Reactor cell neutron dose for the molten salt breeder reactor conceptual design☆
Introduction
Liquid-fueled molten salt reactors (MSRs) are a promising advanced reactor technology noted for their potential safety (Elsheikh, 2013), fuel cycle (Davidson et al., 2019), and high-level waste reducing (Betzler et al., 2018) characteristics. Operational experience with the 8 MWt Oak Ridge National Laboratory (ORNL) Molten Salt Reactor Experiment (MSRE) in the 1960s demonstrated the feasibility of the reactor technology, investigating dynamic response, chemistry control, and material corrosion issues (Haubenreich and Engel, 1970, Grimes, 1970). Efforts in subsequent years were directed to the development of plants with and without fuel processing (Robertson, 1971, Engel et al., 1980) and smaller demonstration reactors (Bettis et al., 1972). Additional design studies in the following decades have provided the technology landscape with a diverse set of potential designs (Serp et al., 2014, Rykhlevskii et al., 2019) that modern reactor developers are currently pursing in hopes of near-term reactor deployment for commercial operation.
A critical component of reactor operation is radiological protection: the radiological dose during normal operating conditions is a major consideration. Liquid-fueled MSRs have a distributed radiological source during normal operation, which imposes stringent handling requirements of all primary loop components (e.g., piping, heat exchangers, pumps). This setup requires remote operations and maintenance (Holcomb et al., 2018). Significant, persistent neutron dose on primary coolant-adjacent components leads to accelerated radiation damage and activation. This shortens design lifetime and precludes handling of components following reactor operation. In addition, the significant amount of water used for visibility and shielding in most operating reactors (Blumberg, 1967) is generally not compatible for use in high-temperature MSR systems.
Operating experience from the MSRE provides some technical context on maintenance and remote handling approaches for this lower power experimental reactor (Blumberg and Hise, 1968, Holz, 1969). Lessons from the 8 MWt MSRE were integrated into the 2250 MWt Molten Salt Breeder Reactor design (Robertson, 1971), which used a combination of permanent, removable, and portable shielding to develop a full-scale maintenance and operations approach for commercial deployment. This publicly available conceptual design has a relatively complete conceptual operational approach to reactor shielding and maintenance. The MSBR was designed to have a breeding ratio of 1.06 to give an annual fissile yield of 3.3% (Robertson, 1971).
Assessing the radiological conditions during operations, after reactor shutdown, and during maintenance operations requires accurate simulation of various physical aspects of the model and the transport calculation: (1) spatially distributed radioactive sources, (2) calculation of activation in salt and structural materials, and (3) efficient generation of transport solutions. Modern high-fidelity modeling and simulation methods and tools can provide for an improved understanding of radiological dose for these complex problems, and they can also streamline the analyses. Some applications of modern Monte Carlo and deterministic tools have been focused on MSR shielding design and assessment against flux or component dose limits (Zhang, 2013, Merk and Konheiser, 2014). Over the last decade, modeling and simulation for radiological protection was identified as a key objective for future MSR deployment (Betzler et al., 2019, Betzler et al., 2018). The original program plan for the development of molten-salt breeder reactors (McNeese et al., 1971) published about 45 years ago called for a physical test reactor; however, over the years, the emphasis has shifted from setting up test reactors to high-fidelity modeling and simulation. This shift in approach is due to the rising cost associated with setting up test and demonstration reactors, and decreasing costs associated with computing and simulation time due to significant advancements made in high performance computing and the development of state-of-the-art high-fidelity modeling and simulation tools. These high-fidelity tools can be used to model very complex reactors under various conditions prior to setting up an expensive demonstration facility, thereby helping to decrease costs and potentially identifying issues with a particular design early on.
This article discusses application of a modern high-fidelity radiation transport tool to the publicly available facility model from the Molten Salt Breeder Reactor (MSBR) conceptual design. No facility model of similar maturity and scope exists in the open literature that can perform this type of analysis. This article focuses on development of the model itself, testing of the core model, and neutron dose calculations within the reactor cell. Follow-on analyses will focus on assessing additional radiological conditions for specific operating scenarios.
The MSBR model integrates two models: (1) a Monte Carlo N-Particle (MCNP) MSBR core model (Skirpan et al., 2017), and (2) an MCNP model generated from a CAD model of the external components (e.g., pipes carrying fuel salt, primary heat exchangers, salt pumps), as well as the entire reactor building. As is typical, significant clean-up, material assignment (Robertson, 1971), and testing were required for the converted CAD-to-MCNP model to improve human readability and provide for accurate simulation. A fission source mesh was generated using ORNL’s Shift (Pandya et al., 2016) Monte Carlo code to run fixed source calculations within the reactor cell to determine neutron flux, neutron ambient and personnel dose equivalent rate, and reactor vessel iron displacements per atom (dpa). The gamma radiation sources and transport calculation for the MSBR are not included in this paper and will be part of future studies. While it may not be necessary to run a detailed reactor facility model during the preliminary design process, it may be necessary during the final stages of the design process to run detailed facility models that are consistent with the engineering CAD models. These detailed simulations will be necessary to ensure that the dose rates to personnel and equipment in the facility meet the safety criteria and requirements.
As expected, the fuel salt–adjacent components used in liquid-fueled MSRs experience significant neutron dose that is orders of magnitude above doses from typical light-water reactors. Radiological conditions within the MSBR reactor cell during operation are uninhabitable. In the context of high-fidelity ex-core reactor vessel dose methods, the MSBR presents a unique problem because the biological shield surrounds the large reactor cell area that encloses the primary heat exchangers, fuel salt pumps and the reactor core, with little shielding around the reactor core itself. Whereas shielding in typical light-water reactors (LWRs) is provided by a combination of the moderator in the downcomer region between the barrel and the reactor vessel, and the thick concrete biological shield which surrounds the reactor core. These simulations demonstrate a framework in which these critical calculations can be performed using a state-of-the-art computational tool set—Shift—which is nearing acceptance for industry applications (Pandya et al., 2016).
Details and steps taken to calculate the parameters outlined above are provided in the following sections. Section 2 discusses the high-fidelity methods and tools leveraged for this analysis; Section 3 details the geometric model of the MSBR reactor cell and facility; Section 4 discusses the reactor cell neutron dose results; Section 5 discusses the implications of the findings; and Section 6 provides some concluding remarks.
Section snippets
Methods
In a liquid-fueled molten salt reactor, radioactive salt is distributed throughout the fuel salt loops, pumps, and heat exchangers. The reactor building is divided into shielded rooms for the reactor and primary coolant systems, drain tanks, flush tanks, processing equipment, and power conversion machinery. The ideal tool for most efficiently modeling and simulating dose in an MSR facility is a high-fidelity radiation transport tool with hybrid method capabilities.
Approaches for radiation
Model
The fully integrated MSBR model is based on information provided in (Robertson, 1971). This model includes a detailed reactor core (i.e., in-vessel) (Skirpan et al., 2017), primary system, maintenance hatch and area, and reactor building. It captures the components and elements most relevant to reactor dose calculations and more mature aspects of the MSBR conceptual design (Table 1, Table 2). Please see Table S.1 in ORNL-4541 (Robertson, 1971) for more details regarding the individual
Results
As mentioned earlier, the fully integrated MSBR model described in this paper was run with Shift, which was released with SCALE 6.2 Beta 13. During the initial testing of the fully integrated model, eigenvalue calculations were run with MCNP and Shift. These keff calculations yielded sufficiently similar results (Table 10). Both MCNP and Shift results presented in Table 10 were generated with ENDF/B-VII cross section libraries. The MCNP eigenvalue runs were completed with 50,000 particles per
Discussion
Neutron fluxes (n/cm2s) for structural and salt-facing components are significantly higher than those for typical reactor technologies with solid fuels (Table 12). For comparison, in the Watts Bar Nuclear Unit 1 (WBN1) test results (BWXT, 2001), the neutron flux calculated for energies greater than 0.1 MeV in the reactor vessel in Cycle 1 is 5.476 n/cm2s along 45°of the azimuthal angle. The MSBR flux is much higher at the outer edge of the reactor vessel. The iron dpa rate calculated for
Conclusion
The fully integrated MSBR model described herein compromises a full-core, primary system, reactor building model for development and assessment of radiation protection methods and tools critical to informing reactor maintenance and operations procedures for MSR deployment. This model represents the most complete facility model in the open literature and was successfully run with the Shift Monte Carlo tool. The work presented in this paper demonstrates the successful conversion of MSBR CAD
CRediT authorship contribution statement
Eva E. Davidson: Methodology, Formal analysis, Writing - original draft, Visualization. Georgeta Radulescu: Methodology, Formal analysis, Writing - review & editing, Visualization. Kurt Smith: Methodology, Visualization. Jinan Yang: Methodology, Formal analysis, Visualization. Stephen Wilson: Methodology, Formal analysis, Project administration. Benjamin R. Betzler: Conceptualization, Writing - original draft, Supervision, Funding acquisition.
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.
Acknowledgments
The authors would like to acknowledge David Holcomb, Katherine Royston, and Joel Risner at ORNL for their invaluable feedback on this work and paper.
This research was supported by the US Department of Energy (DOE) Office of Nuclear Energy (NE) and the US Nuclear Regulatory Commission (NRC) under various programs. The Molten Salt Breeder Reactor CAD model generation, the initial conversion to Monte Carlo N-Particle (MCNP) geometry, and the initial scoping studies on a surrogate MSR transport
References (55)
- et al.
Fuel cycle and neutronic performance of a spectral shift molten salt reactor design
Ann. Nucl. Energy
(2018) - et al.
Modeling and simulation functional needs for molten salt reactor licensing
Nucl. Eng. Des.
(2019) Stability and Accuracy of 3D Neutron Transport Simulations Using the 2D/1D Method in MPACT
J. Comput. Phys.
(2016)- et al.
Modeling a fast spectrum molten salt reactor in a systems dynamics fuel cycles code
Ann. Nucl. Energy
(2019) Safety assessment of molten salt reactors in comparison with light water reactors
J. Rad. Res. Appl. Sci.
(2013)- et al.
Status of the McCad Geometry Conversion Tool and Related Visualization Capabilities for 3D Fusion Neutronics Calculations
Fusion Eng. Des.
(2013) - et al.
Neutron shielding studies on an advanced molten salt fast reactor design
Ann. Nucl. Energy
(2014) - et al.
Implementation, capabilities, and benchmarking of shift, a massively parallel monte carlo radiation transport code
J. Comput. Phys.
(2016) - et al.
The molten salt reactor (MSR) in generation IV: Overview and perspectives
Prog. Nucl. Energy
(2014) - et al.
The Virtual Environment for Reactor Applications (VERA): Design and Architecture
J. Comput. Phys.
(2016)
Design studies of a molten-salt reactor demonstration plant, Report ORNL-TM-3832
Oak Ridge National Laboratory
Molten salt reactor campaign modeling and simulation program plan, Report
Oak Ridge National Laboratory
Maintenance development for molten-salt breeder reactors, Report ORNL-TM-1859 United States 10.2172/4355531 Dep. CFSTI. ORNL English
Oak Ridge National Laboratory
MSRE design and operations report. Part X. Maintenance equipment and procedures, Report ORNL-TM-910 United States 10.2172/4527944 Dep. CFSTI. ORNL English
Oak Ridge National Laboratory
Denovo—A new three-dimensional parallel discrete ordinates code in SCALE
Nucl. Technol.
Molten-salt reactor chemistry
Nuclear Appl. Technol.
Experience with the Molten-Salt Reactor Experiment
Nuclear Appl. Technol.
Feasibility study of remote cutting and welding for nuclear plant maintenance, Report ORNL-TM-2712
Oak Ridge National Laboratory
Cited by (5)
Layered CAD/CSG geometry for spatially complex radiation transport scenarios
2023, Annals of Nuclear EnergyLayered CAD/CSG Geometry for Neutronics Modeling of Advanced Reactors
2022, Proceedings of the International Conference on Physics of Reactors, PHYSOR 2022
- ☆
Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.
- 1
Work completed while employed at Oak Ridge National Laboratory.