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Neutron streaming analysis for shield design of FMIT Facility

Description: Applications of the Monte Carlo method have been summarized relevant to neutron streaming problems of interest in the shield design for the FMIT Facility. An improved angular biasing method has been implemented to further optimize the calculation of streaming and this method has been applied to calculate streaming within a double bend pipe.
Date: December 1, 1980
Creator: Carter, L.L.
Partner: UNT Libraries Government Documents Department

Preliminary upper axial shield design for space reactor ground test

Description: A prototype of the SP-100 space reactor will be tested in a vacuum environment on the ground to verify the design prior to flight applications. Neutronic calculations are under way to design shielding that will provide the appropriate operational protection for both personnel and instrumentation, and will not compromise any important flight-type conditions of the reactor and shield system. This document describes the preliminary design of the shielding system. 1 ref., 11 figs.
Date: November 1, 1987
Creator: Carter, L.L. & Bunch, W.L.
Partner: UNT Libraries Government Documents Department

Certification of MCNP Version 4A for WHC computer platforms. Revision 7

Description: MCNP is a general-purpose Monte Carlo code that can be used for neutron, photon, or coupled neutron/photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces (elliptical tori).
Date: May 3, 1995
Creator: Carter, L.L.
Partner: UNT Libraries Government Documents Department

Computational benchmark for deep penetration in iron

Description: A benchmark for calculation of neutron transport through iron is now available based upon a rigorous Monte Carlo treatment of ENDF/B-IV and ENDF/B-V cross sections. The currents, flux, and dose (from monoenergetic 2, 14, and 40 MeV sources) have been tabulated at various distances through the slab using a standard energy group structure. This tabulation is available in a Los Alamos Scientific Laboratory report. The benchmark is simple to model and should be useful for verifying the adequacy of one-dimensional transport codes and multigroup libraries for iron. This benchmark also provides useful insights regarding neutron penetration through iron and displays differences in fluxes calculated with ENDF/B-IV and ENDF/B-V data bases. (GHT)
Date: October 1, 1979
Creator: Carter, L.L. & Hendricks, J.S.
Partner: UNT Libraries Government Documents Department

Particle-transport simulation with the Monte Carlo method

Description: Attention is focused on the application of the Monte Carlo method to particle transport problems, with emphasis on neutron and photon transport. Topics covered include sampling methods, mathematical prescriptions for simulating particle transport, mechanics of simulating particle transport, neutron transport, and photon transport. A literature survey of 204 references is included. (GMT)
Date: January 1, 1975
Creator: Carter, L.L. & Cashwell, E.D.
Partner: UNT Libraries Government Documents Department

Nuclear data relevant to shield design of FMIT facility

Description: Nuclear data requirements are reviewed for the design of the Fusion Materials Irradiation Test (FMIT) facility. This accelerator-based facility, now in the early stages of construction at Hanford, will provide high fluences in a fusion-like radiation environment for the testing of materials. The nuclear data base required encompasses the entire range of neutron energies from thermal to 50 MeV. In this review, we consider neutron source terms, cross sections for thermal and bulk shield design, and neutron activation for the facility.
Date: April 1, 1980
Creator: Carter, L.L.; Morford, R.J. & Wilcox, A.D.
Partner: UNT Libraries Government Documents Department

Monte Carlo applications at Hanford Engineering Development Laboratory

Description: Twenty applications of neutron and photon transport with Monte Carlo have been described to give an overview of the current effort at HEDL. A satisfaction factor was defined which quantitatively assigns an overall return for each calculation relative to the investment in machine time and expenditure of manpower. Low satisfaction factors are frequently encountered in the calculations. Usually this is due to limitations in execution rates of present day computers, but sometimes a low satisfaction factor is due to computer code limitations, calendar time constraints, or inadequacy of the nuclear data base. Present day computer codes have taken some of the burden off of the user. Nevertheless, it is highly desirable for the engineer using the computer code to have an understanding of particle transport including some intuition for the problems being solved, to understand the construction of sources for the random walk, to understand the interpretation of tallies made by the code, and to have a basic understanding of elementary biasing techniques.
Date: March 1, 1980
Creator: Carter, L.L.; Morford, R.J. & Wilcox, A.D.
Partner: UNT Libraries Government Documents Department

Neutron environment in d + Li facilities

Description: A microscopic d + Li neutron yield model has been developed based upon classical models and experimental data. Using equations suggested by the Serber and evaporation models, a generalized least squares adjustment procedure generated angular yields for E/sub d/ to 40 MeV using the available experimental data. The HEDL-UCD experiment at E/sub d/ = 35 was used to adjust parameters describing the neutron spectra. The model is used to predict yields, spectra, and damage responses in the FMIT Test Cell.
Date: January 1, 1980
Creator: Mann, F.M.; Schmittroth, F. & Carter, L.L.
Partner: UNT Libraries Government Documents Department

Nuclear data relevant to shield design of FMIT facility

Description: Nuclear data requirements are reviewed for the design of the Fusion Materials Irradiation Test (FMIT) facility. This accelerator-based facility, now in the early stages of construction at Hanford, will provide high fluences in a fusion-like radiation environment for the testing of materials. The nuclear data base required encompasses the entire range of neutron energies from thermal to 50 MeV. In this review, we consider neutron source terms, cross sections for thermal and bulk shield design, and neutron activation for the facility.
Date: January 1, 1980
Creator: Carter, L.L.; Morford, R.J. & Wilcox, A.D.
Partner: UNT Libraries Government Documents Department

Thermal neutron flux contours from criticality event

Description: The generation of thermal neutron flux contours from a criticality event is demonstrated for an idealized building with a criticality event in one of the rooms. The MCNP Monte Carlo computer code is used to calculate the thermal neutron flux.
Date: August 1, 1996
Creator: Carter, L.L., Westinghouse Hanford
Partner: UNT Libraries Government Documents Department

MCNP Perturbation Capability for Monte Carlo Criticality Calculations

Description: The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k{sub eff} in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward.
Date: September 20, 1999
Creator: Hendricks, J.S.; Carter, L.L. & McKinney, G.W.
Partner: UNT Libraries Government Documents Department

Certification of MCNP version 4A for WHC computer platforms

Description: MCNP is a general-purpose Monte Carlo code that can be used for neutron, photon, or coupled neutron/photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces (elliptical tori).
Date: May 7, 1996
Creator: Carter, L.L., Westinghouse Hanford
Partner: UNT Libraries Government Documents Department

Verification of the MCNP (TM) Perturbation Correction Feature for Cross-Section Dependent Tallies

Description: The Monte Carlo N-Particle Transport Code MCNP version 4B perturbation capability has been extended to cross-section dependent tallies and to the track-length estimate of Iqff in criticality problems. We present the complete theory of the MCNP perturbation capability including the correction to MCNP4B which enables cross-section dependent perturbation tallies. We also present the MCNP interface as an upgrade to the MCNP4B manual. Finally, we present test results demonstrating the validity of the perturbation capability in MCNP, particularly cross-section dependent problems.
Date: October 1, 1998
Creator: Hess, A. K.; McKinney, G. W.; Hendricks, J. S. & Carter, L. L.
Partner: UNT Libraries Government Documents Department

Spatial dependence of flux and damage in the FMIT test cell

Description: Experimental Li(d,n) thick target yields have been combined with nuclear models to determine the microscopic Li(d,n) cross section as a function of incoming deuteron energy (E/sub d/ < 40 MeV), of outgoing neutron energy (0 less than or equal to E/sub n/ less than or equal to 50 MeV), and of outgoing neutron angle (0 less than or equal to theta less than or equal to 180/sup 0/). A generalized least squares adjustment procedure using all the experimental data for 14 less than or equal to E/sub d/ less than or equal to 50 provided the overall normalization and the angular distribution, while the Serber stripping model and the evaporation model provided the neutron energy dependence. The cross sections are applied to the conditions appropriate to the FMIT (Fusion Materials Irradiation Testing) facility to determine flux and damage parameter levels inside the test cell.
Date: January 1, 1979
Creator: Mann, F.M.; Schmitroth, F.; Carter, L.L. & Schiffgens, J.O.
Partner: UNT Libraries Government Documents Department

Spatial variations of damage parameters in FMIT and their implications

Description: The major conclusion is that the variation in damage rates in FMIT will be dominated by changes in flux, not spectrum. Throughout the test region where the flux is greater than 10/sup 14/ n/cm/sup 2/.s, the flux varies by a factor of about 20, while the spectral-averaged displacement and helium production cross sections for copper vary by less than factors of two and four, respectively. The corresponding helium-to-dpa ratios bracket a fusion reactor first wall value for copper (i.e., 7.7 appm He/dpa). With the Li(d,n) yields and copper damage energy and helium production cross sections used in this study, the test volumes for which the displacement and total helium production rates are greater than those at a D-T fusion reactor first wall, with a loading of 1.25 MW/m/sup 2/, are about 100 and 130 cm/sup 3/, respectively.
Date: December 1, 1978
Creator: Schiffgens, J. O.; Simons, R. L.; Mann, F. M. & Carter, L. L.
Partner: UNT Libraries Government Documents Department

Monte Carlo applications for the design and operation of nuclear facilities

Description: The computational capabilities of current supercomputers enable the application of rigorous Monte Carlo methods to solve day-to-day neutronics and shielding problems. Experience at Westinghouse Hanford Company has included applications to: reactor operations, decommissioning of a reactor facility, and the design of a space reactor; intermediate energy accelerators; and high-level waste facilities and casks. These practical applications are typically computationally intensive because of the amount of information required. A number of practical examples are discussed. An increase in effective computer capabilities would further enhance the use of Monte Carlo methods. 16 refs., 4 figs., 2 tabs.
Date: June 1, 1988
Creator: Carter, L.L.; Bunch, W.L.; Morford, R.J.; Wootan, D.W. & Schwarz, R.A.
Partner: UNT Libraries Government Documents Department

Neutron and gamma characterization within the FFTF reactor cavity

Description: Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables.
Date: August 1, 1980
Creator: Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D. & Wood, M.R.
Partner: UNT Libraries Government Documents Department