input group, while in the other a shape is deduced from the flux values in adjacent groups. In both cases the total flux is preserved in the conversion.
DIFFUSION (WRS module No. 7539) Solves a set of multigroup diffusion equations for a system represented in one-dimensional plane, cylindrical or spherical geometry.
The system is divided into regions each of which is assigned a material throughout which the diffusion length, the diffusion coefficient and the cross sections are uniform. The module may be supplied with group parameters appropriate to the continuous slowing down model namely the diffusion coefficient, the absorption parameter and the inverse slowing down length.
It then calculates the diffusion length and the cross section for transfer to the next lower group from simple relationships involving the above parameters.
There is provision to allow for buckling due to lateral leakage by supplying a buckling value for each group and region. Void regions may be included. The module may also be used to solve the adjoint problem.
In order to solve the differential equation for each group, a spatial mesh is superimposed on the system and the equation at each mesh point is expressed using finite difference approximations for the derivatives.
These, together with extra equations for the internal and external boundary conditions, form a set of linear equations which, being almost tri-diagonal, are easily solved by elimination and back substitution.
DOCTOR (WRS module No. 1827) Calculates any number of reaction rates given a set of multigroup fluxes at a number of spatial mesh points in one spatial dimension.
The flux values may be read from cards by the module, or may be in a data group which is already open when the module is called. For each energy group, values are supplied at a number of specified points in one spatial dimension, these points being divided into regions in which they are not necessarily spaced evenly. Average cross sections in each group can be supplied for each detector in the input data, or they may be in a data group which is already open. In the former case it is also possible to make use of a small library of cross sections which is incorporated in the module and which contains data for a number of neutron detectors in two standard group schemes. If these cross sections are used there is provision to adjust the values for the thermal group and the lowest epithermal group to allow for variations due to temperature. Results are printed out at the set of points at which the flux was given, and the contribution from each group may be included if desired.
GAMSOURCE (WRS module No. 38474) Calculates the source strength of gamma rays arising from neutron capture in a system represented by one-dimensional geometry.
The gamma ray sources are 'prompt' or 'delayed', 'prompt' including all gammas produced by neutron capture (but not from fission or inelastic scattering), 'delayed' comprising gammas released in the decay of irradiation products.
To calculate the prompt source, values of the average microscopic absorption cross section for each element present in a region are calculated for each neutron energy group from a cross section library obtained by processing cross sections from UKNDL. In this library the epithermal absorption cross section is represented by a 100 step histogram between 1 MeV and 0.07 eV.
Values for the thermal group are obtained by averaging over a Maxwellian spectrum.
The photons produced are assigned to prompt energy groups. Two standard prompt group schemes are available: one of 5, the other of 13 groups. For the 5 group scheme, the gamma spectra are those used by the COMPRASH code, while the 13 group spectra are those compiled by Side otham. The spectra may also be inputted if a different group scheme is used.
GAMSOURCE also allows independent gamma sources to be read in.
The method for delayed gamma sources is similar to that of the prompt gammas, but differs in that there are no standard group schemes. It is therefore necessary to supply values of the number of photons emitted 