Deep-Burn Molten-Salt Reactors

2006

– One of the pending questions concerning Molten Salt Reactors based on the 232Th/233U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/233U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233U. A particular reactor configuration is used, called ‘unique channel’ configuration in whi...

2010

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled until consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utilit...

Nuclear Engineering and Design, 2019

The reactor performance and safety characteristics of mixed thorium mononitride (ThN) and uranium mononitride (UN) fuels in a pressurized water reactor (PWR) are investigated to discern the potential nonproliferation, waste, and accident tolerance benefits provided by this fuel form. This paper presents results from an initial screening of mixed ThN-UN fuels in normal PWR operating conditions and compares their reactor performance to UO 2 in terms of fuel cycle length, reactivity coefficients, and thermal safety margin. ThN has been shown to have a significantly greater thermal conductivity than UO 2 and UN. Admixture with a UN phase is required because thorium initially contains no fissile isotopes. Results from this study show that ThN-UN mixtures exist that can match the cycle length of a UO 2-fueled reactor by using 235 U enrichments greater than 5% but less than 20% in the UN phase. Reactivity coefficients were calculated for UO 2 , UN, and ThN-UN mixtures, and it was found that the fuel temperature and moderator temperature coefficients of the nitride-based fuels fall within the acceptable limits specified by the AP1000 Design Control Document. Reduced soluble boron and control rod worth for these fuel forms indicates that the shutdown margin may not be sufficient, and design changes to the control systems may need to be considered. The neutronic impact of 15 N enrichment on reactivity coefficients is also included. Due to the greatly enhanced thermal conductivity of the nitride-based fuels, the UN and ThN-UN fuels provide additional margin to fuel melting temperature relative to UO 2 .

Progress in Nuclear Energy, 2008

The performance of natural uranium and thorium-fueled fast breeder reactors (FBRs) for producing 233 U fissile material, which does not exist in nature, is investigated. It is recognized that excess neutrons from FBRs with good neutron economic characteristics can be efficiently used for producing 233 U. Two distinct metallic fuel pins, one with natural uranium and another with natural thorium, are loaded into a large sodiumcooled FBR. 233 U and the associated-U isotopes are extracted from the thorium fuel pins. The FBR itself is self-sustained by plutonium produced in the uranium fuel pins. Under the equilibrium state, both uranium and thorium spent fuels are periodically discharged with a certain discharge rate and then separated. All discharged fission products are removed and all discharged actinides are returned to the FBRs except the discharged uranium utilized for fresh fuel of the other thorium-cycled reactors. 233 U-production rate of the FBRs as a function of both the uraniumethorium fuel pins fraction in the core and the discharge fuel burnup is estimated. The result shows that larger fraction of uranium pins is better for the FBR criticality while larger fraction of thorium fuel pins and lower fuel burnup give higher 233 U production rate.

Research Reactor Fuel Management, 2003

2007

– Molten salt reactors, in the configuration presented here and called Thorium Molten Salt Reactor (TMSR), are particularly well suited to fulfil the criteria chosen by the Generation IV forum, and may be operated in simplified and safe conditions in the Th/233U fuel cycle with fluoride salts. Amongst all MSR configurations in the thorium cycle, many studies have highlighted the configurations with no moderator in the core as simple and very promising. Since 233U does not exist on earth and is not being produced today, we aim at designing a critical MSR able to burn the Plutonium and the Minor Actinides produced in the current operating reactors, and consequently to convert this Plutonium into 233U. This leads to closing the current fuel cycle thanks to TMSRs started with transuranic elements on a Thorium base, i.e. started in the Th/Pu fuel cycle, similarly to fast neutron reactors operated in the U/Pu fuel cycle. We will detail optimizations of this transition between the reactors...

The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6th Framework Program, is mainly aimed at providing additional key elements for the utilisation and transmutation of plutonium and minor actinides in contemporary and future (high temperature) gas-cooled reactor design, which are promising tools for improving the sustainability of the nuclear fuel cycle, hereby also contributing to the reduction of Pu and MA stockpiles, and to the development of safe and sustainable reactors for CO2-free energy generation. The project runs from

1966

This report was .prepared as a n account of work sponsored by an agency of t h e United States Government. Neither t h e United States Government nor any agency thereof, nor any of their employees, make any warranty, express or implied, or assumes a n y legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by t h e United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

Nuclear Science and Techniques, 2017

The molten salt reactor (MSR), as one of the Generation IV advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning (B&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th= 233 U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period (RP) case and about 1.047 for the 10-day RP case. The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing (RP is 180 days), and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning (PB&B) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.

2012

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 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.

Annals of Nuclear Energy, 2010

This paper discusses the use of 241 Am as proliferation resistant burnable poison for light water reactors.

International Journal of Energy Research, 2016

Spent nuclear fuel out of conventional light water reactors contains significant amount of even plutonium isotopes, so called reactor grade plutonium. Excellent neutron economy of Canada deuterium uranium (CANDU) reactors can further burn reactor grade plutonium, which has been used as a booster fissile fuel material in form of mixed ThO 2 /PuO 2 fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates incineration of nuclear waste and the prospects of exploitation of rich world thorium reserves in CANDU reactors. In the present work, the criticality calculations have been performed with 3-D geometrical modeling of a CANDU reactor, where the structure of all fuel rods and bundles is represented individually. In the course of time calculations, nuclear transformation and radioactive decay of all actinide elements as well as fission products are considered. Four different fuel compositions have been selected for investigations: ① 95% thoria (ThO 2) + 5% PuO 2 , ② 90% ThO 2 + 10% PuO 2 , ③ 85% ThO 2 + 15% PuO 2 and ④ 80% ThO 2 + 20% PuO 2. The latter is used for the purpose of denaturing the new 233 U fuel with 238 U. The behavior of the criticality k ∞ and the burnup values of the reactor have been pursued by full power operation for~10 years. Among the investigated four modes, 90% ThO 2 + 10% PuO 2 seems a reasonable choice. This mixed fuel would continue make possible extensive exploitation of thorium resources with respect to reactor criticality. Reactor will run with the same fuel charge for~7 years and allow a fuel burnup~55 GWd/t.

Radioisotopes - Applications in Physical Sciences, 2011

This chapter describes the manufacturing technology of fuel used in research reactors that produce radioisotopes. Besides this production, the research reactors are also used for materials testing. The most common type of research reactors is called "MTR"-Materials Testing Reactor. The MTR fuel elements use fuel plates, which are quite common around the world. There was a historic development in that fuel type over the years to reach the current state-of-art in this technology. The basic MTR fuel element is an assembled set of aluminum fuel plates. It consists of regularly spaced plates forming a fuel assembly. These spaces allow a stream flow of water that serves as coolant and also as moderator to nuclear reaction. The fuel plates have a meat containing the fissile material, which is entirely covered with aluminum. They are manufactured by adopting the traditional assembling technique of dispersion fuel briquette inserted in a frame covered by aluminum plates, which are welded with subsequent rolling. This technique is known internationally under the name "picture-frame technique". Powder metallurgy techniques are used in the manufacture of the fuel plate meats, making briquettes using ceramic or metallic composites. The briquette is made with powdered nuclear material and pure aluminum powder, which is the structural material matrix of the briquette. Using UF 6 in the chemical plant, it is able to produce several intermediate compounds of uranium. One of these compounds is UF 4 , which is the main raw material to produce metallic uranium. It could be made by several routes. The production of metallic uranium uses the UF 4 reduction through calcio-and magnesiothermic reaction. The metallic uranium is alloyed with Al, Si or Mo. Previously, stable uranium oxides were used as MTR fuels, but they had very small densities to accomplish a good operational performance of the reactors. The fuel material candidate mostly prone to be used in nuclear research reactors is based on alloys carrying more U-density toward the fuel meat. In present state, the U-Mo alloys are good candidates, but it would not be subject of the present chapter since it on its path to be certified to future use in research reactors. Currently, the most used material is U 3 Si 2 LEU, which is low enriched uranium enriched up to 20% of 235 U isotope, which is the nuclear fissile material.

Energy, 2010

Nuclear power can play a substantial role in countering global warming. There are still unsolved problems such as safety, nuclear proliferation, radioactive-waste under using U-Pu system. Transition from U-Pu LWR (Light Water Reactor) system to Th-233 U MSR (Molten-Salt Reactor) system has been analysed in view of the utilization of fissile in form of Pu fuel salt applying the simplified FREGAT process to the spent fuel of LWR. AMSB (Accelerator Molten-Salt Breeder) was also applied as a fissile producer. All fissile in spent fuel can be used by Th-U MSR system so as not to remain storage of spent fuel after retirement of LWR system. The maximum capacity of Th-U MSR system will reach to about 20 Â 10 3 GWe. However storage of spent fuel will remain for the case of rapid growth of Th-U MSR system even though the maximum capacity is large enough. AMSB will start operation about 20 years after the beginning of Th-U MSR system but the timing can be greatly advanced with the scenario of LWR system. Th-U MSR system can be implemented by using the fissile material in spent fuel from LWRs. Detailed assessment of other materials, performance of facilities, strategies of non-proliferation will be needed for the future improvement.

Nuclear Engineering and Design, 2000

Energy production in nuclear power plants on the basis of fission processes lead inevitably to fission products and to the generation of new actinide isotopes. Most of these fission products are rather shortlived and decay within less than about 500 years to stable nuclides. However, a few of them, e.g. 99 Tc and 129 I, are longlived and may contribute to the radiotoxicity and hazard associated with an envisaged repository for their long-term disposal in a stable geologic formation, e.g. a salt dome. The majority of the generated actinide isotopes are fairly longlived, e.g. 239 Pu with a halflife of more than 20 000 years. Therefore, their direct storage poses a heavy burden on the capacity and the possible environment impact of a repository. Furthermore, the energy content of these actinides could be deployed for producing additional nuclear fission energy after recovering them from unloaded irradiated fuel by suitable reprocessing techniques. Various possibilities exist for burning these actinides in different types of reactors, e.g. in light water reactors (LWRs), or LMFRs, adhering to available technology, or in actinide burners particularly designed for the purpose of their efficient incineration. The different options will be discussed in the paper. Transmutation of the manmade actinides and longlived fission products will require advanced technologies e.g. regarding reprocessing losses, remote fabrication techniques, and most probably, isotope separation processes. However, the almost complete elimination of these nuclides resulting from fission energy production in a continued recycling process may be the only feasible way to limit the effects of nuclear power generation to a tolerable and fair level for generations to come.