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Aleksandra Luks, Krzysztof Pytel, Mikołaj Tarchalski, Nikołaj Uzunow and Tomasz Krok

Abstract

Results of numerical calculations of heat exchange in a nuclear heating detector for nuclear reactors are presented in this paper. The gamma radiation is generated in nuclear reactor during fission and radiative capture reactions as well as radioactive decay of its products. A single-cell calorimeter has been designed for application in the MARIA research reactor in the National Centre for Nuclear Research (NCBJ) in Świerk near Warsaw, Poland, and can also be used in the Jules Horowitz Reactor (JHR), which is under construction in the research centre in Cadarache, France. It consists of a cylindrical sample, which is surrounded by a gas layer, contained in a cylindrical housing. Additional calculations had to be performed before its insertion into the reactor. Within this analysis, modern computational fluid dynamics (CFD) methods have been used for assessing important parameters, for example, mean surface temperature, mean volume temperature, and maximum sample (calorimeter core) temperature. Results of an experiment performed at a dedicated out-of-pile calibration bench and results of numerical modelling validation are also included in this paper.

Open access

Sanduni Y. Ratnayake, Anoma K. Ratnayake, Dieter Schild, Edward Maczka, Elzbieta Jartych, Johannes Luetzenkirchen, Marek Kosmulski and Rohan Weerasooriya

,2-dibromo-3-chloropropane (DBCP) and nitrate by iron powder and by H 2 /Pd/Al 2 O 3 . In American Chemical Society National Meeting, Washington, DC, April 2–6, 1995. American Chemical Society. 16. Ratnayake, S., Schild, D., Maczka, E., Jartych, E., Luetzenkirchen, J., Kosmulski, M., Makehelwala, M., Weragoda, S. K., Bandara, A., Wijayawardana, R., Chandrajith, R., Indrarathne, S. P., & Weerasooriya, R. (2016). A novel radiation-induced grafting methodology to synthesize stable zerovalent iron naoparticles at ambient atmospheric conditions. Colloid Polym. Sci

Open access

Grzegorz Kępisty and Jerzy Cetnar

Abstract

In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5). The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR) system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

Open access

Manuel Barrera, Melquiades Casas-Ruiz, José J. Alonso and Juan Vidal

Abstract

A methodology to determine the full energy peak efficiency (FEPE) for precise gamma spectrometry measurements of environmental samples with high-purity germanium (HPGe) detector, valid when this efficiency depends on the energy of the radiation E, the height of the cylindrical sample H, and its density ρ, is introduced. The methodology consists of an initial calibration as a function of E and H and the application of a self-attenuation factor, depending on the density of the sample ρ, in order to correct for the different attenuation of the generic sample in relation to the measured standard. The obtained efficiency can be used in the whole range of interest studied, E = 120–2000 keV, H = 1–5 cm, and ρ = 0.8–1.7 g/cm3, being its uncertainty below 5%. The efficiency has been checked by the measurement of standards, resulting in a good agreement between experimental and expected activities. The described methodology can be extended to similar situations when samples show geometric and compaction differences.

Open access

Mikołaj Oettingen and Przemysław Stanisz

Abstract

This paper describes the methodology developed for the numerical reconstruction and modelling of the thorium-lead (Th-Pb) assembly available at the Department of Nuclear Energy, Faculty of Energy and Fuels, AGH University, Krakow, Poland. This numerical study is the first step towards integral irradiation experiments in the Th-Pb environment. The continuous-energy Monte Carlo burnup (MCB) code available on supercomputer Prometheus of ACK Cyfronet AGH was applied for numerical modelling. The assembly consists of a hexagonal array of ThO2 fuel rods and metallic Pb rods. The design allows for different arrangements of the rods for various types of irradiations and experimental measurements. The intensity of the fresh neutron source intended for integral experiments is about 108 n/s, which corresponds to the mass of about 43 μg 252Cf. The source was modelled in the form of Cf2O3-Pd cermet wire embedded in two stainless steel capsules.

Open access

Mikołaj Oettingen and Jerzy Cetnar

Abstract

In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB) in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238) and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242). The main results were presented as a calculated-to-experimental ratio (C/E) for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55). The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

Open access

Wojciech Głuszewski, Bartłomiej Boruc, Hieronim Kubera and Dinara Abbasowa

Abstract

Radiation preservation of objects of historical significance is an interesting proposition for museums, archives, libraries and private collectors. In this paper, we have limited ourselves to studying the effects of ionizing radiation on the paper. The radiation resistance of various grades of paper was examined in INCT. Irradiations were done by electron beam (10 MeV, 10 kW) and by gamma radiation (7 kG/h), for the purpose of comparison. Yields of hydrogen and absorption of oxygen were determined by gas chromatography (GC). For this purpose, the first time in an original way was used diffuse reflection spectroscopy (DRS). Described as the dose, dose rate, and lignin were found to affect degradation processes of cellulose. Examined the protective effect of lignin in the process of radiation degradation of paper. Proposed research methodology can be successfully applied to study other materials relevant to the conservation of works of art.

Open access

Przemysław Stanisz, Jerzy Cetnar and Mikołaj Oettingen

Abstract

The highest efficiency in the usage of nuclear energy resources can be implemented in fast breeder reactors of generation IV. It is achieved thanks to the ability of consuming minor actinides (MAs) in energy production. One of the options to use this benefit is full recycling of MAs to close the nuclear fuel cycle. Monte Carlo burn up (MCB), an integrated burn-up calculation code, deals with the complexity of the burn-up process which is applied to the European Lead-cooled Fast Reactor (ELFR). MCB uses continuous energy representation of cross section and spatial effects of full core reactor model; however, it automatically calculates nuclide production in all possible reactions or decay channels. Multi-recycling of MAs can cause an intensified build-up of curium, berkelium and californium. Some of their isotopes are strong neutron emitters from spontaneous fission, which hinders fuel recycling. The implementation of a novel methodology for trajectory period folding allows us to trace the life cycle of crucial MAs from the beginning of the reactor life towards the state of adiabatic equilibrium. The result of the analysis performed is presented, showing the sources of strong contribution to the neutron production rate. The parametric sensitivity analysis method for selected nuclide reactions is applied, revealing sensitivity of transmutation chains for the production of neutron emitter isotopes.

Open access

Przemysław Stanisz, Jerzy Cetnar and Grażyna Domańska

Abstract

The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR) was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System) and LEADER (Lead-cooled European Advanced Demonstration Reactor) projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA) are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs), and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB) code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

Open access

Tomasz Bury

). Applicability of AREVA NP containment response evaluation methodology to the U.S. EPR ™ for large break LOCA analysis . AREVA NP Inc. (Technical Report No. 10299-NP, Revision 1). 11. Kostka, P., Techy, Z., & Sienicki, J. (2002). Hydrogen mixing analyses for a VVER containment. In Proceedings of 10th International Conference on Nuclear Engineering, 14–18 April 2002 (paper ICONE10-22206). Arlington, Virginia, USA. 12. AREVA. (2006). U.S. EPR severe accident evaluation topical report . AREVA NP Inc. (ANP-10268NP, Revision 0). 13. OECD/NEA. (1999). State