Journal Issue: Nuclear Energy and Technology
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Volume
9
Number
3
Issue Date
Journal Title
Journal ISSN
2452-3038
Том журнала
Том журнала
Статьи
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Economic advantages of starting up of inherently safe fast reactors with a closed fuel cycle on fortificated uranium
(2023) Orlov, M. A.
The publication substantiates the economic advantages of using in the starting loads of inherently safe fast reactors with a closed fuel cycle of enriched uranium instead of uranium-plutonium regenerate obtained by reprocessing of thermal reactors spent nuclear fuel (SNF). The justifications are given taking into account both the preliminary technical and economic assessments carried out by the basic enterprises of TVEL JSC and SHK JSC, and the neutron-physical and system-economic studies performed at the Private Institution of the ITCP Proryv (Breakthrough). It is shown that the starting-up of a fast reactor on enriched uranium instead of uranium-plutonium fuel, taking into account the costs of preliminary reprocessing of thermal reactors spent fuel, allows achieving a significant economic gain at the stage of construction and commissioning of nuclear power plants. It is also shown that even at moderately high values of the discount coefficient, the uranium start of a fast reactor with a closed fuel cycle is economically preferable in comparison with the option of starting on uranium-plutonium fuel from the positions of the break-even tariff.
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Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison
(2023) Chereshkov, D. G.; Тихомиров, Георгий Валентинович; Терновых, Михаил Юрьевич; Рыжков, Александр Александрович; Tikhomirov, G. V.; Ternovykh, M. Y.; Ryzhkov, A. A.
The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO2, MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead- and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.
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Effects of evaluated nuclear data libraries on the calculation results for fuel burnup with minor actinides in a VVER reactor
(2023) Karpovich, G. W.; Kazansky, Yu. A.; Bakhantsov, K. A.; Isanov, K. A.; Kushnir, N. O.
The paper deals with assessing the effects of the ENDF/B-VI.8, ENDF/B-VII.0, JEFF 3.1 and JEFF 3.1.1 nuclear data libraries on the results of calculating a number of functionals for a system based on a VVER reactor with fuel with a large fraction of minor actinides (up to 10%). Key estimates have been obtained for the errors introduced by libraries in calculations of systems with minor actinides (MA) based on a VVER-1200 reactor:
– for reactivity, σρ = 0.3 βeff;
– for isotopic compositions with minor actinides, ≤ 5% (the error for each particular isotope is different);
– for the total mass of accumulated MAs, εm = 0.8%.
Conclusions have been made with respect to the need for the further refinement of the library MA data proceeding from the nature of the calculation tasks that dictate the requirements for the accuracy of nuclear constants. It has been shown that systems based on VVER-1000/1200/1300 reactors with MAs need to be calculated using several libraries of evaluated nuclear data created at different organizations and based on the largest possible number of non-recurrent sets of experimental data.
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The effect of the importance function resolution on the accuracy of calculating the functionals of the neutron kinetics in water critical assemblies by Monte Carlo method
(2023) Arkhangelsky, D. M.; Daichenkova, Y. S.; Kalugin, M. A.; Oleynik, D. S.; Shkarovsky, D. A.
The paper considers a computational study of the importance function effect on the accuracy of calculating the effective fraction of delayed neutrons, βeff, and generation time of instantaneous neutrons using the MCU Monte Carlo code based on the example of three criticality experiments from the ICSBEP handbook. In the MCU code, the importance function has a piecewise constant form: the computational model is broken down into a finite number of registration objects, and the neutron importance is calculated in each. The obtained importance values are used then to calculate the kinetic functionals due to which the calculation accuracy for the latter depends on the resolution. Three types of the importance function spatial partition (axial, radial, combined) have been studied. The numerical simulation results have shown that the axial component of the neutron importance function in all experiments has practically no effect on the calculation accuracy for βeff and Λ: the difference between the obtained values is less than 1%. The radial component has a notable effect (of up to 15.9%) on the Λ calculation accuracy while having almost no effect on the βeff estimate. Using combined partition, as compared with radial partition, improves the calculation accuracy insignificantly (< 1%).
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Evaluation of neutronic performance for the VVER-1000 reactor core with regenerated uranium-plutonium fuel
(2023) Semishin, V. V.
The possibility has been considered for the VVER-1000 reactor fuel loading to be formed based on regenerated fuel with the use of the spent fuel accumulated in reactors of the same type. A study was undertaken to investigate the change in the isotopic composition of the plutonium discharged from a thermal reactor in the course of its multiple reprocessing and recycle in a thermal neutron reactor. To obtain an equilibrium isotopic composition of the reactor-grade plutonium, 3D neutronic calculations were performed for the stationary fuel cycles of a VVER-1000 serial reactor with conventional oxide fuel and oxide fuel based on regenerated uranium and based on an undivided mixture of uranium and plutonium oxides from SNF. The neutronic performance of reactor cores was compared for the above mentioned fuel types in the course of the fuel company, including the following: in-core radial power density shaping, values of reactivity coefficients for various thermal parameters, reactivity control system efficiency, etc.