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Monte-Carlo Nuclear Reactor Code

#1086


Monte-Carlo Code Development for Spatial Neutron Transient Calculations of NPP Core

Tech Area / Field

  • FIR-MOD/Modelling/Fission Reactors

Status
8 Project completed

Registration date
15.08.1997

Completion date
10.04.2007

Senior Project Manager
Pradas-Poveda J I

Leading Institute
VNIIEF, Russia, N. Novgorod reg., Sarov

Project summary

It is supposed to develop a code for calculation of nuclear power plant cores neutron dynamics by Monte-Carlo technique. (TDMCC: Time-Dependent Monte-Carlo Code).

In the latest decade, for calculating neutron transient processes in nuclear power plant (NPP) cores, high development has been achieved by codes based on three-dimensional group (by neutrons' energies) diffusion approximation. Coming-to-be and development of such codes became a significant headway in increasing accuracy of describing processes taking place in NPP cores. Previously, more simple models were used for description of NPP core dynamics, starting from "point" model and up to two-dimensional multi-layered one. Transition to more accurate three-dimensional group diffusion model became possible owing to development of computational equipment. Let's dwell on the features of this model in more detail.


1. An NPP core is pided into cells. The calculation mesh is constructed for the core, each of its points or elements belonging to a certain cell. A cell is characterized by its group neutron constants. The state of each cell changes versus time in many parameters (fuel and coolant temperatures, coolant density, poisoning, burnup), that's why preliminary calculation of neutron constants is necessary for all types of cells in the state mesh. In many cases, consideration of adjacent cells' influence upon the constants of the cell being calculated is needed. Wide-range multi-dimensional space interpolation program with controllable accuracy is also required. The work on preparation of dynamic constants is performed beforehand for a certain reactor type and require large computation and calendar time outlay (it may take about one year of calendar time to develop a good library of wide-range dynamic neutron constants).
2. In calculation of group cellular constants, approximate physical models are used, including calculation of resonance region.
3. The diffusion approximation is quite applicable for the most of calculations of reactor plant (RP) cores neutron dynamics, but its accuracy is questionable, in particular, in the following cases:

- small reactors with large leakage, e.g. cosmic reactors;
- situations with local (or non-local) loss or boil-up of coolant in a RP core;
- calculations with core detailization.


In order to obtain acceptable accuracy in calculations of such situations or, at least, to test techniques, programs are needed based not on diffusion, but on transport (kinetic) approximation.

Transition from 3D group diffusion approach to solution of 3D group transport equation eliminates the greater part of physical approximations, but preserves a- number of approximations used in preparation of group cellular constants, like homogenization, consideration of adjacent cells, cross-section processing in resonance region, etc. These and other approximations require thorough and careful approach in each case. Interpolation of cellular constants also requires careful and painstaking work every time.

As mentioned before, preparation of dynamic constants requires large amount of calculations and further processing, which may take a calendar period up to half a year and more. This long and effort-consuming procedure is absolutely absent if Monte-Carlo techniques are used. All neutron elementary isotopic constants, existing in elementary evaluated neutron data libraries and included in a Monte-Carlo code library can be used in calculations without any modification.

Previously, no code development work on solution of non-stationary 3D transport equation by Monte-Carlo techniques was conducted for RP dynamics calculations, since there were no sufficient computer resources. Now the computers appeared with >109 floating-point operations per second productivity and the task of developing such a code became actual and scientists start to show interest in it /3/.

As a result of the Project implementation, a code will be developed and tested for solving non-stationary three-dimensional transport equation by Monte-Carlo technique and core burnup kinetics for transient-process calculations (in both operation and accident modes) in reactor plant cores. The code can be delivered to RSIC computer code & data collection.

The code to be developed will permit to solve all major problems in the area of modeling dynamic neutron processes in RP cores, towards which the main efforts of scientists are directed at present:


- physical approximations, used in physical models and codes, will be minimized;
- the code will be suitable for operative conduction of steady-state and transients calculations of any kind of RP.

As a result of the work the rich experience on the development of Monte-Carlo codes for multiprocessor computers will be obtained. Evidently, the code will be suitable and convenient for operation on Super-computers with 1012 floating-point operations per second productivity.

Implementation of the Project will permit to redirect efforts of 29 scientists and 3 specialists involved in weapons development, to solution of the important problem of ensuring atomic power plant safety.

Technical approach

Vast experience is used of code development for numerical modeling of complex processes of neutron dynamics in reactor plant cores by both Monte-Carlo and difference methods.

As a basis for the code to be created, the existing C-90 code will be used /1/. A well-known MNCP code /2/ is the most close to it, in terms of possibilities and purpose.

Potential role of foreign collaborators could be to choose proper guiding lines and trends of the work.

References


1. "Monte-Carlo Technique at VNDEF", N.V. Donskoy, V.A. Yeltsov, A.K. Zhitnik et al. VANT (Mathematical Modeling of Physical Processes), 1993, issue 2, pp. 61-64.
2. "MCNP-A General Monte Carlo Code for Neutron and Photon Transport". LA-7396-M.
3. "Progress in Time-Dependent, Three-Dimensional Neutron Transport Method Development", C.L. Bentley, S. Goluoglu, M.E. Dunn, R.E. Pevey, and H.L. Dodds, Proceedings of the International Conference on the Physics of Reactors, Mito, Ibaraki, Japan, (September 16-20, 1996).


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