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Microreactors, or very small, transportable or mobile nuclear reactors with a capacity of less than 20 MW(thermal), are being developed to provide heat and power for myriad applications in remote areas, military installations, emergency operations, humanitarian missions, and disaster relief zones. A wide variety of reactor types are under consideration, including sodium-cooled fast reactors, molten-salt reactors, very high-temperature gas reactors, and heat pipe reactors. One issue common to all microreactor designs is the need to remove heat from the core. The objective of this paper is to identify a spectrum of diverse approaches to thermal management that can be used develop advanced, high-performance heat removal systems to further enhance the expected performance of a 1- to 20-MW(thermal) nuclear reactor. The focus here is on concepts that can provide a passive means of heat removal and are new to nuclear reactors. Different types of passive heat removal strategies for microreactors are examined, including latent heat-transfer devices, such as various types of heat pipes, natural convection and conduction-radiation cooling, and other thermal devices, such as thermoelectrics and thermoacoustics, that can be used to provide power for auxiliary cooling. Many of these concepts have already been fielded in renewable energy systems. Concepts at different stages of technical maturity are outlined to present ideas that can push the boundaries of thermal management in present-day nuclear technology. Practical considerations relative to the integration of these concepts into nuclear systems are given. [ABSTRACT FROM AUTHOR]

New nuclear reactor designs that incorporate heat pipes are being investigated for possible near-term deployment in terrestrial applications. This study explores the use of screen-covered axially grooved sodium heat pipes and their applicability for providing heat removal for microreactors. A sodium working fluid is appropriate for microreactors operating in the 5 to 20 MW(thermal) range at approximately 650°C. HTPIPE, a legacy software code, was validated for the case of screen-covered grooves and used to perform steady-state analyses to determine the performance limits of a proposed heat pipe design. The performance limits of a sodium heat pipe with a screen-covered square grooved wick structure is compared to that of an equivalent heat pipe with an annular wick. In a horizontal orientation at an operating temperature of 650°C,the performance limits for the heat pipe with an annular wick configuration are 15% higher than for the screen-covered grooved wick. At operating temperatures below 777°C, the annular wick outperforms the screen-covered grooved wick, and at temperatures above 777°C, the screen-covered grooved wick outperforms the annular wick. However, the marginal performance gain at higher temperatures may not justify the use of heat pipes with a screen-covered grooved wick structure due to increased manufacturing costs. [ABSTRACT FROM AUTHOR]

Integrated models are being developed to represent the physics occurring within the high-level and low-activity waste melters that will be used to vitrify legacy tank waste at the Hanford site. These models couple the melt pool, cold cap, and plenum region within a single computational domain. Validation of the models is essential to ensure the reliability of the numerical predictions of the operational melters. Experimental data from laboratory- and pilot-scale tests are thus being used to inform and validate various aspects of the melter model. This paper presents a tiered approach to model validation consisting of a series of progressively more complex test cases designed to model the physics occurring in the full-scale system. A hierarchical methodology has been developed to segregate and simplify the physical phenomena affecting the multiphase flow and heat transfer within a waste glass melter. Four hierarchical levels are defined in a validation pyramid and built up in levels of increasing complexity from unit problems to subsystem cases, to pilot-scale systems, and then to the full-scale system. [ABSTRACT FROM AUTHOR]

Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and postirradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory and then later shipped to the Hot Fuel Examination Facility located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping, and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and nonfueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. From a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads. [ABSTRACT FROM AUTHOR]

The performance of the refractory lining in glass melters used for nuclear waste vitrification is critical to the melter reliability for long-term continuous operation. Monofrax® K-3, a high Cr2O3 fused cast refractory material, has been widely used to build the liners of nuclear waste glass melters in the United States. Corrosion behaviour of Monofrax® K-3 refractory has been evaluated based on crucible-scale testing, inspection of the refractory components following scaled melter testing, and inspections of the Defense Waste Processing Facility (DWPF) melter refractory after service. The literature generally consists of empirical models based on short-term testing to describe refractory corrosion dependence on glass composition. Corrosion data from tests with longer testing times, at various temperatures, in the presence of molten salts, and with different redox reactions in the plenum atmosphere exist, may be insufficient to provide accurate refractory service life estimates. Additionally, the corrosion data collected under actual and scaled melter operating conditions are limited. Recommendations to achieve more direct correlation between the laboratory refractory corrosion data predictions and the observed melter service life are discussed to allow for more accurate predictions of the useful life of melter refractory linings. [ABSTRACT FROM AUTHOR]

The article offers information on the International Topical Meeting on Advances in Thermal Hydraulics, held in Reno, Nevada, from June 15-19, 2014, and also mentions several nuclear technology industry executives including Kurshad Muftuoglu, Seungjin Kim and Jong H. Kim.