Supercritical Reactor Research Articles

Evaluating Nuclear Forensic Signatures for Advanced Reactor Deployment: A Research Priority Assessment

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

Studies of the Thermalhydraulics Subchannel Code ASSERT-PV 3.2-SC for Supercritical Applications

Abstract Over the last decade, several international thermalhydraulics benchmarking efforts have been carried out to support the development of the Generation IV supercritical water-cooled reactor (SCWR) concept. These benchmarking efforts aimed to assess the readiness of computer codes to predict the thermalhydraulics behavior of supercritical fluids for nuclear fuel assembly applications. The results from the benchmarking also shed light on knowledge gaps. Throughout the years, several advancements in this area have been achieved, resulting in relevant conclusions and observations. Furthermore, experimental campaigns have been carried out worldwide to further our knowledge on the thermalhydraulics of supercritical fluids. The nuclear industry uses the subchannel approach to study the thermalhydraulics behavior of nuclear fuel assemblies in detail. In Canada, the subchannel code advanced solution of subchannel equations in reactor thermalhydraulics—pressure velocity (ASSERT-PV) is the qualified code for subchannel applications. ASSERT-PV was modified to handle supercritical conditions, resulting in an interim code version. This publication presents relevant subchannel analyses using the interim supercritical version of ASSERT-PV for fuel assemblies cooled with supercritical fluids.

Searching for optimal reactivity management using a new burnable absorber for SCLWR with (Th, 233U)O2

This work investigates the efficiency of some new burnable absorbers (BAs) suggested to provide possible improvements in the fuel management of a supercritical light water reactor (SCLWR) assembly fueled with (Th, 233U)O2. Four BAs including gadolinium, Erbium, Iridium and Lutetium are suggested as integral burnable absorber (IBA) rods. A three-dimensional model of SCLWR has been modelled using MCNPX version 2.7. A complete neutronic analysis has been carried out for the suggested BAs and compared with the gadolinium vector which is the common BAs. The IBA rods were distributed through the SCWR fuel assembly in a way that provided a flat power distribution. Various concentrations of the suggested BAs have been investigated to obtain the optimum value that can suppress the excess reactivity at the beginning of the cycle and flatten the infinite multiplication factor. The burnup parameters including the fertile, fissile, BAs and most effective fission product concentration as a function of effective full power days (EFPDs) of the SCLWR have been examined for the suggested BAs. The reactivity temperature coefficients have been studied to ensure the viability of the suggested BAs. A further analysis is performed to study the axial, radial power distribution and local power peaking factor.

Searching for an optimal fuel for a supercritical water reactor (SCWR) to dispose of minor actinides resulting from the nuclear waste

This work investigates the optimal treatment of Minor Actinides (MAs) produced in the spent fuel of nuclear reactors operating around the world. This process is very important in addressing the challenges associated with nuclear waste and reducing environmental impact. A three-dimensional model of a supercritical water reactor (SCWR) has been designed using MCNPX to make a comprehensive neutronic analysis. MAs with a concentration of 1 % have been added to UO2, (Th, U)O2, (Th, 233U)O2 and (Th, rgPu)O2. These fuels have been uploaded in the SCWR assembly and burned in a separate fuel cycle. The infinity multiplication factor (kinf) of the suggested has been investigated with and without adding the minor actinides to analyze the effect of MAs on the reactor reactivity. The fuel constituents, plutonium concentration, MAs concentration and transmutation rate have been tracked with fuel burnup. The reactivity temperature coefficients have been calculated for the suggested cases to ensure the validity of the suggested fuels. The power peaking factor (PPF) and radial power distribution have been calculated for the suggested fuels. The neutronic analysis confirms the suitability of the suggested fuel in burning a significant amount of MAs.

Microfluidic supercritical CO2 applications: Solvent extraction, nanoparticle synthesis, and chemical reaction

Supercritical CO2, known for its non-toxic, non-flammable and abundant properties, is well-perceived as a green alternative to hazardous organic solvents. It has attracted considerable interest in food, pharmaceuticals, chromatography, and catalysis fields. When supercritical CO2 is integrated into microfluidic systems, it offers several advantages compared to conventional macro-scale supercritical reactors. These include optical transparency, small volume, rapid reaction, and precise manipulation of fluids, making microfluidics a versatile tool for process optimization and fundamental studies of extraction and reaction kinetics in supercritical CO2 applications. Moreover, the small length scale of microfluidics allows for the production of uniform nanoparticles with reduced particle size, beneficial for nanomaterial synthesis. In this perspective, we review microfluidic investigations involving supercritical CO2, with a particular focus on three primary applications, namely, solvent extraction, nanoparticle synthesis, and chemical reactions. We provide a summary of the experimental innovations, key mechanisms, and principle findings from these microfluidic studies, aiming to spark further interest. Finally, we conclude this review with some discussion on the future perspectives in this field.

Development of a transient analysis code for trans-critical simulation of SCWR

Supercritical water-cooled reactors (SCWR) experience pressure reductions below the critical point during a loss of coolant accident (LOCA). During the reactor depressurization, boiling crises may occur, leading to pressure fluctuations and significant increases in wall temperatures, posing a severe threat to the cladding. In this study, a one-dimensional transient analysis code has been developed to incorporate one-dimensional steady-state flow equations in the fluid domain and transient thermal conductivity equations in the solid domain. The code also includes a wall heat transfer model and a model for the velocity of the moving quench front to simulate transcritical depressurization transient processes. The simulation successfully predicts the typical experimental phenomena. It is reasonable to use the critical temperature as the demarcation point between the dry and wet conditions of the axial wall of the rod bundle under transcritical conditions. By combining the experimental data with the program calculations, the critical interface for the occurrence of boiling crisis under transcritical conditions is determined using mass flow rate, heat flow density and fluid temperature as inputs. The criterion for the occurrence of CHF under transcritical conditions is obtained, which provides a reference to the safety analysis of SCWRs.

Searching for optimum design and burnable absorber for controlling the reactivity of U.S. Supper critical water reactor (SCWR)

Numerous research endeavours are currently underway to advance supercritical water reactors (SCWR), acknowledged as a cornerstone among Generation IV nuclear reactor designs. Evaluation and managing the reactivity of the reactor is a vital issue in the reactor operation. This research seeks to identify efficacious burnable absorber (BA) materials and determine an optimal spatial distribution within the fuel assembly to regulate reactivity levels effectively. Gadolinium, erbium and Lutetium have been suggested as BA in the form of integral burnable absorber (IBA) rods ((UO2 + Gd2O3), (UO2 + Er2O3) and (UO2 + Lu2O3)). Two SCWR assembly models, each featuring varied quantities and distributions of BA, have been examined. Various concentrations of the suggested BAs have been examined in the suggested models to verify the optimum cases. Burnup analyses have been conducted to evaluate the proposed cases. Different alloys of BAs including B4C+Dy2O3 and B4C+Sm2O3 have been investigated in the control rod and compared with the standard alloy B4C.

Assessment of Look-up tables for the prediction of heat transfer coefficient distribution in rod bundles cooled by supercritical water

The evaluation of available Look-Up Tables for prediction of heat transfer coefficient distribution in rod bundles cooled by supercritical water with the aim of their further use in computational analyses of various Fuel Assembles of Supercritical Water-Cooled Reactors is made. The comparison between the calculations based on Look-Up Tables with the values from empirical correlations and experimental data for smooth and wire-wrapped rod bundles was presented. The obtained results showed that Look-Up Table of the University of Ottawa, which was created to describe improved and deteriorated heat transfer regimes in round tubes, allows describing available data points with 30 % of the mean square deviation. It is noted that the presence of wire intensifies heat transfer exchange near pseudocritical temperature region but existing versions of Look-Up Tables cannot take into account this effect. Nevertheless, there is potential for further improvement in predicting the heat transfer coefficient using Look-Up Table by introducing additional correction factors.

Revenue Requirements Method for Assessing the Cost Impact of Fuel Cladding Corrosion in a Supercritical Water-Cooled Small Modular Reactor: A Methodological Review on Life Cycle Costing Corrosion

Abstract Canadian Nuclear Laboratories (CNL) is collaborating in the Joint European Canadian Chinese Development of Small Modular Reactor Technology (ECC-SMART) project to understand the corrosion behavior of the most promising candidate materials for a future supercritical water-cooled – small modular reactor (SCW-SMR). To support this aim and the project's requirements, the present study develops a costing method for assessing the impact of corrosion in a power generation cost model. This cost model builds on a methodological study of various corrosion engineering economics topics in nuclear power generation, such as the expected fuel cladding corrosion phenomena in a supercritical water-cooled reactor (SCWR) concept and estimating the main corrosion costs categories. This understanding is incorporated in a power generation cost model that applies the revenue requirements approach to life cycle costing (LCC). The LCC includes the main corrosion cost categories and a reliability factor used in assessing power generation costs, the costing of chemical species for controlling corrosion, and the present worth of revenue requirements. The method and model, therefore, provide a framework for understanding the kind of information available and needed for taking economical preventative corrosion measures for the current generation of water-cooled reactors and advanced reactors, such as the SCWR.

Characterization of the oxide film formed on alumina-forming austenitic stainless steel in deaerated, DH, DO, and OT supercritical water

The effects of water chemistry (deaerated (DO < 10 ppb), dissolved hydrogen (DH 1.6 ppm), dissolved oxygen (DO 2 ppm), and oxygen treatment (OT 300 ppb O2 and 100 ppb NH3) on the corrosion behavior of alumina-forming austenitic stainless steel (AFAs) were systematically investigated. A dual-layer oxide scale composed of an inner Fe/Cr/Al oxide layer and an outer Fe-rich oxide layer was observed on the AFAs. Water chemistry was found to have a negligible effect on the inner Fe/Cr/Al oxide layer; however, the morphology of the outer Fe-rich oxide layer was found to depend significantly on the chemistry. No facetted and dense hematite, which effectively protected the alloy and led to lower weight gain (less oxidation) over prolonged exposure, was formed in the outer oxide layer in oxidizing environments (DO and OT). Conversely, columnar magnetite with a porous structure was formed in outer oxide layer in reducing environments (deaerated and DH) in the absence of oxygen, with a higher weight gain observed. The addition of NH3 increased the pH of the supercritical water (SCW), thereby promoting alumina dissolution and rendering the OT conditions undesirable for use in a supercritical water-cooled reactor (SCWR). This study provides valuable insight into the behavior of AFAs in SCW and highlights the importance of carefully controlling SCW chemistry to optimize material performance.

Development of a new criterion for predicting critical heat flux in rod bundles with supercritical water flowing upward

Investigations of heat transfer characteristics in channels with supercritical fluids are crucial for power cycle applications, such as supercritical water-cooled reactors. To ensure the safety of the system, heat transfer deterioration (HTD) should be avoided. The criteria to identify the onset of HTD in heated rod bundles are still not determined. This study summarizes indicators for the onset of HTD in a rod bundle with supercritical water flowing upward based on observations from experimental data. Existing criteria for the tube flow are assessed by comparing predicted critical heat flux (CHF) values with experimentally identified values. Considering effects of different operating parameters, a new criterion for predicting HTD behavior in rod bundles is developed. The results indicate that the new criterion has a better performance than all existing criteria, with a mean average relative deviation of 0.26 %. Predicted CHF values are within ± 20 % of the respective experimentally identified CHF values.

Analysis of a test facility for transients in a supercritical water-cooled reactor using dynamic system scaling methodology

Depressurization in Supercritical Water-Cooled Reactors (SCWRs) presents challenges due to their unique operational conditions. Scaled-down facilities are preferred for analyzing depressurization phenomena due to the high costs associated with test facilities operating at 25 MPa. This study proposes the use of Dynamical System Scaling (DSS) methodology to analyze supercritical fluid depressurization processes and verify the similarity between the prototype and the model. A pressure vessel model is proposed, and fluid-to-fluid scaling indicates that carbon dioxide is the most suitable for similarity in this type of process. The analysis of depressurization compares a prototype that uses supercritical water with seven scaled models that use carbon dioxide as a model fluid. The similarity of these processes is assessed based on the temporal displacement rate. The results indicate error percentages below 4.5 %, validating the effectiveness of the proposed sizing method and highlighting the similarity of the depressurization processes over time.

Study of stress corrosion cracking susceptibility of irradiated ferrite-martensitic stainless steel 07Kh12NMFB in supercritical water. Part 2. Development of corrosion crack identification technique and analysis of autoclave test results

The studies of stress corrosion cracking have been carried out for the stainless ferritic-martensitic steel with chromium content of 12% irradiated to a damage neutron dose of ~12 dpa. This steel was chosen as a candidate material for the internals of supercritical water-cooled reactors (SWCR).The first part of the paper was devoted to autoclave testing of specially designed disk specimens under constant load in a supercritical water environment (at 450°C and 250 atm pressure). In this part of the article the developed technique for corrosion cracks identification is presented and analysis of autoclave tests results is performed.As a result of the experiments performed, the threshold stress values below which stress corrosion cracking initiation doesn’t occur. The values of threshold stresses are determined for the studied steel irradiated at temperatures of 390 and 550°C. Possible mechanisms of stress corrosion cracking of the studied steel are analyzed and directions for further research are proposed.

Comparison and mechanism research of corrosion behavior of materials used in supercritical water-cooled reactors

In this study, the corrosion characteristics of essential materials used in supercritical water-cooled reactors (SCWRs) were investigated. Austenitic stainless steels 304 SS and 316 SS and nickel-based alloys 625 and 800 were exposed to supercritical water (SCW) at 500 °C and 25 MPa. The results showed that 304 SS, 316 SS, and Alloy 800 formed a double oxide layer, with the outer and inner layers enriched in Fe and Cr/Ni, respectively. Alloy 625 formed outer Ni-rich oxides (mainly NiO) and inner Cr-rich oxides and Ni-Cr oxides (mainly Cr2O3 and NiCr2O4, respectively). The results indicated that the difference in the protective performance of the corrosion products was mainly due to the Cr content, and the order of corrosion resistance was Alloy 625 > Alloy 800 > 304 SS > 316 SS. In addition, a corrosion mechanism model was established, which lays a good foundation for corrosion prediction and protection in SCWRs.

Molecular dynamics simulation and experimental analysis of nucleation and growth mechanism of mixed inorganic salts in supercritical water

Supercritical water gasification (SCWG) technology is one of the more promising approaches for processing oily wastewater. However, there needs to be more research on the sedimentation of mixed inorganic salts in the supercritical water reactor while treating oily wastewater. In this study, the clustering process of mixed inorganic chlorides containing NaCl, KCl, and CaCl2 in supercritical water was simulated through molecular dynamics simulation. Molecular dynamics simulation was mainly conducted for mixed inorganic salts within the parameter range of 673 K-1073 K and 22 MPa-28 MPa. Combined with Yasuoka and Matsumoto's cluster theory, it can be concluded that the nucleation process of mixed inorganic salts in the supercritical water environment is mainly affected by density. Although the ion diffusion coefficient increases with temperature, the nucleation rate of inorganic salts gradually decreases due to the decrease in density. In the temperature range of 673 K-1073 K, the nucleation rate of mixed inorganic salts decreased from 34.96 to 1.65 1036 m−3·s−1. At the same pressure, the crystal growth rate parameter decreased gradually with increasing temperature, reaching about 168.25–60.09 m·s−1. Finally, the simulation results were verified experimentally, confirming that the crystal growth rate decreases with increasing temperature.

Control system design for a pressure-tube-type supercritical water-cooled nuclear reactor via a higher order sliding mode method

Control system design for a pressure-tube-type supercritical water-cooled nuclear reactor via a higher order sliding mode method

Direct numerical simulation of flow and heat transfer of supercritical water with different heat fluxes

The flow and heat transfer characteristics of supercritical water (SCW) are important considerations in the thermal-hydraulic design of the Supercritical Water-Cooled Reactor (SCWR). In the present study, the direct numerical simulation (DNS), developed with OpenFOAM, was utilized to investigate the effect of heat flux on the flow and heat transfer of SCW in a vertical circular pipe. The investigation involves detailed analyses of the wall temperature, velocity distribution, and turbulent statistics at various heat fluxes. It was found that buoyancy first impairs heat transfer, and then recovers heat transfer. This is attributed to the variation of buoyancy production and turbulence production, which are resulted from the buoyancy direct effect and indirect effect, respectively. As the heat flux increases, the recovery occurs earlier with higher magnitude. The vortex structures based on Q-criterion clearly illustrate the same process as well. Several heat-transfer correlations were assessed based on the DNS results. It indicates that Dittus and Boelter correlation (1930) cannot accurately predict the heat-transfer characteristics of SCW, while Chen and Fang correlation (2014) exhibits a higher level of predictive accuracy for Nusselt number. In addition, several turbulence models were evaluated against the DNS data. The results show significant deviations in predicting the heat transfer of SCW. The large discrepancy may be due to the inappropriate model coefficients, the failed predication of turbulence kinetic energy, as well as the constant turbulent Prandtl number commonly applied.

Review of SCWR research in Japan

SCWR (supercritical water cooled reactor) is one of the Generation IV reactors. This review summarizes the results of SCWR design concept development through numerical simulations carried out by the author-led team at the University of Tokyo and Waseda University from 1989 to 2014. They are core design, subchannel analysis, statistical thermal design, fuel rod design, development of fuel integrity criteria, plant system and heat balance, plant dynamics analysis, plant control, startup system, stability analysis, safety principle, safety criteria, safety analysis, transient subchannel analysis and fast reactor SCWR. A brief summary of experimental results on thermal hydraulics, materials and water chemistry follows. Discussion includes SCPR study by Japanese BWR manufacturers, comments on the 2005 INEEL SCWR report, misconceptions about SCWR and commercialization challenges. The SCWR is an innovation in light water reactors based on supercritical coal-fired power plant technology that has been in use worldwide for over half a century. This review covers most of the SCWR design and analysis. For researchers, it is a good subject to understand the design and analysis of light water reactors.

Nonlinear analysis of coupled neutronic-thermohydraulic stability characteristics of supercritical water-cooled reactor

The Supercritical Water-Cooled Reactor (SCWR) exhibits significant changes in thermophysical properties of the coolant, but its dynamic characteristics are different from existing reactors, due to the supercritical conditions of the coolant. This necessitates the study of the stability behaviour of SCWR, which requires a dynamic model of the reactor. In this work, a simple unsteady lumped parameter model (LPM) for SCWR has been developed. The LPM includes point reactor kinetics for neutron balance and a two-region model for fuel and coolant thermal hydraulics. The two regions are separated from each other by a boundary that depends on the pseudocritical temperature of the coolant. Regions of stable and unstable operation are identified in the parameter space by linear stability analysis. Bifurcation analysis is carried out to capture the non-linear dynamics of the system. Generalised Hopf (GH) points are located, and parameter ranges for supercritical (soft and safe) and subcritical (hard and dangerous) Hopf bifurcation are identified. The type of transient behaviour of the system for finite (though small) perturbations is predicted based on this analysis. The existence of predicted limit cycles is confirmed by numerical simulation of the nonlinear ODEs, using the shooting technique in case of unstable limit cycles. Furthermore, the effects of geometric, control and neutronics parameters on the stability of the system are studied.

Measurement system for in-situ estimation of instantaneous corrosion rate in supercritical water

Current research of cladding and in-core materials for the supercritical water reactor concept is focused on Fe-Cr-Ni alloys (e.g. 800 H, 310 S). Along the standard measurements of general corrosion rate of materials by exposure testing in autoclaves, where poor reproducibility is an issue, especially when comparing results from different experimental facilities, a good alternative may be the use of in-situ electrochemical measurements. The use of electrochemical methods is, however, more challenging due to complicated instrumentation as well as due to the supercritical water conditions. This paper is focused on the design of an experimental facility for in-situ electrochemical testing at supercritical water conditions, i.e. main circulating loop with autoclave, including the design of the electrode system, electrode leads and their surface insulation coating. Stability of electrochemical measurements has been tested using 2- and 3-electrode set-ups. Finally, surface insulation coating of the electrode leads with the best chemical and mechanical stability during experiment has been chosen for the presented application.