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Method for processing ENDF/B photon form factor data

Description: A method is described for processing ENDF/B photon data to generate group-to-group scattering matrices. The method has these salient features: 1. It is tailored toward treating the full energy and angular detail with which the cross sections are represented in ENDF/B; 2. It is simple to program; 3. It closely parallels a treatment developed for producing multigroup neutron matrices; 4. The time required to execute the method on a computer varies linearly with the number of energy groups as opposed to double numerical integration schemes which tend to vary as the square of the number of groups. (GHT)
Date: January 1, 1975
Creator: Lucius, J.L. & Greene, N.M.
Partner: UNT Libraries Government Documents Department

Half-value thickness measurements of ordinary concrete for neutrons from cyclotron targets

Description: From AIHA conference; Boston, Massachusetts, USA (20 May Half-value thicknesses of a wall of ordinary concrete bricks were determined for fast neutrons emitted from a variety of cyclotron beam-target combinations. The neutrons used in the study were produced at the Oak Ridge Isochronous Cyclotron by bombarding thick targets of carbon, aluminum, copper and tantalum with beams of protons, deuterons, alpha particles and carbon ions. Attenuated and unattenuated dose rate measurements of these neutrons were made at an 80 cm thick concrete block wall which served as one side of the target room where the irradiations occurred. The ratios provided by these shielded and unshielded dose rate measurements were translated into half-value thicknesses. (auth)
Date: January 1, 1973
Creator: Butler, H.M.; Wallace, K.M. & Fulmer, C.B.
Partner: UNT Libraries Government Documents Department

Available computer codes and data for radiation transport analysis

Description: The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results. (auth)
Date: January 1, 1975
Creator: Trubey, D.K.; Maskewitz, B.F. & Roussin, R.W.
Partner: UNT Libraries Government Documents Department

Final report on a benchmark experiment for neutron transport in thick sodium

Description: An experiment concerning deep neutron penetration in sodium is described, and experimental results are presented which provide a basis for verification of the accuracy of sodium cross sections to be used in transport calculations. The experiment was performed at the Tower Shielding Facility of ORNL and included measurements of both the neutron fluence and neutron spectra through a large diameter sample of sodium up to 15 ft thick. Ca1culated resu1ts for the experiment are presented for comparison with the experimental measurements. These results were obtained using the multigroup Monte Carlo code, MORSE, and a twodimensional discrete ordinates code, DOT III. One-hundred group data sets were developed from both a preliminary and the final version of the ENDF/III set (MAT = 1156) for sodium for use in the calculations. Comparisons of the calculations with experiment indicate that the preliminary version is slightly superior to the flnal version and that using the preliminary set the total neutron leakage above thermal energies penetrating up through 15 ft of sodium can be calculated to within approximates 15% and the absolute spectra penetrating up through 12.5 ft of sodium can be calculated to within 20%. Using the final set, the corresponding comparisons are 30% and 60%. (20 references) (auth)
Date: January 1, 1974
Creator: Maerker, R.E.; Muckenthaler, F.J.; Childs, R.L. & Gritzner, M.L.
Partner: UNT Libraries Government Documents Department

Sensitivity analysis development and applications program at ORNL

Description: The cross-section sensitivity analysis program at ORNL is reviewed with emphasis on present computer code capabilities and fast successful applications in the radiation shielding area. The FORSS sensitivity code system is discussed in regard to objectives, methodology, and code specifications. Examples of past shielding applications of FORSS emphasize the success of fine energy grid sensitivity studies and group structure selection, the use of evaluated error file and problem uncertainty estimation, two-dimensional shield sensitivity analysis and integral experiment design for fast reactors, data studies for the LMFBR program related to sodium and iron evaluations and iron data problems in CTR shielding design. Conclusions are drawn about the adequacy of present ENDF/B data files for sodium and iron and the general applicability of sensitivity studies in future design and analysis. 16 figures, 3 tables (auth)
Date: January 1, 1975
Creator: Oblow, E.M.
Partner: UNT Libraries Government Documents Department

Fast reactor analytical shielding progress report for July and August 1973. 189a No. 10318, activity No. HN 04 01 01 1

Description: Analytical shielding work performed for the fast reactor program during the months of July and August included calculations for both the FFTF and the Demo plant, computer code development, analyses of TSF experiments, and a cross- section sensitivity study of a TSF experiment. Most of the FFTF studies were aimed at determining the effect of various design changes on the dose rates at the maintenance deck during reactor operation, in particular, the effects of penetrations through the radial cavity shield and the effect of a new design for the head temperature control system shield. Gamma-ray dose rates due to the activated sodium coolant were also calculated for two locations in the system: around one of the branch-arm piping shields and near the entrances of the 28-in. and 16-in. coolant lines into the IHX cell. The main Demo plant study was for the lower axial shield region; another study has been initiated to calculate the Demo stored-fuel power and the ex-vessel detector response. The code development consisted in setting forth a set of specifications for the fourth version of the discrete ordinates code DOT. The TSF experiments that were analyzed were an experiment in which secondary gammaray production in iron-polyethylene configurations was investigated and another in which the transmission of the Westinghouse spectrum modifier and 5-in. radial blanket was measured. The sensitivtty study was performed to demonstrate that estimates of the accuracy of calculations for the FFTF and Demo plant can be deduced from a sensitivity study of an experimental configuration that only roughly mocks up the real system. (auth)
Date: November 1, 1973
Creator: Abbott, L.S.; Childs, R.L.; Engle, W.W. & Maerker, R.E.
Partner: UNT Libraries Government Documents Department