A STUDY OF THE RADIATION SHIELDING CHARACTERISTICS OF BASIC CONCRETE STRUCTURES AT THE TOWER SHIELDING FACILITY

In the first of a series of experiments performed to investigate the protection afforded by various typical structures against prompt weapons radiation, radiationintensity measurements were made at the Tower Shielding Facility in two concrete-shielded bunkers and in an interconnecting tunnel. Prompt weapons radiation was simulated by the Tower Shielding Reactor II, which was operated 100 ft above the ground. The distance between the reactor and the bunkers was approximately 700 ft. The bunkers were each 12-ft cubes and were constructed so that the shield thickness on the front face of one and on the top face of the other could be varied in 4-in. steps from 0 to 20 in. The thickness of concrete and dirt surrounding all other faces was sufficient to make them black to incident radiation. The immediate goals of the experiment were to study (1) the attenuation of radiations by various thicknesses of ordinary concrete slabs, (2) the buildup of radiation intensities within the cavities by scattering of radiation in the walls, and (3) the transmission of radiation down a tunnel with two right-angle bends. The gamma-ray and fastneutron dose rates and thermal- neutron fluxes measured at various positions within the bunkers and in the tunnel and the pulse-height spectra from a 3-in. sodium iodide crystal determined at one position in the top bunker and one position in the tunnel are reported. (auth

CALCULATIONS OF THERMAL-NEUTRON FLUX DISTRIBUTIONS IN CONCRETE-WALLED DUCTS USING AN ALBEDO MODEL WITH MONTE CARLO TECHNIQUES

BS>When radiation shields are penetrated by ducts, the gamma rays resulting from the radiative capture of lowenergy neutrons in the duct walls can be principal contributors to the total dose along the duct. In order to calculate the effect of these capture gamma rays, the distribution of low-energy neutrons in the ducts must be known. This report presents calculations of low- energy neutron distributions in concrete-walled ducts by a method in which an albedo model and a Monte Carlo technique were used. One series of calculations was performed for straight ducts so that comparisons could be made with the Simon- Clifford analytic approximation. Another series was for threelegged rectangular ducts similar to those used in a Tower Shielding Facility experiment. When au albedo similar to that to be expected for a pure thermal-neutron source was used, the results agreed very closely wilh the TSF data. A digital computer code was written to perform the calculations. (auth

MEASUREMENTS OF RADIATION INTENSITIES IN VERTICAL CONCRETE-LINED HOLES AND AN ADJOINING TUNNEL AT THE TOWER SHIELDING FACILITY

Fast neutron and gamma-dose rates within 4-ft-dia., 20-ft-deep, concrete- lined holes were measured at the Tower Shielding Facility. The radiation source was the Tower Shielding Reactor II (TSR-II) enclosed in a shield that modified the neutron to gamma ratio of the reactor leakage spectrum to more closely resemble that of a weapon spectrum. The holes were located at horizontal distances of 100, 228, and 450 ft from the reactor. From the hole at 100 ft extended a reinforced concrete-lined tunnel, 6 ft high, 21/2 ft wide, and 20 ft long, with its ceiling 10 ft below ground level. The experimental measurements consisted of vertical traverses in the three holes and horizontal traverses in the tunnel. The parameters varied included distance from the reactor, the angle of elevation of the reactor with respect to the horizontal at the hole, and the material and thickness of the shield over the hole. Reactor elevation angles ranged from 15 to 90 deg . The shields over the holes were concrete, iron, and laminated iron and concrete slabs. (auth

Bias and Uncertainty of Critical Experiment Models with CSAS25 from SCALE4.4a for Criticality Safety Analyses On the HP J-5600 (CMODB) Workstation

This report documents establishment of bias, bias trends and uncertainty for validation of the CSAS25 control module from the SCALE 4.4a computer code system for use in evaluating criticality safety of uranium systems. The 27-group ENDF/B-IV, 44-group ENDF/B-V, and 238-group ENDF/B-V cross-section libraries were used. The criticality validation calculations were performed using over 500 benchmark cases from Volumes II and IV of the ''International Handbook of Evaluated Criticality Safety Benchmark Experiments,'' published by the Nuclear Energy Agency Organization for Economic Cooperation and Development (NEA/OECD). Based on statistical analysis of the calculation results, the bias, bias trends and uncertainty of the benchmark calculations have been established for these benchmark experiments. Numerical methods for applying margins are briefly described, but the determination of appropriate correlating parameter and values for additional margin, applicable to a particular analysis, must be determined as part of process analysis. As such, this document does not specify upper subcritical limits as has been done in the past. A follow-on report will be written to assess the methods for determination of an upper safety limit in more detail, provide comparisons, and recommend a preferred method. Analysts using these results are responsible for exercising sound engineering judgment using strong technical arguments to develop a margin in k{sub eff} or other correlating parameter that is sufficiently large to ensure that conditions (calculated by this method to be subcritical by this margin) will actually be subcritical. Documentation of determination and justification of the appropriate margin in the analyst's evaluation, in conjunction with this report, will constitute the complete Validation Report in accordance with ANSI/ANS-8.1-1998, Section 4.3.6(4)