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| | výběr: kategorie IYNC 2008| nalezeno záznamů: 130 | |
| [1] | A Four Group Reference Code for Solving Neutron Diffusion Equation in a VVER-440 Core| S. Saarinen | | Fortum Nuclear Services Ltd., Espoo, Finland | | Nuclear reactor core power calculation is essential in the analysis of the nuclear power plant and especially the core. Currently, the core power distribution in Loviisa VVER-440 core is calculated using nodal code HEXBU-3D and pinpower reconstruction code ELSI-1440 that solve the two group neutron diffusion equation. The computer power available has increased significantly during the last decades allowing us to develop a fine mesh code HEXRE for solving the four group diffusion equation. The diffusion equations are discretised using piecewise linear polynomials. The core is discretised using one node per fuel pin cell. The axial discretisation can be chosen freely. The boundary conditions are described using diffusion theory and albedos. Burnup dependence is modelled by tabulating diffusion parameters at certain burnup values and using interpolation for the intermediate values. A two degree polynomial is used for the modeling of the feedback effects. Eigenvalue calculation for both boron concentration and multiplication factor control has been formulated. A possibility to perform fuel loading and shuffling operations is implemented.HEXRE has been thoroughly compared with HEXBU-3D and ELSI-1440. The effect of the different energy and space discretisations used is investigated. Some safety criteria for the core calculated with the HEXRE and HEXBU-3D/ELSI-1440 have been compared. From the calculations (e.g. the safety criteria) we can estimate whether there exists systematic deviations in HEXBU- 3D/ELSI-1440 calculations or not. | kategorie:IYNCpublikováno:IYNC 2008, 21-26 September 2008, Interlaken, Switzerland - No.A1,1 | | umístění:| Tomáš Vytiska | | | | | | | |
| [2] | VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment| P. Borodkin, N. Khrennikov | | Scientific and Engineering Centre for Nuclear and Radiation Safety, Moscow, Russia | | Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime.
Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER -440 by MCNP-5 [1] code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT [2] code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. | kategorie:IYNCpublikováno:IYNC 2008, 21-26 September 2008, Interlaken, Switzerland - No.A1,2 | | umístění:| Tomáš Vytiska | | | | | | | |
| [3] | EPR: High Load Variation Performances with the TMode Core Control| F. Pairot | | AREVA NP | | The load variation performances on a PWR are directly linked to the core control design. This design is mainly characterized by the definition of the control rod banks and the way to both perform the banks movements and to modify the core boron concentration by injection of boric acid or water. The following paper presents the principles of the T mode, the new fully automatic core control mode for the EPR which provides high performance in terms of maneuverability and optimizes the effluents. First, the paper describes the division of the control rods into two control banks (Pbank for temperature and Hbank for power distribution). Then typical movements of these banks during power changes are shown. Then, the principle of the 3 control loops (Tave, AO, Pmax), used to obtain these desired control rod movements, is given. Finally, a load following transient simulation is presented. | kategorie:IYNCpublikováno:IYNC 2008, 21-26 September 2008, Interlaken, Switzerland - No.A1,3 | | umístění:| Tomáš Vytiska | | | | | | | |
| [4] | On the Calculation of Reactor Time Constants Using the Monte Carlo Method| J.I. Leppänen | | VTT Technical Research Centre of Finland | | Full-core reactor dynamics calculation involves the coupled modelling of thermal hydraulics and the time-dependent behaviour of core neutronics. The reactor time constants include prompt neutron lifetimes, neutron reproduction times, effective delayed neutron fractions and the corresponding decay constants, typically divided into six or eight precursor groups. The calculation of these parameters is traditionally carried out using deterministic lattice transport codes, which also produce the homogenised few-group constants needed for resolving the spatial dependence of neutron flux. In recent years, there has been a growing interest in the production of simulator input parameters using the stochastic Monte Carlo method, which has several advantages over deterministic transport calculation. This paper reviews the methodology used for the calculation of reactor time constants. The calculation techniques are put to practice using two codes, the PSG continuous-energy Monte Carlo reactor physics code and MORA, a new full-core Monte Carlo neutron transport code entirely based on homogenisation. Both codes are being developed at the VTT Technical Research Centre of Finland. The results are compared to other codes and experimental reference data in the CROCUS reactor kinetics benchmark calculation. | kategorie:IYNCpublikováno:IYNC 2008, 21-26 September 2008, Interlaken, Switzerland - No.A1,4 | | umístění:| Tomáš Vytiska | | | | | | | |
| [5] | Foreign Material Exclusion Program at CNE CERNAVODA Nuclear Generating Station| D. Urjan | | CNE Nuclearelectrica - Cernavoda Nuclear Power Plant | | In the face of a continuing attention to operations and maintenance costs at nuclear power plants, the future of the industry depends largely upon increasing plant availability and improving operating efficiency. The success in achieving these objectives is dependent upon the success of each plant's equipment maintenance program. Preventing the introduction of foreign materials into a nuclear power plant system or component requires a careful, thoughtful, and professional approach by all site personnel. This paper describes a proactive approach to prevent the introduction of foreign material into systems and components, by providing an overview of technical considerations required to develop, implement, and manage a foreign material exclusion program at CNE Cernavoda Unit 1&2 Nuclear Power Station. It is also described an example of Foreign Material Intrusion which happened during the 2003 planned maintenance outage at Cemavoda Unit 1. This paper also defines personnel responsibilities and key nomenclature and a means for evaluating prospective work tasks and activities against standardized criteria, in order to identify the appropriate level of the required FME controls. | kategorie:IYNCpublikováno:IYNC 2008, 21-26 September 2008, Interlaken, Switzerland - No.A5,1 | | umístění:| Tomáš Vytiska | | | | | | | |
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