Dane F Wilson

Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications

Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700°C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.Copyright

Assessment of Corrosivity Associated With Exhaust Gas Recirculation in a Heavy-Duty Diesel Engine

A high-resolution corrosion probe was placed within the airhorn section of the exhaust gas recirculation (EGR) loop of a heavy-duty diesel engine. The corrosion rate of the mild-steel probe elements was evaluated as a function of fuel sulfur level, EGR fraction, dewpoint margin, and humidity. No significant corrosion was observed while running the engine using a No. 2 grade, < 15ppm sulfur diesel fuel; however, high corrosion rates were observed on the probe elements when operating the engine using a standard grade No. 2 diesel fuel (~350 ppm sulfur) while condensing water in the EGR loop. The rate of corrosion on the mild steel elements was found to increase with increasing levels of sulfate in the condensate. However, the engine conditions influencing the sulfate level were not clearly identified in this study.

Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

Since 2002, the Department of Energys (DOEs) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOEs Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOEs structural materials research activities being conducted to support VHTR development. By far, the largest portion of materials R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water reactor steels for anticipated VHTR off-normal conditions must be determined, as well as the effects of aging on tensile, creep, and toughness properties, and on thermal emissivity. (b) Large-scale fabrication process for higher temperature alloys, such as 9Cr-1MoV, including ensuring thick-section and weldment integrity must be developed, as well as improved definitions of creep-fatigue and negligible creep behavior. (5) High-Temperature Alloys: (a) Qualification and codification of materials for the intermediate heat exchanger, such as Alloys 617 or 230, for long-term very high-temperature creep, creep-fatigue, and environmental aging degradation must be done, especially in thin sections for compact designs, for both base metal and weldments. (b) Constitutive models and an improved methodology for high-temperature design must be developed.

Embedded Sensors and Controls to Improve Component Performance and Reliability Conceptual Design Report

The overall project objective is to demonstrate improved reliability and increased performance made possible by deeply embedding instrumentation and controls (IC adequate performance was obtained through over-design. This report describes the progress and status of the project and provides a conceptual design overview for the embedded I&C pump.

Potential effects of gallium on cladding materials

This paper identifies and examines issues concerning the incorporation of gallium in weapons derived plutonium in light water reactor (LWR) MOX fuels. Particular attention is given to the more likely effects of the gallium on the behavior of the cladding material. The chemistry of weapons grade (WG) MOX, including possible consequences of gallium within plutonium agglomerates, was assessed. Based on the calculated oxidation potentials of MOX fuel, the effect that gallium may have on reactions involving fission products and possible impact on cladding performance were postulated. Gallium transport mechanisms are discussed. With an understanding of oxidation potentials and assumptions of mechanisms for gallium transport, possible effects of gallium on corrosion of cladding were evaluated. Potential and unresolved issues and suggested research and development (R and D) required to provide missing information are presented.

Experimental Study of DRACS Thermal Performance in a Low-Temperature Test Facility

Abstract The Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride salt–cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS heat exchanger (DHX) and the natural draft heat exchanger (NDHX). To experimentally investigate the thermal performance of the DRACS, a scaled-down low-temperature DRACS test facility (LTDF) has been constructed. The design of the LTDF is obtained through a detailed scaling analysis based on a 200-kW prototypic DRACS design developed at The Ohio State University. The LTDF has a nominal power capacity of 6 kW. It employs water pressurized at 1.0 MPa as the primary coolant, water near the atmospheric pressure as the secondary coolant, and ambient air as the ultimate heat sink. Three accident scenarios simulated in the LTDF are discussed in this paper. In the first scenario, startup of the DRACS system from a cold state is simulated with no initial primary coolant flow. In the second scenario, a reactor coolant pump trip process is studied, during which a flow reversal phenomenon in the DRACS primary loop occurs. In the third scenario, the pump trip process is studied with a simulated intermediate heat exchanger in operation during the simulated core normal operation. In all scenarios, natural circulation flows are developed as the transients approach their quasi steady states, demonstrating the functionality of the DRACS. The accident scenarios in the prototypic FHR design corresponding to the simulated ones in the LTDF are also predicted by following a scaling-up process. The predictions show that at any time during the simulated transient, the salt temperatures will be higher than their melting temperatures and that therefore there will be no issue of salt freezing in the three projected accident scenarios. However, the scaled-up primary salt temperatures indicate that the prototypic DHX may have been undersized and may need to be redesigned.

A study of the early stages of erosion of 1100 aluminum using a mechanical properties microprobe

Abstract The incubation period during erosion of 1100 Al has been studied by monitoring the changes occurring in the immediate subsurface layers using a mechanical properties microprobe. The results show that the very near surface regions reach maximum hardness well before steady state, concurrent with the development of a characteristic surface ripple structure. This is interpreted as support for models of material removal in steady state based on some form of critical strain criteria rather than a fracture flow stress. It is also observed that material under valleys in the ripple structure is hardened to shallower depths. The significance of this result is not yet clear but appears to be linked to the development of the ripple structure.

Design of Fluidic Diode for a High-Temperature DRACS Test Facility

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive heat removal system proposed for the Advanced High-Temperature Reactor (AHTR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops relying completely on buoyancy as the driving force. A fluidic diode has been proposed in the DRACS primary loop to maintain the passive feature. Fluidic diodes are passive flow control devices with low flow resistance in one direction and high flow resistance in the opposite direction. The fluidic diode is orientated such that during reactor normal operation the primary salt flow in the DRACS is restricted, thus preventing excessive heat loss from the reactor to the DRACS. However, when the DRACS is functioning during reactor accidents, the primary salt flow is in the forward flow direction of the diode that features low flow resistance.To investigate the reliability and thermal performance of the DRACS, a high-temperature DRACS test facility (HTDF) is being designed and constructed at The Ohio State University (OSU). In this HTDF, a conventional vortex diode has been proposed. In this paper, a detailed design process of the vortex diode for the HTDF is presented. Design parameters, such as the desired flow rates in and pressure drops across the fluidic diode, were first determined for both the forward and reverse flow directions, following which was the parametric CFD study of multiple vortex diodes with variant nozzle size, chamber size, and inlet flow rates. Flow structures inside the diode, and the effects of the nozzle size, chamber size, and Reynolds number on the Euler number were examined for both flow directions. Correlations of the forward and reverse Euler numbers and the diodicity were developed and used to develop a vortex diode design that would be applicable to the HTDF.Copyright

Evaluation of Manufacturability of Embedded Sensors and Controls with Canned Rotor Pump System

This report documents the current status of fabrication and assembly planning for the magnetic bearing, canned rotor pump being used as a demonstration platform for deeply integrating I&C into nuclear power plant components. The report identifies material choices and fabrication sequences for all of the required parts and the issues that need to be either resolved or accommodated during the manufacturing process. Down selection between material options has not yet been performed. Potential suppliers for all of the necessary materials have also been identified. The assembly evaluation begins by logically subdividing the pump into modules, which are themselves decomposed into individual parts. Potential materials and fabrication processes for each part in turn are then evaluated. The evaluation process includes assessment of the environmental compatibility requirements and the tolerances available for the selected fabrication processes. A description of the pump power/control electronics is also provided. The report also includes exploded views of the modules that show the integration of the various parts into modules that are then assembled to form the pump. Emphasis has been placed on thermal environment compatibility and the part dimensional changes during heat-up. No insurmountable fabrication or assembly challenges have been identified.