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[en] The safety of fuel loading of VVER reactors is justified by calculations of the neutronic characteristics of the forthcoming campaign. These calculations are based on the design parameters of fuel assemblies (FA) — fuel enrichment, materials, design features, etc. However, during operation, some parameters change in an uncontrolled manner. In particular, FA can deform — bend or twist, this leads to the appearance of increased gaps between the fuel assemblies. These regions filled with a moderator lead to an increase in comparison with the calculations for the generation of thermal neutrons and, as a result, to a surge in the power of the fuel rods surrounding these regions. Safety requirements limit the power of fuel rods. Therefore, design capacities are increased by means of the so-called engineering margin factor to account for random outbursts. The deviation of the size of the water gaps between the fuel assemblies from the design ones should be known to calculate this coefficient, for example, the size distribution function of the, gaps. This information is most often obtained by modeling the mechanical state of fuel assemblies in the, reactor core. Other approaches are based on experimental data. Measurements in the core during discharging campaign are not possible. Therefore, the geometric parameters of the fuel assembly after the the, from the core are measured. The presented paper uses the data of such measurements obtained after 24th fuel campaign of ZNPP unit mechanical It is assumed that the fuel assemblies tend to retain the form they have in the “free” state, and analyzed., interaction with neighboring fuel assemblies leads to a certain equilibrium state that can be easily proposed In contrast to similar calculations, the elastic energy functional of interacting fuel assemblies is were, whose minimum gives the required size distribution function of the gaps. 24 and 25 campaigns studied., modelled; the role of inter-sector FA shuffling was the, The distribution of the gaps between the fuel assemblies in VVER-1000 core is calculated based on of, measured deformations of the fuel assemblies discharged from the core and the elastic characteristics not, the fuel assemblies. It was demonstrated that 95 % of gaps in the cores both with FA-A and FA-WR do the, exceed 7.6 mm. The results can be used to calculate the engineering margin factor in determining release., peaking factors of energy release.
[en] FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry that is carried out in the critical facility EOLE. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffle and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7*10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR. (authors)
[en] In this paper, a spatial conductometric measuring system is presented, with which it is possible to study the mixing of flows in equipment of complex geometry. The paper presents a description of the experimental conditions, including a matrix of operating parameters for which studies have been carried out, as well as the results of the experiments
[ru]В данной работе представлена пространственная кондуктометрическая измерительная система, при помощи которой можно исследовать смешения потоков в оборудовании сложной геометрии. В работе представлены описание проведенных экспериментальных режимов, в том числе показана матрица режимных параметров, при которых проводились исследования, а также полученные результаты
[en] The general objective of the CORONA II project is to enhance the safety of nuclear installations through further improvement of the training capabilities for providing the necessary personnel competencies in VVER area. More specific objective of the project is to continue the development of a state-of-the-art regional training network for VVER competence called CORONA Academy. The project aims at continuation of the European cooperation and support in this area for preservation and further development of expertise in the nuclear field by improvement of higher education and training. The efforts are focused on using the most advanced techniques of providing training to the trainees, saving cost and time – distance learning and e-learning approaches which will be tested during CORONA II Project implementation. The knowledge management portal will integrate the information on VVER web into a single communication system and develop and implement a semantic web structure to achieve mutual recognition of authentication information with other databases. That will enable the partners to share the materials available in each specific training center. This project has received funding from the EURATOM research and training programme 2014-2018 under grant agreement No 662125. (author)
[en] Highlights: • 153Sm production in MNSRs investigated trough multi-stage approach. • A good agreement between simulation and experiment was founded (6%). • Production of 153Sm reached 852.26 mCi g−1 in 20 successive cycles. • The results are promising for the 153Sm production in MNSRs reactors. - Abstract: The main objective of this study was to explore the feasibility of producing 153Sm radioisotope in miniature neutron source reactors (MNSRs) in Isfahan-Iran. As the first step of this study, the MNSR's geometry was created by using the MCNP6.2 simulation code and afterwards a validity check was performed by comparing the results with the experimental data. Then, by applying values obtained through simulation, the production process was followed up to 20 irradiation cycles using different irradiation and cooling periods (irradiation setups). The results showed that the proposed simulation technique has an acceptable agreement with the experiments (with a difference of nearly 6%). In spite of limitations, such as irradiation time and flux in such reactors, our results showed that by choosing the correct irradiation setup, it is possible to produces 153Sm up to 852.26 mCi g−1 in 20 successive irradiation cycles. However, after the 10th cycle, the production reached 90% of the maximum point. Nevertheless, the continuance of the irradiation process with a new target (by 10 plus 10 discrete irradiation) can double the total activity in comparison with 20 successive irradiation cycles, without any increase in the fuel consumption of the reactor. These findings increase the prospect of a large-scale production of the life-saving 153Sm radioisotope in MNSR reactors.
[en] The article refers to the “steady-state crisis” experiment carried out in the channel of the MIR reactor, in which a heat transfer crisis has been recorded on the three-element assembly of shortened VVER-1000 fuel rods (the length of the fuel column is 1000 mm) at parameters close to the calculated values. The experimental data are used to calculate the temperature conditions for testing fuel assemblies in the MIR reactor
[ru]В статье говорится о проведенном в канале реактора МИР эксперименте “Кризис стационарный”, в котором на трехэлементной сборке укороченных твэлов ВВЭР-1000 (длина топливного столба 1000 мм) при параметрах, близких к расчетным значениям, зафиксирован кризис теплоотдачи. Данные эксперимента используются для расчета температурных условий испытания сборок твэлов в реакторе МИР
[en] Today, objective preconditions have been formed to find the ways on how to increase cost-effectiveness of NPPs operation, while providing the required safety level. One of such ways to increase thermal nominal power of power unit. The paper provides for the results of reactor behavior analysis at increased thermal power above nominal received using a one-dimensional system computer code RELAP5/MOD3.2 and relevant model of VVER-1000 (V-320) power unit. Calculation analyses are performed for quasi-static reactor operating conditions and transients using realistic approach in terms of initial performance parameters of reactor installation. In researches, representative initial events for transients have been selected according to the principle described further. For an abnormal operation, an event has been selected based on its high frequency and consequences, which require decreasing reactor power down to 50 % of nominal thermal power. For emergency conditions an event has been selected which is caused by external extreme impacts typical for Ukrainian NPP sites resulting in the worst consequences. Thus, the transients are represented by events associated with failure of a single turbine-driven feed water pump and total station blackout unit. To analyze emergency conditions caused by long-term blackout, they were additionally accompanied by a leakage through reactor coolant pump seals. Given that increase of steam flow in a turbine at increased thermal power above nominal requires additional studies on residual service life assessment of its critical components, a 3-D model of high-pressure rotor of a full speed turbine is proposed for further studies. Based on the calculations a comparative analysis of major parameters of the reactor at rated and increased thermal power is performed with assessment of significant factors to be considered in further studies on increase of installed thermal output of NPP unit.
[en] The development and implementation of new design, methodical and engineering solutions allows enlarging the capabilities in testing structural materials and FA components of water-cooled reactors to justify their performance and to get experimental data to verify and certify calculation codes. In particular, the developed equipment and methodical approaches to the testing of structural materials and FA components of water-cooled reactors in the MIR reactor. lt allows data to be obtained on the kinetics of changes in the mechanical properties of structural materials under irradiation. Moreover, a possibility to adjust and control the coolant water chemistry during irradiation and to provide the required neutronic and thermo-hydraulic parameters allow for a more full compliance to the testing conditions of structural materials and FA full-size components.