Neutron-physical Characteristics Research Articles

VVER-440 fuel cycles based on fuel assemblies with larger uranium loads and higher uranium content

A long series of computational works validating possibilities for improving of VVER-440 fuel cycles has been completed at the Russian Science Center Kurchatov Institute. Work was performed on upgrading and improving VVER-440 fuel cycles on the basis of second-generation fuel assemblies which make it possible to increase the thermal power to 107-108% nominal 1375 MW(t) while preserving the operating time of the fuel loads. Fuel cycles based on second-generation fuel assemblies with a larger fuel load and average fuel enrichment to 4.87% are now under development. The fuel load in fuel assemblies is enlarged by eliminating the central opening in the fuel pellet of a fuel element and increasing the pellet diameter by decreasing the cladding thickness. The present article summarizes the work done on upgrading VVER-440 fuel cycles and shows the possibility and prospects for further improving the neutron-physical and operational characteristics of reactors by additionally optimizing fuel-assembly design and fuel-load arrangement.

Modeling of the fuel burnup of research reactors on switching to low enrichment

The modeling of the neutronics of fuel assemblies on switching research reactors to low-enrichment fuel is examined. The specifications and computational results for three test problems of modeling the isotopic composition of spent nuclear fuel from a research reactor are presented. The test problems simulate fuel burnup in TVS-2M and -4M fuel assemblies of the IRT-1 reactor. It is shown that the contribution of plutonium isotopes to the energy release and their role in the neutron-physical characteristics of TVS-4M with deep burnup increase considerably on switching to low-enrichment fuel.

Heat- and neutron-physical characteristics of power-generating units of nuclear power plants with VVER-1000 during operation above nominal power

The margin to critical level of heat exchange of thermal power and coolant temperature at the core exit as well as the nonuniformities of the energy release under the operating conditions of a nuclear power plant with VVER-1000 above nominal power is analyzed for the No. 2 unit of the Balakovo nuclear power plant. It is confirmed that the safety criteria comply with OPB-88/97. The content of each of the three main stages of operation is presented. Conclusions are drawn and recommendations are made concerning updating the technical design of the reactor facility in order to increase the power level.

MCNP calculation of the neutron-physical characteristics of a reactor and radiation protection in a nuclear power facility for use in spacecraft

MCNP calculation of the neutron-physical characteristics of a reactor and radiation protection in a nuclear power facility for use in spacecraft

Neutron-physical characteristics and fuel utilization in a sodium-cooled fast reactor with high-density fuel at the initial stage of the transition to a closed cycle

The results of neutron-physical investigations of a fast reactor using high-density fuel (UC) at the initial stage of the transition to a closed fuel cycle are presented. Validation is given for the possibility of making the transition to a closed fuel cycle with self-supply of fissile nuclei starting with the first recycle.

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis.

MCU-FCP program for first-collisions probability calculations of neutron transport

A general scheme, methods, and algorithms, implemented in the MCU-FCP program, making it possible to calculate the neutron-physical characteristics of two-and three-dimensional RBMK cells and polycells taking account of fuel burnup during reactor operation by the first-collisions probability method, are described. Test calculations are preformed and the results are compared with the MCU-REA program.

Neutron-physical characteristics of a VVER-1000 core with 100% fuel load consisting of a mixture of recovered uranium and plutonium and enriched uranium

The results of a calculation of the neutron-physical characteristics of different variants of the fuel load of a VVER-1000 core are presented. The load completely consists of either enriched natural uranium (standard fuel) or recovered fuel, consisting of a uranium-plutonium mixture of recovered material, recovered plutonium, and highly enriched uranium. The mass fraction of plutonium isotopes in the recovered fuel ranged from 2 to 5%. All load variants examined have the same energy potential. The neutron-physical characteristics of a core with a full load of uranium are compared for a core with recovered uranium-plutonium fuel. Analysis of the neutron-physical characteristics did not show any fundamental limitations on the possibility of a VVER-1000 core load consisting entirely of recovered uranium-plutonium fuel.

Effect of 3He and 6Li accumulation in beryllium blocks on the neutron-physical characteristics of the MIR reactor

Reactions resulting in the accumulation of 3He and 6Li, whose thermal neutron capture cross-section is large, occur under the action of neutron radiation in the beryllium blocks of the MIR reactor core. When a neutron absorber accumulates in the moderator of a reactor, important physical characteristics change: reactivity access, efficiency of safety and control rods, and reactivity effects; in addition, energy release is redistributed. An algorithm for calculating 3H, 3He, and 6Li in each beryllium block of the core has been developed and implemented. This algorithm makes it possible to follow the change in the concentration of these nuclides during reactor operation and shutdown. The 3He and 6Li concentrations are used as initial data for calculating the neutron-physical characteristics of the MIR reactor using the MCU and BERCLI programs. The computational results for the effect of the accumulation of the nuclides indicated on the neutron-physical characteristics of the core are presented.

Evaluated critical benchmark VVéR physics experiments at the Russian Science Center Kurchatov Institute

To develop new-generation enhanced-safety VVER-1200 reactor with improved technical performance and cost effectiveness for AES-2006, it is necessary to increase the accuracy of the programs used to calculate the neutron-physical characteristics. Benchmark experiments with a detailed description of the geometry and material composition of the configurations studied are needed to verify such computational programs. Sixty-six evaluated critical experiments on VVER physics performed on benches at the Russian Science Center Kurchatov Institute are presented. They are combined into six assessments; they have passed an analysis by an international group of experts; and, they are included in the International Collection of Evaluated Benchmark Experiments ICSBEP.

Use of weapons plutonium in VVÉR-1000 reactors

One scenario for using excess Russian weapons plutonium is to load it into VVER-1000 reactors. It is proposed that up to 40% of the fuel assemblies with uranium fuel be replaced with structurally similar fuel assemblies with mixed uranium-plutonium fuel. The stationary regime for burning fuel has the following characteristics: the run time is about 300 or 450 eff. days, the yearly plutonium consumption reaches 450 kg, the neutron-physical characteristics are close to the corresponding regimes with uranium fuel. The nuclear safety criteria and the irradiation dose for workers handling fresh and spent mixed fuel remain within the limits of the normative values. The use of mixed fuel makes it necessary to upgrade certain systems at nuclear power plants. A substantial quantity of weapons plutonium can be loaded every year into VVER-1000 reactors, effectively using the energy potential of this plutonium.

Multipurpose electron accelerator driven electronuclear system based on a subcritical cascade reactor

An electronuclear system consisting of an electron accelerator, a neutron-producing target, and dual-zone subcritical blanket with fast and thermal neutron spectra is proposed. Some general mechanisms of cascade multiplication of neutrons in a dual-zone subcritical blanket are examined. The results of calculations of the electron-photon-neutron interactions of an electron beam with the target material and of the neutron-physical and heat-engineering characteristics of the system are presented. It is shown that a cascade neutron amplification factor of 2.5–3 can be obtained with system subcriticality 2%. The power of the system reaches 50 MW with electron beam power 4 MW.

Estimation of the hydrogen emission from a hydride moderator by measuring the reactivity and using mathematical statistics

The emission of hydrogen from zirconium hydride blocks during long thermal tests is discussed. The procedure used to perform thermal tests on four blocks is described. The results of reactivity-measurement experiments which were performed during the same period of time on a critical assembly simulating on the Nartsiss stand the neutron-physical characteristics of the Topaz-2 reactor to be used in space are given. A statistical analysis of the significance of the results obtained from the reactivity measurements is performed. It is noted that virtually no hydrogen was emitted from the testing blocks over the first five years of the thermal tests.

Possibility of operating GT-MHR with different fuels

The results of a computational investigation of the neutron-physical characteristics and parameters of GT-MHR for different fuels are presented: weapons-grade plutonium, reactor plutonium, low-enrichment uranium, mixed fuel based on uranium and weapons-grade plutonium, fuel based on high-enrichment uranium and thorium, and fuel with added americium and curium. Analysis has shown that GT-MHR can operate on different fuels; this is determined by the structural implementation incorporated in the design and the neutron-physical characteristics of the reactor. This feature makes it possible to use different fuels in this reactor without changing the basic structural components: fuel assemblies, size of the core and reflectors, and the number of reactivity compensation rods.

Computational and experimental investigations of the neutron-physical characteristics of the GT-MHR core

The main results of computational and experimental investigations of the neutron-physical characteristics of the GT-MHR high-temperature gas-cooled reactor are presented: analysis of the possible reasons for the uncertainties in the calculations of the neutron-physical characteristics of a core with plutonium fuel using different computer programs, computational and experimental investigations taking account of the experiments on the Astra critical assembly simulating the ring-shaped GT-MHR core, and the effect of using different nuclear data on the temperature coefficient of reactivity computed with high-accuracy computational codes such as MONTEBURNS-MCNP 5-ORIGEN 2.1.

Neutron-physical characteristics of a VVÉR core with 100% load of reprocessed uranium and plutonium fuel

The results of calculations of the neutron-physical characteristics of three variants of the fuel load in a VVER-1000 core are presented: a load consisting completely of enriched natural uranium (standard fuel) or reprocessed uranium-plutonium fuel from the first and second recycles. The calculations were performed for a stationary load of the core with a four-year fuel run. The difference between the neutron-physical characteristics of a core with a full load of uranium and reprocessed uranium-plutonium fuel is discussed. An analysis of the neutron-physical characteristics did not show any fundamental limitations for a possible 100% load of reprocessed uranium-plutonium fuel in a VVER-1000 core.

Calculation and experiment to determine the operational characteristics of a research reactor

The special features in comparing the computational neutron-physical characteristics with experiment and the applications of the computational data in operating a research reactor are examined. Specifically, the interaction of the calculation with operational procedures, in particular, with the procedure for calibrating the working organs of the safety and control system, is examined.

Development and adoption of measures for increasing the safety of power-generating units of nuclear power plants with RBMK reactors in 1986–2005

The basic stages and directions of upgrading and reconstruction of the power-generating units of nuclear power plants with RBMK reactors over a period of 20 yr since the Chernobyl accident are reflected in this paper. The greatest amount of work was done in 1986–1990 on all units with RBMK-1000 and-1500 reactors which were operating at that time in the USSR. The purpose of the upgrade was to improve the neutron-physical characteristics of the reactor cores, increase the response speed of the safety and control system and decrease the positive effect of water outflow from the cooling loop of this system, increase the flow capacity of the systems performing emergency discharge of the steam-gas mixture from the reactor, and improve the basic operating documentation.

Increasing the nominal safety of the no. 5 unit of the Kursk nuclear power plant

The main technical measures for increasing the nominal safety of the No. 5 unit of the Kursk nuclear power plant are presented: optimization of the neutron-physical characteristics of the core, adoption of a comprehensive monitoring, control, and protection system, switching to cluster-type control rods, adoption of a system for limiting radioactive emissions during anticipated accidents with pipeline rupture outside the containment system.

Computational method for and analysis of the application of granular absorbers in VVÉR reactors

Questions concerning the compensation of excess reactivity in pressurized-water reactors by using consumable granular absorbers are examined. A method of computing the spatial-energy distribution of the neutrons in cells with a granular absorber is presented. The neutron-physical and thermophysical characteristics of fuel assemblies with fuel elements based on homogenized and heterogeneous arrangements of gadolinium in them are compared. It is shown that granular absorbers have certain advantages, specifically, they decrease the gadolinium content in the fuel elements and at the same time increase the total number of gadolinium-containing fuel elements in the fuel assemblies. This decreases the maximum power released in the gadolinium-containing fuel elements and the temperature of the fuel during the entire run.