ML16015A052

Univ. of Texas-Austin - Response to Request for Additional Information Regarding Renewal Request, July 31, 2015 Correspondence

ML16015A052
Person / Time
Site:	University of Texas at Austin
Issue date:	12/22/2015
From:	Whaley P University of Texas at Austin
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME7694
Download: ML16015A052 (65)

Text

Deparrnen* of' Mechanical Engineering THE UNIVERSITY OF TEXAS AT AUSTIN Nuclear E'ngineering JbachingLaboratory-'Austin. Ii'xas 78758 512-232-5370" FAX 512-471-4589- httpI/!www, me.utexas.ee'4d-netl/

December 22, 2015 ATT-N: Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 M. Balazik Project Manager Research and Test Reactors Licensing Branch

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) -- July 31, 2015 correspondence Sir:

Attached are responses to request for additional information for 2.4, 4, 5, 12, 16.1, 22.2, 22.4, 22.5, 22.6, 22.7, 27.1, 27.2, 27.2, 27.3, 27.4, 27.5, 27.6, 28.1, 29.1, 29.2, 29.3, 32.3, 32.4, 32.5, 33, 34.1, 34.3, 36.1, 36.3, 36.5, 36.6, 36.7, 36.8, 36.9, 37.1, 37.2, 37.2, 37.2, 37.2, 37.2, 37.3, 37.4, 37.5, 37.6, 37.6, 37.6, 37.6, 38.1, 38.2, 38.3, 38.4, 38.5, 39.1, 39.2, 39.3, 39.4, 40.2, and 40.4.

In review, two issues from previous submissions were identified. First, the September 9, 2013, request identified a response to RAI 20.2, while the response applied to RA1 20.1. Second, the October 23, 2015 request identified ML15114A433 as answering RAI 18 but neglected that the document also resolves the related RAI 36.9.

We are requesting 90 days to complete the work (RAI 19 and the financial questions are still being addressed). It is likely that these items will be completed before the requested time, and the responses will be submitted on.

Please contact me by phone at 512-232-5373 or email whalev@~mail.utexas.edu if you require additional information or there is a problem with this submittal.

Thank you, Associate Director Nuclear Engineering Teaching Laboratory The University of Texas at Austin I declare under penalty of perjury that the foregoing is true and correct.

pot Executed on December 22, 2015 Steven R. Biegalski NETL Director

RAI 12

UT SAR Section 4.5.4, Subsection B provides Figure 4.22 for the power within a fuel element.

the NRC Staff notes that the power distribution in the figure continues to the center of the fuel element indicating that this curve is not applicable to stainless steel clad fuel that has a zirc rod in the center. Please confirm and revise accordingly.

RESPONSE

Neutronic analysis has been revised, and appropriate figures will replace the original.

RAI 16.1 The pooi dimensions of a "tall tank formed by the union of two half-cylinders with a radius of 6% ft.

(1.9812 m) with 6% feet separating the half-cylinders," appears to be inconsistent with the stated tank nominal water volume of 40.57 cubic meters. Please confirm and revise accordingly.

RESPONSE

A sketch of the pool surface area:

146 R0.9906 (X2l The surface are of the pool is therefore 5.056 m2 POOL VOLUME Elevation Volume above Level (in) Pool Floor (in 3 )

Top of Tank 8.236 41.64 Normal water level 8.179 41.35 Minimum (TS) water level 6.50 32.86 Core Top 0.51 2.58 Core Bottom 0.2519 1.27 With a reflector height of approximately 0.54 m, a reflector radius of approximately 0.6 m, and the hexagonal core metal to water ratio of approximately 1:3, the volume occupied by the reflector and core is on the order of 0.5 m and can be neglected in evaluating pool water volume. Therefore during normal operations pool volume is 41.35 in3 , with approximately 38.7 in3 over the core elevation.

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 19.1 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/p4 1.4 and not as it is stated in the UT SAR.

RESPONSE

The correct form of the correlation is AH = D (Vs/i)'"4

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

RESPO NSE The 45.72 cm value is a diameter, the 0.4012 m2 value the area of the flow from the stack.

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

RESPONSE

The height of the stack required additional work to account for possible building wake effects. Analysis has been revised as indicated in response to RAI 22.

RAI 2.4 UT SAR Section 4.2.1 provides Figures 4.2A and 4.2B. The source of this information is not referenced nor is their applicability to the particular operating conditions and fuel depletion of UT TRIGA demonstrated. Please demonstrate the applicability of these figures to the UT TRIGA.

RESPONSE

The figures will be removed.

RA1 21.2 UT SAR Section 9.4.2 states, "A 5-tonne crane is used in conjunction with fuel handling tool and the transfer cask to allow remote handling of irradiated fuel." Please describe the physical or administrative precautions employed to minimize the potential for fuel or core damage due to malfunction, such as loss of electrical power, or dropped loads.

RESPONSE

A loss of electrical power to the crane engages a mechanical brake. Heavy loads are not lifted over the control rod drives.

RAI 22: The guidance in NUREG-1537 Section 11.1.1, "Radiation Sources," requests that the licensee include airborne dose information for characterization of 41Ar, including providing best estimates of the maximum annual dose and the collective dose for the major radiological activities for the full range of normal operations for facility staff and members of the public.

RAI 22.2: UT SAR Section 11.1.1.1.1 describes the production of 41Ar and provides very conservative estimates of the concentration, but does not provide values for the occupational dose. Please provide the 41A occupational exposure including stay times and the effect of ventilation, and how these compare to the limits of 10 CFR Part 20 and the commitments of the UT TRIGA ALARA program.

RAI 22.3 UT SAR Section 11.1.1.1.1 does not describe the whole body dose to facility staff. Please provide a discussion of facility worker doses, and whether these doses are ALARA.

RAI 22.4: UT SAR Section 11.1.1.1.2 provides a conservative estimate of offsite 41Ar 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.

RAI 22.5: UT SAR Section 11.1.1.1.2 provides a discussion of use of the CAP88 PC computer program to estimate the dose to the maximally exposed individual. However, no information is provided regarding the location of this individual, whether the location represents the nearest residence, or whether the location is at a location of special interest. 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.

RAI 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.

RAI 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.

RESPONSE

RAI 22.2 See attached, Section 2 RAI 22.3 See attached, Section 2 RAI 22.4 See attached, Section 2.4 RAI 22.5 See attached, Section 2.3 RAI 22.6 There is no discrepancy; limiting a maximum release of 100 Ci per year results in acceptable dose.

RAI 22.7 The predicted dose to the maximally exposed offsite individual in Section 11.1.1.1.2 is actually the predicted maximum public dose (see response to question RAI 22.5).

RAI 22 REPONSE: ATTACHMENT 1 This analysis is predicated on continuous operation at the maximum licensed power level, while the operating schedule is virtually never 100% of the available time and a large fraction of the time is less than 100% power level.

Normal operation with the purge system removes most of the Ar 41 from experiment facilities and the surface above the pool. Experiments in 2015 demonstrated Ar 41 concentrations on the top deck of the reactor pedestal to be on the order of 0.12% of the concentrations in the space above the pool and below the pool grating. At the base of the reactor there was no detectable Ar 41 except for very small concentrations near BP3. The Ar 41 HVAC stack effluent (exhaust from reactor bay atmosphere) is a small fraction (8.2X10 4 ) of the purge system effluent, indicating that the argon purge is removing most of the Ar 41 generated during reactor operation (when both systems are operating).

1. PRODUCTION OF ARGON 41 AND SPECIFIC ACTIVITY There are three production sources for Ar 41, argon 40 dissolved in water flowing through the core, in beam port air, and air inside the rotary specimen rack (RSR). Beam ports are sealed when not supporting an experiment and when used generally have equipment installed that displaces air. The RSR is normally unvented during operation. The RSR insertion port is normally sealed during operation, and the operating shaft has minimal clearance. Therefore the Ar 41 production with the largest potential for introducing Ar 41 into the reactor bay environment is activation of naturally occurring argon in the water flowing through the reactor. The contribution from each source, the ventilation of individual contributions, and the effects of operation of the argon purge system and the HVAC system are considered.

With no removal other than radioactive decay, the argon 41 production rate at equilibrium (Table 2, taken from the SAR Chapter 11 with pool surface calculated separately) is equal to the decay rate (or activity):

Ni,eq" Zi=Z Where Ni,eq represents the number density of Ar 41 at equilibrium with k the decay constant (1.05x10 4 s-1), Xris the macroscopic cross section for neutron absorption of Ar 40, and t* is the neutron flux.

Activity and volume for the pool water and the experiment facilities from the proposed UT SAR are provided inTable 1.

1.1 Production of Argon 41 in the Pool and Transport to Air The Ar 41 is assumed to be distributed in equilibrium with argon. The air saturation-concentration of Argon at the surface of fresh water at standard pressure and temperature (STP) is 0.556 mg/L1 . For the normal temperature range of UT TRIGA pool water at full power (approximately 25°C to 30°C), solubility is slightly less at 0.4841 +/-3% mg/L.

Variability forsolubility with small changes in pressure near atmospheric pressure is minimal; however, the weight of water from the surface of the pool to the bottom causes hydrostatic pressure to vary SDissolved Gas Concentration in Water, 2 nd Ed., J. Colt (J. Colt), ISSBN 978-0-12-415916-7

significantly with depth. Solubility as a function of depth at STP increases 0.55 (mg/L) per meter from the surface to the floor of the pooi. Water density at 30°C is 3% lower than at 20°C, so that hydrostatic pressure at depth varies only slightly over a range from the reference temperature to normal pool temperature at full power.

The weight fraction of argon in air is 0.934%. The density of air at sea level is nominally 1.225x10-3 g/cm3 , so the density of argon in the air above the pool surface is taken as 1.144x10-s g/cm 3. The ratio of the argon concentration in air above the pool surface to the argon concentration in pool water at the surface is:

1.144x10-s g/cm3 *lxlO 3 mg/g, lxlO3 cm 3 /L

= 23.63 0.4841 mg/L If ventilation is not considered, the Ar 41 concertation above the pool is a factor of 23.63 higher than the Ar 41 at the surface of the pool. The average concentration of argon in fresh water at STP from 0 to 8 meters is 0.776 mg/L, and at the surface is 0.556; the surface concentration is 71.6% of the average concentration. Therefore the Ar 41 concentration above the pool surface is '19.92 times the average Ar 41 concentration in the pool.

1.2 Air Space above the Pool Surface The air above the pool is bounded by the pool surface and Plexiglas on the bottom surfaces of the pool grating (Fig. 1). A rectangular space is formed above the pool tank, and an oval space is formed by the pool surface and the top of the pool tank. Total volume of the air space is 3.28x10 5 cm 3.

R0.9906 (X2J v I .

0.4609

--..T*.- I I

. f I  !

II I -0.2.159 VOLUME:

3.2800,58 mA3 Figure 1: Geometry of the Air Space above the Pool 1.3 Summary of Production Source Terms

Information taken from the proposed SAR for source term production and the volumes of experimental facilities is summarized in Table 1. Pool surface air calculations are based on information above. Pool surface air activity is the specific activity at the pooi surface (16.92 times the average poll water activity),

and total activity in the air above the pool is the air volume times the specific activity.

Table 1: Ar 41 Production Source Terms Cmoet Co n p nInA Activity Volume

,__3,.

Sp.

,n_

Activity

• DqJ [cm I~q/cm Pool water 2.10E+09 5.77E+07 3.64E+01 Pool surface air 2 1.65E+09 2.68E+06 6.16E+02 Beam Port air 9.7OE+09 5.90E+05 1.64E+O4 RSR air 3.30E+10 3.30E+04 1.00E+06 TOTAL AIR 4.43 E+10 3.80E+06 NA 1.3 Ventilation Effects The time rate of change in the concentration of Ar 41 (dNt) isclultd dNi(t)

Where Z is the macroscopic cross section for neutron absorption of neutrons in argon 40,

• is the neutron flux, 2, is the radioactive decay constant (0.693/110 min-1 , or 1.05x10-4 s'), and )Lv is the fractional removal constant. The time dependent concentration is therefore:

N1(t) -

Where 2, is the radioactive decay constant and 4, is the removal constant from ventilation, calculated by the flow rate divided by the volume. For convenience, the term with the removal constant will be referred to as "reduction factor." The equilibrium specific activity when ventilation is used (ar) is related to the equilibrium activity with no ventilation (ac) by:

ti~+ *l There are two ventilation systems that create removal constants, the argon purge system and the reactor bay confinement ventilation system (HVAC). These systems are used in three configurations for reactor operation: (1) both systems operating (2) Purge system only, and (3) HVAC system only. Each configuration defines the flow through the reactor bay and experiment facilities. SAR Chapter 9 provides flow rates associated with the argon purge and reactor bay ventilation systems. SAR Chapter 11 indicates the volumes of the reactor bay and experiment facilities, with the exception of the air space 2 Specific activity based on activity of Ar 41 in the pool (A.N41) and ratio of Ar 41 at pool surface to Ar 41 in air above pool surface in 1.1; total activity is specific activity times volume of air above pool. Air ye the pool is based on Fig 1.

above the pool (previously calculated). Ventilation flow rate, component volume, transport coefficient, and reduction factor are compiled in Table 2 for the two ventilation systems.

Table 2: Physical Parameters Reducing Concentration Vetiato FowRaeComponent Removal Reduction VenilaionFlwnRte Volume Constant Factor Cfm cm 3/s cm3 s Dilution 525 2.48E+05 NA NA NA Bay (HVAC) 7200 3.40E+06 4.12E+09 8.25E-4 1.13E-1 Pool surface 3 525 2.48E+05 2.68E+06 9.26E-02 1.13E-03 Beam Ports 20 9.44E+03 5.90E+05 1.60E-2 6.52E-3 RSR 4 1.89E+03 3.30E+04 5.72E-2 1.83E-3 1.4 Specific Activity (Ar 41) in Reactor Bay For the cases where the purge system is operating, the Ar 41 in the reactor bay atmosphere is insignificant. For the case where only the HVAC system is operating, all of the activity enters the reactor bay atmosphere and is removed through the stack by HVAC flow, so that the Ar 41 concentration in the reactor bay (aBQ*) is calculated as the sum of the activities (Ar, where A is the activity for the contribUting component x) divided by the total volume of the reactor bay (Vx where x is the volume of the reactor bay, air above the pool, in the beam port and in the RSR) modified by the reduction factor associated with HVAC flow rate (-P-VAc):

Apoo1 +/- ABp +/- ARSR asy=VBay + Vpooi air +/- VBP +/- VRSR RFHvAc Calculations of specific activity in the reactor bay with only the HVAC system is operating is provided in Table 3 for conditions with individual purge system components contributing to the source term.

Table 3: Ar 41 in Reactor Bay (HVAC, Bq/cm 3)

Pool surface air YES YES YES Beam Port air YES YES NO RSR air YES NO NO Activity (Bq/cm3) 1.21 0.31 0.05 1.5 Specific Activity (Ar 41) in Stack Effluent The configurations that use the purge system (purge only, and purge in conjunction with the HVAC system) have contributions to purge system flow from dilution flow (from the reactor bay), flow from the pool surface, flow from the beam ports, and flow from the RSR exhausted as purge flow. The configuration that uses only the HVAC system discharges reactor bay air from HVAC flow.

SCalculated in Table 1, with remaining volumes are taken from the SAR

1.5.1 Purge System (Only) Operating Where aeff,purge is the contribution of argon 41 from the purge system to the stack effluent ae,x is the equilibrium specific activity with no ventilation for component x, and v= is the ventilation flow rate for component x:

aepoZVoo *RF~a +/- aeBp *VBp.*Fp+aSR"RR"RFs aeff,purge - a Fp+aRRV= Fs Vdizution ¢ V2Poot +} VBp +] 12RSR 1.5.1 Purge System and HVAC Operating For the case where the HVAC and the purge system are both operating, stack effluent is weighted average of the HVAC flow and the purge system flow:

_aeff ,purge .Veff,purge 12 aejjf stack~ Yeff,purge -+/- HVAC 1.5.2 HVAC (Only) Operating For the case where only the HVAC system is operating, effluent activity is the reactor bay specific activity previously calculated (aBe,).

1.5.3 Summary of Effluent Concentration Calculations Calculations of specific activity in the reactor bay with only the HVAC system is operating is provided in Table 3 for conditions with individual purge system components contributing to the source term.

Table 4, Ar 41 in Stack Effluent (Bq/m 3)

Pool surface air YES YES YES Beam Port air YES YES NO RSR air YES NO NO Purge System (Only) 9.16 2.35 0.35 Purge System & HVAC 6.5E-01 1.7E-01 2.5E-02 HVAC (Only) 1.2 0.31 0.05

2. CONSEQUENCE ANALYSIS 2.1 Worker Doses Only the configuration with the HVAC system operating (and purge secured) contributes significant concentrations of Ar 41 to the reactor bay atmosphere. The 10CFR20 Derived Air Concentration for Ar 6

41 is 3xlOY pCi/ml (1.11xlO-2 Bq/ml, 1.11xlO4 Bq/m 3). Exposure to a DAC for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> will result in a

5 rem dose. Full power, steady state reactor operations with the purge system operating are unrestricted with respect to radiation exposure to Ar 41. Table 5 indicates the dose rates that will occur with the equilibrium Ar 41 concentrations calculated to occur at full power for each experiment use configuration (indicated by YES in Table 5).

Table 5: Impact on Worker Doses, 1.1 MW Operation, HVAC Only Pool surface air YES YES YES Beam Port air YES YES NO RSR air YES NO NO Dose Rate 274 mrem/h 70 mrem/h 10 mrem/h Stay Time 18.3 h 71.4 h 491.3 h With dose rates in the reactor bay associated with only the HVAC system operating, a Radiation Work Permit would be required to control worker exposure if all beam ports are being used in a completely open condition and the RSR is open to reactor bay atmosphere (extremely unlikely, as previously noted).

If beam ports are completely open but the RSR is closed, routine personnel monitoring is adequate to control worker doses. In both these cases, Area radiation monitors would alert workers to radiation dose rates. With beam port and RSR secured, routine personnel monitoring is adequate to control worker doses.

2.2 Effluent Activity The effluent limit is 1x10 8 I*Ci/ml (3.70x10-4 Bq/ml). The effluent limit is based on an annual dose of 50 mrem, 1/ of the maximum permitted dose to the general public. A constraint is imposed by 10CFR20 for no more than 10 mrem/year exposure from effluents.

2.3 Offsite Exposure to Ar 41 Effluent Actual releases are a small fraction of the possible values calculated under the extremely conservative assumptions. Applying CAP88- PC to actual 4lAr release rates measured over the past several years predicts an annual dose to the maximally exposed individual of less than 0.02 mrem which is well within the 10 CFR Part 20 limits and the NETL ALARA goals.

CAP88-PC uses a modified Gaussian plume model to estimate the average dispersion of released radionuclides and appropriate dose conversion factors to calculate expected doses from these releases.

One option the computer program offers is to determine the location and the dose to a hypothetical maximally exposed individual. Individuals present at any other locations would receive a dose no greater than that of the maximally exposed individual. A CAP88-PC calculation was performed that indicates a dose response to effluent activity of 5.95 mrem/(Bq/cm 3 ) for the maximally exposed individual. CAP88- PC determined that the maximally exposed individual was located 200 meters north-northwest of NETL. This location corresponds to the parking lot of a bank and is not a location of special interest (i.e., continual occupancy). For reference, the nearest residence is approximately 500 meters north-northwest of NETL. Table 5 summarizes the effluent concentration, annual exposure for continuous 1.1 MW operations, the dose rate at that exposure level, the ratio of the dose to the 10 mrem constraint and the 100 mrem limit, and the number of hours of operation that reaches the constraint and limit for each configuration.

Table 6, Dose/Dose Rates based on CAP88-PC from Effluents Effl. Exposure Dose Rate Constraint-Based Limit-Based Configuration Bq/cm 3 Mrem mrem/h Ratio Hours Ratio Hours Purge, all sources 9.16 54.5 6.22E-03 0.18 1608 0.5451 NA Purge, Pool & BP 2.35 14.0 1.59E-03 0.72 6279 0.1396 NA Pool, Pool only 0.349 2.1 2.37E-04 4.81 NA 0.0208 NA HVAC, all sources 0.65 3.9 4.41E-04 2.59 NA 0.0386 NA HVAC, Pool & BP 0.17 1.0 1.13E-04 10.11 NA 0.0099 NA HVAC, Pool only 0.02 0.1 1.68E-05 67.94 NA 0.0015 NA HVAC, all sources 1.2 7.2 8.24E-04 1.38 NA 0.0723 NA HVAC, Pool & BP 0.31 1.8 2.11E-04 5.41 NA 0.0185 NA HVAC, Pool Only 0.05 0.3 3.07E-05 37.20 NA 0.0027 NA Where the constraint-based ratio exceeds 1.0, the receptor dose limits can be met either by limiting hours of operation in the configuration or limiting the total annual effluent release. Where R is the total activity released, aeff is the effluent activity concentration, Veff is the effluent flow rate, and T'S is the number of seconds in a year, thetotal annual release can be calculated as:

R = aeff *Veff

• 7' In Table 7, total annual release is calculated for each configuration based on effluent concentration and flow rate. The limiting release is the calculated annual release divided by the limit based ratio of Table

5. There are 2 abnormal configurations identified where continuous operations at 1.1 MW have the potential to cause more than 10 mrem in a year, and none will cause more than 100 mrem.

Table 7, Annual Ar 41 Release Bq/cm 3 Bq Ci Purge, all sources 9.16 7.49E+13 2.02E+03 Purge, Pool & BP 2.35 1.92E+13 5.19E+02 Pool, Pool only 0.35 2.85E+12 7.72E+0l HVAC & Purge, all sources 0.65 1.04E+13 2.81E+02 HVAC & Purge, Pool & BP 0.17 2.66E+12 7.19E+01 HVAC & Purge, Pool only 0.02 3.96E+11 1.07E+01 HVAC, all sources 1.21 9.50E+12 2.57E+02 HVAC, Pool & BP 0.31 2.43 E+12 6.57E+01 HVAC, Pool Only 0.05 3.53E+11 9.55E+00 2.4 Building Wake Effects Building wake effects can limit atmospheric dispersion of effluent at the building perimeter. The building wake equation as used in the original UT-Austin TRIGA Safety Analysis Report (May 1991) and the U.S. Geological Survey TRIGA Safety Analysis Report (May 2008, Section 11.1.1.4) as presented in

DOE/TIC-27601, Flow and Diffusion Near Obstacles Cheater 7, "Atmospheric Science and Power Production" (D. Rnaderson, U.S. Department of Energy; 1984). At a wind speed of 1 rn/s the reduction in Ar 41 concentration in a building wake condition will be reduced by 35.11 (effluent compared to exposure), and with at 4 rn/s (UT TRIGA site annual average wind speed) the reduction will be 140. The dose rate (DR) associated with the effluent AR 41 specific activity is calculated:

50torero/(364.25.24) Seii tvt DRmrem = Seii tvt h 3.70xlO-4 Bq/cm 3 Table 8: Effluent Concentrations and Building Wake Dose Rates, 1.1 MW Sak1 rn/s 4 m/s Effl. Dose Dose Wake Rate Wake Rate Bq/cm 3 Bq/cm 3 mrem/h Bq/cm 3 mrem/h Purge, all sources 9.16 2.61E-01 4.02 6.54E-02 1.01 Purge, Pool & BP 2.35 6.68E-02 1.03 1.68E-02 0.26 Purge, Pool only 0.35 9.94E-03 0.15 2.49E-03 0.04 HVAC & Purge, all sources 0.65 1.85E-02 0.28 4.64E-03 0.07 HVAC & Purge, Pool & BP 0.17 4.73E-03 0.07 1.19E-03 0.02 HVAC & Purge, Pool only 0.02 7.05E-04 0.01 1.77E-04 0.00 HVAC, all sources 1.21 3.46E-02 0.53 8.68E-03 0.13 HVAC, Pool & BP 0.31 8.85E-03 0.14 2.22E-03 0.03 HVAC, Pool Only 0.05 1.29E-03 0.02 3.23E-04 0.00 In the normal configuration (HVAC and purge systems, purge venting pool surface) operations at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, 7 days a week at full power do not have the potential to exceed exposure limits considering maximum building wake effects. Normal operational schedule, limits on operation with only one system operating, normal weather conditions, and routine radiological monitoring are adequate to assure operations in abnormal conditions are controlled to meet exposure limits.

2.5 Comparison with Measurements The argon purge system is instrumented to measure Ar 41 effluent. The Argon monitor is calibrated annually, the total argon activity and average argon effluent activity (normalized to energy generation).

From 2007 to 2014 the Ar 41 activity per kW-h was an average 20.17 iiCi/kW-h (2.69x106 Bq/MW-s) with a standard deviation of 3.69. For the typical flow rate noted in the SAR (0.52 m3/s, or 5.2x10 5 cm 3/s) at 0.95 MW the effluent concentration of AR 41 is therefore 4.91 Bq/cm 3. Normal configuration is 1 beam port and the pool surface vented, with the purge and HVAC systems operating. Table 4 indicates when the purge system and the HVAC are operating with the pool surface and 4 beam ports vented, stack effluent is 0.17 Bq/cm 3. The calculated value of stack effluent for the configuration using both the HVAC and argon purge systems 35% of the measured effluent concentration.

RAI 27

The guidance in NUREG-1537 Section 13.1.1, "Maximum Hypothetical Accident," requests that the licensee provide a maximum hypothetical accident (MHA) and demonstrate that it bounds all potential credible accidents at the facility. Under this guidance the MHA for TRIGA reactors is the failure of one fuel element in the air with the release of gaseous fission products. The purpose of this analysis is to ensure that this accident would not lead to unacceptable radiological consequences to the occupational and non-occupational workers and the environment.

RAI 27.1 UT SAR Section 13.3 analyzes a fuel element failure in the open air of the reactor bay. The analysis provides fission product inventory for a rod power of 3.5 kW which is not consistent with using a saturated inventory in the hottest rod for 1.1 MW operation. Please provide an analysis of the MHA for 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 confirmation of these results.

27.2 UT SAR, Section 13.3 provides a discussion of the atmospheric dispersion employed and identifies the various parameters and assumptions used to determine the concentrations of nuclides at the nearest site boundary. For the case when the reactor bay ventilation is secured and the auxiliary purge system is used to discharge the reactor bay effluent, the UT SAR describes an elevated release through the building stack.

27.2.1 UT SAR, Section 13.3 (p. 13-19), the building stack is located on the roof of the reactor building and its exit is at about 14 feet above the roof leading to a total height of about 63 feet above the ground level that surrounds the facility. The calculations are then performed for distances from 10 to 100 meters from the building. Because, the reactor building is both tall and wide, any release from the stack could be accumulated in the building wake. Therefore, the applicability of the assumption of elevated release is appears inaccurate. Please justify the use of the elevated release values for dose estimates at nearby distances from the facility.

27.2.2 In addition, if there is an error in the correlation used for the plume rise (see RAI 0), 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 For the dispersion calculations of ground releases using RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Regulatory Position 1.3.1, the licensee uses a building wall cross section area, which appears to be 432 m2. 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.

RESPONSE

27.1 - See Analysis following; (a) New fission product inventory assumption based on maximum burnup of 50% 235U (b) Doses to workers and individuals in unrestricted areas calculated (c) Airborne activity within DAC in the reactor bay, and effluent limits off site. Therefore maximum doses within 10CFR20 limits.

27.2.1/2 - New analysis does require atmospheric dispersion.

27.3 - New analysis does not require analysis of building overpressure.

27.4 - New analysis does not require atmospheric dispersion modeling ANALYSIS The maximum hypothetical accident is a release of radioactive noble gas and halogen fission products from a TRGIA fuel elements following discharge. The maximum fission product inventory will occur in a fuel element with the maximum burnup. The maximum burnup is taken to be the burnup that results in a loss of 50% of the initial uranium 235 mass (from 38 grams to 19 grams). This is extremely conservative as the TRIGA fuel temperature reactivity deficit associated with operation at power does not allow support under these conditions.

Depletion calculations using the SCALE T-6 sequence was used to determine burnup and the fission product inventory. A SCALE model of the TRIGA core was configured with a two fuel material composition sets, one representative of a single element to be depleted and the other representing the remaining elements in the core.

A set of SCALE (T-6 depletion) calculations was performed to deplete all elements in the core, generating a fission product inventory for the larger set of elements. A 50% burn interval at 1.5 MW was evaluated, reducing the u235 mass in the single element from 38 grams to approximately 19 grams. In determining the 50% burn, the number of fuel elements was adjusted to result in a calculated flux similar to the nominal UIT TRIGA full power flux. The uranium, transuranic, and fission product concentration were used to develop a material composition simulating the core average at the end of the interval for the 50% single element burn.

The SCALE model was configured to a single fresh fuel element and the remainder of elements at the core average at the end of the 50% burn interval. Radioactive noble gas and halogen activity was calculated for the 50% burn. A similar calculation was performed except that the constant flux option was used. The maximum value for the activity of the isotopes from the two calculations (constant power and constant flux) was taken as the source term for the radioactive noble gas and halogen inventory for a single fuel element at the maximum burnup.

Using a release fraction of lX10-4, and the free volume of the rector bay (4120 in3 ), the concentration of the activity of each isotope (A,, in IICi/ml) in reactor bay atmosphere based on the source term for each isotope (C, ,in Curies) is calculated as:

Ai-4.12x10_9 The average activity of isotopic concentration (A,,(t)), where Ai is the isotope decay constant, over some time interval (t) following the release of isotopes from a fuel element into the reactor bay is calculated:

At)=4.12x10- 9 ' i t Each isotope has limits (based on continuous activity concentrations over one year) on activity concentration for exposure of workers (Derived Air Concentration, DAC) and the general public (Effluent Limit). For mixtures of isotopes, compliance with the limits is demonstrated if the sum of the ratio of each activity concentration to its individual limit is less than unity. Since the hypothetical accident is not a continuous process, the concentration of each isotope is normalized over a year following the release.

The ratio of the average activity ( <A1>) for the year following release of the source term from a fuel element into the reactor bay for occupational exposure (2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />) to the isotopic DAC (DACI) is calculated as:

(At) _ C'.10- (1 - e-iYears) 2000 9

DACi 4.12x10- Ati Years Yearht The fraction of average activity over a year to the effluent limit (EL) is calculated:

(A 1) _C 1

• 10- (1_-e-illYears)

ELi 4.12x10- 9 2*Years The sum of the ratios of concentration to limit is 7.52x10-5 for occupational exposure, and 3.30x10-2 for effluents. Since a exposure to a DAC for a year results in 5 rem, if the releases is completely contained in the reactor bay and an individual works 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> in the bay then a dose of 0.4 mnrem will result. Since exposure to an effluent limit for a year will result in 50 mnrem exposure, an individual exposed at the exit of the reactor bay ventilation for a year following the hypothetical release will receive a dose of 1.65 mrem.

Table X, Summary of MHA Data Fuel Eff release reactor 1 year DAC *Eff Limit Isotope A s1 DAC from bay ave Frcin rato (s)Lmt fuel Ci IICi/ml IICi/ml br83 1342 8.02E-05 3E-05 9E-08 1.34E-1 3.26E-11 1.29E-14 9.79E-11 1.43E-07 br84m 51 1.93E-03 1E-07 1E-09 5.13E-3 1.24E-12 2.05E-17 4.67E-11 2.05E-08 br84 2344 3.64E-04 2E-05 8E-OS 2.34E-1 5.69E-11 4.96E-15 5.66E-11 6.20E-08 br85 3373 3.98E-03 1E-07 1E-09 3.37E-1 8.19E-11 6.51E-16 1.49E-09 6.51E-07

Table X, Summary of MHA Data FulEt release reactor 1 year DAC Eft Limit Isotope Fue (sA DAC Ef from bay ave Frcin rato C* ~Limit fuel Ci ~id/ml iidi/mlI rcin Fato i13 1 8167 9.99E-07 2E-08 2E-10 8.17E-1 1.98E-10 6.28E-12 7.17E-05 3.14E-02 i132m 56 1.39E-04 4E-06 3E-08 5.57E-3 1.35E-12 3.09E-16 1.76E-11 1.03E-08 i132 12130 8.39E-05 3E-06 2E-08 1.21 2.94E-10 1. 11E-13 8.46E-09 5.56E-06 i133 17440 9.24E-06 1E-07 1E-09 1.74 4.23E-10 1.45E-12 3.31E-06 1.45E-03 i134m 1157 3.12E-03 1E-07 1E-09 1.16E-1 2.81E-11 2.85E-16 6.50E-10 2.85E-07 i134 20200 2.20E-04 2E-05 6E-08 2.02 4.90E-10 7.06E-14 8.06E-10 1.18E-06 i135 16420 2.93E-05 7E-07 6E-09 1.64 3.99E-10 4.32E-13 1.41E-07 7.19E-05 i136m 3366 1.47E-02 1E-07 1E-09 3.37 E-1 8.17E-11 1.76E-16 4.01E-10 1.76E-07 i136 6740 8.31E-03 1E-07 1E-09 6.74E-1 1.64E-10 6.24E-16 1.42E-09 6.24E-07 kr85m 3254 4.30E-05 2E-05 1E-07 3.25E-1 7.90E-11 5.82E-14 6.64E-10 5.82E-07 kr85 9 2.22E-10' 1E-04 7E-07 8.60E-4 2.09E-13 2.08E-13 4.74E-10 2.97E-07 kr87 6275 1.51E-04 5E-06 2E-08 6.28E-1 1.52E-10 3.19FE-14 1.45E-09 1.59E-06 kr88 8489 6.78E-05 2E-06 9E-09 8.49E-1 2.06E-10 9.63E-14 1.10E-08 1.07E-05 kr89 10820 3.67E-03 1E-07 1E-09 1.08 2.63E-10 2.27E-15 5.18E-09 2.27E-06 kr90 11580 2.15E-02 1E-07 1E-09 1.16 2.81E-10 4.14E-16 9.45E-10 4.14E-07 kr9 1 7940 8.09E-02 1E-07 1E-09 7.94E-1 1.93E-10 7.55E-17 1.72E-10 7.55E-08 xe131m 80 6.77E-07 4E-04 2E-06 8.01E-3 1.94E-12 9.09E-14 5.19E-11 4.55F-08 xe133m 184 3.65E-06 1E-04 6E-07 1.84E-2 4.48E-12 3.89E-14 8.87E-11 6.48E-08 xe133 17080 1.53E-06 1E-04 5E-07 1.71 4.15E-10 8.59E-12 1.96E-08 1.72E-05 xe135m 2224 7.55E-04 9E-06 4E-08 2.22E-1 5.40E-11 2.26E-15 5.74E-11 5.66E-08 xe135 375 2.11E-05 1E-05 7E-08 3.75E-2 9.09E-12 1.37E-14 3.12E-10 1.95E-07 xe137 15870 3.03E-03 1E-07 1E-09 1.59 3.85E-10 4.03E-15 9.21E-09 4.03E-06 xe138 16030 8.20E-04 4E-06 2E-08 1.60 3.89E-10 1.50E-14 8.57E-10 7.51E-07 xe139 12560 1.75E-02 1E-07 1E-09 1.26 3.05E-10 5.53E-16 1.26E-09 5.53E-07 xel40 8948 5.10E-02 1E-07 1E-09 8.9SE-i 2.17E-10 1.35E-16 3.08E-10 1.35E-07

RAI 27.5 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 (Assumption f on page 11-28 of the 1991 SAR).

RESPONSE

There are substantial differences between the current methodology and that used in the 1991 UT SAR, including:

(1) The 1991 UT SAR assumes continuous operation at 1.5 MW for 4-years, the current analysis assumes operation until the fuel is depleted to 50%.

(2) The 1992 UT SAR in 11.3.2 assumes a 100% release fraction from a TRIGA fuel element for noble gases and halogens, the current analysis assumes a conservative 1X10-4 release fraction based on NUREG/CR-2387 (PNL-4028).

(3) The 1991 UT SAR calculates off-site doses based on atmospheric diffusion models; the current analysis calculates worker and off-site doses based on 10CFR20.

RAI 27.6 27.6 SAR Appendix 13.1, SCALE 6.1 input file, cites an input value 1.6 for the weight fraction of the ZrH 1 .6 U fuel. Is this input value for the weight fraction of hydrogen in the fuel? Please confirm and revise accordingly.

RESPONSE

SCALE input values (mass fractions) are calculated based on assay values. Weight fractions have been recalculated in updated analysis.

RAI 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.

RESPONSE

New analysis was submitted for thermal hydraulic and neutronic analysis, but this was not previously identified as a response to RAI 28.

RAI 32.3

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 ANSl/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.)

32.3 The TS defines the term "immediate" as, "Without delay and not exceeding one hour" and includes an attached note which states "IMMEDIATE permits activities to restore required conditions for up to one hour; this does not permit or imply either deferring or postponing the action." Please revise to the following: when IMMEDIATELY is used as a COMPLETION TIME, The REQUIRED ACTION should bepursued without delay and in a controlled manner.

RESPONSE

The definition will be changed to:

IMMEDIATE Without delay, and not exceeding one hour.

NOTE:

When IMMEDIATE is used as COMPLETION TIME, the REQURIED ACTION should be pursued without delay and in a controlled manner

RAI 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.

RESPONSE

IOCFR50 -- (2) Limiting conditionsfor operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

The definition of reactor shutdown uses a minimum reactivity which is the acceptable minimum allowed subcritical condition.

RAI 32.5

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.)

32.5 The regulatory guidance of NUREG-1537 Appendix 14.1 states that all controlrods must be inserted to achieve REACTOR SECURED MODE. 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.

RESPONSE

ANSI/AN5-15.1-2007 defines:

reactorsecured: A reactoris secured when (1) Either there is insufficient moderator availablein the reactor to attain criticality or there is insufficient fissile materialpresent in the reactorto attain criticalityunder optimum available conditions of moderation and reflection:

(2) Or the following conditions exist:

(a) The minimum number of neutron absorbingcontrol devices is fully inserted or other safety devices are in shutdown position, as requiredby Technical Specifications (b) The console key 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 decouple d from the control rods (d) 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

' The proposed definition as submitted is:

REACTOR SECURED MODE The reactor is secured when the conditions of either item (1) or item (2) are satisfied:

(1) There is insufficient moderator or insufficient fissile material in the reactor to attain criticality under optimum available conditions of moderation and reflection (2) All of the following:

a. At least three control rods are fully inserted

b. The console key is it the OFFposition and the key is removedfrom the lock

c. No work is in progress involving core fuel, core structure, installed control rods, or control rod drives (unless the drive is physically decoupledfrom the control rod)

The limiting shutdown margin is designed to "provide confidence that the reactorcan be made subcriticalby means of control and safety systems startingfrom any permissible operating condition and with the most reactive rod in the most reactive position..." Since the UT reactor has 4 control rods, the shutdown margin provides confidence that the reactor can be made subcritical with three control rods.

Therefore, the phrase "The minimum number of neutron absorbing control devices is fully inserted or other safety devices are in shutdown position, as required by Technical Specifications" for the UT TRIGA reactor is explicitly three control rods.

With three control rods fully inserted, there is confidence that the reactor can be maintained subcritical.

RA1 33 The basis provided in support of the TS 2.1 references Chapter 4, Section 4.2.1 Z which does not exist.

Please discuss this error and/or revise accordingly.

RESPONSE

The information will added to 4.2.1 (3) and referenced appropriately in the basis statement.

RAI134.1 The basis provided in support of the UT TRIGA TS 2.2 references Chapter 4 Section 4.6 B which does not exist. Please provide a basis for the LSSS.

RESPONSE

New analysis was submitted for thermal hydraulic and neutronic analysis for other RAIs, and the nexus to this RAI not previously identified.

RA1 34.3 UT TRIGA TS 2.2, B.2 refers to the statement "verify the measurement value is not correct."

Please describe how this is verified.

RESPONSE

Methods for determining that a temperature channel is not reading correctly include (but is not limited to) comparison with other instrumentation (temperature and power), observing channel operation for spurious or erratic operation), testing with installed function, calibration, and trouble shooting.

RAI 35.3 Section 3.1 of the guidance describes that limits be placed on the shutdown margin and states that this value "should be large enough to be readily determined experimentally, for example, >0.5% Ak/k or >0.50 dollar."

Please provide an analysis and evaluation that demonstrates the ability to repeatedly measure core reactivity with sufficient accuracy to justify this small value of the shutdown margin.

RESPONSE

Reactivity changes are evaluated using a calibration of reactivity worth and position of control rods. Excess reactivity is evaluated prior to each day of reactor operation as well as following changes in experiment configuration. Since UT reactor operations on weekends are extremely rare, the first measurement of each week*

is essentially a cold, clean critical position at the current burnup. If no experiments are installed in the core, the first measurement of each week reflects the reference core condition less reactivity associated with burnup since the last reference core condition measurement (performed concurrently with control rod reactivity worth calibrations). Reactivity values of sequential first-of-the-week reactivity measurements should be comparable if burnup since the previous measurement is reasonably small, and a reasonable gage of repeatability.

All of the first-of-the-week reactivity measurements with the current core configuration (i.e., since installation of a 3-element facility) were reviewed. Measurements of excess reactivity that did not have experiments installed were tabulated along with core burnup (Table 1, Excess Reactivity Measurement Data, 3k(r¢): Excess, and A*MWD:

Total, referenced to the initial reading). The number of days between each measurement and the previous measurement was tabulated (A*T: Day's). A graph of the reactivity data (i.e., excess reactivity at first operation of the week, no experiments installed) shows the relationship between excess reactivity and burnup.

EXCESS REACTIVITY AND BURNUP 0 50 10 10 20 5 0

.10NU-(MWO Basd o brnu vaue an xcs rectvty-ifeene bewensqunia. eaigswreclclt. n tauatdinTbl (ifeecebtwe ecssraciit-6k¢: 3Se., unu ve heitevl-M D:Sq Ineral.Thi dfernc eatiiy etee eqenil eauemnt arssal at i ls tan5.0 wt forexeton.Th ifeecs nreciit eaueensocu t nevasof3 ay n 1.93W,61dy an 413MW,74das nd1.4 8 MWad-1dy n 05 W.Lgraig soitdwt h exetin er xmiebu ecrsdontprvd a biosexlntinfr h esls

Changes in sequential cold, clean excess reactivity measurements are expected to be minimal if the burnup between measurements in small, and/or if the time between the measurements is small. The 13 measurements with intervals less than 35 days and the 7 measurements for which burnup in the interval is less than 2.12 MWD have reactivity differences less than $0.05. All of the differences greater than $0.05 occurred at higher values of burnup and long times between the measurements, although reactivity differences for the remainder of the 22 intervals regardless of time interval or burnup are all small. The agreement of the other measurements across a wide range of intervals and burnup values suggests that those specific measurements may be outliers this comparison.

Excess reactivity calculations are routinely conducted to assure experiment reactivity limits are met. If the rod position data is captured before the delayed neutron population is stable, excess reactivity calculation leads to an overestimate of experiment reactivity worth. This overestimate is conservative, although the measurement is less precise. A likely explanation for the anomalous (difference in) reactivity values is that excess reactivity worth in these operations, although conservative and acceptable, was not determined with precision comparable to the other efforts.

Table 1, Excess Reactivity Measurement Data AMWD AT 6k(¢)

Sneq.a Days Excess A(Seq.)

06/29/10 0.00 0.00 na 0.00 09/27/10 2.22 2.12 90 -2.61 -2.61 11/03/10 3.71 1.59 37 -4.86 -2.26 01/07/11 5.26 1.54 65 -4.06 0.80 01/10/11 5.26 0.00 3 -4.55 -0.49 02/21/11 6.54 1.28 42 -7.13 -2.58 03/21/11 7.56 1.02 28 -10.95 -3.82 05/02/11 9.35 1.79 42 -14.56 -3.61 06/29/11 9.59 0.24 Rod Cal 0.00 na 08/29/11 13.72 4.13 61 -8.03 -8.03 10/03/11 19.89 6.18 35 -11.69 -3.67 01/12/12 30.82 10.93 101 -22.26 -10.57 03/26/12 45.68 14.85 74 -30.08 -7.83 04/02/12 61.37 15.70 7 -34.75 -4.67 04/09/12 77.83 16.46 7 -36.68 -1.93 05/14/12 97.42 19.59 35 -43.78 -7.10 05/29/12 118.16 20.73 15 -44.27 -0.49 06/11/12 140.06 21.90 13 -48.98 -4.71 06/18/12 162.50 22.43 7 -50.54 -1.56 07/09/12 186.08 23.59 21 -51.45 -0.91 07/13/12 209.69 23.61 Rod Cal 0.00 na 07/23/12 234.20 24.51 10 -4.26 -4.26 09/10/12 263.14 28.94 49 -9.23 -4.97 09/17/12 292.36 29.21 7 -11.85 -2.63

Table 1, Excess Reactivity Measurement Data AMWD AT 6k(¢)

Seq. Das Ees (e.

Date Total Interval Das Ees A(q.

09/24/12 321.82 29.47 7 -11.59 0.27 10/01/12 ....- 7 -13.93 -2.35 It is clear that for small burnup intervals and short intervals between measurements, reactivity calculations for sequential measurements are well within a few cents. Reactivity measurements using calibrated control rods at the UT TRIGA reactor are repeatable well within 5 cents.

RAI 35.4 Section 3.2 of the guidance describes that a limit be established for the maximum reactivity 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 time, are typically justified through the analysis of an uncontrolled, control rod withdrawal transient.

RESPONSE~

See response to RAI 11, 7/2015

RAI 36: ANSI/ANS-15.1-2007 Section 3, 'Limiting conditions for operations," requests that the licensee provide [COs for constraints and operational characteristics that shall be adhered to during operation. The ISG states that the applicable TlSs 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.

36.1 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) which are listed as SCRAM channels in UT SAR Table 4.6.

RESPONSE

The associated scram occurs when communications are interrupted for greater than 10 seconds as a "health" monitor, not a measuring channel. The 10 second interval is not related to safety, and could be a different interval and serve the same function. A measuring channel is the combination of sensor, line, amplifier, and output devices that are connected for the purpose of measuring the value of a parameter.

Neither the data acquisition and control system nor the control system computer are measuring channels as they do not have sensors, amplifiers, or output devices related to the scram and do not measure the value of a parameter. it should be noted that sections of the data acquisition and control system are incorporated into many of the measuring channels enumerated, and failure of communications would interrupt multiple scram channels.

RAI 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.

RESPONSE

The statement will be revised to state that operating experience has demonstrated reliable operation with the high voltage at least 80% of the nominal operating value.

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 ] 1 pulsing while in YES NA

_________________________STEADY STATE MODE ____________

RAI 36.5/6 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.

RESPONS (1) The basis statement in the currently approved Technical Specifications will replicated in the SAR and referenced in the Technical Specifications.

(2) The design section on confinement indicates criteria for the HVAC system. The HVAC control system at UT is a single switch, with no options for configuring fans, dampers or louvers. The purge system is a switch for the system with capabilities only for securing flow to the pool ventilation, beam ports, or experiments and a damper for dilution flow, configured based on operational needs. There are no specific criteria for dampers, louvers, or doors to be configured in a specific position.

(3) Completion times and labeling of the required actions will be corrected.

RAI 36.7 Proposed UT TRIGA TS 3.5, "Gaseous Effiuent Control," 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.

RESPONSE

The basis statement is "Analysis shows 100 Ci per year results in a maximum dose to individuals in the effluent plume of 0.142 mrem in a year, well within the 10CFR20 limit of 10 mrem/year for stack effleunts."

This will be revised to indicate annual limit of 100 torero.

RAI 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.

RESPONSE

Although the consequences of the maximum hypothetical accident do not exceed limits, routine operations with fuel that leaks fission products without limits is not considered acceptable by UT.

Step labeling will be corrected.

RAI 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."

RESPONSE

The request will be incorporated.

RAI 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.

RESPONSE

The mode cannot be entered prior to establishing the condition, therefore in practice there is no confusion.

RAI 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).

RAI 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.

RESPONSE

REQUIRED ACTIONS A.1.1, A.1.2, A.3.1, and A.3.3 as stated do not exist. However, this and other RAIs indicate a fundamental error in the intent and method of some action statements in the proposed Technical Specifications.

In general, the intent of the specification is for the operator to make choice, shutdown or attempt to restore the channel. Ifthe channel cannot be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the intent is to require the operator to shut down.

For example, should the particulate CAM initiate confinement ventilation and the operator notes that the unit appears to be without power then the operator is required to either shutdown the reactor or restore the channel.

If (1)the operator determines the appropriate path is to restore the channel and (2) if adequate technical support is available, the operator may direct channel restoration. If support personnel discover a blown fuse, then the channel may be restored without delay if a replacement is in stock. If personnel are unable to identify an immediately available supply, the reactor is required to be shutdown to meet the requirements of Technical Specifications within the same time constraint (without delay and not to exceed 1hour).

Therefore, the structure allows the operator to decide the appropriate solution and ensures that if the selected solution path cannot be completed then the alternate path is completed within the same time frame.

RAI 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).

RESPONSE

A statement that the reactor will not be used for normal operations until the criteria is met will be developed.

RAI 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.

RESPONSE

Labeling will be corrected. However, the intent is as written with all actions to be taken without delay, not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

RAI 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."

RESPONSE' The intent is as written with all actions to be taken without delay, not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Fuel element elongation and bend is tested out of the core, and "discharge fuel elements" is not apllicable.

RAI 37.6.1 REQUIRED ACTION(S) A.1.1 through A.1.3 are in reverse order. The COMPLETION TIME(S) are all IMMEDIATE which is contradictory.

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

RAI 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.

RAI 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.

RESPONSE

(1) The orders are intentional.

(2) All actions are required to be completed without delay, not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

This and other RAIs indicate a fundamental error in the intent and method of some action statements in the proposed Technical Specifications. In general, the intent of the specification is for the operator to make choice, shutdown or attempt to restore the channel. If the channel cannot be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the intent is to require the operator to shut down.

Actions authorized under a specific path/statement are considered by definition not considered a delay, and corrective actions not successful in attempting restoration are not considered a delay in completing a reactor shutdown. Given the confusion in review, this will be incorporated in the Action guidance in Definitions.

For example, should the particulate CAM initiate confinement ventilation and the operator notes that the unit appears to be without power then the operator is required to either shutdown the reactor or restore the channel.

If (1) the operator determines the appropriate path is to restore the channel and (2) if adequate technical support is available, the operator may direct channel restoration. If support personnel discover a blown fuse, then the channel may be restored without delay if a replacement is in stock. If personnel are unable to identify an immediately available supply, the reactor is required to be shutdown to meet the requirements of Technical Specifications within the same time constraint (without delay and not to exceed 1hour).

Therefore, the structure allows the operator to decide the appropriate solution and ensures that if the selected solution path cannot be completed then the alternate path is completed within the same time frame, without delay and not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

(3) CONDITION C will be ressated as less than 6.5 meters above the bottom of the pool (4) The basis for Specification D is stated.

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

RESPONSE

An annual functional test for the DAC and CSC watchdog scram will be added as a surveillance requirement.

RAI 38.2 There are no SRs for the reactor bay differential pressure for CONDITION A.3 in proposed UT TRIGA TS 3.3 "Measuring Channels."

RESPONSE

An annual calibration of the reactor bay differential pressure sensor will be added.

RAI138.3 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

RESPONSE

The "Upper Level Radiation Monitor" surveillance requirements will be changed to "Pool Area Radiation Monitor."

The fuel temperature measuring channel LCO is Specification 3.3.3 A, with Condition, Required Action, and Completion Time in 3.3.4 B

RAI 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."

RESPONSE

The power level scram is tested by 4.3.2 channel test. A channel test or channel check will be added for the manual scram and fuel temperature scram.

RAI 38.5 There are no SRs to suppor-t proposed UT TRIGA TS 3.7 "Fuel Integrity," CONDITION C.

RESPONSE

Fission product release will be detected by radiation monitoring systems. The particulate CAM and the argon cam are specified for operation. The Emergency Plan addresses detection of fuel element failure. The basis for 3.7/4.7 will be revised to credit detection systems that are continuous monitors.

RA[ 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.

RESPONSE

The stoichiometry reflects data from phase diagram; and is approved for other TRIGA reactors.

RAI 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 burn-up limits. These design features are not stated in the proposed UT TRIGA TS.

RESPONSE

A section will be added to Design Features, Reactor Core Applicability:

This specification applies to reactor core configurations Objective The objective is to ensure that reactor core configurations are bounded by the safety analysis.

Specification

1. The limiting core configuration will be a minimum of 90 (standard and instrumented) fuel elements with three Fuel Element Follower Control Rods (FFCR) and 1 Transient (control) Rod (TR)

2. Control rods will be

a. Adequate to maintain reactivity controls

b. Installed in locations designated by design (i.e., full penetrations in the lower grid plate accommodating fuel follower control rod movement)

c. Limited in travel by a safety plate below the grid plate

3. Positions with full penetrations in the lower grid plate may contain fuel by installing adapters that support the fuel element

4. The 120 Core positions may be occupied by

a. Standard TRIGA Fuel Elements (SFE)

b. Instrumented Fuel Elements (IFE)

c. Fuel Element Follower Control Rods (FFCR)

d. A central thimble in positon A-i

e. A neutron source

f. Sample transfer systems

g. Three element irradiation facilities

h. Seven element irradiation facilities

i. Approved experiments

j. Water voids Bases Thermal hydraulic analysis demonstrates the critical heat flux ratio remains greater than 2.0 and the maximum fuel element temperature remains below the maximum permitted for TRIGA fuel with 90 fuel elements during steady state operations.

Neutronic analysis and operating experienc:e demonstrates that reactivity limits can be met with the current complement of control elements installed.

The safety plate installed below the core grid plate prevents control elements from falling below positions where they contribute to shutdown reactivity.

Full penetrations in the lower gird plate do not support fuel elements in the core, and a physical support is required to hold the fuel element in position.

The UT TRIGA core was designed to accommodate the equipment as identified.

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

RESPONSE

A statement to the effect that fuel storage will be accomplished by mechanical devices such as racks or stands that hold fuel in position will be made in the Bases section.

RAI 39.4 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.

RESPONSE

From 10CFR50.36:

(4) Design features. Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.

(c) Technical specifications will include items in the following categories:

(1) Safety limits, limiting safety system settings, and limiting control settings. (i)(A) Safety limits for nuclear reactors 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.

(ii)(A) Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.

(2) Limiting conditionsfor operation. (i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

(3) Surveillance requirements. Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Specification 5.3.4 identifies design criteria for experiments such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety. The criteria are not safety limits, limiting safety system settings, or limiting control settings. The criteria are not experiment functional capabilities or performance criteria that meet the criterion for a limiting condition for operation, i.e., not installed instrumentation for the reactor coolant pressure boundary, not an initial condition of a design basis accident or transient that assumes a failure of or challenges the integrity of a fission product barrier, does not mitigate a design basis accident, and is not a structure system or component show to be significant risk to public health and safety. In fact, the experiment design criteria specifically prevent an experiment from encroaching on the criterion that would require an [CO. Placing the criterion into the Design section and requiring experiment reviews to assure the design criterion preserves experiment design requirements. Finally, placing the experiment design criteria in an LCO does not lead to a surveillance requirement except that the design criteria were implemented, i.e., the experiment is designed to be acceptable or. not.

RAI 4

The guidance in NUREG-2537 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-2537 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 pooi, connected pipes or beam ports.

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

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

RAI 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-15.11-1993 (R2004). This individual or group shall report to Level 1 or Level 2. The proposed UT TRIGA TS contains no such section.

RESPONSE

Administrative controls - Functional Responsibility identifies a "Radiation Safety Officer" who "acts as the delegated authority of the Radiation Safety Committee in the daily implementation of policies and practices regarding the safe use of radioisotopes and sources of radiation." And, "The Radiation Safety Committee is established through the Office of the President of The University of Texas at Austin."

.4 RAI 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.

RESPONSE

Section 6.1, Organization and Responsibilities of Personnel, part c) Staffing, states "Operation of the reactor an activities associated with the reactor, control system, instrument system, radiation monitoring system, and engineered safety system will be the function of staff personnel with the appropriate training and certification, with a footnote, "Selection and Training of Personnel for Research Reactors," ANSI/ANS-15.4 - 1970 (N380). The footnote will be changed from 1970 to 1988.