ML20097C637

Univ. of Texas - Analysis of the Neutronic Behavior

ML20097C637
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Site:	University of Texas at Austin
Issue date:	12/31/2019
From:	Oregon State University
To:	Office of Nuclear Reactor Regulation
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ANALYSIS OF THE NEUTRONIC BEHAVIOR OF THE NUCLEAR ENGINEERING TEACHING LABORATORY REACTOR AT NETL Neutronic Analysis THE UNIVERSITY OF TEXAS Submitted By:

Radiation Center Oregon State University Corvallis, Oregon December 2019 December 2019

Table of Contents

1.

Introduction......................................................................................................................... 4

2.

Summary and Conclusions of Principal Safety Considerations......................................... 4

3.

Reactor Fuel........................................................................................................................ 4

4.

Reactor Core.................................................................................................. :.................... 6

5.

Model Bias.......................................................................................................................... 7

6.

Bum up Calculations......................................................................................................... I 0

7.

Current Core Configuration.............................................................................................. 11

8.

Limiting Core Configuration............................................................................................ 17

9.

Summary........................................................................................................................... 22 NETL Neutronic Analysis 2

December 2019

List of Figures Figure 1 - TRI GA Stainless Steel Clad Fuel Element Design used in the NETL Core................ 5 Figure 2-Schematic Illustration of the NETL Showing the Current Core Configuration............. 6 Figure 3 -Horizontal and Vertical Cross-sections of the NETL MCNP Model at BOL................ 7 Figure 4 - Reactivity (including bias) of 80 Different BOL Critical Core Configurations............. 8 Figure 5 - Reactivity (including bias) of 36 Different BOL Critical Core Configurations............. 9 Figure 6 - Vertical Cross-section of Current Core Configuration MCNP Model......................... 11 Figure 7 - Current Core Power-Pyr-Element (in kW) Distribution at 1.1 MW............................ 11 Figure 8 - Current Core Configuration Prompt Temperature Coefficient, ap, as a Function of Temperature................................................................................................................. 14 Figure 9 - Current Core Configuration Moderator Void Coefficient............................................ 15 Figure 10 - Current Core Configuration Moderator Temperature Coefficient............................. 15 Figure 11 - Vertical Cross-section of Limiting Core Configuration MCNP Model..................... 17 Figure 12-Limiting Core Configuration Power-Per-Element Distribution at 1.1 MW............... 18 Figure 13 - Limiting Core Configuration Prompt Temperature Coefficient, ap, as a Function of Temperature................................................................................................................. 20 Figure 14 - Limiting Core Configuration Moderator Void Coefficient........................................ 20 Figure 15 - Limiting Core Configuration Moderator Temperature Coefficient........................... 21 List of Tables Table 1 - Characteristics of Stainless Steel Clad Fuel Elements.................................................... 5 Table 2 - Summary of Bumup Steps.............................................................................................. 10 Table 3 - Beff and Prompt Neutron Lifetimes for Current Core Configuration............................. 12 Table 4-Current Core Rod Worth Calculations.......................................................................... 13 Table 5 - Current Core Configuration Prompt Temperature Coefficient...................................... 14 Table 6-K-Effective Calculations Used to Determine Current Core Power Defect................... 16 Table 7 - Beff and Prompt Neutron Lifetimes for Limiting Core Configuration........................... 18 Table 8 - Limiting Core Configuration Rod Worth Calculations................................................. 19 Table 9-Prompt Neutron Lifetimes in Limiting Core Configuration......... Error! Bookmark not defmed.

Table 10- Limiting Core Configuration Prompt Temperature Coefficient.................................. 20 Table 11 -K-Effective Calculations Used to Determine Limiting Core Power Defect................ 21 Table 12-Limiting Core Hot Channel Power Summary............................................................. 22 NETL Neutronic A~alysis 3

December 2019

1.

Introduction This report contains the results of investigation into the neutronic behavior of the Nuclear Engineering Teaching Laboratory reactor (NETL) at the University of Texas Austin. The objectives of this study were to: 1) create a model of the NETL to study the neutronic characteristics, and 2) demonstrate acceptable reactor performance and safety margins for the NETL core under normal conditions.

2.

Summary and Conclusions of Principal Safety Considerations The conclusion of this investigation is that the MCNP model does an acceptable job of predicting behavior of the NETL core. As such, the results suggest that the current NETL core can be safely operated within the parameters set forth in the technical specifications. Discussion and specifics of the analysis are located in the following sections. The final sections of this analysis provide suggestions for a limiting core configuration.

3.

Reactor Fuel The fuel utilized in the NETL is standard TRI GA fuel manufactured by General Atomics. The use of low-enriched uranium/zirconium hydride fuels in TRIGA reactors has been previously addressed in NUREG-1282 [l]. This document reviews the characteristics such as size, shape, material composition, dissociation pressure, hydrogen migration, hydrogen retention, density, thermal conductivity, volumetric specific heat, chemical reactivity, irradiation effects, prompt-temperature coefficient of reactivity and fission product retention. The conclusion of NUREG-1282 is that TRI GA fuel, including the fuel utilized in the NETL, is acceptable for use in reactors designed for such fuel.

The design of standard stainless steel clad fu~Lutilized in the NETL is shown in Figure 1. Stainless steel clad elements used at NETL all have fuel alloy length of 38.1 cm. The characteristics of standard fuel elements are shown in Table 1.

NETL Neutronic Analysis 4

December 2019

STAINLESS STEa TOP EN) OF ATTING ei

~

1B GRAPHITE 2.6 1N.

=

STAINLESS STEEL TUBE CI..AOOING THICKNESS 0.02 1N.

ZIRCONIUM HYDRIDE-8.5WT%

URANILJM...

15 IN.

1.475 IN.

MOLYDISC 0.08 MM THleK"=

qf

~

=

1_1 GRAPHITE 3.7 IN.

STAINLESS STEEL

) ~I BOTTOM END ATTING J

Figure 1-TRIGA Stainless Steel Clad Fuel Element Design used in the NETL Core Table 1 - Characteristics of Stainless Steel Clad Fuel Elements Uranium content mass %

8.5 BOL 235U enrichment mass % U 19.75 Ori inal uranium mass 37 0.25 1.435 1.475 T

e 304 SS 0.020 15 1.43 2.6 3.7 0.8 NETL Neutronic Analysis 5

December 2019

4.

Reactor Core The NETL core is a seven-ringed hexagonal grid array (labeled A through G) with 121 positions mostly composed of stainless-steel-clad standard TRI GA fuel elements. The current core configuration contains 113 fuel elements (including three fuel-followed control rods, i.e. FFCRs).

The core also contains an air-followed transient rod in C-1, a central thimble in A-1, several non-fueled locations that allow for a larger irradiation facility (in positions E-11, F-13 and F-14 ), a startup source in G-32, and a pneumatic transfer (Rabbit) irradiation facility in G-34, and an empty position G-26. The reactor is controlled by three electromagnetic control rods (Shim I, located in D-6; Shim II, located in D-14; and Regulating, located in C-7) and a pneumatic air-followed control rod (Transient, located in C-1 ), which utilize borated graphite (84C) as a neutron poison.

Fuel temperature is measured by an instrumented fuel element (JFE) located in 8-3. The current core configuration is shown in Figure 2.

Rabbit G21 635 2928 G20 10815 G36 2925 FOl 3504 G18 3496 G2 6142 G17 G3 5919 2960 F-04 6143 G4 3700 G6 2952 Figure 2 - Schematic Illustration of the NETL Showing the Current Core Configuration Detailed neutronic analyses of the NETL core were undertaken using MCNP6.2 [2]. MCNP6.2 is a general purpose Monte Carlo transport code which permits detailed neutronic calculations of complex 3-dimensional systems. It is well suited to explicitly handle the material and geometric heterogeneities present in the NETL core. The original input deck for the NETL model was developed at UT Austin and modified by Oregon State University. Facility drawings provided by the manufacturer at the time of construction of the facility were used to define the geometry of the core and surrounding structures. The geometry of the stainless steel clad fuel elements and control NETL eutronic Analysis 6

December 2019

rods were based upon the manufacturing drawings. Representative cross-sectional views of the MCNP model (of the initial core loading) are shown in Figure 3.

Figure 3-Horizontal and Vertical Cross-sections of the NETL MCNP Model at BOL The NETL reactor initially achieved criticality in March of 1992, however all of the fuel ( except for the fresh FFCRs) was previously used at other facilities. Most ofit came from a previous reactor on campus at Taylor Hall, but there were other sources as well. This made the beginning-of-life (BOL) fuel isotopic determination difficult. UT Austin performed a SCALE analysis to bum the fuel in conjunction with the given bumup records. The SCALE outputs were used to create BOL fuel isotopics for the MCNP runs.

5.

Model Bias Using critical rod height data from the first few months of NETL operation, a series of MCNP analyses based upon various critical rod heights were performed to determine the bias of the model.

This bias represents such things as differences in material properties that are difficult to determine or unknown (i.e., exact composition of individual fuel meats and trace elements contained therein) or applicability of cross section data sets used to model the reactor (i.e., interpolation between temperatures). As a result, the validation of the model was based upon the ability of the code to accurately predict criticality as compared with measurements made on the reactor in early 1992.

NETL Neutronic Analysis 7

December 20 I 9

A criticality calculation was performed using cold clean critical core configuration information from 3/23/1992. The k-effective of this configuration was 0.99393 +/- 0.00013, or -$0.87 +/- $0.04.

Eighty different critical core configurations were then analyzed to determine how they bounded around the bias of this initial critical configuration. Figure 4 shows these 80 configurations with respect to the bias run. All of these kcode calculations utilized 500,000 neutrons per cycle for 200 total cycles (175 active cycles).

Reactivity (including bias)

$1.00 l

• • ,... _.,,,,_ ~_.,....._..,__.._... _;*_

$(1.00)

$(2.00)

$(3.00)

$(4.00)

$(5.00)

$(6.00) 0 10 20 30 40 50 60 70 80 Critical Configuration Number Figure 4 - Reactivity (including bias) of 80 Different BOL Critical Core Configurations There appears to be significant deviation in the first 40 configurations. Note that most of these configurations are at low power but some are at high power. Most of the configurations with significant deviation are the high power runs, which would indicate that either the model is inaccurate or there is evidence of another problem. If the first 44 runs are ignored (if runs after 5/5/92 are observed), the data looks more accurate (see Figure 5), with an average of -$0.23.

NETL Neutronic Analysis 8

December 2019

$0.60

$0.40

$0.20

$(0.20)

$(0.40) *

$(0.60)

$(0.80) 45 50 Reactivity (including bias) 55 60 65 Critical Configuration Number 70 75 80 Figure 5 - Reactivity (including bias) of 36 Different BOL Critical Core Configurations Note that these latter 36 configurations include some full power operations ( cases #70-72, 76, 78 and 80). There is only one outlier over +/-$0.60 ( case #51 ), which would indicate that there were inconsistencies between high power operations during the first few months of operation. Other evidence, such as lower-than-expected fuel temperatures at these supposed high-power levels, would also indicate that something was inconsistent during the first few months of operation.

Thus the model bias that will be used for this study is -$1.10 (the -$0.23 bias plus -$0.87 bias).

This bias represents such things as differences in material properties that are difficult to determine or unknown (i.e., lack of manufacturer mass spectroscopy data on the exact composition of individual fuel meats and trace elements contained therein) or applicability of cross section data sets used to model the reactor (i.e., interpolation between temperatures). A large source of error is the uncertainty of the contents of the BOL fuel meats, as all of the fuel (except for the FFCRs) was previously irradiated. Without knowing the exact bumup and previous grid location of these elements, it is nearly impossible to accurately determine their fuel compositions.

This bias will be used to determine reactivity values in the following sections.

9 December 2019

6.

Burnup Calculations After performing the initial model bias calculations, a series of MCNP BURN calculations were performed to burn the NETL fuel to its current core configuration which was established in February 2018. This was a very detailed process as NETL is a very active facility and experienced many different core configurations. Using the fuel move logs, it was determined that there were 18 significant different core configurations that needed to be modeled (see Table 2). Each burnup step involved the fuel bumup for the specified amount of MW-days, parsing of the output fuel isotopics, then subsequent core model reconfiguration.

Table 2-Summary ofBurnup Steps Bumup From To MW-days Total FEs Note Step MW-days 1

3/19/1992 10/12/1995 9.201 9.201 87 Initial Fuel Load 2

10/12/1995 1/20/1998 5.276 14.477 87 NewIFE 3

1/20/1998 6/19/1998 2.789 17.266 87 Fuel Swapped Out/Add Rabbit 4

6/19/1998 3/4/1999 6.376 23.642 87 New IFE 5

3/4/1999 11/12/1999 7.671 31.3 13 90 Add 3 Fuel Elements 6

4/6/2000 6/29/2000 3.444 34.757 89 Core Reload 7

6/29/2000 1/29/2001 1.919 36.676 92 3L Experiment 8

1/29/2001 7/30/2001 9.138 45.814 92 3L Experiment with New IFE 9

7/30/2001 7/22/2002 21.508 67.322 95 Add 3 Fuel Elements 10 7/22/2002 11/13/2002 13.966 81.288 95 Fuel Shuffle 11 11/13/2002 4/1/2004 24.933 106.221 103 Add 8 New Fuel Elements 12 7/26/2004 7/13/2005 15.71 121.931 102 3L Experiment Core Reload 13 7/13/2005 7/11/2006 22.983 144.914 104 Add 2 Fuel Elements 14 7/11/2006 7/24/2007 41.732 186.646 104 Fuel Shuffle 15 7/24/2007 6/12/2008 18.347 204.993 108 Add 4 Fuel Elements 16 6/12/2008 6/24/2010 21.288 226.281 110 7L Experiment 17 6/24/2010 1/15/2016 73.587 299.868 114 Remove 7L Experiment 18 1/15/2016 2/22/2018 38.026 337.894 114 NewIFE NETL Neutronic Analysis IO December 2019

7.

Current Core Configuration Once the burnup calculations were complete, the core was reconfigured to the current core configuration (as of 2/22/2018, see Figure 6). The next series of calculations were then performed to determine various neutronic characteristics of the NETL.

Figure 6 - Vertical Cross-section of Current Core Configuration MCNP Model Core Power Distribution F4 flux tallies were used to determine the power-per-element. The tallies output as a fluence per fission neutron. These units were converted to power density (W /cm3) which were then converted to power-per-element. The individual power-per-element values (in kW) are shown in Figure 7.

626 Empty 627 5.74 628 5.61 630 13.75 624 5.13 F21 6.80 F22 7.61 F23 7.89 F24 7.34 F26 6.61 632 Source 623 5.49 F20 7.34 E17 8.98 E18 10.10 E19 10.40 E21 9.07 F27 7.42 633 6.31 622 5.76 F19 7.77 E16 E22 10.28 F28 8.44 634 Rabbit F18 7.79 ElS 8.37 635 6.38 7.03 E14 9.61 BOS 806 15.12 D18 13.20 E24 10.60 F30 7.79 636 5.74 E13 8.39 D10 10.94 C07 13.72 804 15.4 AOl CT COl Trans D01 12.70 EOl 9.62 FOl 7.18 7.23 E12 10.56 D09 12.26 C06 13.50 803 15.82 802 14.14 D02 13.18 E02 11.38 F02 8.23 62 5.91 F14 Empty Ell Empty DOB 12.16 COS 13. 75 C04 14.91 F03 8.93 63 6.52 616 6.53 F13 Empty ElO 10.95 D07 9.07 64 6.84 615 5.59 F12 7.44 E09 8.71 10.01 E07 10.95 EOS FOS 8.39 GS 6.57 614 4.96 Fll 6.36 FlO 7.47 F09 8.34 FOS 8.06 F06 7.35 66 5.95 612 5.20 611 5.79 610 6.14 GS 5.71 Figure 7 - Current Core Power-Per-Element (in kW) Distribution at 1.1 MW NETL Neutronic Analysi 11 December 20 I 9

The red highlighting indicates the hottest fuel element locations, which are in B-1 and B-2, with a maximum power of 15.93 kW (at a total maximum core power of 1.1 MW). B-2 is actually slightly higher than B-1 (15.931 kW vs. 15.929 kW) but both are within the 2-sigma error of 0.04 kW.

Effective Delayed Neutron Fraction and Prompt Neutron Generation Time MCNP outputs effective delayed neutron fraction (~eff) and prompt neutron lifetime when using the KOPTS card. Nine different MCNP calculations (the same calculations used in the following Core Excess section) were used to determine Petiand prompt neutron lifetime (see Table 3).

Table 3 - Perr and Prompt Neutron Lifetimes for Current Core Configuration Case Prompt Neutron Error (s)

Peff Generation Time (s)

Trans fully in 47.62 7.543 0.00705 Trans fully out 46.868 7.111 0.00716 Reg fully in 48.08 7.824 0.00707 Reg fully out 46.718 6.961 0.00707 Shim I fully in 48.023 7.748 0.00702 Shim I fully out 46.777 6.974 0.00705 Shim II fully in 48.104 7.684 0.00717 Shim II fully out 46.708 7.086 0.00713 All Rods Out 45.824 6.626 0.00720 Average 47.191 7.284 0.00710 The average effective delayed neutron fraction PetTwas calculated to be 0.00710 +/- 0.00007. This is in reasonable agreement with values predicted in other LEU TRI GA cores (i.e., Oregon State University Peff= 0.0076 [3], University of Maryland ~eff= 0.007 [ 4]) and also the value historically used for the NETL of Peff = 0.007. The value Peff = 0.007 will be used to express all dollar values of reactivities in this report.

The average prompt neutron generation time is 4 7.1 91 +/- 7.284 seconds.

NETL Neutronic Analysis 12 December 2019

Core Excess, Control Rod Worth and Shutdown Margin Nine different MCNP calculations were performed to determine core excess, control rod worth, and shutdown margin. Core excess is calculated as the reactivity of all rods withdrawn from the core. Control rod worths and shutdown margin were calculated by determining a critical state of the reactor with one rod full inserted and the other three rods banked at the same height, then fully withdrawing the previously-inserted rod. The resulting values (with comparison to values measured at NETL) are shown in Table 3.

Table 4 - Current Core Rod Worth Calculations MCNP MCNP MCNP Experimental Case k-effective k-effective Difference Rod Full-In Rod Full-Out Rod Worth Reactivity Transient 1.00035 1.02354

$3.24

$3.44

-$0.20 Re1n1lating 0.99978 1.02214

$3.13

$3.18

-$0.05 Shim 1 1.00078 1.02248

$3.03

$3.09

-$0.06 Shim2 1.00014 1.0211

$2.93

$2.94

-$0.01 All Rods Out 1.04118

$6.75

$6.06

$0.69

( Core Excess)

MCNP appears to accurately calculate the individual rod worths. The Regulating, Shim 1 and Shim 2 rods are all within the margin of error (which is approximately +/-$0.06 for each case).

These calculations show a core excess of $6.75 +/- $0.03. This is below the technical specification limit of$7.00. The core excess was measured by NETL to be $6.06 on 3/6/18. MCNP appears to have over-estimated core excess by approximately $0. 70. This could be due to a variety of reasons, such as only modeling the fuel elements as one single material per element, thus some bumup resolution is lost as the fuel does not bum uniformly throughout.

The technical specification definition of shutdown margin is the minimum reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems starting from any permissible operating condition (the highest worth MOVEABLE EXPERIMENT in its most positive reactive state, each SECURED EXPERIMENT in its most reactive state), with the most reactive rod in its most reactive position, and that the reactor will remain subcritical without further operator action." The most reactive rod is the Transient rod.

NETL Neutronic Analysis 13 December 2019

Total rod worth minus the Transient rod is $9.09 +/- $0.06. NRC shutdown margin is this value minus the core excess, which would be $2.34 +/- $0.06, which is far above the technical specification limit of$0.29.

Prompt Fuel Temperature Coefficient The prompt-temperature coefficient associated with the NETL fuel, ap, was calculated by varying the fuel meat temperature while leaving other core parameters fixed. The MCNP model was used to simulate the reactor with all rods out at 293, 600, 900, 1200 and 2500 K. The prompt-temperature coefficient for the fuel was calculated at the mid-point of the four temperature intervals. The results are shown in Figure 8 and tabulated in Table 5. Results from GA were added to show similarity [5]. The prompt-temperature coefficient is observed to be negative for all evaluated temperature ranges with decreasing magnitude as temperature increases. The coefficient has a value of -1.3¢/°C at 446.8 K, which is similar to the value of -0.01 %/°C stated in the original SAR [6].

~

... =,-.

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ic: -$0.020

~...

C. ~

~ ~ -$0.015

~-

- =

~.2! -$0.010

~~

t ~ -$0.005

=U i=..

$0.000 200

-.~

+ NETL

• GA 700 1200 1700 2200 Temperature (Kelvin)

Figure 8 - Current Core Configuration Prompt Temperature Coefficient, aF, as a Function of Temperature Table 5 - Current Core Configuration Prompt Temperature Coefficient Fuel Temperature [K]

Prompt Temperature Coefficient [$/°C]

446.8

-$0.0130 750

-$0.0208 1050

-$0.0092 1850

-$0.0010 NETL Neutronic Analysis 14 December 2019

Moderator Void Coefficient The moderator void coefficient of reactivity was also determined using the MCNP model. The voiding of the core was introduced by uniformly reducing the density of the liquid moderator in the entire core. The calculation was performed from 0% to 1 00% voiding at 10% intervals. The void coefficient was negative for every interval and steadily decreased, as can be seen in Figure 9.

$0.00 "O

-$0.20 "O

~ =

• = -$0.4o

~ -~ ;... -$0.60 i= ~

f ~ c.. -$0.80 Qi = Qi "g U c.. -$1.00

~

~ -$1.20

-$1.40 0

20 40 60 80 100 Percent Void Figure 9 - Current Core Configuration Moderator Void Coefficient Moderator Temperature Coefficient The moderator temperature coefficient of reactivity, UM, was determined by varying the moderator density with respect to temperature within the MCNP model from the expected operating temperature range of 20°C to 50°C (using Engineering Toolbox [7] to determine water density).

The results are shown in Figure 10. The moderator temperature coefficient is calculated to be slightly positive from 25°C to 30 °C and from 45 °C to 50 °C, but these changes are less than

$0.01/°C and both points (with 2-sigma error) are bounded around zero. The moderator temperature coefficient appears to be negligible.

Qi.. = -

~..

$0.020

$0.015

$0.010 Qi -

c.. = -

$0.005 e -~~

~ ~ ~ $0.000

.. ~ C. $

= = "' - 0.005

°e U '-' -$0.010 Qi "O =

~

-$0.015

-$0.020 20 0

0

-~

~r 25 30 35 40 45 50 Moderator Temperature (°C)

Figure 10 - Current Core Configuration Moderator Temperature Coefficient NETL Neutronic Analysis 15 December 2019

Power Coefficient of Reactivity The power coefficient of reactivity, otherwise known as power defect, is the amount of reactivity required to overcome the temperature feedback during the rise to power. This is modeled by analyzing two MCNP decks that are similar except for the neutron cross-sections used. Two k-effective calculations were performed with all rods out, one using cross sections at 293K (low power) and one using cross sections at 600K (full power). The results are seen in Table 6.

Table 6-K-Effective Calculations Used to Determine Current Core Power Defect Case MCNP k-effective Standard Deviation Reactivity Error (2-sigma)

Low Power 1.04118 0.00012

$6.75

$0.03 Full Power 1.01327 0.00010

$2.94

$0.03 Power defect is simply the difference in reactivity between these two cases; thus the power defect is $3.81 +/- $0.05.

NETL Neutronic Analysis 16 December 2019

8.

Limiting Core Configuration This section will suggest a limiting core configuration that utilizes fresh fuel to improve reactor efficiency while maintaining proper safety margins. The NETL limiting core configuration is a core that completely consists of fresh fuel.

Figure 11 shows the suggested limiting core configuration. For this analysis, it is suggested that the core is loaded with 84 fresh fuel elements (including FFCRs), which will provide just under the license limit of$7.00 core excess ($6.93 +/- $0.07). This is comparable to the original 1992 BOL core configuration, which was measured to have a $6.38 core excess on a core of 87 lightly-irradiated fuel elements. This configuration will provide maximum flux to the beam port facilities while maintaining safety margins.

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1_*-'+

~-

~..... *~*r1

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1,'!ij Figure 11-Vertical Cross-section of Limiting Core Configuration MCNP Model ETL Neutronic Analysis 17 Decem her 20 l 9

Core Power Distribution Figure 12 shows the power-per-element (in kW) in the suggested limiting core configuration.

Figure 12 - Limiting Core Configuration Power-Per-Element Distribution at 1.1 MW The hottest fuel element in now in location B-5. This makes sense as the core is more shifted to the northwest, which would better centralize the location of the maximum power production around B-5. Also, the hottest power-per-element at I.I MW is now 22.14 +/- 0.06 kW, which is higher than the current core hot channel, due to a lower fuel loading concentrating more power at the center of the core.

Effective Delayed Neutron Fraction and Prompt Neutron Generation Time Once again using the "KOPTS" card and running nine cases, the effective delayed neutron fraction Betr and prompt neutron generation times were calculated Table 7 - Petr and Prompt Neutron Lifetimes for Limiting Core Configuration Case Prompt Neutron Generation Time (s)

Error (s)

~eff Trans fully in 42.828 5.531 0.00743 Trans fully out 42.721 5.024 0.00725 Reg fully in 43.764 5.502 0.00732 Reg fully out 41.951 4.985 0.00742 Shim I fully in 43.546 5.616 0.00737 Shim I fully out 42.407 5.104 0.00737 Shim II fully in 43.614 5.458 0.00733 Shim II fully out 42.261 5.200 0.00728 All Rods Out 42.024 4.965 0.00742 Average 42.791 5.265 0.00735 NETL Neutronic Analysis 18 December 2019

The average ~etrwas calculated to be 0.00735 +/- 0.00007. There is a slight increase in ~etrcompared to the current core configuration, but for consistency, 0.007 will continue to be used to express all dollar values of reactivities in this report.

The average prompt neutron generation time is 42.791 +/- 5.265 seconds.

Core Excess, Control Rod Worth, and Shutdown Margin The same nine MCNP rod worth calculations were performed again for the limiting core configuration: Core excess, shutdown margin, and individual rod worths were calculated from these outputs and the reactivity values (with the bias taken into account) of each of these calculations are shown in Table 7.

Table 8-Limiting Core Configuration Rod Worth Calculations Case MCNP k-effective MCNP k-effective MCNPRod Rod Full-In Rod Full-Out Worth Transient 0.99886 1.02191

$3.22 Regulating 1.00024 1.03222

$4.43 Shim 1 1.00003 1.02431

$3.39 Shim2 1.0003 1.02857

$3.93 All Rods Out (Core Excess) 1.04257

$6.93 These calculations show a core excess of $6.93 +/- $0.07. This is below the technical specification limit of$7.00.

Now the most reactive rod is the Regulating, due to having more fuel near its vicinity and the power shifted to the northwest side of the core. Total rod worth minus the Regulating Rod is $10.53

+/- $0.16. NRC shutdown margin is this value minus the core excess, which would be $3.60 +/- $0.1 6, which is still far above the technical specification limit of $0.29.

Prompt Fuel Temperature Coefficient NETL Neutronic Analysis 19 December 2019

The results of the limiting core configuration prompt fuel temperature coefficient calculations are shown in Figure 13 and tabulated in Table 9.

-$0.025 t

-$0.020

=-

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I,;

t t -$0.015

=~

~ '-'

E-4.... -$0.010

~ ~

=...

. -. ~

~~

c.. 't -$0.005 ga t

$0.000 200 700 1200 Temperature (Kelvin)

+ NETL

• GA 1700 2200 Figure 13 - Limiting Core Configuration Prompt Temperature Coefficient, aF, as a Function of Temperature Table 9 - Limiting Core Configuration Prompt Temperature Coefficient Fuel Temperature rKl Prompt Temperature Coefficient r$/°Cl 446.8

-$0.01302 750

-$0.02081 1050

-$0.00928 1850

-$0.00105 These values are similar to the original BOL coefficients.

Moderator Void Coefficient Figure 14 shows the moderator void coefficient in the suggested limiting core configuration.

$0.00 - * * *

-$0.20 -

-=

-= $

~-

~....... - 0.40

= 0

~ *o > -$0.60 --

.... i:::~

=.,_ C

i

~ i.. -$0.80 -

-= 0

~

~ U ;-$1.00

-$1.20 -

-$1.40 0

20 40 60 80 100 Percent Void Figure 14 - Limiting Core Configuration Moderator Void Coefficient NETL Neutronic Analysis 20 December 2019

The void coefficient was negative for every interval and steadily decreased, similar to the current core configuration. The void coefficient is slightly more negative in the limiting core configuration, likely due to having more moderator in the core configuration.

Moderator Temperature Coefficient Figure 15 shows the moderator temperature coefficient m the suggested limiting core configuration.

CII -= -

OS -

$0.020

$0.015

$0.010 i - ---

e -~ ~ $0.oos

~ !5 t $0.000 ell Q.,

~ 8 e,$0.005

~

-$0.010 0

E

-$0.015

-$0.020

~-

20

~*

~*

• r 25 30 35 40 45 50 Moderator Temperature (°C)

Figure 15 - Limiting Core Configuration Moderator Temperature Coefficient Once again the moderator temperature coefficient appears to be negligible as it bounds around

$0.00 at all observed temperature ranges.

Power Coefficient of Reactivity The power coefficient of reactivity results are seen in Table 10.

Table 10 - K-Effective Calculations Used to Determine Limiting Core Power Defect Case MCNP k-effective Standard Deviation Reactivity Error (2-sigma)

Low Power 1.04231 0.00015

$6.90

$0.04 Full Power 1.01921 0.00010

$3.79

$0.03 Thus the power defect is $3.11 +/- $0.05. This is lower than the current core configuration's power defect, likely due to less resistance at the point-of-adding-heat due to the lower amount of zirconium-hydride in the core.

NETL Neutronic Analysis 21 December 2019

Hot Channel Power Summary The hot channel in the limiting core configuration was determined to be B-5. An fmesh calculation was performed to analyze a 20 by 20 mesh array to determine axial and radial power distributions.

Table 11 summarizes the results of this calculation.

Table 11 - Limiting Core Hot Channel Power Summary Hot Rod Hot Rod Hot Rod Hot Rod Core Hot Rod Thermal Peak Factor Axial Peak Radial Peak Effective Configuration Location Power [kW]

[Pmax/Pavg]

Factor Factor Peak Factor fPmaJPavgl fPmaJPavgl Limiting Core B6 22.14 1.691 1.296 1.017 2.229

9.

Summary MCNP6.2 was used to calculate fundamental and operational parameters for the Nuclear Engineering Teaching Laboratory Reactor to demonstrate the reactor's adherence to safety margins in the technical specifications. Values of fundamental parameters agree well with theoretical values. Values of operational parameters agree well with measured values, giving confidence in the model's ability to predict the viability of future core configurations. The results of this study indicate that the NETL can be operated safely within the Technical Specification bounding envelope and that its MCNP model can be used to predict future core configuration changes.

REFERENCES

[1]

NUREG-1282, "Safety Evaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors,' USNRC, August 1987.

[2]

C.J. Werner, et al., "MCNP6.2 Release Notes", Los Alamos National Laboratory, report LA-UR-18-20808 (2018).

[3]

"Safety Analysis Report for the Conversion of the Oregon State University TRI GA Reactor from HEU to LEU Fuel," Submitted by the Oregon State University TRIGA Reactor (2007).

NETL Neutronic Analysis 22 December 2019

[4]

"Analysis of the Neutronic Behavior of the Maryland University Training Reactor,"

Submitted by the Oregon State University Radiation Center to the Department of Energy (July 2017).

[5]

GA-7882, Kinetic Behavior of TRIG A Reactors, General Atomics (1967).

[6]

"Safety Analysis Report" Submitted by the University of Texas at Austin Nuclear Engineering Teaching Laboratory (January 2012).

[7]

Engineering Toolbox. Web. Accessed May 3rd, 2017.

Link: http://www.engineeringtoolbox.com/water-thermal-properties-d _ 162.html NETL Neutronic Analysis 23 December 2019

WALKER DEPARTMENT OF MECHANICAL ENGINEERING Nuclear Engineering Teaching Laboratory Pickle Research Campus R-9000

• Austin, Texas 78758

• 512-232-5380 *FAX 512-471-4589 nuclear. engr. utexas. edu

• wcharlton@austin. utexas. edu ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Geoffrey Wertz, P.E.

Non-Power Production and Utilization Facility Licensing Branch Division of Advance Reactors and Non-Power Utilization Nuclear Reactor Regulation March 31, 2020

SUBJECT:

Docket No. 50-602, Facility Operating License R-129 - Submission of Neutronic and Thermal Hydraulic Analysis for the University of Texas at Austin Research Reactor

REFERENCE:

October 18, 2018 letter: University-of Texas at Austin - Summary of Site Visit and Request for Schedule for Completion of the Reactor Analyses RE: Renewal of Facility Operating License No.

R-129 for The University of Texas at Austin Research Reactor (EPID NO. L-2017-RNW -0032)

Sir:

We respectfully submit neutronics and thermal-hydraulic analysis, attached. If you have any questions, please contact me at 512-232-5373 or whaley@mail.utexas.edu.

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

W. S. Charlton ATT:

(1)

Analysis of the Neutronic Behavior of the Nuclear Engineering Teaching Laboratory at the University of Texas (2)

Thermal Hydraulic Analysis of the University of Texas (UT) TRIGA Reactor