ML19165A022

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Oregon State Univ. - Analysis of the Neutronic Behavior of the Maryland University Training Reactor
ML19165A022
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Site: University of Maryland
Issue date: 07/31/2017
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Office of Nuclear Reactor Regulation
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Text

ANALYSIS OF THE NEUTRONIC BEHAVIOR

  • OF THE MARYLAND UNIVERSITY TRAINING REACTOR MUTR Neutronic Analysis Submitted By:

Radiation Center Oregon State University Corvallis, Oregon July 2017 July 2017

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..............................*......................................................................................... 10

6.

Beginning-of-Life (BOL) Core Configuration....................................................... :......... 10 Effective Delayed Neutron Fraction................................................,................................ 10 Core Excess, ControlRod Worth and Shutdown Margin....... :.......... :.............................. 11 Prompt Fuel Temperature Coefficient.................................................................. ~......... :. 12 Moderator Void Coefficient.............................................................................................. 13 Moderator Temperature Coefficient................................................................................. 13 Core Po_wer Distribution.................-................... ~.............................................................. 14 Bumup............................................................................................................................... 15

7.

Current Core Configuration.............................................................................................. 15 CoreExcess, Control Rod Worth and Shutdown Margin.................................................. 15 Prompt Fuel Temperature Coefficient..................................................... ;........................ 16 Moderator Void Coefficient.............................................................................................. 18 Moderator Temperature Coefficient.................................... :......... :.................................. 18 Core Power Distribution................................................................................................... 19 Control Rod Calibration Curves............................................................................ _........... 20

8.

Suggested New Core Configuration......,.......................................................................... 22 Effective Delayed Neutron Fraction..................................... :.......,................................... 23 Core Excess, Control Rod Worth, and Shutdown Margin................................................ 23 Prompt Fuel Temperature Coefficient.............................................................................. 24 Moderator Void Coefficient..................................................... -..................................-....... 25 Moderator Temperature Coefficient................................................................................. 25 Core Power Distribution........................ -........................................................................... 26

9.

Suggested Changes to Technical Specifications............................................................... 27 Change to Section 3.1...................................................................... '................................. 27 Change to Section 3.3.......................................................... *.............................................. 27 Change to Section 5.3....................................-................................................................... 28

10.

Summary..........................................................................................................,............-.... 29 MUTR Neutronic Analysis 2

July 2017

List of Figures Figure 1 - TRI GA Stainless Steel Clad Fuel Element Design used in the MUTR Core [2]......... 5 Figure 2 - Schematic Illustration of the MUTR Showing the Current Core Configuration............ 6 Figure 3 -Horizontal Cross-section of the MUTRMCNP Model................................ :*****************7 Figure 4 - Vertical Cross-section of the MUTR MCNP Model.................... ;................................. 8 Figure 5 -BOL Prompt Temperature Coefficient, m, as a Function of Temperature.................. 12 Figure 6 - BOL Moderator Void Coefficient................................................................................ 13 Figure 7 - BOL Moderator Temperature Coefficient.................................................................... 14 Figure 8 - Current Prompt Temperature Coefficient, <iF, as a Function of Temperature.............. 17 Figure 9 - Current Moderator Void Coefficient................................................. :.......................... 18 Figure 10- Current Moderator Temperature Coefficient.................................. ;........................... 19 Figure 11 - Shim I Rod Calibration Curves (Experimental vs. MCNP Prediction)................'......20 Figure 12-Shim II Rod Calibration Curves (Experimental vs. MCNP Prediction).................... 21 Figure 13....,... Regulating Rod Calibration Curves (Experimental vs. MCNP Prediction)............... 21 Figure 14-Historic MUTR Control Rod Worth Data.............,.................................................... 22 Figure 15 - New Core Prompt T~mperature Coefficient, aF, a:s a Function of Temperature.......24 Figure 16 -New Core Configuration Moderator Void Coefficient..............................................25 Figure 17 -New Core Configuration ModeratorTemperature Coefficient.................,................26 List of Tables Table 1 - Characteristics of Stainless Steel Clad Fuel Elements..................................................... 5 Table 2.-Physical Densities and Mass Fractions for Selected Core Components in the MCNP Model of the MUTR*.............................................. '........................................................ 9 Table 3 -BOL Rod Worth Calculations........................................................................................ 11 Table 4 - BOL Prompt Temperature Coefficient.............................................,.......................,.... 13 Table 5 - BOL Core Power Distribution....................................................................................... 14 Table 6 - Current Rod Worth Calculations................................................................................... 15 Table_ 7 - Current Prompt Temperature Coefficient............... :............... :.......... :........................... 17.

Table 8 - Current Core Power Distribution................................................................................... 19 Table 9-Suggested New Core Configuration (109 Element _Core)............................................. 23 Table 10 ~ New Core Configuration Rod Worth Calculations:.....................................................23 Table 11 - New Core Prompt Temperature Coefficient................................................................25 Table 12 - New Core Power Distribution................................,..................................................... 26 MUTR Neutronic Analysis 3

July 2017.

1.

Introduction This report contains the results of investigation into the neutronic behavior of the Maryland University Training Reactor (MUTR). The objectives of this study were to: 1) create a model of the MUTR to study the neutronic characteristics, 2) demonstrate acceptable reactor performance and safety margins for the MUTR core under normal conditions, and 3) suggest ways to improve performance of the MUTR.

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 MUTRcore. As such, the results suggest that the current MUTR core can be safely operated within the parameters set forth in the technical specifications; however, the MUTR core as currently loaded is unable to operate at 250 kW due to depleted fuel, which the MCNP model confirmed. Discussion and specifics of the analysis are located in the following sections. The final sections of this analysis provide suggestions for a new core configuration and suggested changes to the technical ~pecifications to accommodate the new core configuration.

3.

Reactor Fuel The fuel utilized in the MUTR is standard TRI GA fuel manufactured by General Atomics. The use of low-enriched uranium/zirconium hydride fuels in TRI GA reactors has been previously addressed in NUREG-1282 [1]. 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 MUTR, is acceptable for use in reactors designed for such fuel.

The design of standard stainless steel clad fuel utilized in the MUTR is shown in Figure 1 [2].

Stainless steel clad elements used at MUTR all have fuel alloy length of 38.1 cm.

The characteristics of standard fuel elements are shown in Table 1.

MUTR Neutronic Analysis 4

July 2017

Top fuel fltti~

Fuel polw (3)

Molyt:dcm.. rn pois.en disc Bottom fu&l filtil"YJ t.4J4 DIA.

NOM.(REF) 23.125 (REF) 1.370:g;:1..o.

cRE.F) 25.ars:f;ff?

Figure 1 - TRIG A Stainless Steel Clad Fuel Element Design used in the MUTR Core [2]

Table 1 - Characteristics of Stainless Steel Clad Fuel Elements Uranium content [ mass % ]

8.5 BOL 235U enrichment f mass % Ul 19.75 Original uranium mass f gm l 37 Zirconium rod diameter fin l 0.225 Fuel meat inner diameter fin l 0.25 Fuel meat outer diameter fin l 1.374 Cladding outer diameter fin l 1.414 Cladding material Type 304 SS Cladding thickness finl 0.020 Fuel meat length [in]

15 Graphite slug outer diameter fin]

1.291 Graphite slug length fin l 3.47 Molybdenum disc thickness f mml 0.8 MUTR Neutronic Analysis 5

July2017

4.

Reactor Core The MUTR core is a five-by-nine rectangular grid array (labeled B through F, 1 through 9) composed of stainless-steel-clad standard TRIG A fuel and aluminum-clad graphite reflector clusters (located in E2 and D2). The core also contains several non-fueled locations that house instrumentation (fission chamber in F9, ion chamber in F2), a plutonium-beryllium startup source in B6, and a pneumatic transfer (Rabbit) irradiation facility in C4. The current configuration established in 1974 includes 93 standard stainless-steel-clad fuel elements.

The reactor is controlled by three electromagnetic control rods (Shim I, located in E4; Shim II, located in E7; and Regulating, located in C6) which utilize borated graphite (B4C) as a neutron poison. Fuel temperature is measured by an instrumented fuel element (IFE) located in the southeast comer of the D8 fuel cluster. The current core configuration is shown in Figure 2.

THERMAi. COLUMN 9

8 7

MUTRCORE Figure 2 - Schematic Illustration of the MUTR Showing the Current Core Configuration MUTR Neutronic Analysis 6

July 20 17

Detailed neutronic analyses of the MUTR core were undertaken usmg MCNP6.1. l [3].

MCNP6. l. l 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 MUTR core. The original input deck for the MUTR model was developed by Dr. Ali Mohamed [ 4]. 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 rods were based upon the manufacturing drawings. Representative cross-sectional views of the MCNP model are shown in Figure 3 and Figure 4.

Figure 3 - Horizontal Cross-section of the MUTR MCNP Model MUTR Neutronic Analysis 7

July20 17

Figure 4 - Vertical Cross-section of the MUTR MCNP Model Fuel element meats were modeled as a homogeneous mixture of 235U, 238U, natural zirconium and hydrogen. Compositions of all significant materials are shown in Table 2.

MUTR Neutronic Analysis 8

July 2017

Table 2 - Physical Densities and Mass Fractions for Selected Core Components in the MCNP Model of the MUTR Material Physical Density [g/cm3]

Nuclide Mass Fraction 90zr 0.462087 91zr 0.100770 92Zr 0.154029 Standard 8.5 94zr 0.156095 wt%

5.95 96zr U-ZrH fuel 0.025148 lH 0.016871 23su 0.017000 23su 0.068000 Natural Mn 0.020 Type 304 SS Natural Cr 0.190 (Fuel Clad) 7.92 Natural Ni 0.095 Natural Fe 0.695 Graphite 1.6 12c 1.0000 Zirconium 6.5 Natural Zr 1.0000 Rod Pure 2.700 27Al 1.0000 Aluminum 27Al 0.975114 Natural Cr 0.003837 Aluminum 2.700 Natural Cu 0.005861 6061-T6 Natural Mg 0.008970 Natural Si 0.006218 lH 0.11 Water 1.000 160 0.89 14N 0.752308 15N 0.002960 160 0.231687 Air l.29E-03 170 0.000094 12c 0.000124 40Ar 0.012827 MUTR eutronic Analysis 9

July 201 7

5.

Model Bias Using critical rod height data from control rod calibrations performed at initial fuel loading, a series ofMCNP analyses based upon various core configurations 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 nnknown (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). 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 the summer of 1974.

Kcode calculations were performed using eleven different critical rod height configurations from control rod calibrations performed on June 201h, 1974. These MCNP calculations utilized 100,000 neutrons per cycle for 200 total cycles (150 active cycles). The final k-effectives were compiled and averaged, yielding a model bias of $1.97 +/- $0.05 (2-sigma error). While this model bias appears to be high, it is similar to bias observed at the NRAD facility [2]. MCNP appears to over-estimate criticality in small MTR-conversion-type TRlGA reactors, which may also be attributed to the zirconium cross sectional data [6]. This bias will be used to determine reactivity values in

  • the following sections.
6.

Beginning-of-Life (BOL) Core Configuration Effective Delayed Neutron Fraction The effective delayed neutron fraction for the MUTR core was calculated with MCNP6.1.1 by utilizing the expression:

k

/J =l--

p ejf k

p+d where kp is the system eigenvalue assuming fission neutrons are born with the energy spectrum for prompt neutrons, and kp+d is the system eigenvalue assuming fission neutrons are born with the appropriately weighted energy spectra for both prompt and delayed neutrons. Using the "totnu" card and running two identical cases, the effective delayed neutron fraction ~ eff was calculated to be 0.007035 +/- 0.000165. This is in reasonable agreement with values predicted in other LEU MlJTR Neutronic Analysis 10 July 2017

TRIGA cores (i.e., Oregon State University ~err= 0.0076 [5], Washington State University ~eff=

0.0075) and also the value historically used for the MUTR of ~eff = 0.007. The value ~ eff = 0.007 will be used to express all dollar values of reactivities in this report.

Core Excess, Control Rod Worth and Shutdown Margin Five different MCNP calculations were performed: (1) All control rods in, (2) Shim I rod out, Shim II and Reg rods in, (3) Shim II rod out, Shim I and Reg rods in, (4) Reg rod out, Shim I and Shim II rods in, and (5) All control rods out. Core excess, shutdown margin, and individual rod worths were calculated from these outputs and the reactivity values of these five MCNP calculations (with the bias taken into account) are shown in Table 3.

Table 3 - BOL Rod Worth Calculations Case MCNP k-effective Standard Deviation Reactivity Error (2-sigma)

All Rods In 0.97373 0.00018

-$5.82

$0.05 Shim I fully out 0.99922 0.00020

-$2.08

$0.06 Shim II fully out 0.99919 0.00019

-$2.09

$0.05 Reg fully out 0.99231 0.00020

-$3.08

$0.06 All Rods Out 1.03453 0.00019

$2.80

$0.05 These calculations show a core excess of $2.80 +/- $0.05. This is well-below the original technical specification limit of $3.50 but higher than the current technical specification limit of $1.12.

Individual rod worths are simply the absolute value of "all rods in" minus the absolute value of the worth of the individual rod. Thus the control rod worths of the rods are Shim I: $3.74 +/- $0.06,

. Shim II: $3.74 +/- $0.05, Reg: $2.75 +/- $0.06 and the total rod worth of$10.23 +/- $0.10. Experimental control rod worth calculations were unavailable for comparison, however the console log book indicates a "shutdown margin" (total rod worth) calculation was performed on 6/20/1974 using a prompt drop approximation. The three control rods were dropped from heights of 99.6%, 26.7%

and 100%, respectively. Power promptly dropped from 570 mW to 60 mW and an approximate "shutdown margin" of $8.50 was calculated. Although this is labeled "shutdown margin", we interpret the data to mean rod worth. Additionally, the accuracy of this datum is unknown. In comparison, the MCNP-predicted rod worths in this configuration, adding the full value of Shim I and Reg ($3.74 + $2.75 = $6.49) and approximating the value of Shim II at 26.7% withdrawn

( approximately $0. 7 5), yields a total shutdown margin of $7.24.

MUTR Neutronic Analysis 11 July 2017

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 OPERA TING condition and with the most reactive rod in its most reactive position." The most reactive rod is the Shim II rod. Total rod worth minus the Shim II rod is $6.49 +/- $0.10. NRC shutdown margin is this value minus the core excess, which would be $3.69 +/- $0.10, which is far above the technical specification limit of $0.50. It should be noted that the implication of the results is that the MUTR was initially loaded with more than enough fuel to achieve 250 kW.

Prompt Fuel Temperature Coefficient The prompt-temperature coefficient associated with the MUTR fuel, a F, 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 Error! Reference source not found. and tabulated in Table

4. Results from GA were added to show similarity [7]. 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 -1.6¢/°C at 446.8 K, which is similar to the value of -1.2¢/°C stated in the original SAR.

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-$0.004

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+ MUTR

  • GA T

700 1200 1700 2200 Temperature (Kelvin)

Figure 5 - BOL Prompt Temperature Coefficient, aF, as a Function of Temperature MUTR Neutronic Analysis 12 July 2017

Table 4 - BOL Prompt Temperature Coefficient Fuel Temperature [K]

Prompt Temperature Coefficient [$/0C]

446.8

-$0.0162 750

-$0.0l81 1050

-$0.0161 1850

-$0.0010 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 100% voiding at 10% intervals. The void coefficient was negative for every interval and steadily decreased, as can be seen in Figure 6.

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20 40 60 80 Percent Void Figure 6 - BOL Moderator Void Coefficient Moderator Temperature Coefficient 100 The moderator temperature coefficient ofreactivity, aM, was determined by varying the moderator density with respect to temperature within the MCNP model MUTR core from the expected operating temperature range of 20°C to 50°C (using Engineering Toolbox [8] to determine water density). The results are shown in Figure 7. The moderator temperature coefficient is calculated to slightly increase from 2o*c to 25 *c and from 45 *c to 50 *c, but these changes are on the order of $0.01/°C and all points (with 2-sigma error) are bounded around zero. The moderator temperature coefficient appears to be negligible.

MUTR Neutronic Analysis 13 July 20 17

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-'-.L-.L.-+--'--'--'--'-+-'--'--'--'-+-'--'--'--'--!--'--'--'--'--l 20 25 30 35 40 45 Moderator Temperature ('C)

Figure 7 - BOL Moderator Temperature Coefficient Core Power Distribution so 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 Table 5.

Table 5 - BOL Core Power Distribution 8

7 6

5 4

3 7390 7391 7378 7379 7354 7355 7395 7393 7168 7169 7333 7335 F

1.34 1.68 2.08 2.41 2.66 2.77 2.76 2.63 2.37 2.07 1.69 1.37 7389 7392 7377 7380 7353 7356 7397 7396 7167 7166 7334 7336 1.53 1.95 2.51 3.15 3.21 3.31 3.30 3.14 3.11 2.45 1.94 1.53 7161 7026 7398 259 7368 7365 7374 7375 304 7406 7342 7343 E

1.73 2.22 3.06 0.00 3.86 3.78 3.77 3.77 0.00 2.99 2.20 1.67 7028 7027 7399 7400 7367 7373 7376 7404 7405 7341 7344 1.82 2.37 3.05 3.79 3.88 3.95 3.77 3.64 2.92 2.31 1.75 7408 7409 7345 7346 7382 7383 7371 7372 7290 7330 7164 7165 D

1.79 2.37 2.98 3.42 3.75 3.89 3.85 3.59 3.23 2.76 2.23 1.70 7407 7160 7348 7347 7381 7384 7370 7369 7332 7331 7163 7162 1.63 2.14 2.68 3.10 3.40 3.71 3.45 3.18 2.88 2.51 2.00 1.54 7360 7357 7352 7349 7401 260 7388 7385 7362 7363 C

1.33 1.74 2.20 2.54 2.94 0.00 2.96 2.64 1.74 1.31 7359 7358 7403 7402 7387 7386 7361 7364 1.56 1.40 7339 7340 0.82 0.94 MUTR Neutronic Analysis 14 July 2017

The orange bundles indicate control rod clusters, yellow indicates the IFE, and red indicates the hottest fuel element location. The hottest fuel element in the core is located in the southeast comer of the bundle in E6, with a maximum power of 3.99 kW (at a total core power of 250 kW). The IFE produces greater than 50% of the hottest fuel element so it is sufficient in its current location.

Bumup The MUTR reports bumup each year (in MW-hr and grams 235U) as part of their annual report [9].

Summing the annually-reported bumup yields a total bumup of 557.272 MW-hr, or 23.22 MW-days, with a reported bumup of21.93 grams of 235U. Three annual reports were unable to be found

( 1996, 2001, 2003) but based upon operating trends around those years, the expected unaccounted-for burnup should not be significant. Thus, for a burnup calculation, a total bumup of 25 MW-days was assumed. The MCNP BURN option was then utilized to perform a 25 MW-day bumup calculation.

The results of the bumup calculation indicate very minimal bumup. The average bumup of fuel was 0.92% of original fuel loading (3441 grams of 235U), or 31.76 grams of 235U. This is slightly higher than the reported total of 21. 93 grams, but is still far below the technical specification limit of 50% 235U burnup (1720.5 grams of 235U).

7.

Current Core Configuration After performing the bumup calculation, the burned-up fuel isotopics were parsed from the bumup outputs and re-inserted into the MCNP deck. The previous safety analyses were performed again in the current core configuration.

Core Excess, Control Rod Worth and Shutdown Margin The five configurations were performed again and the results are shown in Table 6.

Table 6-Current Rod Worth Calculations Case MCNP k-effective Standard Deviation Reactivity Error (2-sigma)

All Rods In 0.95703 0.00019

-$8.38

$0.05 Shim I fully out 0.98201 0.00019

-$4.59

$0.05 Shim II fully out 0.98248 0.00020

-$4.52

$0.06 Reg fully out 0.97544 0.00020

-$5.57

$0.06 All Rods Out 1.01637 0.00017

$0.33

$0.05 MUTR Neutronic Analysis 15 July 2017

These calculations show a core excess of $0.33 +/- $0.05. This would appear to show agreement with current reactor status, as the MUTR is currently unable to reach full licensed power of 250 kW.

Thirty-three cents of core excess would not be sufficient to overcome the negative temperature coefficient at the point of adding heat.

Individual rod worths are simply the absolute value of "all rods in" minus the absolute value of the worth of the individual rod. Thus the control rod worths of the rods are Shim I: $3.80 +/- $0.05, Shim II: $3.87 +/- $0.06, Reg: $2.82 +/- $0.06 and the total rod worth of $10.48 +/- $0.10. These values appear to be greater than values calculated during rod calibrations. The last available rod calibration data yielded rod worths of$1.81 for Shim I, $2.58 for Shim II and $2.54 for Reg, for a total rod worth of $6.93.

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 and with the most reactive rod in its most reactive position." The most reactive rod is the Shim II rod. Total rod worth minus the Shim II rod is $6.61 +/- $0.10. NRC shutdown margin is this value minus the core excess, which would be $6.28 +/- $0.10, which is far above the technical specification limit of $0.50.

Prompt Fuel Temperature Coefficient The prompt-temperature coefficient calculation results are shown in Figure 8 and tabulated in Table 7 with the GA results added to show similarity [7]. The prompt-temperature coefficient is observed to be negative for all evaluated temperature ranges with decreasing magnitude as

  • temperature increases. The coefficient once again has a value of -l.6¢/°C at 446.8 K, which is similar to the value of -1.2¢/°C stated in the original SAR.

MUTR Neutronic Analysis 16 July 2017

-$0.020

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f

~

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+ MUTR

  • GA A

200 700 1200 1700 2200 Temperature (Kelvin)

Figure 8 - Current Prompt Temperature Coefficient, O.F, as a Function of Temperature Table 7 - Current Prompt Temperature Coefficient Fuel Temperature [K]

Prompt Temperature Coefficient ($/

0 C]

446.8

-$0.01651 750

-$0.01776 1050

-$0.01595 1850

-$0.00086 MUTR Neutronic Analysis 17 July 20 17

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 100% voiding at 10% intervals. The void coefficient was negative for every interval and steadily decreased, as can be seen in Figure 9.

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20 40 60 80 Percent Void Figure 9 - Current Moderator Void Coefficient Moderator Temperature Coefficient 100 The moderator temperature coefficient of reactivity, aM, was determined by varying the moderator density with respect to temperature within the MCNP model MUTR core from the expected operating temperature range of 20°C to 50°C (using Engineering Toolbox [8] to determine water density). The results are shown in Figure 10. The moderator temperature coefficient is calculated to slightly increase from 25°C to 30 °C and from 40 *c to 45 *c, but these changes are on the order of $0.01/"C and all points (with 2-sigma error) are bounded around zero.

The moderator temperature coefficient appears to be negligible.

MUTR Neutronic Analysis 18 July20l7

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0 25 30 35 40 45 Moderator Temperature ('C)

Figure 10 - Current Moderator Temperature Coefficient Core Power Distribution so The power-per-element calculations were once again performed and the values are seen in Table

8.

Table 8 - Current Core Power Distribution 8

7 6

5 4

3 7390 7391 7378 7379 7354 7355 7395 7393 7168 7169 7333 7335 1.30 1.65 2.03 2.35 2.56 2.70 2.70 2.58 2.35 2.06 1.67 1.36 F I 7389 7392 7377 7380 7353 7356 7397 7396 7167 7166 7334 7336 1.48 1.89 2.4 1 3.07 3.09 3.20 3.19 3.09 3.07 2.45 1.93 1.55 7161 7026 7398 259 7368 7365 7374 7375 304 7406 7342 7343 1.68 2.16 2.97 0.00 3.74 3.66 3.66 3.71 0.00 2.97 2.20 1.69 E I 7028 7027 7399 7400 7367 7366 7376 7404 7405 7341 7344 1.78 2.29 2.95 3.65 3.74 3.85 3.67 3.58 2.89 2.29 1.75 7408 7409 7345 7346 7382 7383 7371 7372 7290 7330 71 64 7165 1.76 2.27 2.89 3.33 3.63 3.77 3.73 3.51 3.16 2.73 2.21 1.70 D I 7407 7160 7348 7347 7381 7384 7370 7369 7332 733 l 7163 7162 1.60 2.09 2.63 3.02 3.33 3.64 3.36 3.13 2.84 2.47 2.00 1.55 7360 7357 7352 7349 7401 260 7388 7385 7362 7363 1.31 1.71 2.16 2.48 2.88 0.00 2.89 2.60 1.75 1.32 C I

.-,"'lern 7358 7351 7350 7403 7402 7387 7386 7361 7364 1.57 1.42 7339 7340 0.82 0.94 MUTR Neutronic Analysis 19 July 2017

After 25 MW-days ofburnup, the power distribution does not significantly change. The maximum power per element slightly shifts to the southwest comer of bundle ES ( one element east of the previous hot channel).

Control Rod Calibration Curves University of Maryland provided control rod calibration information (critical and super-critical rod heights) from their 2015 calibrations. Due to the low core excess in the current core, MUTR is unable to calibrate the full length of the control rod and need to extrapolate the available data in order to determine the full control rod worths. Thus, only the given critical/super-critical rod heights were used to perform a series ofMCNP criticality calculations. Figure 11, Figure 12, and Figure 13 show the comparison between the experimental rod calibrations and the MCNP prediction.

Shim I Worth Curve

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$1.20

$1.00 in'

$0.80 0

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$0.60

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$0.20

$0.00 50 60 70 80 90 100 Percent Withdrawn

....._ Experimental MCNP Figure 11 - Shim I Rod Calibration Curves (Experimental vs. MCNP Prediction)

MUTR Neutronic Analysis 20 July 2017

Shim II Worth Curve

$0.80

$0.70

$0.60 ii,

$0.50 t1I 0

$0.40

~

$0.30

  • s; u

$0.20 t1I QI c:::

$0.10

$0.00 65 70 75 80 85 90 95 100 Percent Withdrawn

-.- Experimental MCNP Figure 12 - Shim II Rod Calibration Curves (Experimental vs. MCNP Prediction)

Regulating Rod Worth Curve

$0.90

$0.80 ii, $0.70

$0.60 t1I 0

$0.50

~

$0.40

  • s;

+'

$0.30 u

t1I QI

$0.20 c:::

$0.10

$0.00 60 65 70 75 80 85 90 95 100 Percent Withdrawn

-+- Experimental MCNP Figure 13 - Regulating Rod Calibration Curves (Experimental vs. MCNP Prediction)

There appears to be a disagreement between the MCNP compared to the experimental control rod data, but there may be some uncertainty associated with the experimental control rod data. Figure 14 shows historic MUTR rod worth data compiled over three decades.

MUTR Neutronic Analysis 21 July 20 17

$3.30

$3.10

$2.90

$2.70

..c I... $2.50 f--

0 I

"'O $2.30 0

er::

$2.10

$1.90

$1.70

$1.50 1980 Shim 1 Shim 2

-.- Regulating 1990 2000 Year 2010 Figure 14 - Historic MUTR Control Rod Worth Data 2020 Control rod worth is not expected to vary much between years, especially when there is low fuel burnup and no core reconfiguration. The large changes in rod worth over time are as yet unexplained but likely contribute to the difference observed between the calculated and measure rod worths.

8.

Suggested New Core Configuration The University of Maryland received 19 lightly-irradiated standard TRIG A fuel elements from Idaho National Laboratory (INL) during the spring of 2017. This fuel was originally irradiated at the University of Wisconsin in the 1970s and had been in storage at INL since then. This section will suggest a new core configuration that utilizes this fuel to improve reactor efficiency while maintaining proper safety margins. Idaho National Laboratory provided burnup information from Wisconsin and this data was incorporated into the MCNP decks.

Table 9 shows the suggested new core configuration. For this analysis, it is suggested that sixteen fuel elements be added to the core to increase the core inventory to 109 total fuel elements. The MUTR Neutronic Analysis 22 July 201 7

green highlighted clusters are the new fuel additions. Column 2 (not shown) maintains graphite reflectors in locations E2 and D2.

Table 9 - Suggested New Core Configuration (109 Element Core) 8 7

6 5

4 3

7390 7391 7378 7379 7354 7355 7395 7393 7168 7169 7333 7335 F

7389 7392 7377 7380 7353 7356 7397 7396 7167 7166 7334 7336 7161 7026 7398 259 7368 7365 7374 7375 304 7406 7342 7343 E

7028 7027 7399 7400 7367 7366 7373 7376 7404 7405 7341 7344 7408 7409 7345 7346 7382 7383 7371 7372 7290 7330 7164 7165 D

7407 7160 7348 7347 7381 7384 7370 7369 7332 7331 7163 7162 7360 7357 7352 7349 7401 260 7388 7385 7362 7363 C

Rabbit 7359 7358 7351 7350 7403 7402 7387 7386 7361 7364 6286 6284 5861 6281 6287 6289 7338 7337 6282 6277 B

'(

PuBe Source 6283 6285 5862 5864 6279 6290 7339 7340 6288 6268 Effective Delayed Neutron Fraction Once again using the "totnu" card and running two identical cases, the effective delayed neutron fraction PeffWas calculated to be 0.007244 +/- 0.000160. There is a slight increase in Peffcompared to the beginning-of-life, but 0.007 will continue to be used to express all dollar values of reactivities in this report.

Core Excess, Control Rod Worth, and Shutdown Margin The same five MCNP rod worth calculations were performed again for the new core configuration:

(1) All control rods in, (2) Shim I rod out, Shim II and Reg rods in, (3) Shim II rod out, Shim I and Reg rods in, (4) Reg rod out, Shim I and Shim II rods in, and (5) All control rods out. 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 five calculations are shown in Table 10.

Table 10 - New Core Configuration Rod Worth Calculations Case MCNP k-effective Standard Deviation Reactivity Error (2-sigma)

All Rods In 0.97872 0.00020

-$5.08

$0.06 Shim I fully out 0.99926 0.00020

-$2.08

$0.06 MUTR Neutronic Analysis 23 July 2017

Shim II fully out 0.99992 0.00016

-$1.98

$0.05 Reg fully out 1.00106 0.00021

-$1.82

$0.06 All Rods Out 1.03513 0.00018

$2.88

$0.05 These calculations show a core excess of $2.88 +/- $0.05. This is below the original technical specification limit of $3.50 but higher than the current technical specification limit of $1.12.

Individual rod worths are simply the absolute value of "all rods in" minus the worth of the individual rod. Thus the control rod worths of the rods are Shim I: $3.00 +/- $0.06, Shim II: $3.09

+/- $0.05, Reg: $3.26 +/- $0.06 and the total rod worth of $9.35 +/- $0.09.

Now the most reactive rod is the Regulating Rod, due to having more fuel near the through tube, making the Regulating Rod more valuable. Total rod worth minus the Regulating Rod is $6.09 +/-

$0.09. NRC shutdown margin is this value minus the core excess, which would be $3.22 +/- $0.10, which is still far above the technical specification limit of $0.50.

Prompt Fuel Temperature Coefficient The results of the new core configuration prompt fuel temperature coefficient calculations are shown in Figure 15 and tabulated in Table 11.

-$0.020 C:

-$0.018 111

  • u
E

-$0.016 111 0

-$0.014 u

+ MUTR

  • GA T

111 E _ -so.012 Ill ~

lii lii -$0.010 c.. c..

E ~ -$0.008 111-

~

-$0.006

~

-$0.004 c..

E

-$0.002 0

C:

$0.000 200 700 1200 1700 2200 Temperature (Kelvin)

Figure 15 -New Core Prompt Temperature Coefficient, <lF, as a Function of Temperature MUTR Neutronic Analysis 24 July 2017

Table 11 - New Core Prompt Temperature Coefficient Fuel Temperature [K]

Prompt Temperature Coefficient [$/

0 C]

446.8

-$0.01605 750

-$0.01739 1050

-$0.01 504 1850

-$0.00095 These values are similar to the original beginning-of-life coefficients.

Moderator Void Coefficient Figure 16 shows the moderator void coefficient in the suggested new core configuration.

$0.00

-$0.20

~

-$0.40

  • u
E

-$0.60 QI =ti' 8 'ci -$0.80

-c >

  • o '*' -$1.00 o [ -$1.20

... ~

~ -

-$1.40

-c

~

-$1.60

-$1.80

-$2.00 0

20

.A.

40 60 80 100 Percent Void Figure 16 - New Core Configuration Moderator Void Coefficient The void coefficient was negative for every interval and steadily decreased, similar to the beginning-of-life configuration.

The void coefficient is slightly less negative in this core configuration but is still negative overall.

Moderator Temperature Coefficient Figure 17 shows the power-per-element (in kW) in the suggested new core configuration.

MUTR Neutronic Analysis 25 July 2017

$0.02 C

C1I *o

$0.01 C1I 8

$0.01 C1I...

~ JJ' $0.00 C1I C1I

a. a.

E 11). -$0.01 Cll-1-

S

-$0.01 n:J...

~

-$0.02 0

~

-$0.02 20 0

~~

~~

j~

25 30 35 40 45 so Moderator Temperature ('C)

Figure 17 - New 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.

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

Table 12 - New Core Power Distribution 8

7 6

5 4

3 7390 7391 7378 7379 7354 7355 7395 7393 7168 7169 7333 7335 1.12 1.40 1.71 1.99 2.19 2.28 2.27 2.16 1.98 1.73 1.41 1.16 F

7389 7392 7377 7380 7353 7356 7397 7396 7167 7166 7334 7336 1.28 1.63 2.08 2.64 2.65 2.73 2.71 2.62 2.58 2.06 1.63 1.32 7161 7026 7398 259 7368 7365 7374 7375 304 7406 7342 7343 1.48 1.90 2.60 0.00 3.24 3.17 3.15 3.18 0.00 2.52 1.86 1.46 E 7028 7027 7399 7400 7367 7366 7373 7376 7404 7405 7341 7344 1.63 2.09 2.67 3.29 3.34 3.44 3.41 3.25 3.16 2.54 2.02 1.55 7408 7409 7345 7346 7382 7383 7371 7372 7290 7330 7164 7165 1.68 2.15 2.71

3. l l 3.36 3.47 3.42 3.22 2.91 2.51 2.03 1.56 D 7407 7160 7348 7347 7381 7370 7369 7332 7331 7163 7162 1.63 2.10 2.62 2.99 3.22 3.23 3.01 2.76 2.42 1.93 1.50 7360 7357 7352 7349 7401 260 7388 7385 7362 7363 1.49 1.90 2.37 2.66 3.01 0.00 3.02 2.71 1.82 1.38 C 7359 7358 7351 7350 7403 7402 7387 7386 7361 7364 1.27 1.63 2.00 2.21 2.37 2.60 2.33 2.30 1.6 l 1.18 6286 6284 5861 6281 6287 6289 7338 7337 6282 6277 i : 1.0 I 1.28 1.55 1.78 1.86 l.76 l.76 1.59 1.2 l 0.93 6283 6285 5862 5864 6279 6290 7339 7340 6288 6268 0.73 0.92 1.13 1.36

- 1.43 1.28 1.19 1.05 0.86 0.68 MUTR eutronic Analysis 26 July20 17

The hottest fuel element in now FE 7384 in the southeast comer oflocation D6. This makes sense as the core geometry is closer to a square, which would better centralize the location of the maximum power production. Also, the hottest power-per-element is 3.52 kW, which is far lower than the previous high from BOL of 3.99 kW, due to a higher fuel loading spreading out the power.

9.

Suggested Changes to Technical Specifications This section contains suggested changes to technical specifications, using the previous sections' analyses, in order to improve MUTR operation and efficiency.

Change to Section 3.1 3.1 Reactor Core Parameters

1. The EXCESS REACTIVITY relative to the REFERENCE CORE CONDITION, with or without experiments in place shall not be greater than $1: 12.

This report shows that the MUTR can be safely operated with a core excess of $2.88, as the shutdown margin would exceed $2.00. It is suggested that this specification be rewritten as such:

1. The EXCESS REACTIVITY relative to the REFERENCE CORE CONDITION, with or without experiments in place shall not be greater than '$3.50.

This would return the core excess specification to the previous technical specification limit of

$3.50. This value allows the MUTR the ability to add new fuel in order to improve operational efficiency.

Change to Section 3.3 3.3 Primary Coolant System Ob;ectives

4. The pool water temperature shall not-exceed 90°C, as measured by thermocouples located in the pool.

While the reactor could safely operate at this temperature, there is potential for tank degradation.

It is suggested that this specification be rewritten as such:

MUTR Neutronic Analysis 27 July 2017

4. The pool water temperature shall not exceed 49°C (120°F), as measured by thermocouples

, located in the pool.

Change to Section 5.3

5. 3 Reactor Core and Fuel
1. The core shall consist of 93 TRIGA fuel elements assembled into 24 fuel bundles, 21 bundles shall contain four fuel elements and 3 bundles shall contain three fuel elements and a CONTROL ROD guide tube
2. The fuel bundles shall be arranged in a rectangular 4 x 6 configuration, with one bundle displaced for the in-core pneumatic experimental system.
3. The reactor shall not be operated at power levels exceeding 250 kW.
4. The reflector shall be a combination of two graphite reflectors.

This report shows that the MUTR can accommodate more fuel than was originally loaded. The addition of fuel can allow the MUTR to return to 250kW operation as well as improve the flux in the rabbit and beam port facilities. It is suggested that this specification be rewritten as such:

1. The core shall consist of TRIG A fuel elements assembled into three or four element fuel bundles.
2. The fuel bundles shall be arranged in a close-packed rectangular 5 x 9 configuration, with bundles displaced for the in-core pneumatic experimental system, PuBe source, neutron detectors, and graphite reflector elements.
3. The reactor shall not be operated at power levels exceeding 250 kW.,
4. The reflector shall be a combination ofgraphite reflectors and water.

This technical specification change allows the MUTR to incorporate all of the fuel received from INL. This analysis shows that the MUTR can safely operate with all newly-received fuel elements added to the current core configuration. Due to the short core life of standard TRI GA fuel elements (approximately 100 MW-days [1]), the core needs to be over-loaded to compensate for reactivity loss due to fuel depletion and poison buildup. A similar over-loading was performed at Texas A&M in the 1960s [10]. When Texas A&M increased maximum power from 100 kW to 1 MW, they increased their fuel element inventory to a 126-element (completely full) core.

MUTR Neutronic Analysis 28 July 2017

A close-packed "5 X 9" configuration allows MUTR the freedom to place fuel and graphite reflectors in any available position within columns 1 through 9 and rows B through F, as long as other limits (such as shutdown margin) are maintained.

Finally, "the reflector shall be a combination of graphite reflectors and water" allows MUTR the freedom to place future reflector elements on the periphery of the core to increase core efficiency as long as other limits ( such as shutdown margin) are maintained.

It is also suggested that the reactor power limit be increased to 300 kW so that a regular operating power of 250 kW can be produced without fear of technical specification violation during full-power operations or power calibrations, but the thermal hydraulic calculations to support this increase are beyond the scope of this report.

10.

Summary MCNP6. l.1 was used to calculate fundamental and operational parameters for the Maryland University Training 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. The results of this study indicate _that the MUTR can be operated safely within the Technical Specification bounding envelope. These results also confirms that the MUTR currently cannot operate at full power of 250 kW.

Further analysis was performed to determine a more efficient core configuration, incorporating 16 lightly-irradiated fuel elements delivered from Idaho National Laboratory. This analysis indicates that the MUTR can safely operate with 16 extra fuel elements installed, though technical specifications will need to be revised to allow this operation. This new core configuration increases the flux in irradiation facilities (beam ports, pneumatic transfer facility) and provides more core excess reactivity, which will prolong operating lifetime.

MUTR Neutronic Analysis 29 July 2017

REFERENCES

[1]

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

[2]

  • J.D. Bess, et. al. "Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel".

INL/EXT-10-19486.

March 2011.

Idaho National Laboratory.

[3]

T. Goorley, "MCNP6.l.l-Beta Release Notes", LA-UR-14-24680 (2014).

[4]

A. Mohamed. "Flux Maps Obtained from Core Geometry Approximations: Monte Carlo Simulations and Benchmark Measurements for a 250 kW TRI GA Reactor." Dissertation.

University of Maryland. 2009.

[5]

"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 TRI GA Reactor (2007).

[6]

L. Snoj, G. Zerovnik, A. Trkov, "Analysis of Cross Section Libraries on Zirconium Benchmarks," Proceedings of the International Conference on Nuclear Criticality (ICNC 2011), Edinburgh, Scotland, September 19-22 (2011).

[7]

GA-7882, Kinetic Behavior ofTRIGA Reactors, General Atomics (1967). *

[8]

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

Link: http://www.engineeringtoolbox.com/water-thermal-properties-d_162.html

[9]

"University of Maryland Annual Operating Report: July 1, 2015-June 30, 2016". Web.

Accessed May 17th, 2017. Link: https://www.nrc.gov/docs/ML1627/ML16279A078.pdf

[10]

Randall, J.D. et. al. "The Improvement in Operating Characteristics Resulting from the Addition of FLIP Fuel to a Standard TRI GA Core." Texas A&M University, College Station, Texas, USA ivIUTR Neutronic Analysis 30 July 2017

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