Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 4-8
Author(s):  
M. N. Zizin ◽  
V. F. Boyarinov ◽  
V. A. Nevinitsa ◽  
P. A. Fomichenko ◽  
Yu. N. Volkov ◽  
...  
Keyword(s):  
Cross Sections ◽  
Nuclear Reactors ◽  
Reaction Rates ◽  
Automatic Generation ◽  
Hydraulic Calculation ◽  
Monte Carlo Code ◽  
Fission Reaction ◽  
Development Environment ◽  
Transient Conditions ◽  
Control Rods

Abstract Coupled neutronic and thermal hydraulic calculation codes are verified for calculating the design of modern and prospective types of nuclear reactors. This verification is done by comparing experimental and calculated results for stationary and transient conditions. This paper presents ShIPR (Shell of Intelligent Package for Reactor) Integrated Development Environment with automatic generation of head programs based on the chain of computational modules. The aim of this study is to find the reason of a discrepancy in the modelling of sub-critical states that was found in previous work. The comparison of ShIPR stationary module with Monte-Carlo code (MCU) and experimental results on ASTRA HTGR critical facility was presented in the paper. To compare the detector readings and MCU calculation with the ShIPR module, the interpolation cross-section procedure was performed. This procedure allows simulating 235U fission reaction rates (detector readings) in the complicate, annular core, using the micro-cross sections prepared by the cell-lattice code. We found that the calculation accuracy of stationary ShIPR module is on an acceptable level but the macro constants for control rods need to be prepared independently, given with surroundings.

scholarly journals Uncertainty Propagation Analysis for PWR Burnup Pin-Cell Benchmark by Monte Carlo Code McCARD

2012 ◽  
Vol 2012 ◽  
pp. 1-6 ◽  
Author(s):  
Ho Jin Park ◽  
Hyung Jin Shim ◽  
Chang Hyo Kim
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Sampling Method ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Monte Carlo Code ◽  
Uncertainty Propagation Analysis ◽  
Propagation Analysis ◽  
Cross Correlations

In the Monte Carlo (MC) burnup analyses, the uncertainty of a tally estimate at a burnup step may be induced from four sources: the statistical uncertainty caused by a finite number of simulations, the nuclear covariance data, uncertainties of number densities, and cross-correlations between the nuclear data and the number densities. In this paper, the uncertainties ofkinf, reaction rates, and number densities for a PWR pin-cell benchmark problem are quantified by an uncertainty propagation formulation in the MC burnup calculations. The required sensitivities of tallied parameters to the microscopic cross-sections and the number densities are estimated by the MC differential operator sampling method accompanied by the fission source perturbation. The uncertainty propagation analyses are conducted with two nuclear covariance data—ENDF/B-VII.1 and SCALE6.1/COVA libraries—and the numerical results are compared with each other.

scholarly journals BORON. PROPERTIES AND APPLICATION IN NUCLEAR ENERGY

2017 ◽  
pp. 91-94
Author(s):  
V. A. Palekha ◽  
A. A. Getman
Keyword(s):  
Nuclear Energy ◽  
Nuclear Reactors ◽  
Chemical Elements ◽  
Fission Reaction ◽  
Control Rods

Boron is one of the available chemical elements actively influencing the properties of the alloys. The main use of boron is in the alloys of the control rods of nuclear reactors to stop or slow down the fission reaction.

Hydraulic model of a water cooling system in a chemical plant

1988 ◽  
Vol 53 (4) ◽  
pp. 788-806
Author(s):  
Miloslav Hošťálek ◽  
Jiří Výborný ◽  
František Madron
Keyword(s):  
Flow Rate ◽  
Cooling System ◽  
Cooling Water ◽  
Graph Algorithm ◽  
Automatic Generation ◽  
Hydraulic Calculation ◽  
Return Flow ◽  
Flow Rates ◽  
Pressure Losses ◽  
Controlled Flow

Steady state hydraulic calculation has been described of an extensive pipeline network based on a new graph algorithm for setting up and decomposition of balance equations of the model. The parameters of the model are characteristics of individual sections of the network (pumps, pipes, and heat exchangers with armatures). In case of sections with controlled flow rate (variable characteristic), or sections with measured flow rate, the flow rates are direct inputs. The interactions of the network with the surroundings are accounted for by appropriate sources and sinks of individual nodes. The result of the calculation is the knowledge of all flow rates and pressure losses in the network. Automatic generation of the model equations utilizes an efficient (vector) fixing of the network topology and predominantly logical, not numerical operations based on the graph theory. The calculation proper utilizes a modification of the model by the method of linearization of characteristics, while the properties of the modified set of equations permit further decrease of the requirements on the computer. The described approach is suitable for the solution of practical problems even on lower category personal computers. The calculations are illustrated on an example of a simple network with uncontrolled and controlled flow rates of cooling water while one of the sections of the network is also a gravitational return flow of the cooling water.

ABUNDANCE ANOMALIES PRODUCED BY NUCLEAR REACTIONS IN STELLAR SURFACE LAYERS: I. NUCLEAR-REACTION RATES

1967 ◽  
Vol 45 (10) ◽  
pp. 3275-3296 ◽  
Author(s):  
P. J. Brancazio ◽  
A. Gilbert ◽  
A. G. W. Cameron
Keyword(s):  
Statistical Theory ◽  
Nuclear Reaction ◽  
Cross Sections ◽  
Nuclear Reactions ◽  
Level Density ◽  
Preliminary Investigation ◽  
Computer Time ◽  
Reaction Rates ◽  
Reaction Cross ◽  
Surface Layers

A preliminary investigation of the effects on abundances in stellar surfaces of extensive nuclear bombardment required the calculation of more than 105 nuclear-reaction cross sections. It was necessary to develop simplified methods for using the statistical theory of nuclear reactions to make these calculations in order that the computer time should not be prohibitive. These methods are described here and the results are compared with experiment. The accuracy of the calculations is, in general, about as good as, or somewhat better than, that obtained in previous applications of the statistical theory, probably because the use of an accurate level density formula outweighed the crudity of other approximations.

3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code

2013 ◽  
Author(s):  
Cécile-Aline Gosmain ◽  
Sylvain Rollet ◽  
Damien Schmitt
Keyword(s):  
Monte Carlo ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Reaction Rate ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Reaction Rates ◽  
Surveillance Program ◽  
Monte Carlo Code ◽  
Neutron Sources

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

scholarly journals Production of radioisotopes within a plasma focus device

2005 ◽  
Vol 20 (1) ◽  
pp. 33-37 ◽  
Author(s):  
Ergisto Angeli ◽  
Agostino Tartari ◽  
Michele Frignani ◽  
Vincenzo Molinari ◽  
Domiziano Mostacci ◽  
...  
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Plasma Focus ◽  
Reaction Rates ◽  
Experimental Production ◽  
Endogenous Production ◽  
The Us ◽  
Further Development ◽  
Radio Isotopes ◽  
Very High

In recent years, research conducted in the US and in Italy has demonstrated production of radioisotopes in plasma focus devices, and particularly, on what could be termed "endogenous" production, to wit, production within the plasma it self, as opposed to irradiation of tar gets. This technique relies on the formation of localized small plasma zones characterized by very high densities and fairly high temperatures. The conditions prevailing in these zones lead to high nuclear reaction rates, as pointed out in previous work by several authors. Further investigation of the cross sections involved has proven necessary to model the phenomena involved. In this paper, the present status of research in this field is re viewed, both with regards to cross section models and to experimental production of radio isotopes. Possible out comes and further development are discussed.

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