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Creators/Authors contains: "Oak Ridge National Lab., TN"
  1. The integrated Earth system model (iESM) has been developed as a new tool for projecting the joint human/climate system. The iESM is based upon coupling an integrated assessment model (IAM) and an Earth system model (ESM) into a common modeling infrastructure. IAMs are the primary tool for describing the human–Earth system, including the sources of global greenhouse gases (GHGs) and short-lived species (SLS), land use and land cover change (LULCC), and other resource-related drivers of anthropogenic climate change. ESMs are the primary scientific tools for examining the physical, chemical, and biogeochemical impacts of human-induced changes to the climate system. Themore » iESM project integrates the economic and human-dimension modeling of an IAM and a fully coupled ESM within a single simulation system while maintaining the separability of each model if needed. Both IAM and ESM codes are developed and used by large communities and have been extensively applied in recent national and international climate assessments. By introducing heretofore-omitted feedbacks between natural and societal drivers, we can improve scientific understanding of the human–Earth system dynamics. Potential applications include studies of the interactions and feedbacks leading to the timing, scale, and geographic distribution of emissions trajectories and other human influences, corresponding climate effects, and the subsequent impacts of a changing climate on human and natural systems. This paper describes the formulation, requirements, implementation, testing, and resulting functionality of the first version of the iESM released to the global climate community.« less
  2. The materials requirements in a high-power spallation neutron source like the SNS are particularly demanding. Materials at the target station are of special concern; these include the containment vessel and protective shroud for the mercury target material, beam windows, support structures, moderator housings, and beam tubes. The material chosen for the mercury containment vessel is Type 316 stainless steel (316SS). This choice is based on the extensive background of experience with 316SS and on its good fabricability and availability. While much has been learned about the effect of radiation on the properties of 316SS, the preponderance of this information stemsmore » from fission reactor irradiations, where the average neutron energy is only 1 to 2 MeV and relatively few neutrons are present with energies above {approximately}10 MeV. By contrast, the energies of the spallation neutrons at the SNS extend up to the 1000-MeV energy of the incident protons. However, because of the design of the target module, irradiation creep is not expected to be a significant problem. The major concern is for radiation embrittlement. Mercury is known to be an aggressive medium, and corrosion and compatibility studies for 316SS and INCONEL 718 are included in the research and development program on SNS materials. INCONEL 718 is under consideration as a beam window material. Two issues are receiving particular attention: liquid-metal embrittlement and temperature gradient mass transfer. Constant-strain-rate tensile tests were conducted in mercury and mercury-gallium at 23 C for 316SS and INCONEL 718 and at 100 C for 316SS; in all the tests there was no evidence of liquid metal embrittlement. In addition, preliminary mass transfer tests on 3165SS or INCONEL 718 at temperatures up to 350 C reveled no significant effects. Fatigue tests in mercury have recently been initiated. As a partial simulation of SNS conditions, they provide a more severe probe for possible ductility loss associated with a liquid-metal environment.« less
  3. The Spallation Neutron Source (SNS) is a neutron source providing intense neutron fluxes that will be used for performing a large variety of neutron scattering experiments. SNS is to be completed and start operation in 2005. Protons will be accelerated to 1 GeV, stored in an accumulator ring, and then injected into a neutron-producing target. After leaving the target (Hg in the ca/se of SNS), the neutrons are prepared for experiments by first using a moderator to impose energy and width requirements on the neutron pulse. One of the most important ingredients is the moderator material. Four materials that aremore » commonly used and that were considered for use in SNS are liquid hydrogen (L-H{sub 2}), liquid water (L-H{sub 2}O), liquid methane (L-CH{sub 4}), and solid methane (S-CH{sub 4}). The spectra (neutron current versus neutron energy) for these four materials are shown. As may be seen, at low neutron energies (<10 MeV), the most effective moderator is S-CH{sub 4}, which produces up to four times as many neutrons in this energy range as L-H{sub 2}. The problem with the material is the internal storage of energy that can be spontaneously and explosively released. At energies of just above 10 MeV, the most effective moderator material is L-CH{sub 4}. Polymerization problems, however, preclude its use at high powers (again such as in SNS), where the buildup of undesirable materials becomes prohibitive. This is, however, an important energy range for neutron experiments. Preliminary consideration is being given to a composite moderator that contains two adjacent sections, one of L-H{sub 2} and one of L-H{sub 2}O, which produces a spectrum that is very similar to L-CH{sub 4}.« less
  4. Monte Carlo methods are extremely powerful and heavily utilized for many applications in nuclear criticality safety. Accurate criticality calculations are possible because of the global nature of neutron multiplication. The stochastic approach has limitations, however, and is not appropriate for specialized applications that require differential fluxes or accurate neutron density distributions. The NEW Transport Algorithm (NEWT) computer code, developed at Oak Ridge National Laboratory (ORNL), has the ability to closely model nonorthogonal two-dimensional geometries that are traditionally left to Monte Carlo analyses. Because it is based on the discrete ordinates formalism, it can provide an accurate prediction of neutron distributionsmore » in space and energy. However, unlike most discrete ordinates methods, NEWT solves fluxes on a grid of arbitrary polygons, which can be used to closely approximate complex configurations. Results of multidimensional depletion and sensitivity/uncertainty analyses will be reported in the future after significant testing has been completed. Herein the authors focus on a recent study performed at ORNL to understand discrepancies noted for the Wigner-Seitz cell approximation often applied in lattice calculations.« less
  5. Different severe accident sequences employing the MELCOR code, version 1.8.4 QK, have been simulated at the Grand Gulf Nuclear Station (Grand Gulf). The postulated severe accidents simulated are two low-pressure, short-term, station blackouts; two unmitigated small-break (SB) loss-of-coolant accidents (LOCAs) (SBLOCAs); and one unmitigated large LOCA (LLOCA). The purpose of this study was to calculate best-estimate timings of events and source terms for a wide range of severe accidents and to compare the plant response to these accidents.
  6. A mercury target will be used to produce neutrons from pulsed 1-GeV protons for the Spallation Neutron Source (SNS). Mercury was selected because of its high source brightness, its low freezing temperature, and the improved radiation damage lifetime expected. The mercury target vessel will be made of thin stainless steel walls. One of the key issues associated with using such a target is the ability to withstand the thermal shock loads caused by the pulsed proton beam of 17,000 J in {approximately}0.5 {micro}s. The resulting pressure waves in the mercury associated with the enormous rate of temperature rise ({approximately}10{sup 7}more » C/s), will interact with the walls of the mercury target. This interaction could lead to excessive stresses in the target vessel, thereby limiting the lifetime of the target. To address the thermal shock issues, a series of single-pulse tests on a mercury target was conducted in the Los Alamos Neutron Science Center (LANSCE). Additional single-pulse thermal shock tests were conducted in the LANSCE facility on January 30--31, 1999.« less
  7. The author proposes preconditioning as a viable acceleration scheme for the inner iterations of transport calculations in slab geometry. In particular he develops Adjacent-Cell Preconditioners (AP) that have the same coupling stencil as cell-centered diffusion schemes. For lowest order methods, e.g., Diamond Difference, Step, and 0-order Nodal Integral Method (ONIM), cast in a Weighted Diamond Difference (WDD) form, he derives AP for thick (KAP) and thin (NAP) cells that for model problems are unconditionally stable and efficient. For the First-Order Nodal Integral Method (INIM) he derives a NAP that possesses similarly excellent spectral properties for model problems. The two mostmore » attractive features of the new technique are:(1) its cell-centered coupling stencil, which makes it more adequate for extension to multidimensional, higher order situations than the standard edge-centered or point-centered Diffusion Synthetic Acceleration (DSA) methods; and (2) its decreasing spectral radius with increasing cell thickness to the extent that immediate pointwise convergence, i.e., in one iteration, can be achieved for problems with sufficiently thick cells. He implemented these methods, augmented with appropriate boundary conditions and mixing formulas for material heterogeneities, in the test code APID that he uses to successfully verify the analytical spectral properties for homogeneous problems. Furthermore, he conducts numerical tests to demonstrate the robustness of the KAP and NAP in the presence of sharp mesh or material discontinuities. He shows that the AP for WDD is highly resilient to such discontinuities, but for INIM a few cases occur in which the scheme does not converge; however, when it converges, AP greatly reduces the number of iterations required to achieve convergence.« less
  8. An analytical study of the solid angle subtended at a point by objects of first and second algebraic order has been made. It is shown that the derived solid angle for all such objects is in the form of a general elliptic integral, which can be written as a linear combination of elliptic integrals of the first and third kind and elementary functions. Many common surfaces and volumes have been investigated, including the conic sections and their volumes of revolution. The principal feature of the study is the manipulation of solid-angle equations into integral forms that can be matched withmore » those found in handbook tables. These integrals are amenable to computer special function library routine analysis requiring no direct interaction with elliptic integrals by the user. The general case requires the solution of a fourth-order equation before specific solid-angle formulations can be made, but for many common geometric objects this equation can be solved by elementary means. Methods for the testing and application of solid-angle equations with Monte Carlo rejection and estimation techniques are presented. Approximate and degenerate forms of the equations are shown, and methods for the evaluation of the solid angle of a torus are outlined.« less
  9. Available nuclear structure information for all nuclei with mass numbers A=248,252,256,260 and 264 are presented. Various decay and reaction data are evaluated and compared. Adopted data, levels, spin, parity and configuration assignments are given.
  10. This paper presents updated analyses of the cylinder specimen being used in the international Network for Evaluating Steel Components (NESC) large-scale spinning-cylinder project (NESC-1). The NESC was organized as an international forum to exchange information on procedures for structural integrity assessment, to collaborate on specific projects, and to promote the harmonization of international standards. The objective of the NESC-1 project is to focus on a complete procedure for assessing the structural integrity of aged reactor pressure vessels. A clad cylinder containing through-clad and subclad cracks will be tested under pressurized-thermal shock conditions at AEA Technology, Risley, U.K. Three-dimensional finite-element analysesmore » were carried out to determine the effects of including the cladding heat-affected zone (HAZ) in the models. The cylinder was modeled with inner-surface through-clad cracks having a depth of 74 mm and aspect ratios of 2:1 and 6:1. The cylinder specimen was subjected to centrifugal loading followed by a thermal shock and analyzed with a thermoelastic-plastic material model. The peak K{sub 1} values occurred at the clad/HAZ interface for the 6:1 crack and at the HAZ/base interface for the 2:1 crack. The analytical results indicate that cleavage initiation is likely to be achieved for the 6:1 crack, but questionable for the 2:1 crack.« less
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