Nicholas R. Brown

Operational Results of a Closed Brayton Cycle Test‐Loop

A number of space and terrestrial power system designs plan to use nuclear reactors that are coupled to Closed‐loop Brayton Cycle (CBC) systems to generate electrical power. Because very little experience exists regarding the operational behavior of these systems, Sandia National Laboratories (through its Laboratory Directed Research and Development program) is developing a closed‐loop test bed that can be used to determine the operational behavior of these systems and to validate models for these systems. Sandia has contracted Barber‐Nichols Corporation to design, fabricate, and assemble a Closed‐loop Brayton Cycle (CBC) system. This system was developed by modifying commercially available hardware. It uses a 30 kWe Capstone C‐30 gas‐turbine unit (www.capstoneturbine.com) with a modified housing that permits the attachment of an electrical heater and a water cooled chiller that are connected to the turbo‐machinery in a closed loop. The test‐loop reuses the Capstone turbine, compressor, and alternator. Th...

Accelerator–Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of accelerator-driven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

Thorium fuel cycles with externally driven systems

Abstract Externally driven subcritical systems are closely associated with thorium, partially because thorium has no naturally occurring fissile isotopes. Both accelerator-driven systems (ADSs) and fusion-driven systems have been proposed. This paper highlights key literature related to the use of thorium in externally driven systems (EDSs) and builds upon this foundation to identify potential roles for EDSs in thorium fuel cycles. In fuel cycles with natural thorium feed and no enrichment, the potential roles are (1) a once-through breed-and-burn fuel cycle and (2) a fissile breeder (mainly 233U) to support a fleet of critical reactors. If enriched uranium is used in the fuel cycle in addition to thorium, EDSs may be used to burn transuranic material. These fuel cycles were evaluated in the recently completed U.S. Department of Energy Evaluation and Screening of nuclear fuel cycle options relative to the current once-through commercial nuclear fuel cycle in the United States. The evaluation was performed with respect to nine specified high-level criteria, such as waste management and resource utilization. Each of these fuel cycles presents significant potential benefits per unit energy generation compared to the present once-through uranium fuel cycle. A parametric study indicates that fusion-fission–hybrid systems perform better than ADSs in some missions due to a higher neutron source relative to the energy required to produce it. However, both potential externally driven technology choices face significant development and deployment challenges. In addition, there are significant challenges associated with the use of thorium fuel and with the transition from a uranium-based fuel cycle to a thorium-based fuel cycle.

Analysis Model for Sulfur-Iodine and Hybrid Sulfur Thermochemical Cycles

Abstract This paper presents a transient control volume modeling scheme for both the sulfur-iodine (SI) and Westinghouse hybrid sulfur (HyS) thermochemical cycles. These cycles are very important candidates for the large-scale production of hydrogen in the 21st century. In this study, transient control volume models of the SI and HyS cycles are presented, along with a methodology for coupling these models to codes that describe the transient behavior of a high-temperature nuclear reactor. The transient SI and HyS cycle models presented here are based on a previous model with a significant improvement, namely, pressure variation capability in the chemical reaction chambers. This pressure variation capability is obtained using the ideal gas law, which is differentiated with respect to time. The HyS model is based on a time-dependent application of the Nernst equation. Investigation of the new pressure assumption yields a peak pressure rate of change of 5.877 kPa/s for a temperature-driven transient test matrix and 2.993 kPa/s for a mass flow rate–driven transient test matrix. These high rates of pressure change suggest that an accurate model of the SI and/or HyS cycle must include some method of accounting for pressure variation. The HyS model suggests that the hydrogen production rate is directly proportional to the SO2 production rate.

SIMULATION OF SULFUR-IODINE THERMOCHEMICAL HYDROGEN PRODUCTION PLANT COUPLED TO HIGH-TEMPERATURE HEAT SOURCE

Abstract The sulfur-iodine (SI) cycle is one of the leading candidates in thermochemical processes for hydrogen production. In this paper a simplified model for the SI cycle is developed with chemical kinetics models of the three main SI reactions: the Bunsen reaction, sulfuric acid decomposition, and hydriodic acid decomposition. Each reaction was modeled with a single control volume reaction chamber. The simplified model uses basic heat and mass balance for each of the main three reactions. For sulfuric acid decomposition and hydriodic acid decomposition, reaction heat, latent heat, and sensible heat were considered. Since the Bunsen reaction is exothermic and its overall energy contribution is small, its heat energy is neglected. However, the input and output streams from the Bunsen reaction are accounted for in balancing the total stream mass flow rates from the SI cycle. The heat transfer between the reactor coolant (in this case helium) and the chemical reaction chamber was modeled with transient energy balance equations. The steady-state and transient behavior of the coupled system is studied with the model, and the results of the study are presented. It was determined from the study that the hydriodic acid decomposition step is the rate-limiting step of the entire SI cycle.

A Summary of the Department of Energy’s Advanced Demonstration and Test Reactor Options Study

Abstract An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the U.S. Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions, and advanced reactor technology needs were identified based on recent DOE and international studies. A broad team of stakeholders from industry, academia, and government was assembled to develop a comprehensive set of goals, criteria, and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study’s evaluation criteria and metrics, an independent group of experts from industry, universities, and national laboratories scored each of the point designs. Pathways to deployment for concepts of varying technical maturities were estimated for the different demonstration systems with regard to cost, schedule, and possible licensing approaches. This study also presents the trade-offs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development and to serve potential secondary missions. The main findings of the study indicate the following: (1) for industrial process heat supply, a high-temperature gas-cooled reactor is the best choice because of the high outlet temperature of the reactor and its strong passive and inherent safety characteristics; (2) for resource utilization and waste management, a sodium-cooled fast reactor (SFR) is best because of the use of a fast flux to destroy actinides; (3) to realize the advantages of a promising but less-mature technology, a fluoride salt-cooled high-temperature reactor and a lead-cooled fast reactor fare about the same; (4) for fulfilling the needs of a materials test reactor, a SFR is considered best because of its ability to produce high fast flux, incorporate test loops, and provide additional large volumes for testing.

Effect of Shim Arm Depletion in the NBSR

The cadmium shim arms in the NBSR undergo burnup during reactor operation and hence, require periodic replacement. Presently, the shim arms are replaced after every 25 cycles to guarantee they can maintain sufficient shutdown margin. Two prior reports document the expected change in the 113Cd distribution because of the shim arm depletion. One set of calculations was for the present high-enriched uranium fuel and the other for the low-enriched uranium fuel when it was in the COMP7 configuration (7 inch fuel length vs. the present 11 inch length). The depleted 113Cd distributions calculated for these cores were applied to the current design for an equilibrium low-enriched uranium core. This report details the predicted effects, if any, of shim arm depletion on the shim arm worth, the shutdown margin, power distributions and kinetics parameters.

Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

Analysis of Reactivity Insertion Accidents for the NIST Research Reactor Before and After Fuel Conversion

Reactivity insertion accidents have been analyzed for the 20-MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains highly enriched uranium fuel, and for a proposed equilibrium core with low-enriched uranium fuel. The time-dependent analysis of the primary system is performed with a RELAP5 model that includes the reactor vessel, primary coolant pump, heat exchanger, fuel element geometry, and flow channels for both the 6 inner and 24 outer fuel elements. Postprocessing of the simulation results has been conducted to evaluate minimum critical heat flux (CHF) ratio and minimum onset of flow instability (OFI) ratio using the Sudo-Kaminaga correlations and Saha-Zuber criteria, respectively. Evaluations are carried out for the control rod withdrawal start-up accident and the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that no damage to the fuel will occur and there is adequate margin to CHF and OFI because of sufficient coolant flow through the fuel channels and the negative reactivity insertion due to scram.

The Mechanical Response Evaluation of Advanced Claddings During Proposed Reactivity Initiated Accident Conditions

This study investigates the failure mechanisms of advanced oxidation resistant FeCrAl nuclear fuel cladding at high-strain rates, similar to conditions characteristic of design basis reactivity initiated accidents (RIAs). During a postulated RIA, the nuclear fuel cladding may be subjected to complex loading which can cause multiaxial strain states ranging from plane-strain to equibiaxial tension. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses, simulating strains rates occurring in a postulated RIA. The mechanical response of the advanced claddings, in the unirradiated state with ample ductility, was compared to that of hydrided zirconium-based nuclear fuel cladding. The hoop strain evolution pulses were collected in situ; the permanent diametral strains of both accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture. Both zirconium-based alloys and FeCrAl alloys exhibited ductile behavior. FeCrAl model alloys without microstructural control and strengthening mechanism were used in this demonstration study that showed reduced diametral strain (less than 0.15) compared to the diametral strain for the unirradiated zirconium-based alloy (approximately 0.2).