Using Green's Functions to Compute 3 Dimensional Neutron Flux Maps - 16016
- 1. EDF-DP2D, 154 avenue Thiers, CS 60018, 69458 Lyon Cedex 06 (France)
Description
The knowledge of the radionuclide content of radioactive waste is of utmost importance for safety and waste management reasons. Numerical simulations are used by EDF-DP2D to anticipate dismantling and radioactive waste management. The main numerical activation schemes include the computing of a 3 dimensional neutron flux map, obtained on the basis of a neutron propagation calculation. The codes used solve the transport equation called the Boltzmann equation without simplification. The Monte Carlo method is used to solve the Boltzmann equation. Random series of numbers are used to simulate the lives of millions of neutrons. The codes follow each neutron individually, from its birth to its disappearance by leakage, absorption, or fission, and then the neutron flux map is calculated at the nominal power rating conditions. The main drawback of the Monte Carlo codes is the large computing time needed to converge within acceptable statistical criteria. This extended time period is incompatible with the multiple fuel managements used by EDF to operate its NPPs because it is necessary to make a Monte Carlo calculation for each fuel loading pattern. The solution developed by EDF-DP2D is to use the importance factors of Green's Functions to simplify the computing of the 3 dimensional neutron flux maps. Physicists talk about Green's Functions where mathematicians talk about elementary solutions of linear equations with constant coefficients. With an activation scheme based on the Green's Functions approach the 3D neutron flux is obtained on the basis of the convolution of the normalized source contributions (i.e. normalized response matrices) with the neutrons emitted by the fuel assemblies. The response matrices to unitary sources are computed with the Monte Carlo code MCNP for different tallies of interest and the neutron sources emitted by the fuel are computed with a core code for different fuel loading patterns. The principal interest of this approach is the calculation of the response matrices to unitary sources that can be used for all power distributions, thus limiting the Monte Carlo computing time. (authors)
Availability note (English)
Available from: WM Symposia, Inc., PO Box 27646, 85285-7646 Tempe, AZ (US)Additional details
Identifiers
Publishing Information
- Imprint Pagination
- 14 p.
- Report number
- INIS-US--19-WM-16016
Conference
- Title
- 42. Annual Waste Management Symposium
- Acronym
- WM2016
- Dates
- 6-10 Mar 2016
- Place
- Phoenix, AZ (United States)
INIS
- Country of Publication
- United States
- Country of Input or Organization
- France
- INIS RN
- 50083287
- Subject category
- S12: MANAGEMENT OF RADIOACTIVE WASTES, AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- ABSORPTION; BOLTZMANN EQUATION; COMPUTERIZED SIMULATION; EMISSION; FISSION; FUEL ASSEMBLIES; FUEL MANAGEMENT; MATHEMATICAL SOLUTIONS; MONTE CARLO METHOD; NEUTRON FLUX; NEUTRON SOURCES; NEUTRONS; NUCLEAR FUELS; NUCLEAR POWER PLANTS; RADIOACTIVE WASTE MANAGEMENT; RADIOACTIVE WASTES; RADIOISOTOPES; REACTOR SAFETY; THREE-DIMENSIONAL CALCULATIONS; TRANSPORT THEORY
- Descriptors DEC
- BARYONS; CALCULATION METHODS; DIFFERENTIAL EQUATIONS; ELEMENTARY PARTICLES; ENERGY SOURCES; EQUATIONS; FERMIONS; FUELS; HADRONS; INTEGRO-DIFFERENTIAL EQUATIONS; ISOTOPES; KINETIC EQUATIONS; MANAGEMENT; MATERIALS; NUCLEAR FACILITIES; NUCLEAR MATERIALS MANAGEMENT; NUCLEAR REACTIONS; NUCLEONS; PARTIAL DIFFERENTIAL EQUATIONS; PARTICLE SOURCES; POWER PLANTS; RADIATION FLUX; RADIATION SOURCES; RADIOACTIVE MATERIALS; REACTOR MATERIALS; SAFETY; SIMULATION; SORPTION; THERMAL POWER PLANTS; WASTE MANAGEMENT; WASTES
Optional Information
- Notes
- 3 refs.; available online at: http://archive.wmsym.org/2016/index.html