Elsevier

Annals of Nuclear Energy

Volume 103, May 2017, Pages 49-59
Annals of Nuclear Energy

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Core design and safety analysis

https://doi.org/10.1016/j.anucene.2017.01.003Get rights and content

Highlights

  • Engineering demonstration reactors are built as proof of concept for technology options.

  • This paper details a preconcept for a fluoride salt-cooled engineering demonstration reactor (FHR DR).

  • Core design, fuel cycle, and safety of the FHR DR preconcept have been evaluated.

  • Preliminary safety analysis indicates that the reactor could be used to demonstrate inherent safety potential of FHRs.

Abstract

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor II (EBR-II) for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor technologies. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable.

Core design characteristics, fuel cycle performance, and safety analysis of the FHR DR preconcept have been evaluated. The FHR DR core design features a negative or negligible void coefficient throughout a reactor operating cycle. Both single-batch (cartridge) and multiple-batch fuel cycles can be demonstrated in the FHR DR. The single-batch cycle length of the FHR DR core is estimated at between 12 and 18 months, assuming the successful qualification of composite carbon (C/C) or silicon carbide (SiC/SiC) structural fuel block tie rod material. Fuel cycle performance of the FHR DR is similar to a high temperature gas-cooled reactor. Preliminary safety analysis of the FHR DR indicates that the reactor could be used to demonstrate the inherent safety characteristics of FHR designs.

Introduction

This paper is a companion paper to Qualls et al. (2017), which presents a high-level overview of a preconceptual design philosophy for a fluoride salt-cooled engineering demonstration reactor. This paper describes details of the reactor core design and preliminary safety analysis for that engineering demonstration reactor.

In 2015, the U.S. Congress authorized the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to initiate the Advanced Demonstration and Test Reactor (ADTR) study. The ADTR study evaluated advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy. National laboratories, industry, and other relevant stakeholders of an advanced nuclear reactor conducted the ADTR study.

The fluoride high temperature salt-cooled reactor (FHR) engineering demonstration reactor (DR) is a preconceptual design for a small-scale demonstration reactor with the goal of increasing the technology readiness level (TRL) of the overall system (Qualls et al., 2016). The FHR DR was developed as a part of the broader ADTR study. The notional FHR DR concept, also known as a point design, described in this paper was developed for evaluation as part of the ADTR study. The potential missions of the demonstration reactor concepts in the ADTR study include: (1) process heat applications to reduce the carbon footprint of U.S. industry, (2) closing the nuclear fuel cycle and extending natural resource utilization, and (3) deploying a small scale engineering demonstration reactor with the goal of increasing the TRL of the overall system. Point designs were developed for several advanced reactor technology options (e.g. FHR, HTGR, SFR, and others) and considered against each potential mission. The FHR DR targets the third potential mission for an advanced engineering demonstration reactor. The methodology, approach, and findings of the ADTR study are outside the scope of this paper, which is focused on the technical details of the point design for the reactor core and safety analysis.

FHRs comprise a class of reactor concepts that use fluoride salts as low-pressure coolants to produce high-temperature heat with a high degree of inherent safety. In recent years, Oak Ridge National Laboratory (ORNL), the University of California, Berkeley (UCB), and the Massachusetts Institute of Technology (MIT) (among others) have developed salt-cooled FHR concepts.

The FHR DR will demonstrate many key features of FHRs including:

  • Tritium management, salt handling, and passive safety;

  • Active heat exchangers proposed for heat removal during maintenance outages for integral FHRs; and

  • Safety system operation to enable licensing and deployment of inherently safe commercial FHRs.

The FHR DR will demonstrate the inherent safety features of FHR designs, as Experimental Breeder Reactor II (EBR-II) demonstrated the potential for inherently safe sodium fast reactor designs with the shutdown heat removal tests (Planchon et al., 1986).

The FHR DR is designed to be an affordable, near-term system for demonstrating technology solutions that bridge remaining gaps to established technical viability. To meet that goal, lower risk technologies identified from previous experimental and design efforts are incorporated into the design. However, the flexibility of the facility also allows inclusion, demonstration, and qualification of other technologies being pursued for commercial development efforts.

Technology gaps for FHRs include demonstration of a fuel form for a commercial plant, including the ability to efficiently install and remove a core, in-core structural material performance, and inherent safety performance. Other technical uncertainties applicable to salt reactors pertain to salt procurement, chemistry control, and tritium management of lithium- and beryllium-containing salts. Operational uncertainties include demonstration of reliable pump and heat exchanger performance at temperatures of interest using relevant salt coolants. In addition, demonstration of control rod drive mechanisms will be an important aspect of the FHR DR.

The operating scheme for the FHR DR will resemble Shippingport, the historic light water reactor test bed that demonstrated three different core designs within the same vessel. Another relevant example of an engineering demonstration reactor is the EBR II. Similar flexibility in the FHR DR allows for testing of different fuel forms and cores as they become available. As part of the present work, the neutronics and thermal hydraulics of multiple fuel forms in an FHR DR core were evaluated and deemed acceptable. Based on an assessment of technology readiness, the baseline fuel form selected for the FHR DR was TRISO fuel compacts within a prismatic graphite block. This fuel form was previously demonstrated in high-temperature gas-cooled reactors.

Section snippets

FHR DR core neutronic design

The FHR DR uses prismatic block-type fuel with integral fuel compacts and coolant channels as its base fuel form. The fuel is a typical high temperature gas-cooled reactor (HTGR) particle geometry with 15.5% UC0.5O1.5 TRISO kernels with packing fractions of 0.35 within compacts. Selection of prismatic block-type fuel was motivated by the manufacturability of the prismatic fuel for HTGRs (Petti et al., 2009). The recent experience in the advanced gas reactor fuel irradiation program is another

FHR DR thermal-hydraulic design and core analysis

Several tools were used to perform thermal-hydraulic design calculations: COMSOL (COMSOL, 2015) and RELAP5-3D (RELAP5-3D Code Development Team, 2014). COMSOL was used to calculate assembly-level fuel and graphite temperatures, and the RELAP5-3D code was used to perform overall system calculations. In addition, the U.S. Nuclear Regulatory Commission (NRC) tool TRACE (U.S. NRC, 2012) was used to confirm the RELAP5-3D predictions of thermal feedback during transients. The PARCS regulatory-grade

Safety analysis of the FHR DR

The FHR DR has a variety of safety features, including inherently negative reactivity feedback coefficients and a robust fuel form (TRISO). The well-established tools RELAP5-3D and TRACE were applied in a preliminary safety analysis of the FHR DR. Design and licensing strategies by Scarlat et al. (2014) provided the basis for selecting the initiating events and transient scenarios studied for the FHR DR safety analysis. Analyses were performed to determine the potential response of the FHR to a

Tritium management in the FHR DR

FLiBe features relatively attractive coolant properties and is also a neutron moderator. This combination results in coolant temperature and void coefficients that are either negative or negligible, which makes it the coolant of choice for FHR concepts. A challenge with FLiBe coolant is the production of tritium (3H), a radioactive isotope of hydrogen. An economically attractive and technically viable tritium management solution is necessary for the commercial FHR deployment, and a vital role

Summary

The FHR DR is a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. Neutronic and fuel cycle design calculations were performed using the Serpent Monte Carlo reactor physics tool and the PARCS nodal diffusion tool. Thermal-hydraulic calculations performed with the analysis tools COMSOL and RELAP5-3D were used to determine the potential thermal and hydraulic characteristics of the concept. The TRIDENT code was used for scoping evaluation of tritium transport in the

Acknowledgements

This effort was supported by the U.S. DOE-NE in an evaluation of advanced demonstration and test reactor concepts utilizing a variety of technology options. We gratefully acknowledge Tom O’Connor (DOE-NE), Jess Gehin (ORNL), Bob Hill (Argonne National Laboratory), and David Petti (Idaho National Laboratory).

The FHR DR effort benefits greatly from the efforts of many individuals and institutions performing research on Fluoride Salt-Cooled High Temperature Reactors. We would like to credit the

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    This manuscript has been authored by UT-Battelle, LLC under Contract No. 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. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-planhttp://energy.gov/downloads/doe-public-access-plan).

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