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Wigeland, R.

Publications and source records attributed to Wigeland, R..

11 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

SAS4A/SASSYS-1 VERSION 5.2 SAFETY ANALYSIS CODE SYSTEM

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolant flow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical behavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.«

Fanning, Thomas↗
DOE OSTI2017

Fuel Cycle and Nuclear Technology Seminars for DOE/NE – Molten Salt Reactors

This report provides the information presented at the Molten Salt Reactor (MSR) Seminar held at DOE Headquarters in Germantown, MD on February 16, 2017. DOE/NE-43 had requested an informational seminar to provide a detailed technical background on the technology of molten salt reactors, the history of molten salt reactor development, the current status of the technologies, and the identified issues requiring R&D. To meet this need, the seminar had the following objectives: 1. Provide information specifically for DOE/NE Federal managers on the concepts, technologies, fuel cycles, and challenges associated with MSRs. - Provide a review of various MSR technologies - Domestic and international, historic and current - Describe the reasons behind the recent interest in MSR technologies - Potential advantages, disadvantages, and challenges 2. Provide a summary of the current status of MSRs and the major challenges to development and deployment. The MSR seminar consisted of a series of presentations mainly developed by Oak Ridge National Laboratory, reviewed and coordinated by the Fuel Cycle Options – Systems Analysis and Integration campaign to meet the seminar objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fuel cycle impacts of accident tolerant fuels

This paper discusses the impact of several accident-tolerant fuel (ATF) options on the fuel cycle performance of once-through LWRs compared to current uranium oxide-Zircaloy fueled commercial LWRs. Desirable characteristics of proposed ATF concepts include higher thermal conductivity, enhanced fission product retention, and reduced reaction kinetics and hydrogen generation in loss of coolant conditions. ATF concepts that have been proposed target changes to the fuel and/or cladding. Broadly speaking, the options can be summarized as follows: -) UO{sub 2} pellet fuels with increased thermal conductivity to lower temperatures, -) high-density pellet fuels (e.g., nitrides and/or silicides), -) fully ceramic microencapsulated (FCM) fuels, -) metallic fuel with high fissile density and high thermal conductivity, -) coatings on Zircaloy cladding, -) total elimination of the Zircaloy as cladding and replacement with a ceramic, and -) composite cladding including for instance layered Zircaloy and SiC and molybdenum. We can draw 4 conclusions. First, other than SiC, the other cladding options will result in a reactivity penalty requiring either thinner cladding or increased fissile loading. Secondly, silicide fuels have a higher density than UO{sub 2} and can allow longer cycle lengths for the same enrichment or a reduction in the enrichment for a given cycle length. Thirdly, while the pellet fuels yield very similar results, the FCM fuel is dramatically different due to the low fuel loading. Fourthly, we have to note that in order to achieve the maximum benefit from a particular fuel-clad combination, it requires consideration of the characteristics of each. For example, a cladding with a relatively high absorption (e.g., FeCrAl) would benefit from a high density fuel such as U{sub 3}Si{sub 2}.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

VCTR: A Versatile Coupled Test Reactor Concept

The United States has currently no domestic fast neutron test and irradiation capabilities necessary to pursue a strong research and development program related to fast neutron reactors. This lack prevents the testing of fast reactor fuels in a prototypic environment, the testing of new safety features and of new coolants. A new irradiation test reactor is a costly investment that will operate for several decades and the initial specifications are crucial to ensure the relevance of such a facility in the long run. This new test reactor should be highly reconfigurable and allow operation as a fast test reactor or a thermal test reactor or a cou-pled fast-thermal test reactor. This report presents the on-going effort at Idaho National Laboratory (INL) to design a versatile coupled test reactor (VCTR—pronounced Victor) that could fulfill the R&D needs requiring high fast neutron fluxes but also high thermal neutron fluxes in configurations that could not be met in current thermal neutron test reactors such as ATR. The objective of this report is to summarize the efforts undertaken at INL from May 2015 to March 2016 aimed at understanding the main options for designing and building a versatile test reactor, with a primary mission to provide thermal and fast neutrons for irradiation of fuels and materials in a manner complementary to the ATR and HFIR reactors. Other missions can also be considered, including the validation of computer codes and the development of specific reactor technologies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Economic Evaluation of Promising Options

This report presents the efforts, undertaken in FY15, aimed at enhancing the Fuel Cycle Options (FCO) Campaign’s capability to perform credible fuel cycle cost estimates and evaluation of financial risk. This effort builds on the work performed in FY14 (Ganda 2014), which focused extensively on understanding the principal drivers of Light Water Reactors (LWR) construction costs, based on historical data. That framework has received further confirmation from the work and the extensive data uncovered in FY15.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Investigation of Benefits from U/TRU Recycle - Quantification and Comparison to U/Pu Recycle

The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there was no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may adversely affect the ability to reduce HLW volume. The short-term radioactivity and long-term radiotoxicity of the HLW is also affected, which may be of more or less importance depending on the specific geologic disposal environment. To study these potential effects, a range of waste forms and disposal environments were used in the analysis, documenting to what extent the recycle of minor actinides in addition to plutonium may offer further benefit. Another area of investigation concerned the recycle fuel, for the fast reactor and for the thermal reactors they may support. Information to date indicates that U/Pu fuel may be simpler to fabricate and has a much more extensive database than U/TRU fuel, one of the reasons for the increased challenge for developing and deploying a U/TRU fuel cycle, and also indicates that heterogeneous recycle of the minor actinides may be even more difficult as compared to homogeneous recycle. This information was reviewed and updated to reflect the most recent studies for the purpose of informing on all aspects of the differences between U/Pu and U/TRU recycle. The results of all of these investigations will be presented to provide information on the findings concerning the value of U/TRU recycle.

Fast reactor recycle↗
DOE OSTI2015

Evaluation of waste arising from future nuclear fuel cycle - 15197

A comprehensive study was recently completed at the request of the US Department of Energy Office of Nuclear Energy (DOE-NE) to evaluate and screen nuclear fuel cycles. The final report was issued in October 2014. Uranium- and thorium-based fuel cycles were evaluated using both fast and thermal spectrum reactors. Once-through, limited-recycle, and continuous-recycle cases were considered. This study used nine evaluation criteria to identify promising fuel cycles. Nuclear waste management was one of the nine evaluation criteria. The waste generation criterion from this study is discussed herein. The fundamental characteristics of nuclear fuel cycles were used to create a comprehensive set of fuel cycle options. These options were categorized into 40 Evaluation Groups based on similarities in their physics-based performance. These Evaluation Groups, listed below, comprehensively represent all fuel cycle options: - 8 once-through Evaluation Groups: 5 uranium-based fuel cycles and 3 thorium-based fuel cycles - 10 limited-recycle Evaluation Groups: 6 uranium-based fuel cycles and 4 thorium-based fuel cycles, and - 22 continuous-recycle Evaluation Groups: 14 uranium-based fuel cycles and 8 thorium-based fuel cycles. The study focused on the quantity and characteristics of radioactive wastes generated by the different fuel cycles, examining 5 metrics characterizing waste generation: - mass of spent nuclear fuel (SNF) plus high level waste (HLW) disposed per energy generated, - activity of SNF + HLW (at 100 years) per energy generated, - activity of SNF + HLW (at 100,000 years) per energy generated, - mass of depleted uranium (DU) + recovered uranium (RU) + recovered thorium (RTh) disposed per energy generated, and - volume of low level radioactive waste (LLW) per energy generated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Nuclear Fuel Cycle Options Evaluation to Inform R&D Planning

An Evaluation and Screening (E&S) of nuclear fuel cycle options has been conducted in fulfilment of a Charter specified for the study by the U.S. Department of Energy (DOE) Office of Nuclear Energy. The E&S study used an objective and independently reviewed evaluation process to provide information about the potential benefits and challenges that could strengthen the basis and provide guidance for the research and development(R&D) activities undertaken by the DOE Fuel Cycle Technologies Program Office. Using the nine evaluation criteria specified in the Charter and associated evaluation metrics and processes developed during the E&S study, a screening was conducted of 40 nuclear fuel cycle evaluation groups to provide answers to the questions: (1) Which nuclear fuel cycle system options have the potential for substantial beneficial improvements in nuclear fuel cycle performance, and what aspects of the options make these improvements possible? (2)Which nuclear material management approaches can favorably impact the performance of fuel cycle options? (3)Where would R&D investment be needed to support the set of promising fuel cycle system options and nuclear material management approaches identified above, and what are the technical objectives of associated technologies?

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

The Indian Advanced Heavy Water Reactor (AHWR) and Non-Proliferation Attributes

This report describes the Indian Advanced Heavy Water Reactor (AHWR), focusing on the reactor design and operational features that are relevant to proliferation risk, for both the AHWR intended for domestic use and the AHWR-LEU for export. The AHWR is the key component of the final stage of the three-stage Indian Nuclear program, “Thorium Utilization for Sustainable Power Programme,” that has been in development since the 1950’s. The nuclear program in India has historically assumed a heavy reliance on the ultimate transition to a thorium-based fuel cycle due to the abundant domestic thorium resources and the scarcity of domestic uranium. Both the AHWR and AHWR-LEU use thorium-based oxide fuels, with the AHWR using both UO 2 /ThO 2 and PuO 2 /ThO 2 fuels simultaneously, while AHWR-LEU uses only UO 2 /ThO 2 fuel. The AHWRs are essentially derivatives of the advanced CANDU reactor, with the fuel oriented vertically instead of horizontally and refueling still occurring on-line, but only from the top of the reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Innovative Separations Technologies

Reprocessing used nuclear fuel (UNF) is a multi-faceted problem involving chemistry, material properties, and engineering. Technology options are available to meet a variety of processing goals. A decision about which reprocessing method is best depends significantly on the process attributes considered to be a priority. New methods of reprocessing that could provide advantages over the aqueous Plutonium Uranium Reduction Extraction (PUREX) and Uranium Extraction + (UREX+) processes, electrochemical, and other approaches are under investigation in the Fuel Cycle Research and Development (FCR&D) Separations Campaign. In an attempt to develop a revolutionary approach to UNF recycle that may have more favorable characteristics than existing technologies, five innovative separations projects have been initiated. These include: (1) Nitrogen Trifluoride for UNF Processing; (2) Reactive Fluoride Gas (SF6) for UNF Processing; (3) Dry Head-end Nitration Processing; (4) Chlorination Processing of UNF; and (5) Enhanced Oxidation/Chlorination Processing of UNF. This report provides a description of the proposed processes, explores how they fit into the Modified Open Cycle (MOC) and Full Recycle (FR) fuel cycles, and identifies performance differences when compared to 'reference' advanced aqueous and fluoride volatility separations cases. To be able to highlight the key changes to the reference case, general background on advanced aqueous solvent extraction, advanced oxidative processes (e.g., volumetric oxidation, or 'voloxidation,' which is high temperature reaction of oxide UNF with oxygen, or modified using other oxidizing and reducing gases), and fluorination and chlorination processes is provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗