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Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

Description: This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.
Date: October 3, 2005
Creator: Yoder, G.L.
Partner: UNT Libraries Government Documents Department

Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options

Description: The Nuclear Waste Policy Act requires the Secretary of Energy to inform Congress before 2010 on the need for a second geologic repository for spent nuclear fuel. By that time, the spent fuel discharged from current commercial reactors will exceed the statutory limit of the first repository. There are several approaches to eliminate the need for another repository in this century. This paper presents a high-level analysis of these spent fuel management options in the context of a full range of possible nuclear energy futures. The analysis indicates the best option to implement varies depending on the nuclear energy future selected.
Date: October 3, 2004
Creator: Dixon, B.W. & Piet, S.J.
Partner: UNT Libraries Government Documents Department

Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

Description: The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; ...
Date: October 3, 2004
Creator: Shropshire, D.E. & Herring, J.S.
Partner: UNT Libraries Government Documents Department

A Review of Thorium Utilization as an option for Advanced Fuel Cycle--Potential Option for Brazil in the Future

Description: Since the beginning of Nuclear Energy Development, Thorium was considered as a potential fuel, mainly due to the potential to produce fissile uranium 233. Several Th/U fuel cycles, using thermal and fast reactors were proposed, such as the Radkwoski once through fuel cycle for PWR and VVER, the thorium fuel cycles for CANDU Reactors, the utilization in Molten Salt Reactors, the utilization of thorium in thermal (AHWR), and fast reactors (FBTR) in India, and more recently in innovative reactors, mainly Accelerator Driven System, in a double strata fuel cycle. All these concepts besides the increase in natural nuclear resources are justified by non proliferation issues (plutonium constrain) and the waste radiological toxicity reduction. The paper intended to summarize these developments, with an emphasis in the Th/U double strata fuel cycle using ADS. Brazil has one of the biggest natural reserves of thorium, estimated in 1.2 millions of tons of ThO{sub 2}, as will be reviewed in this paper, and therefore R&D programs would be of strategically national interest. In fact, in the past there was some projects to utilize Thorium in Reactors, as the ''Instinto/Toruna'' Project, in cooperation with France, to utilize Thorium in Pressurized Heavy Water Reactor, in the mid of sixties to mid of seventies, and the thorium utilization in PWR, in cooperation with German, from 1979-1988. The paper will review these initiatives in Brazil, and will propose to continue in Brazil activities related with Th/U fuel cycle.
Date: October 3, 2004
Creator: Maiorino, J.R. & Carluccio, T.
Partner: UNT Libraries Government Documents Department

Answering Key Fuel Cycle Questions

Description: Given the range of fuel cycle goals and criteria, and the wide range of fuel cycle options, how can the set of options eventually be narrowed in a transparent and justifiable fashion? It is impractical to develop all options. We suggest an approach that starts by considering a range of goals for the Advanced Fuel Cycle Initiative (AFCI) and then posits seven questions, such as whether Cs and Sr isotopes should be separated from spent fuel and, if so, what should be done with them. For each question, we consider which of the goals may be relevant to eventually providing answers. The AFCI program has both ''outcome'' and ''process'' goals because it must address both waste already accumulating as well as completing the fuel cycle in connection with advanced nuclear power plant concepts. The outcome objectives are waste geologic repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety. The process objectives are rea diness to proceed and adaptability and robustness in the face of uncertainties.
Date: October 3, 2004
Creator: Piet, S. J.; Dixon, B. W.; Bennett, R. G.; Smith, J. D. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

RECENT DEVELOPMENTS OF THE IRIS PROJECT OF INTEREST FOR LATIN AMERICA

Description: The IRIS (International Reactor Innovative and Secure) reactor design is being developed by an international consortium of 21 organizations from ten countries, including three members from Brazil and one from Mexico. This reflects the interest that Latin America has for a project which addresses the energy needs of the region. Presented here are some of the most recent developments in the IRIS project. The project's highest priority is the current pre-application licensing with the US NRC, which has required an investigation of the major accident sequences and a preliminary probabilistic risk assessment (PRA). The results of the accident analyses confirmed the outstanding inherent safety of the IRIS configuration and the PRA analyses indicated a core damage frequency due to internal events of the order of 2E-8. This not only highlights the enhanced safety characteristic of IRIS which should enhance its public acceptance, but it has also prompted IRIS to consider the possibility of being licensed without the need for off-site emergency response planning which would have a very positive economic implication. The modular IRIS, with each module rated at {approx} 335 MWe, is of course an ideal size for developing countries as it allows to easily introduce a moderate amount of power on limited electric grids. IRIS can be deployed in single modules in regions only requiring a few hundred MWs or in multiple modules deployed successively at time intervals in large urban areas requiring a larger amount of power increasing with time. IRIS is designed to operate ''hands-off'' as much as possible, with a small crew, having in mind deployment in areas with limited infrastructure. Thus IRIS has a 48-months maintenance interval, long refueling cycles in excess of three years, and is designed to increase as much as possible operational reliability. For example, the project has recently adopted internal ...
Date: October 3, 2004
Creator: Carelli, M.D. & Petrovic, B.
Partner: UNT Libraries Government Documents Department

Overview of the INPRO Project

Description: During the last fifty years remarkable results are achieved in the application of nuclear technology for the production of electricity. Looking ahead to the next fifty years it is clear that the demand for energy will grow considerably and also the requirements for the way the energy will be supplied. Within the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), the future of the energy demand and supply was explored and several scenario's identified. A leading requirement for energy supply is coming up and will play a crucial role: sustainability of the way the energy supply will be realized. Fulfilling the growing need for energy in developing countries is as well an important issue. Based on these scenario's for the next fifty years, an inventory of requirements for the future of nuclear energy systems has been collected as well a methodology developed by INPRO to assess innovative nuclear systems and fuel cycles. On the base of this assessment, the need for innovations and breakthroughs in existing technology can be defined. To facilitate the deployment of innovative nuclear systems also the infrastructure, technical as well as institutional has to be adjusted to the anticipated changes in the world such as the globalization. As a contribution to the conference the main messages of INPRO will be presented.
Date: October 3, 2004
Creator: Kupitz, J.; Depisch, F. & Zou, Y.
Partner: UNT Libraries Government Documents Department

On-Going Comparison of Advanced Fuel Cycle Options

Description: This paper summarizes the current comprehensive comparison of four major fuel cycle strategies: once-through, thermal recycle, thermal+fast recycle, fast recycle. It then proceeds to summarize comparison of the major technology options for the key elements of the fuel cycle that can implement each of the four strategies - separation processing, transmutation reactors, and fuels.
Date: October 3, 2004
Creator: Piet, S. J.; Bennett, R. G.; Dixon, B. W.; Herring, J. S.; Shropshire, D. E.; Roth, M. et al.
Partner: UNT Libraries Government Documents Department

IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

Description: The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.
Date: October 3, 2004
Creator: Alonso, G.; Ramirez, R.; Gomez, C. & Viais, J.
Partner: UNT Libraries Government Documents Department

The Optimum Plutonium Inert Matrix Fuel Form for Reactor-Based Plutonium Disposition

Description: The University of Florida has underway an ongoing research program to validate the economic, operational and performance benefits of developing an inert matrix fuel (IMF) for the disposition of the U.S. weapons plutonium (Pu) and for the recycle of reprocessed Pu. The current fuel form of choice for Pu disposition for the Department of Energy is as a mixed oxide (MOX) (PuO2/UO2). We will show analyses that demonstrate that a Silicon Carbide (SiC) IMF offers improved performance capabilities as a fuel form for Pu recycle and disposition. The reason that UF is reviewing various materials to serve as an inert matrix fuel is that an IMF fuel form can offer greatly reduced Pu and transuranic isotope (TRU) production and also improved thermal performance characteristics. Our studies showed that the Pu content is reduced by an order of magnitude while centerline fuel temperatures are reduced approximately 380 degrees centigrade compared to MOX. These reduced temperatures result in reduced stored heat and thermal stresses in the pellet. The reduced stored heat reduces the consequences of the loss of coolant accident, while the reduced temperatures and thermal stresses yield greatly improved fuel performance. Silicon Carbide is not new to the nuclear industry, being a basic fuel material in gas cooled reactors.
Date: October 3, 2004
Creator: Tulenko, J.S.; Wang, J. & Acosta, C.
Partner: UNT Libraries Government Documents Department

Review of the margins for ASME code fatigue design curve - effects of surface roughness and material variability.

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. The Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of the existing fatigue {var_epsilon}-N data for carbon and low-alloy steels and wrought and cast austenitic SSs to define the effects of key material, loading, and environmental parameters on the fatigue lives of the steels. Experimental data are presented on the effects of surface roughness on the fatigue life of these steels in air and LWR environments. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue evaluations. A critical review of the margins for ASME Code fatigue design curves is presented.
Date: October 3, 2003
Creator: Chopra, O. K.; Shack, W. J. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Primary system temperature limits and transient mitigation

Description: Results of a study to determine the limiting temperature conditions in a large reactor system are presented. The study considers a sodium-cooled breeder reactor system having a loop-type primary system configuration. A temperature range of 930 to 1050/sup 0/F in reactor outlet temperature is covered. Significant findings were that the use of the materials for the 930/sup 0/F reference design, i.e., a core material of 20% cold-worked 316 stainless steel, a primary piping material of 316SS, and a steam generator material of unstabilized 2-1/4 Cr - 1 Mo resulted in limiting conditions in component performance at the higher temperatures. Means to circumvent these limits through the use of alternate materials, mitigation of thermal transients, and/or design changes are presented. The economic incentive to make some materials changes is also presented.
Date: October 3, 1978
Creator: Drucker, G.S. & Bost, D.S.
Partner: UNT Libraries Government Documents Department

TS-1 and TS-2 transient overpower tests on FFTF fuel

Description: The TS-1 and TS-2 TREAT transient experiments subjected a low burnup (2 MWd/kg) and a medium burnup (58 MWd/kg), respectively, FFTF irradiated fuel pin to unprotected 5 cents/s overpower transient conditions. The fuel pin failure response was similar in the two tests, which demonstrated a large margin to failure (P/P/sub 0/ > 3) and a favorable upper level failure location. Thus, for these transient conditions, burnup effects on transient performance appeared to be minimal in the range tested. Pin disruption in the medium burnup TS-2 test was more severe due to the higher fission gas pressurization, but failure occurred at only a 5% lower power level than for the low burnup TS-1 fuel pin. Both tests exhibited axial extrusion of molten fuel to the region above the fuel column several seconds before pin failure, demonstrating a potentially beneficial inherent safety mechanism to delay failure and mitigate accident consequences.
Date: October 3, 1985
Creator: Pitner, A.L.; Ferrell, P.C.; Culley, G.E. & Weber, E.T.
Partner: UNT Libraries Government Documents Department