Pressurized Water Reactor Assembly Research Articles

Development and application of the added fluid mass and substructure techniques for integrated pressurized water reactor assembly

The SMART (system-integrated advanced modular reactor), one of iPWRs (integrated pressurized water reactors), has been on the process of design in Republic of Korea. For the structural integrity of the 330MWt SMART reactor assembly against earthquake, this paper presents FE (finite element) models, and dynamic and seismic analysis results. Prior to the dynamic and seismic analysis for the iPWR assembly, elementary techniques such as added fluid mass and substructure techniques were developed for simplified models used in this work. For the added fluid mass techniques, it is found that the use of the fluid element is more realistic and rational for the fluid effect on dynamic behavior than the addition of the lumped mass. For the substructure techniques using super-elements, it is found that these techniques can avoid a long calculation time and reduce the data storage volume. The dynamic analysis was implemented for the iPWR assembly using the aforementioned elementary techniques. As a result of the dynamic analysis, natural frequency at each mode is reduced with the consideration of the fluid mass and the super-element generates the very same natural frequencies as the normal element model generates. The response spectrum and time history-based seismic analyses were performed for the iPWR assembly using the elementary techniques. It is concluded that the iPWR assembly was designed with sufficient seismic safety margins.

RANS modeling for flow in nuclear fuel bundle in pressurized water reactors (PWR)

This paper presents use of Reynolds-averaged Navier–Stokes (RANS) based turbulence model for single-phase CFD analysis of flow in pressurized water reactor (PWR) assemblies. An open source code called OpenFoam was used for computational fluid dynamics (CFD) study using computational meshes generated using Shari Harpoon. The PWR assembly design used in this analysis represents a 5 × 5 pin design including structural grid equipped with mixing vanes. The design specifications used in this study were obtained from the experimental setup at Texas A&M University and the results obtained are used to validate the CFD software, algorithm, and the turbulence model used in this analysis.

Neutronic and burn-up calculations of heterogeneous Thorium/Uranium fuel in pressurized water reactors

Abstract The purpose of this work is to study the feasibility of the Thorium/Uranium fuel cycle in heterogeneous Pressurized Water Reactors (PWR) core design. This paper focuses on the neutronic and burn-up analysis of the Thorium/Uranium fuel using the computer codes MCNPX and WIMS. The design is based on the Whole Assembly Seed and Blanket (WASB) concept, in which the individual seed (Uranium) and blanket (Thorium-Uranium) units occupy one full-size PWR assembly in a checkerboard core configuration. The results of the present models were compared with the solution of benchmark problems and satisfactory agreement was found.

Time-Spectral Analysis Methods for Spent Fuel Assay Using Lead Slowing-Down Spectroscopy

Nondestructive techniques for measuring the mass of fissile isotopes in spent nuclear fuel is a considerable challenge in the safeguarding of nuclear fuel cycles. A nondestructive assay technology that could provide direct measurement of fissile mass, particularly for the plutonium (Pu) isotopes, and improve upon the uncertainty of today's confirmatory methods is needed. Lead slowing-down spectroscopy (LSDS) has been studied for the spent fuel application previously, but the nonlinear effects of assembly self shielding (of the interrogating neutron population) have led to discouraging assay accuracy for realistic pressurized water reactor fuels. In this paper, we describe the development of time-spectral analysis algorithms for LSDS intended to overcome these self-shielding effects. The algorithm incorporates the tabulated energy-dependent cross sections from key fissile and absorbing isotopes, but leaves their mass as free variables. Multi-parameter regression analysis is then used to directly calculate not only the mass of fissile isotopes in the fuel assembly (e.g., Pu-239, U-235, and Pu-241), but also the mass of key absorbing isotopes such as Pu-240 and U-238. Modeling-based assay results using this self-shielding relationship indicate that LSDS has the potential to directly measure fissile isotopes with less than 5% average relative error for pressurized water reactor assemblies with burnup as high as 60 GWd/MTU. Shortcomings in the initial self-shielding model and potential improvements to the formulation are described.

Actinide Transmutation in PWRs Using CONFU Assemblies

Expansion of domestic use of nuclear power to provide energy security and environmental sustainability requires minimization of the nuclear waste. To achieve this goal in the short term, transmutation of transuranic (TRU) elements in COmbined Non-Fertile and UO2 (CONFU) Generation-III pressurized water reactor (PWR) assemblies is evaluated. These assemblies are composed of a mix of standard UO2 fuel pins and pins made of recycled TRU in an inert matrix and are designed to fit in currently deployed PWRs. Previous studies have shown the feasibility of a CONFU-Equilibrium (CONFU-E) assembly design with a net TRU balance between production and destruction and a CONFU-Burndown (CONFU-B) assembly design with net destruction of TRU coming from several reactors. In this paper, a CONFU-self-Contained (CONFU-C) assembly is shown to achieve net TRU destruction in a self-contained TRU multirecycling system. Both the CONFU-B and CONFU-C designs are presented in this paper in detail.For these designs a detailed assembly-level neutronic analysis has been performed using CASMO-4 to investigate cycle length, TRU management performance, and key reactor reactivity parameters, along with detailed intraassembly power peaking factors (IAPPFs). Various fuel mixing schemes and cooling times were evaluated. Using the IAPPF results, a full core thermal-hydraulic analysis using VIPRE was performed to validate thermal margins, and a loss-of-coolant-accident event was assessed using RELAP5. Based on the TRU management characteristics of these designs, metrics were developed to reflect the material handling difficulties of the multirecycled fuel, along with its repository impact. These parameters were compared to a standard once-through UO2 cycle, along with other Pu or TRU multirecycling schemes [mixed oxide with enriched uranium (MOX-UE) and COmbustible Recyclage A ILot (CORAIL)]. Finally, an economic analysis has been conducted to compare the fuel cycle cost (FCC) associated with these designs.TRU management results of CONFU-B and CONFU-C showed a net TRU destruction of 2 to 20 kg/TW·h(electric) generated, with an FCC of 12 to 15 mills/kW·h(electric), depending on the mixing strategy and cooling time chosen. Reactor control parameters and thermal margins were found to be comparable to an all-UO2 assembly. While both designs offer significant repository benefits, the accumulation of minor actinides may limit the practicality of fuel multirecycling.

Acceleration of Response Matrix Method Using Cross-Section Scaling

In this paper, an acceleration scheme for the red-black response matrix iteration is proposed. The proposed method is easily applied not only to newly developed codes but also to existing ones; cross-section sets are input by multiplying by a scaling factor, without requiring any code modification. The proposed method is called the cross-section scaling acceleration (CSA) method and is applicable to accelerate inner iteration of the response matrix calculation of second-order partial differential equations (e.g., diffusion, simplified PN, and PN). An eigenvalue analysis of the proposed method was carried out for one-group homogeneous problems. The analysis showed that the maximum eigenvalue of the red-black response matrix strongly depends on the scaling factor, and that the convergence of iteration becomes faster when an appropriate scaling factor is used. In the derivation of the response matrix, it was found that the CSA method is viewed as an alternative form of the acceleration method proposed by Lewis and Palmiotti. Although their method requires modifications of the response matrix, application of the CSA method is much easier. The CSA method was used for three test problems that cover a wide range of applications: a simple one-group, one-dimensional problem; a multigroup pressurized water reactor (PWR) assembly problem; and a more realistic multigroup PWR quarter-core problem. The calculation results of the test problems showed that the number of iterations can be reduced from 30 to 80% by utilizing the CSA method.

A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy

For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 × 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

Calculated Actinide and Fission Product Concentration Ratios for Gaseous Effluent Monitoring Using Monteburns 3.01

Techniques have been developed at Los Alamos National Laboratory for accurately calculating certain spent-fuel isotope concentration ratios for pressurized water reactor assemblies using a linked MCNP/ORIGEN2 code named Monteburns 3.01, without resorting to an assembly or full-core calculation. The effects of various fuel parameters such as the number of radial fuel regions per pin, burnup step size, reactor power, reactivity control mechanisms, and axial profiles have been studied. The significance of each factor was determined. A method was also proposed for calculating spent-fuel inventories as a function of burnup for a wide range of reactors and fuel types. It was determined that accurate calculations can be obtained using a three-dimensional, modified pin cell with seven radial fuel regions and two (flat-flux) axial fuel regions calculated with 2000 MWd/tonne U burnup steps for burnups ranging from 0 to 50 000 MWd/tonne U. The calculational technique was benchmarked to measured values from the Calvert Cliffs Unit 1 reactor, and good agreement from the point of view of calibrating a monitoring instrument was found for most cases.

Transport Synthetic Acceleration for Long-Characteristics Assembly-Level Transport Problems

We apply the transport synthetic acceleration (TSA) scheme to the long-characteristics spatial discretization for the two-dimensional assembly-level transport problem. This synthetic method employs a simplified transport operator as its low-order approximation. Thus, in the acceleration step, we take advantage of features of the long-characteristics discretization that make it particularly well suited to assembly-level transport problems. Our main contribution is to address difficulties unique to the long-characteristics discretization and produce a computationally efficient acceleration scheme.The combination of the long-characteristics discretization, opposing reflecting boundary conditions (which are present in assembly-level transport problems), and TSA presents several challenges. We devise methods for overcoming each of them in a computationally efficient way. Since the boundary angular data exist on different grids in the high- and low-order problems, we define restriction and prolongation operations specific to the method of long characteristics to map between the two grids. We implement the conjugate gradient (CG) method in the presence of opposing reflection boundary conditions to solve the TSA low-order equations. The CG iteration may be applied only to symmetric positive definite (SPD) matrices; we prove that the long-characteristics discretization yields an SPD matrix. We present results of our acceleration scheme on a simple test problem, a typical pressurized water reactor assembly, and a typical boiling water reactor assembly.

A New Self-Shielding Method Based on a Detailed Cross-Section Representation in the Resolved Energy Domain

The calculation of a dilution cross section {bar {sigma}}{sub e} is the most important step in the self-shielding formalism based on the equivalence principle. If a dilution cross section that accurately characterizes the physical situation can be calculated, it can then be used for calculating the effective resonance integrals and obtaining accurate self-shielded cross sections. A new technique for the calculation of equivalent cross sections based on the formalism of Riemann integration in the resolved energy domain is proposed. This new method is compared to the generalized Stamm`ler method, which is also based on an equivalence principle, for a two-region cylindrical cell and for a small pressurized water reactor assembly in two dimensions. The accuracy of each computing approach is obtained using reference results obtained from a fine-group slowing-down code named CESCOL. It is shown that the proposed method leads to slightly better performance than the generalized Stamm`ler approach.

A Quasi-Isotropic Reflecting Boundary Condition for the TIBERE Heterogeneous Leakage Model

The influence of assembly or cell heterogeneity on neutron leakage has been consistently taken into account in the TIBERE simplified heterogeneous B1 model. The assumption adopted within the TIBERE model that neutrons are specularly reflected on the boundary introduces two problems. Calculations with this model may become rather time consuming and even unnecessarily long in the case of a Canada deuterium uranium reactor cell, and the peripheral or total coolant voiding of a pressurized water reactor assembly leads to infinite leakage coefficients. These problems have been overcome by the development of another simplified heterogeneous B1 leakage model, TIBERE-2, which has quasi-isotropic reflecting boundary conditions. The TIBERE-2 model uses similar approximations as the TIBERE model and yields an iterative scheme to simultaneously compute multigroup scalar fluxes and directional currents in a heterogeneous geometry. These values enable the evaluation of directional space-dependent leakage coefficients. This new model requires the classical and directional escape and transmission probabilities in addition to the classical and directional first-flight collision probabilities calculated for an open assembly. The TIBERE-2 model has been introduced for general two-dimensional geometry into the DRAGON multigroup transport code. The numerical results obtained by DRAGON show that the TIBERE-2 model represents leakages much better than the homogeneous B1 leakage model. Moreover, the TIBERE-2 model yields results that are extremely close to those obtained by the TIBERE model with considerably shorter computing times.

Subspace Iteration for Nonsymmetric Eigenvalue Problems Applied to the λ-Eigenvalue Problem

Often in reactor dynamics, higher eigenfunctions of the multigroup diffusion equation must be determined. An algorithm to calculate higher eigenfunctions (modes) of the [lambda]-eigenvalue problem corresponding to the steady-state two-group neutron diffusion equation is presented. The method is based on a special type of subspace iteration for large sparse nonsymmetric eigenvalue problems. Having been tested using an International Atomic Energy Agency benchmark problem and also applied to a VVER-1000 pressurized water reactor assembly, the algorithm was found to work very effectively and reliably. In its application, the algorithm presented is not restricted to the [lambda]-eigenvalue problem only but is also generally applicable to large sparse nonsymmetric eigenvalue problems even with multiple and complex eigenvalues.

A Consistent Technique for the Pin-by-Pin Homogenization of a Pressurized Water Reactor Assembly

Proposals are made for improving the heterogeneous diffusion procedure for reactor design and operating calculations. The procedure is based on the use of pin-by-pin properties for the assemblies a...

A Consistent Technique for the Global Homogenization of a Pressurized Water Reactor Assembly

A Consistent Technique for the Global Homogenization of a Pressurized Water Reactor Assembly

Generalization of the Stamm’ler Method for the Self-Shielding of Resonant Isotopes in Arbitrary Geometries

The self-shielding treatment of resonant isotopes is currently performed in most lattice codes using the Stamm’ler method on simplified one-dimensional geometries. A generalization of this procedure is proposed for self-shielding calculations over the arbitrary two- and three-dimensional geometries typical of most advanced reactor designs. Numerical results are presented for a simple two-region cylindrical cell and for a small pressurized water reactor assembly exhibiting true two-dimensional behavior. The absorption rates obtained after self-shielding are compared with exact values obtained using an elastic slowing-down calculation where each resonance is modeled individually in the resolved energy domain. It is shown that the generalized Stamm’ler method can be applied without loss of accuracy to multidimensional domains.

The Effect of Boron and Gadolinium Burnable Poisons on the Hot-to-Cold Reactivity Swing of a Pressurized Water Reactor Assembly

Control requirements for advanced pressurized water reactor designs must be met with heavy loadings of burnable poison rods, the required reactivity hold-down typically amounting to 30% or more in ...

Refinement of the Substructure Method for Integral Transport Calculations

A new generalization of the interface-current method for coupling two-dimensional heterogeneous assemblies, called substructures, has been developed. The method has been designed for fine-structure...