Graphite for Advanced Nuclear Reactors: Deployment Readiness Review

Smith, Rebecca E [Idaho National Laboratory] (ORCID:0000000242059751); Windes, William E [Idaho National Laboratory] (ORCID:0000000188030885); Craft, N. [Dominion Engineering, Inc.]; Marks, C. [Dominion Engineering, Inc.]; Metcalf, Martin [The University of Manchester]; Mack, Andrea L [Idaho National Laboratory] (ORCID:0000000238965540); Campbell, Anne [Oak Ridge National Laboratory]; Gallego, Nidia [Oak Ridge National Laboratory]; Geringer, Josina Wilna [Oak Ridge National Laboratory]

DOE OSTI · 2515603

Graphite for Advanced Nuclear Reactors: Deployment Readiness Review

Smith, Rebecca E [Idaho National Laboratory] (ORCID:0000000242059751)·Windes, William E [Idaho National Laboratory] (ORCID:0000000188030885)·Craft, N. [Dominion Engineering, Inc.]·Marks, C. [Dominion Engineering, Inc.]·Metcalf, Martin [The University of Manchester]·Mack, Andrea L [Idaho National Laboratory] (ORCID:0000000238965540)·Campbell, Anne [Oak Ridge National Laboratory]·Gallego, Nidia [Oak Ridge National Laboratory]·Geringer, Josina Wilna [Oak Ridge National Laboratory]

View original source ↗Read source document ↗Export metadata ↓

Abstract

Historically, graphite has been used in numerous reactor technologies, including research/test reactors and, commercially, in advanced gas reactors. These non-metallic materials can play a key role as internal core structures, reflectors, or neutron moderators, making them important for deployment of certain advanced reactor technologies. This report explores the industry readiness for graphite material deployment, documenting Codes and Standards applicable to their design, qualification, and manufacturing. Discussions also examine the manufacturing processes, aging/degradation, inspection techniques, and disposal options. The report covers the current status, documents gaps, and proposes some conclusions about approaches to managing some of the current gaps.

Explore related subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS 99 GENERAL AND MISCELLANEOUS Graphite Graphite for Advanced Nuclear Reactors Graphite for Advanced Nuclear Reactors Deployment Readiness Review

Keep this discovery

Explore connections, maps & timelines

BibTeX RIS

Smith, Rebecca E [Idaho National Laboratory] (ORCID:0000000242059751), Windes, William E [Idaho National Laboratory] (ORCID:0000000188030885), Craft, N. [Dominion Engineering, Inc.], Marks, C. [Dominion Engineering, Inc.], Metcalf, Martin [The University of Manchester], Mack, Andrea L [Idaho National Laboratory] (ORCID:0000000238965540), Campbell, Anne [Oak Ridge National Laboratory], Gallego, Nidia [Oak Ridge National Laboratory], Geringer, Josina Wilna [Oak Ridge National Laboratory]. 2025-02-13. Graphite for Advanced Nuclear Reactors: Deployment Readiness Review. https://www.osti.gov/biblio/2515603

Cite the original work for its findings. Save a collection to share your selection of sources.

KEEP EXPLORING

Related reports

DOE OSTI2026

Evaluating transient fission gas release in high burnup light water reactor fuel during loss of coolant accident conditions via new capabilities

In this work, the role of transient fission gas release (tFGR) in the cladding burst behavior of high burnup fuel during a loss-of-coolant accident (LOCA) in commercial light water reactors was further investigated via the use of a new apparatus. During the LOCA-related temperature ramps of high burnup fuel, the release of fission gases exceeds the steady-state release observed under normal operating conditions. An enhancement was made to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) to probe the various factors influencing tFGR. Experiments were performed on commercially irradiated, zirconium-clad uranium dioxide fuel, and this paper details the design of the experimental setup, the initial test results, and the subsequent post-test analyses. Notably, the first test on high burnup fuel demonstrated a LOCA-relevant tFGR of 5.3% from an unpressurized fuel segment. The ultimate tFGR was 10.7% for beyond LOCA conditions. A follow-up test on similar fuel revealed a tFGR of 12.6% under comparable conditions. Microstructural analysis and an analysis of the released gas provide some insight regarding the source of tFGR in the fuel. Finally, a grain boundary bubble model may aid in the interpretation of the results and offer a guide for future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Nondestructive Assessment of Potentially Contaminated DUF6

Explore the source record for details and available documents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Characterization of ceramic fuel powder packing fractions to support INFLUX

Triply periodic minimal surface (TPMS)-based structures show marked potential in novel nuclear reactor fuel designs, as their high surface area-to-volume ratio increases the efficiency of heat transfer out of the fuel, enabling safer, more innovative reactor designs. This milestone report addresses the role of dUO 2 powder processing route on the fill behavior of TPMS-based cladding shells to understand and advance the feasibility of manufacturing TPMS-based nuclear fuel forms. dUO 2 powder was processed through either a dry granulation route, varying consolidation pressure, or through milling, varying milling time, milling method and milled size distribution. The lowest tapped bulk densities (TBD), but best powder flowabilities, were obtained when testing unprocessed dUO 2 powder which was prone to self-agglomeration and formed low-density spheroids. The highest TBD and lowest flowabilities were obtained when using powder produced by hammer-milling dUO 2 powder to pass through a 200-mesh sieve, which led to particles with angular morphologies. Powder produced by dry granulation exhibited TBD that varied according to the consolidation pressure used to form the initial pellets and exhibited improved flowabilities when compared to hammer-milled material. Because of the large span of granule sizes formed as well as the irregular shape associated with the granules, a packing fraction of 0.69 was achieved, exceeding the analytical solution for random close packing of mono-sized spheres. TPMS polymer shells were loaded with unprocessed, granulated, and hammer-milled dUO 2 powders, and their qualitative packing behaviors were analyzed using x-ray computed tomography (xCT). TBDs calculated after loading TPMS polymer shells were 10-20% lower when compared to tapped bulk density measurements taken in a glass graduated cylinder, indicating a non-trivial impact on the tapped bulk density of either the TPMS channel size, TPMS channel surface material, powder cohesiveness, or a combination of the two parameters. A metallic zircaloy-4 TPMS shell will be loaded with hammer-milled dUO 2 powder upon receipt of the shell from Oak Ridge National Laboratory (ORNL) and shipped to Idaho National Labs (INL) for subsequent hot isostatic pressing (HIP) densification experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗