NEA-0625 DINE.
DINE, Neutron Flux, Neutron Dose Rate in Multi-Region Slab Reactor Shield by Removal Diffusion
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROBLEM 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 AUTHOR, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| DINE | NEA-0625/01 | Report | 10-FEB-1993 |
Machines used:
No item found
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3. DESCRIPTION OF PROBLEM OR FUNCTION
DINE is a multigroup removal diffusion program which calculates the neutron flux and dose rate distribution in a reactor shield with multiregion slab geometry due to a cylindrical reactor core. From the integral transport equation for linear anisotropic scattering a difference equation has been derived, in which the mesh distance is proportional to the mean free path. This method gives an improvement of the deep penetration of diffusing neutrons.
DINE is a multigroup removal diffusion program which calculates the neutron flux and dose rate distribution in a reactor shield with multiregion slab geometry due to a cylindrical reactor core. From the integral transport equation for linear anisotropic scattering a difference equation has been derived, in which the mesh distance is proportional to the mean free path. This method gives an improvement of the deep penetration of diffusing neutrons.
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4. METHOD OF SOLUTION
The removal flux is calculated in the same lethargy group structure and at the same mesh points as used in the multigroup diffusion calculation following. To calculate the re- moval flux, numerical integration of the first collision source equation is carried out over the reactor core.
The essence of the method for the treatment of the diffusion equa- tion is that on the integral diffusion equations for the flux and current, a difference operation with respect to the space coordin- ates is performed, due to which the series expansion of the scat- tering integral becomes much more convergent and can be approximated more easily.
The removal flux is calculated in the same lethargy group structure and at the same mesh points as used in the multigroup diffusion calculation following. To calculate the re- moval flux, numerical integration of the first collision source equation is carried out over the reactor core.
The essence of the method for the treatment of the diffusion equa- tion is that on the integral diffusion equations for the flux and current, a difference operation with respect to the space coordin- ates is performed, due to which the series expansion of the scat- tering integral becomes much more convergent and can be approximated more easily.
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8. RELATED AND AUXILIARY PROGRAMS
DINE can use the output of TEDDI-M - a two-dimensional diffusion code for core calculations, to de- termine the neutron source distribution in the core and the flux values at the core boundary.
RAGA - a Monte Carlo code which calculates the transport of cap- ture gamma rays, can use the reaction rates in the shield delivered by DINE.
DINE can use the output of TEDDI-M - a two-dimensional diffusion code for core calculations, to de- termine the neutron source distribution in the core and the flux values at the core boundary.
RAGA - a Monte Carlo code which calculates the transport of cap- ture gamma rays, can use the reaction rates in the shield delivered by DINE.
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10. REFERENCES
- K.A. Verschuur:
Approximation of the Transport Equation with Difference Equations, Proc. of International Conf. on the Physics Problems of Reactor
Shielding, Harwell (Sept. 1967) AERE-R 5773, Vol. 1,
Paper RS/1.1/6.
- K.A. Verschuur:
DINE - A One-Dimensional Neutron Shielding Program with Physically Determined Space Mesh Distances and Linear Anisotropic Scattering
RCN-153, Petten (June 1971).
- K.A. Verschuur:
Approximation of the Transport Equation with Difference Equations, Proc. of International Conf. on the Physics Problems of Reactor
Shielding, Harwell (Sept. 1967) AERE-R 5773, Vol. 1,
Paper RS/1.1/6.
- K.A. Verschuur:
DINE - A One-Dimensional Neutron Shielding Program with Physically Determined Space Mesh Distances and Linear Anisotropic Scattering
RCN-153, Petten (June 1971).
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Keywords: anisotropic scattering, dose rates, multigroup, neutron diffusion equation, neutron flux, one-dimensional, reactor cores, removal-diffusion, shielding.