The case for the thorium molten salt reactorE. D. Greaves, K. Furukawa, L. Sajo-Bohus, and H. Barros Citation: AIP Conf. Proc. 1423, 453 (2012); doi: 10.1063/1.3688845 View online: http://dx.doi.org/10.1063/1.3688845 View Table of Contents: http://proceedings.aip.org/dbt/dbt.jsp?KEY=APCPCS&Volume=1423&Issue=1 Published by the American Institute of Physics. Additional information on AIP Conf. Proc.Journal Homepage: http://proceedings.aip.org/ Journal Information: http://proceedings.aip.org/about/about_the_proceedings Top downloads: http://proceedings.aip.org/dbt/most_downloaded.jsp?KEY=APCPCS Information for Authors: http://proceedings.aip.org/authors/information_for_authors

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The Case for the Thorium Molten Salt Reactor

E.D. Greavesa,1, K. Furukawab, L. Sajo-Bohusa, H. Barrosa

aLaboratorio de Física Nuclear. Universidad Simón Bolívar, Caracas, Venezuela 1egreaves20002000@yahoo.com

bThorium Tech Solution Inc., Japan .

Abstract. Shortcomings of current PWR and BWR, solid uranium-fuel, nuclear power reactors are summarized. It is shown how the Molten Salt Reactor (MSR) created and operated at Oak Ridge National Laboratory (ORNL), USA (1960s-1970s) and developed as FUJI reactor by Furukawa and collaborators (1980s - 1990s), addresses all of these shortcomings. Relevant properties of the MSR regarding to simplicity, its impact on capital and operating costs, safety, waste product production, waste reprocessing, power efficiency and non proliferation properties are reviewed. The Thorium MSR within the THORIMS-NES fuel cycle system is described concluding that the superior properties of the MSR make this the technology of choice to provide the required future energy in the South American region.

Keywords: Thorium, molten salt, nuclear reactor, nuclear energy, safety, waste processing. PACS: 28.50.Hw; 28.41.Vx; 28.41.Kw

INTRODUCTION

There isn’t a more effective measure for averting social disorder and solving human poverty as ensuring an adequate supply of clean and cheap energy. The growth rate of primary energy in the world is estimated at 2.3% yearly resulting in a doubling time of 30 years. Reliance on available fossil fuels implies greenhouse CO2 gas production and danger of global warming increase. The development and introduction of wind and solar renewable energy is advisable. However, they are low in energy density, irregular in output and still uneconomical and impractical for large industrial scale power plants. Nuclear fusion as a practical power source remains in the distant future. Hence the only technologically developed source capable of supplying the world’s energy demand is nuclear fission provided the chosen technology has an adequate short (~10 years) doubling time [1,2]. Current fission power generation (BWR and PWR) suffers from a number of shortcomings: Their non-acceptance by society even after 60 years of development, the danger of nuclear weapons proliferation, the extreme complexity of solid fuel reactors, the inefficient (5%) utilization of the energy content of the fuel requiring periodic nuclear fuel rod exchange with the result of accumulation of highly toxic spent fuel waste and the use of very high pressure vessels which constitutes a mayor accident safety issue. In this paper we describe the THORIum Molten Salt Nuclear Energy System (THORIMS-NES) which is a concept designed to overcome most of the stated shortcomings.

IX Latin American Symposium on Nuclear Physics and ApplicationsAIP Conf. Proc. 1423, 453-460 (2012); doi: 10.1063/1.3688845

© 2012 American Institute of Physics 978-0-7354-1003-9/$30.00

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THORIMS-NES Nuclear System

The THORIum Molten Salt Nuclear Energy System (THORIM-NES) is a complete fuel cycle concept [2]. It proposes a power reactor (FUJI) radically different from current practice. It uses: A) Thorium instead of uranium as the fertile element to breed the fissile 233U. B) Liquid fuel instead of solid fuel elements. C) It separates the nuclear power production from fuel breeding by a very simple thorium molten salt reactor (Th-MSR) used exclusively for energy generation, initially with 235U or 239Pu and eventually with 233U. D) An Accelerator Molten Salt Breeder (AMSB) devoted exclusively to the production of fissile 233U and E). It incorporates fuel reprocessing in Regional Centers. It is a “Symbiotic” system with each function optimized by its simplicity. The THORIMS-NES concept includes a planned timetable: The first stage is construction of the miniFUJI. A 10 MWe small power reactor to recover the know-how of the Oak Ridge National Laboratory (ORNL), USA, obtained in the period 1964-1969 when the molten salt reactor experiment (MSRE) took place [3]. The miniFUJI is a demonstration reactor that may be developed in a short time estimated at 7 years. The second stage is the FUJI reactor. A 150 MWe thorium molten salt reactor planned to go operational in 14 years as an affordable, simple, safe and reliable power reactor burning either 235U or 239Pu from dismantled nuclear weapons or from spent fuel reprocessing. The third stage estimated some 25 years in the future is the establishment of regional Breeding and Chemical Processing Centers for production of 233U by thorium spallation in AMSB to supplant the use of uranium or plutonium and enter into the thorium nuclear power age.

Why thorium? Thorium-based reactor fuels have a number of advantages over uranium–based

fuels. Th is geochemically three to four times more abundant in the Earth than U. Resources of about 2 M tons have been confirmed with estimated amounts of about 4 M tons [4]. Natural Th has only one isotope, 232Th, and 100% abundant except for about 10ppm 230Th. Hence in the production of a fuel no “enrichment” of the fuel is required. Chemically refined thorium is added directly to the molten salt as discussed below. 232Th in the reactor fuel is converted to the fissile 233U by the reaction:

232Th (n,γ) 233Th (β−: 22.3 m half-life) 233Pa (β−: 27 d half-life) 233U

Fissile 233U is suitable for thermal reactors with the advantage that with fertile 232Th

it can largely eliminate the production of long lived trans-uranium elements (TRU, or actinides) including Pu isotopes. These elements have exceedingly long half lives of the order of 10 000 years or more. Actinide production in a thorium-fueled reactor is estimated to be 2 or 3 orders of magnitude smaller than in a uranium-fueled reactor. This is due to the lighter nature of 232Th against 238U. The negligible production of plutonium makes the thorium-fueled reactor a nuclear weapons proliferation-resistant technology. Plutonium is the ideal isotope for the manufacture of atomic weapons due to the weak accompanying radioactivity. 233U can also be used to make nuclear weapons. However, it is extremely difficult due to the associated high radioactivity.

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The reason is that 233U fuel is accompanied by very strong gamma activity from 232U requiring sophisticated remote handling or a liquid-fuel technology for easy handling. The gamma activity is due to the production of the 232U isotope which takes place in a thorium-fueled system by several n-capture reactions [5]. “No nuclear weapons have ever been deployed using fissile 233U” [6]. Transport of significant amounts of 233U with more than 10 ppm level of 232U require remote handling operations and constitutes a high radiological hazard that requires lead or concrete shielding. Note that it is the daughter products 212Bi (1.8 MeV gamma) and 208Tl (2.6 MeV gamma) isotopes that are very strong gamma emitters and not 232U itself.

Why Liquid Fuel? The idea for a liquid fuel reactor is due to Nobel-prize winner Dr. Eugene Wigner

[7]. This concept was later developed by Oak Ridge National Laboratory (ORNL), USA, through the Molten-Salt Reactor Program (MSRP) during 1957-1976 [8] under the able guidance of his successor Dr. Alvin Weinberg. In the course of this program a Molten Salt Reactor (MSR) operated at ORNL during the four years between 1964 and 1969. The operation was successful; it ran its course without any accident or incident and the developments were fully documented. This extensive and invaluable literature is freely available in the WEB site established by Kirk Sorensen in 2010 [9]. The operation of a power reactor with a liquid fuel as opposed to the well established practice of using solid fuel elements has a large number of advantages. These advantages are most apparent with the liquid media that was developed during the MSRP: A eutectic mixture of lithium fluoride and beryllium fluoride called FLIBE, with fertile thorium and fissile uranium or plutonium dissolved in the fluoride molten salt. (7LiF-BeF2-ThF4-233UF4 ; 73,78 -16 – 10 - 0,22 mol %) This fluid serves a triple function: 1.- As liquid fuel element, 2.- As heat transfer medium, 3.- As fuel processing medium.

As Liquid Fuel Element. In a molten salt reactor the fissionable isotopes, the fertile isotopes and the products of the nuclear reactor operation: both, fission products and heavy elements resulting from neutron capture reactions, reside as ionic elements dissolved in the molten salt. The liquid is forced to circulate in such a fashion that when it enters the reaction chamber, the presence of the graphite moderator material creates conditions for nuclear reaction criticality. The fuel generates heat as the fission reaction proceeds. The heated liquid fuel exits the reaction chamber and the criticality of the fuel ceases while it circulates through the pump, heat exchangers and other devices before returning to the reaction chamber. Solid fuel elements suffer radiation damage. This radiation damage determines a very short life for solid fuel elements such that safety determines an obligatory exchange when only 5%, to at most 10%, of the useful energy has been burned.

On the other hand, a molten liquid fuel is free from structural radiation damage. This property determines that there is no need for fuel element replacement during the life of the reactor. The chemistry of the liquid fuel may be monitored and may be adjusted by very simple addition of components in an external section outside of the reactor vessel. It is easy to add additional fuel salt containing fissile 233U, 235U or 239Pu in order to maintain an optimum fuel composition or likewise to remove deleterious components such as radioactive gases 133Xe and 135Xe. Gases that act as neutron

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poisons due to the huge cross section for thermal neutrons of 135Xe of 2.6 × 106 barns [10]. These gases are removed by helium injection as a carrier gas and collected in active charcoal, stored and allowed to decay before final disposal. With poison gases removed, reactor power can be reduced or increased at will allowing it to follow the power load demand without the limitation that 135Xe buildup imposes on solid fuel reactors. In a molten salt reactor the reactor vessel may operate at low pressure. The container housing the liquid metal or molten salt only requires resisting just the necessary pressure to ensure fuel circulation. Pressure range contemplated in the MSR is about 0.5 MPa (4.93 Atm or 72,5 PSI) which contrasts to pressures in the range of 15 MPa (148 Atm or 2180 PSI) as are used in PWR´s. Hence no large pressure sealing flange is required. This constitutes a significant safety and cost advantage. The possibility of catastrophic reactor vessel failure completely disappears in a liquid fuel reactor. A molten salt reactor, in common with liquid-metal cooled reactors, may operate at a high temperature, several hundred degrees higher than any water cooled reactor. This implies significantly higher thermal efficiency for electrical energy production as well as the possibility of hydrogen production relevant to the establishment of a hydrogen economy which is being currently considered. [11].

As Heat Transfer Medium. The important thermo-physical properties of molten FLIBE-based fuel salt indicate that this solvent salt has good characteristics as a working fluid and coolant. Such excellent characteristics are based on (1) low pressure, (2) highest heat capacity due to the main constituent ions being the smallest possible, (3) low viscosity fluid, and (4) suitable Prandtl number of 10-20 in the fuel-salt. The parameter for heat-transfer per unit pump power has the highest value for FLIBE among other possible molten salt systems. Molten FLIBE is a single phase liquid (MP 480-530ºC, BP about 1400°C) being ideally suited as primary heat exchange fluid from the reactor vessel.

As Fuel Processing Medium. In the THORIMS-NES concept the fuel processing media is FLIBE the same media that is used as a molten salt fluid fuel. Processing is done by high temperature methods called by the generic terms “Pyroprocessing or Dryprocessing” as an alternative to PUREX or other wet hydrometallurgical procedures for spent fuel processing. Advantages are: 1.- They do not use neutron moderator solvents containing hydrogen and carbon, creating risks of criticality accidents, 2.- They are more compact than aqueous methods, 3.- They can separate many or almost all of the elements contained in spent fuel: remaining fertile or fissile uranium and plutonium, fission products and transuranic actinides, 4.- Simplicity of the separating equipment, 5.- No radiation damage is expected on the processing media: the liquid molten salt. Separation of components is achieved by chemical processes such as: 1.- Electro deposition [12]; 2.- Absorption into a liquid metal cathode (cadmium or bismuth) [13]; 3.- By the production of volatile compounds which can be separated by fractional distillation [14] or 4.- By selective precipitation of oxides [15]. Pyroprocessing is the ideal procedure for processing fuel from a molten salt reactor. All the materials to be processed, separated or recovered are already in a suitable molten salt medium which can be used for the recovery process.

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FUJI Reactor The FUJI-series power reactors are modeled on the successful molten salt reactor program (MSRP) developed at ORNL [16]. However important differences are incorporated: FUJI power reactors are simpler, size-flexible and fissile-fuel self-sustaining, which allows a most simple/stable operation and requires minimum maintenance work.

FIGURE 1. Cross section of the primary system of Molten-Salt Power Reactor FUJI [17]

Such idealistic performance is almost realized by the FUJI concept, eliminating the

continuous chemical processing in situ and periodical core-graphite replacement, both of which were needed for the ORNL Molten Salt Breeder Reactor (MSBR). [18]. Figure 1 shows a vertical cross section of the reactor core and primary fuel salt circuit system of FUJI. A standard conceptual design of FUJI [19] is 350 MW thermal and 160 MW electric. The reactor-vessel is cylindrical 5.4 m diameter and 4.0 m high, inside of which it is filled by graphite (93.9 vol.%) and fuel-salt as shown in Figure 2. The reactor-vessel is weld-sealed in the factory and does not need opening during its entire life. The core is made of liquid fuel flowing directly inside central hexagonal graphite rods surrounded by a graphite neutron reflector. Graphite inventory is 161 tons and spatially arranged to get a best performance attaining an initial conversion-ratio of 1.002. The graphite moderator is not expected to be replaced in the life of the reactor. To prevent damage, a neutron irradiation limit of 3 x 1022 nvt is selected (<50 keV). Therefore the maximum core flux should be less than 6 x 1013 n cm-2s-1 in a 30 year life with 60% load as a local power station. High quality graphite with high irradiation resistance and small pore size (< 1 μm) is used. FUJI employs control rods which are made of graphite for the control of power. It employs shut-down control rods using B4C, which are always withdrawn when the reactor is in operation. When inserting the graphite control rod into the core, the graphite functions as a moderator

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which promotes the fission reaction, contrary to the scram control rods. Graphite rods will be drawn by floating power in a fail-safe mode.

FIGURE 2 Cross section of the graphite reactor core. Molten-Salt Power Reactor FUJI [19]

The standard fuel salt of FUJI is 7LiF-BeF2-ThF4-UF4 (69.78-18-12-0.22 mol%).

The total volume of fuel salt is 13.7 m3 flowing upward at a rate of 33.2 m3/min. The inner diameter of the main fuel piping is 25 cm. The structural Ni alloy, Hastelloy N is appropriate for up to 1170K or more. Hence it can operate as industrial-heat supply up to 930K, and in the future 1030K will be feasible. As such it allows for hydrogen production, as well as cogeneration, desalination, and district heating. Centrifugal pumps transfer the outlet fuel salt to heat exchangers, where the heat is transferred to a secondary coolant salt of NaBF4-NaF, which transports the heat to a super-critical steam generator for electric generation, resulting in a thermal efficiency of more than 44%. Several analyses have been carried out to establish nuclear characteristics of FUJI series cores which use several kind of fissile materials (233U, 235U and 239Pu) with denominations such as FUJI-233U or FUJI-Pu [20] and several output powers. A full schematic image of the FUJI molten salt reactor is shown in Ref [20]. It includes the reactor containment building, primary heat transfer and cooling circuit, secondary cooling salt circuit, supercritical steam generator, turbines and electrical generators. The reactor design has a three level containment security: The reactor core is contained in a primary Hastelloy N vessel which is inside the high-temperature containment. The tertiary level is the reactor containment building. The design is extremely safe as the fuel is only critical inside the core. In the unlikely event of reactor fuel salt leakage the molten salt will be caught by a spill pan and flow into a drain tank preventing any release of radioactive material. Overheating protection is provided by a freeze-valve melting and draining the reactor vessel.

CONCLUSIONS Safety. It is practically impossible to have a severe accident or explosion in FUJI as the pressure of the nucleus is very low: 0.5 MPa (72 psi, 5.1 kg/cm2) as mentioned earlier. The molten salt is chemically inert and it does not react with water or air.

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FLIBE’s boiling point is 1673K much higher than the reactor operating temperature of 973K. The fuel (< 1% 233U) is only critical in the graphite inside the nucleus hence re-criticality is impossible. There is no possibility of “failure”, “rupture” or “fusion” of fuel elements which do not exist. Permanent removal of radioactive gases: Tritium, Krypton, Xenon do not accumulate in the fuel. It makes impossible their escape in an accident. Furthermore, with no “Xenon poisoning” it allows changing the reactor power to follow the load. In the case of a loss of electrical power (case of the accident at Fukushima I, Japan, 11th March 2011) a freeze valve releases fuel to a passively cooled drain tank, surrounded by borated water, ensuring automatic shut down. Nuclear waste. The fuel in the reactor (the molten salt) remains permanently in the reactor (30 years) there is no need for spent fuel pools. Thorium as fertile material produces very little actinides (Trans Uranics) resulting in 2 to 3 orders of magnitude less long lived nuclear waste. The molten salt is an ideal media for reprocessing and recovering uranium and plutonium from nuclear waste hence it is a single medium for power production and for fuel processing. Nuclear Non-Proliferation and Terrorism. The thorium MSR has negligible production of plutonium 239Pu burning weapons stockpiles of 235U and 239Pu. It produces 233U with which it is very difficult to make weapons and it is very difficult to transport due to its very high radioactivity. With 233U it is very difficult to produce a “Significant Quantity” Simplicity. There is no need for fuel rod fabrication plants. No fuel elements that have to be exchanged or re-arranged regularly. This results in low construction cost and low operating cost leading to economy on short and long term. Advantages for South America. The Thorium Molten Salt Reactor will help ensure the maintenance of the continent free of nuclear weapons. It will help achieve energy independence of foreign (Imperial) companies and countries. It can make use of indigenous sources of thorium (Cerro Impacto in Venezuela & Brazil Th resources). The small size, relatively simple technology and high security allow deployment near population centers. There is interest in thorium and MSR development in Japan, Czech Republic, Russia, France, Turkey, Singapore, Venezuela and China [21].

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6. Sorensen, Kirk (2010) Remark during the “Media Session” at the Thorium Energy Conference 2010, London, UK, October 17 – 20. Hosted by itheo.org.

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21. The China Academy of Science announced (January 30, 2011) $ 300 million investment for MSR development.( http://energyfromthorium.com/2011/01/30/china-initiates-tmsr/) Acceded Nov.2011.

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