Core Thermal-hydraulic Analysis Research Articles

A subchannel analysis code SACOS-Na for sodium-cooled fast reactor

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm.For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes.

Parallelization and optimization for a pin-by-pin whole core thermal–hydraulic analysis code

With increasingly developing high-performance computing (HPC) technology, pin-by-pin thermal–hydraulic simulations are now feasible. The in-house subchannel analysis code KMC-FBc has been extended to perform pin-by-pin calculations. To expedite the whole-core thermal–hydraulic analysis process, various parallelization and optimization strategies are proposed and implemented at the single-assembly and multi-assembly levels, including acceleration methods for solving large sparse matrix equation systems, memory and file input/output (I/O) optimization, hybrid MPI/OpenMP-based whole-core acceleration method and so on. Parameters for evaluating the parallel performance, such as the running time, speedup, and parallel efficiency, are tested on an HPC cluster. Results show that the maximum acceleration ratio for the whole-core problem with a mesh size in the tens of millions can reach approximately 120, with a parallel efficiency of around 32%, which demonstrates the rationality and effectiveness of the parallel algorithm and optimization methods and lays the foundation for future high-resolution and high-fidelity thermal–hydraulic analysis.

Core flow distribution optimization of a natural circulation reactor using genetic algorithm, simulated annealing and characteristic statistic algorithm

Optimization of core flow distribution is a critical issue in nuclear engineering, particularly for natural circulation reactors with limited driving force. By appropriately arranging the resistance at fuel assembly inlets, core flow distribution could be well matched with core power distribution, thereby reducing the non-uniformity of core outlet temperature distribution. This, in turn, could enhance the performance and safety of the reactor. However, the flow distribution optimization throughout the fuel life cycle is a complicated global optimization problem that entails a vast solution space. In this paper, the flow distribution optimization with minimizing the maximum-to-minimum outlet temperature difference as the primary optimization objective, has been carried out incorporating various meta-heuristic algorithms with the core thermal-hydraulic analysis code COBRA, for a 200 MW natural circulation light water reactor. Minimizing the maximum-to-average outlet temperature difference is also chosen as an optimization objective since a smaller maximum outlet temperature is desirable. After the parametric sensitivity analysis of the genetic algorithm, the simulated annealing, and the characteristic algorithm, a set of parameters of each algorithm for the current flow distribution optimization problem has been determined. The results show that different algorithms yield similar resistance schemes under both optimization objectives, and all exhibit satisfying optimization effects, among which simulated annealing converges to the global optimum with significantly less computational costs.

CorTAF: A nuclear reactor core three-dimensional thermal-hydraulic characteristics analysis code based on OpenFOAM

In this paper, a thermal hydraulic analysis method based on open source CFD platform OpenFOAM was proposed. The models of coolant flow and heat transfer, fuel rod heat conduction and coupled heat transfer were established according to the bundle structure characteristics of PWR core, and then a nuclear reactor three-dimensional thermal–hydraulic characteristics analysis code CorTAF was developed based on finite volume method with the spatialresolution of subchannel level. The international benchmarks, including GE3 × 3, Weiss and PNL2 × 6 fuel assembly flow and heat transfer experiments, were selected to carry out the validation. The calculation results were basically consistent with the experimental data, illustrating that CorTAF is suitable for predicting the flow and heat transfer characteristics of coolant in the rod bundle fuel assembly. The thermal–hydraulic characteristics of fuel assembly and PWR core under full power operation and blockage accident conditions were simulated using CorTAF. The steady-state and transient spatial distribution of physical quantities in subchannel-scale including coolant and fuel rod temperature were obtained, and the influence of blockage condition on flow and heat transfer characteristics was analyzed. This work has reference significance for the development of core thermal hydraulic analysis tools for PWR.

Whole core thermal-hydraulic analysis considering inter-wrapper flow phenomena in the liquid metal cooled fast reactor

The inter-wrapper flow (IWF) phenomenon is identified as an integral part of the full core thermal-hydraulic simulation for liquid metal cooled fast reactors. In this work, a whole core thermal-hydraulic simulation code based on the subchannel analysis method, named KMC-FBc, is developed. Both the “two-step” and pin-by-pin level methodologies are investigated together with several inter-wrapper flow and heat transfer models, which are numerically solved simultaneously. To validate the code, it is compared with the experimental data from KALLA-IWF test and the numerical simulation results from computational fluid dynamics (CFD). Both normal conditions and flow blockage scenarios show that the pin-by-pin level method has better accuracy than the two-step method. An in-depth assessment of the influence of different IWF models on the medium-power lead-cooled fast reactor M2LFR-1000 is performed. Current results indicate that IWF heat transfer has little effect on the maximum temperatures and overall thermal-hydraulic characteristics under nominal conditions, while in the internal flow blockage scenario, IWF heat transfer will greatly reduce the maximum temperatures of fuel pellets and cladding, which can improve the accuracy of the whole core thermal-hydraulic analysis and provide margins for future reactor design and optimization.

Parallelization and application of SACOS for whole core thermal-hydraulic analysis

SACOS series of subchannel analysis codes have been developed by XJTU-NuTheL for many years and are being used for the thermal-hydraulic safety analysis of various reactor cores. To achieve fine whole core pin-level analysis, the input preprocessing and parallel capabilities of the code have been developed in this study. Preprocessing is suitable for modeling rectangular and hexagonal assemblies with less error-prone input; parallelization is established based on the domain decomposition method with the hybrid of MPI and OpenMP. For domain decomposition, a more flexible method has been proposed which can determine the appropriate task division of the core domain according to the number of processors of the server. By performing the calculation time evaluation for the several PWR assembly problems, the code parallelization has been successfully verified with different number of processors. Subsequent analysis results for rectangular- and hexagonal-assembly core imply that the code can be used to model and perform pin-level core safety analysis with acceptable computational efficiency.

Physical and thermal coupling calculation for accelerator driven subcritical core

The lead Anode (LBE) cooling ADS core is taken as the research object, and the ADS core axial one-dimensional physical thermal coupling model is established. The deterministic method is used to calculate the subcritical flux neutron flux distribution based on Monte Carlo. The method calculates the accelerator beam and the external neutron source generated by the spallation target. The ADS core thermal hydraulic analysis is carried out by a single-channel model, and the fuel elements are respectively divided into nodes in the axial direction and the radial direction, and the influence of thermal feedback on the neutronics characteristics of the subcritical core is considered. The one-dimensional physical thermal coupling model established in this paper can accurately calculate the spatial distribution of key parameters such as neutron flux density, fuel and cladding temperature in ADS core.

Calculation and experimental studies of coolant hydrodynamics in the inlet region of fuel assembly

An innovative core with an increased energy resource was used when designing RITM-200 reactor unit. The paper presents the results of experimental and numerical simulation of hydrodynamic processes occurring in the inlet region of the RITM reactor fuel assembly model. Computational mesh of fuel assembly, containing ∼ 22 million elements was created using Ansys ICEM CFD. The values of relative axial velocities in several cross sections at the inlet to the bundle of fuel elements are obtained. At the inlet to the fuel rods bundle the velocity field is not uniform, due to the complex geometry of the fuel assembly. The obtained results of CFD-simulation can be used to determine the input boundary conditions for subchannel programs of the core thermal-hydraulic analysis. This allows taking into account uneven flow rate distribution in subchannels due to the complex geometry of the fuel assembly inlet region.

Experimental studies and numerical simulation of coolant hydrodynamics in the inlet area of nuclear reactor fuel assembly

An innovative core with an increased energy resource was used when designing RITM-200 reactor unit. The paper presents the results of experimental and numerical simulation of hydrodynamic processes occurring in the inlet region of the RITM reactor fuel assembly model. Computational mesh of fuel assembly, containing ~ 22 million elements was created using Ansys ICEM CFD. The values of relative axial velocities in several cross sections at the inlet to the bundle of fuel elements are obtained. At the inlet to the fuel rods bundle the velocity field is not uniform, due to the complex geometry of the fuel assembly. The obtained results of CFD-simulation can be used to determine the input boundary conditions for subchannel programs of the core thermal-hydraulic analysis. This allows taking into account uneven flow rate distribution in subchannels due to the complex geometry of the fuel assembly inlet region.

Thermal-hydraulic design and analysis of a small modular molten salt reactor (MSR) with solid fuel

Summary A 20 MWth, 540 EFPD once through fuel cycle small modular molten salt reactor with solid fuel is proposed by Massachusetts Institute of Technology for off-grid applications. In this paper, various thermal-hydraulic analysis methods including computational fluid dynamics, Reactor Excursion Leak Analysis Program (RELAP5), and DAKOTA are adopted step-by-step for the reactor design based on the neutronic analysis results. First, 1/12th full core thermal hydraulic analysis is performed by using STAR CCM+ with most conservative considerations. Second, the transient safety behaviors of reactor system with risky assumptions are conducted by using REALP5. Finally, due to the unknown factors affecting reactor thermal-hydraulic characteristics, the uncertainty quantification and sensitivity analysis for the designed reactor is performed with DAKOTA code coupled with RELAP5. Numerical results show that a more uniform temperature distribution with reduced peak temperatures of fuel and coolant across the reactor core has been achieved. Enough safety margin is maintained even under most severe transient accident. The uncertainties in the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. All above results preliminarily indicate the feasibility of the current small modular molten salt reactor design and provide the further optimization direction from reactor thermal-hydraulic prospective.

Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP) VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD) codes and for detailed cellwise analyses of the VVER-100 reactor cores.

Fuel element and full core thermal–hydraulic analysis of the AHTR for the evaluation of the LOFC transient

The Advanced High Temperature Reactor (AHTR) is a fluoride-cooled and graphite-moderated reactor concept designed by Oak Ridge National Laboratory (Holcomb et al., 2011). The modeling and optimization of the heat removal system and the core structure is required, in order to obtain an adequate heavy metal loading and to provide effective cooling capability. The single channel MATLAB model provides a simple tool to evaluate the steady state conditions for the coolant and the fuel plate and the effects of the power distribution; sensitivity studies on the main design parameters of the fuel element are performed. A RELAP5-3D single channel model is developed for the validation and comparison with the MATLAB model; this model is the starting point for the development of a full core model, enabling the study of transients. A one-third fuel assembly model is then analyzed, consisting of six fuel plates and modeling the heat conduction of graphite through RELAP5-3D conduction enclosures. Since the assembly model is not suitable for the implementation in a full core model with the same level of detail, several simplifications have been evaluated, involving the modeling of the plate through a single heat structure and the modeling of different plates through a single plate. A SCALE model of the fuel assembly was developed for the evaluation of the reactivity feedback and the power distribution in the core. The results from the neutronic evaluations and the assembly model were implemented in a full core model, involving the core, the main reactor structures, the cooling system and the safety system (DRACS). The RELAP5-3D core model was used for the evaluation of the steady state conditions and the effects of a loss of forced cooling accident (LOFC).

Validation of a Monte Carlo based depletion methodology via High Flux Isotope Reactor HEU post-irradiation examination measurements

The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-I, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-I were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes. A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core. The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel.

Actinide Transmutation in PWRs Using CONFU Assemblies

Expansion of domestic use of nuclear power to provide energy security and environmental sustainability requires minimization of the nuclear waste. To achieve this goal in the short term, transmutation of transuranic (TRU) elements in COmbined Non-Fertile and UO2 (CONFU) Generation-III pressurized water reactor (PWR) assemblies is evaluated. These assemblies are composed of a mix of standard UO2 fuel pins and pins made of recycled TRU in an inert matrix and are designed to fit in currently deployed PWRs. Previous studies have shown the feasibility of a CONFU-Equilibrium (CONFU-E) assembly design with a net TRU balance between production and destruction and a CONFU-Burndown (CONFU-B) assembly design with net destruction of TRU coming from several reactors. In this paper, a CONFU-self-Contained (CONFU-C) assembly is shown to achieve net TRU destruction in a self-contained TRU multirecycling system. Both the CONFU-B and CONFU-C designs are presented in this paper in detail.For these designs a detailed assembly-level neutronic analysis has been performed using CASMO-4 to investigate cycle length, TRU management performance, and key reactor reactivity parameters, along with detailed intraassembly power peaking factors (IAPPFs). Various fuel mixing schemes and cooling times were evaluated. Using the IAPPF results, a full core thermal-hydraulic analysis using VIPRE was performed to validate thermal margins, and a loss-of-coolant-accident event was assessed using RELAP5. Based on the TRU management characteristics of these designs, metrics were developed to reflect the material handling difficulties of the multirecycled fuel, along with its repository impact. These parameters were compared to a standard once-through UO2 cycle, along with other Pu or TRU multirecycling schemes [mixed oxide with enriched uranium (MOX-UE) and COmbustible Recyclage A ILot (CORAIL)]. Finally, an economic analysis has been conducted to compare the fuel cycle cost (FCC) associated with these designs.TRU management results of CONFU-B and CONFU-C showed a net TRU destruction of 2 to 20 kg/TW·h(electric) generated, with an FCC of 12 to 15 mills/kW·h(electric), depending on the mixing strategy and cooling time chosen. Reactor control parameters and thermal margins were found to be comparable to an all-UO2 assembly. While both designs offer significant repository benefits, the accumulation of minor actinides may limit the practicality of fuel multirecycling.

Development of Statistical Thermal Design Procedure to Evaluate Engineering Uncertainty of Super LWR

The thermal performance of a nuclear reactor core contains various engineering uncertainties. These uncertainties often arise from calculation, measurement, instrumentation, manufacture, fabrication and data processing. Statistical techniques are useful to evaluate and combine these uncertainties in the thermal design of nuclear reactors. In this paper, a statistical method is developed and employed in the thermal design of the supercritical pressure light water reactor (Super LWR) to evaluate the statistical engineering uncertainties. This method is referred as the Monte Carlo Statistical Thermal Design Procedure for Super LWR (MCSTDP). This method uses the maximum cladding surface temperature (MCST) as a crucial criterion and a sub-channel code is utilized to perform the core thermal hydraulic analysis. The engineering uncertainties are considered with strict respect to the 95/95 limit of Super LWR. The main purpose of this paper is to establish the statistical evaluation methodology. The engineering uncertain for the thermal design of Super LWR is evaluated by using this method to get an approximate quantification. The results are compared with those of the Revised Thermal Design Procedure (RTDP) of Super LWR.

Reactor Subchannel Analysis–Electric Power Research Institute Perspective

One of the basic objectives of subchannel flow simulation and analysis effort sponsored by the Electric Power Research Institute was the development of a computer code for subchannel analysis and its verification and validation for applications to reactor thermal margin evaluation under steady and transient conditions. A historical perspective is given of the development of specifications for a reactor core subchannel thermal-hydraulics analysis code for utility applications in the evaluation of reactor safety limits during normal operation and accident scenarios. The subchannel analysis capabilities of the VIPRE-01 code based on the homogeneous equilibrium with the algebraic slip model of two-phase flow are presented. The code, which received a safety evaluation report from the US Nuclear Regulatory Commission in 1986, is in wide use in the utility industry for fuel reload safety analysis, critical heat flux correlation development and testing, thermal margin analysis, and core thermal-hydraulic analysis. A considerable amount of work has been done during the past few years on the development of VIPRE-02, an advanced subchannel analysis code based on the two-fluid model of two-phase flow capable of simulating reactor cores, vessels, and internal structures. The functional specifications, development of VIPRE-02, and current applications for VIPRE-02, such as boilingmore » water reactor mixed fuel core evaluation, are also discussed. Code is also used for PWR`s.« less

Pressurized water reactor thermal-hydraulic core analysis with the FLICA computer code

The FLICA computer code is devoted to steady state and transient thermal analysis of nuclear reactor cores. For this purpose, it is now widely used by different French companies working in the nuclear field. This paper deals essentially with the physical validation and applications of the FLICA-3M (Version 3.1) code for Pressurized Water Reactor Core thermal-hydraulic analysis. First, the FLICA-3M thermal-hydraulic modeling which is based upon subchannel analysis, is briefly described. The second section deals with the physical improvement of the model and the associated test loops. In the third section, we give results obtained for typical class II and III transients. Then, control rods ejection transient computations performed with the CRONOS-1 computer code, which results from the link between the FLICA computer code and a 3-D neutron kinetic model are given.

多目的高温ガス実験炉炉心のクロス流れ Ｉ ２ブロッククロス流れ実験

In course of the design development of the VHTR (very high-temperature gas-cooled reactor) core, special attention has been given to the crossflow through the interface gaps between fuel blocks because of its adverse effect on thermal hydraulic performance of the core.Experiments were carried out to determine crossflow rate on full-scale two-block models based on a proposed design of fuel element. Air at atmospheric pressure and ambient temperature was used as a fluid. Crossflow rate and pressure difference were measured for various gap configurations and widths. Measurements of the crossflow rate that entered into individual coolant channel were also made.The experimental results were expressed by the relation between crossflow loss coefficient factor and Reynolds number with interface gap as a parameter, allowing the prediction of crossflow in the core thermal hydraulic analysis.

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