The MORSE code is a multipurpose neutron and gamma-ray xransport Monte Carlo code. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems may be obtained in either the forward or adjoint mode. Time dependence for both shielding and criticality problems ir pxovided. General three-dimensional geometry, as well as specialized one-dimensional geometry descriptions, may be used with an albedo option av.ulable at any material surface. Standard multigroup cross sections such as those used in discrete ordinates codes may be used as input; either ANISN or DTP-IV cross-section formats are acceptable. Anisotropic scattering is treated for each groupto-group transfer by utilizing a generalized Gaussian quadrature technique. The modular form of the code with built-in analysis capability for all types of estimators makes it possible to solve a complete neutron-gammaray problem as one job and without the use of tapes. A detailed discussion of the relationship between forward and adjoint flux and collision densities, as well as a detailed description of the treatment of the angle of scattering, is given in the appendices. Logical flow charts for each subroutine add to the understanding of the code. viii

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Fast-neutron and gamma-ray dose rates within 4-ft-diam, 20-ftdeep, concrete-lined holes have been measured at the ORNL Tower Shielding Facility. The radiation source was the Tower Shielding Reactor II (TSR-II) enclosed in a shield which modified the neutron to gamma-ray 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, 2 1/2 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°. The shields over the holes were concrete, iron, and laminated iron and concrete slabs.

Measurements and calculations were made of fast-neutron dose rates transmitted through a LiH SNAP shield surrounded by a collar shield of iron and oil. This was done in order to evaluate the Monte Carlo techniques used to design the experimental configurations for SNAP shielding experiments at the Tower Shielding Facility. Comparisons were made for a niimber of typical configurations and the calculated and measured fast-neutron dose rates for neutrons leaving both the LiH shield and the collar shield are in excellent agreement. This establishes the validity of this technique for analyzing future experiments.

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 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.