Neutron-physical Characteristics Research Articles

Experimental study of heat pipe start-up characteristics and development of an enhanced model considering gas diffusion effects

Heat pipe-based cooling reactors, offering passive heat transfer and modularity, are becoming the preferred type for micro-reactors. Alkali metals such as sodium, potassium and lithium are usually used as the working medium of the heat pipe, which is in a frozen state at normal atmospheric temperature. After heating the heat pipe, the working medium will melt and enter the gas chamber, and its flow state will undergo a transition from discontinuous flow to continuous flow. The start-up of the heat pipe affects the heat export and neutron physical characteristics of the heat pipe cooling reactor. Building a model of heat pipe can predict the start-up of heat pipe effectively, which can be helpful to study the start-up and safety analysis of heat pipe cooling reactor. In this paper, a heat pipe flat-front model considering the gas diffusion effect is established, and the gas temperature distribution characteristics in the non-continuous flow region are obtained through the start-up experiment of heat pipe. The calculated results of the flat-front model considering gas diffusion effect and the vertical flat-front model are compared with the experiment. The results show that the model in this paper can better describe the start-up of heat pipe. At the same time, the evaporation and condensation coefficient model based on temperature change is used to deal with the evaporation and condensation mass flow at the phase change interface of the heat pipe, which can better describe the temperature difference at the interface of the heat pipe.

Physical start-up of the Nuclear Subcritical Facility “Neutron Source” of NSC KIPT

National Science Center “Kharkov Institute of Physics and Technology” jointly with the Argonne National Laboratory developed, designed and constructed the Subcritical Nuclear Facility “Neutron source based on the subcritical assembly driven by electron linear accelerator”. The article presents the main results obtained during the physical start-up of the plant, the program of which involved the staged loading of fuel assemblies into the core with subsequent measurements of its neutron-physical characteristics. In total, 37 fuel assemblies of WWR-M2 type were loaded into the facility core. The values o f the effective neutron multiplication factor keff and reactivity ρ of the subcritical assembly were measured after each loading step using the multiplication and area ratio methods. In addition, the neutron flux to electron current ratio method was tested for on-line reactivity monitoring. The investigations were performed when the accelerator was operated with 100 MeV electron beam energy, 2.7 μs electron pulse duration, 20 Hz pulse repetition rate and ≤ 40 mA beam pulse current. The obtained results were analyzed in terms of ensuring the facility nuclear safety at the next stage of its commissioning which is the pilot operation.

Критичність паливовмісних матеріалів в об'єкті "Укриття"

The article presents the results of the analysis of the validity of simplifications and assumptions accepted by researchers in the models of neutron-physical characteristics of fuel-containing materials (FCM) in the "Shelter" object (SO). The analysis and substantiation of the FCM nuclear safety will be one of the important issues during the SO operation for a long time, until the moment of retrieval and disposal of nuclear materials from the ChNPP Unit 4 destroyed in 1986. A significant number of nuclear materials in the SO are uncontrolled, and the geometry of their location is also little known. Therefore, the use of traditional methods of nuclear safety analysis to determine the conditions for reaching criticality in the volume of FCM can lead to significant uncertainties, depending on the assumptions and simplifications made in FCM models. In contrast to traditional nuclear facilities with design characteristics and control systems, there is no possibility to deterministically influence the FCM, and accordingly to determine their dynamic characteristics in the SO. In such conditions, it is necessary to take a responsible approach to obtain and interpret the values of FCM parameters by control systems of the SO. For example, in 1990, one of the neutron flux detectors recorded an abnormal change in the neutron pulse count rate for about 100 hours. This event is often called a “neutron incident”, and it formed the basis of some models, in which the possibility of reaching criticality in one of the FCM clusters is assumed. The article presents the results of the study on the physical possibility of the anomalous event in 1990, as well as the results of the analysis of simplifications and assumptions made in models of FCM criticality. An important result of the conducted research is the justification of the need to include in the FCM criticality model the effects of reactivity on fuel and water temperature, as well as to take into account delayed neutrons.

MTIR-SCP nuclear safety concept

The MTIR-SCP reactor is a research reactor with a light-water coolant having a fast neutron spectrum in nominal operation. None of the existing or projected reactors has a set of these features, therefore, special attention should be paid to the choice of the design of control rods and their location. The article considers various designs of control rods characteristic of the VVER-SCP power reactor currently under development, as well as existing research reactors with a liquid metal coolant having a fast neutron spectrum, taking into account the features of the MTIR-SCP: a seven-element control rod assembly design, as well as a ring-type absorbing element. Efficiency calculations are carried out for each of the structures under consideration and it was determined that the design of a control rod with an annular type of absorbing element has the highest specific efficiency. This design is chosen as the main one for the MTIR- SCP reactor. Criteria are proposed and optimization of the placement of control rods in the reactor core is carried out in order to reduce the uncertainty of obtaining neutron-physical characteristics in the central autonomous loop channel and increase the flow density in it. When optimizing the placement of control rods in the research reactor, their influence on the main neutron-physical characteristics, and especially on their distribution during operation under nominal conditions, is taken into account. Emergency protection and compensating rods that compensate for temperature and density effects when the reactor reaches the nominal power level have been removed, therefore their effect on the neutron-physical characteristics of the reactor and the loop channel during operation at power is not significant, and the compensating rods are located in the core and are removed as the fuel burns out, therefore their effect on the spatial The distribution of neutron-physical characteristics will be significant. To reduce this effect, burnout compensators and automatic regulators are proposed to be moved outside the core into the reflector. For various reactor conditions, the efficiency of the control and protection system was evaluated. It is shown that in various states, the control rods transfer the reactor to the required level of subcriticality, in accordance with NP-009-17.

Calculation and experimental analysis of BN-800 neutron-physical characteristics during the transition to mixed oxide fuel loading

Calculation and experimental analysis of BN-800 neutron-physical characteristics during the transition to mixed oxide fuel loading

Neutron-physical characteristics of UO2 and UN/U3Si2 fuels with Zr, SiC and APMT accident tolerant claddings

H2 production due to zirconium hydration was the primary source of explosion in the Fukushima Daiichi nuclear accident. To improve accident tolerance of the new reactor designs, advanced fuel and cladding materials are being extensively investigated. Neutronic evaluation at both the pin-cell and assembly levels are performed for the conventional UO2 and the candidate UN/U3Si2 fuels with traditional cladding (Zr) as well as the accident tolerant claddings (SiC and advanced powder metallurgic ferritic, APMT). The neutronic performance, plutonium inventory, cycle length, radial and pin power distribution, spectrum distribution, spatial self-shielding analysis, plutonium radial distribution, and reactivity coefficients are evaluated and analyzed. Combination of UN/U3Si2 and APMT produces 68% and 66 % more 239Pu and 135Xe than the UO2 and APMT systems. In overall, the reactivity change versus burnup and other main neutronic parameters of SiC cladding are quite similar to the current Zr cladding. Using SiC resulted in an average increase of 0.85 % in reactivity in comparison to Zr conventional cladding (without any change in the enrichment of the fuel). Compared with the traditionally used UO2 fuel assembly, the UN/U3Si2 fuel assembly exhibit a higher (0.2 % at BOC and 0.4 % at EOC) pin power peak value in an assembly, which should be a concern from safety aspects. The negativity of FTC and MTC for the proposed UN/U3Si2 is higher than that of UO2 fuel. These higher and enhanced FTC and MTC values of UN fuels mean an improved inherent safety feature of UN-like high U density fuels.

INVESTIGATION OF NEUTRON-PHYSICAL CHARACTERISTICS OF THE IRRADIATION CAPSULE DESIGNED FOR RADIATION COLORING OF TOPAZ IN THE WWR-K REACTOR

This paper presents the results of the study of neutron-physical characteristics of an irradiation capsule designed for radiation coloring of topazes. The studies were performed in the core of the critical facility. Sandwich screen made of boron carbide powder and tantalum foil was used to cut off thermal neutrons inside the capsule. It is shown that using sandwich screen the thermal neutron flux is reduced by 8 times, and the fast neutron flux density is practically unchanged. The activation of the tantalum monitor is reduced by more than a factor of 2. The reactivity effect from loading the irradiation capsule with the screen into the core is minus 0.9% Δk/k.

Results of Experiments under the Physical Start-Up Program of the IVG.1M Reactor

In 2022, the physical start-up stage of the IVG.1M research reactor was successfully completed, initiated by reducing the fuel enrichment of 235U. This phase included the loading of nuclear fuel into the reactor core and conducting experiments to determine the neutron-physical characteristics of the reactor. Prior to the physical start-up, preliminary calculations were performed using the computational code MCNP6 and a full-scale model of the IVG.1M reactor with low-enriched uranium fuel (LEU). During the start-up series, the reactivity worth curves of the reactor control and protection system’s operating and compensating elements were determined. Additionally, experiments were performed to measure the reactivity effects of technological channel draining and to obtain activation reaction rates in the central experimental channel using nickel and gold activation indicators. The results of determining the neutron-physical characteristics of the IVG.1M reactor have confirmed the operability of the reactor core with LEU fuel.

Стан наукової проблеми щодо дослідження характеристик продуктів взаємодії розплаву ядерного палива з бетоном

Стан наукової проблеми щодо дослідження характеристик продуктів взаємодії розплаву ядерного палива з бетоном

Influence of Fuel Temperature Distribution Dependence on Neutron-Physical Characteristics of Nuclear Core with WWER-1000 (1200)

This paper presents the results of the influence of reactivity on fuel temperature distribution for the neutron-physical characteristics of WWER 1000-1200 cores, which in turn affects the parameters of Xenon oscillation. The improved method for calculating Resonance Absorption in a 238 UO 2 (Doppler reactivity effect) with a Nonuniform Temperature Profile are developed. The parameters calculation results for Xenon oscillations have been calculated using complex program PROSTOR. The validation of the improved method was examined by comparing the calculated xenon parameters to measurement results obtained during commissioning tests of WWER-1000 reactor power units. Results showed a more accurate calculation for the Resonance Absorption in a 238 UO2, they significantly reduce the stabilization of Doppler Effect broadening, hence these facts, are vital importance in the case of load following mode accompanied by Xenon processes in the core. The improved method is implemented into the computer-software of the Full-Scale Simulator of the 3 rd Power Unit of the Rostov NPP and the Full-Scale Simulator of the 4 th Power Unit of the Kalinin NPP. The method was tested during acceptance tests of these Full-Scale Simulators.

Simulating the fuel cycle of a lead-cooled fast reactor

The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was used, which has all the necessary capabilities to simulate the operation of the reactor in the closed nuclear fuel cycle mode, taking into account the stages of fuel storage and refabrication. The article presents a technique for modeling the fuel cycle, implemented for the operation of fast reactors with a lead coolant. To demonstrate methodology, a closed nuclear fuel cycle was simulated for the BREST-OD-300 and BR-1200 reactors for the design life. The article describes the scenarios in which the calculation of the burnup of reactor was carried out. In the considered scenarios, it is assumed that the unloading of fuel at the end of the micro campaign is conducted according to the maximum burnup. During the computational modeling the ranges of changes in fuel density and enrichment, reactivity margin, breeding ratio and isotopic composition of plutonium were determined.

SPECIFIC FEATURES OF APPLICATION OF CADMIUM AS A NEUTRON-ABSORBING SCREEN WHILE DOPING SILICON IN A NUCLEAR REACTOR

The gradient of the neutron field in a nuclear reactor and the requirements for the permissible spread of the specific electrical resistance over the volume of the silicon ingot makes it necessary to develop an irradiation device. This is especially true for large silicon ingots. One of the options for reducing the gradient of the neutron flux along the height of the ingot is the use of neutron-absorbing screens in the design of the irradiation device. At the WWR-K reactor, cadmium with a natural isotopic composition is used as a neutron-absorbing screen material. The paper presents the results of a study of an irradiation device with a cadmium screen. The effect of a cadmium screen on the neutron-physical characteristics of an irradiation device for silicon doping is shown.

Using the Few-Group Approximation for Calculating Some Neutron-Physical Characteristics of VVER-1000 Core by Means of the Monte Carlo Universal Code

Abstract The main purpose of this work is to study the possibility of using the few-group approximation for calculation of some neutron-physical characteristics of VVER-1000 core by means of special version of Monte Carlo Universal code. The Monte Carlo method for VVER-1000 core neutron-physical characteristics calculation using the few-group approximation with an estimate of neutron cross sections “by location” was provided and tested in this research. The reduction of calculation time due to the transition from a pointwise model of representation of cross sections to the few-group approximation and methodical error of this approach were evaluated. Optimal number of energy groups was determined. It was found that consideration of the scattering anisotropy leads to a significant decrease in methodical error. Ways of further reduction of methodical error were worked out.

ПРОГРАММА СПЕКТРОМЕТРИИ НЕЙТРОННЫХ ПОЛЕЙ В РЕАКТОРЕ СМ ПОСЛЕ МОДЕРНИЗАЦИИ

A program for refining the energy spectra of neutron radiation in the neutron trap, core and reflector of the SM-3 reactor after modernization using the certified equipment and verified methods is presented in the article. The relevance of the article lies in familiarization with the experimental implementation of the methods in order to compare the experimental neutron-physical characteristics of the reactor before and after the last reconstruction. During the program creation, the previous experience of spectrometric measurements was taken into account, as well as experiments were carried out under conditions close to those of spectrometry after reconstruction. The results of the research are necessary for the planning of irradiation experiments and verification of calculation methods for determining the neutron-physical characteristics. The results of the described researches cover most science and technology areas, many aspects of human practice, including healthcare. The authors presented results of accompanying experiments, which made it possible to work out the regulations and features of short-term measurements. The developed and optimized program of neutron spectrometry in the SM-3 reactor channels will allow to obtain reliable information on the density of neutron fluxes for the first time after modernization to verify analytical data and to optimize the use of the SM-3 reactor facility to increase the production of transplutonium elements and radionuclide products.

ЯДЕРНЫЕ ДАННЫЕ ДЛЯ РАСЧЕТОВ БЫСТРЫХ РЕАКТОРОВ - БИБЛИОТЕКА ФАЙЛОВ РОСФОНД И СИСТЕМА КОНСТАНТ БНАБ-РФ

The work is devoted to one of the most important scientific and technical problems in reactor physics related to the development and verification of codes and nuclear data that provide reliable and highly accurate calculations of neutron-physical characteristics of fast reactors and radiation shielding, including nuclear fuel cycle, criticality and radiation safety parameters. The designed neutronics characteristics of fast reactors should be based on certified, qualified sets of codes and nuclear constants: the calculation tools should be related to the modern state of scientific knowledge and computational techniques, and used nuclear physics constants should be adequate to the most reliable evaluations of nuclear data. In connection with rapid development of the computing engineering and all greater introduction in practice of calculation Monte Carlo codes, the methodical constituent of calculation error falls down substantially. In these terms the nuclear constant’s constituent of error of calculations becomes fully qualificatory. A situation is intensifyed by the fall-off of financing of experimental works, why in this connection the amount of fast critical stands in the world diminishes sharply. The paper consider the state of art of the constant’s providing system CONSYST/ABBN, created on the basis of the national library of neutron data files ROSFOND and libraries of multigroup constants ABBN-93 and ABBN-RF. One of the most important problems under consideration here also is the methodical and software for estimating of errors of the calculated physical characteristics. System of codes and nuclear data for reactor neutronics calculations is based on the unified methodological basis that ensures the transparency of the procedure for obtaining the data used in the calculations, the reliability of their verification and the obtaining of guaranteed accuracy of the calculated physical reactor characteristics.

Аналіз впливу вигоряння ядерного палива реактора ВВЕР-1000 на швидкість утворення 16N в теплоносії першого контуру

The developed model of the WWER-1000 reactor using MCNP6.2 (Monte Carlo N-Particle Transport Code) includes the detailed core taking into account the design of the fuel assemblies, as well as the baffle, the lower plenum, the fuel support columns, the core barrel, a downcomer, and the reactor pressure vessel. It allows implementing multifunctional calculations such as recriticality with various fuel configurations, the critical concentration of boric acid, determination of the axial and radial peaking factor in the reactor core, etc. For obtaining the more precise result of the cumulation nitrogen-16 formation rate, the contribution from different water volumes was taken into account: in the core, above the fuel and the top nozzle, in the top nozzle of the fuel assembly, in the bottom nozzle, between the fuel and the bottom nozzle, in the axial channels of the baffle, in the reflector. In order to obtain the realistic boundary conditions, the change of the isotopic composition in the fuel assemblies during one fuel cycle was calculated using the ORIGEN-ARP of SCALE software. Therefore, the influence of the nuclear fuel depletion of fuel assemblies in the WWER-1000 reactor on the change of the basic neutron-physical characteristics was determined such as the distribution of the neutron flux density with the energies necessary to initiate the 16O(n,p)16N reaction, the average number of neutrons per fission, the neutron spectrum and average fission energy. As a result, the dependence of the nitrogen-16 formation rate in the primary coolant system on the nuclear fuel burnup is obtained.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

CROSSER - ПРОГРАММНЫЙ МОДУЛЬ ПОДГОТОВКИ ГРУППОВЫХ КОНСТАНТ ДЛЯ ИНЖЕНЕРНЫХ РАСЧЕТОВ БЫСТРЫХ РЕАКТОРОВ

The purpose of this work was to test the CROSSER high-speed constant preparation module for use in the practice of engineering calculations of the neutron-physical characteristics of fast reactors. Computational models of fast reactors BN-600, BN-800 and BN-1200 were used as test tasks. The following neutron-physical parameters were calculated: the criticality of the reactor model, the efficiency of the CPS and the sodium void coefficient of reactivity. The main reactor functionals were calculated using the MMKK engineering program, the constants for which were prepared using the CROSSER module. The initial sets of nuclear data for the MMKK program were the ABBN-93 library of constants and the constants in the ABBN-RFE formats obtained on the basis of the ROSFOND-2010 library. The results obtained using the MMKC program with detailed tracking of the particle energy were used as the reference calculation results. Detailed data in ACE format obtained from the ROSFOND 2010 library were used as the initial nuclear data for this program.

ВЛИЯНИЕ ЗАМЕНЫ КОНСТРУКЦИОННОЙ СТАЛИ ТВС НА ЗАПАС РЕАКТИВНОСТИ В РЕАКТОРЕ БН-600

The current program in Russia to increase the fuel consumption of fast reactors and increase its burn-out causes the transition to new structural materials, which, in turn, leads to changes in the neutron-physical characteristics of reactors. In particular, the drop in the reactivity reserve noted in the BN-600 reactor of the Beloyarsk NPP at the end of 76 operational cycles, as will be shown below, is due to the transition to a new type of shell steel with an increased content of nickel, which strongly affects the reactivity. Design support for the operation of the BN-600 and BN-800 fast reactors, as well as the experiments carried out on them, is performed by IPPE. This article presents the results of a calculated analysis of the expected changes in the reactivity reserve at the end of 76 operational cycles when replacing the shell steel in BN-600. In addition, the influence of experimental assemblies located in the core on the reactivity reserve of the BN-600 is analyzed. Analysis of calculations of the actual loading of the BN-600 reactor at 76 operational cycle using the methods of the 1st-order perturbation theory, strict perturbation theory, and the Monte Carlo method showed that a partial transition at 76 operational cycle to EK-164 shell steel leads to a decrease in the reactivity margin by 0.030±0.004 %Δk/k. Replacement of steel for the entire core will reduce the reactivity margin by ~0.12 %Δk/k, which is confirmed by Monte Carlo calculations. The calculated reactivity margin obtained at the end of 76 operational cycles for the hot state of the BN-600 reactor is in good agreement with the measured reactivity margin.

Effect of gadolinium absorber radial profiling in fuel pins on VVER-1000 assembly neutron-physical characteristics

The paper raises the problem of non-optimal use of resources. The purpose is to extend the fuel campaign. An alternative task is to increase the duration of the reactor campaign to increase the flexibility of decision-making during the operation of a nuclear reactor. The paper presents calculations of the profiling of a burnable absorber in fuel assemblies for VVER-1000 reactors. A quantitative assessment of the effect of alternative placement of gadolinium in fuel elements is made. Changes in the burnup of fuel elements in fuel assemblies are noted, as well as an advantage expressed in a change in the kinf. The results showed a strong dependence of the isotopic composition of the peripheral layer on the neutron flux density and a weak dependence of the isotopic composition of the central fuel element with gadolinium layers on the neutron flux density. The burnup when using the iterative process turned out to be impossible to align along the radius presumably due to the production of plutonium. A significant advantage of profiling a fuel cell with gadolinium is expressed in the period before the first overload, with each subsequent overload the efficiency of the technology decreases. However, kinf of the profiled assembly throughout the entire fuel campaign exceeds the kinf of the standard assembly, which indicates the efficient use of a burnable absorber.