Pin Power Distribution Research Articles

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.

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.

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.

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 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.

A Source-expansion-based method for transient Pin-Power reconstruction

In this paper, the transient pin-power reconstruction based on source-expansion method has been proposed. The transient calculation is sufficient to simulate the reactor-physics process in second-time scale, such as the load-dump process and dynamic rod worth measurement. During the transient process, the pin-power distribution is essential for the determination of the power-peak factors, including Fq and FΔH. However, with the widely-applied transverse-integrated nodal methods for transient simulation, the detailed pin-averaged results can’t be provided directly. To address this problem, the transient pin-power reconstruction was proposed. It was derived from the transient fixed-source problem assuming the total-source and the transient fixed-source terms could be respectively expanded by the fourth-order Legendre polynomials and the biquadratic Legendre polynomials constructed from the nine-node problem. Moreover, to improve the accuracy of pin-averaged results, the corner-point condition and corner-discontinuity factors were employed in the reconstruction process. This proposed transient pin-power reconstruction method has been implemented in our self-developed code named SPARK and verified with several benchmarks. Results of the verification indicated that this proposed method for transient pin-power reconstruction can provide a satisfactory accuracy in transient process.

DEVELOPMENT OF PIN-LEVEL NEUTRONICS/THERMAL-FLUID ANALYSIS COUPLED CODE SYSTEM FOR A BLOCK-TYPE HTGR CORE

Recently, the coupling between computer codes that simulate different physical phenomena has attracted for more accurate analysis. In the case of high-temperature gas-cooled reactor (HTGR), the coupling between neutronics and thermal-fluid analysis is necessary because of large change of temperature in the reactor core. Korea Atomic Energy Research Institute (KAERI) has developed the coupled code system between a reactor physics analysis code CAPP and a thermal-fluid system safety analysis code GAMMA+ for a block-type HTGR. The CAPP/GAMMA+ coupled code system provides more accurate block-wise distribution data than CAPP or GAMMA+ stand-alone analysis. However, the block-wise distribution data has the limitation in order to predict safety parameters such as the maximum temperature of the nuclear fuel. It is necessary to calculate refined distribution, for example, pin-level (fuel compact level) distribution. In this study, we tried to solve this problem by coupling CAPP and a high-fidelity thermal-fluid analysis code CORONA. CORONA can perform a high-fidelity thermal-fluid analysis of Computational Fluid Dynamics (CFD) level by dividing a block-type HTGR core into small lattices. On the other hand, CAPP can provide a pin power distribution. It is expected that the refined, more accurate distribution data for a block-type HTGR can be obtained by coupling these two codes. This paper presents the development of coupled code system between CAPP and CORONA, and then it is tested on a simple HTGR column problem with encouraging results.

OpenNTP: Implementation of the S[formula omitted] method in cartesian 2D geometry and the CP method in cylindrical and spherical 1D geometry

This paper presents the implementation of the discrete ordinates method (SN) in 2D cartesian geometry and the collision probability method (CP) in cylindrical and spherical 1D geometry in OpenNTP code (Open Neutron Transport Package). This code is a pedagogical tool for computer analysis of nuclear reactors. Its main features are as follows: a free software with an open source, it solves the neutron transport equation to a few steady-state groups on a grid structured in one or two spatial dimensions with an isotropic and anisotropic dispersion source, and any new calculation method and algorithm would be easy to implement in this proposed code. Also, the code offers the possibility to calculate the main parameters of nuclear reactors such as the multiplication factor and the distribution of the scalar and angular neutron fluxes. Additional parameters, like the reaction rates, the pin power distribution and the boundary currents are also calculated. Moreover, a graphical user interface written in Python 3 programming language has been developed to simplify the use of OpenNTP. Some applications of the OpenNTP code have been compared on the one hand with the Monte Carlo OpenMC and MCNP6.1 codes and the WIMSD-5B lattice transport code, on the other hand. Numerical results are given to illustrate the accuracy of the OpenNTP code. Program summaryProgram Title: OpenNTP, version 1.2CPC Library link to program files:https://doi.org/10.17632/gm2b6799j6.1Developer’s repository link:https://github.com/mohamedlahdour/OpenNTPLicensing provisions: GPLv2Programming language: fortran90 and Python 3External routines/libraries: NumPy, Matplotlib, PyQt5, f2pyNature of problem: Solving the steady-state multigroup neutron transport equation by different methods in one, or two spatial dimensions.Solution method: the characteristic method (MOC), the discrete ordinate (SN) method and the collision probability method (CP)

Neutronics Assessment of Accident-Tolerant Fuel in Advanced Power Reactor 1400 (APR1400)

Safety and reliability are the most desirable conditions that each nuclear power plant should improve. Since the Fukushima Daiichi accident, Accident-Tolerant Fuel (ATF) has been extensively researched to improve the performance of the nuclear fuel system. This paper presents the investigation of the ATF system from a neutronics perspective, which positively reflects on the performance of the APR-1400 nuclear power plant. Several advanced fuel candidates such as UC, U 3 Si 2 , and UN, which have better thermophysical properties than current UO 2 nuclear fuel, have been considered. Meanwhile, advanced cladding candidates such as FeCrAl, Zr-alloy with coating, and SiC which can reduce or even eliminate the oxidation rate of current Zr-alloy cladding have been adopted in this study. The Monte Carlo Serpent code, in conjunction with ENDF/B-VII.1 nuclear data library, has been used to calculate and evaluate the important neutronics parameters at the assembly level such as the fuel residence time, discharge burnup, kinetics parameters, pin power distribution, temperature reactivity feedbacks, and fissile evolution. The advanced fuels show better neutronics performance than the current UO 2 fuel. In addition, SiC provides optimum neutronic performance as cladding.

Study of neutronic characteristics for the VVER reactor assembly with five various burnable poisons using MCNPX code

The present work investigates the neutronic characteristics of VVER reactor assembly with various burnable poisons. The MCNPX code is used to design VVER-1000 LEU Assembly model. At first gadolinium oxide (Gd2O3) as integral burnable absorber is used, the results are compared with the results of the reference benchmark. Different fuels (uranium oxide, thorium oxide and thorium plutonium oxide) and five various burnable poisons are used in the study. Gadolinium Oxide (Gd2O3) and Erbium Oxide (Er2O3) are used together in the same assembly as IBAs that are mixed homogenously with 30 uranium fuel rods and Dysprosium Titanite (Dy2O3–TiO2) is coated the fuel rod with 0.001 cm thickness. Also, Iridium Oxide (IrO2) used in guide tube as BPR and finally Minor Actinides are used as both IBAs or BPRs. A number of parameters are determined at BOL such as thermal neutron flux, axial power distribution and pin power distribution. Also, four physical parameters are studied, Keff, βeff, MTC and FTC. The burnup calculation is performed for VVER-1200. The comparison of the multiplication factor variation with depletion (for 730 days) illustrated that Iridium Oxide in Th–F assembly has a faster rate of depletion than other Burnable absorbers. Using Iridium Oxide shortens the core life time due to the transformation process of Ir-191 to Ir-192. Moreover, the concentration of actinides for the different assemblies at EOC were discussed. In addition, the effect of using BPs in different fuel assemblies on the concentration of 135Xe and 149Sm production are also performed.

Impact of Radial Reflector Fidelity on Neutronics and Vessel Fluence Simulations

The Consortium for Advanced Simulation of Light Water Reactors is developing the Virtual Environment for Reactor Applications (VERA), and the MPACT code, which is the primary deterministic neutron transport solver in VERA, provides sub-pin level flux and power distributions as part of full-scale cycle depletion and analysis. In such calculations, an important aspect is the radial reflector treatment. To improve the fidelity of the radial reflector treatment, MPACT was extended to approximate the modeling of the reactor’s structural components such as the core shroud, barrel, neutron pads, and vessel. This work explores several modeling configurations with varying levels of fidelity and computational burden and assesses the importance of modeling fidelity on the eigenvalue and pin power distribution. Two two-dimensional (2-D) problems were analyzed to assess the impact on eigenvalue and pin power distributions with low-fidelity, coarse square cell reflector representations: (1) a Watts Bar Nuclear Plant Unit 1 (WBN1) quarter-core slice with depletion and (2) an AP1000 quarter-core slice. The analyses showed that the effect on eigenvalue is fairly small, but the effect on pin power is more pronounced, especially locally in the assemblies closest to the periphery, where the maximum pin power difference is nearly 3.5% in the AP1000 case. Two additional 2-D problems were used to assess the comparison between the low-fidelity coarse square cell treatment and a high-fidelity geometric representation that uses subpin material specification: (1) the same WBN1 quarter-core slice and (2) a representative model of the NuScale small modular reactor (SMR), which features a solid reflector design with moderator holes. These results demonstrate that even a coarse, low-fidelity representation adequately captures the necessary simulation characteristics. Last, these capabilities were applied to the 2-D WBN1 quarter-core depletion to assess the impact on vessel fluence using VeraShift. From adjoint calculations, pins along the periphery were observed to be of highest importance for fluence calculation, so the impact of the reflector representation in MPACT could theoretically substantially affect the predicted result. However, it was observed that the change in pin powers along the periphery minimally impacts the maximum vessel fluence with a difference within the statistical uncertainty but provides terrific insight on the sensitivity of the peripheral pins.

Steady-State Thermal-Hydraulic Model for Fluoride-Salt-Cooled Small Modular High-Temperature Reactors

A whole-core, steady-state, thermal-hydraulic model for the cylindrical pin-type fluoride-salt-cooled small modular advanced high-temperature reactor (SmAHTR) is developed. In this preconceptual reactor design initially proposed by Oak Ridge National Laboratory, each fuel assembly in the graphite-moderated core has the FLiBe coolant flowing parallel to a hexagonal array of fuel and moderator pins. The present study considers a slightly modified fuel assembly design with a hexagonal inner housing compared to the original cylindrical housing. Burnable poison pins and control rods are also included in the fuel assembly considered here. The thermal-hydraulic model employs finite volumes to solve three-dimensional conduction in the pins and the hexagonal graphite reflector regions in the core. Heat transfer between the fuel assemblies is also addressed. The finite volumes in the fluid region are modeled using a subchannel approach in which the fluid is discretized into edge, corner, and interior subchannels and the resulting mass, momentum, and energy equations are systematically solved. The subchannel model also includes the transport between adjacent subchannels both due to radial pressure gradient–driven cross flow and turbulent mixing. Appropriate closure models from the literature are used to quantify axial and lateral flow resistances, heat transfer from solid to fluid, and turbulent mixing. The resulting thermal-hydraulic model provides detailed temperature and flow information for the entire core at a modest computational cost. Preliminary verification studies are also performed and reported. Whole-core, steady-state results are presented for this SmAHTR core configuration for different power profiles. The effect of grid refinement and total mass flow rate into the core on the peak fuel temperature is also investigated. Fuel temperatures from a preliminary analysis with pin power distributions from a neutronic model are also included. The peak fuel temperature of ~1229°C in this illustrative case is below the steady-state operation limit for the SmAHTR.

Investigating the Performance and safety features of Pressurized water reactors using the burnable poisons

Investigating the performance and safety features of Pressurized water reactors (PWRs) using the burnable poisons (BPs) are evaluated in this paper. Modeling and simulating of PWR assembly from Vera core physics benchmark are carried out using the MCNPX code. The effect of using different burnable poisons (eight) on the neutronic parameters are studied. Moreover, the types of BPs either as integral burnable absorbers (IBAs) or as burnable poison rods (BPRs) are investigated. The effect of these poisons on the effective multiplication factor is the first parameter that simulated. Gadolinium oxide (Gd2O3) and Erbium oxide (Er2O3) are used as IBAs that mixed homogenously with uranium fuel. The results indicated that Erbium is more effective than Gadolinium due to the fast depletion of gadolinium that shortens the core life. Zirconium diboride (ZrB2), Lutetium Oxide (Lu2O3) and Protactinium Oxide (PaO2) are also investigated as IBAs that are used as coating the outer surface of fuel rod. The results demonstrated that Protactinium is more effective than other poisons. Moreover, the impact of Pyrex (SiO2-B2O3), WABA (Al2O3-B4C) and Hafnium di-boride (HfB2) are included and analyzed. Effective multiplication factor and beta effective are evaluated at the beginning of life of reactor operation. Finally, via the Pin by pin power distribution at BOC and EOC on the assembly level, maximum and minimum power values for the different cases are compared.

Uncertainty analysis of the OECD/NEA LWR UAM benchmark Phase I using CASMO5 with JENDL-4.0 and ENDF/B-VII.1

ABSTRACT An uncertainty analysis of the OECD/NEA LWR UAM benchmark Phase I was performed by CASMO5 using JENDL-4.0 and ENDF/B-VII.1, and the individual influence of nuclide/reaction covariance data was estimated to preliminary discuss the important covariance data on the uncertainties of lattice physics parameters used in core calculations. In the boiling water reactor and pressurized water reactor UO2 fuel assembly problems, the random sampling was used to estimate the uncertainties of lattice physics parameters. The difference in the covariance data between JENDL-4.0 and ENDF/B-VII.1 was discussed by comparing the influence of nuclide/reaction covariance data. In terms of the k-infinity uncertainty, it was confirmed that the large discrepancy between the two nuclear data libraries was caused by the covariance data of 235U [-] (i.e. the magnitude of the 235U nubar (an average number of neutrons per fission) uncertainty), which was one of the dominant covariance data. In addition, even though the covariance data of 238U [elastic-(n,n’)], 90Zr [elastic-elastic], and 157Gd [(n,γ)-(n,γ)] had the large influence on the assembly-homogenized 2-group macroscopic cross section and pin power distribution, those data were lack in JENDL-4.0. Thus, it is important to take into account the discrepancy and lack of such covariance data in future uncertainty analyses.

Development of the triangle-based nodal algorithm for reconstructing pin power distributions

The fuel pin power is an essential parameter for increasing the safety and reliability features of the reactor. Although the high-order triangle-based polynomial expansion nodal (TPEN) algorithm had been suggested for incredible accuracy and computational speed in hexagonal core analyses, the pin power reconstruction of this algorithm is not developed and assessed. This paper presents the characteristics and performances of the TPEN algorithm for pin power reconstruction. The converged nodal information obtained from sweep between coarse-mesh finite difference (CMFD) and TPEN methods is used to reconstruct the neutron flux distribution in a homogeneous fuel assembly. The modulation technique is used to obtain the heterogeneous distribution of power density. In this technique, the power density homogeneous distribution, calculated with the reconstructed neutron flux is multiplied by a form function. These functions are generated by DRAGON5. The results obtained by this algorithm are verified for various core configurations of a VVER-1000 reactor. The pin power factors show good agreement with the reference solution obtained by heterogeneous fine mesh finite element method. The largest and average relative errors found were of the order of 5% and 0.5% for TPEN, in a peripheral cell of a fuel element with the faces towards the region of the baffle/reflector. We also compared the results with those obtained from the nodal expansion method (NEM). The maximum and mean relative error of 10% and 1% are found by the NEM method.

Verification of the 3D capability of OpenMOC with the C5G7 3D extension benchmark

Due to its numerical accuracy, geometrical flexibility, potential parallel degrees, and implementary simplicity, the Method of Characteristics (MOC) is a very promising candidate for high-fidelity full-core transport solving. OpenMOC, a general purpose open source neutron transport solver, was originally a 2D criticality calculation solver, and has been extended to have 3D capability recently. This paper presents the solutions obtained with OpenMOC to the C5G7 3D extension benchmark problems along with the sensitivity analysis of simulation parameters. All three configurations are calculated, and the results are compared with the benchmark reference with respect to both the multiplication factor and the pin power distribution. Good agreements with the reference results were reached, verifying the solution accuracy of the 3D capability of OpenMOC.

Comparison of the performance of heterogeneous variational nodal method and 2D/1D-MOC on pin-resolved neutron transport calculations of PWR

Heterogeneous Variational Nodal Method (HVNM) and the 2D/1D Method of Characteristics (2D/1D-MOC) are two promising methods for high-fidelity neutronics calculation. HVNM starts from even-parity form of neutron transport equation and employs iso-parameteric finite elements, orthogonal polynomials and piece-wise constants for spatial discretization. Additionally, spherical harmonics are applied for the angular expansion. In terms of 2D/1D-MOC, it starts from first-order transport equation and splits the three-dimensional transport equation into one 2D and one 1D equations which are coupled by the radial and axial leakage terms. Then space and angle are discretized by characteristic lines in radial plane and 1D Sn calculation is employed in axial direction. Both HVNM and 2D/1D-MOC have the capability of dealing with pin-resolved geometry and strong transport effect in PWR core problem. Thus, we developed two codes PANX and NECP-X which based on the methods of HVNM and MOC respectively. In this paper, to compare their performance on pin-resolved problems in PWR, the KAIST problem, the NuScale problem and the Beavrs problem are calculated. Both computational accuracy and efficiency are evaluated. The numerical results shown that both PANX and 2D/1D-MOC can obtain accurate eigenvalues with the error less than 100 pcm. However, PANX obtains more accurate pin power distribution and achieves higher computational efficiency than 2D/1D-MOC.

Feasibility Study of Thorium-Plutonium Mixed Oxide Assembly In Light Water Reactors

Thorium-plutonium mixed oxide, (Th,Pu)OX, is currently used as an alternative fuel in the light water reactors in the world. The main objective of this paper is not only to show the benefits of using the thorium, but mainly to study how the way thorium is introduced in the fuel affects the neutron parameters. Among these benefits is the possibility of extending the operating cycle length and the reduction of the increasing stockpiles of plutonium. The first investigated method is introducing thorium as (Th,Pu)OX. The second one is a homogeneous model of thorium plutonium oxide. It is carried out by adding an amount of plutonium separated from the uranium oxide cycle at 50 GWd/ton of heavy metal to the same amount of thorium. Thus, we studied three assemblies; the reference assembly is uranium oxide of 4.2% enrichment containing borated water as a moderator of concentration 500 ppm (part per million) of B-10. The second is a (Th,Pu)OX and the third one is an assembly with homogenized thorium plutonium. All three assemblies are modeled using MCNPX. A comparison is held between the results of the three lattices. The factors compared are the effective multiplication factor, the inventory of plutonium and uranium isotopes, and the depletion of B-10, the pin by pin power distribution at 0 and 60 GWd/ton and the relative pin radial power for the three lattices. The comparison is aimed to show the effect on the cycle length, the reduction in the Pu content and the power flattening across the assembly. It is found that the evolution of the multiplication factors shows a similar behaviour using (Th-Pu)OX fuel in the assembly as UOX cycle inspite of lowering the K-eff of fresh (Th-Pu)OX fuel (1.19847). The power flattening which is favorable in reactor operation is clearer in (Th,Pu)OX fuel. It is noticed that the mass of Pu-239 decreases by 1.07% from its initial value at the end of life. For homogeneous (Th,Pu)OX, the mass decreases by 0.0832%. The high power peaking factor for (Th,Pu)OX is not expected to cause significant effects during reactor operation but it can be reduced by adding burnable poisons.