Neutronics-thermal Hydraulics Research Articles

Anderson acceleration for PWR whole-core pin-by-pin Nu-TH coupling calculation in NECP-Bamboo2.0

For PWR whole-core pin-by-pin neutronics/thermal-hydraulics (Nu-TH) coupling analysis, the Picard iteration strategy is widely employed benefitting from the convenience of implementation. Nevertheless, it is limited to the global multi-physics convergence rate regarding its oscillatory convergence behaviors in the local solution of each single-physics field. In this paper, the Anderson Acceleration (AA) scheme is employed and investigated to enhance the convergence performance of the pin-by-pin Nu and subchannel TH coupling analysis code NECP-Bamboo2.0. Different from the classical relaxation scheme where the relaxation factor is predetermined and fixed, the AA scheme utilizes the linear combination of several previous solutions to correct the current solution. The coefficients of this linear combination can be obtained by solving the residual equation, making the approach more robust. In addition, the selection of the accelerated solution vector is investigated based on the tightly coupling strategy of Nu-TH in NECP-Bamboo2.0. It is illustrated that the neutron-transport solution vector could have superior theoretical acceleration performance compared to the commonly used TH solution vector in the practical PWR analysis scenario. Finally, the AA scheme in NECP-Bamboo2.0 is validated using the VERA#6 3D single-assembly transport/TH coupling benchmark problem and the VERA#7 3D whole-core multi-physics coupling benchmark. It is demonstrated by the numerical results that the AA scheme possesses good robustness and computational efficiency. For the 3D whole-core VERA#7 benchmark, compared with the classic Picard iteration with a relaxation factor of 0.5, the AA scheme based on the neutron-transport solution vector can reduce the number of global iterations and calculation time by around 80%.

Neutronics and thermal-hydraulics coupling analysis for a small lead-based fast reactor based on the discrete ordinate nodal and parallel channel method

The full-core neutronics/thermal-hydraulics (NE/TH) coupling analysis for a small lead-based fast reactor (SLFR) was presented. The NE-to-TH coupling occurs via power distribution, whereas the TH-to-NE coupling occurs via the material cross section, which is related to the temperature distribution. In the NE module, a discrete ordinate (SN) nodal solver is implemented based on the hexagonal prism mesh. In the TH module, a parallel channel model is implemented with each hexagonal fuel assembly as a channel. The mass flow of each channel is searched according to the same pressure drop, the temperature and density of the coolant within each channel is calculated with a one-dimensional axial heat convection model, and the temperature distribution along the fuel pin is calculated with a one-dimensional radial heat conduction model. In order to update the cross section, a temperature-dependent homogenized microscopic cross-section library is prepared in advance, and based on the temperature distribution obtained from the TH calculation, the cross-section data required by the NE calculation are obtained by interpolation. The results of the NE/TH coupling analysis for SLFR indicate that the TH parameters (the maximum temperature of the fuel, cladding, and coolant and the maximum velocity of the coolant) are all within the design limits, and the TH feedback provides negative reactivity feedback (about −200 pcm) and has a fairly small effect on the power distribution (<1.0%).

LbmNTH: A unified lattice Boltzmann framework for coupled neutronics-thermal-hydraulics analysis

This work establishes the unified lattice Boltzmann (LB) framework, namely lbmNTH, to simulate the fully coupled neutronics-thermal-hydraulics (NTH) process in nuclear reactor. The neutron transport and delayed neutron precursor balance are simulated using the finite-Boltzmann schemed LB models with considering the temperature feedback. The fluid flow is simulated using the large-eddy-based LB model, and the convection heat transfer with fission generation also is considered under the LB framework. All the LB models have the same data structure and implementation, which simplifies the data exchange between different physical fields, and no extra data interpolation is required. Numerical results show that the proposed lbmNTH framework can simulate the coupled NTH process flexibly. The temperature feedback has a strong effect in high-temperature conditions, while the high flow velocity can effectively improve the heat transfer capability. This work can provide an optional multi-physics technique to analyze the coupled neutronics-thermal-hydraulics process in nuclear reactor engineering.

Investigation of Anderson acceleration in neutronics-thermal hydraulics coupled direct whole core calculation

The Anderson Acceleration (AA) scheme is applied and investigated in neutronics and thermal-hydraulics (T/H) coupled direct whole core calculation as an effort to improve the convergence of the alternating calculation of the neutronics and T/H fields which are nonlinearly coupled with each other. This is an approach to overcome the limited convergence rate of the conventional Gauss-Seidel type fixed-point iteration scheme which might reveal oscillatory convergence behaviors in local solutions. The scheme is to represent the solution at a fixed-point iteration as a linear combination of the previous solutions. The AA solution scheme is implemented in the direct whole core calculation code nTRACER in which both a simple closed channel T/H module and the pin-level drift-flux based T/H solution module ESCOT are embedded. A series of coupled 3-dimensional problems are solved with increasing complexity starting from a single assembly steady-state problem to a full core depletion problem via checker-board problems. The convergence behavior is examined in terms of the true error reduction by comparing the intermediate fission source distribution with the fully converged reference solution obtained applying a very tight convergence criterion. It turns out that the number of neutronics-T/H iterations is reduced considerably because the local oscillatory behaviors can be smoothed and convergence is reached earlier so that the computing time of the coupled calculations is reduced by about 25% retaining the solution accuracy.

Analysis of unprotected loss of heat sink accident in compact high temperature reactor cooled under natural circulation of LBE coolant

India is developing 100kWth Compact High Temperature Reactor (CHTR) to demonstrate the technology enabling high-temperature process heat applications such as hydrogen production. The reactor core is envisaged with 233U-Th based TRISO coated fuel particles and beryllium oxide moderator in prismatic fuel assemblies for high-temperature operation. The core heat is designed to be removed by lead–bismuth eutectic (LBE) coolant under natural circulation in vertical channels of CHTR at nominal power. The heat generated in the primary circuit is to be transferred to the secondary side using high-temperature sodium heat pipes. The design and safety of the prototype core necessitate comprehensive, integrated neutronics-thermal hydraulics analyses of anticipated transients. In this context, the Unprotected Loss of Heat Sink Accident (ULOHSA) at full power during the beginning of the core life cycle is investigated with indigenous 3D multi-physics code ARCH-TH. The key parameters defining neutronics and thermal-hydraulics of the core have been assessed in the steady-state as well as during the transient. The findings, along with the model of the simulation and the code validation are discussed in this paper. The analyses indicate that the core is being shut down passively due to reactivity feedbacks and the loss of heat sink leads to reverse flow of the coolant in the low powered channels during the transient. However, the coolant and fuel peak temperatures are observed to stay well below the leak-tightness limit of TRISO particles and boiling point of LBE, even considering the failure of shutdown systems. The modifications in the design of CHTR are also investigated to pacify the condition of flow reversal during such transients.

Low-order multiphysics coupling techniques for nuclear reactor applications

The accurate modeling and simulation of nuclear reactor designs depends greatly on the ability to couple differing sets of physics together. Current coupling techniques most often use a fixed-point, or Picard, iteration scheme in which each set of physics is solved separately, and the resulting solutions are passed between each solver. In the work presented here, two different coupling techniques are investigated: a Jacobian-Free Newton-Krylov (JFNK) approach and a new methodology called Coarse Mesh Finite Difference Coupling (CMFD-Coupling). In this work, both techniques were applied to the low-order CMFD system of equations. This allows for the multiphysics feedback effects to be captured on the low-order system without having to perform a neutron transport solve.The JFNK and CMFD-Coupling approaches were implemented in the MPACT neutron transport code, which is being developed for the Consortium for Advanced Simulation of Light Water Reactors (CASL). These methods were tested on a wide range of practical reactor physics problems, from a single 3D fuel pin to a massively parallel 3D full core problem. When coupled neutronics-thermal hydraulics problems were investigated with both JFNK and CMFD-Coupling, it was concluded that CMFD-Coupling outperformed JFNK in terms of both accuracy and runtime for every problem. When applied to large full core problems with multiple sources of strong feedback enabled, CMFD-Coupling reduced the overall number of transport sweeps required for convergence.

3D space-time analysis of ATWS in compact high temperature reactor with integrated thermal-hydraulic model in code ARCH-TH

The Compact High Temperature Reactor (CHTR) is being designed as a technology demonstrator for comprehensive Indian high temperature reactor programme for hydrogen production and similar process heat applications. The 100kWth CHTR core consists of 233U-Th based TRISO coated fuel particles, BeO moderator and LBE coolant with natural circulation in vertical prismatic fuel assemblies. The CHTR being a new conceptual design, necessitates comprehensive integrated Neutronics/Thermal-Hydraulics (N-TH) analyses and study. For design and safety studies, 3D space-time analyses of anticipated transient without scram (ATWS) in CHTR have been carried out in detail with indigenous code system with various cases of temperature feedbacks. One of the striking features observed in the analysis is that in case of weak Doppler feedback due to high fissile content and high temperature core conditions in CHTR, the BeO moderator plays a crucial role to limit the rise in nuclear power as well as peak fuel and coolant temperatures during such transient. For these studies, 1D-radial heat conduction in multi-channel based TH module has been developed in 3D space-time code ARCH. The viability of neutronics and adiabatic Doppler feedback capability of the code has also been examined with AER benchmark problems (AER-DYN01 & 02) and results are discussed. The analysis shows that the transient peak fuel and coolant temperatures are limiting at values much below the fuel safety criteria of TRISO particle (∼1600°C) and boiling point of LBE coolant (1670°C) even with scram failure. The significance of temperature feedback effects of BeO moderator in CHTR is seems to be a first of its kind observation and is not reported in the literature.

Sensitivity analysis of feedback factors by neutronics and thermal-hydraulics coupling code based on FLUENT software

With the great improvement of computer performance, analyzing the complex flow and heat transfer phenomena under transient condition by coupling CFD and neutronics has attracted lots of attention nowadays. We developed a neutronics-thermal hydraulics coupled code for transient analysis of pool type lead cooled fast reactors based on FLUENT UDF. The coupled code FLUENT/PK was validated by critical and sub-critical reactor under unprotected transient condition. The validation results show the correctness and feasibility of CFD method in safety analysis of reactors. The coupled code was used to analyze the lead cooled fast reactor SNCLFR-100 (which had been proposed by USTC) under different transient conditions. The calculated results show that there exists great difference if the introducing time of reactivity is different. The sensitivities of the feedback factors were analysed for the reason that these factors will be quite different at different operation time of the reactor core, and the result shows that Doppler constant has the most important effect on reactor safety compared with coolant temperature coefficient, fuel rod axial expansion coefficient and cladding expansion coefficient.

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Test case for VVER-1000 complex modeling using MCU and ATHLET

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

Neutronics and sub-channel thermal-hydraulics analysis of the Iranian VVER-1000 fuel bundle

In the present research, sub-channel thermal-hydraulic analysis of the Bushehr Nuclear Power Plant (BNPP) core, a Russian VVER-1000 type, at hot full power (HFP) and steady state BOC conditions is presented herein using modified COBRA-EN code. Required power distribution and hot channel factors were computed by our past neutronics calculations using MCNP-5 code. Due to increasing temperature, modification on the continuous-energy cross sections in the MCNP libraries are carried out using NJOY code. Maximum and average fuel temperature, enthalpy, void fraction, coolant temperature and density, coolant mass flow rate and pressure drop are calculated using different models. Thermal-hydraulics calculations of the most rated channel (hottest sub-channel), which is determined based on coupled neutronics-thermal hydraulics calculations are investigated and results in temperature, enthalpy, critical heat flux and MDNBR of the hottest sub-channel are found. Finally, our results are compared with analytical approaches and the reactor FSAR.

Fuzzy-like PD controller for spatial control of advanced heavy water reactor

Spatial oscillations in the neutron flux distribution due to xenon reactivity feedback requires stringent control in large nuclear reactors, like advanced heavy water reactor (AHWR). If the spatial oscillations in the power distribution are not controlled, power density and rate of change of power at some locations in the reactor core may exceed limits of fuel failure due to ‘flux tilting’. Further, situations such as on-line refueling might cause transient variations in flux-shape from the nominal flux-shape. For analysis and control of spatial oscillations in AHWR, it is necessary to design a suitable control strategy, which will stabilize these oscillations. In this paper, a simplified scheme to design a conventional fuzzy logic controller for spatial control of AHWR is presented. This scheme known as fuzzy-like proportional derivative (FZ-PD) controller, uses robust PD (proportional derivative) type rule base. Due to robust rule base structure, tuning of scaling factors is greatly reduced. The non-linear coupled core neutronics-thermal hydraulics model of AHWR considered here represented by 90 first order differential equations. Through the dynamic simulations, it is observed that the designed FZ-PD controller is able to suppress spatial oscillations developed in AHWR and its performance is found to be robust.

The axial power distribution validation of the SCWR fuel assembly with coupled neutronics-thermal hydraulics method

SCWR (super critical water reactor) is one of the IV generation nuclear reactors in the world. In a typical SCWR the water enters the reactor from the cold leg with a temperature of 280°C and then leaves the core with a temperature of 500°C. Due to the sharp change in temperature, there is a huge density change of the water along the axial direction of the fuel assembly (FA), which will affect the moderating power of the water. So the axial power distribution of the SCWR FA could be different from the traditional PWR FA.In this paper, it is the first time that the thermal hydraulics code CFX and neutronics code MCNP are used to analyze the axial power distribution of the SCWR FA. First, the factors in the coupled method which could affect the result are analyzed such as the initialization value or the partition method especially in the MCNP code. Then the axial power distribution of the Europe HPLWR FA is obtained by the coupled method with the two codes and the result is compared with that obtained by Waata and Reiss. There is a good agreement among the three kinds of results. At last, this method is used to calculate the axial power distribution of the Chinese SCWR (CSR1000) FA. It is found the axial power profile of the CSR1000 FA is not so sensitive to the change of the moderator density.

Application of Time-Dependent Neutron Transport Theory to High-Temperature Reactors of Pebble Bed Type

We introduce a new coupled neutronics/thermal-hydraulics code system for analyzing transients of high-temperature gas-cooled reactors (HTGRs), based on a neutron transport theory approach. At the heart of the coupled code system resides the DORT-TD code, a time-dependent extension of the well-known DORT discrete ordinates code. DORT-TD uses a fully implicit time integration scheme and is coupled via its generalized thermal-hydraulics interface to the THERMIX-DIREKT code, an HTGR-specific heat conduction/convection code for pebble bed-type reactor cores. Feedback is accounted for by interpolating multigroup cross sections from libraries pregenerated with appropriate spectral codes. These libraries are structured for user-specified discrete sets of thermal-hydraulic parameters, e.g., fuel and moderator temperatures. The coupled code system is applied to a pebble bed HTGR model case, i.e., the PBMR 268 MW design. Steady-state studies are performed, and several design-basis and beyond-design-basis transients are simulated in an effort to assess the adequacy of using neutron diffusion theory against the more accurate but yet computationally more expensive neutron transport approach. Relatively small but significant differences arise from using either theoretical approach, from which it is concluded that transport theory as the more versatile tool should be used as reference to quantify the effects of the approximations inherent in diffusion and to gain confidence in its predictive power, especially in safety analyses. In an effort to validate the DORT-TD/THERMIX code system, the neutronics stand-alone solver is benchmarked against available transient benchmark exercises, and the coupled code system is applied to the Organisation for Economic Co-operation and Development/Nuclear Energy Agency/Nuclear Science Committee PBMR 400 MW Coupled Neutronics Thermal Hydraulics Transient Benchmark, demonstrating its remarkable viability for a wide range of safety cases. The final product is a high-fidelity, highly flexible, and well-validated state-of-the-art computer code system, with multiple capabilities to analyze HTGR safety-related transients in an accurate and efficient manner.

A neutronics–thermalhydraulics model for preliminary studies on TRADE dynamics

In the last few years the possible role of accelerator driver systems (ADS) for effective transmutation strategies with fully closed cycles has received increased attention due to their potential to improve the flexibility and safety characteristics of transmutation systems. The substantial difference between the neutron kinetics and dynamic behavior of ADS and conventional critical reactors has given rise to a wide international consensus on the need of an experimental program to improve their knowledge and to validate calculation methods. To this end the international cooperation TRADE proposed a sub-critical experiment based on the coupling of a TRIGA reactor in sub-critical core configuration with a proton accelerator (cyclotron) by means of a neutron spallation target. The experiment was initially conceived in the RC1-TRIGA reactor located at the ENEA Center of CASACCIA (Rome, Italy) to demonstrate the feasibility of the accelerator driven system (ADS) concept at a representative power. This article presents a preliminary study performed with the RELAP5/PARCS code on the dynamic behavior of such a system in order to demonstrate the code capability to support the design of the experiment and the safety analysis. The specific code version used joins the well known capability of RELAP5 to treat light water reactors with the potentiality of PARCS modified by ENEA to simulate the three-dimensional neutronics of sub-critical systems, i.e. to treat external neutron sources. PARCS modifications are preliminary assessed against a simple analytical solution of the sub-critical neutronics of the experiment based on the kinetics pseudo-potentials method. A quite detailed model for the coupled code is developed in order to realistically evaluate both the thermal feedback effects, the control rod action and the external source strength changes. A wide range of operational and accidental transients of the sub-critical reactor are simulated with the coupled model in order to obtain a first system response to a number of reactor elementary events at different subcriticality levels. The calculation results show a high qualitative agreement with the sub-critical system physical theory underlining how the numerical model developed could be a useful tool for the definition of the operational procedures and the investigation of accidental conditions; moreover the accidental transient trends highlight the inherent safety behavior of the TRIGA research reactors that makes them extremely suitable for the coupling of the different components with a quite simple licensing procedures.

TRAC-BF1/NEM Modeling and results of OECD/NEA BWR core transient benchmarks

TRAC-BF1/NEM is a coupled neutronics-thermal hydraulics code for BWR transient and accident analysis developed at the Penn State University. Research has been carried out at Penn State to further verify this coupled code by using the (OECD/NEA) 3-D BWR core transient benchmarks. These benchmarks provide an opportunity to compare TRAC-BF1/NEM predictions with the results of many other coupled codes as well as to establish a well-defined set of test problems for future TRAC-BF1/NEM modifications and code distributions. The paper gives a description of the developed neutronics and thermal-hydraulic models and presents a comparison and discussion of the obtained steady-state results as well as the transient results for the inlet sub-cooling transient.