Core Inlet Research Articles

Neutron noise analysis of simulated mechanical and thermal-hydraulic perturbations in a PWR core

KWU pre-Konvoi PWRs (SIEMENS design) are commonly exhibiting high neutron noise levels which can lead to costly operational issues (i.e. activation of SCRAM system, operation of unrated core power, etc.). The frequency region of interest of neutron noise is below 1 Hz, which is the typical frequency range of thermal-hydraulic phenomena. This feature seems to indicate that, coolant flow and temperature oscillations can have a key impact on neutron noise phenomena. Moreover, an increasing neutron noise trend (in term of normalized root mean square) was recently observed in many KWU-PWRs. This increasing trend has been speculated to be correlated with the introduction of a new fuel design type in the KWU-PWRs. This fact indicates that there should be some correlation between neutron noise spectral characteristics and fuel assemblies’ performance. In order to advance in understanding this phenomenon, the transient nodal code SIMULATE-3K (S3K) has been used to simulate mechanical vibrations of fuel assemblies and thermal-hydraulic fluctuations of the core inlet flow and temperature. The simulated neutron detectors responses are analysed with noise analysis techniques and compared to real plant data. This analysis indicates that the cross-feedback between the mechanical and thermal-hydraulic disturbances complicate the identification of the origin of the perturbation source. The simulated results indicate that the neutron noise spectral characteristics can be associated separately to different causes. In this sense, the results of this work seem to indicate that the spectral features of the neutron noise are a consequence of both mechanical perturbations and thermal-hydraulic fluctuations.

Numerical analysis of the ROCOM boron dilution benchmark experiment using the CUPID code

Abstract The Rossendorf Coolant Mixing Model (ROCOM) boron dilution benchmark experiment was assessed using the CUPID code in the framework of a Coordinated Research Project by the International Atomic Energy Agency, namely, “Application of Computational Fluid Dynamics Codes for Nuclear Power Plant Design.” Two-equation Reynolds-Averaged Navier–Stokes (RANS) turbulence models, such as the standard k-e model, Re-Normalization Group (RNG) k-e model, and Shear Stress Transport (SST) k-ω model, were tested, and predictions of boron concentration were compared to measurements at the downcomer and core inlet in the reactor pressure vessel. The analysis result showed that the de-borated water slug injected from one cold leg was diffused by the turbulence mixing effect as it flowed downward in the downcomer and passes through a perforated drum in the lower plenum. The applied turbulence model affected the mixing patterns of the de-borated water, which resulted in a local distribution of boron concentration at the core inlet. This was the main concern of the boron dilution phenomena from the viewpoint of safety analysis. The result showed combining the CUPID code with the RANS turbulence models could reasonably predict the boron dilution phenomena in the reactor vessel when the proper y-plus wall treatment was assured, considering the characteristics of the turbulence models.

Preliminary hydraulic analysis of the distribution zone in the Dual Fluid Reactor concept

This article focuses on the validation of the basic design of the distribution zone of the Dual Fluid Reactor (DFR), which is based on the Generation IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The distribution zone located under the core joins the core and input pipes of both fluids. On the top of the core, another distribution zone specified for output is also symmetrically set up; however it will not be mentioned in the paper. This distribution zone partitions the inlet flow of the salt, determines the pressure and velocity distribution of the core inlet and influences the neutronic behaviors in the core. This article uses COMSOL as the tool to develop models and to study the pressure and velocity differences of fuel salt and coolant lead in the “original” and “alternative” configurations/designs of the distribution zone in the DFR concept. COMSOL is also used to determine the initial/boundary conditions of configuration variations of the distribution zone by hydraulic modelling and calculations. The simulation results are analyzed and the optimized design is delivered.

Flashing-driven natural circulation characteristics analysis of a natural circulation integrated pressurized water reactor

Abstract In this paper, the simulation model of a flashing-driven natural circulation integrated pressurized water reactor is established based on RELAP5/Mod3.4 code. The distinctive characteristics involved in the reactor, such as flashing-driven natural circulation and self-pressurization, are revealed. Based on the thermal-hydraulic characteristics of the reactor, the operation scheme that maintains both core outlet temperature and steam pressure constant is determined. A temperature-pressure joint control strategy is preliminarily determined. Based on the established reactor model, the thermal-hydraulic characteristics of the IP100 reactor at different thermal load are analyzed. The results indicate that, at lower load, the onset of flashing will move downwards and the coolant flow rate decreases, while the core inlet temperature is higher. However, the mutual coupling effect leads to the thermal-hydraulic parameters of the reactor non-linearly varies with the needed load. In the secondary loop, the steam temperature is higher at lower load. Furthermore, the effect of main system parameters is analyzed and the criteria for reactor operation parameters are discussed. The obtained research results are significant for deeper understanding the thermal-hydraulic performance of the flashing-driven NC-IPWR.

Numerical Study of Integral Inherently Safe Light Water Reactor in Case of Inadvertent DHR Operation Based on the Multiscale Method

In detailed previous work by the authors, an innovative decay heat removal (DHR) system has been proposed and designed for the Integral Inherently Safe Light Water Reactor (I2S-LWR). The current paper studies the inadvertent actuation of one DHR system train during I2S-LWR normal operation due to a false signal or operator action. The RELAP5 code is used to perform a one-dimensional study, and important thermal-hydraulic characteristics, including primary loop coolant flow rate, pressure, temperature, DHR primary-side flow rate, and coolant temperature, are achieved during this transient. Then, a detailed computational fluid dynamics simulation utilizing STARCCM+ is carried out to investigate the coolant mixing characteristics in the downcomer and lower plenum and obtain the local thermal-hydraulic conditions at the reactor core inlet. It is found that as a consequence of inadvertent DHR actuation, the maximum overcooling at the reactor core inlet is about 3 K, which would not result in significant reactivity insertion. Furthermore, a more severe transient of inadvertent DHR operation with intermediate loop break is studied, and the results show that this would not lead to more significant overcooling to the I2S-LWR core compared with inadvertent DHR operation without intermediate loop break. This work is an indispensable supplement for DHR system comprehensive assessment in the I2S-LWR project.

Best estimate calculation and uncertainty quantification of sodium-cooled fast reactor using MARS-LMR code

The safety analysis of nuclear power plants has been performed using conservative approach based on conservative assumptions and boundary conditions to evaluate the safety margin of plant operation. However, this conservative approach could lead to unrealistic behavior predictions and eventually distort some phenomena in reactor systems. Therefore, the nuclear field moved towards an alternative best-estimate approach with uncertainty quantification in order to improve the phenomena prediction and to decrease the excessive conservatism in safety margins. In this study, the best estimate methodology is applied to improve the accuracy and reliability in safety analysis of an SFR. The applied best estimate methodology is based on the CSAU. This methodology is composed of three unique steps for evaluation of code capability, assessment and range of parameters, and sensitivity and uncertainty analysis. The primary purpose of this study is to evaluate the appropriateness of sensitivity parameters and its ranges, which have been determined through intensive experts’ panel discussion, by use of the data obtained from the EBR-II Unprotected Loss of Flow (ULOF) experiment. The MARS-LMR thermal–hydraulic code and the parallel computing platform integrated for uncertainty and sensitivity analysis (PAPIRUS) become the basic calculation tools in the study. Confirmation of data coverage is performed through the evaluation of coolant temperature in the instrumented subassemblies XX09. The appropriateness of parameters and its ranges are evaluated for three different cases: original parameters and ranges suggested in the MIRT, ±10% increased parameter ranges, and 200% increased axial reactivity feedback coefficient only. The case with the original parameters and ranges does not result in a valid data coverage, which means inadequate modeling accuracy for the ULOF scenario. The other two cases give complete coverage of EBR-II temperature data measured at the core top, which suggest the need of further refinement of reactivity models. The relative importance of the parameters is confirmed through the sensitivity analysis with respect to the Figures of Merit (FoM). The selected dominant parameters are the sodium density reactivity, above core load pad strain coefficient, core radial expansion reactivity coefficient and fuel axial expansion reactivity coefficient. The pump coastdown curve and the core inlet form loss are also found to be significant parameters during the transient.

Численное исследование температурного поля в активной зоне РУ БРЕСТ-ОД-300 при частичной блокировке проходного сечения на входе теплоносителя

A reactor core design with shroudless fuel assemblies (FAs) has been adopted in the BREST-OD-300 reactor’s basic design. The choice of shroudless FAs is dictated by the fact that with this reactor core design, there are no separate isolated fuel rod cooling channels shaped by FA shrouds. Therefore, emergencies involving partial blocking of the core flow cross section at the core inlet that are considered for the newly designed lead-cooled reactor, will have less severe consequences for the fuel rods than they may be in the case of using shrouded FAs. Thus, with the shrouded FA’s flow cross section being fully blocked, the heat produced by the fuel rods can only be removed from them to the coolant flowing in the gap between the shrouds (both due to natural convection of coolant inside the FA and due to thermal conductivity throughout the fuel rod bundle in the affected FA, predominantly in the radial direction). Obviously, if the core power is not reduced to a significant extent, overheating of fuel rods and loss of their cladding tightness will be unavoidable. A hypothetical situation involving partial blocking of the BREST-OD-300 reactor plant’s core flow cross section was investigated by carrying out 3D calculations using a porous body model. The case of the FA flow cross section being fully blocked at the level corresponding to the first spacer grid from the bottom is considered. A conclusion can be drawn from the obtained study results that the growth of fuel rod cladding temperature that will take place when the flow cross section of even seven central FAs is blocked will not cause them to fail immediately (the cladding temperature will in this case make 790 °C), but in all likelihood, it will only result in their service life to become shorter. To obtain better accuracy of the reactor core analysis performed according to the porous body model, a more detailed reactor core model representing individual FAs and gaps between them should be used. Nonetheless, the results obtained from the performed numerical analysis show that the reactor core with shroudless fuel assemblies has a significant advantage over the design with shrouded fuel assemblies.

Mechanical analysis of the bow deformation of a row of fuel assemblies in a PWR core

Abstract Fuel assembly (FA) bow in pressurized water reactor (PWR) cores is considered to be a complex process with a large number of influencing mechanisms and several unknowns. Uncertainty and sensitivity analyses are a common way to assess the predictability of such complex phenomena. To perform such analyses, a structural model of a row of 15 FAs in the reactor core is implemented with the finite-element code ANSYS Mechanical APDL. The distribution of lateral hydraulic forces within the core row is estimated based on a two-dimensional Computational Fluid Dynamics model with porous media, assuming symmetric or asymmetric core inlet and outlet flow profiles. The influence of the creep rate on the bow amplitude is tested based on different creep models for guide tubes and fuel rods. Different FA initial states are considered: fresh FAs or FAs with higher burnup, which may be initially straight or exhibit an initial bow from previous cycles. The simulation results over one reactor cycle demonstrate that changes in the creep rate and the hydraulic conditions may have a considerable impact on the bow amplitudes and the bow patterns. A good knowledge of the specific creep behavior and the hydraulic conditions is therefore crucial for making reliable predictions.

Investigation into the impacts of distributed inlet temperature on thermal limit during LFWH event in Chinshan plant

Abstract The Condensate and Feedwater System of the Chinshan BWR units is used to provide reliable and high-quality water to maintain the reactor water level during operation. If a Loss of Feedwater Heating (LFWH) event occurs, the core inlet subcooling increases and then induces corresponding power excursion and the reactor pressure rise. In the Chinshan Final Safety Analysis Report (FSAR), a loss of the feedwater temperature of 55.6°C (100°F) is conservatively assumed in the event. This study analyzes the integral reactor system response with RETRAN. Furthermore, a partial vessel model (PVM) of CFD is used to acquire the conditions of the fuel channel inlet to compensate the weakness of the RETRAN system model to generate detailed thermal-hydraulic conditions. The evaluation shows that the feedwater temperature drop is about 40°C which is lower than the FSAR value. In addition, the sensitivity study shows that the hot channel method underestimates the ΔCPR about 0.025, and there is no direct relation between ΔCPR and either of inlet subcooling or power fraction in transient, which is quite different from the conclusion of the hot channel method. Finally, the sensitivity study also proves the ΔT of 55.6°C (100°F) used in FSAR analysis conservative enough to cover the worst channel with a margin of 0.015 in ΔCPR.

Разномасштабные задачи тепломассообмена в атомной энергетике

This paper presents a review of the results obtained in OKB «GIDROPRESS» taking into account the experience gained in the field of numerical modeling of heat and mass transfer in various nuclear power facilities. The benefits of using CFD technology in the NPP design is the ability, based on a limited number of experimental data, to study in detail the heat and mass transfer processes in the facilities in order to confirm or improve design solutions at the advanced scientific and technical level. A number of problems are presented, for solution of which the modern numerical methods of computational hydrodynamics with high-performance computer technology were used. New possibilities of computational modeling with the use of modern supercomputer computing technologies, as well as accompanying computational complexity and difficulties of analysis of results are shown. Examples of use of the considered technology for modeling experimental rigs and full-scale objects under different operating conditions are given. The automatic post-processing of the results is shown. The procedure allows to analyze large problems with mesh size up to 1 billion control volumes by integral parameters, such as mass flow distribution at the core inlet and outlet, distribution of coolant heat-up over the core fuel assemblies, etc.

Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

Abstract The 7th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7th AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7th AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

The stability-analysis code FIAT development for density wave oscillations and its application to PV/PT SCWR

SuperCritical Water Reactor owns a high operation pressure and a wide coolant temperature range. The thermo-physical properties and transport properties of the coolant in the core would change greatly both in the operation condition and startup process, which may causes density wave oscillations. In this paper, a stability analysis code named FIAT is developed based on frequency domain method to analyze the density wave oscillations (DWO) for SCWR. Frequency domain method and assumptions of constant pressure drop between core inlet and outlet are applied in the code. The neutron/thermal-hydraulic coupling characteristic equation is derived through using linearization and laplace transformation for the mass, momentum and energy conservation equations. With analyzing the characteristic equation, the onset of DWO instabilities can be confirmed with non-dimensional parameters plane. The prediction accuracy of the FIAT code is verified through experiment data and code-to-code comparison under subcritical and supercritical pressure. The FIAT code is applied to evaluate the effects of basic core model on the stability for pressure vessel type SCWR and optimize the startup procedure for pressure tube type SCWR. The results show that for pressure vessel type SCWR, fuel rods heat transfer model and neutron kinetics model help stabilize the system while water rods model produces an opposite effect. For pressure tube type SCWR, two pressurization phases are required during sliding pressure start-up process to meet both the thermal–hydraulic and stability requirements.

Computational fluid dynamics analysis on orifice structure inside valve core of pilot-control angle globe valve

A novel pilot-control angle globe valve is proposed, and it has an obvious advantage of energy conservation during its opening and closing process. In pilot-control angle globe valve, the opening and closing forces are related to the orifice located inside the valve core. In this paper, the effects of orifice diameter are thoroughly studied under different working conditions such as valve core displacements and inlet velocities. To begin with, the numerical model is validated by comparing similar angle valves, and then the flow and loss coefficients under different orifice diameters are discussed. It is found that the effects of orifice diameter on force acting on valve core depend on valve core displacement and inlet velocity. Thus different valve core displacements and inlet velocities combined with different orifice diameters are further studied. It is also found that when the orifice diameter is larger than 12 mm, pilot-control angle globe valve cannot be used under small inlet velocity or large valve core displacement. In addition, formulas to calculate forces on valve core are proposed for further orifice design. This work can be referred in process industries especially in a piping system with orifice plates or globe valves.

Experimental Evaluation of Injection Pressure and Flow Rate Effects on Geological CO2 Sequestration Using MRI

Deep Saline aquifers are promising long-term storage sites for CO2 emissions. The injection pressure is one of major factor that determines CO2 migration phase equilibrium, and the flow rate influences the injectivity behavior and plume migration. However, data and knowledge of the injection pressure and the variation of velocity in displacement process are scarce.In this paper, we describe a set of laboratory tests on two–phase flow drainage of CO2 and brine. Fundamental characteristics such as the CO2 front, the breakthrough time and sweep efficiency, residual water saturation, due to the injection pressure and rate of displacement fluid was explored. The measurement is conducted at 40°C under two different pressure conditions and four CO2 injection rate, representative of a CO2 storage operation. Magnetic resonance imaging (MRI) technique was used to observe pore-scale events in drainage process.After the injection of 3-5PV CO2, saturation of CO2 keeps constant indicating that the rest part remains inaccessible. The CO2 saturation at the last time for 8 MPa is higher than 6 MPa. Residual saturation and breakthrough time of different injection rate displacement procedures are different. The packed porous media is not perfectly homogeneous, so the CO2 run through the high permeability regions vertically in a short period. With CO2 injection, the brine saturation decreased gradually from the inlet of the bead-pack core, and the MRI image turned into dark gradually. The CO2 front proceeding in the porous media could be measured clearly, we can find CO2 “channeling” phenomena obviously in the case of gaseous CO2 displacement due to the difference of fluid viscosities and densities, and it can be seen from the profiles of 76 and 80 min. This phenomenon causes premature breakthrough of the CO2 to occur, thereby reducing the displacement efficiency of the water. And yet, the piston-like displacement occurred in the supercritical CO2 drainage experiment, due to the viscosity of supercritical CO2 is higher than gaseous CO2. According to capillary number definition, the Darcy velocity U keep invariable with constant CO2 flow rate, and the non-wetting phase (CO2) viscosity increase with the injection pressure increase. In other words, the displacement process capillary number of 8 MPa is higher than 6 MPa. The sweep efficiency in 8 MPa is slightly higher compared to the CO2 injection in 6 MPa. Therefore, supercritical CO2 displacement process is more practical than gaseous CO2 displacement process.The sweep efficiency will decrease for higher capillary number. At a small CO2 injection rate, the breakthrough time is long. A longer break through time resulted higher sweep efficiency. The CO2 flooding increases the displacement efficiency by raising the capillary number due to the relatively low interfacial tension values between the brine and the injected CO2. At high injection rate, especially 0.8 ml/min, the MRI image near the outlet region remain at high signal intensity in a relatively long time, indicating that the CO2-brine distribution does not change, since the displacement is steady. The sweep efficiency decreases as the injection flow rate increase, but the decrease extent reduces gradually. This means there is a critical capillary number for CO2 sequestration, which results in minimum sweep efficiency. In future experimental work, we will go into more detail about the critical capillary number. Within the capillary number scope of 10-9 to10-10, higher capillary number will decrease the CO2 sweep efficiency.

Reducing CO2 Flow Using Foams

As part of the MiReCOL three-year European project (www.mirecol-co2.eu) on storage remediation technologies, we studied in the laboratory the capacity of foams to reduce gas flow for CO2-brine systems in rock core sample with common surfactants, as a function of interstitial velocity and gas to water fraction. Two different types of experimental set-up are used. They both allow local measurement of the water saturation for low and high pressure/temperature condition. A small MRI core-flood set-up is used to perform experiments at 35°C and 10 bars pore pressure, in CO2 non-dense condition. In order to work under reservoir condition, at 40°C and 130 bars, with dense super-critical CO2 we used a classical core-flood system coupled to an X-ray detection set-up.All experiments were carried out in similar Clashach sandstones with permeability between 220 and 1500 mD, and porosity in the range 10-20%. The gas and the surfactant-brine solution were co-injected at the core inlet face with a gas fraction around 0.7. We vary the interstitial velocity within two decades from about 3 ft/day up to 100 ft/day.The performance of the generated foams was evaluated from the relative foam viscosity, the ratio of the measured pressure drop in the presence of foam to the pressure drop in single phase condition for the same interstitial velocity. Whatever the pressure and permeability/porosity, the relative foam viscosity can be described as a power law vs. the shear rate evaluated from the interstitial velocity, permeability and porosity. The exponent is close to -1 describing the shear-thinning behavior.

Assessment of post-LOCA long term cooling considering the in-vessel downstream effect

A method is discussed to calculate the long-term cooling phase following a Loss-Of-Coolant-Accident (LOCA) considering the In-vessel Downstream Effect (IDE) using a thermal-hydraulic code, MARS-KS1.3. There are two major concerns in considering the in-vessel downstream effect; the first is a simulation of blockage effect of debris at the core inlet and the second is an incorporation of the effect of chemical deposition at the fuel surface. Regarding the debris blockage effect, the present method takes into account the results of the tests which were conducted using a prototypical fuel assembly to determine the maximum pressure drop which corresponds to the maximum potential debris accumulation. And the result of the calculation by LOCA Deposition Model (LOCADM), the thickness and thermal conductivity of the debris layer, was used. Also, additional calculations using a transient fuel specific performance code, FRAPTRAN is conducted to confirm the validity of the present method. Using the present method the peak cladding temperature during post-LOCA long-term cooling phase considering in-vessel downstream effect can be predicted for evaluating the compliance to the acceptance criteria.

Magnetic-Resonance Imaging of Fines Migration in Berea Sandstone

Summary Fines migration is a phenomenon of practical importance in the petroleum-production and drilling industry. The movement of clay particles, induced by incompatible aqueous-phase chemistry or high flow rate, obstructs pore throats downstream of the fluid flow, leading to permeability reductions that can be as large as two orders of magnitude. Magnetic-resonance-imaging (MRI) methods derived from the Carr-Purcell-Meiboom-Gill (CPMG) method (Meiboom and Gill 1958) can map T2 distributions in porous rocks, hence showing the spatial variation of the pseudo-pore-size distribution. In this work, the traditional water-shock experiment was used to mobilize clay particles in the aqueous phase flowing in Berea core plugs. Spin-echo single-point imaging (SE-SPI), a phase-encoding MRI method derived from the CPMG method, was used to determine spatially resolved T2 spectra of the samples, and therefore the pseudo-pore-size distributions. The shift in the T2 spectra of the core inlet and outlet showed opposite trends. The pore-size distribution of the inlet and outlet, inferred from T2 distributions, were shifted to larger and smaller values, respectively. Therefore, the average pore size was increased at the inlet of the core and reduced at the outlet of the core. This MRI method provides a new analytical approach to screen reservoirs for potential fines-migration problems.

Risk assessment of thermal striping in MYRRHA research reactor

The thermal striping is one of the problems encountered in nuclear facilities. It may affect the reactor main components structural integrity and, sometimes, it can cause their rapid failure.The thermal striping problem touches, in general, mixing zones of industrial facilities. While it affects several areas as the core of liquid metal reactors that is due to the larger difference between core inlet and core outlet temperature, these reactors are more vulnerable to striping compared to water reactors.It is true that a better understanding of the induced thermal loads and associated damage mechanisms are obtained nowadays, thanks to recent developments in different domains such as CFD calculation and fatigue modelling. However, in the MYRRHA reactor project, the coolant used will be the Lead Bismuth Eutectic (LBE), i.e. a liquid-metal with a high thermal conductivity. Such conditions are broadly different compared to the most conventional coolant that is water.The present paper presents an assessment of the consequences of thermal striping under the severe conditions of the MYRRHA reactor. The results of the study can be considered as satisfactory since the risk to structural integrity appears to be globally low. A number of aspects will however require further investigation as the reactor design progresses.

Stand-alone core sensitivity and uncertainty analysis of ALFRED from Monte Carlo simulations

The aim of this paper is the sensitivity and uncertainty analysis of a Lead-Cooled Fast Reactor (LFR) based on Monte Carlo simulation of sizes up to 2000. The methodology developed in this work considers the uncertainty of sensitivities and uncertainty of output variables due to a single-input-variable variation. The Advanced Lead fast Reactor European Demonstrator (ALFRED) is analyzed to determine the behavior of the essential parameters due to effects of mass flow and temperature of liquid lead. The ALFRED core mathematical model developed in this work is fully transient, which takes into account the heat transfer in an annular fuel pellet design, the thermo-fluid in the core, and the neutronic processes, which are modeled with point kinetic with feedback fuel temperature and expansion effects. The sensitivity evaluated in terms of the relative standard deviation (RSD) showed that for 10% change in the core inlet flow, the response in thermal power change is 0.58%, and for 2.5% change in the inlet lead temperature is 1.87%. The regression analysis with mass flow rate as the predictor variable showed statistically valid cubic correlations for neutron flux and linear relationship neutron flux as a function of the lead temperature. No statistically valid correlation was observed for the reactivity as a function of the mass flow rate and for the lead temperature. These correlations are useful for the study, analysis, and design of any LFR.

A study on different thermodynamic cycle schemes coupled with a high temperature gas-cooled reactor

With gradual increase in reactor outlet temperature, the efficient power conversion technology has become one of developing trends of (very) high temperature gas-cooled reactors (HTGRs). In this paper, different cycle power generation schemes for HTGRs were systematically studied. Physical and mathematical models were established for these three cycle schemes: closed Brayton cycle, simple combined cycle, and reheated combined cycle. The effects and mechanism of key parameters such as reactor core outlet temperature, reactor core inlet temperature and compression ratio on the features of these cycles were analyzed. Then, optimization results were given with engineering restrictive conditions, including pinch point temperature differences. Results revealed that within the temperature range of HTGRs (700–900°C), the reheated combined cycle had the highest efficiency, while the simple combined cycle had the lowest efficiency (900°C). The efficiencies of the closed Brayton cycle, simple combined cycle and reheated combined cycle are 49.5%, 46.6% and 50.1%, respectively. These results provide insights on the different schemes of these cycles, and reveal the effects of key parameters on performance of these cycles. It could be helpful to understand and develop a combined cycle coupled with a high temperature reactor in the future.