Engineering topics

Li, Yunzhao

Publications and source records attributed to Li, Yunzhao.

5 records

DOE OSTI2022

Simulation of SPND Beta-Current in AP1000 core

The Self-Powered Neutron Detectors (SPNDs) are widely used in Generation-III commercial PWR due to its small size and external-power free character. It can measure the in-core neutron flux density immediately to obtain the power profile in the core. To better understand the actual state of the simulated core, a SPND beta-current simulation module has been established in the NECP-Bamboo code. Firstly, neutron simulation is carried out in the lattice code Bamboo-Lattice together with nuclide depletion simulation to obtain the depleted composition of the entire lattice including the SPND, and the few-group parameters including density and cross section of emitter material. Secondly, the depleted lattice can then be piped into a Monte-Carlo code for neutron-photon-electron simulation to obtain the electron escape efficiency from the emitter of the SPND to the corresponding collector. Thirdly, the electron escape efficiency can be parameterized into the function of the lattice state parameters similarly to the common few-group constants. Finally, the beta-current was calculated by the core code Bamboo-Core simultaneously with the simulation of the core operation process. The beta-current of the Vanadium SPND in the AP1000 core was simulated and analyzed in this paper. Comparison between simulation and measurement of the beta-current distribution in the specific core state is shown in this paper to verify the effectiveness of the method and the code. The radial normalized current distribution is in good agreement with the measurement. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Heterogeneous Neutron Leakage Model for PWR Pin-by-pin Calculation

In the last few decades, the nodal diffusion calculation accompanied with the assembly homogenization has been widely employed. The single assembly calculation with the reflective boundary condition (BC) would be carried out to determine the assembly-homogenized parameters. Because of the reflective BC, no axial and/or radial leakage rates have been taken into accounted for the lattice calculation. To represent the global influence of the reactor core on an assembly calculation, one mostly frequently uses a procedure called leakage model to approximately take into account the absorption due to the macroscopic structure of the physical flux in a heterogeneous situation. For the assembly homogenization, this absorption is introduced as the product of the leakage coefficient and the buckling obtained for an equivalent homogeneous medium using the well-known homogeneous B1 theory. Clearly, this is an imperfect representation of leakages in a heterogeneous medium. However, the errors introduced by the foregoing procedure are small for the whole? -assembly homogenization in a PWR. When it comes to the pin-by-pin calculation, it is not the case. The leakage rate? will be affected by the neutron streaming effect caused by the heterogeneity of the assembly. Especially when poison pins or empty channels are present, the streaming effect becomes important. The errors introduced appear particularly visible. The heterogeneous leakage model? which assumes that the neutron flux is variational and the leakage rate is not uniform over the lattice pitch is a more appropriate choice. When the heterogeneous leakage model is applied in the lattice calculation for the pin-cell homogenization, the pin-homogenized problem would be changed compared with the original assembly problem without any homogenization technique. So, the Cooperation of the heterogeneous leakage model with the pin-cell homogenization theory is unsettled. In this paper, both homogenous and heterogeneous leakage models are evaluated in pin-by-pin calculation. Their implements used in pin-cell homogenization of the lattice calculation are studied. The diffusion coefficient obtained from transport cross-section and the space-dependent leakage coefficient together with the two leakage models are analyzed for the practical reactor-core applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Pn-FEM Method for Solving the First-Order Neutron Transport Equation

A deterministic method was investigated and developed for the photon-neutron transport calculation for fusion reactor blanket. Firstly, considering the fact that the second order transport equation cannot describe void region accurately due to the appearance of zero total cross section in the denominator, the first order equation was selected as the starting point of the solution. The angular discretization uses the spherical harmonics (Pn) method. Although discrete ordinates (Sn) method has been most commonly applied due to its efficiency, it suffers from the ray effect. To describe unstructured geometry, the finite element method was used to discrete spatial variable. It is well known that Bubnov-Galerkin discretization of the first-order equation is inappropriate which will create oscillations. To obtain robust solutions, Petrov- Galerkin methods have been developed such as the Streamline-Upwind Petrov-Galerkin (SUPG) method and the Generalized Least-Squares (GLS) method. These methods require a stabilization term added to its test functions, which will incur additional computational cost. Another finite element formulation that usually used for first order equation is the Discontinuous-Galerkin (DG) method. In this method, each element has a unique and separate set of basis function. Then the adjacent elements can be connected by using the in-coming flux. Its main drawback is the number of unknowns, much larger than Continuous-Galerkin (CG) method. The Sub-Grid Scale (SGS) finite element method combines the advantages of CG and DG methods together. It inserts the DG method in each element to correct the CG solution, and provides numerically stable solution of first-order equation without increasing the number of unknowns. The boundary conditions and interface conditions for Pn method is a hard nut to crack. We use Riemann method to deal with this problem. It decomposes the neutron angular flux into in-coming and out-going information through an interface, and applies boundary condition to the in-coming one. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

Improved Variable-Transformation Algorithm for the Hexagonal Variational Nodal Method

Hexagonal Variational Nodal Method (VNM) has been widely employed in fast reactor core calculation codes. It provides the weak form solution of the second order transport equation. It employs a direct three-dimensional expansion instead of the transverse integration technique, with the singular term caused by transverse integration over the hexagonal nodal avoided. However, as the increasing Pn expansion order in angle, the number of moments increases dramatically, which significantly increases the requirements in both memory and computational time. As a consequence, Yang proposed the Variable-Transformation Algorithm. It firstly used the reverse Cuthill-McKee (RCMK) reordering method to transform the response matrices into block-diagonal. Then, for the unreducible matrices, orthogonal transformation was applied, after which the matrices were sparse enough to be reduced by using the RCMK reordering once more. In this summary, based on a newly developed hexagonal VNM code NECP-VioletHex, the Variable Transformation Algorithm was investigated and improved by adopting more constraints to the basis functions to further take advantage of the symmetric properties of the hexagonal-z nodes. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

NECP-Cypress: Development and Benchmarking of a 3D PWR Core-Analysis Code

As one of the key components of nuclear reactor design and safety analysis software for Pressurized Water Reactor (PWR), the advance of the core-analysis code indicates the nuclear power development of the corresponding country. Hence, Nuclear Engineering Computational Physics Lab. (NECP) developed a PWR core-analysis code NECP-Cypress. Currently, it contains several modules including a generalized cross section module, a diffusion module using the Variational Nodal Method (VNM), a depletion-calculation using the traditional macroscopic module, a thermal-hydraulics module solving the single channel heat transfer equation and 1D heat conduction equation. With the VNM employed, NECP-Cypress can eliminate the control rod cusping effect and provide a better pin power reconstruction. In this paper, we firstly introduce the modules of the code NECP-Cypress. Secondly, two large-scale PWR core benchmarks were employed to demonstrate the ability of the code by comparing with SIMULATE-3, including the BEAVRS problem and a pseudo CAP1400 reactor core problem. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗