On the Employment of a Chloride or Fluoride Salt Fuel System in Advanced Molten Salt Reactors, Part 2; Core Inventory, Fuel Burnup, and Salt Clean-Up System

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On the Employment of a Chloride or Floride Salt Fuel System in Advanced Molten Salt Reactors, Part Two; Core Inventory, Fuel Burnup, and Salt Clean-Up System

Abstract: Breed and Burn fuel cycle in molten salt reactors qualifies this reactor type as one of the best candidates to be developed for the Generation Four R and D program. This feature can be approached by employing a closed fuel cycle and application of a molten salt reactor as a spent nuclear fuel burner; the features promise sustainable and clean energy in the future. In this study, a complete package has been developed to calculate core inventory, fuel burnup, and salt clean-up systems of molten salt reactors during their lifetime. To achieve this, the iMAGINE-Three-BIC package ("IMAGINE Three-D-Reg Burnup and Inventory Calculator package") has been developed in MATLAB R two thousand twenty-three a by employing a CINDER ninety module of MCNPX two point seven for burnup-calculation and multi-linear regression method. The package can estimate the core inventory (concentration of twenty-five actinides and two hundred forty-five non-actinides elements) and the burnup of the reactor core during molten salt reactor lifetime (up to one hundred years) while optimizing the computational resources (time, CPU and RAM), and it can even be hassle-freely executed on standalone PCs in an appropriate time due to its generous database. In addition, the salt clean-up module of the iMAGINE-Three-BIC package can be employed to evaluate the effects of the salt clean-up system on the above parameters over the molten salt reactors' lifetime. Finally, the iMAGINE-Three-BIC package has been applied to an iMAGINE reactor core design (University of Liverpool, UK-chloride-based salt fuel system) and an EVOL reactor core design (CNRS, Grenoble, France, fluoride-based salt fuel system) to evaluate and compare the performance of chloride/fluoride-based salt fuel molten salt reactors from the point of burnup, core inventory, and salt clean-up systems. The results confirm that while a chloride-based salt fuel system has some advantages in less dependency on the salt clean-up system and fewer poisoning elements inventory, the fluoride-based system can achieve higher burnup during the reactor lifetime. The outcome of this study, along with the first part of this article, provides evidence to support the neutronic decision matrix as well as the pros and cons of employing chloride- or fluoride-based fuel systems in molten salt reactor cores.

One. Introduction

Generation Four nuclear reactors represent a remarkable leap forward in the evolution of nuclear energy technology, introducing a host of innovative features and design concepts that promise to address many of the challenges faced by earlier generations of reactors. Molten salt reactors-as one of the Generation Four candidates-represent a revolutionary approach to nuclear energy generation that offers numerous advantages over traditional solid-fuel reactors. One of the key aspects that sets molten salt reactors apart is their remarkable fuel burnup capabilities. Fuel burnup, in the context of molten salt reactors, refers not only to the efficient utilization of nuclear fuel, where a significantly higher percentage of the fissile material is consumed before it is discarded as waste but also to their breeding cycle, which produces additional fissile material. This efficiency is a fundamental characteristic of well-designed molten salt reactors and is central to their promise of safer, more sustainable, and potentially game-changing nuclear energy technology.

Burnup of nuclear reactors can directly affect the core inventory, breeding capabilities, and their neutronic parameters and indirectly affect thermal-hydraulic ones and even radiation damage to the fuel and structural materials. There is a strong need for a comprehensive assessment of these parameters throughout the reactor lifetime since their evolution can significantly affect the safety margins of the reactor. Considering the importance of evaluation of the molten salt reactors burnup and its reactor core-related parameters, various research studies have been conducted that belong to one of the following categories: all employing burnup as the main parameter; (i) burnup and core inventory calculations, (ii) burnup optimization, (iii) salt clean-up systems, and, finally, (iv) waste transmutation.

The first step in evaluating fuel utilization in molten salt reactors is calculating the burnup and core inventory during their life cycle and considering the effects of variables. A wide range of codes and methods are usually employed for this purpose; ORIGEN, SCALE, SERPENT, perturbation theory, and linear chain method are more popular, among others. The neutron spectrum of the reactor has different effects on the core inventory of molten salt reactors and needs to be evaluated individually. For small-scale molten salt reactors, the difference in fuel utilization under thermal and fast spectra (both IMAGINE and EVOL) is small. However, for large-scale reactors, the achievable burnup under the fast spectrum is significantly higher than that under the thermal spectrum.

A molten salt fast reactor is predicted to work in a closed uranium/plutonium or thorium/uranium-based fuel cycle with a full reprocessing of all actinides in the core. To reach this point, Ashraf et al. modeled the primary circuit of the molten salt fast reactor (European model) to optimize the concentration of the start-up liquid fuel using the code SERPENT two point zero. They have found that a molten salt fast reactor was self-sustained regardless of the type of fissile materials used. The plutonium trifluoride (plutonium in plutonium trifluoride is a vector of plutonium two hundred thirty-nine, zero point six nine zero two, plutonium two hundred forty, zero point two six seven, plutonium two hundred forty-one, zero point zero one seven six, and plutonium two hundred forty-two, zero point zero two five two as the pair of isotope and weight fraction, respectively) and trifluoride of trans-uranium elements fuels appear to be prospective fuels compared to the uranium two hundred thirty-three tetrafluoride fuel. One of the other parameters that need to be optimized is the fuel-salt combination and geometry of the molten salt fast reactor core structure to identify the best candidate of fuel-salt composition from burnup optimization points and even the proportional ratio of salt (reactor core) to the moderator (usually graphite). It is important to define the salt composition with a special focus on the amount of heavy metal that can be carried in the salt. Strong computational resources are usually employed to simulate these salt combination parameters and burnup; thus, the development of artificial intelligence and machine learning methods can support the expensive computations (both time and CPU) of burnup to optimize the parameters. A fast-filtering model for burnup equilibrium state and a fast prediction model for equilibrium neutronic properties were developed based on the machine learning technique by Chen et al. Considering the various performance metrics for measuring the predicted performances of machine learning models in the classification and regression, the LightGBM model looks the most favorable for filtering the burnup state and predicting the neutronic parameters in molten salt fast reactors.

Molten salt fast reactors allow for continuous online fuel treatment and processing using a variety of subsystems. Among these subsystems, the salt clean-up system continuously removes fission products from the primary fuel salt and has a paramount role in the development of molten salt fast reactors. The first step in this process should be identifying the key poisoning elements to be separated in molten salt fast reactors' salt clean-up systems, as the removal of these elements that prevent the reactor from long-term operation is a vital step. A series of calculations of the amounts of the specific elements appearing in the core after a burnup of one hundred gigawatt-days per metric ton of heavy metal have identified ruthenium and molybdenum as elements with the highest influence on criticality, with each formed after one hundred gigawatt-day per metric ton of hypermetallic burnup at the reasonably high concentration of two thousand five hundred to almost four thousand parts per million. The next elements identified were caesium, neodymium, and palladium. A salt clean-up system can be simulated through the batch execution of burnup calculation and removing the positioning elements by a manager kernel (exchanging data of modules). The effects of removal then can be investigated through results in comparison to the continuous burnup simulation without the removal of poisons.

Finally, one of the most important features of MSFRs is to employ this reactor type for the waste transmutation process. This application has been developed in different projects and collaborations such as MOSART, MARS, and MIMOSA. This feature can be achieved through two different approaches, (i) designing MSFRs in a way that can utilize spent nuclear fuel as a fuel source and operate in a closed fuel cycle and (ii) designing waste transmuter based on MSFR technology. In addition to employing MSFR technology for the latter option, Advanced Liquid-Metal Reactors, Particle Bed Reactors, accelerator transmutation of waste systems, and accelerator-driven fast reactor concepts are other systems that have been suggested by different technology developers. Designing such a waste transmutation system needs an entire evaluation and assessment of burnup behavior through the reactor lifetime and finding the best steps available to achieve this goal.

Following the first part of this work, in this study, a new code package entitled "IMAGINE-3BIC" has been developed for an entire evaluation of burnup and core inventory during the MSR/MSFRs' lifetime. The iMAGINE-3BIC package includes its dedicated burnup and core inventory database resulting from using the CINDER ninety module of MCNPX-based on IMAGINE and EVOL (as chloride- and fluoride-based MSFRs). It employs a multi-linear regression method that can optimize the computational cost of burnup/core inventory calculation during a long reactor lifetime (up to one hundred years) while it could impose high computational costs without using this package. The developed salt clean-up module of the iMAGINE-3BIC package can also be used to apply and assess the effects of the salt clean-up system on the burnup and core inventory of MSRs while considering different elements extractions. Eventually, a complete comparison has been performed on the pros and cons of chloride or fluoride salt fuel MSFRs on the burnup and core inventory parameters. In addition, the result of this study evaluates the effects of power scale-up on burnup and core inventory of zero power and demonstrator MSFRs.