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To aid in the determinations of reactivity worths for target materials in a proposed High Flux Isotope Reactor (HFIR) target configuration containing two additional hydraulic tubes, the worths of cadmium rabbits within the current hydraulic tube were calculated using a reference model of the HFIR and the MCNP5 computer code. The worths were compared to measured worths for both static and ejection experiments. After accounting for uncertainties in the calculations and the measurements, excellent agreement between the two was obtained. Computational and measurement limitations indicate that accurate estimation of worth is only possible when the worth exceeds 10 cents. Results indicate that MCNP5 and the reactor model can be used to predict reactivity worths of various samples when the expected perturbations are greater than 10 cents. The level of agreement between calculation and experiment indicates that the accuracy of such predictions would be dependent solely on the quality of the nuclear data for the materials to be irradiated. Transients that are approximated by ''piecewise static'' computational models should likewise have an accuracy that is dependent solely on the quality of the nuclear data.

The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.

The solution of a special type of deep penetration problem is obtained by coupling a deep-penetration forward calculation with a localized adjoint calculation. The system on which the calculation is performed consists of a target far removed from a radiation source. In the absence of the target, the system geometry is simple (i.e., one- or two-dimensional). The problem is to compute some effect of interest (e.g., reaction rate, flux, etc.) within the target. The problem solution consists of a source-centered calculation of the radiation field with the target absent, a targetcentered adjoint calculation on the system with the source absent, and a coupling of the above two calculations. The technique was applied to fissile and non-fissile targets located at various distances from and having various orientations with respect to a unit isotropic point fission neutron source in an infinite air medium. (3 tables) (auth)