NESC0862 FX2-TH.

FX2-TH.

FX2-TH solves the steady-state and time-dependent two-dimensional multigroup neutron diffusion equations with thermal and hydraulic feedback. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation and a more detailed model consisting of an explicit representation of fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes.

The code is designed for nuclear reactor analysis.

The multigroup diffusion equations are discretized in space using mesh-centered finite differences. The steady-state solution of these equations is found by accelerating a fission source iteration through the use of Chebyshev polynomials. The neutron fluxes for each energy group at each power iteration are found using the successive line over-relaxation method. A consistent set of steady-state conditions is established by iterating between the steady-state neutronics and thermal-hydraulics equations. The time-dependent solution is then found using the improved quasistatic method. The shape calculations required by the quasistatic method utilize the same calculational methods as described above for the steady-state calculation.

Variable dimensioning is used throughout the program so that computer storage requirements depend on a variety of problem parameters. The amount of memory required can range from 300K bytes for a small problem up to the maximum limit of computer storage for very large problems.

Running times are dependent on the number of groups, number of mesh points, length of the transient, and amount of change in the problem variables as the transient progresses.

The NESC executed the sample problem in 6 minutes on an IBM3033.

Microscopic cross sections and delayed neutron data are specified in the CCCC formats, ISOTXS and DLAYXS, respectively. A disk editing capability is included.

Input microscopic cross sections and delayed neutron data may be generated using the MC**2-2 program (NESC Abstract 355).

• R. Douglas O'Dell,
  Standard Interface Files and Procedures for Reactor Physics Codes, Version IV, LA-6941-MS.

• D.A. Meneley, Gary Leaf, A.J. Lindeman, T.A. Daly, and W.T. Sha,
  A Kinetics Model for Fast Reactor Analysis in Two Dimensions, Dynamics of Nuclear Systems, The University of Arizona Press, Tucson, Arizona, pp.483-500, 1972.

FX2-TH will dynamically allocate problem variables either on disk or in computer memory depending on the total amount of computer memory available. For any problem a certain minimum amount of memory is required; this amount is dependent on the size of the problem. In general, FX2-TH operates more efficiently if sufficient computer memory is available to retain most of the variables permanently. The sample problem executed by NESC required 900K bytes of memory.

OS/370.

FX2-TH may be compiled using either FORTRAN H or Extended compilers with OPT=2 optimization. Skeleton versions of standard routines as specified by the Committee of Computer Code Coordination (CCCC) and in the document LA-6941-MS are part of the FX2-TH package. These standard routines are REED, RITE, LRED, LRIT, TIMER, and SEEK; they may be replaced by suitable local environment routines for enhanced operation.

R.A.Shober and A. Daly
Applied Physics Division
D.R. Ferguson
Fast Reactor Safety Technical Management Center
Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439

USA