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Feasibility of beam driven semi-catalyzed deuterium fusion neutron sources for hybrid reactor applications

Description: The assessment is based on the estimation of the fusion device plasma properties using a simple point model, on general energy balance considerations for hybrid reactors for power generation, and on the estimation of expected performance of a specific type of hybrid reactor-a natural uranium fueled, light water moderated breeding hybrid power reactor driven by a semi-catalyzed deuterium fusion neutron source. Beam-driven semi-catalyzed deuterium and D-T fusion devices for hybrid power reactor applications are compared, and potential advantages of the former are identified. It is found that the plant efficiency of hybrid power reactors driven by a semi-catalyzed deuterium neutron source might exceed that attainable with a D-T neutron source when the fraction of the fusion neutrons that reach the blanket is smaller than about 0.8. A beam-driven deuterium fusion device can be operated in the semi-catalyzed mode and provide an intense source of neutrons even for plasma electron temperatures as low as 5 keV. Such a fusion neutron source might be useful for experimental facilities.
Date: December 1, 1977
Creator: Greenspan, E.
Partner: UNT Libraries Government Documents Department

Conceptual innovations in hybrid reactors

Description: A number of innovations in the conception of fusion-fission hybrid reactors, including the blanket, the fusion driver, the coupling of the fusion and the fission components as well as the application of hybrid reactors are described, and their feasibility assessed.
Date: January 1, 1980
Creator: Greenspan, E. & Miley, G.H.
Partner: UNT Libraries Government Documents Department

SWANS: A Prototypic SCALE Criticality Sequence for Automated Optimization Using the SWAN Methodology

Description: SWANS is a new prototypic analysis sequence that provides an intelligent, semi-automatic search for the maximum k{sub eff} of a given amount of specified fissile material, or of the minimum critical mass. It combines the optimization strategy of the SWAN code with the composition-dependent resonance self-shielded cross sections of the SCALE package. For a given system composition arrived at during the iterative optimization process, the value of k{sub eff} is as accurate and reliable as obtained using the CSAS1X Sequence of SCALE-4.4. This report describes how SWAN is integrated within the SCALE system to form the new prototypic optimization sequence, describes the optimization procedure, provides a user guide for SWANS, and illustrates its application to five different types of problems. In addition, the report illustrates that resonance self-shielding might have a significant effect on the maximum k{sub eff} value a given fissile material mass can have.
Date: January 11, 2001
Creator: Greenspan, E.
Partner: UNT Libraries Government Documents Department

Preliminary report on the promise of accelerator-driven natural-uranium-fueled light-water-moderated breeding power reactors

Description: A new concept for a power breeder reactor that consists of an accelerator-driven subcritical thermal fission system is proposed. In this system an accelerator provides a high-energy proton beam which interacts with a heavy-element target to produce, via spallation reactions, an intense source of neutrons. This source then drives a natural-uranium-fueled, light-water-moderated and -cooled subcritical blanket which both breeds new fuel and generates heat that can be converted to electrical power. The report given presents a general layout of the resulting Accelerator Driven Light Water Reactor (ADLWR), evaluates its performance, discusses its fuel cycle characteristics, and identifies the potential contributions to the nuclear energy economy this type of power reactor might make. A light-water thermal fission system is found to provide an attractive feature when designed to be source-driven. The equilibrium fissile fuel content that gives the highest energy multiplication is approximately equal to the content of /sup 235/U in natural uranium. Consequently, natural-uranium-fueled ADLWRs that are designed to have the highest energy generation per source neutron are also fuel-self-sufficient; that is, their fissile fuel content remains constant with burnup. This feature allows the development of a nuclear energy system that is based on the most highly developed fission technology available (the light water reactor technology) and yet has a simple and safe fuel cycle. ADLWRs will breed on natural uranium, have no doubling time limitation, and be free from the need for uranium enrichment or for the separation of plutonium. It appears that ADLWRs could also be efficiently operated with thorium fuel cycles and with denatured fuel cycles.
Date: July 1, 1977
Creator: Greenspan, E.
Partner: UNT Libraries Government Documents Department

Constrained sensitivity theory

Description: In sensitivity and uncertainty analysis of to-be-built reactors it is customary to use k-reset sensitivity functions - accounting for the combined effects of the change (or uncertainty) in the input data and of the alteration in some design variable applied to maintain criticality. Critical reactors are usually subjected to several constraints, such as power peaking factor and breeding ratio constraints, in addition to the criticality constraint. Perturbation theory formulations which can account, simultaneously, for several constraints both in critical reactors and in source driven systems (such as radiation shields and blankets of fusion devices) are presented. All the sensitivity and uncertainty analyses of source driven systems carried out so far used unconstrained sensitivity functions despite the fact that such systems can be also subjected to a variety of constraints.
Date: January 1, 1980
Creator: Greenspan, E. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Use of Solid Hydride Fuel for Improved long-Life LWR Core Designs

Description: The primary objective of this project was to assess the feasibility of improving the performance of PWR and BWR cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the bus-bar cost of electricity (COE). Additional performance measures considered are safety, fuel bundle design simplicity – in particular for BWR’s, and plutonium incineration capability. It was found that hydride fuel can safely operate in PWR’s and BWR’s without restricting the linear heat generation rate of these reactors relative to that attainable with oxide fuel. A couple of promising applications of hydride fuel in PWR’s and BWR’s were identified: (1) Eliminating dedicated water moderator volumes in BWR cores thus enabling to significantly increase the cooled fuel rods surface area as well as the coolant flow cross section area in a given volume fuel bundle while significantly reducing the heterogeneity of BWR fuel bundles thus achieving flatter pin-by-pin power distribution. The net result is a possibility to significantly increase the core power density – on the order of 30% and, possibly, more, while greatly simplifying the fuel bundle design. Implementation of the above modifications is, though, not straightforward; it requires a design of completely different control system that could probably be implemented only in newly designed plants. It also requires increasing the coolant pressure drop across the core. (2) Recycling plutonium in PWR’s more effectively than is possible with oxide fuel by virtue of a couple of unique features of hydride fuel – reduced inventory of U-238 and increased inventory of hydrogen. As a result, the hydride fuelled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it incinerates in one pass is double that of the MOX fuel. The fissile fraction of the Pu in ...
Date: April 30, 2006
Creator: Greenspan, E
Partner: UNT Libraries Government Documents Department

Nuclear elastic scattering effects on fusion product transport in the FRM

Description: Large energy transfer (LET) events such as nuclear elastic scatterng (NES) are shown to have significant effects on fusion product transport in the field-reversed mirror. The method used and preliminary results obtained from the study on NES effects on f/sub p/ orbits are described. (MOW)
Date: January 1, 1981
Creator: DeVeaux, J.C.; Greenspan, E. & Miley, G.H.
Partner: UNT Libraries Government Documents Department

Tritium catalyzed deuterium tokamaks

Description: A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the /sup 3/He from the D(D,n)/sup 3/He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general).
Date: April 1, 1984
Creator: Greenspan, E.; Miley, G.H.; Jung, J. & Gilligan, J.
Partner: UNT Libraries Government Documents Department

Small Liquid Metal Cooled Reactor Safety Study

Description: The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S ...
Date: November 2, 2005
Creator: Minato, A; Ueda, N; Wade, D; Greenspan, E & Brown, N
Partner: UNT Libraries Government Documents Department

SWAN: a code for the analysis and optimization of fusion reactor nucleonic characteristics

Description: This report is intended as a User's Manual for SWAN''--- a code written for perturbation theory analysis and optimization of the nucleonic characteristics of fusion reactor blankets. SWAN is also applicable to any problem described by the inhomogeneous linear transport equation, e.g., radiation shield problems. The optimization method is based on variational techniques. The variables of the optimization are material densities, with no restriction on their number. One joint restraint may be imposed in addition to upper and lower limits on each density. The parameter to be extremized and the restraint may be either a weighttype (linear) or nucleonic (bilinear) functional. The transport calculations for SWAN are performed with the one-dimensional discrete-ordinate code ANISN. (auth)
Date: November 1, 1973
Creator: Greenspan, E.; Price, W. G., Jr. & Fishman, H.
Partner: UNT Libraries Government Documents Department

Developments in sensitivity theory

Description: A review of recent developments in sensitivity theory is presented with an emphasis on (a) extensions to new areas such as thermal hydraulics, reactor depletion, multi-constraint and extrema problems, and (b) recent mathematical refinements to and extensions of the basic theory. The diverse new areas of application are discussed from a unified theoretical viewpoint based on nonlinear functional analysis. Several new applications of sensitivity theory are presented for problems in constrained reactor physics calculations, irradiation experiment analysis, reactor burnup calculations, and transient thermal-hydraulic analysis. Future directions of the research are suggested.
Date: January 1, 1980
Creator: Cacuci, D.G.; Greenspan, E.; Marable, J.H. & Williams, M.L.
Partner: UNT Libraries Government Documents Department

Code development incorporating environmental, safety and economic aspects of fusion reactors; Annual progress report

Description: This document is a proposal to continue the authors work on the Environmental, Safety and Economic (ESE) aspects of fusion reactors under DOE contract DE-FR03-89ER52514. The grant objectives continue those from the previous grant: (1) completion of first-generation Environmental, Safety and Economic (ESE) computer modules suitable as integral components of tokamak systems codes. (2) continuation of work on special topics, in support of the above and in response to OFE requests. The proposal also highlights progress on the contract in the twelve months since April, 1992. This has included work with the ARIES and ITER design teams, work on tritium management, studies on materials activation, and calculation of radioactive inventories in fusion reactors.
Date: December 31, 1993
Creator: Fowler, T.K.; Greenspan, E. & Holdren, J.P.
Partner: UNT Libraries Government Documents Department

Monte Carlo simulations of neutron well-logging in granite and sand to detect water and trichloroethane (TCA)

Description: The Monte Carlo code MCNP is used in simulations of neutron well logging in granite to detect water and TCA (C{sub 2}H{sub 3}Cl{sub 3}), a common ground contaminant, in fractures of 1 cm and 1 mm thickness at various distances and orientations. Also simulated is neutron well logging in wet sand to detect TCA and lead (Pb) at various uniform concentrations. The {sup 3}H(d,n) (DT) and{sup 2}H(d,n) (DD) neutron producing reactions are used in the simulations to assess the relative performance of each. Simulations are also performed to determine the efficiency of several detector materials such as CdZnTe, Ge and NaI as a function of photon energy. Results indicate that, by examining the signal from the 6.11 MeV gamma from the thermal neutron capture of Cl in TCA, trace amounts (few ppm) are detectable in saline free media. Water and TCA filled fractures are also detectable. These results are summarized in Tables 7--21. Motivation for this work is based on the need for detection of trace environmental pollutants as well as possible fracture characterization of geologic media.
Date: January 1, 1998
Creator: Hua, D.D.; Donahue, R.J.; Celata, C.M. & Greenspan, E.
Partner: UNT Libraries Government Documents Department

Development of coupled SCALE4.2/GTRAN2 computational capability for advanced MOX fueled assembly designs

Description: An advanced assembly code system that can efficiently and accurately analyze various designs (current and advanced) proposed for plutonium disposition is being developed by {open_quotes}marrying{close_quotes} two existing state-of-the-art methodologies-GTRAN2 and SCALE 4.2. The resulting code system, GT-SCALE, posses several unique characteristics: exact 2D representation of a complete fuel assembly, while preserving the heterogeniety of each of its pin cells; flexibility in the energy group structure, the present upper limit being 218 groups; a comprehensive cross-section library and material data base; and accurate burnup calculations. The resulting GT-SCALE is expected to be very useful for a wide variety of applications, including the analysis of very heterogeneous UO{sub 2} fueled LWR fuel assemblies; of hexagonal shaped fuel assemblies as of the Russian LWRs; of fuel assemblies for HTGRs; as well as for the analysis of criticality safety and for calculation of the source term of spent fuel.
Date: May 1, 1995
Creator: Vujic, J.; Greenspan, E.; Slater, Postma, T.; Casher, G.; Soares, I. & Leal, L.
Partner: UNT Libraries Government Documents Department

Optimal Neutron Source & Beam Shaping Assembly for Boron Neutron Capture Therapy

Description: There were three objectives to this project: (1) The development of the 2-D Swan code for the optimization of the nuclear design of facilities for medical applications of radiation, radiation shields, blankets of accelerator-driven systems, fusion facilities, etc. (2) Identification of the maximum beam quality that can be obtained for Boron Neutron Capture Therapy (BNCT) from different reactor-, and accelerator-based neutron sources. The optimal beam-shaping assembly (BSA) design for each neutron source was also to e obtained. (3) Feasibility assessment of a new neutron source for NCT and other medical and industrial applications. This source consists of a state-of-the-art proton or deuteron accelerator driving and inherently safe, proliferation resistant, small subcritical fission assembly.
Date: April 30, 2003
Creator: Vujic, J.; Greenspan, E.; Kastenber, W.E.; Karni, Y.; Regev, D.; J.M. Verbeke, K.N. Leung et al.
Partner: UNT Libraries Government Documents Department

The Secure, Transportable, Autonomous Reactor System

Description: The Secure, Transportable, Autonomous Reactor (STAR) system is a development architecture for implementing a small nuclear power system, specifically aimed at meeting the growing energy needs of much of the developing world. It simultaneously provides very high standards for safety, proliferation resistance, ease and economy of installation, operation, and ultimate disposition. The STAR system accomplishes these objectives through a combination of modular design, factory manufacture, long lifetime without refueling, autonomous control, and high reliability.
Date: May 27, 1999
Creator: Brown, N.W.; Hassberger, J.A.; Smith, C.; Carelli, M.; Greenspan, E.; Peddicord, K.L. et al.
Partner: UNT Libraries Government Documents Department

ENHS : the encapsulated nuclear heat source - a nuclear energy concept for emerging worldwide energy markets.

Description: A market analysis is presented which delineates client needs and potential market size for small turnkey nuclear power plants with full fuel cycle services. The features of the Encapsulated Nuclear Heat Source (ENHS) which is targeted for this market are listed, and the status of evaluation of technological viability is summarized.
Date: February 26, 2002
Creator: Wade, D.C.; Feldman, E.; Sienicki, J.; Sofu, T.; Brown, N.W.; Hossain, Q. et al.
Partner: UNT Libraries Government Documents Department

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

Description: This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report ...
Date: June 23, 2008
Creator: Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P. et al.
Partner: UNT Libraries Government Documents Department