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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

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

MONTE CARLO CHARACTERIZATION OF PRESSURIZED WATER REACTOR SPENT FUEL ASSEMBLY FOR THE DEVELOPMENT OF A NEW INSTRUMENT FOR PIN DIVERSION DETECTION

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

CHARACTERIZATION OF SURPLUS PLUTONIUM FOR DISPOSITION OPTIONS

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

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

CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

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

Advanced Burner Reactor Preliminary NEPA Data Study.

Description: The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because ...
Date: October 15, 2007
Creator: Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R. et al.
Partner: UNT Libraries Government Documents Department

Thermal and nuclear design review of the NRX-A5 Reactor

Description: A thermal and nuclear NRX-A5 Design Review was held at WANL on April 10, 1964. Analyses and planning performed to date on the NRX-A5 and subsequent reactors were presented. Systems analyses were performed on various "cooled periphery".and "hot periphery" seal configurations. A number of inner reflector designs were examined, as well as means for cooling the lateral support springs. Core component redesign incl,uded methods for controlling inter-element pressure distribution by undercutting or chamfering of fuel elements. Changes in tie rod diameters, use of as-extruded fuel elements, and performance analysis of the various systems were also described. Nuclear analysis was confined to parametric studies on excess reactivity and control drum span for inner reflector material combinations of graphite, aluminum, and/or beryllium. Changes required to more precisely define the NRX-A5 were discussed together with the effects associated with the twice power reactor capability envisioned for subsequent reactors.
Date: May 15, 1964
Creator: Black, D L; Minners, W; Stefanko, J S & Vijuk, R M
Partner: UNT Libraries Government Documents Department

Guide for prioritizing power plant productivity improvement projects: handbook of availability improvement methodology

Description: As part of its program to help improve electrical power plant productivity, the Department of Energy (DOE) has developed a methodology for evaluating productivity improvement projects. This handbook presents a simplified version of this methodology called the Availability Improvement Methodology (AIM), which provides a systematic approach for prioritizing plant improvement projects. Also included in this handbook is a description of data taking requirements necessary to support the AIM methodology, benefit/cost analysis, and root cause analysis for tracing persistent power plant problems. In applying the AIM methodology, utility engineers should be mindful that replacement power costs are frequently greater for forced outages than for planned outages. Equivalent availability includes both. A cost-effective ranking of alternative plant improvement projects must discern between those projects which will reduce forced outages and those which might reduce planned outages. As is the case with any analytical procedure, engineering judgement must be exercised with respect to results of purely mathematical calculations.
Date: September 15, 1981
Partner: UNT Libraries Government Documents Department

Technical study for the chemical cleaning of Dresden-1 DNS-D1-016. Volume VII. Appendices IX thru XIV

Description: Appendices are presented which contain information concerning the decontamination of Dresden-1; consultant's opinions; proceedings of the American Power Conference, Volume 37, 1975; health physics reports; toxicological properties and industrial handling hazards of Dow solvent NS-1; and expected radiation dose rates in the new Dresden station radioactive waste processing building.
Date: June 15, 1977
Partner: UNT Libraries Government Documents Department

Technical study for the chemical cleaning of Dresden-1. Volume I, Section 1 and 2

Description: A feasibility study has been completed to decontaminate the primary system of the Dresden-1 Nuclear Power Unit operated by Commonwealth Edison Company of Illinois. Available data initially were searched to determine the state of the art. Solvents based on organic acids and chelates gave unsatisfactory deontamination factors or unacceptable corrosion rates when evaluated for cleaning of specimens from the Dresden-1 primary system, under static and dynamic conditions. A new proprietary cleaning solution, Dow Solvent NS-1, was successfully applied in these laboratory studies.
Date: June 15, 1977
Partner: UNT Libraries Government Documents Department

Cover-gas-seal component development: dynamic inflatable-plug seal improvement

Description: This report documents the 1) radial compliance and 2) low friction coating tests conducted on the CRBRP Rotating Plug Inflatable Seals per test plan N707TR810014. Test results show that narrowing the seal blade from 0.25 to 0.12 in. will effectively reduce dynamic drag from 30 to 20 lb/ft under nominal conditions and will increase seal radial compliance from 0.12 to 0.30 in. without an unacceptable rise in dynamic drag. Tests also demonstrated that application of a teflon coating to the seal wear surface reduced breakaway drag by 25% based on results of comparison dwells.
Date: September 15, 1977
Creator: Horton, P.H.
Partner: UNT Libraries Government Documents Department

Activation of electrical machinery. Supplement 1. [Preliminary evaluation; not applicable to ground tests]

Description: The following analysis of the induced radioactivity in SNAP-50/SPUR electrical machinery having a high cobalt content is submitted. Induced radioactivity in the flight vehicle will contribute negligibly to allowable radiation levels. This is especially so due to the low neutron to gamma ratio of assumed radiation damage tolerances to semiconductors. A calculation to estimate the order of magnitude of induced radioactivity in cobalt is attached. The calculation is based on a best guess of the neutron spectrum directly behind a lithium hydride shield. The resulting low cobalt activity and associated dose rate of about 1 mr/hr at 10 ft from a generator or a motor is insignificant. Although the evaluation indicates insignificant levels of induced radioactivity, this conclusion is not applicable to a ground test. Neutron moderation and scattering from a containment vessel and biological shield would greatly perturb the neutron environment behind the flight shield. Posttest handling of all components within the vacuum test chamber will undoubtedly be a problem. Notwithstanding the importance of limiting induced radioactivity, other considerations such as economy, cooling and vacuum requirements will largely dictate the final facility design. In summary, an activation analysis involves the overall facility design and will not be readily resolved. For a 10,000 hr. test the Co/sup 60/ activity may range from 100 curies per lb of cobalt where no shielding is provided to 10/sup -3/ curies per lb of cobalt where the equivalent of a flight shield is provided.
Date: November 15, 1963
Creator: Smolen, J.R.
Partner: UNT Libraries Government Documents Department

Thermal and hydraulic analyses of the System 81 cold traps

Description: Thermal and hydraulic analyses of the System 81 Type I and II cold traps were completed except for thermal transients analysis. Results are evaluated, discussed, and reported. Analytical models were developed to determine the physical dimensions of the cold traps and to predict the performance. The FFTF cold trap crystallizer performances were simulated using the thermal model. This simulation shows that the analytical model developed predicts reasonably conservative temperatures. Pressure drop and sodium residence time calculations indicate that the present design will meet the requirements specified in the E-Specification. Steady state temperature data for the critical regions were generated to assess the magnitude of the thermal stress.
Date: June 15, 1977
Creator: Kim, K.
Partner: UNT Libraries Government Documents Department

IMPACT OF WATER PH ON ZEBRA MUSSEL MORTALITY

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

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