Supercritical Reactor Research Articles

Kinetics of formic acid decomposition in subcritical and supercritical water – a Raman spectroscopic study

The decomposition of formic acid is studied in a continuous sub- or supercritical water reactor at temperatures between 300 and 430 °C, a pressure of 25 MPa, residence times between 4 and 65 s, and a feedstock concentration of 3.6 wt%. In situ Raman spectroscopy is used to produce real-time data and accurately quantify decomposition product yields of H2, CO2, and CO. Collected spectra are used to determine global decomposition rates and kinetic rates for individual reaction pathways. First-order global Arrhenius parameters are determined as log A (s−1) = 1.6 ± 0.20 and EA = 9.5 ± 0.55 kcal/mol for subcritical decomposition, and log A (s−1) = 12.56 ± 1.96 and EA = 41.90 ± 6.08 kcal/mol for supercritical decomposition. Subcritical and supercritical Arrhenius parameters for individual pathways are proposed. The variance in rate parameters is likely due to changing thermophysical properties of water across the critical point. There is strong evidence for a surface catalyzed free-radical mechanism responsible for rapid decomposition above the critical point, facilitated by low density at supercritical conditions.

Assessment of heat transfer correlations in the sub-channels of proposed rod bundle geometry for supercritical water reactor

There are heat transfer correlations for heat transfer analysis in single tube geometries after several experimental and theoretical heat transfer studies in these single tube geometries. This is not the case for heat transfer analysis in rod bundle geometry with regard to proposed square fuel assembly of the Supercritical-Water-Cooled Reactor (SCWR) European Atomic Energy (EURATOM) design. Thus limited heat transfer studies exist on rod bundle geometry at supercritical pressures. Heat transfer correlations with accurate prediction capabilities of coolant and wall temperatures will be helpful in carrying out heat transfer studies at supercritical pressures. This paper presents the performance of twelve selected heat transfer correlations assessed on the 1/8th bare square fuel assembly of the SCWR EURATOM design using Simulation of Turbulent flow in Arbitrary Regions Computational Continuum Mechanics C ++ based code (STAR-CCM + CFD code). The obtained numerical results were compared with the results obtained by Waata numerical experimentation. Overall, the Cheng et al. correlation provided the most satisfying prediction for the wall temperatures in all the sub-channels and captured closely Wataa's Numerical data. The maximum wall temperature was obtained in sub-channel 9, the hottest sub-channel and exceeded the design limit 620 °C by 60 °C for the Cheng correlation. The difference in temperature between the hottest and coldest sub-channels 9 and 1 respectively was approximately 80 °C. It was found that Cheng correlation is best suited for heat transfer prediction in rod bundle geometry at supercritical pressures with regard to the proposed square fuel assembly of the SCWR EURATOM design. It was also found that the different numerical tools adopted for this study and Waata study were able to capture the trends of normal, enhanced and deteriorated heat transfer regimes normally observed at supercritical pressures. Nevertheless, experimental investigations involving rod bundles adopted in this study should be conducted to validate the results obtained numerically and address the inconsistency of the conclusions drawn when compared with Waata data and other similar studies.

Heat transfer to water near the critical point: evaluation of the ATHLET thermal-hydraulic system code

Abstract The heat transfer coefficient is an essential measure in the predesign of supercritical water-cooled reactors (SCWRs). At supercritical pressures, three distinct heat transfer modes exist: normal, improved, and deteriorated. The heat transfer behavior of supercritical water in the pseudo-critical range is different from that of single-phase fluids in the subcritical range. These heat transfer modes differ from those of single-phase flow at subcritical pressures, resulting in an unusual behavior of the heat transfer coefficients. Moreover, during accidental scenarios, when the operating pressure is reduced from supercritical to subcritical conditions, a boiling crisis may occur. During pressure reduction, temporary phenomena such as superheating of the cladding temperature can endanger the safe operation of SCWRs. In order to analyze operational and accidental scenarios of SCWRs, thermal-hydraulic system codes such as ATHLET are applied. However, the prediction capabilities of thermal-hydraulic system codes rely on a comprehensive validation work based on experimental data. This study presents an extensive analysis of the applicability of ATHLET at the near-critical pressure range. ATHLET is assessed against the LESHP-database and two trans-critical transient experiments. At supercritical pressures, the heat transfer coefficient correlations are evaluated with regard to their prediction accuracy and numerical problems including the “multiple solutions problems”. The trans-critical transient experiments are used to test the prediction capability of ATHLET with respect to transient heat transfer phenomena including critical heat flux, film boiling and return to nucleate boiling.

Experimental investigation on heat transfer to supercritical water in a three-rod bundle with spacer grids

Accurate prediction of the heat transfer performance in tight rod bundles is important for the design and safety analysis of supercritical water-cooled reactors (SCWRs). However, the error of the current correlations for predicting supercritical heat transfer coefficient in rod bundle and grid spacer is large under certain conditions. In the present paper, an experimental study of heat transfer to supercritical water in a three-rod bundle with spacer grids is carried out at Shanghai Jiao Tong University. The rod bundle is composed of three heated rods and three unheated rod-like fillers with an outer diameter of 15 mm and a pitch-to-diameter ratio of 1.2. Honeycomb grid spacers with a blockage ratio of 0.158 are equipped in the bundle. The system pressure is 23 MPa and the mass flux ranges from 800 to 1600 kg/(m2·s). Moreover, the heat flux varies from 400 to 700 kW/m2 and the fluid temperature is in the range of 285–382 °C. Wall-temperature gradients are observed along the circumference direction of heated rods. The wall temperature is higher in the gap area between heated rods and it is lower in the cold-wall area. When approaching the pseudo-critical point, the non-uniformity of the wall temperature is significantly reduced. Six existing correlations for fully-developed heat transfer at supercritical pressure are assessed against the experimental data, among which the Bishop correlation gives the best predictions. A new correlation is developed based on the experimental data and most of them are predicted within ±15% with a mean absolute deviation of 6.0% and a root-mean-square deviation of 7.3%. The enhancement ratios of heat transfer downstream of spacers are substantially under-predicted by the commonly used correlations. Based on the observation that the maximum enhancement ratio increases linearly with the Reynolds number in pseudo-liquid regions, a new correlation is proposed to predict the heat transfer augmentation downstream of spacers, correlating most of the data within ±20% with a mean absolute deviation of 6.2% and a root-mean-square deviation of 8.4%.

РАЗВИТИЕ НАПРАВЛЕНИЯ SCWR ОТ КОНЦЕПЦИИ ДО ТЕСТОВОГО РЕАКТОРА

The nuclear reactor cooled by supercritical pressure water - SCWR (Supercritical Water-cooled Reactor) is adopted as one of the promising reactors of the IV-generation. SCWR conceptual proposals are developed by more than 45 organizations in 16 countries with developed nuclear energy. The SCWR concept is based on the implementation of a single-circuit direct-flow scheme of nuclear power plants cooled by SKD water. The introduction of reactors of this type will increase the efficiency up to 45 %, increase the coefficient of fuel reproduction, reduce metal consumption and construction volumes, improve economic and environmental performance. Countries participating in the MFP in the direction of SCWR consider the development of a reactor with a thermal spectrum of neutrons and uranium fuel to be a priority, but in the subsequent stages, with an increase in problems with the storage of spent nuclear fuel (SNF) and Junior actinides (MA), it is possible to move to a reactor with a fast neutron spectrum, MOX fuel and a closed fuel cycle (ZTC). For ~10 years, the SSC RF - FEI and OKB "Gidropress" have been working together on the conceptual design of VVER-SKD - single-circuit RP with SCD coolant with a fast-resonance neutron spectrum with a capacity of Ne=1700 MW. This rector is recognized as the prospect of development of VVER technology with the possibility of using uranium fuel and the transition in the future to MOX-based fuel (U-Pu-Th) and to ZTZ. Rosatom recognizes this area as an innovative one, and has signed systemic agreements on Russia's participation in the work of the MFP in the SCWR direction. The paper presents the results of computational studies on the main version of the high-power reactor, as well as the test N=30 MW.

Improving breeding performance of Super FR with fuel shuffling in multi-axial layers

Super FR is the fast reactor version of SuperCritical Light Water Reactor (SCWR), which operates under supercritical condition of light water, so that designing the core with high coolant outlet temperature above the pseudo-critical point is possible. To raise the core outlet temperature of Super FR, axially heterogeneous core concept was proposed by the preceding study. It consisted of alternatively arranged two seed layers and two blanket layers (i.e., four-layers core). Furthermore, the core was axially divided to two sections and fuel shuffling was considered in each of the two axial sections independently to improve breeding performance. However, the study assumed the same fuel depletion history along the axial direction of the core, despite the large coolant density change. The core outlet temperature was only 387 C and the Compound System Doubling Time (CSDT) was about 98 years, which was not as short as desired. Hence, this study aims to reveal more comprehensive understanding of the concept of improving the performance of Super FR with fuel shuffling in multi-axial layers through design analyses. A five-layers core is considered, which consists of the lower blanket, lower seed, inner blanket, upper seed, and the upper blanket layers. Influences of the lower blanket/upper blanket discharge burnup on the core characteristics, such as the CSDT, void reactivity, and core outlet temperature have been studied with different coolant inlet temperatures for the first time. Different fuel depletion histories are considered for each of the axial layers with consideration of the axial coolant density change (which had not been considered in the preceding study). The present results show that it is preferable to keep the lower blanket discharge burnup relatively low, compared with that of the upper blanket to improve breeding performance with respect to achieving short CSDT. However, it leads to slight reduction in the core outlet temperature. Based on the sensitivity analyses, a representative design is proposed, which achieves CSDT of 74 years and average core outlet temperature of 492 C, which show significant improvement from the preceding design.

Analysis of Shutdown System Effectiveness in the Canadian Super Critical Water Reactor Using Coupled Thermal Hydraulics and Three-Dimensional Neutron Kinetics

Abstract The Canadian supercritical water-cooled reactor concept features a re-entrant fuel channel wherein coolant first travels down a center flow tube and then up around the fuel elements. Previous work demonstrated that in cases of sudden coolant flow reduction or reversal (such as that which would result from a large pipe break near the core inlet), the coolant density reduction around the fuel has a positive reactivity effect that results in a power excursion. Such a transient is inherently self-terminating since the inevitable density reduction in the center flow tube has a very large negative reactivity effect. Nevertheless, a brief power pulse would ensue. In this work, the possibility of mitigating the power pulse with a fast-acting shutdown system was explored. The shutdown system model, consisting of bottom-inserted neutron absorbing blades and realistic estimates of insertion rates and trip conditions, was added to a full-core coupled spatial neutron kinetics and thermal-hydraulics model. It was demonstrated that such a system can effectively mitigate both the peak magnitude of the power excursion and its duration.

Entropy generation study for a supercritical water reactor (SCWR)

The main objective of the present work was to analyze thermodynamic performance for a supercritical water reactor (SCWR) concept carried out via an entropy generation approach. In this work, a square array of subchannel was applied. Using powerful variation of thermophysical properties in the supercritical region of water can lead to an efficient nuclear plant design. From the thermodynamic point of view, desirable results can be gained when minimum entropy is produced at the supercritical point, which consequently leads to higher system efficiency. Effects of important thermal hydraulic parameters such as mass flux, flow Reynolds number, system heat flux, system pressure and system temperature on the generation of entropy within a SCWR subchannel were evaluated. It was found that regions closer to the supercritical point caused the entropy generation to decrease significantly.

Desain Inti Reaktor SCWR (Supercritical Water Reactor) Model Teras Silinder (r, z) dengan Bahan Bakar Thorium Hasil Daur Ulang

The Research of the supercritical water reactor (SCWR) core design of the cylindrical core model (r, z) using the SRAC program has been done. The SRAC basic code was PIJ and CITATION. PIJ was used to calculate the fuel level and CITATION was used to calculate the reactor core level. The calculation of the reactor core has been done on the 1/4 cylinder core (r, z) and the geometry of the fuel cell was the cylindrical cell. Reactor fuel material was thorium burned 40 GWd/t and 30 GWd/t. The neutron parameters in this research were fuel enrichment, burn up, reactor core size, reactor core configurations, multiplication factor, and power density distribution. Multiplication factor (k-effective) in this research was 1.000004, which is reactor was in a critical condition. The reactor core in critical condition had the size of radius (r) was 130 cm, height (z) was 270 cm and fuel enrichment 2.8262%. The maximum power density was 130.0808 Watts /cm3 which was located at a radius of 25 cm and 135 cm high. The peak power factor in the radial direction was 1.6063 and the peak power factor in the axial direction was 1.3189.

Current status of research on heat transfer in forced convection of fluids at supercritical pressures

Investigation of heat transfer at supercritical pressures began as early as the 1930s with the study of free-convection heat transfer to fluids at a near-critical point. In the 1950s, the concept of using supercritical “steam” to increase thermal efficiency of fossil-fired thermal power plants became an attractive option. Germany, USA, the former USSR, and some other countries extensively studied supercritical heat transfer during the 1950s till the 1980s. This research was primarily focused on bare circular tubes cooled with SuperCritical Water (SCW). However, some studies were performed with modeling fluids such as SC carbon dioxide and refrigerants instead of SCW. Currently, the use of supercritical “steam” in fossil-fired thermal power plants is the largest industrial application of fluids at supercritical pressures.Near the end of the 1950s and at the beginning of the 1960s, several studies were conducted to investigate a potential of using SCW as a coolant in nuclear reactors. However, these activities were abandoned for some time and regain momentum in 1990s. In support of development of SuperCritical Water-cooled nuclear-Reactor (SCWR) concepts, first experiments have been started in annular and various bundle-flow geometries. At the same time, more numerical and CFD studies have been performed in support of our limited knowledge on specifics of heat transfer at SC Pressures (SCPs) in various flow geometries.As the first step in this process, heat transfer to SCW in vertical bare tubes can be investigated as a conservative approach (in general, heat transfer in fuel bundles will be enhanced with various types of appendages, i.e., grids, end plates, spacers, bearing pads, fins, ribs, etc.). New experiments in 1990s–2000s were triggered by several reasons: 1) thermophysical properties of SCW have been updated from 1950s to 1970s, e.g., a peak in thermal conductivity in the critical/pseudocritical points was “officially” introduced in 1990s; 2) experimental techniques have been improved; 3) in SCWRs various bundle flow geometries will be used instead of bare-tube geometry; and 4) in SC steam generators of thermal power plants larger diameter tubes/pipes (20–40 mm) are used, however, in SCWRs hydraulic-equivalent diameters of proposed bundles will be within 5–12 mm.A comparison of selected SCW heat-transfer correlations has shown that their results may differ from one to another by more than 200%. Based on these comparisons, it became evident that there is a need for a reliable, accurate and wide-range SCW heat-transfer correlation to be used. Therefore, the objective of this paper is to summarise in concise form the work performed concerning with the heat transfer at SCPs, based mainly on examples of water, carbon dioxide, and R-12.

Effect of Steam Temperature and Pressure on the Oxidation of Potential Coating Alloy FeCrAl for Supercritical Water-Cooled Reactor Application

MCrAl (M = Fe, Ni, or Co) alloys have exceptional corrosion and oxidation resistance and can be used as both oxidation resistant structural materials and coatings. As coatings, they protect high temperature steels or Ni based alloys by forming a dense alumina layer on the surface and thus impeding further oxidation. In order to assess its potential usage as an overlay coating on components used in supercritical water-cooled nuclear reactors (SCWRs), an Fe-2 3Cr-5Al alloy in the form of wire was tested under two different super-heated steam (SHS) conditions (625 °C and 800 °C) and also in supercritical water (SCW) (625 °C and 26 MPa), for 500 h. The corrosion behavior of samples was assessed by measuring the weight change per unit surface area and by examining the surface, cross-sectional microstructure and the phase compositions using scanning electron microscopy (SEM) and X-ray diffraction (XRD). The tested samples showed different oxidation behavior after exposure to these three conditions. SEM and XRD results showed that FeCrAl has the ability to form protective Al- and Cr-containing oxide(s) under all three conditions. Based on the findings, it is concluded that the oxidation behavior of Fe–23Cr–5Al is highly influenced by pressure and temperature within the range of testing conditions. SHS exposure at low temperature led to greater weight gain while that in SCW resulted in weight loss. Overall, its performance is better under SHS conditions compared to CoCrWC S16 but worse under the SCW condition.

High Quality Syngas Production with Supercritical Biomass Gasification Integrated with a Water–Gas Shift Reactor

A thermodynamic assessment is conducted for a new configuration of a supercritical water gasification plant with a water–gas shift reactor. The proposed configuration offers the potential for the production of syngas at different H2:CO ratios for various applications such as the Fischer–Tropsch process or fuel cells, and it is a path for addressing the common challenges associated with conventional gasification plants such as nitrogen dilution and ash separation. The proposed concept consists of two reactors, R1 and R2, where the carbon containing fuel is gasified (in reactor R1) and in reactor R2, the quality of the syngas (H2:CO ratio) is substantially improved. Reactor R1 is a supercritical water gasifier and reactor R2 is a water–gas shift reactor. The proposed concept was modelled using the Gibbs minimization method with HSC chemistry software. Our results show that the supercritical water to fuel ratio (SCW/C) is a key parameter for determining the quality of syngas (molar ratio of H2:CO) and the carbon conversion reaches 100%, when the SWC/C ratio ranges between two and 2.5 at 500–1000 °C.

Study on the Movement and Deposition of Particles in Supercritical Water Natural Circulation Based on Grey Correlation Theory

The movement and deposition of particles that occur during their natural circulation in supercritical water exercise an important impact on the safe and stable operation of a supercritical water reactor (SCWR). When supercritical water flows in pipelines, a large number of corrosive particles may be generated due to pipeline corrosion or the purity of the fluid itself. The presence of particulate matter affects the heat transfer efficiency of the pipeline, increasing flow resistance and easily promoting heat transfer deterioration. ANSYS-CFX numerical analysis software was used to simulate the natural circulation loop of supercritical water, and micron particles were added in the initial flow field. The effects of heating power, particle concentration and particle diameter on particle deposition were obtained. Through this analysis, it can be concluded that the heating of the pipeline has a certain inhibitory effect on the deposition of particles. The rise in both initial particle concentration and particle diameter serve to reinforce the deposition of particles in the heating section. Depending on the degree of influence, the contributory parameters to particle deposition include particle diameter, particle concentration and heating power in turn.

Control and thermal analysis for SCWR startup

Startup system, the design of startup sequences analysis is an important part of SCWR design. A thermal hydraulic system analysis code for supercritical water reactor named SCTRAN is used to model the entire startup system based on the circulation loop for startup and once-through direct cycle. The problem of the heat transfer coefficient (HTC) does not accurately capture deterioration phenomenon, the HTC is calculated as a discontinuity in the mode transfer region, its low prediction accuracy above the quasi-critical region have been solved by the new wall heat transfer model. Especially, the look-up table would not be used to obtain the HTC and achieves high prediction accuracy across the critical region, unless the pressure is higher than 19 MPa. After that, to get a smooth recirculation variable pressure startup process, the system model integrates the control system which can controls the temperature, the steam drum water level, the thermal power, and the coolant flow rate. Based on the CSR1000 core and entire once-through direct cycle and circulation loop for startup, four stages under control systems, from low pressure to full power condition in recirculation startup process, were analyzed with code SCTRAN and wall heat transfer model was modified. The calculation results show that the recirculation system can startup from subcritical state to full power state without issue of CHF. The control system can control the parameters quite well and maximum cladding temperature (MCST) can be limited under 650 °C in the startup process. The modified SCTRAN code in this paper can further expand the computational range and computational accuracy. The full-scale control system can meet the needs of parameters expected response.

Formation Mechanism of Epitaxial Palladium–Platinum Core–Shell Nanocatalysts in a One‐Step Supercritical Synthesis

AbstractThe scarcity of platinum group metals provides a strong incentive to optimize the catalytic activity and stability, e.g., through nanoalloys or core–shell nanoparticles. Here, time‐resolved X‐ray total scattering and transmission electron microscopy characterization are used to study the formation of palladium–platinum core–shell nanoparticles under solvothermal conditions. It is shown that Pd rapidly forms small (5–10 nm), disordered primary particles, which agglomerate and crystallize when reaching 20–25 nm. The primary Pd particles provide nucleation sites for Pt, and, with extended reaction time, the Pd cores become fully covered with Pt shells. The observed core–shell material is surprising when considering the Pt–Pd phase diagram and relative surface energies, but it can be rationalized through the kinetics of precursor conversion. To bridge the gap between scientific studies and industrial demand for large‐scale production, the synthesis process is successfully transferred to a continuous flow supercritical reactor providing a simple scalable and green process for production of bimetallic nanocatalysts.

Sub and supercritical water oxidation of pharmaceutical wastewater

Sub and supercritical water oxidation (Sub- CWO & SCWO) technology has gained attention over the past few years due to near complete conversions at shorter residences times. A pharmaceutical sample containing variety of analgesics, antibiotics, antipyretics, antifungal, β-blockers and drug intermediates contributing to overall total organic carbon (TOC) of 2017 ± 49 mg/L was collected from a nearby pharmaceutical industry for this research. The sample was subjected to sub and supercritical water oxidation both in batch and continuous processes. A maximum removal efficiency of 80.1% was achieved with excess oxidant (oxidation coefficient (OC 3)) at 400 °C for 60 min residence time in batch reactor. The pharmaceutical sample was also processed in continuous supercritical water reactor by varying both temperature (400–550 °C) and oxidant coefficient (0–2) at 23 MPa for 60 s residence time. With the operating temperature of 550 °C and OC 2 resulted in maximum removal of the pharmaceutics with efficiency of 97.8%.

On nuclear-coupled thermal-hydraulic instability analysis of Super-Critical-Light-Water-cooled-reactor (SCLWR)

The nonlinear stability analysis of a supercritical light water reactor (SCLWR) is presented using a nuclear-coupled thermal-hydraulic reduced-order model. The analytical model is developed by coupling 1. the point-kinetics equations with one group of delayed neutrons, 2. the fuel heat transfer and 3. a 1-D reduced order model which represents the heat absorption phenomenon during the coolant flow. Unlike the existing studies, which are limited to linear stability analysis, the primary objective of the work is to present the detailed nonlinear dynamics of the SCLWR system. The said goal is achieved at two levels. The first level is the linear stability analysis wherein the linear stability boundaries are shown in two sets of parameter space namely the two intrinsic reactivity feedbacks (Doppler reactivity feedback and density reactivity feedback) and the pseudo-phase-change number and pseudo subcooling number. The parametric effects show the sensitivity of the linear stability boundaries with the system parameters. In the second level, to discuss the nonlinear characteristics of the system, two types of Hopf bifurcations (subcritical and supercritical) are studied with the help of first Lyapunov coefficients of the system. Multiple numerical simulations are performed to verify the resultant limit cycle behavior associated with these bifurcations. Moreover, the occurrence of the generalized Hopf bifurcation is shown which represents the bifurcation between the subcritical Hopf and the supercritical Hopf regions. Further, inside the stable region of the linear stability boundary, the saddle-node bifurcation is found which represents the location of the turning point and the threshold of the globally stable region of the system.

Process modeling and economic analysis for bio-heavy-oil production from sewage sludge using supercritical ethanol and methanol

The solvothermal liquefaction of sewage sludge (SS) has received much attention as an ecofriendly method of liquid fuel production. In this study, we evaluated the economic feasibility of bio-heavy-oil (BHO) production from 100 t/d SS using supercritical ethanol (scEtOH) and methanol (scMeOH). The process included a dryer, supercritical reactor, BHO separator, steam boiler, wastewater treatment unit and storages. Two cases for the techno-economic analysis (TEA) were considered: BHO production with scEtOH (Case 1) and with scMeOH (Case 2). The four-level economic potential approach (4-level EP) was used for the TEA. The total capital investments of Cases 1 and 2 were 19.6 and 19.5 million dollars, respectively. Case 2 showed higher economic potential than Case 1 because of the low price of methanol. The BHO plant using scMeOH was economically feasible with a return on investment of 21%/a.

Minor Actinide Transmutation in Supercritical-CO2-Cooled and Sodium-Cooled Fast Reactors with Low Burnup Reactivity Swings

This paper presents the investigation of minor actinide (MA) transmutation in supercritical CO2-cooled and sodium-cooled fast reactors (S-CO2-FR and SFR) with the thermal output of 600 MW(thermal) for simultaneously attaining low burnup reactivity swings and reducing long-life radioactive waste. Minor actinides are loaded uniformly in the fuel of the cores, and the MA contents are determined to minimize the burnup reactivity swings. In the S-CO2-FR, the burnup reactivity swing is minimized to 0.11% ∆k/kk’ when the MA content is 6.0 wt%. In the SFR, the MA content was determined to reduce the burnup reactivity swing while maintaining sodium void reactivity under a design limitation of 5 $. The burnup reactivity swing of the SFR is reduced to 1.94% ∆k/kk’, whereas sodium void reactivity is about 4.7 $ when 10.0 wt% MAs are loaded. The low burnup reactivity swing enables minimization of control rod operation during fuel burnup. The number of control rods in the two reactors is reduced to ten, which is half of a typical sodium-cooled mixed-oxide fuel MONJU reactor without MA loading. The MA transmutation rates in the S-CO2-FR and SFR are 42.2 and 52.2 kg/year, respectively, which are equivalent to the production rates in seven and nine light water reactors of the same electrical output.

Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program

Existing conditions make possible obtaining information that being discussed openly by wide scientific community could help outlining or even establishing the expediency of a particular area of present and future research. Use link http://www.sciencedirect.com to learn about the topics or areas that most attract researchers from different countries. The Generation IV International Forum (GIF-IV) established in January 2000 has set a goal to improve the new generation of nuclear technologies in the following areas: stability, safety and reliability, economic competitiveness, proliferation resistance and physical protection. The purpose of the present publication is to prepare a discussion of one of the directions of development of fourth-generation NPPs, the groundwork for which has already been laid in thermal power engineering in various countries. The number of papers published annually on this topic is the largest among other similar topics dedicated to nuclear power plants of the fourth generation. Judging from the operating experience of existing nuclear power plants using water as a coolant, it can be ascertained that the tendency of building water-cooled nuclear power plants will remain during the next 30 to 50 years. During the present stage the task in the development of alternative types of reactors will be limited to demonstration of their performance and acceptability for future power engineering and the society. The project of supercritical water-cooled reactor is based on the operating experience of VVER, PWR, BWR reactors (more than 14,000 reactor-years); many years of experience accumulated in operating fossil thermal power plants (more than 400 power units; 20,000 years of operation of power units) using supercritical (25 MPa, 540°C) and super-supercritical (35–37 MPa, 620–700°C) water steam. In Russia more than 140 supercritical pressure units are currently in operation. Numerical calculation and design of supercritical water-cooled reactor (similarly to BR-10 reactor) will allow not only training personnel for future development of this technology, but will also help revealing the most difficult points requiring experimental confirmation with application of independent test facilities, as well as formulating the plan of first priority experimental studies. Knowledge accumulated over the last 10 years in the world allows the following: further specifying the already developed concept; developing a plan of specific priority studies; compiling task order for designing small-power pilot VVER SKP-30 reactor (30 MW-th). The scope of problems that are to be solved to substantiate VVER-SCP reactor and commence designing an experimental reactor with thermal capacity of 30 MW is the same as that in developing any type of nuclear reactor: physics of the reactor core; material related matters (primarily concerned with the reactor pressure vessel, fuel, and fuel rod cladding); thermal hydraulics of rod bundles in the near- and supercritical areas; water chemistry at supercritical pressure; corrosion of materials, development of safety systems. Research must be carried out both in static conditions and under irradiation. The absence in Russia during the extended time period of approved program with allocation of appropriate funding and preservation of the existing status during the coming two or three years will lead to the situation when Russia will be hopelessly lagging behind in the development of SCWR technology.