Yixue Chen

Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs), which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

Development and Validation of an Interactive Efficient Dose Rates Distribution Calculation Program Arshield for Visualization of Radiation Field in Nuclear Power Plants

Point kernel integration (PKI) method is widely used in the visualization of radiation field in engineering applications because of the features of quickly dealing with large-scale complicated geometry space problems. But the traditional PKI programs have a lot of restrictions, such as complicated modeling, complicated source setting, 3Dxa0fine mesh results statistics and large-scale computing efficiency. To break the traditional restrictions for visualization of radiation field, ARShield was developed successfully. The results show that ARShield can deal with complicated plant radiation shielding problems for visualization of radiation field. Compared with SuperMC and QAD, it can be seen that the program is reliable and efficient. Also, ARShield can meet the demands of calculation speediness and interactive operations of modeling and displaying 3D geometries on a graphical user interface, avoiding error modeling in calculation and visualization.

Evaluation of ACPs in China Fusion Engineering Test Reactor Using CATE 2.1 Code

Activated corrosion products (ACPs) are the dominant radiation hazard in water-cooled fusion reactor under normal operation conditions and directly determine the occupational radiation exposure during operation and maintenance. Recently, the preliminary design of China Fusion Engineering Test Reactor (CFETR) has been just completed. Evaluation of ACPs is an important work for the safety of CFETR. In this paper, the ACPs analysis code CATE 2.1 was used to simulate the spatial distribution of ACPs along the blanket cooling loop of CFETR, in which the influence of adopting different pulse handling methods was researched. At last, the dose rate caused by ACPs around the blanket cooling loop was calculated using the point kernel code ARShield. The results showed that the dose rate under normal operation for 1.2 years at contact is 1.02u2009mSv/h and at 1u2009m away from pipe is 0.45u2009mSv/h. And after shutting down the reactor, there will be a rapid decrease of dose rate, because of the rapid decay of short-lived ACPs.

ARES: A Parallel Discrete Ordinates Transport Code for Radiation Shielding Applications and Reactor Physics Analysis

ARES is a multidimensional parallel discrete ordinates particle transport code with arbitrary order anisotropic scattering. It can be applied to a wide variety of radiation shielding calculations and reactor physics analysis. ARES uses state-of-the-art solution methods to obtain accurate solutions to the linear Boltzmann transport equation. A multigroup discretization is applied in energy. The code allows multiple spatial discretization schemes and solution methodologies. ARES currently provides diamond difference with or without linear-zero flux fixup, theta weighted, directional theta weighted, exponential directional weighted, and linear discontinuous finite element spatial differencing schemes. Discrete ordinates differencing in angle and spherical harmonics expansion of the scattering source are adopted. First collision source method is used to eliminate or mitigate the ray effects. Traditional source iteration and Krylov iterative method preconditioned with diffusion synthetic acceleration are applied to solve the linear system of equations. ARES uses the Koch-Baker-Alcouffe parallel sweep algorithm to obtain high parallel efficiency. Verification and validation for the ARES transport code system have been done by lots of benchmarks. In this paper, ARES solutions to the HBR-2 benchmark and C5G7 benchmarks are in excellent agreement with published results. Numerical results are presented which demonstrate the accuracy and efficiency of these methods.

Verification for Ray Effects Elimination Module of Radiation Shielding Code ARES by Kobayashi Benchmarks

ARES is a multi-group of anisotropic scattering transport shielding code based on discrete ordinates method. The 3D radiation transport benchmark problems proposed by Kobayashi were calculated by ARES with sub-module ARES_RayEffect which using first collision method for ray effects mitigation. ARES_RayEffect calculates uncollided flux and first collision source moments for ARES. The uncollided flux is obtained by a ray tracing calculation between a source point and a target mesh center. In addition, ARES_RayEffect has a modifying factor function to improve the quality of uncollided flux calculation. For verification, the results of MCNP code are used as reference solution and the results of TORT with FNSUNCL3 are compared. ARES_RayEffect introduced the modifying factor to reduce the relative difference of meshes near the source region. For example, at the position (15,15,15) in Problem 1 case i, the relative difference of the result of ARES with ARES_RayEffect is −2.34%, while relative difference of the result of TORT with FNSUNCL3 is −11.92%. The calculated total neutron fluxes show good agreement with the MCNP solutions. For the pure absorber cases, the maximum differences are less than 3%. For the half scattering cases, the maximum differences are less than 11%. Numerical results demonstrate that ray effects can be effectively mitigated.Copyright

The high-energy multi-group HEST1.0 library based on ENDF/B-VII.0: development, verification and preliminary application

ENDF/B-VII.0, which was released by the USA Cross Section Evaluation Working Group (CSEWG) in December 2006, was demonstrated to perform much better than previous ENDF evaluations over a broad range of benchmark experiments. A high-energy (up to 150 MeV) multi-group library set named HEST1.0 with 253-neutron and 48-photon groups has been developed based on ENDF/B-VII.0 using the NJOY code. This paper provides a summary of the procedure to produce the library set and a detailed description of the verification of the multi-group library set by several shielding benchmark devices, in particular for high-energy neutron data. In addition, the first application of HEST1.0 to the shielding design of the China Spallation Neutron Source (CSNS) is demonstrated.

Analysis of Spatial Discretization Error Estimators Implemented in ARES Transport Code for Equations

The discrete ordinates method (SN) is one of the mainstream methods for neutral particle transport calculations. Assessing the quality of the numerical solution and controlling the discrete error are essential parts of large-scale high-fidelity simulations of nuclear systems. Three error estimators, a two-mesh estimator, a residual-based estimator, and a dual-weighted residual estimator, are derived and implemented in the ARES transport code to evaluate the error of zeroth-order spatial discretization for SN equations. The difference in scalar fluxes on coarse and fine meshes is adopted to indicate the error in the two-mesh method. To avoid zero residual in zeroth-order discretization, angular fluxes within one cell are reconstructed by Legendre polynomials. The error is estimated by inverting the discrete transport operator using the estimated directional residual as an anisotropic source. The inner product of the forward directional residual and the adjoint angular flux is employed to quantify the error in quantities of interest which can be denoted by a linear functional of forward angular flux. Method of Manufactured Solutions (MMS) is adopted to generate analytical solutions for SN equation with scattering and the determined true error is used to evaluate the effectivity of these estimators. Promising results are obtained in the numerical results for both homogeneous and heterogeneous cases. The larger error region is well captured and the average effectivity index for the local error estimation is less than unity. For the series test problems, the estimated goal quantity error can be contained within an order of magnitude around the exact error.

Goal-Oriented Regional Angular Adaptive Algorithm for the SN Equations

Abstract Angular discretization errors inherent in the discrete ordinates method are a major problem, especially for localized source problems and problems with strongly absorbing media or large-volume void regions, where angular discretization errors would be totally unacceptable. This paper proposes a regional angular adaptive algorithm together with a goal-oriented error estimate to solve the SN equations. Standard angular adaptive refinement techniques are based on estimated local errors. We compare an interpolated angular flux value against a calculated value to generate local errors. The adaptive quadrature sets can be created by subdividing a spherical quadrilateral into four spherical subquadrilaterals that have positive weights and can be locally refined. Techniques for mapping angular fluxes from one quadrature set to another are developed to transfer angular fluxes on the interfaces of different spatial regions. To provide a better detector response, local errors are weighted by the importance of a given angular region toward the computational goal, providing an appropriate goal-oriented angular adaptivity. First collision source methods are employed to improve adjoint flux calculation accuracy. We tested the performance and accuracy of the proposed goal-oriented regional angular adaptive algorithm within the ARES code for a number of benchmark problems and present the results of a one-region test model and the Kobayashi benchmark problems. The reduction of angular number is at least one order of magnitude for adaptive refinement. The benchmarks demonstrate that the proposed goal-oriented adaptive refinement can achieve the same level of accuracy as the SN method, which has significantly higher computation cost. Thus, adaptive refinement is a viable approach for investigating difficult particle radiation transport problems.

Analyses of the TIARA 43 MeV Proton Benchmark Shielding Experiments Using the ARES Transport Code

ARES is a multidimensional parallel discrete ordinates particle transport code with arbitrary order anisotropic scattering. It can be applied to a wide variety of radiation shielding calculations and reactor physics analysis. To validate the applicability of the code to accelerator shielding problems, ARES is adopted to simulate a series of accelerator shielding experiments for 43u2009MeV proton-7Li quasi-monoenergetic neutrons, which is performed at Takasaki Ion Accelerator for Advanced Radiation Application. These experiments on iron and concrete were analyzed using the ARES code with FENDL/MG-3.0 multigroup libraries and compared to direct measurements from the BC501A detector. The simulations show good agreement with the experimental data. The ratios of calculated values to experimental data for integrated neutron flux at peak and continuum energy regions are within 64% and 25% discrepancy for the concrete and iron experiments, respectively. The results demonstrate the accuracy and efficiency of ARES code for accelerator shielding calculation.

Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence ( u2009MeV or u2009MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4u2009cm meshes in plane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1u2009cm meshes in plane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.