Pin Power Research Articles

Neutronic and fuel cycle performance of LEU fuel with different means of excess reactivity control: Impact of neutron leakage and refueling scheme

Low enriched uranium (LEU) fuel with different Integral Burnable Absorbers (IBAs) and Burnable Poison Rods (BPRs) in AP1000 assembly was investigated. These include IBAs of Gd2O3, Er2O3, AmO2, PaO2, and BPRs of Al2O3-B4C. The conventional uranium dioxide (UO2) fuel and two accident tolerant fuels (ATFs) namely UN-U3Si2 and U3Si2 were taken as the base fuel materials. Nine fuel compositions of these fuels with and without burnable poison were simulated in the lattice physics code Dragon, version 5. The burnup-dependent infinite multiplication factor was determined and significant variation was found due to the addition of poison. Reactivity was calculated at different levels of neutron leakage (0–5%). More positive reactivity at the early burnup stage and less negative reactivity at end of life were found for PaO2. The Linear Reactivity Model (LRM) was used to calculate the fuel cycle parameters. Cycle burnup and discharge burnup were determined considering different refueling batch number at each leakage scenario. Both parameters are found to decrease with neutron leakage while a large batch number results in a higher discharge burnup. Reduced discharge burnup was observed for each IBA and BPR except PaO2. Cycle length was also calculated for each of the composition and cycle length penalty was positive only for PaO2. The fuel temperature coefficient of reactivity, pin power distribution, and energy-dependent neutron flux were determined using the Monte Carlo code OpenMC. The burnup-dependent isotopic compositions were also studied.

SiC and FeCrAl as Potential Cladding Materials for APR-1400 Neutronic Analysis

The aim of this study is to investigate the potential improvement of accident-tolerant fuels in pressurized water reactors for replacing existing reference zircaloy (Zr) fuel-cladding systems. Three main strategies for improving accident-tolerant fuels are investigated: enhancement of the present state-of-the-art zirconium fuel-cladding system to improve oxidation resistance, replacement of the current referenced fuel-cladding system material with an alternative high-performance oxidation-resistant cladding, and replacement of the current fuel with alternative fuel forms. This study focuses on a preliminary analysis of the neutronic behavior and properties of silicon carbide (SiC)-fuel and FeCrAl cladding systems, which provide a better safety margin as accident-tolerant fuel systems for pressurized water reactors. The typical physical behavior of both cladding systems is investigated to determine their general neutronic performance. The multiplication factor, thermal neutron flux spectrum, 239Pu inventory, pin power distribution, and radial power are analyzed and compared with those of a reference Zr fuel-cladding system. Furthermore, the effects of a burnable poison rod (Gd2O3) in different fuel assemblies are investigated. SiC cladding assemblies present a softer neutron spectrum and a lower linear power distribution compared with the conventional Zr-fuel-cladding system. Additionally, the SiC fuel-cladding system exhibits behaviors that are consistent with the neutronic behavior of conventional Zr fuel-cladding systems, thereby affording greater economic and safety improvements.

An approach to minimize reactivity penalty of Gd2O3 burnable absorber at the early stage of fuel burnup in Pressurized Water Reactor

The high capture cross-section (σc) of Gadolinium (Gd-155 and Gd-157) causes reactivity penalty and swing at the initial stage of fuel burnup in Pressurized Water Reactor (PWR). The present study is concerned with the feasibility of the combination of mixed burnable poison with both low and high σc as an approach to minimize these effects. Two considered reference designs are fuel assemblies with 24 IBA rods of Gd2O3 and Er2O3 respectively. Models comprise nuclear fuel with a homogeneous mixture of Er2O3, AmO2, SmO2, and HfO2 with Gd2O3 as well as the coating of PaO2 and ZrB2 on the Gd2O3 pellet's outer surface. The infinite multiplication factor was determined and reactivity was calculated considering 3% neutron leakage rate. All models except Er2O3 and SmO2 showed expected results namely higher values of these parameters than the reference design of Gd2O3 at the early burnup period. The highest value was found for the model of PaO2 and Gd2O3 followed by ZrB2 and HfO2. The cycle burnup, discharge burnup, and cycle length for three batch refueling were calculated using Linear Reactivity Model (LRM). The pin power distribution, energy-dependent neutron flux and Fuel Temperature Coefficient (FTC) were also studied. An optimization of model 1 was carried out to investigate effects of different isotopic compositions of Gd2O3 and absorber coating thickness.

Validating the serpent-ants Calculation Chain Using BEAVRS Fresh Core Hot Zero Power Data

Abstract The serpent Monte Carlo code and the serpent-ants two-step calculation chain are used to model the hot zero power physics tests described in the BEAVRS benchmark. The predicted critical boron concentrations, control rod group worths, and isothermal temperature coefficients are compared between serpent and serpent-ants as well as against the experimental measurements. Furthermore, radial power distributions in the unrodded and rodded core configurations are compared between serpent and serpent-ants. In addition to providing results using a best practices calculation chain, the effects of several simplifications or omissions in the group constant generation process on the results are estimated. Both the direct and two-step neutronics solutions provide results close to the measured values. Comparison between the measured data and the direct serpent Monte Carlo solution yields root-mean-square (RMS) differences of 12.1 mg/kg, 25.1×10−5 and 0.67×10−5 K−1 for boron, control rod worths and temperature coefficients, respectively. The two-step serpent-ants solution reaches a similar level of accuracy with RMS differences of 17.4 mg/kg, 23.6×10−5 and 0.29×10−5 K−1. The match in the radial power distribution between serpent and serpent-ants was very good with the RMS and MAXimum (MAX) for pin power errors being 1.31% and 4.99%, respectively, in the unrodded core and 1.67% (RMS) and 8.39% (MAX) in the rodded core.

Validation of the advanced neutron transport lattice code KYLIN-V2.0 with critical and burnup experiments

KYLIN-V2.0 is an advanced lattice code developed in Nuclear Power Institute of China. The subgroup method, MOC method and Chebyshev rational approximation method (CRAM) are applied in the resonance, transport and burnup calculation to provide the macro cross-section library. In this paper, several critical experiments and burnup benchmarks, including B&W critical experiment, KRITZ critical experiment and Takahama-3 burnup benchmark, are applied in the validation of KYLIN-V2.0. KYLIN-V2.0 is validated by 21 cases in B&W-1484, 23 cases in B&W-1810 and 16 samples in Takahama-3. Eigenvalue, pin power distribution and nuclides density results of these cases show the good consistency compared with measured data. The KYLIN-V2.0 code shows good heterogeneous calculation ability for two-dimensional lattices.

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.

Coupling MCNP6/ANSYS codes to calculate axial temperature and power distribution in a VVER-1000 fuel assembly

Abstract The precise simulation of the reactor core either in steady state or during transients is very important for the safety assessment of nuclear power plants. This requires accurate determination of the parameters that influence the reactor operation. Coupling neutronic and thermal hydraulic schemes are developed to calculate these parameters. In the present paper, a coupling scheme between MCNP6 and ANSYS-FLUENT17.2 codes is proposed to obtain accurate radial and axial temperature distribution and hence pin power distribution for VVER-1000 fuel assembly. The Performance of the developed coupling scheme is investigated in steady state calculations. An iterative process is associated with the exchange of data between codes to meet the convergence criteria. The results obtained demonstrate that the proposed coupling scheme is able to simulate the VVER-1000 fuel assembly accurately. It gives information about thermal and neutronic behavior of the assembly and allows the feedback effects to be accurately modeled. This work is a step forward to establish a consistent methodology to be used in transient calculations.

Neutronic and fuel cycle performance of VVER-1000 for dual cooled annular fuel with coated burnable poison

The feasibility of the dual cooled annular fuel rod with coated burnable absorber in the hexagonal assembly of the VVER-1000 reactor was studied. The models consist of a 4.5% enriched dual cooled UO2 fuel rod with a coating of Gd2O3, Er2O3, PaO2, AmO2, and ZrB2 as the burnable absorber. Designed models along with the reference designs were simulated in lattice physics code Dragon and Monte Carlo code OpenMC. Burnup analysis was conducted for 1620 EFPD. The multiplication factor, as well as reactivity, was determined for fuel burnup. Improvement in the multiplication factor and reactivity is observed for dual cooled annular fuel with coated burnable poison. Designed models also showed a significant increase in thermal fission reaction rate compared to the reference design. Linear reactivity model (LRM) was used to calculate cycle burnup, discharge burnup, and cycle length for three and four batch refueling schemes. No degradation is observed in fuel cycle performance using dual cooled annular fuel with coated burnable absorber except for the model of gadolinium. The other absorbers showed better results than gadolinium in both neutronic and fuel cycle performance. PaO2 used model showed the most promising result among them. The value of k-infinity was found to be 1.27089 at BOL and the cycle burnup was around 15 MWD/kg. Also, no cycle length penalty was observed for PaO2. Burnup-dependent atomic concentration was studied. A lower concentration of fission product poison was also noticed in designed models. The analysis of relative pin power distribution and energy-dependent neutron flux is also included in this study.

Pin Power Reconstruction for Hexagonal Geometry in Nodal Neutronics Program Ants

Pin Power Reconstruction for Hexagonal Geometry in Nodal Neutronics Program Ants

Derivation of optimal process MOC parameters and analysis of the 2D C5G7 MOX benchmark using DRAGON5 code

This work uses the Characteristics Methods (MOC) implemented in the lattice DRAGON5 code to analyze the 2D C5G7 MOX Benchmark. This method was performed using the MCCGT: module of DRAGON5, this module includes parameters that affect the results. So, a sensitivity study was necessary to find the optimum parameters. The eigenvalues calculated by DRAGON5 code agree well with the reference Monte Carlo code MCNP, and also with the OpenMC code with a percent error of just 0.001% and 0.004% respectively. Compared to the Monte Carlo code OpenMC, the maximum percentage error of assembly power, the maximum and the average percentage error of pin power are 0.07%, 0.85% and 0.13%, respectively.Based on the accuracy of the results and on the computational cost, the sensitivity study of various MOC parameters reveals that the 2D C5G7 benchmark problem requires: between 16 and 32 for the azimuthal angles, more than 10 cm-1 for the ray density and 8 × 8 for the special mesh.

Refinements of the Pin-Based Pointwise Energy Slowing-Down Method for Resonance Self-Shielding Calculation—I: Theory

The pin-based pointwise energy slowing-down method (PSM), which is a resonance self-shielding method, has been refined to treat the nonuniformity of material compositions and temperature profile in the fuel pellet by calculating the exact collision probability in the radially subdivided fuel pellet under the isolated system. The PSM has generated the collision probability table before solving the pointwise energy slowing-down equation. It is not exact if the fuel pellet has nonuniform material compositions or temperature profile in all the subdivided regions. In the refined PSM-CPM, the pre-generated table is not required for directly calculating the collision probability in all the subdivided regions of the fuel pellet while solving the slowing-down equation. There are an advantage and a disadvantage to the method. The advantage is to exactly consider the nonuniformity of the material compositions and temperature profile in the fuel pellet. The disadvantage is the longer computing time than that of the PSM when the fuel pellet has more than five subdivided regions. However, in the practical use for UO2 pin-cells, it is still comparable for the computation time with the PSM and the conventional equivalence theory methods. In this article, using simple light water reactor 17 × 17 F A problems with a uniform material composition and temperature profile, it is demonstrated that PSMs (PSM and PSM-CPM) exhibit consistent accuracy in calculating the multiplication factor and the pin power distribution with no compromise in the computation time. More detailed accuracy assessments with various test cases, including problems representing the nonuniformity, are presented in the accompanying article.

Refinements of Pin-Based Pointwise Energy Slowing-Down Method for Resonance Self-Shielding Calculation-II: Verifications

The pin-based pointwise energy slowing-down method (PSM) has been refined through eliminating the approximation for using the pre-tabulated collision probability during the slowing-down calculation. A collision probability table is generated by assuming that material composition and temperature are constant in the fuel pellet using the collision probability method (CPM). Refined PSM (PSM-CPM), which calculates the collision probability in the isolated fuel pellet during the slowing-down calculation using CPM, can consider nonuniform material and temperature distribution. For the methods, the extensive comparative analysis is performed with problems representing various possible conditions in a light water reactor (LWR) design. Conditions are categorized with the geometry, material distribution, temperature profile in the fuel pellet, and burnup. With test problems, PSMs (PSM and PSM-CPM) have been compared with conventional methods based on the equivalence theory. With overall calculation results, PSMs show the accuracy in the eigenvalue with differences in the order of 100 pcm compared to the reference results. There was no noticeable difference in the multigroup cross sections, reaction rates, and pin power distributions. However, PSM-CPM maintains the accuracy in the calculation of the fuel temperature coefficient under the condition with 200% power and nonuniform temperature distribution in the fuel pellet. PSM shows the difference in the eigenvalue in the order of 2,000 pcm for the fictitious pin-cell problem with highly steep temperature profiles and material compositions, but PSM-CPM shows the difference in the eigenvalue within 100 pcm.

Preliminary assessments of variations in the form-function-based maximum pin power in CANDU analyses

In the current CANDU full-core physics analysis, the maximum pin power within a bundle is determined by multiplying the bundle power obtained from solving the three-dimensional neutron diffusion equation with a ratio of the highest pin power to the average pin power, obtained from burnup-dependent infinite lattice calculations. Since the geometry of the fuel channel changes during the life of the reactor, the ratio of the highest to the average pin power changes as well. The objective of the current paper is to quantify such variations. Two sources of variation are considered: (i) pressure tube diametral creep (PTDC); and (ii) the placement of the fuel bundle inside the fuel channel in the lattice model. The results from the study indicate that the maximum pin power could vary up to 1.2%, depending on the amount of the PTDC, the fuel burnup, and the fact that a concentric or non-concentric lattice model is used. Therefore, it is recommended that the lattice model used for deriving the form function be as consistent as possible to the core condition being analysed, especially with respect to the PTDC value.

Preliminary neutronic analysis of alternative cladding materials for APR-1400 fuel assembly

This paper presents a preliminary neutronics analysis for Accident Tolerant Fuel (ATF) cladding materials for standard APR-1400 reactor A0 fuel assembly. AMPT, FeCrAl, 304 and 310 stainless steels, and SiC were considered and compared with the original Zircaloy-4 cladding material. A parametric evaluation was done for fuel and cladding materials to confirm the geometry requirements to achieve the end-of-cycle fuel reactivity, and the results were compared with the standard APR-1400 reference fuel-cladding system. Serpent Monte Carlo reactor physics code version 2.31 was utilized to evaluate the associated neutronics penalty when different cladding materials were used such as, fuel reactivity, thermal neutron spectrum, plutonium and isotopic inventory evolution, and fuel assembly linear pin power distribution. In the examined cases, fuel enrichments were fixed, and cladding thickness and pellet diameter were changed for zircaloy, cases 1 – 4. As a result, thermal neutron absorption cross section varies between materials due to higher absorbing cross section leads to hardening of the neutron spectrum in the fuel-cladding system. That leads to a slight increase in the production of actinides in the inventory such as plutonium.

Environment Effect Treatments in PWR Whole-Core Pin-by-Pin Calculation

The environment effect arises when pin-cell homogenized parameters are generated with reflective boundary conditions. To treat it in whore-core pin-by-pin calculation, two works are summarized in this article. Firstly, by analyzing the relative errors of pin-cell homogenized group constants and the relative importance of pin-cell discontinuity factors (PDF) in each group, the importance of correcting the PDF of the thermal group is recognized. Secondly, the least-squares method for a multivariate polynomial is utilized to functionalize the relation of the thermal group PDF and the core parameters, including diffusion coefficient, removal cross-section, neutron source, and normalized surface flux. The C5G7 and KAIST benchmarks are employed to evaluate the performance of the PDF predication. Numerical results indicate its effectiveness in reducing the errors of eigenvalue and pin power, especially for the cases with the fuel pins located near the interface between different assemblies.

High performance 3D pin-by-pin neutron diffusion calculation based on 2D/1D decoupling method for accurate pin power estimation

The methods and performance of a 3D pin-by-pin neutronics code based on the 2D/1D decoupling method are presented. The code was newly developed as an effort to achieve enhanced accuracy and high calculation performance that are sufficient for the use in practical nuclear design analyses. From the 3D diffusion-based finite difference method (FDM) formulation, decoupled planar formulations are established by treating pre-determined axial leakage as a source term. The decoupled axial problems are formulated with the radial leakage source term. To accelerate the pin-by-pin calculation, the two-level coarse mesh finite difference (CMFD) formulation, which consists of the multigroup node-wise CMFD and the two-group assembly-wise CMFD is implemented. To enhance the accuracy, both the discontinuity factor method and the super-homogenization (SPH) factor method are examined for pin-wise cross-section homogenization. The parallelization is achieved with the OpenMP package. The accuracy and performance of the pin-by-pin calculations are assessed with the VERA and APR1400 benchmark problems. It is demonstrated that pin-by-pin 2D/1D alternating calculations within the two-level 3D CMFD framework yield accurate solutions in about 30 s for the typical commercial core problems, on a parallel platform employing 32 threads.

Identification of power PIN diode design parameters: Circuit and device-based simulation approach

This paper aims to present a detailed systematic approach to identify the main design parameters of PIN power diodes. Firstly, the diode physical parameters are initialized using simple analytical equations. The second phase is the optimization of the diode parameters considering PSPICE circuit simulation where an electro-thermal physically based circuit model is utilized depending on a series of dynamic and static measurements. The final optimization step is carried out by using TCAD simulations. First, the diode extracted parameters are used to virtually fabricate the diode by using a process simulator. Then, using the output of the process simulator, a device simulator is used to get the desired output that is validated against experimental data. Three case studies for different power diodes are presented showing a good agreement between circuit/device simulation results and measurements. The presented methodology provides high accuracy like TCAD-based parameter extraction procedure with less time. In addition, it gives higher accuracy than the widely used circuit-based parameter extraction technique.

A New Embedded Analysis with Pinwise Discontinuity Factors for Pin Power Reconstruction

Based on embedded analysis, an accurate pin power reconstruction (PPR) method is proposed for conventional nodal analysis. Unlike the common form function (FF) method, the new PPR method, named the embedded pin power reconstruction (EPPR) method, directly solves a two-group fixed-source problem that is defined with pinwise homogenized group constants (HGCs) and coarse-mesh incoming partial currents on the boundary. In the EPPR scheme, the pinwise HGCs including the pinwise discontinuity factor are predetermined from single-assembly lattice calculations, and the boundary partial currents are obtained from two-step nodal analyses. Two EPPR approaches are proposed: One is a 3×3 extended color-set configuration, and the other is a smaller one considering the half-thickness of the surrounding fuel assemblies. The performance of the EPPR methods is evaluated with various benchmark problems including partially mixed oxide–loaded pressurized water reactor cores, and the results are compared with the conventional FF method. Comprehensive results of this work demonstrate that the new EPPR method can provide much better accuracy than the conventional FF-based PPR method.

Extended Applications of Subgrid Representation in the 2D/1D Method

Recent efforts in MPACT have focused on improving the performance of the 2D/1D subplane implementation to help target computational performance goals. This paper builds on previous efforts that targeted the use of subgrid treatments to improve the accuracy of control rod representation, presenting three additional applications of subgrid treatments with the goal of reducing the computational burden of simulations. These subgrid applications include treatment of spacer grids, thermal feedback, and axial reflector material representation. With these approaches, a single method of characteristics (MOC) plane can contain several different materials axially that are represented explicitly via subgrids on the coarse mesh finite difference (CMFD) mesh but are axially homogenized on the MOC mesh. This allows for a substantial reduction in the number of MOC planes needed in the calculation through the introduction of an approximate treatment, particularly with regard to the self-shielded cross sections and MOC-informed radial current coupling coefficients in CMFD. Several test problems ranging from single rod to quarter core are used to assess the solution accuracy and performance of these various subgrid representations. Overall, the accuracy of the approximations seems very reasonable, with extremely small differences in eigenvalue observed and maximum pin power errors in the 0.5% to 1.0% range. Several cases show substantial value in the compromise between accuracy and computational performance. Others highlight the new computational hurdles that future research will aim to resolve.

GENERATION OF NODAL CORE SIMULATOR UTILIZING VERA

Nuclear core simulators based upon few-group nodal diffusion method currently are widely used to predict light water reactor core behavior. Nodal parameters’ input, e.g. cross-sections, discontinuity factors, and pin form factors, are typically generated utilizing lattice physics codes. In doing so, a number of approximations are introduced related to using zero current boundary conditions, 2-D radial geometry, and uniform thermal conditions in coolant and fuel. Usage of full core models with prediction fidelity typical of lattice physics to predict nodal parameters would eliminate these approximations. The VERA code can serve as such a full core model and was so utilized in this work. Via subsequent processing of VERA predictions, for a range of state points, nodal parameters and their functionalization in terms of coolant density, fuel temperature, and soluble poison concentration, were obtained and input to the NESTLE nodal code. By processing VERA predictions using consistent nodal methodologies as used in NESTLE, when using nodal parameters after functionalization based upon All-Rods-Out (ARO) VERA state points, the maximum reactivity and pin power differences between VERA and NESTLE were 2 pcm and 0.003 for ARO core simulations. For rodded core simulations, these maximum differences grew to 58 pcm and 0.04. Increases in differences were determined to be attributed to usage of unrodded nodal parameters generated using the VERA ARO state points whether the core is partially rodded or not, consistent with lattice physics practice. Obtaining unrodded nodal parameters using the VERA rodded state points reduced maximum differences to 2 pcm and 0.003 in pin powers.