Search

• SACOS-PLATE: Enhanced code for plate fuel analysis. • Validated against RA-6, Bettis experiments. • Unique heat conduction, flow models. Over the past few decades, plate-type fuel elements have been widely used in advanced research reactors, small modular pressurized water reactors, production reactors, etc., with the advantages of compact structure arrangement, good heat transfer characteristics and high fuel consumption. Currently, a thermal–hydraulic subchannel analysis code, namely SACOS-PLATE code, was newly designed to satisfy the calculation of different types of plate fuel elements under various steady-state operating conditions. Based on the original SACOS code, this updated version modifies the flow distribution model, the plate heat conduction model, the heat distribution model as well as the flow heat transfer model, and then realizes the simulation of the special thermal–hydraulic phenomena of the fuel plate and the rectangular flow channel. In this present study, the RA-6 and Bettis experimental data was used to validate the code, and the simulation calculation was performed on the JRR-3 research reactor, which effectively proved the good computing capability of SACOS-PLATE code. [ABSTRACT FROM AUTHOR]

The condensation enhancement research of pure steam and steam with non-condensable gas is of great significance for reactor safety. In this study, to evaluate the influence of chrome-plated tubes on condensation enhancement, pure steam and steam with air condensation heat transfer experiments are conducted, and the plating thicknesses of 1 μm (tube #1) and 10 μm (tube #2) are considered. The results show that dropwise condensation (DWC) occurred in the part of tube #1 surface, while tube #2 surface is filmwise condensation (FWC). For the pure steam condition, the steam condensation heat transfer coefficient (HTC) of tube #2 is higher than that of tube #1, and both are greater than the theoretical analysis of Nusselt. Thus, the condensation enhancement effect of the chrome-plated tubes is remarkable. For the steam with air condition, a new condensation HTC expression is proposed, which is related to gas pressure, wall subcooling, and air mass fraction. Compared with the widely used experimental correlations and the experimental data of polishing tubes, it is found that the condensation enhancement performance is not significant. This is because the dominant thermal resistance affecting heat transfer is the air layer at the gas-liquid interface. Thus, enhancing the disturbance of the gas layer is the key to the condensation enhancement of steam with non-condensable gas. [ABSTRACT FROM AUTHOR]

• The leak rate prediction models of LBB were established for SFR. • The leak rate prediction program (LAPIS) for sodium pipelines was developed. • The effects of crack morphology parameters on leak rate were studied. The Leak Before Break (LBB) technology plays a vital role in the safety evaluation of Sodium-cooled Fast Reactor (SFR). In this study, the leak rate considered in LBB assessment of SFR was predicted and analyzed. Based on the models of fluid resistance in cracks, two-dimensional numerical simulation for leakage process of uniform cross-section axial cracks were conducted with Computational Fluid Dynamics (CFD) to obtain recommended values of crack empirical parameters, including entrance resistance coefficient, effective parameters and friction factors. With consideration of the complex crack morphologies, the leak rate prediction models with a wide range of parameters were established for liquid sodium, and corresponding computation code (LAPIS) was developed. The proposed models were verified and validated by seepage model, numerical results and experimental data with deviations within 1%, 35% and 10%, respectively. The effects of the crack morphology and thermal-hydraulic parameters on the leak rate for tight and wide cracks were studied with the LAPIS. The leak rate increases with Crack Opening Displacement (COD), crack length, ratio of entrance to exit area, entrance resistance coefficient, stagnation pressure and decreases with temperature. This work could provide theoretical and technical support for the study of leak rate considered in LBB assessment of SFR. [ABSTRACT FROM AUTHOR]

• The additional forces caused by motion are considered in the instability model. • The natural circulation instability will be slightly extended affected by motion. • Inclination and rolling tend to destabilize the parallel channels system. • The loop flow fluctuation does not contribute much to the parallel instability. In this study, the mathematical models suitable for natural circulation, parallel channels, and general forms of additional force under moving conditions were established. An analysis code was developed using SIMPLE algorithm. Based on the validated code, the natural circulation and parallel flow instability under typical moving conditions such as inclination, rolling, heaving, and coupled motion were discussed, and the corresponding flow instability boundaries were obtained. The results show that under the moving conditions, the natural circulation flow change is manifested as the superposition of the two-phase instability fluctuation and the motion-induced fluctuation. The heaving condition has the most drastic effect on flow rate with fluctuation amplitude exceeding 30%, but the average flow rate remains unchanged. The boundary of natural circulation flow instability will be slightly extended affected by motions and the variation of the boundary quality is less than 0.03. The inclination and rolling tend to destabilize the parallel channels system and reduce the boundary quality by more than 0.1, whereas the effect of heaving on the parallel instability is not significant. For the parallel flow instability, the adverse effect is mainly attributed to the decrease of the mean flow rate at the entrance of the parallel channels caused by the motions, while the flow re-distribution between parallel channels and natural circulation flow fluctuation induced by the moving conditions do not contribute much. [ABSTRACT FROM AUTHOR]

The SACOS code is thermal-hydraulic subchannel analysis code designed to meet steady-state conditions and operational transients in different types of reactor cores. It uses a homogeneous flow or drift-flux model to simulate single-phase and two-phase flow characteristics. In order to achieve a more refined two-phase flow modeling for LWR cores, a new eight-equation two-fluid based analysis module of SACOS is developed in this study. The phasic equations for mass continuity, momentum, and energy are spatially discretized by the finite volume method (FVM) with a staggered grid structure. To reduce numerical diffusion and improve numerical accuracy, the high-resolution scheme with nonlinear flux limiter is implemented. The code assessment is performed with linear advection flow problem and the preliminary code validation is presented by comparing with the experimental results of Weiss 14 × 14 blockage test, CNEN 4 × 4 mixing test, Studsvik 3 × 3 mixing test, and ISPRA 4 × 4 mixing test covering single/two-phase and unheated/heated conditions. The application of SACOS implies that the code can be used to model and perform assembly-level core safety analysis with visualization capabilities for hexagonal and rectangular geometry. Development of a two-fluid based thermal-hydraulic subchannel analysis code and its application. [Display omitted] • A new eight-equation two-fluid based analysis module of SACOS is developed. • The high-resolution scheme with nonlinear flux limiter is implemented into code. • The validity of the higher-order scheme is verified with linear advection flow problem. • The two-fluid model is validated based on reported bundle experiments. [ABSTRACT FROM AUTHOR]

• The oxidation kinetics models of ATF claddings were incorporated into RELAP5. • Safety analysis of PWR loaded with ATFs was carried out during LOCA. • Optimization schemes of ECCS were proposed based on the performance of ATFs. Based on the excellent performance of accident-tolerant fuels (ATFs), the optimization schemes for the safety system of Chinese advanced three-loop pressurized water reactor during loss of coolant accident (LOCA) were proposed and corresponding feasibility analysis was carried out in this study. The modified system analysis code was used to safety evaluation and four ATF systems were investigated. Several modifications were made to the rod geometries in partial ATF systems to make up the poor neutron economic performance of FeCrAl and Mo MAX. The emergency core cooling system (ECCS) was optimized during LOCA with different break sizes. The results show that it is feasible for all ATF systems to cancel the high pressure safety injection (HPSI) or cancel both HPSI and low pressure safety injection (LPSI) but add an extra accumulator (ACC). Also the safety of reactors fueled with ATFs can be guaranteed with only 0.35 times the original flowrate of LPSI during large break LOCA (LB-LOCA). In the scheme of replacing the original ECCS with two extra ACCs during LB-LOCA, reactors loaded with partial ATFs are safe, while UO 2 -Zr systems are in danger as a contrast. This work could provide reference for the design of reactor safety system in the future. [ABSTRACT FROM AUTHOR]

As one of the Accident Tolerant Fuels (ATFs), Fully Ceramic Microencapsulated (FCM) presents advantageous safety performance in the severe accidents compared with UO 2 owing to its high melting point, superior heat transfer ability and proliferation resistance capability. In this study, preliminary neutronic and thermal-hydraulic analysis of Chinese improved three-loop Pressurized Water Reactor (PWR) with FCM pellet and SiC cladding (FCM-SiC) were conducted to demonstrate the fuel performance of ATFs. First, a three dimensional full core model loaded with FCM-SiC was developed to acquire the cycle length, power distribution and isotopic inventories. And low reactor power and high fuel enrichment were used to satisfy the annual reload requirement. Then, the reactor thermal-hydraulic safety evaluation was carried out by the modified subchannel and system analysis codes, which depicts lower Fuel Centerline Temperature (FCT) and Peak Cladding Temperature (PCT) in FCM-SiC reactor compared to the reactor with traditional fuels. The Minimum Departure from Nucleate Boiling Ratio (MDNBR) of cladding is much higher than the criterion of 1.3 as well. Furthermore, the reactor safety system was simplified and optimized by taking full advantage of the preponderance of ATFs, and the results show that the reactor of FCM-SiC can operate safely with only 10% initial flowrate of Low Pressure Safety Injection (LPSI) or without Accumulators (ACCs) during Large Break Loss of Coolant Accident (LBLOCA), while the conventional zircaloy is failed under the same conditions. This work could provide guidelines for the optimized design of the reactor safety system in the future. • A 3D full core model loaded with FCM-SiC was developed for neutronic analysis. • Thermal-hydraulic evaluation was conducted by modified subchannel analysis code. • Safety assessment during LBLOCA was carried out by modified system analysis code. • Reactor safety system was optimized with the help of ATFs' preponderance. [ABSTRACT FROM AUTHOR]

Critical heat flux (CHF) is one of the most important thermal criteria for nuclear reactor design. To obtain a more widely applicable CHF prediction method, many mechanistic models have been proposed based on single channels with a good precision. Since the annular flow may occur in rod bundle in gas-liquid two-phase region for boiling water reactors (BWRs), it is essential to apply the annular film dryout (AFD) mechanistic models to rod bundle geometry. Among many existing AFD mechanistic models that are available to review, two major categories can be distinguished with considerable attention: the three-fluid model and the film analysis model. Many studies show that it is feasible to couple AFD mechanistic model with sub-channel analysis using suitable coupling strategies, but some crucial issues have not been solved. In view of these facts, this paper initially highlights the basis of AFD mechanistic model and its application to rod bundle. Then some factors which have predominant influence on the accuracy of the mechanistic model are reviewed and analyzed, including cross flow, onset of annular flow, inception fraction for droplet entrainment, boiling and spacer grid effect, dryout criteria. Finally, conclusions and prospects in this research field are summarized and proposed. • This study carried out a comprehensive review on annular film dryout (AFD) mechanistic model. • A detailed discussion was applied on coupling the AFD mechanistic model with subchannel analysis for rod bundle geometry. • Some critical factors were presented in this study and subsequent research direction is evaluated. [ABSTRACT FROM AUTHOR]

• A three-field mechanistic model is developed for dryout prediction in annular flow. • The model is verified by experimental data of the circular and rectangular channel with relatively large parameter range. • Some key constitutive relations in the model are discussed for the better prediction. • The model is effective on prediction of dryout in the rod bundles by coupling with subchannel code. Critical heat flux (CHF) is one of the most important thermal criteria for nuclear power plants. It has traditionally been evaluated using look-up tables or empirical correlations. In this paper an annular film dryout (AFD) mechanistic model has been developed based on the interaction of three fields in annular flow region, i.e. the liquid film, entrained droplets and vapor core. The model describes the mass and momentum conservation equations of three fields together with a series of constitutive relations. The effect of some constitutive correlations (the entrainment and deposition of droplets and the onset of annular flow) on the prediction accuracy of the model is studied. The results are compared with CHF experimental data in flow boiling. Fairly good agreements are observed for the CHF in circular tubes with uniform and axially non-uniform heating, as well as rectangular channels with uniform heating. Through coupling with the subchannel analysis method, the model is used to predict the dryout-type CHF in the rod bundles with a good precision. [ABSTRACT FROM AUTHOR]

Summary: The fuel rods in the lead‐based fast reactor (LFR) are usually supported by wire spacer, and the presence of the wire results in the complexity of the thermal‐hydraulic analysis in rod bundles. In this study, several previous experimental research about lead‐bismuth eutectic flowing through the rod bundle are investigated. Based on the collected experimental data, a series of flow and heat transfer correlations in the wire‐fixed rod bundle are compared and evaluated. The recommended correlations are then implemented into the in‐house sub‐channel analysis code SACOS‐PB. Furthermore, to account for the changes in the channel cross‐section parameters caused by the wire spacer, two optional wire geometric treatments were added to the code: axially‐averaged approach and axially‐varied approach with extra cross‐flow enhancement model. 19‐rod bundle and 61‐rod bundle tests are used to validate the optimized code. The deviations for the temperature differences of cladding and coolant in internal sub‐channels are within 25%, and the axially‐varied approach shows a higher prediction accuracy, which proves that the improvement of the SACOS‐PB are reliable. This work could provide a reference for the subsequent design and development of the LFR core. [ABSTRACT FROM AUTHOR]

The helium-xenon(He–Xe) gas mixture with low Prandtl number and good heat transfer characteristics suits for the advanced high heat flux and compact gas-cooled reactors. A numerical simulation was conducted on the flow, heat transfer, and turbulent mixing characteristics of He–Xe gas mixture in a tightly arranged triangular rod bundle subchannel. The transition characteristics of He–Xe gas mixture in rod bundle structure were obtained. The friction resistance coefficient, Nusselt number(Nu), and turbulent mixing factor β was corrected within the Reynolds number(Re) range of 3 × 102 - 1.8 × 105. The research results indicate that the transition occurs at Re equal to 2000 and forms a transition zone within the Re range of 2000–5000; The Blasius correlation can describe the wholly turbulent flow precisely, but it is suitable to adopt a new formula in the transition zone; The local Nu prediction performance of the modified Peckett correlation is in good agreement with the numerical simulation results, and the Petukhov correlation with shape correction and temperature correction factors has good prediction results on global Nu ; The turbulent mixing factor using Re and shape parameters can better describe the mixing characteristics. This study has reference value for the design of rod bundle reactors using He–Xe gas mixtures as working fluids. [ABSTRACT FROM AUTHOR]