IAEA0929 RICANT.
RICANT, Neutron Transport in X-Y Geometry for Isotropic Scattering
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
[ top ]
[ top ]
To submit a request, click below on the link of the version you wish to order.
Only liaison officers are authorised to submit online requests. Rules for requesters are
available here.
| Program name | Package id | Status | Status date |
|---|---|---|---|
| RICANT | IAEA0929/01 | Tested | 14-FEB-1990 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| IAEA0929/01 | NORSK DATA 560 | DEC VAX 8810 |
[ top ]
[ top ]
4. METHOD OF SOLUTION
The integral form of the neutron transport equation is solved by the interface current method. The problem is broken into small regions. The transport equation is setup in each of these small regions. The outgoing neutron current at one surface is related to the incoming currents at the other surfaces of the region and the neutron sources inside the region. The outgoing current of one region is the incoming current of adjacent regions. Associated Legendre polynomials are used to represent the angular dependence of fluxes. The weighted residual method is used to find the expansion coefficients of the angular fluxes. Making use of the superposition principle of neutron currents, inner iterations are performed over the current components. Outer iterations are performed over groups. Numerical integrations are performed by Gauss quadrature. Neutron conservation is imposed for the calculation of transmission probabilities of current components and also in finding the average flux inside a region.
The integral form of the neutron transport equation is solved by the interface current method. The problem is broken into small regions. The transport equation is setup in each of these small regions. The outgoing neutron current at one surface is related to the incoming currents at the other surfaces of the region and the neutron sources inside the region. The outgoing current of one region is the incoming current of adjacent regions. Associated Legendre polynomials are used to represent the angular dependence of fluxes. The weighted residual method is used to find the expansion coefficients of the angular fluxes. Making use of the superposition principle of neutron currents, inner iterations are performed over the current components. Outer iterations are performed over groups. Numerical integrations are performed by Gauss quadrature. Neutron conservation is imposed for the calculation of transmission probabilities of current components and also in finding the average flux inside a region.
[ top ]
5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
No restrictions exist on the maximum number of energy groups or maximum number of meshes usable. The main storage array is presently 100K words, easily extendible due to variable dimensioning. However, due to the fixed dimension of the built-in matrix inversion routine, the maximum number of angular expansion functions is limited to ten.
No restrictions exist on the maximum number of energy groups or maximum number of meshes usable. The main storage array is presently 100K words, easily extendible due to variable dimensioning. However, due to the fixed dimension of the built-in matrix inversion routine, the maximum number of angular expansion functions is limited to ten.
[ top ]
6. TYPICAL RUNNING TIME
For a 500 flux point problem with flux convergence at any point lower than 1*10**(-3), and eigenvalue convergence lower than 1*10**(-5), CPU time is 15 minutes.
For a 500 flux point problem with flux convergence at any point lower than 1*10**(-3), and eigenvalue convergence lower than 1*10**(-5), CPU time is 15 minutes.
IAEA0929/01
The test case ran at NEADB on a DEC VAX 8810 computer in 3 minutes of CPU time.[ top ]
7. UNUSUAL FEATURES OF THE PROGRAM
Unlike the discrete ordinate method for solving the neutron transport equation (e.g. TWOTRAN), this code does not suffer from negative fluxes and ray effects. Use of P2 half space expansion for angular flux is sufficient for most problems. Use of higher order expansions is yet to be studied which is easily possible using the present code.
Unlike the discrete ordinate method for solving the neutron transport equation (e.g. TWOTRAN), this code does not suffer from negative fluxes and ray effects. Use of P2 half space expansion for angular flux is sufficient for most problems. Use of higher order expansions is yet to be studied which is easily possible using the present code.
[ top ]
[ top ]
IAEA0929/01, included references:
- P. Mohanakrishnan :Choice of Angular Current Approximations for Solving Neutron
Transport Problems in 2-D by Interface Current Approach.
Reprint from "Annals of Nuclear Energy", Vol.9 (1982), pp. 261-274
- P. Mohanakrishnan :
A Guide to the Use of Computer Code - RICANT.
RG/RPD-311 (December 1987)
[ top ]
IAEA0929/01
To run the test case on a VAX 8810, about 400K bytes of main storage are required.[ top ]
IAEA0929/01
VMS V5.1 (VAX 8810).[ top ]
[ top ]
[ top ]
IAEA0929/01
| File name | File description | Records |
|---|---|---|
| IAEA0929_01.001 | Ricant information file | 127 |
| IAEA0929_01.002 | Ricant FORTRAN source file | 1569 |
| IAEA0929_01.003 | Ricant sample problem input | 33 |
| IAEA0929_01.004 | Ricant sample problem output | 351 |
Keywords: neutron transport equation, two-dimensional, x-y.