Molten Salt Reactor Design Research Articles

Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

Abstract In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

ICONE23-1700 SIMULATION OF THE SYSTEM DYNAMICS FOR THE MOLTEN SALT REACTOR EXPERIMENT USING MODIFIED TRACE

The aim of this work is to use a modified version of TRACE for a preliminary study of a molten salt reactor (MSR) system behavior for the steady-state, transient and accidental conditions so as to demonstrate the suitability of the TRACE code for MSR simulation. To this end the analysis and results from the MSRE (Molten Salt Reactor Experiment) were used as a reference case. The basic approach of this work is to couple the 1D thermal-hydraulic model for 3 loops and the 0D neutronic dynamics (point-kinetic model), which was solved by the ODE-solver built by "Control Blocks" offered in TRACE. Additionally, new working fluids, namely the molten salts, were added into the source code of TRACE. Most results of the simulations show good agreement with the ORNL reports as well as the previous study in recent years and the errors were predictable and in an acceptable range. Therefore, the necessary code modification of TRACE appears to be successful and will be refined further in order to investigate new MSR designs, such as MSFR and DFR.

Alloys compatibility in molten salt fluorides: Kurchatov Institute related experience

In the last several years, there has been an increased interest in the use of high-temperature molten salt fluorides in nuclear power systems. For all molten salt reactor designs, materials selection is a very important issue. This paper summarizes results, which led to selection of materials for molten salt reactors in Russia. Operating experience with corrosion thermal convection loops has demonstrated good capability of the “nickel–molybdenum alloys+fluoride salt fueled by UF4 and PuF3+cover gas” system up to 750°C. A brief description is given of the container material work in progress. Tellurium corrosion of Ni-based alloys in stressed and unloaded conditions studies was also tested in different molten salt mixtures at temperatures up to 700–750°C, also with measurement of the redox potential. HN80MTY alloy with 1% added Al is the most resistant to tellurium intergranular cracking of Ni-base alloys under study.

Preliminary study on plutonium and minor actinides utilization in thorims-nes minifuji reactor

Molten salt reactor (MSR) design has been chosen as one of the Generation IV nuclear energy systems, since it has many merits such as proliferation resistance, resource sustainability, and safety improvement. Recently, a concept which called as thorium molten salt nuclear energy synergetic (THORIMS-NES) was being proposed for the safe and sustainable nuclear industry. THORIMS-NES miniFUJI reactor is a small power MSR which originally considers Th/233U or Th/Pu as main fuel. In this study, the utilization of Pu and minor actinides (MA) in 25MWth and 50MWth miniFUJI reactors has been evaluated. The reactor grade plutonium and weapon grade plutonium are employed in the present study. The criticality for 25MWth of miniFUJI can be accomplished by loading 8.76% of reactor grade Pu & MA (RGPuMA), and 3.96% of weapon grade Pu & MA (WGPuMA) in fuel, respectively. While, for that of 50MWth, the reactor can attain its criticality with 9.16% and 4.36% of RGPuMA and WGPuMA, correspondingly. The neutron spectra become harder with the lower grade of fissile plutonium vector as well as the increasing of Pu and MA contents in loaded fuel salt.

Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail.

New breeding gain definitions and their application to the optimization of a molten salt reactor design

The molten salt reactor (MSR) is an attractive breeder reactor. A graphite-moderated MSR can reach breeding because of the online salt processing and refueling. These features give difficulties when the breeding gain (BG) of the MSR is evaluated. The inventory of the core and external stockpiles have to be treated separately in order to quantify the breeding performance of the reactor. In this paper, an improved BG definition is given and it is compared with definitions used earlier. The improved definition was used in an optimization study of the graphite – salt lattice of the core. The aim of the optimization is a passively safe, self-breeder reactor. The fuel channel diameter, graphite volume and thorium concentration were varied while the temperature feedback coefficient of the core, BG – as defined in the paper – and the lifetime of the graphite were calculated. There is a small range of lattices which provide both negative temperature feedback and breeding. Furthermore, breeding is possible only at low power densities in case of the salt processing efficiencies set in this study. In this range of power the lifetime of the graphite is between 12 and 20 years.

Considerations on safety and proliferation-resistant aspects for the MSBR design

Even though a molten salt reactor (MSR) design did not pop up on industrial scale such as the liquid metal fast breeder, it has been retained under the six reference concepts of Generation IV. Some states have definitely interest in it, in particular when considering the use of large Th resources in a U–Th fuel cycle. Japan, Russia and the US reported under the ITHSMO consortium a well-advanced development stage of the FUJI 100 MWe MSR. This paper recalls the reference-design of the Molten Salt Breeder Reactor of ORNL while focusing on two opposing aspects the breeding flexibility and the complex operation safety (in particular reactivity control). Contrary to the safety aspects, for which solutions were uncovered in the European MOST project, it was felt necessary for the safeguards aspects to complete a proliferation sensitivity study and to analyse innovative solutions for proliferation-resistance. The assessment of the Molten Salt Breeder Reactor (MSBR) to the proliferation-resistance criteria of Generation IV reactors leads to propose no Pa and U removal in the reprocessing of the fuel. This proposal exploits the activity level of U 232 as intrinsic barrier against nuclear material proliferation and showed that a 60 ppm U 232 impurity level, with a high impurity precursor level, would enable this. However, without the removal of Pa and U from the fuel mixture, the reactivity starts to fluctuate and needs compensation.

Feasibility of a critical molten salt reactor for waste transmutation

This study assesses the feasibility of designing a Molten Salt Reactor (MSR) using the salt mixture of LiF (15 mol%), NaF (58 mol%) and BeF 2 (27 mol%) to be critical when fuelled with TRU from LWR spent fuel without exceeding the actinides solubility limit and while extracting fission products at realistic rates. The first part of the study investigated the graphite-to-MS volume ratio on the neutron balance, transmutation characteristics and graphite lifetime. It is found that a core without graphite moderator is the preferred design option; it offers the best neutron balance, most compact design and alleviated graphite lifetime problem. The second part of the study investigated sensitivity of the epithermal spectrum core to the feed composition, power density, fission products residence time and actinides loss fraction. It is found that the transmutation effectiveness improves with increasing power density and that the shorter the LWR spent fuel cooling time is, the better becomes the MSR neutron balance. The optimal MSR design offers a remarkably high transmutation capability – fissioning of as high as 99.8% of the TRU fed. The transmutation capability of the MSR is also rated in terms of final waste radiotoxicity, decay heat, spontaneous fission neutrons emission, fissile and 237Np inventory.

Recommendations for a restart of molten salt reactor development

The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often-cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. Strong incentives for the molten salt reactor design are its good fuel utilization, good economics, amazing fuel flexibility and promised large benefits. It can: • use thorium or uranium; • be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have an epithermal neutron spectrum; • fission uranium isotopes and plutonium isotopes; • produces less long-lived wastes than today’s reactors by a factor of 10–100; • operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon grade fissile fuel; • be a breeder or near breeder; • operate at temperature >1100 °C if carbon composites are successfully developed. Enhancing 232U content in the uranium to over 500 ppm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR are enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high-enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/year base program for 10 years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/year over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the number of steps to commercial deployment. Assuming electricity is worth $ 50 per MWe h, then 50 years of 10 TWe power level would be worth 200 trillion dollars. If the MSR could be developed and proven for 10 B$ and would save 10% over its alternative, the total savings over 50 years would be 20 trillion dollars: a good return on investment even considering discounted future savings. The incentives for the molten salt reactor are so strong and its relevance to our energy policy and national security are so compelling that one asks, “Why has the reactor not already been developed?”

Experimental Studies on the Use of Liquid Lead in a Molten Salt Nuclear Reactor

A set of experimental studies is presented as a first approach to the problems that liquid-lead circuits might pose in the molten salt reactor design with cooling by direct contact between the salt and this liquid metal. Technologically it appears that the components of circuits developed for the use of liquid sodium in fast neutron breeder reactors (valves, electromagnetic pumps and flowmeters, pressure transducers, and cold traps) can be used in the presence of liquid lead, though with certain restrictions.Where corrosion is concerned, ferritic steels, although subject to mass transfer phenomena, are much more resistant than austenitic steels at the temperatures currently adopted in the molten salt reactor design. Finally, liquid lead could have a slight embrittling effect on ferritic steels, but this phenomenon needs to be checked more thoroughly.

Graphite and Xenon Behavior and their Influence on Molten-Salt Reactor Design

Existing data an dimensional changes in graphite have been fitted to parabolic temperature-sensitive curves. From these, the graphite life, radiation-induced stresses, and permissible geometries ha...