.0001 4.12E+08 0.0232 Xe-133m 0.0001 4.12E+08 0.0224 Xe-135 0.00001 41200000 0.0191 Xe-135m 0.000009 37080000 0.0205 Xe-138 0.000004 16480000 0.0366 Hg-197 0.000006 24720000 1.15 Hg-197m 0.000004 16480000 1.03 Hg-199m 0.00007 2.88E+08 0.559 Hg-203 5E-07 2060000 0.156

Of the materials analyzed, only chlorine 36 has the potential to exceed the 100 mrem/yr exposure limit for routine non-occupational exposure or the 10 mrem/yr limit on routine effluents. Neutron activation is calculated by:

A = 0 . N -. (l-e-"')

Where A is the specific activity, b is the neutron flux, N is the number density of the target material, a is the cross section, and A is the decay constant. Defining KCas total activity, VC135 as the volume of the target material, and using the standard mass/atomic weight relationship:

AMU VC1: 35 j Vc13 0*cr (1 -e -A,)

or KC AMU m0 . (1 - e-"') NA For an activity that would result in 1000 times the C136 DAC (4.12E6 QCi), neutron flux approximately 1E13 n/cm 2-s, cross section of approximately 60 barns, the number of target chlorine 35 atoms required to achieve the 1000 DAC value (where t is the irradiation interval) is:

(4.12 Ci)(3.7x10 10 Bq)(cn 2

s)(AMU) = 4.22x10- 4 /(1 - eAt) 3 24 (Ci)(101 n)(60x1o_ cn2)(1-e_1t)(6.023x1O23)

The decay constant (2) is 0.693/t%; the denominator of the equation is on the order of 1E-8 for values less than about 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. Therefore the mass of C135 required to achieve 1000 DAC is on the order of 42 kg. The 100 mrem/yr exposure limit for routine operations can be met if the target mass is reduced to 31%, on the order of 13 kg. The annual limit of 10 mrem/yr from routine effluents can be met if the target mass is reduced to 3.1%, or 1.3 kg. The chemical reactivity, physical form, and half-life of chlorine 36 (3E5.years) render generation of C136 quantities that are capable of challenging exposure limits virtually impossible.

Since there is no realistic mechanism to generate a quantity of C136 that could challenge limits, a single specified value of 1000 DAC is preferred (rather than noting the exception) to provide simple and clear guidance.

TRIGA fuel elements are specifically exempted from consideration in experimental limits, and this analysis does not rely on or impact the maximum hypothetical accident (TRIGA fuel element failure).

RAI 32.1

32. The "Interim Staff Guidance for the Streamlined Research Reactor License Renewal Process," (ISG) identifies ANSI/ANS-15.1-2007 and the corresponding regulatory positions in NUREG-1537, Appendix 14.1 are the guidance documents for the review of technical specifications. The guidance in ANSI/ANS-15.1-2007 Section 1.3, "Definitions," recommends definitions commonly used in Research and Test Reactor TS.

The TS definitions noted below were either missing, were not consistent with guidance, or were lacking recommended details. (Note: capitalization for this sequence of RAIs follows the style of the proposed UT TRIGA TS.)

32.1 The proposed UT TRIGA TS definitions section 1.0 does not describe definitions for: core configuration; license; licensee; protective action; reactivity worth of an experiment; reactor operator; reactor operating; responsible authority; safety limit; scram time; senior reactor operator; shall/should/may definitions (only the definition for "shall" is provided); true value; unscheduled shutdown. Please provide definitions for the above or provide justification for not using them.

RESPONSE

The following terms are not used at all: Responsible Authority, Scram time, Should, True Value, and Unscheduled Shutdown. In the absence of the term "Should," May is unambiguous.

The following terms are not used in specification statements, but are unambiguous and understood In context:

Core Configuration, Reactor Operating, and Protective Action. In addition, protective action is used in a generic sense, and includes reactor safety system response as well as actions for protection of personnel.

License and licensee have more than one 10CFR20 definition, and are understood in context.

Reactivity worth of an experiment is unambiguous and understood in context.

Safety Limit is an explicit section of the Technical Specifications, and has specific statements and requirements related to the section.

As per a recent proposed Technical Speciation change, reactor operator and senior reactor operator are currently being changed to the terminology of 10CFR55. Since operators are trained to 10CFR55 the terms should be well understood, and "certified" is not a correct term.

RAI 32.2

32. The "Interim Staff Guidance for the Streamlined Research Reactor License Renewal Process," (ISG) identifies ANSI/ANS-15.1-2007 and the corresponding regulatory positions in NUREG-1537, Appendix 14.1 are the guidance documents for the review of technical specifications. The guidance in ANSI/ANS-15.1-2007 Section 1.3, "Definitions," recommends definitions commonly used in Research and Test Reactor TS.

The TS definitions noted below were either missing, were not consistent with guidance, or were lacking recommended details. (Note: capitalization for this sequence of RAIs follows the style of the proposed UT TRIGA TS.)

32.2 The following definitions are not consistent with the guidance: channel calibration, excess reactivity; experiment; confinement; movable experiment; secured experiment. Please revise accordingly or provide justification for the deviations.

RESPONSE

1. channel calibration ANSI/ANS-15.1-2007 defines:

channel calibration:A channel calibrationis an adjustment of the channelsuch that its output corresponds with acceptableaccuracy to known values of the parameterthat the channel measures. Calibrationshall encompass the entire channel, including equipment actuations,alarm, or trip, and shall be deemed to include a channel test.

The original proposed definition is:

Channel Calibration Channel calibrationis an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameterwhich the channel measures. Calibrationshall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a channel test.

The defined term is taken verbatim from the current approved Technical Specifications. Since this was acceptable in the current (approved) Technical Specifications, it remains unchanged.

Differences between the approved definition and ANSI standard include (1) avoiding multiple uses of the word "that", deletion of "s" for parallel construction, and deletion of a comma. This is an editorial difference, and does not change the meaning of the definition in any way.

On further consideration, the second sentence will be split into two sentences for clarity in a second editorial change that does not change the meaning of the definition in any way, so that the definition will be:

Channel Calibration Channel calibrationis an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameterwhich the channel measures. Calibrationshall encompass the

entirechannel, including equipment actuation, alarm, or trip. Calibrationshall be deemed to include a channel test.

1. excess reactivity ANSI/ANS-15.I-2007 defines:

Excess reactivity: Excess reactivity is that amount of reactivity that would exist if all reactivitycontrol devices were moved to the maximum reactivity conditionform the point where the reactoris exactly critical(keff =1) at reference core conditions or at a specified set of conditions.

The original proposed definition is:

Reactivity, Excess Excess reactivity is that amount of reactivity that would exist if all the controlrods were moved to the maximum reactive conditionfrom the point where the reactoris exactly critical.

The defined term is taken verbatim from the current approved Technical Specifications. Since this was acceptable in the current (approved) Technical Specifications, it remains unchanged.

The last phrase of the ANSI definition ("or at a specified set of conditions") requires additional information to have an unambiguous meaning. Since "reactivity control devices" at the UT TRIGA are "control rods,"

and since "exactly critical" does not need amplification, the equivalent and unambiguous definition is less ambiguous than, but otherwise explicitly agrees with, the ANSI definition.

2. experiment:

ANSI/ANS-15.1-2007 defines:

experiment: any operation,hardware, or target (excluding devices such as detectors,foils, etc.) that is designed to investigate nonroutine reactorcharacteristicsor that is intendedfor irradiationwithin the pool, or on or in a beam port or irradiationfacility. Hardwarerigidly secured to the core or shield structure so as to be part of its design to carry out experiments is not normally consideredan experiment.

The original proposed definition is:

Experiment Any operation, component, or target (excluding devices such as detectors,foils, etc.), which is designed to investigate non-routine reactorcharacteristicsor which is intendedfor irradiationwithin the pool, on or in a beam tube or irradiationfacility and which is not rigidly secured to a core or shield structureso as to be part of their design.

The defined term is taken verbatim from the current approved Technical Specifications. Since this was acceptable in the current (approved) Technical Specifications, it remains unchanged.

The meaning of the word "hardware" in the ANSI standard definition is not as well defined as when the standard was developed; "component" as used in the approved definition is more descriptive. However,

the definition is still difficult to read; after further consideration, the definition will be edited in a way that does not alter the meaning to:

Experiment Any operation, component, or target (excluding devices such as detectors,foils, etc) that is:

(1) designed to investigate non-routine reactorcharacteristics,or intendedfor irradiationwithin the pool, on or in a beam tube or irradiationfacility; and (3) not rigidly secured to a core or shield structureso as to be part of the core or shield structure design.

3. confinement:

ANSI/ANS-15.1-2007 defines:

confinement: Confinement is on enclosure of the overall facility that is designed to limit the release of effluents between the enclosure and its external environment through controlled or defined pathways The proposed definition is equivalent:

Confinement Confinement means an enclosure on the overallfacility which controls the movement of air into it and out through a controlled path.

The defined term is taken verbatim from the current approved Technical Specifications. Since this was acceptable in the current (approved) Technical Specifications, it remains unchanged.

4. moveable experiment:

ANSI/ANS-15.1-2007 defines:

moveable experiment:A moveable 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 reactoris operating The proposed definition:

Experiment, Moveable A moveable 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 reactorwhile the reactoris operating.

The defined term is taken verbatim from the current approved Technical Specifications. Since this was acceptable in the current (approved) Technical Specifications, it remains unchanged.

In comparison to the ANSI standard, the defined term is changed from "moveable experiment" to "experiment, moveable" in the current approved Technical Specifications in order to place the definition

alphabetically near the definition of "experiment." Since this was acceptable in the current (approved)

Technical Specifications, it remains unchanged.

5. secured experiment ANSI/ANS-15.1-2007 defines:

secured experiment: A secured experiment is any experiment, experimental apparatus,or component of an experiment that is held in a stationaryposition relative to the reactorby mechanicalmeans. The restrainingforce must be substantiallygreater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or otherforces that are normal to the operatingenvironment of the experiment, or by forces which can arise as a result of credible malfunctions.

The proposed definition:

Experiment, Secured A secured experiment is any experiment, experimentfacility, or component of an experiment that is held in a stationaryposition relative to the reactorby mechanicalmeans. The restrainingforce must be substantiallygreaterthan those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or otherforces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible conditions.

The defined term is taken verbatim from the current approved Technical Specifications. Since this was acceptable in the current (approved) Technical Specifications, it remains unchanged.

In comparison to the ANSI standard, the defined term was changed from "secured experiment" to "experiment, secured" for the currently approved Technical Specifications in order to place the definition alphabetically near the definition of "experiment." Since this was acceptable in the current (approved)

Technical Specifications, it remains unchanged.

RAI 35.7 Section 3.3 of the guidance describes specifications for leak or loss of coolant detection and a secondary coolant activity limit. No such specifications are found in the proposed UT TRIGA TS.

RESPONSE

The chill water subsystem for the pool cooling system as described in the Safety analysis Report is protected from radioactive contamination by:

1. System design, which maintains chill water pressure greater than pool water pressure so that any potential leakage will be from chill water to pool water, and

2. Automatic closure of a pneumatic valve if differential pressure drops to 2 psid, and

3. Control room alarms if differential pressure falls to 5 psid The alarm and valve actuation are functionally tested prior to each startup. A pool- to chill- water leak would require multiple failures of the heat exchanger, heat exchanger design, and associated automatic controls as well as personnel failure to respond to an abnormal condition. A primary to secondary leak is not credible for the UT reactor.

RAI 35.8 Section 3.8.2 of the regulatory interpretations states that containers for experiments containing known explosive materials shall be designed such that the design pressure of the container is twice the pressure the experiment can potentially produce. Proposed UT TRIGA TS 3.6 "Limitations on Experiments," does not include such a specification.

RESPONSE

The 13.8.2, D provides analysis for over pressure from explosives in a standard capsule used at the UT TRIGA reactor. Experiment design specification (5.4.3, part 9) will be revised to:

9. Use of explosive solid or liquid material with a National Fire Protection Association Reactivity (Stability) index of 2, 3, or 4 in the reactor pool or biological shielding

a. SHALL NOT exceed the equivalent of 25 milligrams of TNT without prior NRC approval.

b. SHALL be irradiated in a container capable of withstanding twice the pressure the experiment can potentially produce.

RAI 36: ANSI/ANS-15.1-2007 Section 3, "Limiting conditions for operations," requests that the licensee provide LCOs for constraints and operational characteristics that shall be adhered to during operation. The ISG states that the applicable TSs should explain why the TSs, including their bases, are acceptable. The following deficiencies and differences are noted with the proposed UT TRIGA LCOs: Please confirm and revise accordingly, or explain why such changes are not necessary.

36.2 The setpoint stated for condition B for Specification B in proposed UT TRIGA TS 3.3, "Measuring Channels," is stated as 2mW. However, the neutron count rate should be stated in terms of

- neutrons per unit time.

RESPONSE

3.3.3 Specifications will be revised to state:

A The MEASURING CHANNELS specified in TABLE 1 SHALL be OPERATING B The startup channel indicates greater than 2 mW And the basis will be revised to include:

Experience has shown that subcriticalmultiplication with the neutron source used in the reactordoes not provide enough neutronflux to correspondto an indicatedpower level of 2x10y7 %. Therefore an indicated power of 2xl0-7 % (or 2 mW) or more indicates operating in a potential criticalcondition, and at least one neutron channel is required with sensitivity at a neutron flux level corresponding to reactorpower levels less than 2x10-7 % ("Startup Channel"). If the indicated neutron level is less than the minimum sensitivity ffor the channel, a neutron source will be used to determine that the channels is responding to neutrons to ensure that the channel is functioning priorto startup.

RAI 36: The guidance in ANSI/ANS-15.1-2007 Section 3, "Limiting conditions for operations," provides guidance and recommendations for the specifications pertaining to the limiting conditions for operation (LCO). This guidance is supplemented by NUREG-1537 Appendix 14.1. Some deficiencies and differences with the proposed UT TRIGA TS are described below. Please discuss these deficiencies and differences and revise accordingly.

36.4: Proposed UT TRIGA TS 3.4 Table 2 does not provided the scram setpoints for the Reactor Power Level, Fuel Temperature, and Pool Water Level SAFETY SYSTEM CHANNELS.

RESPONSE

Table 2 will be revised as:

TABLE 2: REQUIRED SAFETY SYSTEM CHANNELS Minimum Function OPERATING Mode Safety System Channel Number Setpoint/Applicability or Interlock Operable STEADY STATE PULSE MODE MODE Reactor power level 2 Scram 1.1 MW NA Manual scram bar 1 Scram YES YES Fuel Temperature 1 Scram 550°C 550°C Pool water level 1 Scram YES YES CONTROL ROD Prevent withdrawal of (STANDARD) position 1 standard rods in the NA YES interlock PULSE MODE Prevent inadvertent Pulse rod interlock['] 1 pulsing while in YES NA STEADY STATE MODE

RAI 4

The guidance in NUREG-1537 Section 4.2.5, "Core Support Structure," requests that the licensee provide design information pertaining to the core support structure. UT SAR Section 4.2.5 provides some information, but does not address suitability for continued use. Please confirm whether there is any visual evidence of cracking, corrosion, or deformation of the core support structure, and state whether the structure is appropriate for continued use for the operating period being requested.

RAI 5

The guidance in NUREG-1537 Section 4.3, "Reactor Tank or Pool," requests that the licensee provide a description of the reactor tank and associated components including how those components will perform their intended functions to prevent possible leakage associated with chemical interactions, penetration, and weld failures. The UT SAR does not provide sufficient information. Please confirm whether there is any visual evidence of cracking, corrosion, or deformation of the reactor pool liner, connected pipes or beam ports and provide a discussion of preventative measures employed to monitor and maintain the integrity of the connected primary coolant system over the life of the facility.

RESPONSE

(1) In 2004 the reactor reflector was replaced, using support by in-pool divers with helmet mounted cameras.

Although not specifically acquired to inspect the pool and reflector stand, the extensive video provides evidence that there is no detectable degradation in the reflector stand or the pool. Four snapshots taken from the video are provided, strictly as a sample of available graphic information.

(2) There is no current visible evidence of cracking, corrosion or deformation of the core support structure or the reactor pool, connected pipes or beam ports.

(3) Maintenance of conductivity minimizes potential corrosion (RAI 18).

(4) Monitoring pool level provides a means to detect pool and pool cooling system leakage.

RAI 40.1 ANSI/ANS-15.1-2007 Section 6.1 "Organization," recommends organizational structures including levels and reporting authority. UT TRIGA TS 6.1.1 "Structure," Figure 6.1, does not describe the organizational structure as described in ANSI/ANS-15.1-2007, Section 6.1 "Organization."

RESPONSE

Levels and reporting authority are represented on the organizational charts. A correlation between ANSI standard and UT structure will be incorporated in the basis.

RAI 40.3 Although the guidance of ANSI/ANS-15.1-2007 Section 6.1.3 "Staffing," have been included in proposed UT TRIGA TS 6.1.3 "Staffing," the following items are not consistent with the guidance:

A designated senior reactor operator shall be readily available on call. "Readily Available on Call" means an individual who has been specifically designated and the designation known to the operator on duty, can be rapidly contacted by phone, by the operator on duty.

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) other operations personnel.

Events requiring the presence at the facility of the senior reactor operator are (a) initial startup and approach to power, (b) all fuel or control-rod relocations within the reactor core region, (c) relocation of any experiment with reactivity worth greater than one dollar; (d) recovery from unplanned or unscheduled shutdown or significant power reduction.

RESPONSE

The proposed Technical Specifications includes a definition for INITIAL STARTUP:

INITIAL STARTUP A reactorstartup and approach to powerfollowing:

1 Modifications to reactorsafety or control rod drive systems, 2 Fuel element or control rod relocationsor installationswithin the reactor core region, 3 Relocation or installationof any experiment in the core region with a reactivity worth of greaterthan one dollar, or 4 Recovery from an unscheduled (a) shutdown or (b) significant power reductions.

The proposed specification for staffing (section 6.1.3) includes:

Whenever the reactoris not secured, the reactorshall be (1) under the direction of or (2) directly operated by a (USNRC licensed) Senior Operator,designatedas Reactor Supervisor. The Supervisor may be on call if cognizant of reactoroperations and capableof arrivingat the facility within thirty minutes.

The specification statement from section 6.1.3 referenced above will be revised to:

1. Replace "on call" with "capableof being rapidly contacted by phone by the operatoron duty," and

2. Add "The Supervisor will be present at the facility for any INITIAL STARTUP."

The statement "Names and telephone numbers of reactorfacility management personnel,radiation safety personnel, other operationspersonnelshall be listed in the control room for use by the operator" will be added to section 6.1.3.

RAI 40.5 ANSI/ANS-15.1-2007 Section 6.4 "Procedures," recommends that procedures be written for surveillance checks, calibrations, and inspections required by the TS or those that may have an effect on reactor safety.

RESPONSE

The proposed Technical Specifications section 6.3 requires written procedures for:

a. Startup, operation, and shutdown of the reactor

b. Fuel loading, unloading, and movement within the reactor.

c. Controlrod removal or replacement.

d. Routine maintenance, testing, and calibrationof control rod drives and other systems that could have an effect on reactorsafety.

e. Administrative controls for operations, maintenance, conduct of experiments, and conduct of tours of the Reactor Facility.

f. Implementing proceduresfor the Emergency Plan or Physical Security Plan.

and:

a. Personnelradiationprotection, in accordance with the Radiation Protection Programas indicated in Chapter11

b. Administrative controlsfor operationsand maintenance

c. Administrative controls for the conduct of irradiations and experiments that could affect core safety or reactivity The currently approved proposed Technical Specifications section 6.3 explicitly incorporates the ANSI recommendations for written procedures, and the proposed specifications will be revised to:

a. Startup, operation, and shutdown of the reactor.

b. Fuel loading, unloading and movement in the reactor.

c. Routine maintenance of major components of systems that could have an effect on reactor safety.

d. Surveillance calibrationsand tests required by the technical specificationsor those that could have an effect on reactorsafety.

e. Administrative controlsfor operation,maintenance and the conduct of experiments or irradiationsthat could have an effect on reactorsafety.

f. Personnelradiationprotection, consistent with applicableregulationsor guidelines,and shall include a management commitment and programs to maintain exposures and releases as low as reasonablyachievable.

g. Implementation of requiredplans such as the emergency plan or physical security plan.

RAI 40.6 ANSI/ANS-15.1-2007 Section 6.7.2(1) "Special reports," specifies facsimile or similar conveyance of the special report. Proposed UT TRIGA TS 6.8(c) specifies telegraph of similar conveyance.

RESPONSE

The referenced part is:

The SAFETY LIMIT violation shall be reported to the Nuclear Regulatory Commission within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone, confirmed via written statement by email, fax or telegraph The Merriam-Webster online dictionary defines FAX as synonymous with FACSIMILE.

RAI 40.7 ANSI/ANS-15.1-2007 Section 6.8 "Records," provides recommendations for record retention. Section 6.8.2 recommends an administrative control that retraining and requalification records for operators be retained for at least one certification cycle (per 10 CFR 55.55(a) this period is 6 years) and be maintained at all times the individual is employed or until the certification is renewed. The proposed UT TRIGA TS 6.10 is not consistent with this guidance

RESPONSE

10CFR55(a)5 states:

Records. The requalificationprogram documentation must include the following:

(i) The facility licensee shall maintain records documenting the participationof each licensed operatorand senior operatorin the requalificationprogram. The records must contain copies of written examinations administered,the answers given by the licensee, and the results of evaluationsand documentation of operating tests and of any additionaltrainingadministeredin areasin which an operatoror senior operatorhas exhibited deficiencies. The facility licensee shall retain these records until the operator'sor senior operator'slicense is renewed.

The regulation explicitly addresses licenses that will be renewed, and does not address record retention following termination of licenses. Paragraph § 55.55 (Expiration) states:

(a) Each operatorlicense and senior operatorlicense expires six years after the date of issuance, upon termination of employment with the facility licensee, or upon determinationby the facility licensee that the licensed individual no longer needs to maintain a license.

Therefore the proposed Technical Specification will be revised to:

One License Cycle The facility licensee shall maintain records documenting the participationof each licensed operatorand senior operatorin the requalificationprogram. The recordsmust contain copies of written examinations administered,the answers given by the licensee, and the results of evaluationsand documentation of operatingtests and of any additionaltraining administeredin areasin which an operatoror senior operatorhas exhibited deficiencies. The facility licensee shall retain these records until the operator'sor senior operator's license is renewed, expired, or terminated.

UT-Austin, License Renewal June 25, 2012 RAI Response Status Comment Not submitted when planned Not submitted when planned rIltdbt: Ut~bUlUe LIM Iiit:IIVUb Ubet: tU1 bWOaUy bWLU ICUUUIiIt IktMdUY-ýWLW OllU Nil IC-~.) OaU LIE I 101-YUIO-UI Ott-y-l and include comparisons with UT TRIGA measurements that demonstrate that those methods are appropriate to analyze the limits imposed by the UT TRIGA TS.

partial response 8.3 Please provide the technical parameters including analysis of "reactor kinetic behavior, basis reactor criticality, control rod worth, definition of the limiting core configuration (LCC), [etc.]" (NUREG-1 537, Section 4.5.1). State whether the comparison of calculated and measured values demonstrates acceptable model development. partial response 9 The GA-4361 unit cell parameters are displayed and compared with UT TRIGA core parameters. Please provide the 12/19/2012 ML13002A015 P Partially acceptable for the 1992 and LCC, but OC is technical parameters that are applicable to UT TRIGA. missing 10 Given the difference between the "Reference" and 'Current" values of excess reactivity and shutdown margin, which 12/19/2012 ML13002A015 P Partially acceptable for the 1992 and LCC, but OC is values are being used in the UT TRIGA TS. missing 11.1 Please describe any limits or conditions on the evaluation of excess reactivity contributors, such as those due to temperature variations and poisons (e.g., xenon and samarium).Please provide calculations of full power reactivity defects for power, xenon, and samarium.

11.2 Please provide calculations for excess reactivity and control rod worths, and evaluate whether they are in agreement with the analytical model and with UT TRIGA performance. Provide a discussion that describes the evaluation of these calculations to demonstrate acceptable reactor shutdown and shutdown margin. Include consideration of experiment reactivity 11.3 Please provide "a transient analysis assuming that an instrumentation malfunction drives the most reactive control rod N out in a continuous ramp mode," (NUREG-1537, Section 4.5.3) of the reactor using a rate of withdrawal consistent with proposed UT TRIGA TS values of the maximum control rod withdrawal speed, reactivity rate, and the control rod scram time including uncertainties.

Not submitted when planned 11 .4 Please provide all other applicable technical parameters, "excess reactivity, control rod worth, temperature coefficients,

[etc.]" (NUREG-1 537, Section 4.5.3).

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UT-Austin, License Renewal June 25, 2012 RAI Response Status RAt # Question Reply Date ADAMS Accept Comment 13 Please confirm that the Bernath correlation is used to characterize DNBR for UT TRIGA, or demonstrate the applicability of Biasi correlation to UT TRIGA analysis.

14.1 Please provide clarification as to the relationship of the reactivity coefficient with the analysis provided in UT SAR Section 4.6.

14.2 The basis for TS 3.2 "Pulsed Mode Operation," states that the reactivity limits are established so as to meet fuel temperature limits. However, this is inconsistent with the statements in UT SAR Section 4.6 as described above. Please revise the discussion in UT SAR Section 4.6 to support the UT TRIGA TS.

14.3 UT SAR, Section 4.6 (p. 4-46) provides a series of statements regarding pulse reactivities and responses that are not supported by analyses. Please provide analysis supporting these statements in sufficient detail so that a confirmatory analysis can be performed.

15.1 Please describe the analytical methods used to determine the DNBR, including the core inlet and exit conditions

[assumed and other assumptions and correlations employed. This [TRACE DNBR] analysis should describe the parameters determined from the LCC such as peaking factors and limiting coolant inlet temperature and that the inlet temperature used for DNBR is a limiting value by showing how it corresponds to the primary pool water temperature measuring channel value.

15.2 Please provide a comprehensive description of the calculational methods and the results that demonstrate the acceptability of design assumptions and TS for pulsing at UT TRIGA (e.g., the LCC, the approved power level, the pulse of reactivity inserted by the transient rod as allowed by TS, the value of the fuel temperature coefficient, the effective delaved neutron fraction, the Dromot neutron lifetime.

IPlease provide a pH testing TS or provide a justification for not doing so. Not submitted when planned The licensee cites a correlation that determines effective release height above the building exhaust stack due to effluent momentum from the purged air system or the ventilation system. Please confirm that the correct form of the correlation is AH = D (Vs/p)1 .4 and not as it is stated in the UT SAR.

19.2 The licensee uses two different stack exit diameter values for the stack (0.4012 m2 on UT SAR, p. 9-6 and 45.72 cm on UT SAR p. 9-2). Please explain this discrepancy.

19.3 Ensure the impact of the above changes on offsite doses for both normal operation and accident conditions are considered and revised accordingly.

20.1 Please identify all locations covered by the license where fuel elements are stored, identify the types and numbers of fuel elements that are stored, provide details concerning the storage rack or bin geometry, and analysis that demonstrates that such racks or bins provide adeauate conditions for storage.

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UT-Austin, License Renewal June 25, 2012 RAI Response Status RA usinReply Date ADAMS Accept Comment 22.2 Please provide the 4lr occupational xosr includingi stay timeis and theý effect of venitila3tion, and hoihs comIpare to, the limi1tsýof 10- CFR Part 20 and the co,,mmitmnents of the UTITRIGA ALARA program.

22.4 UTISAR Section 11 .1.1.1.2 provides a conservative estimate of offlsite, 41lAr air concentrations using an equation for ground level concentration at the building center. Please provide a reference for the equation cited, and a discussion of its suitability for providing dose calculations for members of the public and their location.

22.5 Please provide a complete description of the maximally exposed individual calculation, including how the estimates compare to the limits in 10 CFR Part 20 and the commitments of the UT TRIGA ALARA program.

22.6 UT SAR Section 11.1.1.1.2 provides conservative dose estimates for the maximally exposed individual of 66 mrem per year using the CAP88-PC computer code. UT TRIGA TS 3.5.3(D) indicates that releases of 41Ar from the reactor bay to an unrestricted environment SHALL NOT exceed 100 Ci per year, and provides CAP88-PC model results indicating that 100 Ci per year release of 41Ar would result in a maximally exposed individual dose of 0.142 mrem per year.

Please resolve this discrepancy between the maximally exposed individual doses in the UT SAR and those provided in the TS.

22.7 UT SAR Section 11.1.1.1.2 provides a discussion of the maximally exposed offsite individual, but does not provide doses to members of the public. Please provide a discussion of potential public doses.

27 Please provide an ana eM Afor the UT TRIGA including doses to the workers and to the individuals in the non-restricted areas that bounds all other accident analyses. Please describe all assumptions, the operating conditions of the HVAC system, and the sequence of events used in calculating the potential radiological consequences and discuss how those consequences are less than the applicable limits in 10 CFR Part 20. Please provide sufficient detail to allow independent onfirmation of these resultsh 27.2.1 The [atmospheac dispersion] calculations are then performed for distances from 10 to 100 meters from the building.

Because, the reactor building is both tall and wide, any re ease from the stack culd be accumulated in the buildings wake. Thereforesytea l e assumption of elevated release is appears inaccurate. Please justity the use of the elevated release values for dose estimates at nearby distances from the facility. 2 27.2.2 if there is an error in the correlation used for the plume rise (see RAt 19.1), the estimated plume rise above the stack height may be inaccurate. Please confirm and revise accordingly.

27.3 For the determination of effluent leakage around doors and HVAC duct vents the licensee employs complicated discussions and assumptions that are not supported or justified. Please revise the discussion and calculations using applicable assumptions for building overpressure.

27.4 UT SAR page 3-7 states that the reactor bay is about 18.3 m on each side, with a total of 4575 m3 of volume. This leads to a wall cross section area of about 250 m2, which is in-line with the value of 234 m2 given in the original application for licensing safety analysis report in 1991 (1991 SAR). Please confirm the building wall cross section area and revise accordingly.

27.5 For the offsite public dose calculations, in the UT SAR it does not appear consistent with the potential for ground release of the reactor bay air content, similar to that evaluated in the 1991 SAR 27,6 UT SAR Appendix 13.1, SCALE 6.1 input file, cites an input value 1.6 for the weight fraction of the ZrH1.6U fuel. Is this input value for the weight fraction of hydrogen in the fuel? Please confirm and revise accordingly.

28.1 It appears that the UT SAR does not provide sufficient information on the peaking factors and other assumptions used to estimate the maximum fuel temperature rise as listed in UT SAR Tables 13.20 and 13.21. Please provide sufficient additional information to allow confirmatory analysis.

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UT-Austin, License Renewal June 25, 2012 RAI Response Status RAI # Question Rel Date ADAMS Acept Comment 29.1 Please provide a LOCA analysis that represents the current licensed power level for the UT TRIGA in sufficient detail to allow confirmatory analysis.

29.2 Please confirm that the LOCA analysis uses peaking factors that are consistent with the LCC and revise the analysis accordingly.

29.3 UT SAR Section 13.5.8 Resultsand Conusion,' states that the maximum fuel temperature in a LOCA event after "long-term operation at full power of 2000 kW is 750°C." Please provide the analysis that supports this temperature.

lot submitted when planned 32.3 Please revise to the following: when IMMEDIATELY is used as a COMPLETION TIME, The REQUIRED ACTION should be pursued without delay and in a controlled manner.

32.4 The proposed UT TRIGA TS definition of REACTOR SHUTDOWN only requires the reactor to be subcritical by $0.29.

Please explain the discrepancy in using the value of an abnormal condition (shutdown margin) for a normal condition, i.e., the definition of Reactor Shutdown.

32.5 The proposed UT TRIGA TS definition of REACTOR SECURED MODE requires that 3 of the 4 control rods be fully inserted. Please either provide analysis demonstrating the acceptability of the insertion of 3 out of 4 rods or revise the definition to require insertion of all 4 control rods in order to satisfy the requirements of this mode.

33 The basis provided in support of the TS 2.1 references Chapter 4, Section'4.2.1 zwihich does not exist Please discuss this error and/or revise accordingly.

34.1 The basis provided in support of the UT TRIGA TS 2.2 references Chapter 4 Section 4.6 B which does not exist. Please

,,- 1A. . h.ii f-, +h. I 1 J4.J veritie*.

35.1 Section 3.1 of the guidance describes having specifications for fuel burnup, core configurations, and reactivity coefficients (ifsuch coefficients establish required conditions). Such specifications are not present in the proposed UT TRIGA TS Pease discuss.

sufficient accuracy to justify this small value of the shutdown margin.

35.4 Section 3.2 of the guidance describes that a limit be established for the maximum control rod reactivity insertion rate for non-pulsed operation. The proposed UT TRIGA TS do not provide such a specification. This rate, and the control rod scram times, are typically justified through the analysis of an uncontrolled, control rod withdrawal transient.

35.5 Section 3.2 of the guidance describes a specification for permitted bypassing of channels for checks, calibrations, maintenance, or measurements. Proposed UT TRIGA TS 3.3, 'Measuring Channels,"

does not specify when it is permitted to bypass channels for checks, calibrations, maintenance or measurements 35.6 Proposed UT TRIGA TS 4.3 'Measuring Channels,' contain Surveillance Requirements for the Fuel Temperature Channel and the Upper Level Radiation Monitor. However, there are no associated LCO specifications.

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UT-Austin, License Renewal June 25, 2012 RAI Response Status Ri# Qu n Rep e ADAMS A t Comment 36.1 'The list of measuring channels presented in Table 1 of proposed UT TRIGA TS 3.3 'Measuring Channels," does not include the data acquisition and control (DAC) and control system computer (CSC) [watchdog] which are [is] listed as I [a] SCRAM channels in UT SAR Table 4.6. Please revise or justify 36.3 'The basis for propose UT TRIGA TS 3.3 contains a statement 'According to General Atomics, detector voltages less

than 80% of required operating value do not provide reliable ..." Please explain how this statement applies to UT TRIGA and how the required conditions for safe operation are ensured by your TS. Such information should be discussed in the SAR and then utilized in the TS basis.

36A PrpsdLT TRIA TS3.4 Table2 dosntpoie h ca epit o h eco PowerLevel Fue 36.5 Proposed UT TRIGA TS 3.5 "Gaseous Effluent Control," Specification A does not establish the conditions that determine HVAC OPERABILITY (e.g., conditions or positions for the fans/louvers/doors); a basis statement is not

provided for the stated value of 10,000 cpm; such information should be discussed in the SAR and then utilized in the TS basis. Also, there are more COMPLETION TIMEs for Specification A than there are REQUIRED ACTIONs. Please explain or revise.

36.6 Proposed UT TRIGA TS 3.5 'Gaseous Effluent Control," Specification B does not provide a basis statement for the stated limit of 10,000 cpm; such information should be discussed in the SAR and then utilized in the TS basis. Also, I there is a missing COMPLETION TIME for REQUIRED ACTION B.3.

36.7 ProPosed UT TIRIGA TS 3.5, "Gaseous Effluent Contro1,' Specification D does not provide a basis statement for the stated limit of 100 Ci/yr; such information should be discussed in the SAR and then utilized in the TS basis.

36.8 !The basis for the proposed UT TRIGA TS 3.7 'Fuel Integrity," does not provide an appropriate basis statement to support the limits in Specification C. Specification B is missing the word "not" in the REQUIRED ACTION. The second occurrence of CONDITION B should be CONDITION C, 36.9 TS 3.8, "Reactor Pool Water" has no specification for maintaining an acceptable pH level.

Question 37 series require review against ANS/ANSI 15.1-2007 section 3 to explain deviation or revise TS.

37.1 Proposed UT TRIGA TS 3.2 "Pulsed Mode Operation," the COMPLETION TIME listed for the REQUIRED ACTION is "immediate." Please consider the COMPLETION TIME to be "prior to commencement of pulsing operation."

37.2.1 CONDITION A.2 the lumping together of COMPLETION TIME(S) under A.2 is confusing as to which REQUIRED ACTION must be completed first.

37.2.2 The REQUIRED ACTION(S) A.1.1 and A.1.2 are, "Restore channel to operation OR ENSURE the reactor is SHUTDOWN.' The COMPLETION TIME is stated as Immediate for both REQUIRED ACTION(S). Please consider a sequence of events (e.g., either restore the channel to operation within an acceptable COMPLETION TIME, OR shutdown).

37.2.3 The COMPLETION TIME(S) for the REQUIRED ACTION(S) A.3.1 through A.3.3 are confusing in that no action is identified to take precedence over another, potentially leaving the operator to make their own assumptions as to the priority of events within one hour of any specified CONDITION.

37.2.4 -CONDITION(S) A4 through A.7 state a series of REQUIRED ACTION(S) that are not sequentially linked. Use of the same COMPLETION TIME for each action is contradictory.

37.2.5 The REQUIRED ACTION(S) A.4.3 and A.4.4 seem to contradict each other.

37.3:!Proposed UT TRIGA TS 3.4 'Safety Channel and Control Rod Operability," Specification B has no associated REQUIRED ACTION(S) or COMPLETION TIME(S).

37.4 Proposed UT TRIGA TS 3.5 'Gaseous Effluent Control," logical "AND/OR" connectors are missing between REQUIRED ACTION(S) C.2.a-C.2.b and C.2.b-C.2.c. COMPLETION TIME(S) are all listed as IMMEDIATE which is contradictory. Please revise providing a clear sequence of the expected steps.

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UT-Austin, License Renewal June 25, 2012 RAI Response Status RAI# Qmtton Riq* DtU ADAMS Accept Comment 37.5 Proposed UT TRIGA TS 3.7 "Fuel Integrity," the COMPLETION TIME listed for all REQUIRED ACTION(S) is IMMEDIATE. Please consider revising the REQUIRED ACTION(S) for Specification A and B to state, "Discharge fuel elements prior to reactor operation."

37.6.1 REQUIRED ACTION(S) A.1 through A.3 are in reverse order. The COMPLETION TIME(S) are all IMMEDIATE which is contradictory.

37.6.2 REQUIRED ACTION(S) B.1 and B.2 are in reverse order.

37.6.3 REQUIRED ACTION(S) C.1 and C.2 are in reverse order. The COMPLETION TIME(S) are all IMMEDIATE which is contradictory. Also, and the CONDITION seems to be improperly stated.

37.6.4 REQUIRED ACTION(S) D.2 and D.3 are in reverse order. The COMPLETION TIME(S) are all IMMEDIATE which is contradictory. A basis to support the established limits in Specification D is not provided. Such information should be discussed in the SAR and then utilized in the TS basis.

38.1 There are no SRs for the DAC or CSC that are listed as SCRAM channels in UT SAR Table 4.6.

38.21There are no SRs for the reactor bay differential pressure for CONDITION A.3 in proposed UT TRIGA TS 3.3 "Measuring Channels."

38.3 Proposed UT TRIGA TS 3.3 "Measuring Channels," contains Surveillance Requirements for the Fuel Temperature Channel and the Upper Level Radiation Monitor but there are no associated LCO specifications 38.4 There are no SRs for the Reactor Power Level scram, the Manual scram, or Fuel Temperature scram to support proposed UT TRIGA TS 3.4 "Safety Channel and Control Rod Operability."

38.5 There are no SRs to support proposed UT TRIGA TS 3.7 "Fuel Integrity," CONDITION C.

39,1 Proposed UT TRIGA TS 5.1.3(1) allows fuel having a stoichiometry of 1.55 to 1.80 in hydrogen to be used in UT TRIGA.

39,2 1) core parameters; 2) conditions for operation of the reactor with damaged or leaking fuel elements; 3) parameters such as maximum core loading, thermal characteristics, physics parameters, etc; and 4) fuel bum-up limits. These design features are not stated in the proposed UT TRIGA TS.

39.3 Please provide a basis for meeting UT TRIGA TS 5.2 "Reactor Fuel and Fueled Devises in Storage," in recommended byANS Standard 15.1, Section 5.4.

39A4 Proposed UT TRIGA TS 5.4 incorporates considerations for experiments into the design features section. These considerations do not meet the regulations of the definition for design features from 10 CFR 50.36.

40.1 ANS/ANS-15.1-2007 Section 6.1 ," recommends tresra inrildig levels and reportng authority. UT TRIGA TS 6.1.1 '$triucture," Figure 6.i1, doenottdescribe the oraiainlstiuctire as described in ANSVANS-15.1-2007, Section 6.1 *Organization.'

40.2 individual or group that shall be assigned responsibility for implementing the radiation protection program using the guidelines of "Radiation Protection at Research Reactor Facilities," ANSI/ANS-1 5.11-1993 (R2004). This individual or group shall report to Level 1 or Level 2. The proposed UT TRIGA TS contains no such section.

40.3 UT TRIGA TS 6.1,3"Staffing," the followng items are not consistent with the g*idanoce 40.4 ANSI/ANS-15.1-2007 Section 6.1.4 "Selection and training of personnel," recommends a section for ensuring that the selection and training of UT TRIGA staff is consistent with ANSI/ANS-15.4-1988. No such section is provided in the UT TRIGA TS.

41 UT SAR does not provide a written statement whether U I is 'owned, controlled, or dominated, Dy an alien, toreign corporation, or foreign government." Please provide a written statement that confirms or denies this status.

42 UT's latest annual financial statements were not included in the UT SAR. Please provide a copy of UT's latest annual financial statements.

43 Assuming review of this application is completed on schedule, and a renewed license granted in FY 2013, please provide the estimated operating costs for each of the FYs 2013 through FY2017.

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UT-Austin, License Renewal June 25, 2012 RAI Response Status RAI # QueonRpIy Date ADAMS Accpt Comment 44.1 !A comparison of the UT TRIGA decommissioning cost estimate to more recently decommissioned research reactors of similar licensed power limit as the UT TRIGA.

44.2 decommissioning cost estimate for the UT TRIGA to meet the NRC's radiological release criteria for decommissioning the facility, which should also include a contingency factor of at least 25 percent. A contingency factor provides reasonable assurance for unforeseen circumstances that could increase decommissioning costs 44.3 Provide a calculation detailing how the rates in Table 15.3 "Escalation Costs." were derived.

44.4 Provide a calculation showing how the escalation factors in Table 15.4 'Calculation Summary," were derived.

45.1 Provide the following information: An updated SOl containing the decommissioning cost estimate in 2013 dollars, and the name of the document(s) governing control of funds 45.2 Provide the following information: Written documentation verifying that the signator of the Sol, Kevin P. Hegarty, Vice President and Chief Financial Officer, is authorized to execute the SOl that binds UT financially.

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