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China is accelerating the development of sodium-cooled fast reactor technology. For nuclear reactors, whether it is a pressurized water reactor or a fast reactor, core flow distribution is a key concern, which directly determines whether the reactor can operate safely and reliably. Sodium-cooled fast reactor core adopts a three-stage flow distribution method consisting of diagrid, distribution headers and assemblies. Distribution headers are installed on diagrid, and various types of assemblies are installed on distribution headers. Pressure drop of the core is composed of distribution header pressure drop and assembly pressure drop. The distribution header pressure drop itself affects the flow distribution of the assemblies, thereby affecting the safety of the core. Therefore, it is of great significance to study the impact of distribution header pressure drop on the flow distribution of assemblies for the sodium-cooled fast reactors. In order to reduce the flow distribution deviation of assemblies caused by the distribution header pressure drop, it is necessary to carry out a reasonable assembly pressure drop design. Based on the mechanism of flow distribution deviation of assemblies caused by distribution header pressure drop, a theoretical calculation model was proposed, and an optimized design of assembly pressure drop was carried out for the China Experimental Fast Reactor (CEFR) core. Based on the actual layout of CEFR core, the maximum deviation of flow distribution of fuel assemblies was obtained, and the optimization direction of nominal assembly pressure drop was determined, indicating that optimization design of nominal assembly pressure drop should be carried out for the first five rings. After adjusting the nominal pressure drop of the first five rings of assemblies from 250 kPa (the original nominal pressure drop) to 249 kPa, 248.5 kPa, and 248 kPa, respectively, the maximum deviation of flow distribution of fuel assemblies firstly decreases from −0.99% to −0.95%, and then increases to −1.02% and −1.08%, which indicates that nominal assembly pressure drop should be elaborately determined to obtain a minimum flow distribution deviation of fuel assemblies. In conclusion, when conducting core thermal hydraulic design for sodium-cooled fast reactors, it is necessary to analyze the optimization direction of nominal assembly pressure drop based on actual core layout, and sensitivity analysis should be conducted to finally determine the nominal assembly pressure drop to reduce the impact of distribution header pressure drop on flow distribution of assemblies to the lowest extent.

This article presents the results of an experimental study of the coolant flow in a fuel rod bundle of a nuclear reactor fuel assembly of a small modular reactor for a small ground-based nuclear power plant. The aim of the work is to experimentally determine the hydrodynamic characteristics of the coolant flow in a fuel rod bundle of a fuel assembly. For this purpose, experimental studies were conducted in an aerodynamic model that included simulators of fuel elements, burnable absorber rods, spacer grids, a central displacer, and stiffening corners. During the experiments, the water coolant flow was modeled using airflow based on the theory of hydrodynamic similarity. The studies were conducted using the pneumometric method and the contrast agent injection method. The flow structure was visualized by contour plots of axial and tangential velocity, as well as the distribution of the contrast agent. During the experiments, the features of the axial flow were identified, and the structure of the cross-flows of the coolant was determined. The database obtained during the experiments can be used to validate CFD programs, refine the methods of thermal-hydraulic calculation of nuclear reactor cores, and also to justify the design of fuel assemblies. [ABSTRACT FROM AUTHOR]

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In the present-day conditions under which the nuclear power industry is developed, a need arises to diversify the designs of new nuclear power plant units, which should differ from the previously constructed ones by featuring flexibility to the customer requirements and by using safety systems based on fully passive safety assurance principles. In 2022, specialists of Experimental and Design Organization (OKB) Gidropress commenced activities on elaborating the draft design of a new integral pressurized water-cooled reactor plant VVER-I with natural circulation of coolant for a basic thermal capacity of 250 MW. The design incorporates passive safety systems able to provide reliable heat removal from the core under the conditions of a long-term NPP blackout and without the operator's participation. The article presents the results obtained from thermal and fluid dynamic computations of the new reactor plant carried out using the KORSAR/GP code that has been certified for safety analyses. A reactor plant thermal-hydraulic model, which can be used for computations of stationary normal operation conditions and, subsequently, also for simulating the accident scenarios evolvement dynamics, has been developed and tested. Computations carried out using the system code have confirmed a correct choice of the reactor's main geometric parameters and the steam generator's heat-transfer surface for operation at the nominal power. Based on the computation results for optimizing the design, it is proposed to use a jacketed steam generator, which will make it possible to exclude stray coolant leaks in bypass of the heat-transfer surface. It is shown that the newly developed reactor plant has a significant potential for increasing the thermal power capacity up to 400 MW without introducing fundamental changes in the design. The study results can be used in designing new VVER reactors with natural coolant circulation, and also in the development of passive safety systems. [ABSTRACT FROM AUTHOR]

Described shortly here is a procedure of demounting, removal, transport and long-term storage of the SM-3 core, based on the previous experience of reactor refurbishment undertaken in 1991. Prior to performing refurbishment, computations and calculated data analysis were performed to prove radiation safety of this work, which included estimation of the activity level for activation products in the structural materials of the nuclear research reactor core and the radiation conditions at different stages of its handling. As evidenced by the calculated data, the activity of the main dose-forming radionuclide 60Co attains equilibrium in about 12 years of radiation exposure. Taking into account the fact that the time period between two refurbishments was longer than 12 years, the calculated values of the equivalent dose rate were normalized to the radiation monitoring data obtained during the previous refurbishment, taking into account the calculated activity of 60Co radionuclide. The normalization made it possible to confirm reliability of estimates. The obtained activity data of activation products and taking into account the time spent during the SM-3 refurbishment in 1991, the radiation impact on personnel was estimated. Calculated values of the anticipated effective radiation exposure doses to the personnel engaged in the refurbishment revealed that the main limits of the personnel radiation exposure established in accordance with NRB-99/2009 were not exceeded. Comparison of the results of calculating the equivalent dose rate with the results of radiation monitoring at various points allowed us to establish that during the calculation and analytical justification of the radiation safety of work, the assessment of reflected radiation was significantly underestimated. But the radiation monitoring data, personal radiation monitoring, as well as recorded data of automatic radiation monitoring system show that all work was performed in compliance with the requirements of regulatory documents in the field of radiation safety.

GFR or Gas-cooled Fast Reactor is one type of fast generation-IV that uses a very high cooling temperature. Thus, it is necessary to have the right reactor core design so that the power distribution of neutrons produced reaches a safe and even limit point. The use of a uniform (homogeneous) reactor core can produce peaking power. This is very avoidable because it will cause a reactor accident. In this study, researchers tried to compare the results of the analysis for two heterogeneous reactor core designs including the configuration of 3 fuel variations and 5 fuel variations using UN-PuN fuel. This study aims to determine the keff value produced by both types of fuel variations during 5 years of burn-up and determine the characteristics of neutron flux, fission rate, and fission product during 15 years of burn-up. This study was started by calculating the homogeneous and heterogeneous core of 3 and 5 fuel variations with neutron transport simulation involving OpenMC. The calculation results show that the heterogeneous core configuration of 5 fuel variations for the keff value is more optimal than 3 fuel variations, because it has the smallest excess reactivity value. The neutron flux and fission rate characteristics for 5 fuel variations are more evenly distributed when compared to 3 fuel variations to maintain neutron lifetime and reactor life in operation. Burn-up residual plutonium material and minor actinide waste for 5 fuel variations have less mass than 3 fuel variations. The results of neutronic analysis of GFR reactors with heterogeneous reactor core designs for 5 fuel variations are better in terms of reactor criticality, neutron power distribution, and waste produced. Finally, optimization of the UN-PuN fuel volume fraction of 60 % provides the optimal keff value. [ABSTRACT FROM AUTHOR]

This article presents the results of an experimental study of the coolant flow in a fuel rod bundle of a nuclear reactor fuel assembly of a small modular reactor for a small ground-based nuclear power plant. The aim of the work is to experimentally determine the hydrodynamic characteristics of the coolant flow in a fuel rod bundle of a fuel assembly. For this purpose, experimental studies were conducted in an aerodynamic model that included simulators of fuel elements, burnable absorber rods, spacer grids, a central displacer, and stiffening corners. During the experiments, the water coolant flow was modeled using airflow based on the theory of hydrodynamic similarity. The studies were conducted using the pneumometric method and the contrast agent injection method. The flow structure was visualized by contour plots of axial and tangential velocity, as well as the distribution of the contrast agent. During the experiments, the features of the axial flow were identified, and the structure of the cross-flows of the coolant was determined. The database obtained during the experiments can be used to validate CFD programs, refine the methods of thermal-hydraulic calculation of nuclear reactor cores, and also to justify the design of fuel assemblies.

In order to extend the operational life of the BN-600 reactor, it is necessary to validate operability of the nonreplaceable reactor elements—including the neutron reflector—that operate under the conditions with high damaging doses and high temperatures. This can be done through irradiating metal of a simulating package inside a materials science assembly (MSA) inserted in the BN-600. The paper describes the main characteristics of the assembly, including the design features and irradiation parameters. Described are the basic principles of selecting the location for the materials science assembly in the reactor and for specimens inside the assembly in order to ensure the required irradiation conditions for the metal specimens. With comparison of the locations of the materials science assembly in the reactor core and in the radial blanket, the effect that the location of the materials science assembly has on the characteristics of the reactor standard fuel assemblies and on the neutron flux field is discussed. To calculate the irradiation parameters and the effect that the location has on the materials science assembly, the same methods and programs are used as for the reactor core design work. The effect of the materials science assembly on the reactor core neutronic characteristics has been shown to be minimal when the assembly is installed in the radial blanket. [ABSTRACT FROM AUTHOR]

Anomaly detection for core temperature has great significance in maintaining the safety of nuclear power plants. However, traditional auto-encoder-based anomaly detection methods might extract the latent space features with redundancy, which may lead to missing and false alarms. To address this problem, the idea of feature disentangling is introduced under the auto-encoder framework in this paper. First, a feature disentangling auto-encoder (DAE) is proposed where a latent space disentangling loss is designed to disentangle the features. We further propose an incrementally feature disentangling auto-encoder (IDAE), which is the improved version of DAE. In the IDAE model, an incremental feature generation strategy is developed, which enables the model to evaluate the disentangling degree to adaptively determine the feature dimension. Furthermore, an iterative training framework is designed, which focuses on the parameter training of the newly incremented feature, overcoming the difficulty of model training. Finally, we illustrate the effectiveness and superiority of the proposed method on a real nuclear reactor core temperature dataset. IDAE achieves average false alarm rates of 4.745% and 6.315%, respectively, using two monitoring statistics, and achieves average missing alarm rates of 6.4% and 2.9%, respectively, using two monitoring statistics, outperforming the other methods. [ABSTRACT FROM AUTHOR]

Advanced pressurized water reactors are the main part of a new generation of nuclear power plant projects under development that provide cost-effective power production for various needs (Yemelyanov et al. 1982, Klimov 2002, Boyko et al. 2005, Baklushin 2011, Bays et al. 2019, Nuclear Technology Review 2019). The innovative technologies are aimed at improving the safety and reliability as well as at reducing the cost of NPPs. At the same time, improvements in design, technological and layout solutions are focused primarily on the reactor core. Assessments of the efficiency of these improvements are preceded by numerical simulations of the processes in the core, in particular heat generation and sink, with account for the difference between the study object and the standard version tested in operational practice. The authors of the article propose a method for calculating the temperature field in the core of a heterogeneous reactor (using the example of a pressurized water reactor), which makes it possible to quickly assess the level of temperature safety of various changes in the core and has the necessary speed for analyzing transients in real time. This method is based on the energy equation for an equivalent homogeneous core in the form of a heat equation that takes into account the main features of the simulated heterogeneous structure. The procedure for recovering the temperature field of a heterogeneous reactor uses the analytical relation obtained in this work for the heat sink function, taking into account inter-fuel element heat leakage losses. Calculations of temperature fields in the model of the PWR type reactor (The Westinghouse Pressurized Water Reactor Nuclear Plant 1984) were carried out in stationary and transient operating modes. The calculation results were compared with the results of CFD simulation. The area of competing use of the temperature field recovery method was indicated.

The reactor cores of the Advanced Gas-cooled Reactors (AGRs) are made up of a large, three-dimensional array of interlocking graphite bricks. To assess the seismic performance of a wider range of reactor cores in a given condition (number, type, location and orientation of cracking of individual graphite components) that would be possible with physical testing, a computational model or digital twin has been developed. Cross-comparison of numerical simulations using this model and output from testing of a quarter-scale physical model have been used to validate the computational model, which has then been used to assess the behaviour of full-sized reactor cores during a seismic event. These assessments aim to demonstrate that the reactor cores can be shut and held down and adequate cooling of the fuel maintained during and after seismic events. [ABSTRACT FROM AUTHOR]

• The porous medium approach was applied to the simulation of the two-pass folded-flow NTP reactor core. • The non-equilibrium thermal model was used to simulate the fuel element and the applicability was verified. • The thermal–hydraulic characteristics of the reactor core and the influence between different factors were studied. Nuclear thermal propulsion (NTP) is a promising advanced technology which has attracted wide attention in recent years. The reactor core is an essential component of an NTP system and the corresponding thermal–hydraulic analysis is necessary. In this study, the porous medium approach was applied to the simulation of a two-pass NTP reactor core which consists of the porous prismatic cermet fuel elements. The thermodynamic property models of hydrogen and the fuel element materials were implemented, as well as the empirical correlations of the heat transfer coefficient and the friction factor. The three-dimensional simulation of a single fuel element was carried out and the results were compared against another code. The code-to-code comparison verified the applicability of the porous medium approach. The three-dimensional model of the two-pass NTP reactor core was established and the steady-state simulation was carried out. The distribution patterns of the parameters are determined by the thermal–hydraulic characteristics of the reactor core, including the nonuniform heat release, contact heat conduction and folded-flow scheme. The full-core heat-flow adaptability analysis is realized, which provides a reference for the thermal–hydraulic safety analysis of the NTP reactor. [ABSTRACT FROM AUTHOR]

Summary: Pebble‐bed HTR utilizes the configuration of randomly distributed graphite and fuel pebbles which contain randomly dispersed TRISO particles, causing the double heterogeneity effect and making simulation get complicated. To establish a high‐fidelity whole core model of the annular core pebble‐bed HTR, this article proposes a two‐step whole core modelling scheme with flexibility. This scheme is verified by comparing the HTR‐10 initial critical benchmark results with the HTR‐10 experiment results. Based on the geometry modelling method and Monte Carlo simulation, this study investigates the effect of the central graphite column dimension and the pebble size upon the nuclear heating power density distribution in annular core pebble‐bed HTR. Results show that the annular core reactor has a more edgy distribution of neutron flux and nuclear heating power density and a higher peak value, compared with the cylindrical configuration core reactor. The annular core reactors with a higher thermal power could realize a higher helium outlet temperature with a precondition that the outlet helium flow is carefully separated and mixed. Accordingly, a higher thermoelectric conversion efficiency could be achieved. Reactors filled with smaller pebbles reach the criticality more quickly. However, the radius of the pebbles in the range from 2.5 to 3.5 cm does less effect than the size of the central graphite column does to the neutron flux and nuclear heating power density distribution. The running‐in phase of the annular configuration core reactor is investigated in the last section. The heating power density gradually flattens as the initial core pebbles fall and new fuel pebbles are loaded into the cavity. In this running‐in phase, we adopt a one‐to‐one mapping technique that sets the temperature of pebbles to their real value, varying from their locations. This enables us to do further work of the neutronics/thermal‐hydraulics analysis and the dynamic simulation which fit the realistic engineering practice, and to explore the fuel management scheme of the annular core high‐temperature gas‐cooled pebble‐bed reactor. [ABSTRACT FROM AUTHOR]

The paper presents calculations of the thermal power distribution in the reactor core (RC) of the WWR-SM research nuclear reactor of the Institute of Nuclear Physics of the Academy of Sciences of the Republic of Uzbekistan, settlement Ulugbek, Tashkent, both for all fuel assemblies loaded into the core and for each fuel element of a separate fuel assembly. These calculations were carried out for RC configurations with a different number of fuel assemblies – 18, 20, and 24. The power distribution reactor core was performed using the IRT-2D code. A detailed simulation of the power distribution in the fuel element was performed using the MCNP4C code, while the fuel elements were modeled as square pipes with straight angles without rounding. The power distribution was calculated for each side of each fuel tube and divided into 15 axial nodes. The results of modeling of the thermal-hydraulic state were obtained using the PLTEMP code for various RC configurations. In the calculations, it was assumed, that the inlet water temperature is 45 and 48 °C for all RC configurations, the heat hydraulic parameters were taken from the calculation of the flow rate of the first circuit through the core 1250 m3/h. An analysis of the thermal power distribution of nuclear fuel in the reactor core of the WWR-SM research reactor showed that even with a conservative approach, permissible operating modes are not exceeded. During the operation of the three main circulation pumps, which provide the coolant flow through the core at the level of 1250 m3/h, the heat exchange crisis does not occur in the most energy-stressed fuel assemblies, namely, the temperature of the fuel rod clad and coolant remains below the permissible limits.

A nonstationary two-dimensional model of numerical calculation of heat transfer in the core of a WWER-1200 reactor has been developed on the basis of representation of the core as a porous body. Nonstationary regimes occurring in the presence of local growth in the heat release in the cage and resulting in the reorganization of the flow field in the core have been modeled. [ABSTRACT FROM AUTHOR]

The electrical power of a thermionic reactor-converter is determined by various parameters of the thermionic fuel element and reactor core, as well as by the formation of the electrical branch that connects thermionic fuel elements according to various criteria. The formation of electric branches is possible in various ways. The number of branches is determined by the specified characteristics of the thermionic fuel element and the required electrical power. In this case, the thermionic fuel elements are connected electrically in series to the branches, and the branches themselves are connected in parallel. The selection of the electrical branch in the core of the reactor is based on the characteristics of each thermionic fuel element. In this article, we evaluate the characteristics of the thermionic reactor-converter with various options for switching thermionic fuel elements. [ABSTRACT FROM AUTHOR]

The results of complex calculation of the solution pulsed nuclear reactor (PNR) VIR-2M are presented. The simulation of the fuel solution dynamics is described, and different pulse modes implemented at VIR-2M are analyzed. The effect of cross-connecting baffles for fixing the rod channel positions with respect to the central channel of PNR VIR-2M was taken into account in calculations for the first time. The numerical model of PNR VIR-2M core vessel was constructed, and the analysis of its strain condition under the action of pulsed load determined by the fuel solution dynamics in fission pulse generation was performed. The studies prove the mechanical and strength parameters of the PNR VIR-2M vessel and substantiate its service properties. [ABSTRACT FROM AUTHOR]

The results of experimental studies of characteristics of the YAGUAR nuclear homogeneous uranium aperiodic reactor obtained in the process of physical startup after its modernization are presented. [ABSTRACT FROM AUTHOR]

The paper presents the results of investigations into the hydrodynamics of a three-dimensional coolant flow motion in the mixed core of a VVER (a.k.a. WWER or PWR) reactor. The experiments were conducted on an aerodynamic research facility test bench with scaled models of core fragments. An investigation has been made into the patterns of the coolant flow motion by experimental finding of the velocity vector. Measurement of the flow pressure field was conducted with a five-channel pneumometric (pressure-tube) probe and was then recalculated and converted into the direction and value of the velocity. To obtain a detailed pattern of a three-dimensional coolant flow motion, the most representative region of the models′ cross section including the space between the fuel (fuel rod bundles) and four rows of fuel elements of each of the adjacent fuel assemblies was identified and investigated. The general pattern of the coolant flow is represented by plots of transverse velocity distribution as a function of a relative coordinate and by cartograms of axial velocity distribution. An analysis of the spatial distribution of the flow velocity projections in characteristic regions of the model made it possible to identify the patterns of the coolant flow around hydraulically nonidentical intensifier grids. [ABSTRACT FROM AUTHOR]

Summary: A real three‐dimensional representation of reactor core with involute fuel plates via Monte Carlo method is still lacking at the present, and usually equivalent homogeneous models of water coolants and simplified fuel plate shapes have been used. This work proposed an algorithm to simulate the reactor core composed of involute fuel plates with an explicit accurate description of its involute surfaces. The description of involute surfaces is through its governing equations. The mathematical function methods to depict the involute curves and the schemes to compute the distance between any neutron particle and an involute surface along a ray are presented and demonstrated step‐by‐step. The algorithm is realized in the Monte Carlo code and validated by the published data of a high flux isotope reactor. This study definitively provides a method to model the involute fuel plates in the Monte Carlo simulation. [ABSTRACT FROM AUTHOR]

Special features of computational assessment of dimensional change of WWER-1000 reactor core baffle during the reactor operation are discussed. The paper gives results of computational analysis of a change in shape of core baffle, which was performed by applying up-to-date approaches of modeling the irradiation-induced swelling of constrained austenitic steels under the action of neutron irradiation and elevated temperatures. The results have been obtained by using median and conservative values of parameters of the temperature and dose dependence of free swelling of austenitic steel Kh18N10T. The authors have set out the fundamental principles of elastic-plastic stress-strain analysis of the reactor core baffle and core barrel, taking into account the irradiationinduced swelling deformation and contact interaction conditions. The finite-element analysis is based on a FEM mixed scheme that ensures a continuous approximation both for displacements as well as stresses and strains, thus providing a high-accuracy determination of the stress-strain state. The calculations were carried out in the two-dimensional definition for the baffle cross-section with the maximum damaging dose and irradiation temperature in the baffle height, assuming a generalized plane-strain deformation. The calculated results are given for the reactor full-power operation and scheduled shutdown for re-fueling at the end of the charge life. According to the calculation data, disregarding the irradiation-induced swelling deformation leads to an incorrect assessment of the change in shape of the core baffle during its operation, while the use of the accepted free swelling model gives too conservative results on the shape change, even within the designed lifetime. The influence of the mean normal stress on the irradiation-induced swelling of metal makes the principal contribution to the change in shape of the core baffle. During the reactor operation beyond the designed lifetime there occurs a local contact between the core baffle and the core barrel in the zone of the largest-diameter circular opening of the longitudinal cooling channel and the port channel for coolant flow between the baffle and the barrel. The paper gives results of the analysis of the change in shape of the baffle in modeling of the contact conditions taking into consideration the temperature re-distribution due to deviation from the design coolant passage conditions in the zone where the baffle is in contact with the barrel. In the case of using the median dependence of the free swelling, no loss of the nominal gap between the core baffle and the core barrel is observed within the designed lifetime. The computational assessment using the conservative parameters of the irradiation-induced swelling leads to a distinct decrease of the gap between the core baffle and the core barrel and a local contact between them at the end of the designed lifetime. The residual gap between the spacer grids of edge fuel assemblies and the baffle faces is ensured in the case of the median and conservative dependence of the irradiation-induced swelling. [ABSTRACT FROM AUTHOR]

The article presents the results from experimental investigations into local coolant flow fluid dynamics that were carried out on the fragmental model of a VVER-type pressurized water reactor's mixed core consisting of two types of fuel assemblies (FAs): TVSA-T (one segment) and TVSA-T.mod.2 (two segments). The coolant flow processes in the fuel rod bundle were simulated on an aerodynamic test bench. The pressure field in the flow was measured using a five-channel pneumometric probe. The obtained pressure field in the flow was recalculated for the coolant velocity vector according to the dependences derived during calibration. For drawing up a detailed flow motion pattern, a characteristic model cross section area was separated, which included the space between the assemblies and four fuel rod rows in each TVSA-type fuel assembly. The spatial distribution of coolant flow velocity projections was analyzed, as a result of which it became possible to reveal the regularities associated with the streamlining of spacer-, mixing-, and combined-spacer grids in TVSA assemblies by coolant; to determine coolant cross flows resulting from streamlining of hydraulically nonidentical grids and determine their localization in the experimental model longitudinal and cross sections; and to reveal the effect of accumulating flow hydrodynamic disturbances in the model longitudinal and cross sections resulting from the staggered arrangement of hydraulically nonidentical grids. The results obtained from investigations of inter-assembly interaction of coolant between the neighboring TVSA-T and TVSA-T.mod.2 assemblies have been adopted for practical use at AO Afrikantov OKBM in estimating the thermal reliability of VVER-type reactor cores and have been included in the database for verification of computation fluid dynamics computer programs (CFD codes) and detailed cell-wise numerical analysis of the core of VVER reactors. [ABSTRACT FROM AUTHOR]

The paper presents calculations of the thermal power distribution in the reactor core (RC) of the WWR-SM research nuclear reactor of the Institute of Nuclear Physics of the Academy of Sciences of the Republic of Uzbekistan, settlement Ulugbek, Tashkent, both for all fuel assemblies loaded into the core and for each fuel element of a separate fuel assembly. These calculations were carried out for RC configurations with a different number of fuel assemblies – 18, 20, and 24. The power distribution reactor core was performed using the IRT-2D code. A detailed simulation of the power distribution in the fuel element was performed using the MCNP4C code, while the fuel elements were modeled as square pipes with straight angles without rounding. The power distribution was calculated for each side of each fuel tube and divided into 15 axial nodes. The results of modeling of the thermal-hydraulic state were obtained using the PLTEMP code for various RC configurations. In the calculations, it was assumed, that the inlet water temperature is 45 and 48 °C for all RC configurations, the heat hydraulic parameters were taken from the calculation of the flow rate of the first circuit through the core 1250 m3/h. An analysis of the thermal power distribution of nuclear fuel in the reactor core of the WWR-SM research reactor showed that even with a conservative approach, permissible operating modes are not exceeded. During the operation of the three main circulation pumps, which provide the coolant flow through the core at the level of 1250 m3/h, the heat exchange crisis does not occur in the most energy-stressed fuel assemblies, namely, the temperature of the fuel rod clad and coolant remains below the permissible limits. [ABSTRACT FROM AUTHOR]

The neutronic parameters are required in the safety analysis of the RSG-GAS research reactor. The RSG-GAS research reactor, MTR (Material Testing Reactor) type is used for research and also in radioisotope production. RSG-GAS has been operating for 30 years without experiencing significant obstacles. It is managed under strict requirements, especially fuel management and fuel burn-up calculations. The reactor is operated under the supervision of the Regulatory Body (BAPETEN) and the IAEA (International Atomic Energy Agency). In this paper, the experience of managing RSG-GAS core fuels will be discussed, there are hundred possibilities of fuel placements on the reactor core and the strategy used to operate the reactor will be crucial. However, based on strict calculation and supervision, there is no incorrect placement of the fuels in the core. The calculations were performed on working core by using the WIMSD-5B computer code with ENDFVII.0 data file to generate the macroscopic cross-section of fuel and BATAN-FUEL code were used to obtain the neutronic parameter value such as fuel burn-up fractions. The calculation of the neutronic core parameters of the RSG-GAS research reactor was carried out for U3Si2-Al fuel, 250 grams of mass, with an equilibrium core strategy. The calculations show that on the last three operating cores (T90, T91, T92), all fuels meet the safety criteria and the fuel burn-up does not exceed the maximum discharge burn-up of 59%. Maximum fuel burn-up always exists in the fuel which is close to the position of control rod.

Experience of initial fuel loading of VVER-type reactors was analyzed prior to initiation of the “first criticality” phase of Novovoronezh NPP-II unit no. 1 commissioning. The analysis demonstrated a number of negative factors which may develop during the commissioning phase under discussion, for instance those associated with fuel assembly stability. Special measures were undertaken to ensure safe initial fuel loading with simultaneous use of loaded and dummy assemblies. Monitoring of deformation, flushing and verification of dummy fuel assemblies were applied for ensuring safe first fuel loading with simultaneous loading of fuel and dummy assemblies in the reactor. Conventional method of nuclear fuel loading ensuring resistance of partially loaded reactor core against internal and external disturbances of natural and man-inflicted character (in particular, against seismic effects) was refined taking into account the revealed issues and the experience of start-up of new power units by the use in the implementation of initial loading of regular nuclear reactor core with fuel assembly imitators. Simultaneous loading of charged FAs and dummy FAs in the reactor core was used as applied to VVER reactors for the first time. A set of measures was suggested allowing formulating the conclusion about applicability of dummy FAs for joint use with regular FAs. Control of deformations, flushing and inspection of FA imitators for ensuring safe initial core load in the case of joint loading of FAs and dummy FAs in the reactor core were implemented. Additional equipment was implemented for controlling coolant level in the reactor core and concentration of boric acid in the process of initial loading of VVER-1200 reactor core, because low level and absence of coolant circulation in the core do not allow using standard control systems. Effects of calculation parameters and high sensitivity of detectors on the control of neutron flux in the course of implementation of nuclear fuel loading were investigated.