Reactor Core Analysis Research Articles

A Next Generation Method for Light Water Reactor core analysis by using Global/Local Iteration Method with SP3

The Global/Local Iteration Method (GLIM) capability was implemented into the in-house developed nodal code, NuCoT, which was based on the Hybrid Nodal Green’s Function Method (HNGFM) with the diffusion/SP3 approximations. A procedure was proposed for obtaining the Even Parity Discontinuity Factor (EPDF) in SP3 method. The accuracy of NuCoT with GLIM was verified by using two benchmark problems, KAIST and C5G7, with reference solutions calculated by NEWT and MCNP respectively. The results of NuCoT show that the accuracy of the GLIM is comparable to the whole core pin-wise calculation. The SP3 calculation with EPDF is better than the diffusion result. For the KAIST problem using GLIM, the Root-Mean-Square (RMS) of pin power error is ∼1%, the maximum pin power error is ∼6%, and the keff error is <83 pcm for All-Rod-In case and <135 pcm for All-Rod-Out case. For the C5G7 problem, the RMS of pin power is <0.8% and the maximum error is <4%. The keff error is <80 pcm in 2D problem and <200 pcm in 3D rodded problem.

Reactor Core Design and Analysis for a Micronuclear Power Source

Underwater Vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micro nuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle (UUV) power system. In this paper, a 2.4MWt lithium heat pipe cooled reactor coreis designed for micro nuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by 6 control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback and core criticality safety. A code MCORE coupling MCNP5 and ORIGEN2 is used to analysis the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micro nuclear power source.

Analysis of a small-scale reactor core with PARCS/Serpent

Cross section homogenization is a challenging task that has been implemented mainly for full-core calculations of nuclear power plants. Small-scale reactors started to be a main point of interest for this kind of analysis during last few years. This paper presents a suitable homogenization strategy for a small core of VR-1 reactor (28.6 × 28.6 cm) realized by Serpent 2.1.28. Homogenized macroscopic data are used in PARCS_v3.2 core simulator for calculations of multiplication factor and radial flux distribution in the core. The PARCS results are compared to reference results calculated by Serpent 2.1.28. Reasons for observed inconsistencies between these two calculation modes are discussed.

Modeling of Resonance-Interference Effect in Depleted Fuel Compositions by Pseudo Resonant Isotopes

In order to develop an efficient resonance self-shielding method that can model the complex resonance-interference effects in depleted fuel compositions, an improved Pseudo Resonant Isotope Model (PRIM) has been developed by incorporating a number density–perturbation technique in the resonance cross-section tables for pseudo isotopes. Numerical results for homogeneous mixtures, pin cells, and pressurized water reactor lattice problems show that the new model is able to produce accurate group cross sections for a wide range of depletion states of different types of fuels, comparable to those obtained from online ultra-fine-group slowing-down calculations. Computational cost analysis shows that the improved PRIM is a promising method applicable to the resonance self-shielding calculations for large-scale reactor core analysis with depletion.

Coupled unstructured fine-mesh neutronics and thermal-hydraulics methodology using open software: A proof-of-concept

The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using free open source software is presented. The proposed contributions go in two different directions: one, is the focus on the open software approach development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code milonga. This concept was applied to model the behavior of a TRIGA-IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using the Serpent code. The results show that this coupled system gives consistent results, encouraging system further development and its use for complex geometries simulations.

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.

New stage on the neutronics and thermal hydraulics analysis of a small modular reactor core

Building on the success of the large nuclear plants, small modular reactors (SMRs) offer the potential to expand the use of clean, reliable nuclear energy to a broad range of customers and energy applications. In this work, a model to describe the neutronics parameters of a SMR core that can produce up to 530 MW of thermal power was developed. Using this model, several configurations of fuel enrichment to obtain the most homogeneous distributions of the power inside the fuel assemblies during all the core lifetime were studied. Temperature reactivity coefficients and mass variation of the principal isotopes for the optimised core were calculated. Finally, a model that allows the thermal hydraulic evaluation of the hottest section of the fuel elements was developed. The outlet coolant temperature, temperature distributions in the pins and the radial temperature distribution in the hottest pin were calculated.

New stage on the neutronics and thermal hydraulics analysis of a small modular reactor core

Building on the success of the large nuclear plants, small modular reactors (SMRs) offer the potential to expand the use of clean, reliable nuclear energy to a broad range of customers and energy applications. In this work, a model to describe the neutronics parameters of a SMR core that can produce up to 530 MW of thermal power was developed. Using this model, several configurations of fuel enrichment to obtain the most homogeneous distributions of the power inside the fuel assemblies during all the core lifetime were studied. Temperature reactivity coefficients and mass variation of the principal isotopes for the optimised core were calculated. Finally, a model that allows the thermal hydraulic evaluation of the hottest section of the fuel elements was developed. The outlet coolant temperature, temperature distributions in the pins and the radial temperature distribution in the hottest pin were calculated.

VHTR core analysis with McCARD and DeCART codes for high temperature engineering test reactor benchmark

This paper presents a Very High Temperature Gas-Cooled Reactor (VHTR) core analysis of the McCARD and DeCART-2D with High Temperature Engineering Test Reactor (HTTR) benchmark problem. The numerical results of the HTTR benchmark problem are demonstrated, the capability for VHTR core analysis of McCARD is validated by using the experiment data, and that of DeCART-2D is verified by comparing the numerical results with those of McCARD. The detail HTTR benchmark problem and specification of core geometry and material are also described. Also, a method of approximation for the complex geometry of the VHTR core is described for modelling in computational code. In terms of the multiplication factor and temperature coefficient, the numerical results of McCARD are easily compared with the HTTR experiment data. Because the DeCART-2D can solve 2-dimensional (2D) geometry, the simulation of 2D HTTR fuel rod, block, and core problem is performed with DeCART-2D and McCARD. The numerical results between them are around 200–400 pcm, meaning that DeCART-2D can be used to analyze VHTR core.

Application of CUPID for subchannel-scale thermal–hydraulic analysis of pressurized water reactor core under single-phase conditions

Abstract There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal–hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Abstract In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1X-A1, U3O8-Al, and U3Si2-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1X-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U3Si2-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

Recent developments in the GENESIS code based on the Legendre polynomial expansion of angular flux method

Abstract This paper describes recent development activities of the GENESIS code, which is a transport code for heterogeneous three-dimensional geometry, focusing on applications to reactor core analysis. For the treatment of anisotropic scattering, the concept of the simplified Pn method is introduced in order to reduce storage of flux moments. The accuracy of the present method is verified through a benchmark problem. Next, the iteration stability of the GENESIS code for the highly voided condition, which would appear in a severe accident (e.g., design extension) conditions, is discussed. The efficiencies of the coarse mesh finite difference and generalized coarse mesh rebalance acceleration methods are verified with various stabilization techniques. Use of the effective diffusion coefficient and the artificial grid diffusion coefficients are found to be effective to stabilize the acceleration calculation in highly voided conditions.

OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.

Three dimensional thermal hydraulic characteristic analysis of reactor core based on porous media method

Thermal-hydraulic performance in the reactor core was an essential factor in the nuclear power plant design. In this study, we analyzed the three-dimensional (3-D) thermal-hydraulic characteristic of reactor core based on porous media method. Firstly, a 3-D rector pressure vessel (RPV) model was built, including the inlet leg nozzle, downcomer, lower plenum, reactor core, upper plenum and outlet leg nozzle. Porous media model was used to simplify the reactor core and upper plenum. The commercial CFD code ANSYS CFX was employed to solve the governing equations and provide the 3-D local velocity, temperature and pressure field. After appropriate parameters and turbulent model being carefully selected, the simulation was validated against the 1:5 scaled steady-state hydraulic test. The predicted hydraulic parameters (normalized flowrate distribution and pressure drop) were in good agreement with the test results. And the predicted thermal parameters agreed well with the designed values. The validation indicated that this method was practicable in analyzing the 3-D thermal-hydraulic phenomena in the RPV. Finally, the thermal-hydraulic features in reactor core were analyzed under the condition of the Steam Generator Tube Rupture (SGTR) accident. The simulation results showed that the coolant temperature increased gradually from the center to the periphery in the reactor core in the accident. But the temperature decreased to safety level rapidly after the reactor shutdown and safety injection operation. The reactor core could keep in a safe state if appropriate safety operations were performed after accidents.

On the diffusion coefficient calculation in two-step light water reactor core analysis

ABSTRACTThis paper presents consistent and rigorous accuracy assessments of various methods for calculating the diffusion coefficients in a two-step reactor core analysis of light water reactors (LWRs). The diffusion coefficients are significantly affected by the transport correction and critical spectrum calculations. There are various methods for the transport corrections (inflow/outflow/hybrid corrections) and critical spectrum calculations (B1/P1/CASMO-4E methods) so that it is necessary to decide the best combination to achieve a high accuracy in the transport/diffusion two-step analysis. Numerical tests are performed step-by-step to search for the best combination of the methods by comparing each other the transport one-step results, transport/diffusion two-step results, and Monte Carlo results. Numerical test results with a large and a small LWR core show that the combination of inflow transport correction and CASMO-4E critical spectrum calculation is most accurate than the other combinations in terms of eigenvalues and assembly power distributions.

Development of an Application Programming Interface for Depletion Analysis (APIDA)

A new utility has been developed with extensive capabilities in identifying nuclide decay and transmutation characteristics, allowing for accurate and efficient tracking of the change in isotopic concentrations in nuclear reactor fuel over time when coupled with a transport solution method. This tool, named the Application Programming Interface for Depletion Analysis (APIDA), employs both a matrix exponential method and a linear chain method to solve for the end-of-time-step nuclide concentrations for all isotopes relevant to nuclear reactors. The Chebyshev Rational Approximation Method (CRAM) was utilized to deal with the ill-conditioned matrices generated during lattice depletion calculations, and a complex linear chain solver was developed to handle isotopes reduced from the burnup matrix due to either radioactive stability or a sufficiently low neutron-induced reaction cross section. The entire tool is housed in a robust but simple application programming interface (API). The development of this API allows other codes, particularly numerical neutron transport solvers, to incorporate APIDA as the burnup solver in a lattice depletion code or reactor core analysis code in memory, without the need to write or read from the hard disk. The APIDA code was benchmarked using several decay and transmutation chains. Burnup solutions produced by APIDA were shown to provide material concentrations comparable to the analytically solved Bateman equations – well below 0.01% relative error for even the most extreme cases using isotopes with vastly different decay constants. As a first order demonstration of the API, APIDA was coupled with the transport solver in the SERPENT code for a fuel pin cell depletion problem. Results show APIDA to be effective and efficient in solving lattice depletion problems, in addition to being successful in terms of portability for users to implement via the API.

A New Numerical algorithm for two-phase flows drift-flux model with staggered grid in porous media

FLICA4 is a 3D compressible code dedicated to reactor core analysis. FLICA4 solves for two-phase flows in porous media using a compressible drift-flux model [2]. To derive convective fluxes, FLICA4 uses a specific finite volume numerical method based on an extension of the Roe’s approximate Riemann colocated solver [3]. The method is developped to solve 1D flows. It requires adjustments to solve for 2D or 3D flows, at low Mach number, on a cartesian mesh. As Mach number goes to zero, in some extreme 2D or 3D cases, the native method does converge but the solution is not consistent with the analytical or experimental solution [4]. A so-called “pressure correction“ is added to the initial solver to derive an accurate solution. However, this extra term produces checkerboard oscillations in space for most applications of interest, even when solving for 1D flows. These checkerboard oscillations are sometimes critical and may lead to unstable solutions or divergence of the solver. The paper presents the analysis of an alternative numerical algorithm to solve for compressible drift-flux model in the low Mach regime. The current study aims at investigating a new compressible scheme that is both accurate and robust at low Mach number on staggered grids. The actual behavior of the new scheme is tested against simplified nuclear core ”boiling channel” study cases in the low Mach regime. The tests are extended to compressible regime for flows exhibiting both shock waves and no shock waves. The paper shows that the alternative numerical algorithm leads to numerical solutions that are very close to analytical solutions. Moreover, no more checkboard oscillations are generated. Therefore, the algorithm is a real improvement to FLICA4 current numerical method.

Analysis of the nuclear reactor core of Chernobyl Power Plant for 5 seconds before explosion with three-dimensional spherical space

This research analyzed the water flow and bubble (void) of nuclear reactor core of Chernobyl Power Plant, for 5 seconds before the explosion, using a mathematical model of two-dimensional spherical coordinates. To solve the problem, we considered this phenomenon as an analogy of Newtonian gravity theory, which had been solved in Schwarzchild Solution. As a result, the calculated radius of the spherical space of water and bubble indicated the maximum limit, at which the reactor core lost the control of reactor power. And, then, a regression analysis, with exponential model, confirmed the influence of bubble and water to the reactor’s power.

1-D dynamic stability analysis of mixed-spectrum supercritical water reactor core

The supercritical water reactor (SCWR) is one of the most prominent Generation IV reactors due to its high efficiencies. However, the stability issues, which are mainly caused by the great changes of thermodynamic properties and transport properties of supercritical water near the pseudo-critical temperature, are a challenge to the system safety and must be studied carefully. This paper is focused on 1-D dynamic stability analysis of mixed-spectrum SCWR (SCWR-M) reactor core. To this end, a frequency-domain model has been developed for linear stability analysis, and marginal stability boundaries under both the fixed inlet flow boundary conditions and the fixed external pressures boundary conditions are generated, which indicate that the system normal operational condition is in stable regions. Parametric sensitivity studies in frequency domain have been carried out. Increasing the wall thermal conductivity and mass flows can increase system stability. The system is more stable if the thermal zone has a lower power fraction. System with the designed non-uniform axial power distribution is also more stable than with the uniform distribution. A time-domain model has also been developed for nonlinear analysis, and the system marginal stability boundaries calculated by this method is consistent with those by frequency domain method. The existence of transitional stable region has been observed. A more reasonable definition for system logarithmic decay ratio has been achieved. The SCWR-M core has a subcritical bifurcation characteristic under fixed external pressures boundary conditions, thus its dynamic behaviors are not only related to systematic parameters, but also to the amplitudes of perturbations.