Reactor Physics Research Articles

The Design of Thermodynamic System Instrument & Control for the Daya Bay Turbine High Pressure Cylinder Retrofit

Before the retrofit, the equipments of the turbine high pressure cylinder of the Daya Bay Nuclear Power Plant is of the old technology. The design of the leaf shape and the steam runner is not advanced, and the equipments are aged after more than 20 years’ operation. From the point of increasing the power and improving the reliability of the equipments, it’s decided to retrofit the turbine high pressure cylinder of Daya Bay Nuclear Power Plant unit 1 and unit 2. The retrofit relates to the design and modification of many secondary loop thermodynamic systems’ instrument & control parameters, in order to match with the first loop reactor control systems, and also the mechanism and electric systems, it’s necessary to verify and optimize these designed parameters of the units after the retrofit. This design and analysis method and experience of the secondary loop thermodynamic systems’ instrument & control parameters can be referred by the similar retrofit of other nuclear power plants.

Reactor physics study of square and hexagonal lattice cell using MANTRA (Method of ChAracteristics based Neutron TRAnsport Code) coupled with resonance treated multi energy group cross section library

Since last few years we have been developing a computer code MANTRA (Method of ChAracteristics based Neutron TRAnsport Code) in phased manner to solve neutron transport equation using Method of Characteristics (MOC) for lattice (i.e. assembly) as well as full core with a flexibility of choosing boundary condition out of vacuum/periodic/reflective options appropriate for the simulation. In literature, we have already reported about the development and performance of the code in triangular mesh network for mainly lattice level calculation with reflective and periodic boundary conditions and to some extent core level calculation with vacuum boundary condition using the processed cross section of benchmark problem. Later, we have explored the advantage of linear source assumption over the commonly used flat source assumption within a mesh in some selected transport calculations (where the variation in neutron flux is substantial) using MANTRA. We have also demonstrated the application of Krylov subspace iteration in a matrix-free framework to accelerate the solution of MANTRA for solving complex problems in effective manner. In continuation of the work, MANTRA is now coupled with WIMSD multi energy group cross section library. For the purpose of coupling, a module is developed in the code to process the cross section data and apply resonance treatment to account for the self-shielding effect of the resonant nuclides. The resonance treatment requires estimation of Dancoff factor, which is carried out using MANTRA for various arrangement of fuel elements. Finally, a number of benchmark problems based on PHWR, BWR, VVER lattices are studied with resonance treated library to compare the performance of MANTRA with other internationally recognized codes like HELIOS, DRAGON, WIMSD, MCNP etc.

A multiphysics model of the versatile test reactor based on the MOOSE framework

The traditional modeling approach for sodium fast reactor cores relies on separate physics models, where the fuel performance, thermal–hydraulics, and neutronics calculations required to predict the core physics characteristics for nominal conditions are decoupled by relying on user-imposed boundary conditions. This paper aims at evaluating the impact of multiphysics simulations for predicting the core characteristics of the Versatile Test Reactor, which is being designed as a 300-MWt sodium-cooled fast reactor. The purpose of the Versatile Test Reactor is to accelerate the testing of advanced nuclear materials in the United States. The proposed multiphysics model relies on the Griffin reactor physics code, the SAM thermal–hydraulic system code, the BISON fuel performance code, as well as generic Multiphysics Object-Oriented Simulation Environment capabilities implemented in the open-source tensor mechanics module. For keff calculations, the introduction of a tight coupling between the neutronics, thermo-mechanical and thermal–hydraulics models induces a change of around 543 pcm in the eigenvalue, compared to the traditional standalone neutronics calculation where approximate temperature profiles are used. The multiphysics model is then employed for quantifying the impact of the thermal conductivity uncertainties on some of the key figures of merit, such as the fuel centerline temperature, assembly powers, and keff for nominal core conditions. As anticipated, uncertainties on fuel thermal conductivity mostly impact the fuel centerline temperature, and to a lesser extend the keff.

Validation of MCS code for shielding calculation using SINBAD

The MCS code is a computer code developed by the Ulsan National Institute of Science and Technology (UNIST) for simulation and calculation of nuclear reactor systems based on the Monte Carlo method. The code is currently used to solve two main types of reactor physics problems, namely, criticality problems and radiation shielding problems. In this paper, the radiation shielding capability of the MCS code is validated by simulating some selected SINBAD (Shielding Integral Benchmark Archive and Database) experiments. The whole validation was performed in two ways. Firstly, the functionality and computational rationality of the MCS code was verified by comparing the simulation results with those of MCNP code. Secondly, the validity and computational accuracy of the MCS code was confirmed by comparing the simulation results with the experimental results of SINBAD. The simulation results of the MCS code are highly consistent with the those of the MCNP code, and they are within the 2σ error bound of the experiment results. It shows that the calculation results of the MCS code are reliable when simulating the radiation shielding problems.

Development of a Trajectory Period Folding Method for Burnup Calculations

In this paper, we present a trajectory period folding method for numerical modelling of nuclear transformations. The method uses the linear chain method, commonly applied for modelling of isotopic changes in matter. The developed method folds two consecutive periods of time and forms linear chain representations. In the same way as in the linear chain method, the mass flow of straight nuclide-to-nuclide transitions following the formation of nuclide transmutation chains in every step is considered over the total period of interest. Therefore, all quantitative data about the isotopic transformations for the period beyond a particular calculation step are preserved. Moreover, it is possible to investigate the formation history of any isotope from the beginning of irradiation to the arbitrary time step, including cooling periods and multi-recycling for any designed nuclear fuel cycle. We present a case study for the transition from 238U to 239Pu and define the properties of the developed method and its possible applications in reactor physics calculations.

A novel SN/MOC collaborative method for few-group cross-section generation with versatile neutron spectral adaptability in advanced reactor physics analysis

This study proposes a novel deterministic discrete ordinate method (SN)/method of characteristics (MOC) collaborative method for few-group cross-section generation to realize versatile spectral computation for different types of neutron energy spectra. The method encompasses fast/intermediate/thermal spectra, which facilitates advanced reactor physics analysis. The SN with multi-direction scanning was initially used to compute the region-wise isotope-dependent background cross-sections to accommodate the complicated spectral features over the entire energy range and obtain the fine-group cross-sections. The MOC was then adopted to acquire accurate weighting spectra over the entire domain for few-group cross-section generation. As a practical application, the proposed method was implemented in AVENUE code development, which was employed for the reactor physics analysis of typical fast/thermal spectra, over-moderated thermal spectra, and complicated intermediate spectra with additional parallelization and acceleration techniques. The verification results revealed that the proposed SN/MOC collaborative method enables accurate few-group cross-section generation with versatile spectral adaptability.

The effect of enriched gadolinia and its concentrations on the neutronic parameters of AP-1000 fuel assembly

This paper investigates methods of effective use of gadolinium-bearing rods as burnable poisons in AP-1000 fuel assembly. To do this, a sensitivity analysis was performed on the effect of using the enriched Gd2O3 and various concentrations of Gd2O3 fuel pins. Different neutronic parameters of AP-1000 FA (that is widely used in pressurized water reactors) such as Kinf, peak pin power, reactivity swing, and the residual binding at different burnups are analyzed. The main goal of this investigation is to effectively use burnable poisons to maintain the reactivity swing as low as possible to decrease the frequent moving of the control rods. DRAGON code is used to develop subassembly designs. The cases with enriched Gd157 have burned out near 11 GWd/MTU and for Gd155 cases, it was near 17 GWd/MTU. The reactivity change of cases with enriched Gd155 is smoother than the cases with enriched Gd157. Our reactor physics study arrives at the flattest reactivity curve and the highest burnout time with the highest value of Gd155 enrichment (100 w/o). This design gives the lowest reactivity swing as well as a better pin power peaking factor. The new suggested design of enriched gadolinia bearing rods performed excellently in less reactivity swing, power peaking factor, and reactivity binding.

Investigation of novel weight window methods in Serpent 2 for fusion neutronics applications

Released in 2009, the Serpent Monte Carlo code has established itself as a highly efficient and powerful simulation code for nuclear systems analysis. Originally developed for reactor physics applications, the scope of the code now extends to coupled multi-physics simulations and photon transport. The latter has allowed adoption of the code by the fusion neutronics community following developments of a coupled neutron-photon capability in 2014 and the ability to handle complex geometry types in 2016. The code is well validated for the energy regimes and geometry types one can expect in fission reactor analysis. Over the course of recent years a benchmarking effort has been undertaken for application of the code to nuclear fusion. Compared to nuclear fission, or accelerator based applications, the underlying particle interaction phenomena differ greatly at the energies expected in a fusion reactor as well as the specific responses that are of interest. In this paper, a novel weight window generation implementation in Serpent is investigated. The applicability of this method is demonstrated for the Frascati Neutron Generator (FNG) bulk blanket and shield experiment, part of the SINBAD database, and a DEMO helium cooled pebble bed (HCPB) computational model. A comparison is performed against MCNP using weight windows generated with ADVANTG. Excellent agreement is found for the specified tallies and the significant efficiency gain using weight windows generated using both methods is comparable. A robust variance reduction method implementation is fundamental to applications to fusion neutronics and as such, this work is an important step in deployment of Serpent for this type of analysis.

Data-Enabled Physics-Informed Machine Learning for Reduced-Order Modeling Digital Twin: Application to Nuclear Reactor Physics

This paper proposes an approach that combines reduced-order models with machine learning in order to create physics-informed digital twins to predict high-dimensional output quantities of interest, such as neutron flux and power distributions in nuclear reactor cores. The digital twin is designed to solve forward problems given input parameters, as well as to solve inverse problems given some extra measurements. Offline, we use reduced-order modeling, namely, the proper orthogonal decomposition, to assemble physics-based computational models that are accurate enough for the fast predictive digital twin. The machine learning techniques, namely, k-nearest-neighbors and decision trees, are used to formulate the input-parameter-dependent coefficients of the reduced basis, after which the high-fidelity fields are able to be reconstructed. Online, we use the real-time input parameters to rapidly reconstruct the neutron field in the core based on the adapted physics-based digital twin. The effectiveness of the framework is illustrated through a real engineering problem in nuclear reactor physics—reactor core simulation in the life cycle of the HPR1000 governed by the two-group neutron diffusion equations affected by input parameters, i.e., burnup, control rod inserting step, power level, and temperature of the coolant—which shows potential applications for online monitoring purposes.

EFFECTS OF FUEL DENSITY ON REACTIVITY COEFFICIENTS AND KINETIC PARAMETERS OF PEBBLE BED REACTOR

EFFECTS OF FUEL DENSITY ON REACTIVITY COEFFICIENTS AND KINETIC PARAMETERS OF PEBBLE BED REACTOR. Few decades ago a large number of nuclear reactors were designed to use HEU as the fuel. But the use of HEU is being discouraged since it can be used as a nuclear explosive material which makes its proliferation potential. Most of the HEU-fueled nuclear reactors in the world are either closed or converted into other types that have LEU fuel with 235U enrichment below 20%. To extend lifetime, LEU fuel with high density is developed. The change in fuel density from low to high will also change core neutronics and thermal-hydraulics of the reactor, and as a result, it affects the transient response of the reactor. This paper studies the effects of fuel density on reactivity coefficients and kinetic parameters of pebble bed reactor through several calculations with MCNP6 code combine with ENDF/B-VII.1 continuous energy cross-section nuclear data library. The overall calculation results show that the Doppler temperature coefficient (DTC) increases with increasing fuel density, but the moderator temperature coefficient (MTC) decreases due to hardening of neutron spectrum. Kinetic parameters such as effective delayed neutron fraction (βeff), prompt neutron lifetime(ℓ) and neutron generation time (Ʌ) which significantly reduced with increasing fuel density will strongly affect the reactor control and safety. The results of this work conclude that the selection of 9–15 g/cm3 fuel density should be considered carefully given that its effect on reactor controllability.Keywords: Fuel density, reactivity coefficients, kinetic parameters, MCNP6, ENDF/B-VII.1.

Preliminary study on TRU transmutation in VVER-1000 fuel assembly using MCNP6

Abstract Transmutation technology has been developed to address the issue of long-term safety of nuclear waste disposal in geological repository. Transuranic (TRU) transmutation in thermal neutron spectrum can be considered more beneficial at the present time due to the larger fission cross section of TRU elements than in fast neutron spectrum. This paper discusses preliminary study on TRU transmutation in Vodo-Vodyanoi Energetichesky Reactor (VVER)-1000 fuel assembly using MCNP6 code. The fuel assembly is configured by 312 fuel cells consist of 300 UO2 fuel rods with 3.7 wt% 235U and 12 UGD fuel rods containing a mixture of UO2 with 3.6 wt% 235U and 4.0 wt% Gd2O3. The calculation results show that the k inf value is higher with the increase in the TRU thickness at the beginning of cycle and at the end of cycle, which means that the addition of TRU will increase the fuel cycle length. The total temperature coefficient of reactivity is negative for all fuel assemblies both without TRU and with TRU. In general, the computed β eff values of the assembly with and without TRU addition are not significantly different. It shows that the coating of the TRU layer in the UGD fuel cell will not complicate the reactor control. The transmutation of TRU recycled from spent nuclear fuel in the VVER-1000 fuel assembly can be considered feasible from the viewpoint of excess reactivity and control safety characteristics. The total transmutation rate of ∼55.73% can be achieved with 239Pu and 241Am isotopes are taking the largest portion of transmuted nuclides.

On development of in-core fuel system for PWR reactors: Part I generation of macroscopic cross sections using SCALE 6.0 for use in nodal calculation

This work presents the description of the first part of a methodology applied to perform In-Core Fuel Management (ICFM) in Pressurized Water Reactor (PWR). The ICFM of a PWR reactor consists on defining the best charging or recharging pattern of fuel assemblies inside a reactor for an operational cycle. This means, finding a suitable arrangement of fuel assemblies that optimizes the performance of the reactor, which complies with all safety criteria. Genetic algorithms (GAs) are used to select the arrangements that interact with the reactor physics simulation code, holding the neutron characteristics of each fuel assembly. Therefore, a reliable and fast code was developed accordingly. The consolidated technique of coarse mesh node code that numerically solves the multigroup diffusion equation for two groups of energy, fast and thermal neutrons, in two dimensions was selected. In this type of code, it is essential that each fuel assembly is homogenized and characterized by its macroscopic cross sections, for each reactor’s burnup condition. The cross sections are generated with the support of SCALE 6.0, computational platform developed by the Reactor and Nuclear Systems Division (RNSD), from the Oak Ridge National Laboratory (ORNL). The completeness of the qualification and validation of the results obtained from the homogenization of the fuel assembly by the SCALE was performed comparing the results with actual data of a benchmark reactor. The fully documented Almaraz Nuclear Power Plant provided by the International Atomic Energy Agency (IAEA)-TECDOC-815, has been used as benchmark with successful results.

Noise Characteristics and Control of High Voltage Shunt Reactor

Noise Characteristics and Control of High Voltage Shunt Reactor

An enhanced formalism for the inverse reactor kinetics problem

Abstract The inverse kinetics problem in reactor physics is a standard formalism to unfold reactivity on the basis of registered power (flux) profile. The classical inverse point kinetics framework has been retrofitted herein to comprise thermal reactivity feedback effects. The instantaneous fuel and coolant temperatures are thus computed by way of the exponential time-differencing scheme and the corresponding thermal reactivity feedback is plugged into the inverse kinetics module. The core external reactivity is therefore unfolded employing only two consecutive time-steps of the power (flux) profile. A history independent yet straightforward numerical routine is accrued enjoying noticeable robustness with regards to the time-step resolution.

Mathematical modeling of point kinetic equations with temperature feedback for reactivity transient analysis in MTR

Abstract The behavior of the nuclear reactor in response to any sudden change in reactivity is very important for reactor control. Positive reactivity insertions causes power excursion and could have a destructive impact on the reactor core. The aim of the study is to investigate the safety features of a material test reactor (MTR) during reactivity transient with emphasis on the capability of the mathematical modeling using programming language. Therefore a mathematical model using Python3.6; high-level programming language is developed to solve the point kinetic equations taking into account Doppler and moderator feedback effects. The model is validated with AIREKMOD_RR; point kinetic computer code for reactivity transient analysis in nuclear research reactors. The results of the Python model demonstrate the inherent safety features of the MTR reactor. Also, there is good agreement between the results of the Python model and AIREKMOD_RR code, illustrating the efficiency of the Python model in simulating the behavior of the reactor core under reactivity transient.

Supplementing fuel behaviour analyses via coupled Monte Carlo neutronics and fission product solution

The accuracy of power distribution solution and burnup algorithms in standalone fuel behaviour codes is limited. For improved accuracy, fuel behaviour codes can be coupled with external reactor physics solvers. A coupled calculation system between the Monte Carlo neutronics and burnup calculation solver Serpent and the fuel behaviour solver TRANSURANUS was developed in an earlier work. The data exchanged between the solvers included fuel pin axial and radial power and temperature distributions, the axial fast flux distribution in the cladding and the changes in the pin radii. In this work the coupling was further improved by developing the capability to transfer and utilize Serpent calculated nuclide compositions in TRANSURANUS within the coupled calculations. Additionally, support for corrector type burnup algorithms in the coupled solution was now implemented for increased accuracy of the nuclide solution. The new capabilities were demonstrated with a coupled single rod burnup calculation utilizing power and coolant history data from the Loviisa nuclear power plant. The focus at this stage was on fission product nuclides as their accumulation adversely affects the fuel performance, making their accurate solution important to ensure safe yet economical use of fuel. The demonstration showed small but visible differences when compared to simulations that were run without using fission products calculated by Serpent. The developed capabilities will now facilitate import of further fission product nuclides to be taken into account in TRANSURANUS.

Jacobian-free Newton Krylov two-node coarse mesh finite difference based on nodal expansion method

A Jacobian-Free Newton Krylov Two-Nodal Coarse Mesh Finite Difference algorithm based on Nodal Expansion Method (NEM_TNCMFD_JFNK) is successfully developed and proposed to solve the three-dimensional (3D) and multi-group reactor physics models. In the NEM_TNCMFD_JFNK method, the efficient JFNK method with the Modified Incomplete LU (MILU) preconditioner is integrated and applied into the discrete systems of the NEM-based two-node CMFD method by constructing the residual functions of only the nodal average fluxes and the eigenvalue. All the nonlinear corrective nodal coupling coefficients are updated on the basis of two-nodal NEM formulation including the discontinuity factor in every few newton steps. All the expansion coefficients and interface currents of the two-node NEM need not be chosen as the solution variables to evaluate the residual functions of the NEM_TNCMFD_JFNK method, therefore, the NEM_TNCMFD_JFNK method can greatly reduce the number of solution variables and the computational cost compared with the JFNK based on the conventional NEM. Finally the NEM_TNCMFD_JFNK code is developed and then analyzed by simulating the representative PWR MOX/UO2 core benchmark, the popular NEACRP 3D core benchmark and the complicated full-core pin-by-pin homogenous core model. Numerical solutions show that the proposed NEM_TNCMFD_JFNK method with the MILU preconditioner has the good numerical accuracy and can obtain higher computational efficiency than the NEM-based two-node CMFD algorithm with the power method in the outer iteration and the Krylov method using the MILU preconditioner in the inner iteration, which indicates the NEM_TNCMFD_JFNK method can serve as a potential and efficient numerical tool for reactor neutron diffusion analysis module in the JFNK-based multiphysics coupling application.

Analysis of several VERA benchmark problems with the photon transport capability of STREAM

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS – also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark.

A study on the acoustic characteristics of the sound-absorbing structure in single-column UHV Shunt reactors

Single-column UHV shunt reactors usually generate large noises with long wavelengths. For this reason, the noises along the substation boundaries may exceed the standards. We aimed at noise control of single-column UHV shunt reactors in the present study. Based on the noise characteristics of single-column UHV shunt reactors, we employed an acoustic impedance test apparatus and studied the influences of different structural parameters on the acoustic performance of the compound impedance sound-absorbing structure. The results showed that the compound impedance sound-absorbing structure had a better sound-absorbing performance at low frequencies if a microperforated plate was introduced with basalt cotton as the filler. Besides, as the surface density, thickness, size of the dorsal cavity, and size of the intermediate cavity increased, the sound absorption coefficient at low frequencies increased. After the surface density increased to a certain level, the increase rate of the sound-absorbing performance dropped suddenly. Moreover, the intermediate cavity achieved a lesser enhancement of sound-absorbing performance at low frequencies than the dorsal cavity.

Rolling Wheel Abrasion Condition Monitoring of Control Rod Drive Mechanism in Nuclear Reactor

During the operation of in nuclear reactor, the reliability is largely influenced by the changes of working states of control rod drive mechanism. The abrasion of rolling wheel is one of the most serve faults leading to the control failure. To monitor the rolling wheel in the control rod drive mechanism in nuclear reactor, a new recognition method based on complexity of the singular spectrum entropy has been introduced. In this method, the singular spectrum entropy of vibration signal of the system is firstly calculated. Then an index of brain wave information complexity is used to define the singular spectrum entropy complexity. The complexity index is more sensitive to the uncertainty of the signal and it can effectively reflect the intensity of the random shock signal by abrasion which makes it possible to assess the abrasion condition. Through the control rod drive mechanism of whole life test, it is proved that this method can be used in nuclear reactor control rod drive mechanism rolling wheel abrasion condition monitoring with proper accuracy.