Density Reactivity Feedback Research Articles

Non linear stability analysis of supercritical water cooled reactor: Parallel channels configuration

The present work is devoted to the study of non-linear dynamics of the supercritical water reactor (SCWR) with a parallel channel configuration. To fulfil this objective, point reactor kinetics has been coupled with a parallel channel thermal-hydraulic model. The reduced ordered model is developed by coupling point kinetics equations with one group of delayed neutrons, fuel heat transfer and 1-D fluid flow model. The existing studies are commonly limited to linear stability analysis, therefore, present work focused on to demonstrate the nonlinear dynamics of the SCWR with parallel channel reactor core. The stability boundaries are shown in thermal hydraulic (pseudo phase-change and pseudo subcooling number) and intrinsic reactivity feedbacks (Doppler and density reactivity feedback) parametric spaces. Several bifurcation characteristics associated to density wave oscillations namely, subcritical, supercritical and generalized Hopf bifurcations are observed, which is also confirmed by calculating first Lyapunov coefficients. Corresponding to these phenomena, several numerical simulations are carried out at different parametric values, which confirms the existence and nature of the limit cycle behavior associated with these bifurcations. In addition, parametric sensitivity analysis on the stability characteristics are also carried out.

Thermal-hydraulic two-phase modeling of reactivity-initiated transients with CATHARE2 – Application to SPERT-IV simulation

Reactivity-Initiated Accidents (RIA) in nuclear reactor cores are very complex multiphysic transients. They consist in a very fast power excursion in the core, leading to wall temperature excursions. Many experiments have shown that the heat exchanges coefficients during fast transients differ from those in steady-states. The thermal-hydraulic phenomenology of such transients remains complex and some experimental studies allowed to better understand and quantify the heat exchanges between the wall and the fluid, for both single and two-phase flows.The current study lies in a continuous effort of developing and validating a coupled neutronic and thermal-hydraulic code for the analysis of protected and unprotected transient behavior of reactors. More specifically, this study deals with the extension of the thermal hydraulic model of the CATHARE2 code to fast transient configurations. The extended CATHARE2 model is validated against experimental results from SPERT-IV. This reactor was driven mainly by the coolant density reactivity feedback. Thanks to that, many flow regimes can be observed in the core during the transient. That allows a more accurate evaluation of the adequacy of available transient two-phase flow models and correlations. The results, that represent the state of modeling, allow concluding that introduced models in CATHARE2 are able to simulate RIA tests such as SPERT-IV involving fast transient boiling.

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.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

Simulations of SPERT-IV D12/15 transient experiments using the system code THERMO-T

Best-estimate codes, which couple neutronic and thermal-hydraulic solvers, are mainly used for safety analyses of nuclear power plants. During the past decade, the application of these codes to research reactors gained considerable interest and many improvements were presented to them. The increasing interest in the application of best-estimate codes to safety analyses of research reactor is largely driven by advancements in this field concerning power reactors and the diffusion of knowledge and capabilities to smaller, more diverse systems. The current study is a continuous effort in this framework and presents the coupled neutronic and thermal-hydraulic code development for the analysis of protected and unprotected transient behavior of research reactors. The coupling between neutronic and thermal-hydraulic processes is realized by considering the mutual feedbacks between them; the fuel and coolant properties (temperatures and density) variation affect the core's reactivity and hence the neutronic fission chain reaction, which in turn affects the fuel and coolant properties via a heat generation model for the reactor's power. More specifically, this study deals with the extension of the thermal-hydraulic model to the two-phase flow regime of the THERMO-T code. The extended THERMO-T model is validated against experimental results from the SPERT-IV, which was driven mainly by the coolant density reactivity feedback. This allows a more accurate evaluation of the adequacy of available and relevant two-phase flow models and correlations, which are selected from the domain of large power reactors. This is done in order to encourage and ensure standardization of modeling procedure of all types of reactors as part of the international community's continuous efforts towards this goal.

Nuclear coupled thermal hydraulic analysis of fast transient depressurization in supercritical channels

PurposeThe objective of the present work is to simulate the nuclear coupled thermal–hydraulic fast transient case studies for a vertically up-flowing supercritical pressure water channel of circular cross section. The emphasis is on analyzing the phenomenon of the deterioration in heat transfer (DHT) inside the channel subjected to sharp pressure variations.Design/methodology/approachThe thermal–hydraulic model, THRUST, is integrated with the neutron point kinetic (NPK) solver to account for the non-linear interactions between the thermal–hydraulic and neutronic temperature and density reactivity feedback effects. The model implemented and studied accounts for the time-dependent reactor power and is used to analyze various steady-state and flow-induced transient case studies (time-dependent and step change in exit pressure).FindingsThere is good agreement in the predicted behavior of the supercritical water pressure system with that of the available experimental data for the steady-state case. The event of DHT in the second transient case (step decrease in exit pressure) is found to be more severe than that of exponential pressure decrease.Originality/valueThis study evaluated a novel implementation of the thermal–hydraulic model, THRUST, integrated with NPKs applied to supercritical pressure water systems for predicting DHT.

Analysis of anticipated transient without scram of a Super Fast Reactor with single flow pass core

Anticipated transient without scram (ATWS) of the supercritical-pressure light water-cooled fast reactor (Super FR) with single flow pass core is analyzed to clarify its safety characteristics. Abnormal reactivity due to loss of feedwater heating and uncontrolled withdrawal of a control rod is mitigated by density and Doppler reactivity feedback. Increase of the coolant density at the upstream of the fuel channel leads the core flow rate to decrease which in turn decreases the power especially due to the density reactivity feedback. Meanwhile, ATWS accidents which are followed by trip of both RCPs are important. Decreased feedwater flow rate leads directly to a decrease of core flow rate in all fuel assemblies. If the ADS is not used, the MCST increases rapidly due to high power density feature of the Super FR. Coolant expansion effect does not appear in the single flow pass core of Super FR while it appears in the second pass of the two-flow pass core leading to mild increase of MCST of the second pass. However, signal of flow low level 3 (6%) is provided for the ADS actuation and the MCST of each fuel channel is effective to be mitigated due to large core coolant flow rate induced by the ADS operation. The power is also decreased by density and Doppler reactivity feedbacks. MCST and core pressure of all the analyzed events satisfy the criteria.

An Autonomous Reactivity Control system for improved fast reactor safety

The Autonomous Reactivity Control (ARC) system is a new safety device that can passively provide negative reactivity feedback in fast reactors that is sufficient to compensate for the positive coolant density reactivity feedback even in large low-leakage cores. The ARC system is actuated by the inherent physical property of thermal expansion, and has a very small effect on core neutronics at standard operating conditions. Additionally, the ARC system does not have an identified failure mode that can introduce positive reactivity in to the core. An ARC system can be installed in conventional fuel assemblies by replacing a limited number of fuel rods with rods that fill a safety function, providing negative reactivity to the core in the event of coolant temperature rise above nominal. These rods are of the same outer dimensions as the fuel rods, but contain smaller-diameter inner rods that are connected to liquid-filled reservoirs at the top and bottom of the assemblies. The reservoirs are filled with two separate liquids that stay liquid and immiscible throughout the applicable temperature range of fast reactor operation. The lower reservoir contains a “neutron poison” liquid with a high neutron absorption cross-section. The upper reservoir is filled with a separate liquid with a small neutron absorption cross-section. As the temperature in the assembly increases, the liquids in the reservoirs thermally expand, effectively pushing the absorbing liquid up toward the active core region while compressing the inert gas that fills the volume above the liquid between the inner and outer tubes of the ARC rods. The ARC system can be installed, or retrofitted in to existing systems, in every fuel assembly in the core. Since ARC installations in individual fuel assemblies operate independently, the system has a high level of redundancy. ARC-systems respond to local transients as well as core-wide accident scenarios. After actuation, the system automatically returns to its initial state as temperatures decrease, without the need for intervention by reactor operators. The ARC system concept and design considerations are described and illustrated.

Safety analysis of a supercritical-pressure water-cooled fast reactor under supercritical pressure

The supercritical-pressure water-cooled fast reactor (SWFR) is a fast spectrum supercritical water-cooled reactor (SCWR) studied by the University of Tokyo. The SWFR is designed as a two-pass core with an outlet temperature 500 °C. The SWFR has fuel channels cooled by downward flow, higher power density, and smaller coolant density reactivity feedback compared with Super LWR. This paper describes the safety analyses of abnormal events for the SWFR. SPRAT-F code is used for the safety analysis at supercritical pressure considering the downward flow cooled seed fuel channel. This code is based on a 1-D node junction model with point kinetics and decay heat calculations. Flow redistribution among parallel paths is calculated by pressure-loss balance and momentum conservation. The initiating events are selected from those of LWRs. For the safety analysis, nine abnormal transients and four accidents are selected with considering types of abnormality. By the numerical analyses, it was found that the loss of flow events can be mitigated by the “water source” effect of the downward flow blanket channels in the abnormal transients and accidents. All the abnormal events satisfy the criteria with margin.

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAlx–Al) containing 4.40gU/cm3 of an MTR was replaced with silicide (U3Si–Al and U3Si2–Al) and oxide (U3O8–Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38°C to 50°C and 100°C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U3O8–Al was about 2% more than the original UAlx–Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

Prospects of using different clad materials in a material test research reactor – Part 2 – The reactivity feedback coefficients

The reactivity feedback coefficients of a material test research reactor using stainless steel-316 and zircaloy-4 as clad were calculated. For this purpose, the aluminum clad of an MTR was replaced with stainless steel-316 and zircaloy-4. Calculations were carried out to find the fuel temperature reactivity feedback coefficient, clad temperature reactivity feedback coefficient, moderator temperature reactivity feedback coefficient and moderator density reactivity feedback coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 38 °C to 50 °C, at the beginning of life, were maximum in magnitude for stainless steel-316 cladded fuel, followed by aluminum and least for the zircaloy-4 cladded fuel. The fuel temperature feedback coefficient increased in magnitude by 47.37% for stainless steel-316 and decreased by 4.72% for zircaloy-4 clad. The moderator temperature feedback coefficient increased in magnitude by 60.41% for stainless steel-316 and decreased by 3.03% for zircaloy-4 clad, while the moderator density feedback coefficient showed an increase in magnitude of 59.18% for stainless steel-316 and a decrease of 7.63% for zircaloy-4 clad. Zircaloy-4 gave a positive value for clad temperature feedback coefficient, while the others two did not have any clad temperature feedback coefficient.

On the accuracy of reactor physics calculations for square HPLWR fuel assemblies

Although the supercritical-pressure or high-performance light water reactor (HPLWR) concept is largely based on the well-established technological experience available with conventional light water reactors, there is still no consensus on various key design features such as an optimal layout for the fuel assembly. This results mainly from the very large density variations of supercritical-pressure water in the core, which render it difficult to ensure reliable values for parameters such as power peaking factors and reactivity worths. The present paper describes studies carried out to compare deterministic and Monte Carlo codes for analysing a representative square HPLWR lattice with uniform 5%-enriched UO 2 fuel. The main purpose has been to assess the prediction accuracies achievable for integral parameters such as the multiplication factor, control absorber effectiveness, moderator/coolant density reactivity feedback and pin power distributions. The results show good agreement between the deterministic and stochastic calculations for the unperturbed lattice. However, for certain perturbed situations involving, for example, local coolant density changes in the assembly or control absorber insertion, the observed discrepancies are large enough to question the basic viability of the reactor physics design, e.g. with respect to the thermal performance.

ＰＷＲのスクラム不作動を伴う外部電源喪失過渡時熱水力挙動

The loss of offsite power transient without scram which is one of important ATWS events of a PWR was analyzed by the thermal-hydraulic transient analysis code RETRAN. Sensitivity studies on important parameters, the moderator density reactivity feedback coefficient and the heat transfer coefficient at SG secondary side, were included in the analysis. From the results, it has been found that DNB may occur in the core, saturation appears in the primary system since an early phase of transient, and natural circulation may be conditionally degraded, especially influenced by the heat transfer rate along uncovered part of SG U-tubes. These findings have not been reported in any of past analyses and therefore it is very necessary for further detailed investigations to be performed.