Molten salt modelling capabilities in SPECTRA and application to MSRE and Mk1-PB-FHR
Introduction
The global energy demand is expected to rise the coming decades (IEA, 2018). At the same time, the need for CO2 emission reduction is called upon by IPCC (2018). Consequently there is an increasing amount of people around the world who believe that nuclear energy production, which is one of the main low CO2 emitting energy sources worldwide, should play a role in our future energy mix. However, predictions of the IAEA/NEA (2020) show that the current reserves of uranium in the world, when applied in the light water reactors, are sufficient for the coming 100–150 years only. In order to make nuclear energy production more sustainable in the future and enlarge the horizon of 100–150 years significantly, two major options are considered in the Netherlands by the nuclear stakeholders (Nuclear Netherlands, 2017). The first option is to use the uranium–plutonium fuel cycle and deploy fast neutron reactors. The second option is to switch to the thorium fuel cycle and deploy molten salt fueled reactors. NRG performs research and development activities in both directions. NRG has become well known for its irradiation activities in the field of molten salt reactors (Hania, 2018) and the design of a molten salt demonstration loop as reported by Stempniewicz et al. (2017a). The loop is intended to be constructed in the flux field next to the core of the research reactor in the Netherlands (the High Flux Reactor of HFR). The loop is designed to be subcritical and powered by the neutron the flux from HFR. Simultaneously, NRG is preparing and using simulation approaches and tools for design support analyses and future safety analyses of such a molten salt loop and molten salt reactors in general. MCNP and SCALE are used to estimate power production, long term depletion, and breeding performance of the thorium fuel cycle as well as to determine the neutronically optimal geometry and materials. Thermal-hydraulic design support and safety analyses at this stage are being performed with the system thermal–hydraulic (STH) code SPECTRA. The code has similar capabilities as other well-known STH codes, for example RELAP/SCDAP, CATHARE or SAS4A. However, most of these codes were developed long time ago and can’t easily be applied to molten salt reactors. Therefore, world-wide, efforts are being taken to create or adapt modern STH tools and make them applicable to molten salt reactors. Important for liquid fueled molten salt reactors is the modelling of the delayed neutron precursors. However, also many other aspects have to be taken into account when a complete simulation model of an MSR has to be developed. Examples of such codes are LiCore (Laureau et al., 2020), an extended version of RELAP5 (Shi et al., 2016), RELAP5-3D (Mesina, 2016), SAM (Fei et al., 2020), TRANSFORM (Greenwood et al., 2020), and TREND (Yu et al., 2021).
Compared to most of these codes, the advantage of the SPECTRA code is the flexibility of defining fluid properties in the input deck. This facilitates studying different salt types. In 2020, a collaboration was initiated with the developers of the SAM code from Argonne National Laboratory in the USA to compare modelling approaches on various innovative nuclear reactors and mutually learn from each other with the aim to improve the code capabilities and user competences on both sides.
As usual in STH codes, a point kinetics model is used in SPECTRA to calculate the neutronic behavior. For MSR applications, SPECTRA was upgraded to obtain a fully integrated system code applicable for safety analyses of liquid fuel reactors, including a point reactor kinetics model for circulating fuel systems. This was done by providing an extension of the point reactor kinetics models. This paper presents a summary of the code extensions as well as verification test results. Furthermore, this paper presents the SPECTRA model for the Molten Salt Reactor Experiment (MSRE) and the molten salt cooled Mk1-PB-FHR design and results obtained with these models contributing to the verification and validation of the code. The following chapter will shortly discuss different types of molten salt reactors and touch upon some financial and economic aspects of such reactors. After that, the SPECTRA code is generally introduced in the third chapter, together with a description of the specific capabilities of the code required for molten salt reactor modelling. Subsequently, the next two chapters deal with the modelling of MSRE and Mk1-PB-FHR, shortly describing the reactor designs and the SPECTRA model before entering into the calculation results. Finally chapters 6 and 7 present code capabilities which are relevant for molten salt reactor analyses which have not been applied yet to one of the reactor designs. Rather, these capabilities have been verified based on scarcely available experimental data. Chapter 8 summarizes all the work described and provides some recommendations for future work.
Section snippets
Molten salt reactors
Molten salt reactors (MSRs) are nuclear fission reactors which use a molten salt mixture as primary coolant. This molten salt mixture may include the fuel dissolved in the molten salt mixture. Such reactors are referred to as molten salt fueled reactors, whereas another class of molten salt reactors used solid type of fuel, often similar to the fuels used in high temperature reactors. Such reactors are referred to as molten salt cooled reactors. To achieve a high thermodynamic efficiency, MSRs
General description
SPECTRA (Stempniewicz, 2020a, Stempniewicz, 2020b, Stempniewicz, 2020c) is a thermal–hydraulic system code developed at NRG, designed for thermal–hydraulic analysis of nuclear power plants. Originally developed for Light Water Reactors (LWRs), the flexible code set-up also allows application to High Temperature Reactors (HTRs), Liquid Metal Fast Reactors (LMFRs), and now also MSRs. The code can be used for thermal accident scenarios involving loss-of-coolant accidents (LOCAs), operational
MSRE description
Fig. 2(a) shows a schematic of the Molten Salt Reactor Experiment (MSRE). The reactor was designed for a power of 10 MWt, and the values shown in this section are for this power. The primary loop with the molten salt circulating with a pump at a flow of 75.7 l/s enters the reactor vessel at a temperature of 908 K and leaves at a temperature of 936 K (Carbajo et al., 2017). The primary loop salt was LiF-BeF2-ZrF4, with the fuel dissolved as UF4. The heat generated in the core was removed in a
Mechanism and modeling
Leaching of chromium from alloys at high temperatures and in the presence of salt was observed in the Molten Salt Reactor Experiment (MSRE). Chromium was selectively removed from the Hastelloy N alloy in high-temperature regions and deposited in low-temperature regions (Engel et al., 1980). Chromium depletion was expected to a depth of less than 0.13 mm/year in metal at 704 °C. The mechanism and mathematical model of chromium leaching from Hastelloy N is described in Zheng et al. (2015).
Mk-PB-FHR description
The Mark 1 Pebble Bed Fluoride Salt Cooled High Temperature Reactor (Mk1-PB-FHR) is a pebble bed reactor design cooled with a molten salt. The design is developed in the Department of Nuclear Engineering at the University of California Berkeley and described by Andreades et al. (2014). The 236 MWth reactor is designed to produce 100 MWe of base load electricity when operated with only nuclear heat at more than 42% net efficiency. Fig. 12 shows a schematic diagram for the Mk1-PB-FHR reactor,
Summary & conclusions
The purpose of the presented analyses was to demonstrate that the SPECTRA code is able to model the main phenomena in molten salt cooled and molten salt fueled nuclear reactors. These phenomena include:
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delayed neutron precursor drift
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heat removal by natural circulation
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fission product transport in (fueled) molten salt reactors
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noble gas and noble metal behavior
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noble metal extraction
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chromium leaching and deposition
These models have been applied, when possible, to the MSRE data or other experimental
CRediT authorship contribution statement
F. Roelofs: Conceptualization, Data curation, Visualization, Writing – original draft. M.M. Stempniewicz: Formal analysis, Investigation, Software, Writing – review & editing.
Declaration of Competing Interest
The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.
Acknowledgment
This work was funded by the Dutch Ministry of Economic Affairs. The authors also wish to acknowledge the involvement and support of the following NRG colleagues: Fabio Alcaro, Edo Frederix, Eric de Geus, Ralph Hania, and Blaz Mikuz.
References (45)
Molten salt reactors: a new beginning for an old idea
Nucl. Eng. Des.
(2010)- et al.
Economics and finance of Molten Salt Reactors
Prog. Nucl. Energy
(2020) - et al.
Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code
Nucl. Eng. Des.
(2016) - et al.
Development of TREND dynamics code for molten salt reactors
Nucl. Eng. Technol.
(2021) - Afruns, J.P., Kitchener, J.A. (1976). The absolute rate of capture of single particles by single bubbles, Flotation....
- Ahmed, K., Scarlat, R., Hu R. (2017). Benchmark Simulation of Natural Circulation Cooling Systems with Salt Working...
Technical Description of the ‘Mark 1’ Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB FHR) Power Plant
Department of Nuclear Engineering
(2014)- Ball S.J., Kerlin T.W. (1965). Stability Analysis of the Molten Salt Reactor Experiment....
- Capelli E., Frederix E., Mathur A., Hania R., Kloosterman J.-L., Konings R. (2018). Noble metal behaviour and...
- Carbajo J., De Wet D., Brown N. (2017). Modelling the Molten Salt Reactor Experiment with the RELAP5-3D Code. 2017 ANS...
Technology Roadmap Update for Generation IV Nuclear Energy Systems
Demonstration of the advanced dynamic system modeling tool TRANSFORM in a molten salt reactor application via a model of the molten salt demonstration reactor
Nucl. Technol.
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2022, Nuclear Engineering and DesignCitation Excerpt :The influence of helium bubbles on the removal of fission products, however, has not been investigated yet and is the subject of the this work. In previous studies, the behavior of gaseous fission products has been modelled considering the effects of the different gas sources and sinks present in the Molten Salt Reactor Experiment (MSRE) and implemented in a system code (Roelofs and Stempniewicz, 2021). In Price et al. (2020) and Price et al. (2020), the impact of xenon has been studied and a dynamic model was developed and tested against experimental data from ORNL.