Analysis of a High Temperature Fission Chamber Experiment for Next Generation. Reactors THESIS

Transcription

1 Analysis of a High Temperature Fission Chamber Experiment for Next Generation Reactors THESIS Presented in Partial Fulfillment of the Requirements for the Degree Master of Science in the Graduate School of The Ohio State University By Neil Rutger Taylor Graduate Program in Nuclear Engineering The Ohio State University 2017 Master's Examination Committee: Dr. Tom Blue, Adviser Dr. Raymond Cao

2 Copyrighted by Neil Rutger Taylor 2017

3 Abstract Fission chambers are a vital instrument for the power monitoring of nuclear reactors, and as the current generation of reactors age, the need for next generation reactors to come online grows. Along with these new reactors, sensors and instruments are required for safe operation. One such instrument is a fission chamber capable of surviving the elevated temperatures and harsh environments that come with the next fleet of reactors. A high temperature fission chamber (HTFC) was designed in collaboration with Oak Ridge National Lab to fill this demand for new technology. This design was modeled using MCNP6 and an experiment was conducted at the Ohio State University Research Reactor to test the measurement capabilities of the design. MCNP6 was utilized to simulate expected fission rates from the HTFC at a variety of power levels and locations around the OSURR. These results were then used to create a reactor test plan utilizing the different locations and power levels to test the operation of the HTFC in both pulse and current mode operation. Along with these simulations, a detailed activation analysis utilizing the ORIGEN module within SCALE 6.1 was performed to determine the expected activity level and dose rates associated with the reactor experiment. A gas sampling system was developed to sample gas flowed through the experimental apparatus to ensure that the HTFC survived the experiment. This sampling system monitored the flow gas for alpha and beta radiation from fission products that would have leaked out of the containment in case of failure. ii

4 Dedication To my parents, Scott and Jan, my brother Alex and my grandparents, Shirley and Ronald. iii

5 Acknowledgments First, I would like to thank my graduate advisor, Dr. Thomas Blue, for his technical expertise, guidance through the graduate school process and all-encompassing support of my work. I would also like to thank my lab mates Dr. Brandon Wilson, Josh Jarrell, Kelly McCary, and Tony Birri. Without their technical knowledge and always present support, none of this work could have happened. I would like to thank Kevin Herminghuysen, Dr. Susan White, and especially Andrew Kauffman of the OSU Research Reactor, for their assistance and guidance of completing this experiment. I would also like to thank Oak Ridge National Lab and more specifically, Dr. N. Dianne Bull Ezell, Dr. Padhraic Mulligan and many other researchers that provided invaluable technical expertise. I would also like to thank the Department of Energy Nuclear Energy Advanced Reactor Technology Program for their support and allowing me to work on this project. Finally, I would like to thank the Ohio State University and the Department of Energy Nuclear Energy University Program for their funding that enabled me to work on this project without distractions. iv

6 Vita June Revere High School May B.S. Engineering Physics The Ohio State University University Fellow The Ohio State University 2017 to present...graduate Fellow (NEUP Funded) The Ohio State University Publications Taylor, N., Wilson, B., McCary, K., Blue, T., Bull-Ezell, D. A Lumped Parameter Model of Heat Flow Through a High Temperature Fission Chamber. Transactions of American Nuclear Society. Vol Washington, DC, USA. (October 2017) (Accepted) Wilson, B., McCary, K., Taylor, N., Blue, T., Cao, R. The Creation of a High Temperature Irradiation Facility in the Ohio State Research Reactor. Transactions of American Nuclear Society. Vol Washington, DC, USA. (October 2017) (Accepted) Fields of Study Major Field: Nuclear Engineering v

7 Table of Contents Abstract... ii Acknowledgments... iv Vita... v Publications... v Fields of Study... v Table of Contents... vi List of Tables... ix List of Figures... xi Chapter 1: Introduction Background Detector Design Reactor Experiment Description The Ohio State University Research Reactor... 4 Chapter 2: MCNP Modeling MCNP OSURR MCNP Model HTFC MCNP Model MCNP Simulations and Results vi

8 Chapter 3: Thermal Modeling Simulink Modeling Simulink Results Chapter 4: Thermal Testing ORNL Mock Dry Tube Experiment Setup ORNL Mock Dry Tube Experiment Results Chapter 5: Reactor Experiment Experimental Overview Proposed Experimental Components Proposed Experimental Plan Experimental Shutdown Criteria Chapter 6: Activation Analysis Neutron Activation ORIGEN 2.2 Activation Analysis SCALE 6.2 ORIGEN Activation Analysis ORIGEN Validation Storage and Cooldown Plan Chapter 7. Summary Conclusions vii

9 7.2 MCNP Modeling Thermal Modeling and Testing Activation Analysis References viii

10 List of Tables Table 1: MCNP Results from Run with Original HTFC Design and Pure U-235 Coatings with room temperature cross-sections Table 2: MCNP Results from Run with Original HTFC Design and Pure U-235 Coatings with room temperature cross-sections with Water Filled Void Box Table 3: MCNP Results from Run with Original HTFC Design and Pure U-235 Coatings with room temperature cross-sections with Air Filled Void Box Table 4: MCNP Results from Run with Original HTFC Design and Pure U-235 Coatings with High Temperature Cross Sections and Air-Filled Void Box Table 5: MCNP Results from Run with Original HTFC Design with Natural Uranium Coatings with High Temperature Cross Section and Air Filled Void Box Table 6:MCNP Results from Neutron Spectrum Simulation with Fission Rate Table 7: MCNP Fuel Card Change Isotope and Atomic Abundancies Table 8: MCNP Results from Fission Chamber Model and Experimental Assembly within 10 Dry Tube Table 9: MCNP Results from HTFC Experimental Assembly Outside Thermal Column with Boral Aluminum Door Closed Table 10: MCNP Results from HTFC Experimental Assembly Outside Thermal Column with Boral Aluminum Door Open Table 11: MCNP Neutron Heating Results from HTFC Experimental Assembly Table 12:MCNP Gamma Heating Rates from HTFC Experimental Assembly Table 13:MCNP Total Heating Rates from HTFC Experimental Assembly ix

11 Table 14: Experimental Failure Scenarios, Reponses and Resolutions Table 15: ORIGEN2.2 Activation Analysis Calculations Table 16: SCALE 6.2 ORIGEN Activation Analysis During Irradiation Using Experiment Power History Table 17: SCALE 6.2 ORIGEN Activation Analysis Post Irradiation Using Experiment Power History Table 18: SCALE 6.2 ORIGEN Activation Analysis with Gamma Ray Dose Constants Applied Table 19: Metal Wire Masses and Irradiation Times for ORIGEN Validation Table 20: Comparison Between SCALE 6.2 ORIGEN and Experimental Rabbit Activation x

12 List of Figures Figure 1: Top Down View of Operating Reactor with Components Labeled... 4 Figure 2: Facility Locations Around the OSURR Core. North is to the right Figure 3: Top Down View of a Cross Section of the MCNP Model of OSURR Core from VISED at Centerline of Core... 9 Figure 4: Top Down View of a Cross Section of 10" Dry Tube and Void Box at Centerline of Core Figure 5: Top Down View of a Cross Section of the Thermal Column at Centerline of Core Figure 6: Top Down View of Original HTFC MCNP Model within 10 Dry Tube at Centerline of Core Figure 7: Cross Sectional View of Original HTFC MCNP Model within 10 Dry Tube through Vertical Midplanes of Core and Fission Chamber Figure 8: Top Down View of First Updated HTFC MCNP Model within 10 Dry Tube at Centerline of Core Figure 9: Cross Sectional View through Axial Midplane of First Updated HTFC MCNP Model within 10 Dry Tube Figure 10: HTFC Experimental Assembly Technical Drawing Figure 11: Top Down View of Fission Chamber MCNP Model and Experimental Assembly within 10 Dry Tube in vertical midplane of HTFC Figure 12: Cross Sectional View of Fission Chamber MCNP Model and Experimental Assembly within 10" Dry Tube through vertical midplanes of core and fission chamber 19 xi

13 Figure 13: Azimuthal Segmentation of Fissile Material Layer with Core Position Shown Figure 14: Vertical Distribution of Fission Rate for Original HTFC Design Figure 15: Azimuthal Distribution of Fission Rate for Original HTFC Design Figure 16: 10 Dry Tube Differential Neutron Flux per MeV Spectrum from SAND-II 29 Figure 17: Differential Neutron Flux Energy per MeV Spectrum from Inside of Electrode Tube Figure 18: Differential Neutron Flux Energy per MeV Spectrum on Outside of Electrode Tube Figure 19: Differential Neutron Flux per MeV Spectrum on Inside of Electrode Tube 2 31 Figure 20: Differential Neutron Flux per MeV Spectrum on Outside of Electrode Tube Figure 21: Neutron Flux of 10 Dry Tube Neutron Flux Spectrum Figure 22: MCNP Model of HTFC Experiment Assembly Outside Thermal Column with Boral-Aluminum Plate in Place Figure 23: MCNP Model of HTFC Experiment Assembly Outside Thermal Column with Boral-Aluminum Plate Removed Open Figure 24: Top Down View of HTFC Experimental Apparatus with Component Labels 40 Figure 25: Inner Thermal Simulink Model Figure 26: Outer Simulink Thermal Model Figure 27: Simulink Block Components and Connections for the Fission Chamber Insulator Block xii

14 Figure 28: Temperature-Time Plot of Inner Model with Conductive Heat Transfer Only Figure 29: Temperature-Time Plot of Outer Model of Conductive Heat Transfer Only.. 48 Figure 30: Temperature-Time Plot of Inner Model with Both Conductive and Radiative Heat Transfer Figure 31: Temperature-Time Plot of Outer Model with Both Conductive and Radiative Heat Transfer Figure 32: Temperature-Time Plot of Inner Model with Neutron and Gamma Heating.. 51 Figure 33: Temperature-Time Plot of Outer Model with Neutron and Gamma Heating. 51 Figure 34: Components of HTFC Temperature Experiment with Labels: 1- Titanium Tube, 2- Furnace and Furnace Insulation, 3- Secondary Containment, and 4-10-inch Aluminum Tube Figure 35:All Thermocouple Temperature Data from Steady State Power and Temperature Experiment Figure 36: Helium Gas Thermocouple Data Figure 37: Titanium Tube Lid Thermocouple Data Figure 38: Furnace Insulation Thermocouple Data Figure 39: 10 Tube Outer Diameter Thermocouple Data Figure 40: Proposed Reactor Power History for HTFC Experiment Figure 41:Proposed Day 3 Power and Temperature Profile for Reactor Experiment Figure 42: Proposed Day 4 Power and Temperature Profile for Reactor Experiment Figure 43: Proposed Path of Movement for the HTFC Experimental Apparatus xiii

15 Figure 44: Proposed Day 5 Power and Temperature Profile for Reactor Experiment Figure 45: Proposed Day 6 Power and Temperature Profile for Reactor Experiment Figure 46: Proposed Day 7 Power and Temperature Profile for Reactor Experiment xiv

16 Chapter 1: Introduction 1.1 Background Fission chambers are a measurement-instrument used in nuclear reactors to monitor the power of the reactor during a wide variety of power ranges. These devices utilize a fissile material coated on cylinders with an electric voltage applied to these cylinders. Fission chambers operate under the same principles as an ionization chamber. In a neutron field, fission will occur within the fissile material, releasing fission fragments, these fragments travel within the fill gas between the electrodes and create ions. These ions can be collected on the electrodes and measured as a series of electrical pulses or as a current as the pulses pile up [7]. These pulses or current can be output and converted into a signal that can be read. By knowing the conversion from fission rate to reactor power and the efficiency of collecting the charge, the signal can be used to calculate the reactor power. Next generation reactors will require their own in-core measurement-instruments that can provide accurate and reliable measurements within the reactor core. These reactors will operate in temperatures ranging well above that of current light water reactors and with harsh corrosive environments, such as molten salts. Oak Ridge National Lab (ORNL) has developed a sensor design that should be capable of withstanding the elevated temperatures that will be present in future reactors. To test this design, ORNL manufactured a test fission chamber that was tested within the Ohio State Research Reactor (OSURR). This document outlines the work that was completed to complete this experiment. 1

17 1.2 Detector Design The high temperature fission chamber design has been the culmination of many years of work and countless designs [4]. The final design consists of three concentric cylinders that have a fissile material coating on the inside of the outermost cylinder, both sides of the middle cylinder and the outside of the innermost cylinder. These cylinders are housed within an electrically insulating tube and a primary containment assembly [4]. The three fissile-coated cylinders have wires welded to the top for the electrical signal output from the detector. 1.3 Reactor Experiment Description Before an experiment can be conducted, detailed simulations need to be completed to test the feasibility of the experiment. To do this, the fission chamber and experimental assembly were modeled within MCNP to determine the neutronic properties associated with the fission chamber. The OSURR MCNP models of both the core and thermal column were improved to allow for accurate simulations of the fission chamber. MCNP was used to determine the expected neutron flux, fission rate and neutron flux spectrum, as well as the spatial dependency of the flux values and the heating rates from both neutrons and gammas. Simulink, a part of MATLAB, was utilized to determine the heat transfer characteristics of the experiment. This first Simulink model explored the transfer of heat from the heating element within the secondary containment to the fission chamber itself, as well as radially outwards to the 10 dry tube. The effects of the neutron and gamma heating terms from the operation of the reactor were included to produce an accurate thermal model of the experiment. 2

18 ORIGEN was chosen to complete a neutron activation analysis of the experiment assembly. ORIGEN is a relatively simple and quick code to run that can provide an accurate calculation of the activity from a neutron irradiation. The other method of determining the activity from the experiment would be to utilize MCNP s Burnup capabilities. The main problems with using this method, is that these runs are extremely computationally intensive. ORIGEN calculations are completed on the order of seconds, while MCNP runs require multiple hours. MCNP s Burnup calculations, regarding fissile materials, also require a power level associated with the fissile material, which is extremely difficult to estimate for thin layers of material on a fission chamber device. The proposed reactor experiment includes irradiation of the HTFC and experimental assembly in both the 10 dry tube and thermal column of the OSURR. The experiment will test the pulse and current modes of operation of the fission chamber and determine the cross over threshold from going from pulse to current modes. The reactor will operate over a power range from 50 Watts to 450 kilowatts. The experiment will have a heater element within the containment, which will provide up to 2500 Watts of heating to the experiment and ensure the elevated temperature is maintained throughout the experiment. To ensure the safety of the experiment, an external radiation monitor will be used with a flowing gas system to sample the fill gas of the containment. This will allow for the detection of fission product release in case of a leakage from the primary containment, which will lead to the shutdown of the experiment. 3

19 1.4 The Ohio State University Research Reactor The Ohio State University Research Reactor (OSURR) is a light water, open pool type reactor that has been operated since March It is fueled by 19.5% enriched U3Si2 fuel that is dispersed in an aluminum matrix. This fuel meat is cladded with aluminum. The reactor has a total thermal power limit of 450 kw [1], including a 10% total setpoint uncertainty. The largest thermal flux at full power of any irradiation facility is 1.4E13 n/cm 2 /sec and occurs in the central irradiation facility near the core horizontal midplane. Figure 1 below shows a top down view of the OSURR during operation. Figure 1: Top Down View of Operating Reactor with Components Labeled 4

20 The reactor core has several experimental facilities associated with it that allow for experiments to be conducted using the OSURR. Within the core, there are three irradiation facilities: the central irradiation facility (CIF), auxiliary irradiation facility (AIF), and peripheral irradiation facility (PIF). Along the perimeter of the core, there is the graphite thermal column, pneumatic transfer facility (rabbit), two beam ports, and two different sized movable dry tubes, the 7 and 10 dry tubes. The CIF, PIF and AIF are dry tubes that run vertically from the pool top into the reactor core. The CIF experiences the greatest neutron flux value of any of the facilities due to its centralized location within the core and reaches a peak flux of 1.4E13 neutrons per cm 2 per second (nv) [1]. The AIF and PIF experience the next highest flux values of the facilities due to their location within the core but outside the center region. The other ex-core facilities experience lower flux values due to their position outside the core, but can fulfill different experimental requirements. The rabbit pneumatic tube system utilizes a vacuum system to insert and extract small samples next to the core for timed irradiations. The beam ports penetrate the north wall of the reactor pool and allow for testing of larger systems that can be positioned easily outside the reactor. Beam port 1 is aligned perpendicularly to the core and beam port 2 is at a 45 angle relative to the core. Beam port 2 is currently being used for the Nuclear Analysis and Radiation Sensor (NARS) Laboratory. The thermal column irradiation facility provides a very thermalized neutron flux at a much lower flux than the other facilities. It is positioned along the west side of the reactor pool. The movable vertical dry tubes can be positioned anywhere within the reactor pool, but are mostly positioned next to the reactor core to 5

21 experience the highest neutron flux. Figure 2 below shows all the facilities except for the movable dry tubes. Figure 2: Facility Locations Around the OSURR Core. North is to the right. 6

22 Chapter 2: MCNP Modeling 2.1 MCNP MCNP6 is a Monte Carlo particle tracking code that can be used to simulate experiments to validate results or model a problem. MCNP has the capability to track and model neutron, electron, photon and heavy ion transport and interactions, as well as many other particle types. The Monte Carlo method tracks individual particles and records their interaction and behavior throughout their lifetime. The fission chamber was modeled in MCNP, like any other geometry, by using a collection of surfaces that define volume boundaries and form cells [3]. Each cell has a specified list of characteristics that include properties like the elemental composition, atom density, mass density and associated cross sections. MCNP has a built in visual editor program called VISED, that can be used to easily verify correct geometries. MCNP then utilizes tallies that record the results of the simulation for quantities such as surface current, surface flux, track length and energy distribution [2]. These tallies can be used to calculate important values for reactor operations simulations as well as detector response simulations. For example, in the case of this project, the surface flux can be used to calculate the theoretical fission rate for the fission chamber. Tallies are normalized to one source particle, which can be used to determine results based on the reactor power and the number of neutrons created per second at that power. Mesh tallies are another feature of MCNP that was utilized in this project [2]. A mesh tally allows for a mesh of specified user input-dimensions to be created and tallied over. These tallies can then be visually plotted using MCNP to generate useful visual aids. 7

23 2.2 OSURR MCNP Model An MCNP model of the OSURR was used to model the response and characteristics of the HTFC. This MCNP model was originally developed by Andrew Kauffman and Ryan Winningham [6]. Modifications to this code were required to add in the experimental apparatus and 10 dry tube. The thermal column portion of the core was originally produced by Ryan Winningham and Larry Baas [6]. Once again, this code was modified to include the HTFC experimental apparatus. A top down view of the reactor core was created using VISED and is shown below in Figure 3. The OSURR contains fuel plates that are comprised of LEU and aluminum. The core is surrounded by a graphite reflector on the South and West sides and water on the North and East sides. The core has 3 control rods and 1 safety rod that are used to control the power level of the reactor. These can be withdrawn or inserted to reach criticality and change the power level of the reactor. 8

24 Figure 3: Top Down View of a Cross Section of the MCNP Model of OSURR Core from VISED at Centerline of Core Figure 4 shows the added 10 Dry Tube positioned on the East side of the core. The 10 Dry Tube is an irradiation facility that has a 9.5 inner diameter that can be used for experiments. The tube sits on the bottom of the reactor pool and the top sticks out above the water so that experiments can be easily loaded. In addition to the tube, a small box, the Void Box, is welded to the side. This box can be evacuated and filled with air to create a faster flux spectrum. 9

25 Figure 4: Top Down View of a Cross Section of 10" Dry Tube and Void Box at Centerline of Core Figure 5 shows the thermal column area of the OSURR, which is on the West side of the reactor. This portion consists of movable graphite stringers, that are used to create a very thermalized neutron spectrum flux spectrum. Outside the thermal column there are two movable doors, a larger concrete door and a thin Boral-Aluminum door. Both can be opened to allow for experiments to be positioned outside the thermal column. This location can reach extremely low neutron flux values, which can be useful to reactor start up instrumentation testing. 10

26 Boral Plate Figure 5: Top Down View of a Cross Section of the Thermal Column at Centerline of Core 2.3 HTFC MCNP Model At first, a very simplified version of the fission chamber was modeled in MCNP. At this time, the chamber consisted of only two coated concentric cylinders. The inner diameter of the larger cylinder and the outer diameter of the inner cylinder were coated in a variety of fissile material layers that were specified for each run. The cylinders were modeled to be approximately six inches long and the largest cylinder to be approximately two inches in diameter. The dimensions and materials are not specified within this 11

27 document as they are export controlled. The model was then placed within 10 Dry tube within the OSURR Core MCNP Model. Figure 6 and Figure 7 below show the first model of the HTFC within the dry tube. Fission Chamber Colors Fill Gas FC Cylinders Figure 6: Top Down View of Original HTFC MCNP Model within 10 Dry Tube at Centerline of Core 12

28 Fission Chamber Colors Fill Gas FC Cylinders Figure 7: Cross Sectional View of Original HTFC MCNP Model within 10 Dry Tube through Vertical Midplanes of Core and Fission Chamber With additional information provided by Oak Ridge National Lab, the model was updated twice to match the prototype that was to be tested at the OSURR. The first model update included adding additional concentric cylinders and lengthening the fission chamber. With the addition of concentric cylinders, the number of layers of fissile material increased as the outermost cylinder was coated on the inside diameter, the middle cylinder was coated on both the inner and outer diameters, and the innermost cylinder was coated on only the outer diameter. The new overall length of the fission chamber was increased to approximately 15 inches in length and the outermost cylinder had the same diameter. This first model is shown below in Figure 8 and Figure 9. 13

29 Air FC Housing FC Cylinders FC Fill Gas Figure 8: Top Down View of First Updated HTFC MCNP Model within 10 Dry Tube at Centerline of Core 14

30 Figure 9: Cross Sectional View through Axial Midplane of First Updated HTFC MCNP Model within 10 Dry Tube The second updated model included a primary containment directly around the fission chamber, a heater, heater insulation, a basket assembly, and secondary containment. The basket assembly is used to help lower the heater, fission chamber and insulation into the secondary containment. The secondary containment is a larger aluminum cylinder with a stainless-steel lid with feedthroughs. This is used as a housing for the entire experiment, and provides an additional layer of protection in case of a leakage of fission fragments from the primary containment. Materials and dimensions of the fission chamber cylinders and surrounding housing were updated, to ensure accuracy 15

31 within the thermal and neutronic modeling. Figure 10 shows a technical drawing of the HTFC assembly provided by Oak Ridge National Lab and Figure 11 and Figure 12 below show the HTFC and experimental assembly MCNP model placed within the 10 Dry Tube. 16

32 Figure 10: HTFC Experimental Assembly Technical Drawing 17

33 Figure 11: Top Down View of Fission Chamber MCNP Model and Experimental Assembly within 10 Dry Tube in vertical midplane of HTFC 18

34 Figure 12: Cross Sectional View of Fission Chamber MCNP Model and Experimental Assembly within 10" Dry Tube through vertical midplanes of core and fission chamber These MCNP models were used to determine neutron flux values, fission rates, and heating rates for the experiment. The following sections describe how these models were used and the results from the simulations. 2.4 MCNP Simulations and Results MCNP was run using the kcode calculations built into the program. These calculations utilize starting source particles within a cell containing fissile material; these neutrons are then absorbed by a fissile material atom, fission and create neutrons. The life histories of all neutrons produced are simulated by the software, and their paths and interactions are noted in the form of tallies. MCNP has the capability to use other particles as the source particles to keep track of, such as electrons, gamma rays or other 19

35 heavy atoms. These tallies are normalized to the total number of particles created, or source particles. This allows for easy conversion from the tallies to real values, by knowing the power level of the OSURR. The operating power level of the reactor can be converted into neutrons per second by using the energy released per fission event and the number of neutrons created per fission event. The first MCNP calculations run were done using the original HTFC model with a layer of pure U-235 as the fissile material coating. This first run did not include the Void Box attached to the 10 Dry Tube as it had not been modeled yet. The coating was chosen to be pure U-235 to establish an upper limit on the fission rate of the chamber. The reactor power was assumed to be at full power, 450kW, for this calculation. The conversion to neutrons per second results in 3.78 ±0.19E16 neutrons per second or number of source particles per second in MCNP terms. To determine the fission rate, the MCNP tally results were additionally multiplied by the volume of the fissile material layer to determine the total fissions per second of that fissile layer. The results from this first MCNP run are summarized below in Table 1. 20

36 MCNP Surface Flux per source particle per area (n/cm 2 /sec per source particle) Surface Flux at 450 kw (n/cm 2 /sec) Reactor Power (kw) 4.50±0.23E+02 Outer Fissile Layer 1.05±0.05E ±0.20E+11 Conversion to Neutrons/sec 3.78±0.18E+16 Inner Fissile Layer 1.06±0.05E ±0.20E+11 MCNP Fission Rate per source particle per volume (fissions per source particle per cm 3 ) 1.74±0.09E ±0.09E-04 Fission Rate at 450 kw (Fissions/sec) 7.13±0.36E ±0.28E+09 Table 1: MCNP Results from Run with Original HTFC Design and Pure U-235 Coatings with room temperature cross-sections These results were an initial baseline for the expected fission rate of the HTFC design. As expected, both cylinders experienced the same flux, and the inner layer of Uranium experienced a lower fission rate due to the difference in volume of the fissile material layers. Most of the uncertainty in the fission rate and neutron flux are from the uncertainty in the power level of the reactor which is assumed to be a flat 5%[1]. The volume of the inner layer is approximately eighty percent the volume of the outer layer, and the fission rates have the same ratio. Next, the Void Box was added to the MCNP model to accurately model the exact reactor experiment. Simulations were run with the Void Box filled with water and air to compare the differences and determine which would result in a higher fission rate. The 21

37 results of both the air and water filled Void Box MCNP runs are shown below in Tables 2 and 3. Water Filled Void Box Results Outer Fissile Layer Inner Fissile Layer MCNP Surface Flux per source 1.11± 0.01E ±0.01E-05 particle per area (n/cm 2 /sec per source particle) Surface Flux at 450 kw (n/cm 2 /sec) 4.19±0.21E ±0.21E+11 MCNP Fission Rate per source particle per volume (fissions per source particle per cm 3 ) Fission Rate at 450 kw (Fissions/sec) 1.84±0.01E ±0.35E ±0.01E ±0.30E+09 Table 2: MCNP Results from Run with Original HTFC Design and Pure U-235 Coatings with room temperature cross-sections with Water Filled Void Box 22

38 Air Filled Void Box Results MCNP Surface Flux per source particle per area (n/cm 2 /sec per source particle) Outer Fissile Layer 3.98±0.02E-05 Inner Fissile Layer 3.97±0.02E-05 Surface Flux at 450 kw (n/cm 2 /sec) 1.50±0.75E ±0.75E+12 MCNP Fission Rate per source particle per volume (fissions per source particle per cm 3 ) Fission Rate at 450 kw (Fissions/sec) 6.33±0.02E ±0.13E ±0.02E ±0.10E+10 Table 3: MCNP Results from Run with Original HTFC Design and Pure U-235 Coatings with room temperature cross-sections with Air Filled Void Box These results show that adding in the Void Box increased the fission rate and neutron flux for the water-filled case. The air-filled case had an even higher fission rate and flux than the water-filled case, because there is less water to moderate the neutron flux between the fission chamber and the core. For the actual experiment, the Void Box will be evacuated and filled with air to increase the fission rate and get a higher signal output from the detector. Following the addition of the Void Box, the cross sections of the fissile material were updated to use an elevated temperature cross section. In the experiment, the fission 23

39 chamber will be brought to an elevated temperature using a furnace, which means the chamber will be significantly above room temperature. MCNP has libraries that have cross section data for certain isotopes at a variety of temperatures. The temperatures library chosen was for 1200 Kelvin, because this temperature was the closest to the temperature used in the experiment. These new libraries were used for pure U-235 and natural Uranium with the Void Box filled with air to evaluate the effects that using high temperature cross sections and switching to natural Uranium would have on the fission rate. The results of these simulations are shown below in Tables 4 and 5. U-235 High Temperature Cross Section (926 C) MCNP Surface Flux per source particle per area (n/cm 2 /sec per source particle) Outer Fissile Layer 3.98±0.02E-05 Inner Fissile Layer 3.99±0.02E-05 Surface Flux at 450 kw (n/cm 2 /sec) 1.50±0.07E ±0.07E+12 MCNP Fission Rate per source particle per volume (fissions per source particle per cm 3 ) 6.15±0.03E ±0.03E-04 Fission Rate at 450 kw (Fissions/sec) 2.52±0.12E ±0.10E+10 Table 4: MCNP Results from Run with Original HTFC Design and Pure U-235 Coatings with High Temperature Cross Sections and Air-Filled Void Box 24

40 Natural Uranium High Temperature Cross Section (926 C) MCNP Surface Flux per source particle per area (n/cm 2 /sec per source particle) Outer Fissile Layer 3.96±0.02E-05 Inner Fissile Layer 3.98±0.02E-05 Surface Flux at 450 kw (n/cm 2 /sec) 1.50±0.07E ±0.07E+12 MCNP Fission Rate per source particle per volume (fissions per source particle per cm 3 ) Fission Rate at 450 kw (Fissions/sec) 4.46±0.02E ±0.09E ±0.02E ±0.07E+08 Table 5: MCNP Results from Run with Original HTFC Design with Natural Uranium Coatings with High Temperature Cross Section and Air Filled Void Box The above results show that the high temperature cross sections for MCNP slightly reduce the total fission rate of the pure U-235 fissile material layer. The much larger change is from pure U-235 to natural Uranium, which is 0.72 weight percent U This decreased the fission rate of the layers by a factor of , a value approximately equal to the weight percent of U-235 in the natural Uranium. This decrease makes sense, as the U-238 in the natural Uranium will not affect the fission rate to any measurable degree. For the remaining runs, the elevated temperature cross sections were used for all available isotopes that are within the heated region of the experiment. 25

41 After completing the cross-section work, the spatial dependency of the fission rate within the chamber was explored. To do this, the fissile material layers were broken up into different cells vertically and azimuthally. A tally was used in each cell to determine the fission rate from that cell, which could then be compared to the result of each other cell. All cells were portioned to be the same volume for each fissile layer. In total, each fissile layer was broken into 80 separate layers, 10 vertical layers and 8 azimuthally, every 45 degrees. Figure 13 below shows the azimuthal sections, numbered and their relative position to the core. CORE Figure 13: Azimuthal Segmentation of Fissile Material Layer with Core Position Shown For these spatial runs, natural Uranium was used as the fissile material coatings on the cylinders. The vertical spatial distribution of fission rate is shown below in Figure 14 by averaging the fission rate of all 8 cells in each vertical layer. The different vertical sections are each 1.5 cm in height and range from positive 7.5 cm above the center to 26

42 Fission Rate (Fissions per sec) Fission Rate (Fissions per sec) negative 7.5 cm below the center. Figure 14 shows the azimuthal distribution of fission rate, averaged over all heights, for each of the 8-45-degree arc azimuthal sections. 2.55E E E E E E E E E E E+06 Vertical Distribution Height Range (cm) Figure 14: Vertical Distribution of Fission Rate for Original HTFC Design 2.65E+06 Azimuthal Distribution 2.55E E E E E E Azimuthal Distribution Figure 15: Azimuthal Distribution of Fission Rate for Original HTFC Design. 27

43 In Figure 14, you can see a slight increase in the fission rate around the center of the fission chamber and a slight decrease as you move farther away from the center. The order of magnitudes of the fission rate are all the same. There is less than a 10% increase in the vertical fission rate distribution from the top to the center of the fission chamber and a slightly lesser decrease in the fission rate from the center to the bottom of the fission chamber. Figure 15 shows about a 10% decrease in the fission rate for the azimuthal sections that are furthest away from the core in comparison to those that are nearest to the core. This trend is expected, as the neutron flux will be lower for these sections, because some neutrons will be absorbed by the fissile material layers on the side closer to the reactor. The axial and azimuthal variations of the fission rate are of the same order of magnitude, and are judged to not be so significant that they will have an appreciable effect on the experiment, in terms of exceeding design limits for the temperature of the fissile material coating At this point, more detailed dimensions and drawings of the HTFC were received and the MCNP model was updated to the first updated model, as described earlier. Included within these updates was a switch to a highly-enriched blend of U-235 (HEU) in the form of U3O8. With this updated model, the neutron flux spectrum of MCNP was examined to determine if MCNP was creating a neutron spectrum similar to the measured OSURR spectrum. This was done by inserting energy bins into the MCNP tallies, that matched the energy bins used in STAYSL to determine the OSURR spectrum. STAYSL and a foil activation technique were used to experimentally determine the OSURR 28

44 Differential Neutron Flux (nv/mev) spectrum. The OSURR staff completed the activation analysis and spectrum unfolding, which involves irradiating multiple wires of different metals in the reactor. These wires then have their gamma spectrum measured using a high purity germanium (HPGe) detector. Programs like STAYSL and SAND can then be used to unfold the neutron spectrum based on the activities from the wires, the duration of the irradiation and an initial guess of the spectrum. Figure 16 shows the differential neutron flux spectrum that was generated using this foil activation technique using SAND-II. 1.00E E E E E E E E E E E E E E E E E+04 OSURR 10" Dry Tube Neutron Flux Spectrum Energy (MeV) Figure 16: 10 Dry Tube Differential Neutron Flux per MeV Spectrum from SAND-II Figure 16 shows that the neutron flux spectrum within the 10 Dry tube. Analysis, that was conducted by the reactor staff, determined that the total neutron flux of the 29

45 1.00E E E E E E E E E E E E E E E E E E E E E E E E E E+01 Differential Flux (nv/mev) spectrum is 1.1E12 nv with a thermal flux of 8.1E11 nv. Using energy bins and the tallies described above, the neutron flux spectrum passing through each layer of fissile material within the 10 Dry Tube was determined using MCNP. The results are shown below in Figure 17, Figure 18, Figure 19, and Figure E E E E E E E E E+04 Neutron Flux (n/cm^2/s) per MeV on Inside of Electrode Tube 3 Energy (MeV) Figure 17: Differential Neutron Flux Energy per MeV Spectrum from Inside of Electrode Tube 3 30

46 1E E E E E E E E E E E E E E E E E E E E E E E E E E+01 Differential Flux (nv/mev) 1.00E E E E E E E E E E E E E E E E E E E E E E E E E E+01 Differential Flux (nv/mev) 1.00E E E E E E E E E+04 Neutron Flux (n/cm^2/s) per MeV on Outside of Electrode Tube 2 Energy (MeV) Figure 18: Differential Neutron Flux Energy per MeV Spectrum on Outside of Electrode Tube E E E E E E E E E+04 Neutron Flux (n/cm^2/s) per MeV on Inside of Electrode Tube 2 Energy (MeV) Figure 19: Differential Neutron Flux per MeV Spectrum on Inside of Electrode Tube 2 31

47 1E E E E E E E E E E E E E E E E E E E E E E E E E E+01 Differential Flux (nv/mev) Neutron Flux (n/cm^2/s) on Outside of Electrode Tube E E E E E E E E E+04 Energy (MeV) Figure 20: Differential Neutron Flux per MeV Spectrum on Outside of Electrode Tube 1 The above figures show that the neutron flux spectrum for each layer is the same, and there is no difference between the different layers. The spectrum also matches the spectrum observed from the activation analysis in shape; however, the MCNP flux has a greater total flux than from activation analysis. The MCNP total flux is 1.46E12 nv, compared to 1.1E12 nv using the spectrum unfolding technique. The MCNP flux is larger by a factor of approximately 1.33 than the unfolded value. Another way to visualize the neutron flux spectrum is to look at the neutron flux spectrum as it is presented in Figure 21, where the area under the curve is proportional to the neutron flux with energies within the energy bounds defining the energy interval. This method of presentation allows for a much more meaningful comparison of the shape of two different flux spectra Figure 21 below shows the flux passing through the outermost fissile layer as calculated using MCNP compared to the unfolded spectrum. 32

48 Neutron Flux[ n/(cm^2 s MeV)] *Energy [MeV]*ln(10) 1.20E+12 OSURR 10 Inch Dry Tube Neutron Flux 1.00E E+11 MCNP Output SAND II 6.00E E E E E E E E E E E+01 Energy (MeV) Figure 21: Neutron Flux of 10 Dry Tube Neutron Flux Spectrum Figure 21 shows that the MCNP model predicts a higher total flux than what was unfolded using SAND-II. Table 6 below shows the MCNP tally results for both neutron flux and fission rate of this run with the corresponding calculated neutron flux value and fission rate. 33

49 Fission Chamber Model MCNP Surface Surface Flux at Flux per Source 450kW Particle (n/cm 2 /sec (n/cm^2/sec) per source particle) MCNP Fission Rate per Source Particle (fissions per source particle per cm 3 ) Fission Rate (Fissions/sec) Fissile Layer ±0.04E ±0.07E ±0.02E ±0.4E+11 Fissile Layer ±0.04E ±0.07E ±0.02E ±0.3E+11 Fissile Layer ±0.04E ±0.07E ±0.02E ±0.2E+11 Fissile Layer ±0.04E ±0.07E ±0.02E ±0.1E+11 Total 2.0±0.1E+12 Table 6:MCNP Results from Neutron Spectrum Simulation with Fission Rate Table 6 shows the total flux value through each layer, the fission rate for each layer, and the total expected fission rate. The fission rate for each layer, is substantially higher than for all the previous runs, due to the volume increase of fissile material, due to the lengthened chamber, as well as the switch to an HEU fissile material layer. At this point the technical drawings of the HTFC and experimental apparatus were received from ORNL and the model was updated to match the Fission Chamber MCNP Model as described in Section 2.3. A change in the MCNP material card for the OSURR fuel was also updated, as a mistake was found in the code by Kevin Herminghuysen. Table 7 below shows the change in atomic percentages of the MCNP fuel card. 34

50 Old OSURR Fuel Card New OSURR Fuel Card Isotope Atomic % Isotope Atomic % U U U U Al Al Si Si Table 7: MCNP Fuel Card Change Isotope and Atomic Abundancies The major change in the composition of the fuel is the switch from mostly aluminum to silicon, within the fuel matrix. The percentage of U-235 in the fuel also decreases by approximately 10%. At this point, final neutronic calculations were done in regards to the HTFC experimental assembly being located within the 10 Dry Tube and Thermal Column locations. The composition of the fissile material layers on the cylinders was changed one final time to a low enriched Uranium oxide (LEU), as was used for the coating of the actual cylinders. The final MCNP run within the 10 Dry Tube results are summarized below in Table 8. 35

51 MCNP Surface Flux per Source Particle (n/cm 2 /sec per source particle) Fission Chamber Model Surface Flux at 450kW (n/cm^2/sec) MCNP Fission Rate per Source Particle (fissions per source particle per cm 3 ) Fission Rate (Fissions/sec) Fissile Layer ±0.03E ±0.06E ±0.04E ±0.09E+11 Fissile Layer ±0.03E ±0.06E ±0.04E ±0.06E+11 Fissile Layer ±0.03E ±0.06E ±0.04E ±0.05E+11 Fissile Layer ±0.03E ±0.06E ±0.04E ±0.03E+10 Total 4.7±0.03E+11 Table 8: MCNP Results from Fission Chamber Model and Experimental Assembly within 10 Dry Tube The above results show a decrease in the neutron flux reaching the fissile material layers and an overall reduction in the fission rate. The reduction in neutron flux can be explained by the additional moderation and absorption, due to the surrounding material and a decrease in fissile material in the OSURR fuel cards. The fission rate is a result of the reduced neutron flux and the switch from HEU to LEU from the previous model. These results are the best simulation of the experiment and are used to predict the response of the fission chamber at various power levels within the 10 Dry Tube. The results are still elevated from those of the experimental results; however, the MCNP model has the flux box touching the edge of the reactor core. In the experiment, the dry tube is positioned such that the flux box is very close but not touching the outside of the core. The lack of this water gap would contribute to a higher flux. Next the HTFC experimental assembly was placed outside the thermal column with the concrete door partially open. The Boral-Aluminum door was left closed and 36

52 open in some runs, to look at the difference in flux in those setups. Figure 22 below is a top-view of the HTFC experimental assembly positioned outside the thermal column with the Boral-Aluminum door closed. Figure 23 is the same setup but with the Boral- Aluminum plate removed. Figure 22: MCNP Model of HTFC Experiment Assembly Outside Thermal Column with Boral-Aluminum Plate in Place 37

53 Figure 23: MCNP Model of HTFC Experiment Assembly Outside Thermal Column with Boral-Aluminum Plate Removed Open With so few neutrons reaching the opening of the thermal column, a very large number of particles were needed to get any counts on the tallies at this location. For this purpose, an ORNL cluster was utilized to simulate over 5 billion particles utilizing 40 cores for multiple days. For these runs, the neutron flux entering the secondary containment was tallied and assumed to be the same neutron flux for the rest of the cylinders. The fission rate within each layer was tallied, but with so few neutrons reaching the layers, the error in the fission rate is too high to reasonably record the output from MCNP. Table 9 below shows the results from the MCNP modeling with the Boral- Aluminum door closed, and Table 10 shows the results from the simulation with the door open. 38

54 Thermal Column with Boral Aluminum Door Closed MCNP Surface Flux per Source Particle (n/cm 2 /sec Surface Flux at 450 kw (n/cm^2/sec) per source particle) Secondary Containment 5.68±0.10E ±0.11E+06 Table 9: MCNP Results from HTFC Experimental Assembly Outside Thermal Column with Boral Aluminum Door Closed Thermal Column with Boral Aluminum Door Open MCNP Surface Flux per Source Particle (n/cm 2 /sec Surface Flux at 450 kw (n/cm^2/sec) per source particle) Secondary Containment 2.61±0.05E ±0.51E+07 Table 10: MCNP Results from HTFC Experimental Assembly Outside Thermal Column with Boral Aluminum Door Open These results are used later to determine the expected count rates from the fission chamber, under the assumption that the fissile material layers experience the same neutron flux as the secondary containment. Both fluxes are substantially lower than in the 10 Dry Tube. Boral is also a great thermal neutron absorber, which explains why the flux is an order of magnitude lower with the Boral plate in place. Lastly, MCNP was used to determine neutron and gamma heating rates in the components within the experimental assembly. F6 tallies were utilized within MCNP that tallies all the energy deposited within a cell by a certain type of particle. This includes heating from collisions, fission events, and any other events that deposit energy within a material. Figure 24 below shows the different material layers that had heating tallies 39

55 added to their cells. Heating tallies were added to the cells of the fissile material layers; however, they are too small to be visibly seen in the diagram. Figure 24: Top Down View of HTFC Experimental Apparatus with Component Labels Some materials that are tallied that are not included in the above diagram are the titanium rods of the basket assembly and the fission chamber top and bottom caps, which seal the fission chamber. The neutron and gamma heating tallies were added to the MCNP model of the HTFC experimental apparatus within the 10 Dry Tube, since this is where the heating rates will be the highest. The heating rates determined by MCNP are presented in MeV per gram per source particle, so once again we multiplied each result by the number of neutrons created per second in the reactor, as well as the total mass in 40

56 grams of each cell. Table 11 shows the neutron heating results, Table 12 shows the gamma heating results, and Table 13 shows the total heating results from MCNP. Component MCNP Result (MeV/gram* neutron) MeV/ second at 450kW J/second (Watts) at 450 kw Primary 0.00E E E+00 Containment Outer Sleeve Tube 5.48±0.05E ±0.06E ±0.10E-01 Electrode Tube E E E+00 Fill Gas Between Electrode Tubes 3 and E E E+00 Electrode Tube E E E+00 Fill Gas Between Electrode Tubes 2 and E E E+00 Electrode Tube E E E+00 Fill Gas within 0.00E E E+00 Electrode Tube 1 Secondary 9.02±0.09E ±0.07E ±0.010E+00 Containment Basket Assembly 0.00E E E+00 Rod Basket Assembly 0.00E E E+00 Rod Basket Assembly 0.00E E E+00 Rod Basket Assembly 0.00E E E+00 Rod FC Bottom Disk 3.57±0.04E ±0.31E ±0.50E-03 Furnace 1.11±0.01E ±0.13E ±0.21E-01 FC Top Disk 3.40±0.03E ±0.30E ±0.47E-03 Fissile Layer ±0.09E ±0.15E ±0.23E+00 Fissile Layer ±0.08E ±0.11E ±0.17E+00 Fissile Layer ±0.09E ±0.09E ±0.15E+00 Fissile Layer ±0.09E ±0.05E ±0.09E+00 Table 11: MCNP Neutron Heating Results from HTFC Experimental Assembly 41

57 Component MCNP Result (MeV/gram per neutron) MeV per second at 450 kw J/second (Watts) at 450 kw Primary 9.74±0.10E ±0.30E ±0.48E+00 Containment Outer Sleeve Tube 1.11±0.01E ±0.13E ±0.20E+00 Electrode Tube ±0.02E ±0.21E ±0.34E+00 Fill Gas Between 1.03±0.01E ±0.06E ±0.09E-03 Electrode Tubes 3 and 2 Electrode Tube ±0.01E ±0.12E ±0.19E+00 Fill Gas Between 1.01±0.01E ±0.32E ±0.05E-03 Electrode Tubes 2 and 1 Electrode Tube ±0.01E ±0.50E ±0.08E+00 Fill Gas within 9.90±0.10E ±0.09E ±0.14E-04 Electrode Tube 1 Secondary 5.70±0.06E ±0.41E ±0.07E+02 Containment Basket Assembly 1.93±0.02E ±0.41E ±0.07E+00 Rod Basket Assembly 1.11±0.01E ±0.24E ±0.38E-01 Rod Basket Assembly 1.13±0.01E ±0.24E ±0.38E-01 Rod Basket Assembly 7.79±0.08E ±0.17E ±0.26E-01 Rod FC Bottom Disk 9.33±0.09E ±0.08E ±0.13E-01 Furnace 1.27±0.01E ±0.15E ±0.24E+01 FC Top Disk 8.55±0.09E ±0.07E ±0.12E-01 Fissile Layer ±0.04E ±0.67E ±0.11E-02 Fissile Layer ±0.04E ±0.49E ±0.08E-02 Fissile Layer ±0.04E ±0.39E ±0.07E-02 Fissile Layer ±0.04E ±0.23E ±0.36E-03 Table 12:MCNP Gamma Heating Rates from HTFC Experimental Assembly 42

58 Component MCNP Result (MeV/gram per neutron) MeV per second at 450 kw J/second (Watts) at 450 kw Primary 9.74±0.10E ±0.30E ±0.48E+00 Containment Outer Sleeve Tube 1.17±0.01E ±0.13E ±0.21E+00 Electrode Tube ±0.02E ±0.21E ±0.34E+00 Fill Gas Between 1.03±0.01E ±0.06E ±0.09E-03 Electrode Tubes 3 and 2 Electrode Tube ±0.01E ±0.12E ±0.19E+00 Fill Gas Between 1.01±0.01E ±0.32E ±0.05E-03 Electrode Tubes 2 and 1 Electrode Tube ±0.01E ±0.05E ±0.08E+00 Fill Gas within 9.90±0.10E ±0.09E ±0.14E-04 Electrode Tube 1 Secondary 5.79±0.06E ±0.41E ±0.07E+02 Containment Basket Assembly 1.93±0.02E ±0.41E ±0.07E+00 Rod Basket Assembly 1.11±0.01E ±0.24E ±0.38E-01 Rod Basket Assembly 1.13±0.01E ±0.24E ±0.38E-01 Rod Basket Assembly 7.79±0.08E ±0.17E ±0.26E-01 Rod FC Bottom Disk 9.69±0.10E ±0.09E ±0.14E-01 Furnace 1.28±0.01E ±0.15E ±0.24E+01 FC Top Disk 8.89±0.09E ±0.08E ±0.22E-01 Fissile Layer ±0.09E ±0.15E ±0.24E+00 Fissile Layer ±0.08E ±0.11E ±0.17E+00 Fissile Layer ±0.09E ±0.09E ±0.15E+00 Fissile Layer ±0.09E ±0.05E ±0.09E+00 Table 13:MCNP Total Heating Rates from HTFC Experimental Assembly The total heating rates are the combination of neutron and gamma heating. The total heating rate of all the components equals 223±11 Watts at 450 kw. This value is a 43

59 little under 10 percent of the total output power of the heater for the experiment. The gamma heating dominates internal heat generation for most of the components, although the neutron heating from the fissile material layers also contributes a significant amount to the total heat generation of the assembly. The gamma heating of the aluminum containment contributes more than 50 percent of the total heating of the experimental assembly. These total values were used in later simulations to determine the temperaturetime behavior of the internal components of the experiment. The values at full reactor power are the only ones of concern, as these are the conditions under which the highest temperatures will be achieved. The heating due to electrons was not explored, as adding these particles into the simulations causes the runs to last so long that it is unreasonable to attempt to quantify this heating. These temperatures will be monitored because to make sure that temperature limits are not exceeded for any components. 44

60 Chapter 3: Thermal Modeling After completing the MCNP modeling of the HTFC, the heat transfer from the furnace heating element through the entire experimental apparatus was studied. Specifically, the time constants for heating each element using the furnace was explored, while also including gamma and neutron heating determined from MCNP. This allowed an accurate thermal model of the experiment to be created and simulated to determine how long the heater element should be turned on to allow all components to reach a steady- state temperature. These results were used to determine how impactful neutron and gamma heating will be for the experiment. To complete this analysis, the Simulink software within MATLAB was used to construct a lumped parameter model to accurately model the heat transfer within the experiment [5]. 3.1 Simulink Modeling Simulink has lumped-parameter thermal modeling capabilities, where one can model thermal masses and conductive, convective and radiative heat transfer between them. It also allows users to supply both temperatures and heat flow sources as boundary conditions to these models. The model is broken up into two separate distinct regions, the inner and outer models. While the inner model includes everything contained within the heating element, the outer model includes all components outside of the heating element. Within these models, masses, thermal capacitances, thermal conductivities, radiative heat transfer coefficients, as well as some dimensional properties, were input to accurately model the FC system. Convective heat transfer was not included within the model as the gas gaps are so small that convective heat transfer very likely will not contribute any appreciable heat 45

61 transfer. A block diagram of the FC inner model is shown below in Figure 25 and a block diagram of the FC outer model is shown in Figure 26. Both have appropriate labels and boundary conditions applied. The water was held at room temperature. All components were set to room temperature as initial conditions. In the inner and outer models, C_# represents conductive heat transfer between two components and R_# represents radiative heat transfer [5]. Each block represents a thermal mass with specific mass and heat capacity. As an example of the details of the input for a block, the components of the FC insulation block are shown in Figure 27. Figure 25: Inner Thermal Simulink Model Figure 26: Outer Simulink Thermal Model 46

62 Figure 27: Simulink Block Components and Connections for the Fission Chamber Insulator Block All models were run with the same heater temperature and initial room temperature. The components of the FC and FC experimental apparatus are comprised of concentric cylinders, where the length of each is much greater than the radius, so axial heat transfer is neglected and only radial heat transfer is assumed to take place. The models assume that a step change in temperature is applied at the location of the heater element at time t=0 +. First, the simulations were run with only the conductive heat transfer elements present. Next the radiative heat transfer terms were included to see their impact on the equilibrium temperature time constants. Lastly, neutron and gamma internal heating heat sources were added to each component. These internal heating values were determined using the MCNP model of the fission chamber. 3.2 Simulink Results The Simulink models were run for long enough time for equilibrium to be established on both sides of the model. Figure 28 shows the inner model response as it 47

63 approaches a steady state temperature, when only conductive heat transfer is considered and Figure 29 shows the same for the outer model. Figure 28: Temperature-Time Plot of Inner Model with Conductive Heat Transfer Only Figure 29: Temperature-Time Plot of Outer Model of Conductive Heat Transfer Only 48

64 These conductive heat transfer models predict an equilibrium temperature will be established within the inner region of the FC in approximately 2 hours and 30 minutes. The outer model takes significantly longer; however, this region is unimportant to the actual experiment. The inner model component temperatures all converge towards the heating element steady state temperature, while the outer model s component temperatures vary in temperature, due to the heat loss to the surrounding water. However, this model is not very realistic in modeling the actual experiment because conduction is not the only heat transfer method that will occur in the experiment. The effects of adding radiative heat transfer into the inner and outer models are shown in Figure 30 and Figure 31. Figure 30: Temperature-Time Plot of Inner Model with Both Conductive and Radiative Heat Transfer 49

65 Figure 31: Temperature-Time Plot of Outer Model with Both Conductive and Radiative Heat Transfer Adding radiative heat transfer to the models, significantly reduces the time it takes for the inner system to reach equilibrium temperature. The innermost cylinders of the fission chamber reach equilibrium temperature in approximately 1 hour and 15 minutes. The outer model also reaches equilibrium temperature at a much faster rate, and reaches lower equilibrium temperatures than the conductive only models. The outer models reach lower temperatures, because the heat lost to the surrounding water is greater with radiative heat transfer included in the model. These models are a much better estimate of the real experiment, because radiative heat transfer provides a significant amount of heating between the metal cylinders. Lastly, neutron and gamma heating were included and are shown in Figure 32and Figure 33. The neutron and gamma heating values used assume that the reactor is operating at 450 kw. 50

66 Figure 32: Temperature-Time Plot of Inner Model with Neutron and Gamma Heating Figure 33: Temperature-Time Plot of Outer Model with Neutron and Gamma Heating The addition of the neutron and gamma heating increases the rate at which equilibrium temperature is established, but very slightly. The bigger takeaway is the 51

67 elevated temperatures at which the components reach equilibrium in the inner model. The innermost FC cylinder establishes equilibrium at nearly 30 C above the temperature of the heating element. These models assume that the heater element instantaneously achieves the selected temperature, when in reality the heater element will also have to increase in temperature as power is supplied to it. However, these models do provide a good baseline estimate for the amount of time required to heat the FC to temperature and will need to be included into the experiment plan. 52

68 Chapter 4: Thermal Testing 4.1 ORNL Mock Dry Tube Experiment Setup A mock dry tube experiment was completed at ORNL to determine the steady state power requirements and temperatures of various components of the HTFC experiment assembly. To do this experiment a six-foot-tall section of 10-inch outer diameter, 9.5-inch inner diameter aluminum tube was used to act as a mock 10-inch dry tube, like the one that will be used in the reactor experiment. The secondary containment, secondary containment lid, furnace and insulation were all used in the experiment and are all the same as those that will be used in the reactor experiment. For the primary containment and primary containment lid, a titanium tube, with the same outer diameter as the primary containment, was used as a stand-in for the actual primary containment. A piece of titanium was spot welded to the top of the tube to represent the primary containment lid. Type K thermocouples were spot welded onto the titanium tube using a spot welder and strips of titanium foil. A total of three thermocouples were spot welded onto the length of the tube. One 2 inches from the bottom of the tube, another one 9 inches from the bottom and the last one 16 inches from the bottom. The middle thermocouple was hooked up to the DC power supply PID controller to be used as the control temperature for the experiment. This same process for attaching thermocouples will be done for the actual primary containment to be used in the reactor experiment. A thermocouple was located approximately one foot below the secondary containment lid to determine the temperature of the Helium gas within the secondary containment. Another thermocouple was located in between the secondary containment and the furnace insulation. Lastly, thermocouples were placed on the outside of the 10-inch aluminum 53

69 tube during the test. The pictures below, Figure 34, show the components of the temperature experiment Figure 34: Components of HTFC Temperature Experiment with Labels: 1- Titanium Tube, 2- Furnace and Furnace Insulation, 3- Secondary Containment, and 4-10-inch Aluminum Tube 54

70 The titanium tube was initially placed within an insulated Nextel sleeve, but was removed for the actual experiment. The titanium tube was placed within the furnace and insulation, which was already set within the titanium basket assembly. The furnace, insulation and tube were then lowered into the secondary containment. The appropriate thermocouple and power supply connections were made, and then the secondary containment lid was put into place and sealed. The lid utilizes an O-ring to create a leak proof seal and is fastened into place with 16 screws. The gas line was then attached to pump down to vacuum and backfill with Helium gas. A pressure relief valve was also attached to the secondary containment lid to ensure the apparatus does not become over pressurized during the test. Once the containment was sufficiently filled with Helium, the test could proceed. 4.2 ORNL Mock Dry Tube Experiment Results To start the test, the data logger was turned on and set to begin recording data. The DC power supply was then turned on and the current limit was set to 20 Amps. The PID temperature controller was then set to 800 C to begin heating the assembly. The DC power supply, output approximately 2300 Watts and 13 Amps as the titanium tube was brought to temperature. It took approximately 1 hour for the center of the titanium tube to reach a steady state temperature of 800 C. The DC power supply output between 600 and 800 Watts to maintain the steady state temperature. The rest of the components continued to increase in temperature over time, before reaching a steady state. The test was continued for another 3 hours, for a total run of approximately 4 hours. The graphs below, Figures 35 through 39, show the temperatures measured during the experiment. Each vertical dashed lined corresponds to 5 hours passing since the start of the 55

71 experiment. 13:06:03 18:06:03 23:06:03 04:06:03 Figure 35:All Thermocouple Temperature Data from Steady State Power and Temperature Experiment 13:06:03 18:06:03 23:06:03 04:06:03 Figure 36: Helium Gas Thermocouple Data 13:06:03 18:06:03 23:06:03 04:06:03 Figure 37: Titanium Tube Lid Thermocouple Data 56

72 13:06:03 18:06:03 23:06:03 04:06:03 Figure 38: Furnace Insulation Thermocouple Data 13:06:03 18:06:03 23:06:03 04:06:03 Figure 39: 10 Tube Outer Diameter Thermocouple Data The bottom of the titanium tube initially was below 800 C, when the middle achieved steady state, but continued to increase to slightly above 800 C. The jump in temperature at the very end, could be the result of restarting the power supply after temporarily turning it off, to make a noise measurement. The Helium gas temperature peaked a little below 120 C and the titanium tube lid also peaked around 110 C. The Helium gas and titanium tube lid temperature profiles follow the same shape for the entire duration of the experiment. The thermocouple located between the furnace insulation and secondary containment in Figure 38 achieved steady state; however, 57

73 something occurred and the thermocouple temperature had a huge spike and then continued to increase. It is believed the thermocouple shifted from the secondary containment to the furnace insulation, this would explain the huge jump in temperature. This thermocouple temperature peaked around 225 C, and even this temperature is well below the temperature limit for the secondary containment. The thermocouple attached to the outside of the 10-inch tube was still slowly increasing in temperature, but was around 70 C, when the power supply was turned off. In this experiment the tube was in open air, while in the reactor experiment there will be cooled water flowing around the tube keeping it cooled. Temperature data was recorded overnight as the assembly cooled, and 18 hours after turning off the power supply, all components were essentially back to room temperature. In comparison to the Simulink modeling, the outermost cylinder achieved steady state in roughly the same time as the time predicted by the conductive and radiative heat transfer model. This experiment did not include any concentric cylinders within the titanium tube, nor will thermocouples be attached to the inner cylinders for the actual experiment. For the actual experiment, it will be assumed that when the outer primary containment reaches an equilibrium temperature, the cylinders within will also have reached a steady state temperature. The temperatures of the outer components such as the insulation, helium gas and 10 tube were measured to be higher than that of the simulated values. The fact that this experiment took place using only air cooling, rather than water kept at a specific temperature, could be a key factor in explaining this discrepancy. 58

74 Chapter 5: Reactor Experiment 5.1 Experimental Overview The proposed reactor experiment at the OSURR is a weeklong experiment that will span a range of powers going from the lowest reliable power of 50 Watts to the full power of the reactor, 450 kilowatts. The purpose of this experiment is to test the functionality of a prototype HTFC within the OSURR and determine the transition range from pulse to current mode operation. The experiment will utilize two of the irradiation facilities available at the OSURR, the thermal column irradiation facility and 10 dry tube. 5.2 Proposed Experimental Components The reactor experiment requires many different systems and instruments to ensure that the experiment operates as expected. The experimental assembly itself includes the aluminum secondary containment, secondary containment lid with appropriate feedthroughs, titanium basket assembly, heating element and associated insulation, primary containment with fission chamber, gas sampling system tubing, icam, thermocouples, thermocouple data logger, DC power supply, and front-end electronics for receiving the output of the fission chamber. Tanks of high purity helium gas will also be used to constantly flow helium throughout the experiment and sample for fission products. The secondary containment lid contains feed-throughs that enable cabling and tubing to reach the fission chamber and apparatus. Two gas penetrations are included that supply both the inlet and outlet for the gas sampling system. A third and final penetration is used for the mineral insulated (MI) cable that feeds the front-end electronics and outputs the signal from the fission chamber. Five thermocouple feed-throughs were 59

75 attached, so that the DC power supply can have an input for controlling the heating element, as well as collect temperature measurements of various components during testing. Lastly, the lid has two electrical feedthroughs for the input and output of the power supply for the heating element. The heating element is a custom ordered Thermcraft VF V-S heating element. It has a 4 diameter and 18 long heated region that is surrounded by a vitreous aluminosilicate fibrous insulation. The insulation is comprised of 42% alumina, 56% silica and 2% other material. The heating element itself utilizes Kanthal A-1 Resistance Wire, which is a FeCrAl alloy [8]. It has a maximum voltage input of 240V, a maximum temperature of 1,100 C and a recommended maximum power in an oxygen environment of 1800 watts. The heating element can be operated at a higher power, because of the inert atmosphere in which it will be operating. To power the heating element, a DC power supply was chosen to supply current to the resistance wire. A DC supply was chosen over an AC supply, to minimize the noise output from the power supply. Any noise within the experiment could affect the signal output from the detector. A Veeco 2500W power supply was chosen, as it can supply up to 2,500 watts of power. As an input, it takes volts AC and outputs volts DC and 0-25 amps DC. As mentioned earlier, it has a Eurotherm 2408 PID controller, that allows for the controlling of the power supply. The PID controller allows for a user to set a temperature setpoint that the power supply will achieve. The power supply itself has limits on the power and current outputs. 60

76 To check that the fission chamber does not rupture or leak fission products, a gas sampling system was devised to sample the gas within the secondary containment for alpha and beta radiation. To detect these particles, the Canberra icam (continuous air monitor) was attached to the secondary containment using a system of tubes and check valves. The icam measures the flow rate of gas passing by the detector and counts the number of alpha and beta particles it detects. The system has alarms for both types of radiation and will alert the experimenters of a fission product release. The gas system begins with a tank of helium, that flows gas through a regulator and then tubes that lead down to the secondary containment. Two check valves are in place along the length of the tube, one located near the gas tank and another near the feedthrough of the secondary containment lid. These are in place in case of a fission product release, in which case the valves can be shut and trap the fission products within the secondary containment. The gas will flow into the secondary containment from the inlet feedthrough and will flow out the outlet feedthrough. The tubing attached to the outlet feedthrough has the same double check valve system for gas isolation. The tubing then feeds into the icam, where it is sampled for radiation and then dispersed into the atmosphere. 5.3 Proposed Experimental Plan Oak Ridge will ship the experimental components to the OSURR the week leading up to the experiment. The fission chamber sealed within the primary containment will be shipped to Office of Radiation Safety for checks, before being shipped to the reactor. The rest of the components including the secondary containment, basket assembly, DC power supply, heater element, insulation and thermocouple data logger will all be shipped directly to the reactor lab. The equipment associated with the 61

77 electronic and signal processing side of the fission chamber will be brought in person to the reactor lab. The experiment will begin with a two-day setup period October 19 th and 20 th, Day 1 and Day 2. On these days, the experimental apparatus will be assembled and tested for the appropriate connections. First, the thermocouples will be positioned and attached to the primary containment, heater insulation, and secondary containment inner diameter. Next, the heater and associated insulation were placed within the titanium basket assembly. The fission chamber will then be placed within the heater and electrical connections will be made to the fission chamber, heater element and associated feedthroughs on the secondary containment lid. The electrical connections will then be tested to ensure that they were properly made. Next, the basket assembly will be lowered into the secondary containment and the lid will be secured onto the containment. At this point, the electrical and thermocouple connections will be tested once again to ensure that the signal outputs are still working correctly. Attaching the gas sampling system to both the secondary containment lid and the icam will happen next. The system will be tested to ensure that the flow rate and system worked properly. The actual experiment will take place over the week of October 23 rd to 27 th, Days 3 through 7. Each day of reactor operation will begin with the reactor staff conducting their startup routines at 7:00 AM, then bringing the reactor up to the starting power at 8:00 AM. The heater will be started around 7:00 AM each day to allow the fission chamber to reach the elevated temperature. The proposed reactor power and temperature profile is shown below in Figure

78 Figure 40: Proposed Reactor Power History for HTFC Experiment The proposed temperature and power profiles are idealized, as they do not include the power transients that will occur between power level. The temperature will also vary from this plan as it will not instantly heat up, and the neutron and gamma heating will increase the actual temperature. For the first days of testing, Day 3 and 4, the HTFC experiment assembly will be positioned outside the thermal column as described in the MCNP modeling section. After Day 4, the apparatus will be moved to the 10 Dry Tube where it will remain for the rest of the experiment. A more detailed look at Day 3 is shown below in Figure

79 Figure 41:Proposed Day 3 Power and Temperature Profile for Reactor Experiment This day of testing was designed to test the pulse mode operation of the fission chamber instrument. The combination of a lower reactor power and the thermal column location, results in an extremely low flux value. This low flux value corresponds to a low fission rate, which leads to a low count rate. This low count rate is required to test the pulse mode operation of the fission chamber. The power starts at the lowest reliable power level of the reactor, 50 Watts. The power is increased five times over the 6 hours of the day and ends at a power of 550 Watts. The temperature of the heater is set to 800 C for the entire duration of the testing. The next day of proposed testing is shown in more detail below in Figure

80 Figure 42: Proposed Day 4 Power and Temperature Profile for Reactor Experiment This day of testing was proposed to test the transition period from pulse mode to current mode of the detector. The experimental assembly will still be positioned outside the thermal column for this day of testing. The reactor starts off at a low power level of 500 Watts, has 5 power increases and ends at 3.0 kilowatts. Over this power range, the neutron flux and count rate should increase through the threshold at which the detector should transition to current mode operation. After this day of testing, the experimental apparatus needs to be moved to the 10 dry tube location. This can either be done at the end of day 4 or the beginning of day 5. Figure 43 below shows the proposed path of movement for the experimental apparatus. The apparatus will be moved using ropes attached to the secondary containment lid and the crane of the reactor lab. An individual 65

81 will walk with the apparatus as it is moved to ensure that it does not collide with anything. Figure 43: Proposed Path of Movement for the HTFC Experimental Apparatus The crane will be used to lower the experimental apparatus to the bottom of the 10 dry tube and then will be used to move the dry tube next to the reactor core for testing on Days 5, 6, and 7. The proposed power history for the first day of testing within the 10 dry tube location is shown below in Figure 44 66