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Report on generation IV technical working group 3 : liquid metal reactors.

Description: This paper reports on the first round of R&D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process.
Date: March 15, 2002
Creator: Lineberry, M. J.; Rosen, S. L. & Sagayama, Y.
Partner: UNT Libraries Government Documents Department

GDH pipe break transient analysis of the RBMK - 1500.

Description: Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. The cases of GDH impact on an adjacent GDH and its attached piping are investigated in this paper. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.
Date: May 15, 2002
Creator: Kulak , R.; Marcherta, A. & Dundulis, G.
Partner: UNT Libraries Government Documents Department

Advanced Pressurized Water Reactor for Improved Resource Utilization: Part I - Survey of Potential Improvements

Description: This document is an interim report under ACDA BOA AC9NX707, Task Order 80-03, which covers the evaluation of certain potential improvements in pressurized water reactor designs intended to enhance uranium fuel utilization. The objective of these evaluations is to seek advanced, non-retrofittable improvements that could possibly be commercialized by the end of the century, and, on the basis of a preliminary evaluation, to select compatible improvements for incorporation into a composite advanced pressurized water reactor concept. The principal areas of investigation include reduced parasitic absorption of neutrons (Task 1), reduced neutron leakage (Task 2), and alternative fuel design concepts (Task 3). To the extent possible, the advanced concept developed in an earlier study (Retrofittable Modifications to Pressurized Water Reactors for Improved Resource Utilization, SSA-128, October 1980) is used as a basis in developing the advanced composite concept. The reference design considered typical of present PWR commercial practice is the system described in RESAR-414, Reference Safety Analysis Report, Westinghouse Nuclear Energy Systems, October 1976.
Date: September 15, 1981
Creator: Turner, S.E.; Gurley, M.K.; Kirby, K.D. & Mitchell, W. III
Partner: UNT Libraries Government Documents Department

Advanced pressurized water reactor for improved resource utilization, part II - composite advanced PWR concept

Description: This report evaluates the enhanced resource utilization in an advanced pressurized water reactor (PWR) concept using a composite of selected improvements identified in a companion study. The selected improvements were in the areas of reduced loss of neutrons to control poisons, reduced loss of neutrons in leakage from the core, and improved blanket/reflector concepts. These improvements were incorporated into a single composite advanced PWR. A preliminary assessment of resource requirements and costs and impact on safety are presented.
Date: September 15, 1981
Creator: Turner, S.E.; Gurley, M.K.; Kirby, K.D. & Mitchell, W III
Partner: UNT Libraries Government Documents Department

Radionuclides in the Arctic seas from the former Soviet Union: Potential health and ecological risks

Description: The primary goal of the assessment reported here is to evaluate the health and environmental threat to coastal Alaska posed by radioactive-waste dumping in the Arctic and Northwest Pacific Oceans by the FSU. In particular, the FSU discarded 16 nuclear reactors from submarines and an icebreaker in the Kara Sea near the island of Novaya Zemlya, of which 6 contained spent nuclear fuel (SNF); disposed of liquid and solid wastes in the Sea of Japan; lost a {sup 90}Sr-powered radioisotope thermoelectric generator at sea in the Sea of Okhotsk; and disposed of liquid wastes at several sites in the Pacific Ocean, east of the Kamchatka Peninsula. In addition to these known sources in the oceans, the RAIG evaluated FSU waste-disposal practices at inland weapons-development sites that have contaminated major rivers flowing into the Arctic Ocean. The RAIG evaluated these sources for the potential for release to the environment, transport, and impact to Alaskan ecosystems and peoples through a variety of scenarios, including a worst-case total instantaneous and simultaneous release of the sources under investigation. The risk-assessment process described in this report is applicable to and can be used by other circumpolar countries, with the addition of information about specific ecosystems and human life-styles. They can use the ANWAP risk-assessment framework and approach used by ONR to establish potential doses for Alaska, but add their own specific data sets about human and ecological factors. The ANWAP risk assessment addresses the following Russian wastes, media, and receptors: dumped nuclear submarines and icebreaker in Kara Sea--marine pathways; solid reactor parts in Sea of Japan and Pacific Ocean--marine pathways; thermoelectric generator in Sea of Okhotsk--marine pathways; current known aqueous wastes in Mayak reservoirs and Asanov Marshes--riverine to marine pathways; and Alaska as receptor. For these waste and source terms addressed, other pathways, such as ...
Date: November 15, 1999
Creator: Layton, D W; Edson, R; Varela, M & Napier, B
Partner: UNT Libraries Government Documents Department

NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''

Description: The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.
Date: March 15, 2002
Creator: Mynatt, Fred R.; Kadak, Andy; Berte, Marc; Miller, Larry; Khan, Mohammed; McConn, Joe et al.
Partner: UNT Libraries Government Documents Department

Evaluating the Effects of Aging on Electronic Instrument and Control Circuit Boards and Components in Nuclear Power Plants

Description: The report describes potentially useful techniques for monitoring the aging of I&C boards. The techniques have been grouped into: periodic testing, reliability modeling, resistance measures, signal comparison, eternal measures, and internal measures, each representing distinct theoretical approaches to detection and evaluation.
Date: May 15, 2005
Creator: Hannaman, G. William & Wilkinson, C. Dan
Partner: UNT Libraries Government Documents Department

Initial Assessment of Environmental Barrier Coatings for the Prometheus Project

Description: Depending upon final design and materials selections, a variety of engineering solutions may need to be considered to avoid chemical degradation of components in a notional space nuclear power plant (SNPP). Coatings are one engineered approach that was considered. A comprehensive review of protective coating technology for various space-reactor structural materials is presented, including refractory metal alloys [molybdenum (Mo), tungsten (W), rhenium (Re), tantalum (Ta), and niobium (Nb)], nickel (Ni)-base superalloys, and silicon carbide (Sic). A summary description of some common deposition techniques is included. A literature survey identified coatings based on silicides or iridium/rhenium as the primary methods for environmental protection of refractory metal alloys. Modified aluminide coatings have been identified for superalloys and multilayer ceramic coatings for protection of Sic. All reviewed research focused on protecting structural materials from extreme temperatures in highly oxidizing conditions. Thermodynamic analyses indicate that some of these coatings may not be protective in the high-temperature, impure-He environment expected in a Prometheus reactor system. Further research is proposed to determine extensibility of these coating materials to less-oxidizing or neutral environments.
Date: December 15, 2005
Creator: Frederick, M.
Partner: UNT Libraries Government Documents Department

Advanced Safeguards Approaches for New Fast Reactors

Description: This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.
Date: December 15, 2007
Creator: Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L. et al.
Partner: UNT Libraries Government Documents Department

Advanced Safeguards Approaches for New TRU Fuel Fabrication Facilities

Description: This second report in a series of three reviews possible safeguards approaches for the new transuranic (TRU) fuel fabrication processes to be deployed at AFCF – specifically, the ceramic TRU (MOX) fuel fabrication line and the metallic (pyroprocessing) line. The most common TRU fuel has been fuel composed of mixed plutonium and uranium dioxide, referred to as “MOX”. However, under the Advanced Fuel Cycle projects custom-made fuels with higher contents of neptunium, americium, and curium may also be produced to evaluate if these “minor actinides” can be effectively burned and transmuted through irradiation in the ABR. A third and final report in this series will evaluate and review the advanced safeguards approach options for the ABR. In reviewing and developing the advanced safeguards approach for the new TRU fuel fabrication processes envisioned for AFCF, the existing international (IAEA) safeguards approach at the Plutonium Fuel Production Facility (PFPF) and the conceptual approach planned for the new J-MOX facility in Japan have been considered as a starting point of reference. The pyro-metallurgical reprocessing and fuel fabrication process at EBR-II near Idaho Falls also provided insight for safeguarding the additional metallic pyroprocessing fuel fabrication line planned for AFCF.
Date: December 15, 2007
Creator: Durst, Philip C.; Ehinger, Michael H.; Boyer, Brian; Therios, Ike; Bean, Robert; Dougan, A. et al.
Partner: UNT Libraries Government Documents Department


Description: This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to describe the latest version of the probabilistic criticality analysis methodology and its application to the entire commercial waste stream of commercial pressurized water reactor (PWR) spent nuclear fuel (SNF) expected to be emplaced in the repository. The purpose of this particular application is to evaluate the 21 assembly PWR absorber plate waste package (WP) with respect to degraded mode criticality performance. The degradation of principal concern is the borated stainless steel absorber plates which are part of the waste package basket and which constitute a major part of the waste package criticality control. The degradation (corrosion, dissolution) of this material will result in the release of most of the boron from the waste package and increase the possibility of criticality. The results of this evaluation will be expressed in terms of the fraction of the PWR SNF which can exceed a given k{sub eff}, as a function of time and the peak value of that fraction over a time period up to several hundred thousand years. The ultimate purpose of this analysis is to support the waste package design which defines waste packages to cover a range of SNF characteristics. In particular, with respect to PWR criticality the current categories are: (1) no specific criticality control material, (2) borated stainless steel plates in the waste package basket, and (3) zirconium clad boron carbide control rods (Ref. 5.4). The results of this analysis will indicate the coverage provided by the first two categories. With these results, this study will provide the first quantitative estimate of the benefit expected from the control measure consisting of borated stainless steel plates. This document is the third waste package probabilistic criticality analysis. The first two (Ref. 5.12 for ...
Date: September 15, 1997
Creator: Goulib, P.
Partner: UNT Libraries Government Documents Department

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

Description: This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.
Date: February 15, 2006
Creator: Lane, MH
Partner: UNT Libraries Government Documents Department

Initial Screening of Thermochemical Water-Splitting Cycles for High Efficiency Generation of Hydrogen Fuels Using Nuclear Power

Description: OAK B188 Initial Screening of Thermochemical Water-Splitting Cycles for High Efficiency Generation of Hydrogen Fuels Using Nuclear Power There is currently no large scale, cost-effective, environmentally attractive hydrogen production process, nor is such a process available for commercialization. Hydrogen is a promising energy carrier, which potentially could replace the fossil fuels used in the transportation sector of our economy. Fossil fuels are polluting and carbon dioxide emissions from their combustion are thought to be responsible for global warming. The purpose of this work is to determine the potential for efficient, cost-effective, large-scale production of hydrogen utilizing high temperature heat from an advanced nuclear power station. Almost 800 literature references were located which pertain to thermochemical production of hydrogen from water and over 100 thermochemical watersplitting cycles were examined. Using defined criteria and quantifiable metrics, 25 cycles have been selected for more detailed study.
Date: December 15, 1999
Creator: Brown, L. C.; Funk, J. F. & Showalter, S. K.
Partner: UNT Libraries Government Documents Department

Utilizing benchmark data from the ANL-ZPR diagnostic cores program

Description: The support of the criticality safety community is allowing the production of benchmark descriptions of several assemblies from the ZPR Diagnostic Cores Program. The assemblies have high sensitivities to nuclear data for a few isotopes. This can highlight limitations in nuclear data for selected nuclides or in standard methods used to treat these data. The present work extends the use of the simplified model of the U9 benchmark assembly beyond the validation of k{sub eff}. Further simplifications have been made to produce a data testing benchmark in the style of the standard CSEWG benchmark specifications. Calculations for this data testing benchmark are compared to results obtained with more detailed models and methods to determine their biases. These biases or corrections factors can then be applied in the use of the less refined methods and models. Data testing results using Versions IV, V, and VI of the ENDF/B nuclear data are presented for k{sub eff}, f{sup 28}/f{sup 25}, c{sup 28}/f{sup 25}, and {beta}{sub eff}. These limited results demonstrate the importance of studying other integral parameters in addition to k{sub eff} in trying to improve nuclear data and methods and the importance of accounting for methods and/or modeling biases when using data testing results to infer the quality of the nuclear data files.
Date: February 15, 2000
Creator: Schaefer, R. W. & McKnight, R. D.
Partner: UNT Libraries Government Documents Department

Methods for incorporating effects of LWR coolant environment into ASME code fatigue evaluations.

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Appendix I to Section HI of the Code specifies design fatigue curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Recent test data illustrate potentially significant effects of LWR environments on the fatigue resistance of carbon and low-alloy steels and austenitic stainless steels (SSs). Under certain loading and environmental conditions, fatigue lives of carbon and low-alloy steels can be a factor of {approx}70 lower in an LWR environment than in air. These results raise the issue of whether the design fatigue curves in Section III are appropriate for the intended purpose. This paper presents the two methods that have been proposed for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations. The mechanisms of fatigue crack initiation in carbon and low-alloy steels and austenitic SSs in LWR environments are discussed.
Date: April 15, 1999
Creator: Chopra, O. K.
Partner: UNT Libraries Government Documents Department

Morphologies of uranium deposits produced during electrorefining of EBR-II spent nuclear fuel

Description: The morphologies of U metal samples from deposits produced by electrorefining of Experimental Breeder Reactor-II (EBR-II) spent fuel were examined using scanning electron microscopy, energy- and wavelength-dispersive X-ray spectroscopy, and metallography. The morphologies were analyzed to find correlations with the chemistry of the samples, the ER run conditions, and the deposit performance. A rough correlation was observed between morphology and Zr concentration; samples with Zr contents greater than approximately 200 ppm showed fine-grained, polycrystalline dendritic morphologies, while samples with Zr contents less than approximately 100 ppm were comprised of agglomerations or linked chains of rhomboidal single crystals. There were few correlations found between morphology, run conditions, and deposit performance.
Date: February 15, 2000
Creator: Totemeier, T. C.
Partner: UNT Libraries Government Documents Department


Description: A novel concept to detect pin-diversion from spent fuel assembly is proposed and described. The instrument will use multiple tiny neutron and gamma detectors in a form of cluster (detector cluster) and high precision driving system to collect radiation signatures inside pressurized water reactor (PWR) assembly. In order to validate our concept, a Monte Carlo study was done using a Monte Carlo code MCNP5. MONTEBURNS, a computational tool that links MCNP and ORIGEN, was used to produce accurate PWR spent fuel isotopic compositions. Monte Carlo simulations, using realistic fuel geometry and actual fuel material information, were performed to study radiation field inside a PWR spent fuel assembly. The preliminary Monte Carlo simulation study shows that indeed 2 dimensional neutron data, when obtained in the presence of missing pins, have data profiles distinctly different from the profiles obtained without missing pins.
Date: June 15, 2006
Creator: Ham, Y S; Maldonado, G I; Yin, C & Burdo, J
Partner: UNT Libraries Government Documents Department

Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

Description: A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of ...
Date: December 15, 2005
Creator: Ingersoll, DT
Partner: UNT Libraries Government Documents Department


Description: The United States (U.S.) has identified 61.5 metric tons (MT) of plutonium that is permanently excess to use in nuclear weapons programs, including 47.2 MT of weapons-grade plutonium. Except for materials that remain in use for programs outside of national defense, including programs for nuclear-energy development, the surplus inventories will be stored safely by the Department of Energy (DOE) and then transferred to facilities that will prepare the plutonium for permanent disposition. Some items will be disposed as transuranic waste, low-level waste, or spent fuel. The remaining surplus plutonium will be managed through: (1) the Mixed Oxide (MOX) Fuel Fabrication Facility (FFF), to be constructed at the Savannah River Site (SRS), where the plutonium will be converted to fuel that will be irradiated in civilian power reactors and later disposed to a high-level waste (HLW) repository as spent fuel; (2) the SRS H-Area facilities, by dissolving and transfer to HLW systems, also for disposal to the repository; or (3) alternative immobilization techniques that would provide durable and secure disposal. From the beginning of the U.S. program for surplus plutonium disposition, DOE has sponsored research to characterize the surplus materials and to judge their suitability for planned disposition options. Because many of the items are stored without extensive analyses of their current chemical content, the characterization involves three interacting components: laboratory sample analysis, if available; non-destructive assay data; and rigorous evaluation of records for the processing history for items and inventory groups. This information is collected from subject-matter experts at inventory sites and from materials stabilization and surveillance programs, in cooperation with the design agencies for the disposition facilities. This report describes the operation and status of the characterization program.
Date: July 15, 2008
Creator: Allender, J; Edwin Moore, E & Scott Davies, S
Partner: UNT Libraries Government Documents Department


Description: The experiments conducted this past quarter have suggested that the bacterium Pseudomonas fluorescens strain CL0145A is effective at killing zebra mussels throughout the entire range of pH values tested (7.2 to 8.6). Highest mortality was achieved at pH values characteristic of preferred zebra mussel waterbodies, i.e., hard waters with a range of 7.8 to 8.6. In all water types tested, however, ranging from very soft to very hard, considerable mussel kill was achieved (83 to 99% mean mortality), suggesting that regardless of the pH or hardness of the treated water, significant mussel kill can be achieved upon treatment with P. fluorescens strain CL0145A. These results further support the concept that this bacterium has significant potential for use as a zebra mussel control agent in power plant pipes receiving waters with a wide range of physical and chemical characteristics.
Date: October 15, 2002
Creator: Molloy, Daniel P.
Partner: UNT Libraries Government Documents Department

Intact and Degraded Criticality Calculations for the Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package

Description: The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the U.S. Department of Energy's (DOE) Shippingport Light Water Breeder Reactor (SP LWBR) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP), which is to be placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (K{sub eff}) for intact- and degraded-mode internal configurations of the codisposal WP containing Shippingport LWBR seed-type assemblies. The results of this calculation will be used to evaluate criticality issues and support the analysis that is planed to be performed to demonstrate the viability of the codisposal concept for the MGR. This calculation is associated with the waste package design and was performed in accordance with the DOE SNF Analysis Plan for FY 2000 (See Ref. 22). The document has been prepared in accordance with the Administrative Procedure AP-3.12Q, Calculations (Ref. 23).
Date: September 15, 2000
Creator: Montierth, L.M.
Partner: UNT Libraries Government Documents Department

India's Worsening Uranium Shortage

Description: As a result of NSG restrictions, India cannot import the natural uranium required to fuel its Pressurized Heavy Water Reactors (PHWRs); consequently, it is forced to rely on the expediency of domestic uranium production. However, domestic production from mines and byproduct sources has not kept pace with demand from commercial reactors. This shortage has been officially confirmed by the Indian Planning Commission’s Mid-Term Appraisal of the country’s current Five Year Plan. The report stresses that as a result of the uranium shortage, Indian PHWR load factors have been continually decreasing. The Uranium Corporation of India Ltd (UCIL) operates a number of underground mines in the Singhbhum Shear Zone of Jharkhand, and it is all processed at a single mill in Jaduguda. UCIL is attempting to aggrandize operations by establishing new mines and mills in other states, but the requisite permit-gathering and development time will defer production until at least 2009. A significant portion of India’s uranium comes from byproduct sources, but a number of these are derived from accumulated stores that are nearing exhaustion. A current maximum estimate of indigenous uranium production is 430t/yr (230t from mines and 200t from byproduct sources); whereas, the current uranium requirement for Indian PHWRs is 455t/yr (depending on plant capacity factor). This deficit is exacerbated by the additional requirements of the Indian weapons program. Present power generation capacity of Indian nuclear plants is 4350 MWe. The power generation target set by the Indian Department of Atomic Energy (DAE) is 20,000 MWe by the year 2020. It is expected that around half of this total will be provided by PHWRs using indigenously supplied uranium with the bulk of the remainder provided by breeder reactors or pressurized water reactors using imported low-enriched uranium.
Date: January 15, 2007
Creator: Curtis, Michael M.
Partner: UNT Libraries Government Documents Department


Description: The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to explain the design of ceramic host systems culminating in a reference ceramic formulation for use in subsequent studies on process optimization and melt property data assessment in support of FY13 melter demonstration testing. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. In addition to the combined CS/LN/TM High Mo waste stream, variants without Mo and without Mo and Zr were also evaluated. Based on the results of fabricating and characterizing several simulated ceramic waste forms, two reference ceramic waste form compositions are recommended in this report. The first composition targets the CS/LN/TM combined waste stream with and without Mo. The second composition targets with CS/LN/TM combined waste stream with ...
Date: May 15, 2012
Creator: Brinkman, K.; Fox, K. & Marra, J.
Partner: UNT Libraries Government Documents Department