Initial fuel possibilities for the Thorium Molten Salt Reactor

Transcription

1 Sholars' Mine Masters Theses Student Researh & Creative Works Spring 2015 Initial fuel possibilities for the Thorium Molten Salt Reator Dustin Gage Green Follow this and additional works at: Part of the Nulear Engineering Commons Department: Reommended Citation Green, Dustin Gage, "Initial fuel possibilities for the Thorium Molten Salt Reator" (2015). Masters Theses This Thesis - Open Aess is brought to you for free and open aess by Sholars' Mine. It has been aepted for inlusion in Masters Theses by an authorized administrator of Sholars' Mine. This work is proteted by U. S. Copyright Law. Unauthorized use inluding reprodution for redistribution requires the permission of the opyright holder. For more information, please ontat sholarsmine@mst.edu.

2 INITIAL FUEL POSSIBILITIES FOR THE THORIUM MOLTEN SALT REACTOR by DUSTIN GAGE GREEN A THESIS Presented to the Faulty of the Graduate Shool of the MISSOURI UNIVERSITY OF SCIENCE AND TECHNOLOGY In Partial Fulfillment of the Requirements for the Degree MASTER OF SCIENCE IN NUCLEAR ENGINEERING 2015 Approved by Ayodeji Alajo, Advisor Xin Liu Gary Mueller

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4 iii ABSTRACT The Generation IV International Forum plaed six reators as priority for researh and development to ompensate for the world s inreasing energy demands. Among the six were Molten Salt Reators (MSRs). These reators utilize the Th/ 233 U fuel yle using molten fluoride or hloride salts as oolants. MSRs also have the possibility to use other fissile fuels espeially with the first fleet of reators given the low amount of Uranium- 233 available ommerially. With the possibility of diverting from using 233 U initially, the researh presented here will benhmark 233 U as a main fuel for MSRs using the Thorium Molten Salt Reator (TMSR) designed by Billebaud Annik and her team at the Laboratory for Subatomi Partiles and Cosmology in Frane. Uranium-235 and Plutonium-239 will then be tested as suitable initial fuels for the first fleet of TMSRs. These fuels will be ompared to the referene data based on neutron flux spetra and breeding apabilities.

5 iv ACKNOWLEDGMENTS A big thanks to Dr. Alajo and the Nulear Engineering Department at Missouri S&T. Thank you to the Nulear Regulatory Commission for funding the researh. I would like to also thank Dr. Mueller and Dr. Liu for aepting the positions to be on my ommittee. Also, a huge hunk of gratitude goes to my lovely wife for putting up with me and my shooling for three extra semesters.

6 v TABLE OF CONTENTS ABSTRACT... iii ACKNOWLEDGMENTS... iv LIST OF ILLUSTRATIONS... vi LIST OF TABLES... vii SECTION 1. INTRODUCTION THORIUM MOLTEN SALT REACTOR CONCEPT GEOMETRY SALT PROPERTIES MCNP MCNP MODEL, CORE SPECIFICATIONS AND ASSUMPTIONS URANIUM URANIUM PLUTONIUM RESULTS URANIUM URANIUM PLUTONIUM DISCUSSION CONCLUSION APENDICES A. MCNP URANIUM 233 INPUT CODE B. MCNP URANIUM 235 INPUT CODE C. MCNP PLUTONIUM 239 INPUT CODE BIBLIOGRAPHY VITA... 48

7 vi Figure LIST OF ILLUSTRATIONS Page 1.1. Simplified Molten Salt Reator shemati Conept design with basi omponents shown of the TMSR LiF-ThF4 phase diagram Basi overview of the hemial proessing system used to purify the fuel and blanket salt Vertial, to-sale ross-ut of the TMSR Vertial ross-ut of the simulated TMSR in MCNP Breeding ratio of the referene TMSR as a funtion of the reproessing rate Exess prodution of 233U as a funtion of time Neutron spetra for a hloride salt (purple) and a fluoride salt (green) Neutron flux spetrum of the 233 U-operated TMSR shown with the ross setion values of Flourine Estimation of the doubling time for the 233 U-operated TMSR Neutron flux spetrum of the 235 U-operated TMSR Estimation of the doubling time for the 235 U-operated TMSR Neutron flux spetrum of the 239 Pu-operated TMSR Estimation of the doubling time for the 239 Pu-operated TMSR... 22

8 vii Table LIST OF TABLES Page 2.1. Composition (at%) of nikel hastelloy used in the referene TMSR Physiohemial properties used for the salts at a mean temperature of 775 C 9

9 1 1. INTRODUCTION Molten Salt Reators (MSRs) have a ertain flexibility when it omes to how they are operated and managed. Molten salt reators have the apability of operating in the thermal, epithermal, and fast neutron spetra and an also use different nulear fuels to produe fission. Figure 1.1 displays a simplified Molten Salt Reator system. MSRs an be operated utilizing the Th/ 233 U fuel yle using molten fluoride or hloride salts. These types of MSRs are alled Thorium Molten Salt Reators (TMSRs) and rely on a thorium blanket to breed 233 U as seen in Equation 1. Using an onsite hemial proessing system, Figure 1.1. Simplified Molten Salt Reator shemati (A Tehnology Roadmap for Generation IV Nulear Energy Systems, 2002)

10 any fission produts along with the produed from the thorium are extrated. Pa Pa 91 has a 27 day half-life before it deays to this U and an be reinserted into the ore. With Th(n, γ) Th β Pa β U (1) hemial proessing system, TMSRs fulfil the riteria of the Generation IV International Forum for nonproliferation onerns and a losed nulear fuel yle. using For the researh presented here, using MCNP, a Thorium Molten Salt Reator U for the initial and operating fuel will be benhmarked based on the data presented by Billebaud Annik and her team at the Laboratory for Subatomi Partiles and Cosmology in Frane. Comparison of average neutron flux as a funtion of neutron energy will be the main basis of benhmarking. Two additional nulear fuels will be ompared with These fuels are U and Pu whih have been hosen due to U nonproliferation onerns. During the old war era, the U.S. and former Soviet Union stokpiled nulear weapons made from and Pu. After the dissolution of the U 92 Soviet Union, disarmament measures in U.S. and Russia provided the opportunity to use these fissile materials as nulear fuel. When used for nulear weapons, these two metals are enrihed to over 90 weight%, however, through fluorination, the metal an easily be dissolved into a salt for use in MSRs (Military Warheads as a Soure of Nulear Fuel, 2014). Along with aiding in the redution of nulear warhead stokpiles, inorporating 235 U and 239 Pu into the first fleet of TMSRs will allow for the reators to power up without an initial fuel of 233 U. The two major aspets looked at will be how well these

11 3 fuels breed U in the blanket and their fluxes aross the system. Ultimately, this thesis will doument whether TMSRs an be initially fueled with these alternate fuels.

12 4 2. THORIUM MOLTEN SALT REACTOR CONCEPT The base design for this projet omes from the Thorium Molten Salt Reator that is being designed by Billebaud Annik and her team at the Laboratory for Subatomi Partiles and Cosmology in Frane as seen in Figure 2.1. The design is a 3000MWth reator that operates on the Th/ 233 U fuel yle with a mean fuel temperature of 750 C. Figure 2.1. Conept design with basi omponents shown of the TMSR (Serp, et al., 2014)

13 5 The entral salt hannel (green) onsists of a LiF, ThF4, and UF4 salt mixture set at a euteti point orresponding to a melting temperature of 565 C as seen in Figure 2.2.The total fuel salt volume is set at 18m 3 with a distribution of half in-ore and half out-of-ore (Merle-Luotte, Heuer, Allibert, Doligez, & Ghetta, Minimizing the Fissile Inventory of the Molten Salt Fast Reator, 2009). The system utilizes a hemial proessing system to remove unwanted fission produts (FPs) from the fuel and blanket salt. The fuel salt reproessing rate is set at 40L per day in order to provide a onstant keff value of one (Merle-Luotte, Heuer, Allibert, Doligez, & Ghetta, Minimizing the Fissile Inventory of the Molten Salt Fast Reator, 2009). Figure 2.3 provides an overview of the hemial proessing system. Figure 2.2. LiF-ThF4 phase diagram (Merle-Luotte, et al., The Non-Moderated TMSR, An Effiient Atinide Burner and a Very Promising Thorium-Based Breeder, 2007)

14 6 Figure 2.3. Basi overview of the hemial proessing system used to purify the fuel and blanket salt (Rubiolo, et al., January 2013) The blanket salt (red) onsists only of LiF and ThF4 and set at the same euteti point. Within the blanket, thorium absorbs neutrons to eventually deay into 233 U (see Equation 1). Every six months, the blanket salt is fully reproessed through the hemial reproessing system to remove the protatinium and uranium and reset it bak to its original omposition. Above and below the ore reside thik nikel hastelloy refletors. The same alloy surrounds the heat exhangers and ore externals. The refletors along with the boron arbide radial refletor that surrounds the blanket salt are apable of bloking over 99% of the esaping neutrons (Rouh, et al., 2014). This allows for a reator operation time of 133 years to produe 100 dpa (displaements per atom) in the most irradiated setion of

15 7 the axial refletor (Merle-Luotte, Heuer, Allibert, Doligez, & Ghetta, Minimizing the Fissile Inventory of the Molten Salt Fast Reator, 2009). The strutural materials surrounding the ore have to bear a high neutron flux oupled with high temperatures. For these omponents nikel hastelloy is used. This Nibased alloy ontains W and Cr and is detailed in Table 2.1. For the simulations used in the literature and in the researh presented here, all materials radially outside of the boron arbide refletor and axially outside of the fuel salt hannel are omposed of nikel hastelloy. Table 2.1. Composition (at%) of nikel hastelloy used in the referene TMSR (Merle- Luotte, Heuer, Allibert, Doligez, & Ghetta, Minimizing the Fissile Inventory of the Molten Salt Fast Reator, 2009) Ni W Cr Mo Fe Ti C Mn Si Al B P S GEOMETRY The main salt hannel of the TMSR is a single ylinder measuring 2.25m high by 2.25m in diameter with a volume of about 9m 3. The blanket that surrounds the main salt hannel is 50m thik. The boron arbide refletor on the perimeter of the blanket salt is 20m thik. The two axial refletors on the top and bottom of the ore measure 50m in height and have a diameter of 2.95m. Figure 2.4 represents a vertial ross-ut of the TMSR.

16 SALT PROPERTIES The initial fuel salt is omposed of LiF-ThF4-233 UF4 and maintains 77.5 mole % LiF throughout operation with a mole % of 7 Li. As seen in Figure 2.2, the heavy nulei proportion is set at 22.5 mole % whih is the euteti point orresponding to a melting temperature of 565 C. The fration of 233 UF4 is set to remain at 2.7 mole % throughout operation of the reator in order to maintain an exatly ritial reativity level. The remaining molar perentage onsists of ThF4. For the radial blanket, an equivalent salt of 77.5 LiF 22.5 ThF4 is used. The physiohemial properties for the TMSR fuel salt are tabulated in Table 2.2 along with the values used for the MCNP simulations. Figure 2.4. Vertial, to-sale ross-ut of the TMSR (Rouh, et al., 2014)

17 9 Table 2.2. Physiohemial properties used for the salts at a mean temperature of 775 C (Merle-Luotte, Heuer, Allibert, Doligez, & Ghetta, Minimizing the Fissile Inventory of the Molten Salt Fast Reator, 2009) Density ρ (g/m 3 ) Formula T ( C) Value at 775 C Value at 630 C Dynami Visosity μ (Pa.s) exp(2812.9/t (K)) Thermal Condutivity λ (W/m/K) T ( C) Calori Capaity (J/kg/K) 1045

18 10 3. MCNP Monte Carlo N-Partile Transport Code (MCNP) is a general purpose software developed by Los Alamos National Laboratory. This ode an be used for transport alulations of neutrons, photons, eletrons, or any ombination of the three. The primary use of the ode involves the simulation of nulear proesses suh as fission, but it has the apability of appliations in areas suh as, but not limited to, radiation protetion and shielding, radiography, medial physis, detetor design, and fusion (X-5 Monte Carlo Team, 2003) MCNP MODEL, CORE SPECIFICATIONS AND ASSUMPTIONS MCNP input files were developed in aordane with the speifiations of the Thorium Molten Salt Reator (TMSR) disussed in Chapter 2. Some speifiations of the TMSR were not provided in urrent literature. In these ases, reasonable assumptions were made based on relative sizes and physial plausibility. MCNP version 5 was used for the neutronis simulations and MCNPx version 2.7 was used for burnup alulations. A height of 185m is assumed for the radial blanket. This assumption is dedued from Figure 2.4 whih is a to-sale ross-ut of the TMSR ore. This sets the volume of the blanket at 8m 3. The 10 B atomi abundane has been shown to flutuate between samples of boron (Baum, Ernesti, Knox, Miller, & Watson, 2009). Given this information, a 10 B atomi abundane of 19.9% is hosen whih orrelates to the most ommon abundane listed in Nulides and Isotopes: Chart of the Nulides (Baum, Ernesti, Knox, Miller, & Watson, 2009). As seen in Figure 3.1, a stainless steel ell (red)

19 11 outlines the perimeter of the reator. Though not in the literature, this ell allows for an aounting of any neutrons that esape past the nikel hastelloy refletors. Figure 3.1. Vertial ross-ut of the simulated TMSR in MCNP. Dark blue is the fuel salt. Light blue is the blanket salt. Green is the boron arbide refletor. Yellow is the surrounding nikel hastelloy. Red is a stainless steel outer shell. In order to first address the neutron flux of the system, the tally f4:n is used. This tally alulates the average neutron flux within the speified ell. In this ase the ell is the entral fuel salt hannel as seen in Figure 3.1. An energy mesh is also used in order to obtain neutron fluxes at eah energy level. This energy range begins at 1.00E-9 MeV and extends to 100 MeV and is broken into 442 bins.

20 12 Figure 3.2. Breeding ratio of the referene TMSR as a funtion of the reproessing rate (Heuer, et al., 2014) For the ritiality alulations, the ommand kode is used. 20,000 initial partiles per yle are used. Using 300 ative yles yields a total of six million histories in the simulation. The total number of histories was suffiient to assure aurate results and reliable statistis. In order to test how well eah fuel salt performs with the blanket salt, MCNPX has to be used. Using MCNPX, the burnup within the blanket an be simulated. For this, a BURN ard is used within the data ard setion of the input ode. For these burn alulations, the main fuel salt remains ritial and no burn is performed on it. This is done beause the fuel salt is reproessed on a daily basis to maintain a ritial reativity value of one. As Figure 3.2 shows, reproessing every day allows for a breeding ratio

21 13 within the ore to be just under This means that the bulk of the breeding omes from the blanket. Sine the ore is initially fueled to be barely ritial and its ability to maintain ritiality heavily relies on reproessing, no burn is required sine it an be assumed that the breeding rate of 233 U is an almost equal ratio to its fission rate. The burn time is set at six months with three intervals of 60 days eah and is simulated at the speified power of 3000MWth URANIUM 233 The 233 UF4 onentration in the fuel salt is set at 2.7 mole %. This equates to about 7 weight % of 233 U. The goal is to benhmark the MCNP data against the data presented within the literature based on the neutron flux in the ore and the amount of 233 U bred from the blanket. As seen in Figure 3.3, it is expeted that a 233 U-started TMSR should be able to breed its initial fissile inventory within 55 years. This expetation oupled with a fast neutron spetrum (see Figure 3.4), will onlude whether the simulated data will be omparable to data presented in the literature URANIUM 235 The TMSR onfiguration fueled with 233 U was set as the referene for a version of the reator fueled with 235 U. It is found that a reator fueled by 235 U requires 4.3 mole % of 235 UF4 in order to maintain a onstant ritiality of one. This equates to 11.3 weight % of 235 U. The neutroni harateristis of the 235 U fueled reator will be ompared with those of the 233 U-based onfiguration.

22 14 Figure 3.3. Exess prodution of 233U as a funtion of time PLUTONIUM 239 For the final simulation, a Plutonium-239 fuel salt will be ompared to the aforementioned fuel ompositions. For this reator, weight % of 239 Pu allows for a maintained ritiality value of one. As in the two previous simulations, the neutron flux along with the amount of fissile inventory produed by the blanket will be ompared to the other simulations and the referene TMSR.

23 Figure 3.4. Neutron spetra for a hloride salt (purple) and a fluoride salt (green) (Rubiolo, et al., January 2013) 15

24 16 4. RESULTS 4.1. URANIUM 233 Figure 4.1 illustrates the average neutron flux for the fuel salt and the blanket salt in the 233 U-operated TMSR. When the green graph in Figure 4.1 is ompared to the green graph in Figure 3.4, it an be seen that the ore flux spetrum are similar. The flux peak from both Figures are in the same order of magnitude. The flux depressions seen between 10 kev and 1 MeV are due to parasiti absorption at resonane energies of 19 F, whih is Figure 4.1. Neutron flux spetrum of the 233 U-operated TMSR shown with the ross setion values of Flourine-19 (Blue). The fuel salt is flux is depited in green whereas a union of the fuel salt and blanket salt is depited in orange.

25 17 abundant in the driver fuel and blanket (see Figs. 4.1, 4.3 and 4.5). However, the results from the alulations performed indiates flux of a higher magnitude than the result from Rubiolo, et al. It should be noted that the flux spetrum averaged over the ative ore and blanket region results in similar spetral magnitude to Figure 3.4. Also, as disussed in further detail in Setion 4.4, the use of an in-house ode by the researhers ould lead to a more refined neutron flux. Sine this is an evaluation of the fuel salt and how it performs, the blanket salt is kept separate in the alulations to produe more aurate results pertaining to the fuel salt. In order to determine the amount of fissile inventory that is bred from the blanket salt of ThF4-LiF, the data is extrapolated based on a six month burn of the system. Sine the blanket salt is reset every six months through the hemial reproessing system, it an be onluded that the data produed from the simulation will provide an exellent estimation for the amount of fissile fuel produed within the blanket. Figure 4.2 illustrates the doubling time of the reator, whih is the estimated time in order to breed enough fissile inventory to replae the initial ore and fuel a seond TMSR. With an initial 233 U inventory of 5,268kg, it will take about 45 years to reah this riteria. This estimate varies by 10 years when ompared to Figure 3.3. This variation an be attributed to the overall higher neutron flux exhibited by the MCNP simulation. A higher neutron rate per area within the blanket allows for more neutrons to be absorbed and thus a faster doubling time of the reator. Overall, it an be onluded that the MCNP simulations presented here are valid given the suessful benhmarking to the 233 U referene TMSR.

26 18 Figure 4.2. Estimation of the doubling time for the 233 U-operated TMSR 4.2. URANIUM 235 The average neutron flux spetra for the 235 U-fueled TMSR is illustrated in Figure 4.3. This spetra has very little variation when ompared to the 233 U model and yields an average relative error of 2.2%. The blanket salt neutron flux is inluded in Figure 4.3 to indiate that a oupling of the two salts neutron flux spetra would produe a slightly lower average neutron flux spetrum. The doubling time for the 235 UF4 fueled TMSR an be seen in Figure 4.4. As in the previous setion, the data is based on one six month burn of the blanket salt with the

27 19 fuel salt hannel maintaining a ritiality value of one for the duration of the burn. With an initial fissile inventory of 8,452kg, the doubling time for the reator is 76 years. This is almost double the required time for a 233 U-started TMSR. This longer doubling time is due to the larger fissile inventory needed to maintain a ritial reator. Figure 5.3. Neutron flux spetrum of the 235 U-operated TMSR. The fuel salt flux is depited in red, whereas the union of the fuel salt and blanket salt is depited in orange PLUTONIUM 239 The neutron flux spetrum for the 239 Pu-operated TMSR is shown in Figure 4.5. The neutron flux spetra follows the same trend as the previous two fuel salts and yields an average relative error of 1.9%. As with the previous figures, the fuel salt and blanket

28 20 salt are separated to emphasize that oupling the two salts together for an average flux spetrum will yield lower fluxes and lead to wrong interpretation of the fuel salt and how it performs. Figure 4.6 shows that the estimated doubling time for a TMSR fueled with 239 Pu is 78 years to replae the initial inventory of 8,580kg needed to maintain a ritiality value of one within the fuel salt hannel. This is longer than either of the previous reators and is due to the higher amount of initial inventory that the blanket must replae. Figure 4.4. Estimation of the doubling time for the 235 U-operated TMSR 4.4. DISCUSSION As seen in Figures 4.1, 4.3, and 4.5, the average neutron flux spetrum is slightly higher than that of the referene TMSR. Along with making the assumption that the referene data is based on a neutron flux of the fuel and blanket salt, it an also be

29 21 dedued that the in-house ode, whih the researhers oupled with MCNP, produed a more aurate neutron energy spetra. With these statements, it an be seen from Setion 4.1 that the MCNP simulation reasonably agrees with the referene TMSR and allows for a good omparison of the 235 U and 239 Pu TMSRs with the 233 U TMSR. Figure 4.5. Neutron flux spetrum of the 239 Pu-operated TMSR. The fuel salt flux is depited in blue, whereas a union of the fuel salt and blanket salt is depited in orange. From the simulations, it an be onluded that a 233 U-fueled TMSR would be the most effiient in terms of breeding more fuel for future TMSRs. However, Figure 3.3 shows that this is the opposite. It shows that the referene TMSR has the longest exess

30 22 prodution time and that a TRU-started TMSR would have the shortest. This is due to the fat that the data is based on the amount of exess 233 U produed in the blanket salt in referene to how muh 233 U has already been used. Sine the TRU and MOx fuels do not require an initial 233 U inventory, their exess prodution of 233 U inreases more rapidly than that of the referene TMSR. Sine the referene TMSR must breed enough fuel to replae its initial inventory before it an show exess prodution, its breeding times are longer. If one ompares the breeding times of the 235 U and 239 Pu reators based on this riteria, it an be onluded that a 239 Pu-operated TMSR would breed 5,200kg of 233 U in just 47.8 years, and a 235 U-operated TMSR would breed the same amount in 48 years. Figure 4.6. Estimation of the doubling time for the 239 Pu-operated TMSR

31 23 5. CONCLUSION A thesis of a Thorium Molten Salt Reator has been presented that benhmarks the referene TMSR and ompares it to two alternate starter fuels maintaining ritiality at an equal euteti point to the referene data. With 235 U and 239 Pu readily available within nulear weapons, these fuels an be used to start the first generation of MSRs and breed 233 U for future generations. It an also be determined from Figures 4.4 and 4.6 that despite the long doubling times for 235 U and 239 Pu, both fissile fuels still breed 233 U at a similar rate to that of a 233 U-operated reator thus onluding that both ould be initial fuels in order to breed more fuel for future TMSRs.

32 24 APPENDIX A. MCNP URANIUM 233 INPUT CODE

33 25 TMSR 2.7% U-233 Cell Cards imp:n=1 VOL= imp:n=1 VOL= imp:n=1 VOL= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n=1 VOL= imp:n=1 VOL= :502:501 imp:n=0 Surfae Cards Central Salt Channel 101 z pz pz Radial Blanket Use 101 as the inner radius 201 z pz pz Boron Carbide Refletor 301 z Use 201 as the inner radius Use 202 & 203 for top and bottom Ni-Alloy 401 z z 250 Use 102 and 103 for the top and bottom 403 pz pz Stainless Steel 501 z 270

34 pz pz Data Cards mode n kode ksr BURN TIME = $ 6 months MAT = POWER = 3000 PFRAC = 1 2r OMIT = AFMIN = 1.0E-10 5r BOPT = MATVOL = Fuel Salt LiF-ThF4-U233F4 m $ 0.001mol% Li $ mol% Li $ 22.5mol% of ThF4-U233F $ 2.7mol% U233F4 Blanket Salt LiF-ThF4 m $ 22.5mol% of ThF Boron Carbide Radial Refletor m $ I used atomi abundane of 19.9% for B $ B10 onentrations an vary from $ 19.5% to 21.5% Nikel Hastelloy Ni wt% m W wt% Cr-5.90 wt%

35 Mo-1.00 wt% Fe-0.5 wt% Ti-0.20 wt% C-0.05 wt% Mn-0.20 wt% Si-0.10 wt% Al-0.02 wt% B-0.01 wt% P-0.01 wt% SS Cr m Ni

36 Mo C S Si N Fe f4:n 1 2 e e-10 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-8 & e e e e e-8 &

37 e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & 29

38 e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e+2 fm E+20

39 31 APPENDIX B. MCNP URANIUM 235 INPUT CODE

40 32 TMSR 4.3% U-235 Cell Cards Cell Cards imp:n=1 VOL= imp:n=1 VOL= imp:n=1 VOL= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n=1 VOL= imp:n=1 VOL= :502:501 imp:n=0 Surfae Cards Central Salt Channel 101 z pz pz Radial Blanket Use 101 as the inner radius 201 z pz pz Boron Carbide Refletor 301 z Use 201 as the inner radius Use 202 & 203 for top and bottom Ni-Alloy 401 z z 250 Use 102 and 103 for the top and bottom 403 pz pz

41 33 Stainless Steel 501 z pz pz Data Cards mode n kode ksr BURN TIME = $ 6 months MAT = POWER = 3000 PFRAC = 1 2r OMIT = AFMIN = 1.0E-10 5r BOPT = MATVOL = Fuel Salt LiF-ThF4-U233F4 m $ 0.001mol% Li $ mol% Li $ 18.2mol% of ThF4-U235F $ 4.3mol% U235F4 Blanket Salt LiF-ThF4 m $ 22.5mol% of ThF Boron Carbide Radial Refletor m $ I used atomi abundane of 19.9% for B $ B10 onentrations an vary from $ 19.5% to 21.5% Nikel Hastelloy Ni wt% m W wt%

42 Cr-5.90 wt% Mo-1.00 wt% Fe-0.5 wt% Ti-0.20 wt% C-0.05 wt% Mn-0.20 wt% Si-0.10 wt% Al-0.02 wt% B-0.01 wt% P-0.01 wt% SS Cr m

43 35.14 Ni Mo C S Si N Fe f4:n 1 2 e e-10 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-8 &

44 e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & 36

45 e e e e e-3 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e+2 fm E+20

46 38 APPENDIX C. MCNP PLUTONIUM 239 INPUT CODE

47 39 TMSR 4.3% Pu-239 Cell Cards imp:n=1 VOL= imp:n=1 VOL= imp:n=1 VOL= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n= imp:n=1 VOL= imp:n=1 VOL= :502:501 imp:n=0 Surfae Cards Central Salt Channel 101 z pz pz Radial Blanket Use 101 as the inner radius 201 z pz pz Boron Carbide Refletor 301 z Use 201 as the inner radius Use 202 & 203 for top and bottom Ni-Alloy 401 z z 250 Use 102 and 103 for the top and bottom 403 pz pz Stainless Steel 501 z 270

48 pz pz Data Cards mode n kode ksr BURN TIME = $ 6 months MAT = POWER = 3000 PFRAC = 1 2r OMIT = AFMIN = 1.0E-10 5r BOPT = MATVOL = Fuel Salt LiF-ThF4-U233F4 m $ 0.001mol% Li $ mol% Li $ 18.20mol% of ThF4-Pu239F $ 4.3mol% Pu239F3 Blanket Salt LiF-ThF4 m $ 22.5mol% of ThF Boron Carbide Radial Refletor m $ I used atomi abundane of 19.9% for B $ B10 onentrations an vary from $ 19.5% to 21.5% Nikel Hastelloy Ni wt% m W wt% Cr-5.90 wt%

49 Mo-1.00 wt% Fe-0.5 wt% Ti-0.20 wt% C-0.05 wt% Mn-0.20 wt% Si-0.10 wt% Al-0.02 wt% B-0.01 wt% P-0.01 wt% SS Cr m Ni

50 Mo C S Si N Fe f4:n 1 2 e e-10 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-9 & e e e e e-8 & e e e e e-8 & e e e e e-8 &

51 e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-8 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-7 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-6 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-5 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-4 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-3 & e e e e e-2 & 43

52 e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-2 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e-1 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+0 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e e e e e+1 & e+2 fm E+20

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