ML102650318

Texas a & M University System, Redacted - Response to NRC Non-Financial Requests for Additional Information Questions 1-37 for the Tees Nuclear Science Center Reactor

ML102650318
Person / Time
Site:	05000128
Issue date:	08/31/2010
From:	Reece W Texas A&M Univ
To:	Document Control Desk, Office of Nuclear Reactor Regulation
DiMeglio A, NRR/DPR/PRTA, 301-415-0894
References
2010-0054
Download: ML102650318 (68)

Text

TEXAS ENGINEERING EXPERIMENT STATION TEXAS A&M UNIVERSITY 3575 TAMU COLLEGE STATION, TEXAS 77843-3575 NUCLEAR SCIENCE CENTER 979/845-7551 FAX 979/862-2667 August 31, 20 10 2010-0054 Document Control Desk U,S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Response to NRC Non-Financial Requests for Additional Information Questions 1-37, from theTexas A&M University System, Texas Engineering Experiment Station, Nuclear Science Center Reactor (NSCR, License No. R-83, Docket 50-128)

To Whom It May Concern:

The Texas A&M University System, Texas Engineering Experiment Station (TEES), Nuclear Science Center (NSC, License No. R-83) operates a LEU, 1MW, TRIGA reactor under timely renewal. In December, 2003 the NSC submitted a Safety Analysis Report (SAR) as part of the license renewal process. In December, 2005 a conversion SAR (Chapter 18) was submitted resulting in an order to convert from the NRC. In July 2009, the NSC submitted an updated SAR, dated June 2009, to the Nuclear Regulatory Commission (NRC). This updated 2009 version of our SAR incorporated the information from the conversion SAR and the startup of the new LEU reactor core. On June 24, 2010 the NRC submitted a Request for Additional Information as a part of the review process. This request included thirty-seven questions related to the NSC's SAR'submittal. Attached to this letter are the NSC's responses to all thirty-seven of the NRC's non-financial RAls.

Ifyou have any questions, please contact J. A. Remlinger or me at 979-845-7551.

I declare under penalty of perjury that-the foregoing is true and correct. Executed on August 31, 2010.

W. D. Reece NSC, Director Xc:

21 I/Central File Duane Hardesy, NRC Project Manager A. Francis DiMegilo, Licensing RESEARCH AND DEVELOPMENT FOR MANKIND httpY/nsc. tamu. edu

REDACTED VERSION Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Nuclear Science Center's Responses to the U. S. Nuclear Regulatory Commission's Non-Financial Requests for Information Questions 1 through 37 Submitted August 31, 2010 REDACTED VERSION

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 4 - Reactor Description

1. According to Title 10 of the Code of Federal Regulations.(J0 CFR) Section 50.36(c)(1)(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 uncontrolled release of radioactivity. NUREG 1537, Part 2, Section 4.2.1 states that maintaining fuel integrity should be the most important design objective.

In Section 4.6.1 of the Texas A&M 2009 Safety Analysis Report (SAR) the steady-state T-H analysis of the reactor core assumes that the reactor is operating at 1.0 MW with the coolant inlet temperature at 30 degrees Celsius. There is no high temperature limit in the Technical Specifications (TS) or alarm signal available to the operator indicating that the coolant temperature is beyond the temperature assumed in the T-H analysis.

Please discuss any potential effect on fuel temperature and Critical Heat Flux assuming that the water temperature is above the 30 degrees Celsius and why there is no temperature alarm required including a TS item describing a potential limiting condition regarding high water temperature.

NSC Response: The attached report "Evaluation of Pool Temperature on Fuel Temperature and MDNBR for the NSC TRIGA Reactor" concludes that the increase of inlet pool temperature to 600 does not cause the fuel centerline on the hottest fuel element to exceed the LSSS and the DNBRwith the most conservative calculations is 1.54. The NSC further concludes there is no need for a TS or alarm for coolant temperature and thejfuel is well protected by our LSSS setting with its alarm and scram. The NSC operators are acutely aware of coolant temperature since they record pool temperature every 30 minutes.

Evaluationof Pool Temperature on Fuel Temperature and MDNBR for the NSC TRIGA Reactor Jesse Johns Nuclear Engineering Department Texas A&M University College Station, TX 77843-3133 August 4, 2010

ABSTRACT The peak fuel temperature was calculated using two different correlations, Dittus-Boelter and natural convection, far pool temperatures of 300C and 60'C. The Dittus-Boelter correlation predicted a fuel temperature rise of 1.85*C and natural convection predicted 25.64*C.

Experimental data shows that the slope of fuel temperature rise with pool temperature is better represented by the natural convection correlation than the Dittus-Boelter. When a scaling factor of 0.674 is applied to the natural convection correlation, the calculated peak fuel temperatures closely match measured data. The resulting fuel temperature at 600C is 4190C. The Bernath formulation was used for departure of nucleate boiling calculations. When the inlet conditions rise from 30TC to 60*C, the minimum departure from nucleate boiling ratio (DNBR) decreases from 3.12 to 2.24.

In comparing with General Atomics' calculations, the minimum DNBR

.becomes 1.54 at an inlet temperature of 600C.

INTRODUCTION Purpose Originally all of the thermal hydraulic calculations done by General Atomic (GA), the designers of the TRIGA Mark III reactor operated by the Nuclear Science Center (NSC), used an inlet coolant temperature of 30 0C. As this inlet temperature increases, the fuel temperature will increase for a given steady state power level. As the coolant temperature increases, this gradient driving heat transfer from the fuel decreases; therefore, the fuel temperature must increase to allow sufficient heat transfer. The concern is that this increase will.possibly exceed the Limiting Safety System Setting (LSSS) of 5240C or with the increase in coolant temperature the departure of nucleate boiling ratio (DNBR) may reach unity. The counter argument for this is that the heat transfer ability of the cooling water will be improved as its properties and velocity change with increasing coolant temperature.

Theory and Previous Work GA did the calculations necessary for license review and acceptance as required by the Nuclear Regulatory Commission for their Mark III design. Reference 1 provides details on the thermal hydraulic calculations done by GA. The Mark III TRIGA core with LEU fuel is designed as a rectangular prism. A top-down view of the NSC core is shown in. Figure 1, where "R" and 'T' indicate other control rod positions other than the primary shims. The core is in a pool that contains 106,000 gallons of water that provides the necessary cooling for the core by. natural convection. While there are pumps, they are for the circulation of the pool 'water and -transport to the heat rejection (shell and tube heat exchanger).

The bulk pool-temperature increases are ignored in the calculations due to the sheer mass of water 2

available providing for more or less a constant inlet temperature.

All of the core dimensions and parameters can be referenced in Table I of Appendix A.

Figure 1: NSCR core layout.

In the thermal-hydraulic modeling of the core, a sub-channel analysis is used so that an.analytical solution is easily produced. To do this, GA developed a proprietary code called STAT.1 Since the grid placement for bundles in the core is not symmetrical (see Figure.2) the sub-channel uses the closest fuel pitch of 0.038862 m to provide the most conservative calculation. The calculation uses the hot rod for the-entire heated portion, producing an even. more conservative result. The model also assumes that

.there is no cross flow between the sub-channels. This suggests that there will be less mixing and therefore the calculated coolant temperature should be higher than actual temperatures. The critical heat flux formulation used by STAT was developed.by Bernatlh. 2 3

GRID PLATE ADAPTER Figure 2-Nominal bundle placement.

The neutronics modeling for the core was done using DIF3D and this information was translated to the analytical model. Important results from DIF3D and STAT can be seen in Table 1in the Appendix.

METHOD Our calculations were done analytically using MATLAB, which is a numerical computational language, using empirically derived correlations with the same sUb-channel concept mentioned above. This allows for the calculation of the bulk fluid and surface temperatures through energy conservation and the.

convective heat transfer relationships.

With a resistive heat transfer network for the fuel, the temperature distribution at various interfaces can be determined. The MATLAB script divides the sub-channel into many axially positioned slices and solves each node maintaining energy, momentum, and mass conservation, see Figure 3 for an example.

4

Q(ili-QUi+1I T(i)

WO~i Figure 3: Nodal representation of sub-channel.

Coolant temperature increases are calculated by, basic energy conservation principles as given by the following equation:

AT Q

where Q is the total energy entering the coolant node,.r is the mass flow rate, and cp is the constant pressure specific heat of the coolant. For each nodal step in the sub-channel, the average temperature of the node is used to calculate the water properties. This is an iterative calculation since the nodal temperatures are initiallyunknown and the initial water properties must be used. The modeledscenario was limited to an inlet temperature of 600C since the reactor has only infrequently operated with a pool temperature greater than 50 0C.

Two different correlations were used for the fuel surface temperature calculations.

These are the Dittus-Boelter (DB) correlation and natural convection with the Grashof number correlations for a vertical cylinder.

The temperature averaged thermophysical properties of the SS304 cladding, fuel, 5

and gas gap were considered to beconstant. These values are shown in Table 2 of Appendix A. Water properties are determined using IAPWS IF-97 tables scripted into MATLAB.

The DB correlation is used for the purposes of comparison with the GA calculations and is probably the most widely known correlation for forced convection cooling, so its formulation is not be reproduced in this report. Note that the conditions. in the NSCR do not exhibit the behavior required for the-Dittus-Boelter correlation to be accurate, so a sub-cooled enhancement must be applied to the Nusselt number.4 This correction is defined as:

P p T = 0.9217 + 0.1478-- 0.1130 e-7(7-71 where P is the fuel pitch and D is the fuel diameter, such that Nu,= WNu where Nu¢ is now the correction Nusselt number.

Normally, the Dittus-Boelter correlation pertains to forced convection systems with circular geometry and the NSCR core is cooled by free convection. Additionally this correction takes into account the strong geometric nonuniformity of rod bundles.4 The. Dittus-Boelter also requires the flow to be fully developed turbulent flow, and it is very probable that this condition does not exist at the NSC. The Reynolds number calculated at outlet -conditions is slightly above 7,000. Conversely, it is commonly accepted that the onset of turbulence starts at 2,300 and fully developed flow starts around 10,000.'

The pressure drop through the core determines the mass flow rate of the bulk volume and is driven purely by.ntural convection arising from the buoyancy force as the coolant increases in temperature.

To do the momentum balance, the buoyancy forces must equate the fluid acceleration (bulk momentum flux) and the hydrostatic resistances caused by the viscous fluid creating drag on the surrounding structures., The hydrostatic resistance has three main components that affect the pressure drop (or head loss) of the fluid: the inlet nozzle, the friction with the fuel element, and the outlet nozzle. The inlet and outlet pressure drop are estimated using K-factor values of 1.640 and. 1.299, respectively.5 The natural convection heat transfer coefficient calculation uses the -correlation for vertical cylinders from Reference 3. The correlation is derived for a single vertical cylinder with a constant heat flux. The Nusselt number can therefore be calculated using the following equations:

NUD=

0.6 (RaD-*

for RaD-:

> 104 D" 016D

-U

= l. 3 7 (RaD-)

for 0.05:5 RaDT < 104

'-D0.93 a

- 0 fo 5

Da 0.05 XUD 0.93 (RaD L)-

for RaD _

.O 6

BaD gf3(T, - T) ) 3 where g is the gravitational acceleration, Tw is the surface temperature of the cylinder, T., is the bulk fluid temperature, D and L are the diameter and heated length, respectively, of the cylinder, 0 is the specific volume of the fluid, a is the thermal diffusivity of the fluid, and f? is the volumetric thermal-expansion coefficient and is defined where:

Tw - T.

7:=

2 It is standard to use the Nusselt number to calculate the average, heat convection coefficient with the following relationship:

-NUD k

where h is the averaged heat convection coefficient, k is the thermal conductivity of the fluid, and Dh is the hydraulic diameter of the sub-channel.

The surface temperature is then described by:

TS T" +

q' hirDf where q' is the linear heat rate of the fuel and D/ is the diameter of the fuel element. The temperature at the inner annular wall of the fuel is the found using a resistive heat transfer networkas defined' by:

Tfi TS +[k1 R Rf tld + 1Ir~

D 2

tctad

(

- 1 where k is the thermal conductivity of the clad or fuel, R/is the fuel element radius, t is the thickness of the clad or gap, and D is the diameter of the fuel meat annulus, where the subscripts fo and fi represent the outer diameter and inner diameter, respectively.

The critical heat flux (CHF) calculation was done using the Bernath formulation.2 This correlation is described with great detail in Reference 2 and therefore is not be repeated here. The Bernath CHF is a standard calculation to define the heat flux at which the transition from nucleate boiling to film boiling is expected to occur. When taken as a ratio with the steady state operating heat flux, as determined by DIF3D and MCNP calculations, a margin of safety can be developed. This is defined by the CHF ratio:

7.

I, qcrit CHFR =

where q" is the surface heat flux, and the subscripts crit and op define the critical and steady state operation heat flux, respectively. This calculation is done at each nodal point of the sub-channel model and only the smallest ratio is then used. To further improve on this methodology, the calculations were done by increasing the power level and.finding the new CHFR. This takes into account the increasing coolant velocity and changing fluid properties. DNB is expected to occur when the operation heat flux equates the critical heat flux. The ratio of the increased power level to the steady state condition of 1MW is then defined, outside of normal convention, asthe DNB ratio:

Power LeveIIcHFR=1 DNBR-1M

.1MW In textbooks, the CHFR and DNBR are the same. This distinction is made to provide better solutions and to model the approach of Reference I more closely.

It is important to calculate when boiling is expected to occur, because with the onset of boiling, large perturbations of power can occur. This is due to the inherent void reactivity coefficient of light water moderated systems. Because the reactor is under-moderated, the collapsing of voids as they enter cooler coolant will cause a rise in power, equal to the drop in power caused by their formation.

Consequently, the resulting power and temperature fluctuations prevent steady state conditions from being maintained and therefore boiling must be avoided. The generated heat flux must be maintained below a pre-determined value to prevent the onset of boiling.

RESULTS AND DISCUSSION The Dittus-Boelter correlation predicts that with a rise of pool temperature from 300 C to 60'C, the fuel centerline temperature of the hottest fuel rod will rise by 1.85°C, from 395.550 C to 397.400 C. This is still far below the 5240C LSSS, which is measured by an IFE that is adjacent to the hot rod. The IFE itself produces 89% of the power that the hot rod does and the thermocouple (TC) is offset by 0.3 inches from the axial centerline. One would expect the IFE TC indicated temperature to be lower than the maximum centerline temperature of the hot rod.

It was also found that the outlet coolant temperature was 100.04'C, so no boiling is expected to occur since the saturation temperature at the core depth is 1160C.

When using the Dittus-Boelter correlation, one concludes that convective cooling properties. of the heated water improved greatly. The boiling curve, shown in Figure 4, suggests that as nucleations form, there is improvement in the convective heat transfer of the cooling fluid up to the point where the

'critical heat flux is reached. The reasoning behind this phenomenon is that when the nucleation departs from the surface it creates turbulent mixing by aggravation of the coolant. However, if the DNBR is far enough away from unity, one can expect that nucleations do not form in enough excess to promote enhanced cooling.

8

1 6 1ý0I

-Dw NeDN 10 110 0

1,000:

10,000

.Figure 4: Boiling curve.6 With the natural circulation correlations, the rise in pool temperature from 30 0C to 60 0C caused a fuel peak centerline temperature rise of 25.64 °C, from 371.26 °C to 396.9 °C.

The extra increase in temperature can be explained by the increased velocity of the cooling fluid as it heats up. This has the same effect as nucleation formation in improving heat transfer; however, the increase in velocity is not substantial in this scenario.

In fact" in the isolated sub-channel model the velocity increase can be described using the continuity balance, such that:

Vo Pi Vi P0 where V is the velocity of the fluid, p is the density of the fluid, and i and o indicate the inlet and outlet conditions, respectively. So, at a pool temperature of 30°C the ratio of the velocity increase is 1.0207 and at 60 0C it is 1.0257. The velocity from inlet to outlet does increase, but only by 0.5%. The gains in heat transfer from fluid velocity changes, therefore, are not significant. This understanding is further defended in References 1, 2, and 5. Due to the pressure drop in the grid plate and bundle, this concept is not completely accurate, but it gives a good qualitative comparison. Overall, this describes the greater increase in -fueltemperature as pool temperature increases for the natural convection correlation over Dittus-Boelter.

The increase in peak fuel temperature with the natural convection correlation better matches the experimental data than the Dittus-Boelter, see Figure 5, that was acquired over various days of operation over several months.

The data was collected as part of the normal Operation's logging 9

requirements and ranges from pool temperatures of 17.78'C to 38.39 'C. The highest operating pool temperature throughout the hottest portion of the summer seldom exceeds 42.2 0C. Additionally, data collection for comparison purposes was limited to a part of core life when fuel had not been manipulated between temperature measurements. Care was taken into trying to gather temperature data when the core was operating with nearly identical reactivity conditions. That means that the control rod were banked within 5% for each measurement and the operating history and reactor experiments were similar. This provides that the flux shape is not skewed between data collection.

Burn-up over a few months insignificantly affects fuel thermal conductivity for such low power cores after the initial fuel burn-in.

Figure 5: Peak fuel temperature vs. coolant inlet temperature.

In taking the measured data into account, the slope of the natural convection data was transposed to fit the experimental data. A proportionality constant of 0.674 was applied to the Nusselt number to decrease-the average.heat transfer coefficient.

This correction factor arises because correlation used in the natural convection model was derived from experimental data using a single vertical cylinder in an infinite fluid medium. In this model, the heated cylinder is surrounded by eight cylinders. One would expect that the fluid will receive more heat energy and that the natural convection is retarded by the increase in hydrostatic drag forces.

The extrapolation of the transposed correlation to a pool temperature of 600C is 4190C, as seen in Figure 6,..which is still below the LSSS setting. This argument holds if DNBR is not at unity, and in fact, the outlet temperature is at 100.03'C, below the saturation temperature of 116 0C.

10

In both calculations,.the LSSS is not reached, so the integrity of the cladding can be maintained. While, initially, the Dittus-Boelter correlation described fuel temperature accurately, it is evident that fuel temperature raises more with coolant temperature increase than Dittus-Boelter predicts. The natural convection correlation under predicts the fuel temperature initially, but increases with a nearly identical slope. The under-prediction can be due to the experimental data used to derive the correlation.

Because the reactor has multiple vertical, heated cylinders, a proportionality constant of 0.674 can be multiplied to the Nusselt number correlation to align the calculated data onto the experimental data.

This calculation gives a higher centerline fuel temperature than the Dittus-Boelter and non-corrected natural convection correlations.

800 790 780

• • DB P

liENC

[

770 E 760 750

' 740 730 720 50 70 90 110 130 150 Coolant Temperature (0 F)

Figure 6: Peak fuel temperature vs. coolantinlet temperature.

The values calculated for the DNBR at an inlet temperature of 300C and 600C are 3.12 and 2.24, respectively. Since the DNBR approach is equivalent to the GA calculations, these results will be considered; however, the DNBR calculated by GA, using the Bernath correlation at an inlet temperature of 300C is.2.42. There is a discrepancy between the results calculated versus those done by GA and is unresolved.. The error between the two 30*C results is 29%, when this is applied to the 60'C calculation, the result is 1.59. This is still significantly above unity.

If a linear relationship is assumed for the decrease in the DNBR from the 30'C and 60'C calculations, resulting in a change of 0.88, and applied to GA's calculated DNBR, the DNBR is 1.54. Still, significantly above unity; the argument can be made that DNB conditions are avoided.

I1

In conclusion, the increase of inlet pool temperature to 600C does not cause the fuel centerline on the hottest fuelelement to exceed the LSSS. The DNBR with the most conservative calculations is 1.54.

ACKNOWLEDGEMENTS I would like to thank Dr. Cable Kurwitz and Dr. Dan Reece for the tremendous amounts of technical support and patience in aiding me in the completion of this report.

REFERENCES

1.

Safety and Accident Analyses Report Conversion from HEU toLEU Fuel. General Atomics,.2005

2. -E. E. Feldman, Fundamental Approach to TRIGA Stead-State Thermal-Hydraulic CHF Analysis.

Argonne National Laboratory, 2007.

3.

W. Janna, Engineering Heat Transfer 2 "d Edition. CRC Press, FL, 2000

4.

N. E. Todreas, M. S. Kazimi, Nuclear Systems I Thermal Hydraulic Fundamentals, Hemisphere

- Publishing Corporation, NY, 1990.

5.

TRIGA Reactor Thermal-Hydraulic Study, TRD 070.01006.04 Rev. A, General Atomics, San Diego, CA, April 2008

6. Nuclear Power Fundamentals, Department of Energy. August 4h 2010. <http://www.tpub.com/

content/doe/h1012v2/css/hIO12v2 64.htm>

12

APPENDIX A Table 1: Core parameters and GA calculated values.

Number of fuel elements Diameter, mm (in.)

Zirconium Rod Outer Diameter, mm Fuel Meat Outer Diameter, mm Clad Thickness, mm Length (heated), mm (in.)

Core flow area, mm2 (ft) 49,703 (0.535)

Core wetted perimeter, mm (ft.)

10,134 (33.25)

Flow channel hydraulic diameter, mm (ft.)

19.62 (0.0644)

Core heat transfer surface, m' (ft2) 3.86 (41.56)

Hot rod factor 1.565 Axial peaking factor 1.26 Hot spot peaking factor*

1.97 Inlet coolant temperature, °C (°F) 30 (86)

Coolant saturation temperature, °C (°F) 117.0 (242.6)

Exit coolant temperature (average), °C (°F) 64.79 (148.6)

Exit coolant temperature (maximum), `C (°F) 76.82 (170.3)

Coolant mass flow, kg/sec (lb/hr) 6.90 (54,731)

Average flow velocity, mm/sec (ft/sec) 140 (0.459)

Peak fuel temperature in average fuel 282.4 (540.41 element, "C (°F)

Maximum wall temperature in hottest 135.9 (276.6) element, T (°F)

Peak fuel temperature in hottest fuel element, 368.1 (694.6)

°C (OF)

Core average fuel temperature, 'C ('F) 206.8 (404.2)

Average heat flux, W/cm2 (BTU/hr-ft2 )

23.24 (73,690)

.Maximum heat flux in hottest element, W/cm2 50.92 (161,438)

(BTU/hr-ft 2l _

Minimum DNB ratio 2.42 13

Table 2-Thermophysical properties.

KfueI (W/m-K) 18 hgap (W/mi-K) 22210 KKd, (Wlm-K) 14.7 14

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 13 - Safety and Accident Analysis

2. NUREG 1537, Part 2, Chapter 13 states that credible accidents should be categorized and the most limiting accident in each group should be analyzed in detail including the potential consequences of the various accident scenarios, among them loss of coolant accident (LOCA) events.

The Loss of Coolant analysis in Chapter 13, Section 13.5.2 of the 2009 SAR shows that even in the unlikely scenario of instantaneous loss of all coolant flow the fuel can be safely cooled by air at the peak-fuel rodpower densities. It is also shown in 2009 SAR Section 13.5.2 Table 13-3 that the reactor pool may drain at a much slower rate between 19 - 477 minutes.

The slow draining process may result in a partially uncovered core, which may not be fully cooled by assuming a continuous circulation of air, but rather a direct convective heat transfer to the remaining reactor pool water possibly resulting in a different fuel temperature as compared to a fully air cooled fuel rod. Please discuss a partial LOCA scenario and indicate whether the fuel temperature in a partially uncovered core is still bounded by the 2009 SAR LOCA analysis.

NSC Response: Work at Argonne National Laboratory on full and partial LOCA scenarios was published in 2009 (Analyses of.Loss-of-Coolant Accidents for TRIGA-Fueled Research Reactors at Washington State University, the University of Wisconsin, and Oregon State University, F.E.

Dunn, E.E. Feldman, T. Sofu, and J.E. Matos, RERTR Program, Nuclear Engineering Division, Argonne National Laboratory 60439). This report is available at no cost, at http://www.osti.gov/bridge. It is also available on paper to the U.S. Department of Energy and its contractors, for a processing fee, from: U.S. Department of Energy, Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831-0062 (repbrts@adonis.osti.gov).

In this report, the Washington State, University of Wisconsin, and Oregon State reactors were analyzed for partial LOCA events. The important parameters as described in this report are the power production in the hottest rod and the amount of fuel meat left covered by the failure of a beam port. For the University of Wisconsin reactor, Argonne found:

"The peak rod power was calculated to be 19.3 kW at this reactor power level. Since the uncertainty in the measured reactor power level is 2%, this rod power is increased by 2% to 19.7 kW. For a decay heat level corresponding to continuous full-power operation for 7 days/week and a drain time of 15 minutes, a fuel rod power level of 24.0 kW is calculated to result in a peak fuel surface temperature of 700 0C. For a rod power of 19.7 kW, the peak fuel surface temperature would be 585°C (see Appendix A, Table A6, linear fit).

For a decay heat curve based on continuous operation at 1.0 MW for 5 days/week for 40 years and a drain time of 15 minutes, Table 2 shows that a peak fuel surface temperature of 700'C would be reached for a fuel rod power of 27.2 kW, which is much higher than the calculated peak rod power of 19.7 kW, including a 2% uncertainty. The peak fuel surface temperature was calculated to be '525°C under these conditions (see Appendix A, Table A6, linear fit)." (Argonne, op. cit., p. 10)

They further analyze for a partial LOCA.

"The centerlines of the four UWNR beam tubes are aligned with the core mid-plane, which is located aM inches above the bottom of the fuel. Since the beam tubes are 6 inches in diameter, the lowest initial water level for the partial LOCA analysis 1 above the bottom of the fuel. For a fuel rod power of 19.7 kW and a drain time of 15 minutes, the analysis in Appendix B predicted a peak fuel temperature of578°C. Since this temperature is 1220 C below a maximum fuel temperature in air of 700'C, a LOCA initiated by a failure of one of the UWNR beam ports will not result in failure of the hottest fuel rod." (Argonne, op. cit., p. 11)

The hottest rod in the NSC reactor is calculated to be 17.4 kW. With a 2% uncertainty, it would' be 17.8 kW. Clearly we are below the power used for the Wisconsin analysis and well below the threshold for fuel damage duri rial LOCAs according to the Argonne analysis. The Argonne analysis showed tha of fuel meat would be left underwater at the Wisconsin reactor during a partial LOCA. The length of fuel meat left covered by water if Beam Port I or 6 is sheared is the same as the analysis of the Wisconsin reactor. Beam Ports 1 and 6 centerlines are at the center of the fuel and the limiting part of the beam ports is a&ich inner diameter pipe. Since the fuel meat is

oong, the part under water is then:

Therefore, for our Beam Ports 1 and 6, we are bounded by the Argonne calculations.

We also have a thermal column as shown in the attached Figure 1. First, it is a most improbable occurrence that the thermal column could be compromised - large lifts are not done over the stall where the column is located. Nevertheless, if the coupler were sheared, an aluminum plate between the coupler and the thermal column extensionsill se arates the ool from the column.

If the extension were sheared, then the pool is drained the pool floor which is below the grid plate, so the full LOCA analysis applies. This full LOCA analysis has already shown that the fuel cladding is not breached for the hottest rod.

rigure i -

inermal uolumn mim Irradiator and Coupler Installation

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 13 - Safety and Accident Analysis

3. The requirements of 10 CFR Sections 20.1201.(a)(1) and 20.1301(a)(J) include limiting the total effective dose equivalent to facility staff and individual members of the public from the licensed operations. Compliance to 10 CFR Section 20.1301 is described in 10 CFR Section 20.1302.

The LOCA events analyzed in 2009 SAR Chapter 13 Section 13.5.2 assumes that the reactor core becomes uncovered with the potential increase in dose rates to the staff and members of the public due to unshieldedgamma-rays.

Please indicate the potential accumulated dose to the facility staff members considering the facility evacuation plan and whether is in agreement with 10 CFR Section 20.1201 requirements.

In addition, please discuss the maximum accumulated dose to a member of the public due to unshielded gamma-rays and sky-shine, and whether it satisfies 10 CFR Section 20.1301 requirements.

NSC Response:

As stated in 10CFR50, in part:

An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(ii) An individual located at any-point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its-passage), Would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

The design basis accident (DBA) performed in section 13 assumes the loss of integrity of the fuel cladding for one fuel element and the simultaneous loss of pool water resulting in fission product release. The calculations were based on the fission product release.

The accumulated dose to staff following the onset of accident is calculated to be 0.25 mrem for the 5 minute duration after release. This is in agreement with the 1 OCFR20.1201 requirements (TEDE of 5 rem)

The maximum exposure to personnel in unrestricted area calculations includes the dose to unshielded gamma rays and ground shine from the plume passage. The whole body dose (TEDE) calculated for the member of the public is 6.2 mrem and it satisfies the 10CFR20.1301 requirements (TEDE of 100 mrem).

MCNP calculations were performed modeling the containment building to calculate the dose at the fence line to members of thegeneral public for dose arising from a completely uncovered (but intact) core. In all cases studied, the maximum dose to a person at the fence line is below1.0 mrem/hr.

The dose in the control room, arising principally from scattered radiation, from the uncovered core is calculated to be 4.27 rem/hr. With a five-minute exit time this would lead to a dose of 3"56 mrem, well below the 1 OCFR20 requirements. In actually, the quickest the pool could drain is on the order of 15 minutes and even 3 feet of water cuts the dose by an order of magnitude or more.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 4 - Reactor Description and Chapter 14 - Technical Specifications

4. NUREG 1537, Part 1, Section 4.3, Reactor Tank or Pool states that the applicant should present all information about the pool necessary to ensure its integrity, should assess the possibility of uncontrolled leakage of contaminated primary coolant, and should discuss preventive and protective features.

Section 4.3 Chapter 4 of the 2009 SAR provides limited information in this area:

a. Please discuss the reactor pool water level monitoring system, alarm levels and required responses from the facility operators and/or university personnel, if a remote alarm signal is present.

b. Please discuss the basis for not requiring a TSfor a pool level alarm.

c. Please discuss potential draining pathways of reactor pool water leakage, facility operator responses to a water leakage, radioactivity monitoring as a result of the leakage, and the normal radioactive material content of the pool before release.

d. Please explain the significance of the 17-minute time for manual cover installation and provide the,basis for the 17 minutes, including assumptions and boundary conditions, such as operating makeup system.

e. Please discuss the scenario when there is an excessive leakage in the heat exchanger piping, how this type of leakage would be identified (low level alarm, temperature alarm, etc.), and how the licensee would respond should the facility be unoccupied.

NSC Response: The NSC considers some details of the pool level alarm system to be Safeguards Information, but the following information is provided.

a. A robust system is in place that indentifies loss of pool level. An alarm is received both locally and at a 24-hour manned facility. Procedures are in place for a rapid response from both University Police Department and NSCR personnel when an alarm level is reached. A "call out" list for NSCR responders is provided to appropriate response entities.

b. Please. note that TS 14.5.7 has the following specification: "4. A pool level alarm shall indicate loss of primary coolant before or equal to the pool level dropping to 10% below the normal operating level."

c. All leakage will end up in a waste water transfer sump. This water ends up in our waste storage tanks. Water in these tanks are analyzed for radioactive contamination prior to their release to a waste water processing facility. The NSC has detailed operating procedures for response to a pool water leak. Based on years of waste water tank releases the pool radioactive material content is normally less than 10% of the limit for release to sewerage as listed in 10 CFR Part.20, Appendix B.

d. The 17 minute time is simply a conservative point of reference. As the TS states, it is the.

e. Again, a low level alarm identifies a loss of pool water. During operation change in pool level is easily noted since we have a large open pool and our control room is on the same level and adjacent to the pool. As.indicated in "a" above the alarm is 'sent to a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> dispatch who notifies appropriate responders.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

5. NUREG-153 7 states that the format and content of the TSfollow that ofAmerican National Standards Institute (ANSI), American Nuclear Society (ANS) 15.1. ANSI/ANS 15.1-2007, Section 1.3, Definitions, describes the definitions for uniform interpretation of terms and phrases used for research and test reactors. Texas A&M University (TAMU) TS 14.1 does not define the following terms: core configuration, excess reactivity, license, licensee, protective action, reactor operator, and senior reactor operator. Please define these terms or provide a basis for not including these definitions.

NSC Response: The Nuclear Science Center will modify the TS 14.1 to include the terms above using the American National Standards Institute (ANSI), American Nuclear Society (ANS) 15.1.

ANSI/ANS 15.1-2007, Section 1.3.

14.1.7 Core Configuration The core configuration includes the number, type, or arrangement of fuel elements, reflector elements, and regulating/control/transient rods occupying the core grid.

14.1.9 Excess Reactivity Excess reactivity is that amount of reactivity that would exist if all reactivity control devices were moved to the maximum reactive condition from the point where the reactor is exactly critical (keff = 1) at reference core conditions or at a specified set of conditions.

14.1.16 License The written authorization, by the responsible authority, for an individual or organization to carry out the duties and responsibilities associated with a personnel position, material, or facility requiring licensing.

14.1.17 Licensee An individual or organization holding a license.

14.1.28 Protective Action Protective action is the initiation of a signal or the operation of equipment within the reactor safety system in-response to a parameter or condition of the reactor facility having reached a specified limit.

14.1.34 Reactor Operator An individual who is licensed to manipulate the controls of a reactor.

14.1.49 Senior Reactor Operator An individual who is licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.

r

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical 'Specifications

6. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS 15.1-2007, Section 1.3, Definitions, describes the definitions for uniform interpretation of terms and phrases used for research and test reactors. The 2009 SAR TSs use the following terms that are not defined: protective devices and operator. Please define these terms or use terms that have been defined in TS 14.11; or provide a basis for not including these definitions.

NSC Response: The NSC proposes eliminating the term protective devices in 14.1.19, redefining the Limiting Safety System Setting while adding definitions for protectiveactions and operator.

14.1.19 Limiting Safety System Setting The limiting safety system setting is the setting for initiating automatic protective action.

14.1.27 Operator See Reactor Operator.

14.1.28 Protective Action Protective action is the initiation of a signal or the operation of equipment within the reactor safety system in response to a parameter or condition of the reactor facility having reached a specified limit.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

7. NUREG-153 7 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-J5.1-2007, Section 1.3 provides a definition of "Reference Core Condition. " TAMU uses the term "Cold Critical" in TS 14.3.1.5 and 14.3.6.1; however the term is not defined Please evaluate the term "Cold Critical" against the ANSI/ANS-15.1 standard definition for "Reference Core Condition" and consider developing a definition that can be included in TAMU TS 14.1 and used in the Limiting Condition for Operations (LCO) for Excess Reactivity and Shutdown Margin in TS 14.3.1.3 and TS 14.3.1.5 respectively.

NSC Response: The NSC will eliminate the use of the term "Cold Critical," and replace it with the term "Reference Core Condition," and add the ANSI standard definition for "Reference Core Condition" to the definitions section.

14.1.38 Reference Core Condition The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible (<0.30 dollar).

14.3.1.5 Maximum Excess Reactivity Applicability This specification applies to the maximum excess reactivity, above reference core condition, which may be loaded into the reactor core at any time.

14.3.6.1 Reactivity Limits Basis

2. The maximum worth of a single experiment is limited so that its removal from the reference core condition reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

8. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 4provides allowable surveillance intervals for varying time periods and specifically states that "weekly" shall be an "interval not to exceed 10 days."

TAMU TS 14.4 does not define allowable surveillance intervals, however defines these time periods in Section 14.1. TAMU TS 14.1.46 defines "Weekly" as "should be performed at least once every calendar week "Please include a 10 day limit within the definition of "weekly" or provide a basis for maintaining the current definition.

NSC Response: The NSC will modify our definition for weekly to follow ANSI/ANS 15.1.

ANSI/ANS-15.1-2007. Additionally, the definition for Biennially was changed to match the standards.

14.1:55Weekly Items which are designated by these specifications or by standard operating procedures to be performed weekly shall be performed at intervals not to exceed 10 days.

14.1.4 Biennially Items which are designated by these specifications or by standard operating procedures to be performed biennially should be performed at least once every 24 months, but at no time shall the time period between consecutive performances of these items exceed 30 months.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

9. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 6. 7.2 discusses special reporting requirements for operational occurrences. TAMU TS 14.1.31(3) lists a reportable occurrence as "A reactor safety system component malfunction that renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdowns. " ANSI/ANS-15. 1, Section 6.7.2 does not include the stipulation for conditions discovered during maintenance. Please consider removing the stipulation or provide a basis for including this stipulation.

NSC Response: The Nuclear Science Center will modify the TS 14.1.31(3) [now TS 14.1.39(3)]

to remove thestipulation and conform to the American National Standards Institute (ANSI),

American Nuclear Society (ANS) 15.1. ANSI/ANS-15.1-2007, Section 6.7.2.

14.1.39 Reportable Occurrence A reactor safety system component malfunction that renders or could render the reactor safety system incapable of performing its intended safety function. If the malfunction or,.condition is caused by maintenance, then no report is required. (Note:

Where components or systems are provided in addition to those required by the technical specifications, the failure of the extra components or systems is not considered reportable provided that the minimum numbers of components or systems specified or required perform their intended reactor safety function.)

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. '50-128 Responses to Request for Additional Information Chapter 14 - Technical'Specifications

10. TAMU TS 14.3.1.1 establishes that, for the purposes of testing and calibration, the reactor may be operated at power levels greater than the licensed power level. Specifically, operation is authorized up to 1.3 M4W, while licensedpower level is 1.0 MW. Please develop a method for meeting testing/calibration requirements without exceeding the licensed power level and eliminate this exemption from the licensed power level in TS 14.3.1.1 or provide a basis for maintaining the TS in its current state.

NSC Response: The NSCR has operated it's reactor for over 30 years with this steady state LCO. The current basis coupled with 30 plus years of experience is sufficient justification for this LCO. And, after a review of a number of other Non-power reactor licensees we find similar allowance for testing power. However, the NSC proposes to rewrite the TS to limit operation for testing at 1.3 MW to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less.

14.3.1.1 Steady State Operation Applicability This specification applies to the energy -generated in the reactor during steady state operation.

Obiective The objective is to ensure that the fuel temperature safety limit will not be exceeded during steady state operation.Specifications The reactor power level shall not exceed 1.3 megawatts (MW) under any condition of operation.

The normal steady state operating power level of the reactor shall be 1.0 MW. However, for purposes of testing and calibration, the reactor may be operated at higher power levels not to.

exceed 1.3 MW during the testing period. The duration of this testing period shall not exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Basis Thermal and hydraulic calculations indicate the TRIGA fuel may be safely operated up to power levels of at least 2.0 MW with natural convection cooling

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

11. NUREG-153 7 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 3.2(1) includes guidance on establishing LCOs for minimum number of operable control rods (defined in terms of scram time). Please discuss whether TAMU TS 14.3.2 is consistent with the standard guidance.

NSC Response: TAMU TS 14.3.2 does not define an LCO for minimum number of operable control rods. We will modify the TS 14.3.2.3 Scram Time to include this LCO.

14.3.2.3 Minimum Number of Operable Scrammable Control Rods Applicability This specification applies to the minimum number of operable scrammable control rods in the core, where operable is specified in terms of maximum scram time from the instant that the fuel temperature safety channel variable reaches the Limiting Safety System Setting.

Objective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.

Specification During standard operation all scrammable control rods shall be operable. The scram time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable rod reaches its fully inserted position shall not exceed 1.2 seconds. During core manipulations, i.e. core loading and unloading, all installed scrammable control rods shall be operable.

Basis This specification ensures that the reactor will be promptly shutdown when a scram signal is initiated. Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

12. TAMU TS 14.3.2, Table 14-2, establishes the function for the Fuel Element Temperature and High Power Level safety channels as "Scram at LSSS" and "Scram at 125%" respectively.

Please consider changing to statements where the safety channels are required to scram "prior to exceeding" rather than "at" the designated limit.

• NSC Response: The NSC proposes to change the technical specification wording to show that the Fuel Element. Temperature and High Power Level safety channels will "Scram < the LSSS" and "Scram < 125%".

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

13. NUREG-153 7 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 3.2(8) includes guidance on establishing permitted bypassing of channels for the purposes of calibrations and maintenance. Please discuss whether TAMU TS 14.3.2 should include acceptable conditions for bypassing channels for this purpose.

NSC Response: An additional point will be added to TS 14.3.2. That will read:

"During periods of maintenance, surveillance, calibration and repair, the reactor may be operated as necessary to calibrate and assure channel operability. For example, after maintenance, the interlock that prevents accidental pulsing above 1 kW may be out ofspecification. The reactor can be operated to calibrate this interlock and to demonstrate that it now meets technical specifications.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical[Specifications

14. NUREG-153 7 states that the format and content of the TSfollow that ofANS1/ANS 15.1.

ANSI/ANS-15.1-2007, Section 3.3 provides guidance for establishing LCOs on coolant systems to include coolant level limits, leak or loss of coolant detection and fission product activity detection. Please discuss whether TAMU TS 14.3 is consistent with the standard guidance.

NSC Response: ANSI/ANS 15.1-2007 states that the Technical Specification shall specify the minimum operating equipment, or operating limits, or both, for the following:

(1) shutdown cooling or pump requirements; (2) isolation valves; (3) coolant level limits; (4) leak or loss-of-coolant detection; (5) fission product activity detection; (6) hydrogen concentration (off-gas) limits; (7) emergency core cooling systems; (8) secondary coolant activity limits; (9) water chemistry requirements (such as conductivity or pH limits averaged over a period of time).

Responses item by item follow:

(1) We are natural convection cooled reactor. We have no pump requirements. There is no need for a limiting condition of operation requirement.

(2) We have covers 'for the primary intake and outflow. This is covered in TAMU TS 14.5.7.

(3) The reactor shall not be operated if the pool level is below the intake for the diffuser system.

(4) No LCO is required. A daily-checked easily visible gauge on the pool wall is satisfactory visual indication of water loss and is used to determine makeup. Daily readings of pressure of the primary and secondary cooling systems ensure non-leakage through the heat exchanger.

(5) This is already in our Technical Specifications. TAMU TS 14.5.4 and new TAMU TS 14.4.4.1 (6) The NSC operates a large surface area non-enclosed pool in a facility with at least 180,000 cubic feet of free space, 'so the concentration of hydrogen is negligible. This standard is not applicable.

(7) Our design basis accident does not include any emergency cooling system. We have no limiting condition of operation requirement.

.(8) By Tech Spec limits, the reactor pool quality, in terms of both chemistry and radioactive contaminants, will always be such that it could be discharged directly to the sanitary

sewer. We do not operate the reactor if the primary system (pool) contains appreciable anions that will activate so to control the spread of contamination. Given this, the chances of contaminating the secondary system are negligible. We do not need a limiting condition of operation requirement.

(9) This is already in our Technical Specifications. TAMU TS 14.3.8

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

15. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Sections 3.4.1 and 3.4.2 define operations that require confinement and equipment needed to achieve confinement. TAMU TSs 14.3.3.1 and 14.3.3.2 include notes that establish exceptions for the operability of the central exhaust fan. If the intent of these notes is for confinement to be maintained as operable while the main exhaust fan is taken out of service, please provide a basis for the exception and a limit on the period of time that the fan can remain inoperable.

NSC Response: The technical specifications 14.3.3.1 and 14.3.3.2 include an extended note discussing a one hour limit and a basis to reflect the change in specifications 14.3.3.1 (1) and 14.3.3.2 (1).

14.3.3.1 Operations that Require Confinement Applicability This specification applies to confinement requirements during operation of the reactor and the handling of radioactive materials.

Objective To maintain normal or emergency airflow into and out of the reactor building during operations that produce or could potentially produce airborne radioactivity.

Specification Confinement of the reactor building will be required during the following operations:

1. Reactor operating Note: For periods of maintenance to the central exhaust fan, entry doors to the reactor building will remain closed except for momentary opening for personnel entry or exit. The central exhaust fan shall not remain off during periods of maintenance for more than one hour.

2. Handling of radioactive materials with the potential for airborne release.

Basis

1. This basis applies during the conduct of those activities defined as reactor operations.

41Ar is produced during operation of the reactor in experimental facilities and in the reactor pool; thus, air control within the building and the exhaust system in necessary to maintain proper airborne radiation levels in the reactor building and release levels in the exhaust stack. Other radioactivity releases to the reactor building must be considered during reactor operation, such as fission product release from a leaking fuel element or a release from fixed experiments in or near the core. Shutdown of the central exhaust fan for periods of less than one hour is for operability verification during weekly ventilation checks. This limit provides enough time to complete these checks.

2. The handling of radioactive materials can result in the accidental or controlled release of airborne radioactivity to the reactor building environment or direct release to the building exhaust system. In these cases, the control of air into and out of the reactor building is necessary.

14.3.3.2 Equipment to Achieve Confinement Applicability This specification applies to-the equipment and controls needed to provide confinement of the reactor building.

Obiective The objective is to ensure that a minimum of equipment is in operation to achieve confinement as specified in Section 14.3.3.1 and that the control panel for this equipment is available for normal and emergency situations.

Specifications

1. The minimum equipment required to be in operation to achieve confinement of the reactor building shall be the 'central exhaust fan.

Note: During periods of maintenance to the central exhaust fan, entry doors to the reactor building will remain closed except for momentary opening for personnel entry or exit. The central exhaust fan shall not remain off during periods of maintenance for more than one hour.

2. Controls for establishing the operation of the ventilation system during normal and einergency conditions shall be available in the reception room.

Basis

1. Operation of the central exhaust fan will achieve confinement of the reactor building during normal and emergency conditions when the controls for air input are set such that the central exhaust fan capacity remains greater than the amount of air being delivered to the reactor building. The exhaust fan has sufficient capacity to handle extra air intake to the building during momentary opening of doors. Shutdown of the

central exhaust fan for periods of less than one hour is for operability verification during weekly ventilation checks. This limit provides enough time to complete these checks.

2. The control panel for the ventilation system provides for manual selection of air input to the reactor building and the automatic or manual selection of air removal. The air supply and exhaust systems work together to maintain a small negative pressure in the reactor building. These controls are available in the reception room for accessibility during emergency conditions.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

16. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 3.5 provides guidance to ensure the minimum number of ventilation fans for normal operation is defined Please provide an evaluation of TAMU TS 3.5 in relation to this standard to include an explanation of the basis for the exemption allowing reactor operation with the ventilation system inoperable for maintenance. Further, please consider establishing a statement that includes a limit for the period of time that the ventilation system can remain inoperable or provide a basis for the current format.

NSC Response: The ventilation system includes the central exhaust and air handlers. While the air handlers are not required for operation, the louvers that isolate the intakes are required to close when the air handler is shut down. This is checked weekly during the ventilation system checks. The technical specification will now explicitly state this difference and the conditions that follow.

14.3.5 Ventilation System Applicability This specification applies to the operation of the facility ventilation system.

Objective The objective is to ensure that the ventilation system can mitigate the consequences of the possible release of radioactive materials resulting from reactor operation and radioactive material handling.

Specification The reactor shall not be operated unless the following parts of the facility ventilation system are operable:

1) Central Exhaust Fan

2) Intake louvers on the air handlers The reactor may be operated during periods of maintenance up to one hour. During this maintenance period the reactor building doors.will remain closed except for momentary opening for personnel entry or exit.

In the event of a substantial release of airborne radioactivity, the ventilation system will be secured automatically by signals from the appropriate facility air monitor.

Basis During normal operation of the ventilation system, the concentration of 41Ar in unrestricted areas is below the effluent concentration (Chapter 11). In the event of a substantial release of airborne radioactivity, the ventilation system will be secured automatically. Therefore, operation of the reactor with the ventilation system shutdown for short periods of time to make repairs ensures the same degree of control of release of radioactive materials. Moreover, facility air monitors within the building independent of those in the ventilation system will give warning of high levels of radiation that might occur during operation with the ventilation system secured. The ventilation system must be shutdown for weekly operability check. The one hour limit provides enough time to complete these checks.,

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

17. NUREG-153 7 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 3. 7.1 provides a time limit for alternate methods of radiation monitoring with a channel out of service. Please review the notes associated with TAMU TSs 14.3.5.1 and 14.3.5.3, and consider adding time limits consistent with the standard guidance or provide a basis for not including time limits.

NSC Response: The NSC proposes to make changes to applicable technical specification(s) to include the following:

Note: When required monitors are inoperable, or for periods of maintenance, the intent of this specification will be satisfied if this monitor is replaced for a period of no more than 1 week with a portable gamma sensitive instrument having its own alarm which shall be kept under visual observation. If two of the above monitors are not operating, the reactor will be shutdown.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

18. NUREG 153 7, Part 1, Section 10.1 Experimental Facilities and Utilization, states that the applicant should provide sufficient information to demonstrate that no proposed operations involving experimental irradiations or beam utilization will expose reactor operations personnel, experimenters, or the general public to unacceptable radiological consequences. Regulatory Guide 2.2, Section C. 1. c. (3) states that the "materials of construction and fabrication and assembly techniques should be so specified and used that assurance is provided that no stress failure can occur at stresses twice those anticipated in the manipulation and conduct of the experiment or twice those which could occur as a result of unintended but credible changes of or within, the experiment."

TAMU SAR Chapter 14, Technical Specifications, Section. 3.6.2(Ia) Material Limitations, allows that explosive materials in quantities less than 25 mg may be irradiated in the reactor in a container "provided that the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container. "Please discuss how TAMU will ensure a safetyfactor of two in TAMU TS 14.3.6.2.

NSC Response: The specification 14.3.6.2 (1 a) now includes the wording that "provided that the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than half the design pressure of the container."

The basis 14.3.6.2 (Ia) now includes a comment to incorporate, in the analysis of the experiment, a safety factor of 2.0 or greater.

Basis

1. This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving explosive materials.

a. This specification is intended to prevent damage to the reactor core and safety related reactor components located within the reactor pool in the event of failure of an experiment involving the irradiation of explosive materials. Limited quantities of less than 25 milligrams and proper containment of such experiment provide the required safety for in-pool irradiation. To ensure this, a safety factor of 2.0 or greater on the containing structure will be included in the analysis.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

19. NUREG-153 7 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 3.8.2 provides guidance for double encapsulation of experiments involving corrosive materials. Please discuss whether TAMU TS 14.3.6. 2 is consistent with the standard guidance.

NSC Response: We will modify TAMU TS 14.3.6.2 to address corrosive materials.

14.3.6.2 Material Limitations Applicability - Unchanged Objective - Unchanged Specifications

3. Significant amounts of corrosive materials used in a reactor experiment shall be double encapsulated. Exceptions may only be made if a detailed analysis and/or prototype testing with small amounts of materials demonstrates that the experiment presents negligible risk.

Basis 3.This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving corrosive materials.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request, for Additional Information Chapter 14 - Technical Specifications

20. The requirements of 10 CFR Section 50.36(c)2(i)B, Criterion 2 state that written technical specification limiting condition of operation must be established for process variable 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. ANSI/ANS 15.1-2007, Section 3.8.1(2) requires establishing a limiting condition of operation for the sum of the absolute values of the reactivity worths of all experiments.

TAMU SAR Chapter 14, Technical Specifications, Section 3.6.1 Reactivity Limits establishes a limit for secured and non-secured single experiments. There is no reactivity limit specification for the sum of the absolute value of all experiments in the reactor as required by 10 CFR Section 50.36(c)2(ii)B with guidance from ANSI/ANS 15.1.3.8.1(2).

Please discuss the basis whether there is a need for a Limiting Condition for Operation (LCO) regarding the sum of absolute value of all experiments in the reactor ensuring that the total maximum reactivity worth limit is not exceeded.

NSC Response: We will modify TAMU TS 14.3.6.1 to address the sum of the absolute value of all experiments in the reactor.

14.3.6.1 Reactivity Limits Applicability This specification applies to the reactivity limits on experiments installed in the reactor and its experimental facilities.

Objective The objective is to ensure control of the reactor during the handling of experiments adjacent to or in the reactor core.

Specifications The reactor shall not be operated unless the following conditions. governing experiments exist.

1. The reactivity worth of any single, non-secured experiments shall be less than one dollar.

2. The reactivity worth of any single experiment shall be less than two dollars.

3. The sum of the absolute reactivity of all experiments, secured and non-secured, shall be less than five dollars.

Basis

1. This specification is intended to ensure that the worth of a single unfastened experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were suddenly inserted. This does not restrict the number of non-secured experiments adjacent to or in the reactor core.

2. The maximum worth of a single experiment is limited so that its removal from the reference core condition reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.

3. This limit poses a restriction on the number of experiments being run at any given time to prevent excessive positive. and negative reactivity loading. A five dollar limit among research reactors is common practice.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

21. TAMU TS Section 14.3.6.4, Xenon Irradiation for Iodine Production, limits the total facility Xenon-125 of all experiments to 3300 curies (Ci). The corresponding bases state that the production ofXenon-125 in excess of this limit is not necessary. Please discuss the basis and methodology for this limit.

NSC Response: The NSC will change Basis 2 to read:

Basis

2. In the NRC's -safety evaluation supporting amendment 15 to our facility license the commission concluded that the NSC's request for a total inventory limit of Xe-125 of 3500 Ci was reasonable based on the facility experimental design and operating plan for Iodine Production.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

22. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 3.3 provides guidance for establishing LCOs on coolant systems to include water chemistry requirements. TAMU TS 14.3.8 establishes pH limits for bulk pool water with a two week exception for deviations. Please provide the basis for two weeks exception for primary coolant chemistry.

Additionally, while TAMU TS 14.3.8 establishes an operating limit on the conductivity andpH of the primary coolant, there is no corresponding surveillance in TAMU TS 14.4. The pH and radioactivity level of the reactor pool water should be monitored on a periodic basis. TAMU TS

.14.3.8 implicitly requires testing not less than two weeks. If this is considered a surveillance program, please discuss why a specific TS item in TAMU TS 14.4 is not required NSC Response: ANSI/ANS-15.1-2007, Section 3.3 states that conductivity or pH can be used to

.monitor water chemistry. Given the challenges of measuring pH in extremely low conductivity water, we will remove references to pH measurement from TAMU TS 14.3.8. We will also modify the TS to include a maximum conductivity beyond which the reactor may not be operated. TS 14.4.4.2 Coolant Conductivity will be added to address the surveillance program associated with this LCO.

14.3.8 Primary Coolant Conditions Applicability This specification applies to the quality of the primary coolant in contact with the fuel cladding.

Obiective The objectives are to minimize the possibility for corrosion of the cladding on the fuel elements and to minimize neutron activation of dissolved materials.

Specifications

1. Conductivity of the bulk pool water shall be no higher than 5 x 10-6 mhos/cm (5 piSiemens/cm) for a period not to exceed two weeks.

2. The reactor shall not be operated if conductivity exceeds 50[tmho/cm.

Basis A small rate of corrosion continuously occurs in a water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is

required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.

By limiting the concentrations of dissolved materials in the water, the radioactivity of neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel exposure during maintenance and operations. A maximum conductivity level of less than 10 times the limit of 5 Vmho/cm ensures an excessive inventory of radionuclides is not produced.

A two week out of specification exception for primary coolant conductivity allows time to determine and correct the cause for the out of specification condition.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

23. NUREG-153 7 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 4 provides guidance for identifying which surveillances can be deferred during shutdown and which must be performed prior to reactor operations. Please discuss whether TAMU TS 14.4 is consistent with the standard guidance.

NSC Response: The following will be added to TAMU TS 14.4 to address surveillance deferment during shutdown.

14.4.8 Surveillance Deferment Applicability This specification describes which surveillance requirements may be deferred during extended reactor shutdown, and which shall be performed prior to resuming reactor operations.

Obiective The objective is to give facility personnel the ability to avoid conducting unnecessary surveillances during periods of reactor shutdown.

Specifications Possible to Defer Required prior to TS During Shutdowns?

operations?

1.

14.4.2.1 Steady State Operation Yes Yes

2.

14.4.2.2 Pulse Mode Operation Yes Yes

3.

14.4.2.3 Shutdown Margin Yes Yes

4.

14.4.2.4 Excess Reactivity Yes Yes

5.

14.4.2.5 Reactor Fuel Elements No N/A

6.

14.4.3.1 Reactor Control Systems No N/A

7.

Yes, except pool 14.4.3.2 Reactor Safety Systems level Yes

8.

14.4.3.3 Scram Time Yes Yes

9.

14.4.4.1 Coolant Radioactivity Analysis Yes - Annual Yes

10.

14.4.4.2 Coolant Conductivity Yes - Annual Yes

11.

14.4.5 Equipment to Achieve Confinement: Ventilation System Yes Yes

12.

14.4.6 Radiation Monitoring

'Systems and Effluents No N/A

13.

14.4.7 Experiments Yes Yes Basis

1. It is impossible to perform a calorimetric while the reactor is shutdown; a calorimetric shall be performed prior to resuming operations.

2. It is impossible to perform pulse surveillance while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

3. It is impossible to determine control rod worth while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

4. It is impossible to determine control rod worth-while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

5. The possibility for cladding degradation or damage still exists while the reactor is shutdown, therefore this surveillance shall not be deferred.

6. The possibility for control rod degradation or damage still exists while the'reactor is shutdown, therefore this surveillance shall not be deferred.

7. With the exception of the pool level alarm, whose surveillance shall not be deferred, it is not necessary to conduct surveillance on reactor safety systems, which are effectively automatic scram systems, while the reactor is shutdown. The surveillance shall be performed prior to resuming operations.

8. It is impossible to determine scram time while the reactor is shutdown; the surveillance shall be performed prior to resuming operations.

9. The possibility for cladding degradation or damage still exists while the reactor is shutdown; however it is much less likely. Therefore the periodicity of this surveillance may be changed to annual. The surveillance shall be performed prior to resuming operations.

10. Fuel cladding degradation is still undesirable while the reactor is shutdown; however chemically contaminating the pool is much less likely. Therefore the periodicity of this surveillance may be changed to annual. The surveillance shall be performed prior to resuming operations.

11. The ventilation system is only required during reactor operations or radioactive materials handling. The surveillance shall be performed prior to resuming operations or handling.

12. The possibility for cladding degradation or damage still exists while the reactor is shutdown, therefore this surveillance shall not be deferred.

13. No experiments can be installed in the reactor while it is shutdown; the surveillance shall -be performed prior to resuming operations.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

24. NUREG-1537 states that the format and content of the TSfollow that of ANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 4.1(1) provides guidance for surveillance requirements on Excess Reactivity. Please discuss whether TAMU TS 14.4.2 is consistent with the standard guidance.

NSC Response: The excess reactivity is maximum at initial core loading and decreases monotonically for 30/20 LEU fuel. (See SAR.) However we agree that it would be helpful to document excess reactivity at least biennially and after any significant core or control rod change. We will add the following Technical Specification: (See 14.4.2.4) 14.4.2.4 Excess Reactivity Applicability This specification applies to the surveillance requirements of reactor excess reactivity.

Obiective The objective is to verify that requirements on excess reactivity are met for operational cores.

Specification The excess reactivity will be determined biennially and following significant core configuration and/or control rod changes.

Basis The excess reactivity of the core is measured to ensure that during all -states of operation criticality can be maintained for licensed operational limits. With the accumulation of fission product poison buildup and fissile material burnup, excess reactivity must be available for power transients and maintaining criticality.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

25. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANSI15. 1.

ANSI/ANS-15.1-2007, Section 4.1.(2) provides guidance for surveillance requirements on Shutdown Margin and includes the stipulation "and following significant core configuration and/or control rod changes. "Please discuss whether TAMU TS 14.4.2.3 is consistent with the standard,guidance.

NSC Response: The TS will be changed as follows:

14.4.2.3 Shutdown Margin Applicability This specification applies to the surveillance requirement of control rod calibrations and shutdown margin.

Objective The objective is to verify that the requirements for shutdown margins are met for operational cores.

Specification The reactivity worth of each control rod and the shutdown margin shall be determined annually and following significant core configuration and/or control rod changes.

Basis The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worth of experiments inserted in the core. Experience with TRIGA reactors gives assurance that measurement of the reactivity worth on an annual basis is adequate to insure no significant changes in the shutdown margin.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

26. NUREG-1537 states that the format and content of the.TS follow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 4.2(1) provides gu.idance for surveillance requirements on reactivity worth of control rods and includes the stipulation "and following significant core configuration and/or control rod changes. "Please discuss whether TAMU TS 14.4.3.1 is consistent with the standard guidance.

NSC Response: The TS will be changed as follows:

14.4.3.1 Reactor Control Systems Applicability These specifications apply to the surveillance requirements for reactor -control systems.

Objective The objective is to verify the condition and operability of system components affecting safe and proper control of the reactor.

Specifications

1. The control rods shall be visually inspected for deterioration biennially.

2. Operability tests of the control rod mechanism will follow modification or repairs.

3. The Transient Rod drive cylinder and associated air supply system shall be inspected, cleaned and lubricated semiannually.

Basis

1. The visual inspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation of the reactor.

2. These tests provide verification that the control rod has full travel and that the rod drop time is within specification.

3. Inspection and maintenance of the Transient Rod drive assembly reduces the probability of failureof the system due to moisture-induced corrosion of the pulse cylinder and piston rod assembly.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

27. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 4.2(9) provides guidance for surveillance requirements on reactor control interlocks. Please discuss whether TAMU TS 14.4.3 is consistent with the standard guidance.

NSC Response: The NSC will add a point to the specifications of 14.4.3.2 Reactor Safety Systems to specifically address reactor control interlocks.

5. Reactor safety systems which include interlocks shall be tested at least annually.

The basis will also be changed to read:

Channel tests will assure that the safety system channels are operable on a daily basis or prior to an extended run. If the period between operations extends beyond a year, then the annual channel test requirement will assure operability.

Additionally, Specification No. 4 will be rewritten to add the word channel for clarity.

4. A channel check of the fuel element temperature measuring channel shall be made daily whenever the reactor is operated by recording a measured value of a meaningful temperature.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

28. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 4.3 provides guidance for surveillance requirements on reactor coolant systems. Please discuss whether TAMU TS 14.4.3 is consistent with the standard guidance.

NSC Response:

1) The NSC has no emergency shutdown or emergency sump pumps. This standard is not applicable.

.2) A test of the emergency coolant sources is not included in TAMU TS 14.4.3 and will be added to follow the format and content of ANSI/ANS 15.1-2007, Section 4.3.

3) The :SAR has no requirements for in-service inspections. This standard is not applicable.

4) Pool water. is analyzed frequently, but there is no TS for this assay at the NSCR. A TS will be added to follow the format and content of ANSI/ANS 15.1-2007, Section 4.3.

'5) The NSC operates a large surface area non-enclosed pool in a facility with at least 180,000 cubic feet of free space, so the concentration of hydrogen is negligible. This standard is not applicable.

6) The NSC proposes to discontinue the testing of pH. pH measurements of low conductivity water are suspect. A test of the conductivity or pH, or both is not included in TAMU TS 14.4.3 and will be added to follow the format and content of ANSI/ANS 15.1-2007, Section 4.3.

14.4.4 Coolant Systems 14.4.4.1 Coolant Radioactivity Analysis Applicability This specification applies to the surveillance requirements for coolant radioactivity analysis.

Objective The objective is to verify a surplus of radioactive materials is not created in the coolant.

Specifications

1. A sample of the coolant shall be collected and analyzed for radioactive contamination at least weekly during periods of operation and at least annually when not operating.

Basis

1. Weekly sampling during operation will allow the establishment of trends of radioactive contamination from fuel or other sources.

14.4.4.2 Coolant Conductivity Applicability This specification applies to the surveillance requirements for coolant conductivity.

Objective The objectives are to minimize the possibility for corrosion of the cladding on the fuel elements and to minimize neutron activation of dissolved materials.

'Specifications

1. Conductivity of the bulk pool water shall be measured and recorded weekly.

Basis A small rate of corrosion continuously occurs in any water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is -required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limits provides acceptable control.

By limiting the concentrations of dissolved materials in the water, the radioactivity from neutron activation products is limited. This is consistent with the ALARA principle, and tends to decrease the inventory of radionuclides in the coolant system, which will limit/minimize personnel exposure during maintenance and operations.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

29. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15. 1-2007, Section 4.5 provides guidance for surveillances on ventilation system filter efficiency measurements and an operability check of any emergency exhaust systems. Please discuss whether the TAMU TS 14.4.4 is consistent with the standard guidance.

NSC Response: The TS states that the ventilation system shall be verified operable weekly.

While the NSC can operate a filter bank, it is not a part of our design basis and therefore not a part of our TS requirement.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-1-28 Responses to Request for Additional Information Chapter 14 - Technical Specifications

30. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-I15.1-2007, Section 5.1 provides guidance for including a "description of the site and of the facility including location and exclusion or restricted areas. "Please discuss whether TAMU TS 14.5 is consistent with the standard guidance.

NSC Response: A tech spec will be added to include a description of the facility location, as well as, the location of exclusion and restricted areas:

The Texas A&M University Nuclear Science Center is located on a rectangular six-acre site on the Texas A&M University campus 1,500 feet from the North-South runway of Easterwood Airport. The facility is six miles south of the city-center of Bryan (pop. 65,660), 3 miles southwest of the main campus of Texas A&M and two and one-half miles west-southwest of the lcfCIlege Station (pop. 67,890) in Brazos County, Texas. The facilityc 6i The NSC is surrounded by land owned and controlled by Texas A&M University and Easterwood Airport. A chain-link steel fence, which provides reasonable restriction of access to the site, defines the confines of the site. This fence also serves to define both the exclusion and restricted areas of the facility. The main entrance into the site is through an electrically operated chain-link steel gate at the east end of the site. Located within the boundaries of the -site are the reactor confinement building, reception room, laboratory building, mechanical equipment room, cooling system equipment, holding tanks, and other storage and 'support buildings.

The attached Diagram showing the site boundary, which defines the exclusion and restricted areas will be added to the SAR.

-~P-tI I

0r ta Note Th stebunar as dfte flieexcusio an retricd areU

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

31. TAMU TS 14.6.1.3(1) includes a note establishing conditions for when a reactor operator or trainee can be replaced by maintenance personnel if there is a SeniorReactor Operator present.

The note references TAMU TS 14.1.29(b). Please clarify what conditions are being exempted in the note associated with this TS.

NSC Response: The reference to TAMU TS 14.1.29(b) should be corrected to TAMU TS 14.1.36. The clarification is that for maintenance procedures that require the reactor to be unsecured, maintenance personnel may replace the Reactor Operator or Reactor Operator Trainee. For example, a control rod inspection requires a control rod to be removed from the core. During this removal the reactor is unsecured, and so the staffing requirement of TAMU TS 14.6.1.3(1) is in force. Rather than requiring a Senior Reactor Operator, a Reactor Operator or Reactor Operator Trainee, AND maintenance personnel to be on site, the note allows only a Senior Reactor Operator and one other person to be on site.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

32. NUREG-153 7 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 6.1.4 references ANSI/ANS-15. 4-1.988 (R1999), "Selection and Training of Personnel for Research Reactors" as the standard for selection and training of personnel at research reactors. Please discuss whether TAMU TS 14.6.1.4 is consistent with the standard guidance.

NSC Response: TAMU TS 14.6.1.4(1) is consistent with the standard guidance. TAMU TS 14.6.1.4(2)(b) will be modified to include medical examinations in the TS.

14.6.1.4 Selection and Training of Personnel

2. Requalification Program

b. Scope: Scheduled lectures, written examinations, medical examinations, and evaluated console manipulations insure operator proficiency is maintained.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

33. NUREG-153 7, Chapter 12.1 Conduct of Operations, Organizations states that the organization shall meet the non-power reactor standard ANSI/ANS 15.1-2007. ANSI/ANS 15.1.6. 2.2, Review and Audit Groups, Quorums states not less than one-half of the membership where operating staff does not constitute a majority is considered as quorums: TAMU TS 14.6. 2.2 does not specify the composition of the Safety Review Committee (SRC), just the number of members. Please discuss the composition of the SRC and the number of members from operating staff NSC Response: We will add the following to our TS 14.6.2.1 Reactor Safety Board (RSB)

The Reactor Safety Board shall be comprised of at least 3 voting members knowledgeable in fields which relate to Nuclear Safety. One of these members will serve as the Chairman. If the Chairman is unable to attend one or a number of committee meetings he may designate a committee member as Chairman Pro-tem. The members are appointed by the Office of the President of TAMU to serve one year terms. It is expected that the members will be reappointed each year as long as they are willing to serve so that their experience and familiarity with the past history of the NSC will not be lost to the committee. The Director of the NSC, TAMU Radiological Safety Officer, Head of the Department of Nuclear Engineering, and a senior member of the NSCR Radiation Protection Office shall be ex-officio members of the RSB.

14.6.2.2 RSB Charter and Rules The operations of the RSB shall be in accordance with a written charter, including provisions for:

1. Meeting Frequency: The RSB shall meet at least semi-annually at intervals not to exceed 8 months. (Note: The facility license requires a meeting at least once per year and as frequently as circumstances warrant consistent with effective monitoring of facility activities).

2. Quorum: A quorum is comprised of 3 voting members of the RSB excluding ex-officio members.

3. Voting Rules: On matters requiring a vote, if only a quorum is present a unanimous vote of the quorum is required; otherwise a majority vote is required.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

34. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 6.4 includes guidance for procedures that shall be prepared and approved to cover a list of specific activities. One of the activities is maintaining exposures and releases as low as reasonably achievable (ALARA). Please discuss whether TAMU TS 14.6.3 is consistent with the standard -guidance.

NSC Response: Maintaining exposures and releases as low as reasonably achievable (ALARA) will be added to TAMU TS 14.6.3.

14.6.3 Procedures Written operating procedures shall be prepared, reviewed, and approved before initiating any of the activities listed in this section. The procedures shall be reviewed and approved by the NSC Director or his designated alternate, theReactor Safety Board, and shall be documented in a timely manner. Procedures shall be adequate to ensure the safe operation of the reactor but shall not preclude the use of independent judgment and action should the situation require such.

Operating procedures shall be in effect for the following items:

1. Startup, operation, and shutdown of the reactor;

2. Fuel loading, unloading, and movement within the reactor;

3. Control rod removal or replacement;

4. Routine maintenance of the control rod, drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety;

5. Testing and calibration of reactor instrumentation and controls, control rod drives, area radiation monitors, facility air monitors, and the ventilation system;

6. Administrative controls for operations, maintenance, and conduct of irradiations and experiments, that could affect reactor safety or core reactivity;

7. Civil disturbances on or near the facility site;

8. Implementation of required plans such as emergency or security plans; and

9. Actions to be taken to correct specific and foreseen potential malfunctions of systems, including responses to alarms and abnormal reactivity changes.

10. Personnel radiation protection program to maintain exposures and releases as low as reasonably achievable (ALARA).

Substantive changes to the above procedures shall be made effective only after documented review and approval by the NSC Director and the RSB. Minor modifications or temporary changes to the original procedures that do not change their original intent may be made by the NSC Director or his designated alternate. All such temporary changes shall be documented and subsequently reviewed by the RSB.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

35. NUREG-1537 states that the format and content of the TSfollow that of ANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 6.4 includes guidance for procedures that shall be prepared and approved to cover a list of specific activities. TAMU TS 14.6. 3 (5) does not include all TS surveillance procedures. For example: testing of the ventilation system is not included (refer to TAMU TS 14.4.4). Please discuss the basis for not including a requirement for these surveillance procedures to be designated by TAMU TS 14.6. 3 as procedures requiring Reactor Safety Board and NSC Director approval:

NSC Response: The NSC will add wording in TS 14.6.3 to include the testing of the ventilation system.

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

36. NUREG-1537 states that the format and content of the TSfollow that ofANSI/ANS 15.1.

ANSI/ANS-15.1-2007, Section 6 6.1 discusses the actions to be taken in case of a safety limit violation and specifically mentions that a report shall be prepared that includes a discussion of the effect of the violation upon the health and safety of the public. Please discuss whether TAMU TS 14.6.3 is consistent with the standard guidance.

NSC Response: TAMU TS 14.6.6.2(1)(a) will be modified to include the words "that includes a discussion of the effect of the violation upon the health and safety of the public."

Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center License No. R-83 Docket No. 50-128 Responses to Request for Additional Information Chapter 14 - Technical Specifications

37. The requirements of 10 CFR Section 50.36 specifically identify the following three conditions as requiring record retention until the Commission terminates the license of the reactor:

a. Exceeding a safety limit.

b. Failure of an automatic safety function to function as required.

c. Failure to meet a Limiting Condition for Operation.

Please discuss whether TAMU TS 14.6. 7 is consistent with the regulation.

NSC Response: TAMU TS 14.6.7 will be modified to include these three conditions.

14.6.7.4 Records to be retained until the Commission terminates the license of the reactor

1. Exceeding a safety limit.

I

2. Failure of an automatic safety function to function as required.

3. Failure to meet a Limiting Condition for Operation.