A new approach to three-dimensional neutron transport solution based on the method of characteristics and linear axial approximation

Zheng, Youqi,Choi, Sooyoung,Lee, Deokjung Elsevier 2017 Journal of computational physics Vol.350 No.-

<P><B>Abstract</B></P> <P>A new approach based on the method of characteristics (MOC) is proposed to solve the neutron transport equation. A new three-dimensional (3D) spatial discretization is applied to avoid the instability issue of the transverse leakage iteration of the traditional 2D/1D approach. In this new approach, the axial and radial variables are discretized in two different ways: the linear expansion is performed in the axial direction, then, the 3D solution of the angular flux is transformed to be the planar solution of 2D angular expansion moments, which are solved by the planar MOC sweeping. Based on the boundary and interface continuity conditions, the 2D expansion moment solution is equivalently transformed to be the solution of the axially averaged angular flux. Using the piecewise averaged angular flux at the top and bottom surfaces of 3D meshes, the planes are coupled to give the 3D angular flux distribution. The 3D CMFD linear system is established from the surface net current of every 3D pin-mesh to accelerate the convergence of power iteration. The STREAM code is extended to be capable of handling 3D problems based on the new approach. Several benchmarks are tested to verify its feasibility and accuracy, including the 3D homogeneous benchmarks and heterogeneous benchmarks. The computational sensitivity is discussed. The results show good accuracy in all tests. With the CMFD acceleration, the convergence is stable. In addition, a pin-cell problem with void gap is calculated. This shows the advantage compared to the traditional 2D/1D MOC methods.</P>

Comparisons of S_N and Monte-Carlo methods in PWR ex-core detector response simulation

Zheng, Youqi,Lee, Deokjung,Zhang, Peng,Lee, Eunki,Shin, Ho-cheol Elsevier 2017 Annals of nuclear energy Vol.101 No.-

<P><B>Abstract</B></P> <P>The ex-core detector response calculation is an important part in reactor design. However, the response function cannot be measured by experiments quantitatively. Ex-core detector response simulation is therefore required. For decades, the S<SUB>N</SUB> code has been used as the dedicated tool. Nowadays, more and more engineers are expressing an interest in using the Monte-Carlo method instead of the S<SUB>N</SUB> method in simulations, as it is expected that the Monte-Carlo method will give higher accuracy. In this paper, the modeling and simulation of ex-core detector responses is briefly reviewed based on the Korean Kori Unit 1 reactor. Then, the differences between the S<SUB>N</SUB> simulation and Monte-Carlo simulation are compared. The sensitivity of computational conditions is also discussed. It is shown that the problem dependence of cross sections and meshing dependence of spatial discretization in the ex-core detector response calculations are not as strong as expected. However, the ray effect is the main shortcoming for the S<SUB>N</SUB> calculation. Based on the analysis, two benefits are shown by using MCNP for the direct 3D calculation. Firstly, the impact of ray effect is eliminated without using the S<SUB>N</SUB> angular discretization. Secondly, the direct 3D calculation is easier to perform based on the powerful ability of 3D modeling and parallel computing of the Monte-Carlo code. The new DRF values are adopted in the dynamic control rod reactivity measurement of Kori Unit 1 reactor. The results show that the new DRF values improve the error of measured control rod worth by a percentage of 3.</P> <P><B>Highlights</B></P> <P> <UL> <LI> The impact of ray effect and synthesis approximation is analyzed for the S_N code. </LI> <LI> The measured data is applied to prove the benefit obtained from MCNP 3D calculation. </LI> </UL> </P>

Effect of high-energy neutron source on predicting the proton beam current in the ADS design

Youqi Zheng,Xunzhao Li,,Hongchun Wu 한국원자력학회 2017 Nuclear Engineering and Technology Vol.49 No.8

The accelerator-driven subcritical system (ADS) is driven by a neutron source from spallation reactionsintroduced by the injected proton beam. Part of the neutron source has energy as high as a few hundredMeV to a few GeV. The effects of high-energy source neutrons (En > 20 MeV) are usually approximated byenergy cut-off treatment in practical core calculations, which can overestimate the predicted protonbeam current in the ADS design. This article intends to quantize this effect and propose a way to solvethis problem. To evaluate the effects of high-energy neutrons in the subcritical core, two models areestablished aiming to cover the features of current experimental facilities and industrial-scale ADS in thefuture. The results show that high-energy neutrons with En > 20 MeV are of small fraction (2.6%) in theneutron source, but their contribution to the source efficiency is about 23% for the large scale ADS. Basedon this, a neutron source efficiency correction factor is proposed. Tests show that the new correctionmethod works well in the ADS calculation. This method can effectively improve the accuracy of theprediction of the proton beam current.

Verification of SARAX code system in the reactor core transient calculation based on the simplifi ed EBRII benchmark

Xiaoqian Jia,Youqi Zheng,Xianna Du,Yongping Wang,Jianda Chen 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.5

This paper shows the verification work of SARAX code system in the reactor core transient calculationbased on the simplified EBR-II Benchmark. The SARAX code system is an analysis package developed byXi'an Jiaotong University and aims at the advanced reactor R&D. In this work, a neutron-photon coupledpower calculation model and a spatial-dependent reactivity feedback model were introduced. To verifythe models used in SARAX, the EBR-II SHRT-45R test was simplified to an ULOF transient with an inputflowrate change curve by fitting from reference. With the neutron-photon coupled power calculationmodel, SARAX gave close results in both power fraction and peak power prediction to the referenceresults. The location of the hottest assembly from SARAX and reference are the same and the relativepower deviation of the hottest assembly is 2.6%. As for transient analysis, compared with experimentalresults and other calculated results, SARAX presents coincident results both in trend and absolute value. The minimum value of core net reactivity during the transient agreed well with the reported results,which ranged from 0.3$ to 0.35$. The results verify the models in SARAX, which are correct and ableto simulate the in-core transient with reliable accuracy.

Conceptual design of a MW heat pipe reactor

Wu Yunqin,Zheng Youqi,Chen Qichang,Li Jinming,Du Xianan,Wang Yongping,Tao Yushan 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.3

In recent years, unmanned underwater vehicles (UUV) have been vigorously developed, and with the continuous deepening of marine exploration, traditional energy can no longer meet the energy supply. Nuclear energy can achieve a huge and sustainable energy supply. The heat pipe reactor has no flow system and related auxiliary systems, and the supporting mechanical moving parts are greatly reduced, the noise is relatively small, and the system is simpler and more reliable. It is more favorable for the control of unmanned systems. The use of heat pipe reactors in unmanned underwater vehicles can meet the needs for highly compact, long-life, unmanned, highly reliable, ultra-quiet power supplies. In this paper, a heat pipe reactor scheme named UPR-S that can be applied to unmanned underwater vehicles is designed. The reactor core can provide 1 MW of thermal power, and it can operate at full power for 5 years. UPR-S has negative reactive feedback, it has inherent safety. The temperature and stress of the reactor are within the limits of the material, and the core safety can still be guaranteed when the two heat pipes are failed.

On the equivalence of reaction rate in energy collapsing of fast reactor code SARAX

Xiao, Bowen,Wei, Linfang,Zheng, Youqi,Zhang, Bin,Wu, Hongchun Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.3

Scattering resonance of medium mass nuclides leads complex spectrum in the fast reactor, which requires thousands of energy groups in the spectrum calculation. When the broad-group cross sections are collapsed, reaction rate cannot be completely conserved. To eliminate the error from energy collapsing, the Super-homogenization method in energy collapsing (ESPH) was employed in the fast reactor code SARAX. An ESPH factor was derived based on the ESPH-corrected SN transport equation. By applying the factor in problems with reflective boundary condition, both the effective multiplication factor and reaction rate were conserved. The fixed-source iteration was used to ensure the stability of ESPH iteration. However, in the energy collapsing process of SARAX, the vacuum boundary condition was adopted, which was necessary for fast reactors with strong heterogeneity. To further reduce the error caused by leakage, an additional conservation factor was proposed to correct the neutron current in energy collapsing. To evaluate the performance of ESPH with conservation factor, numerical benchmarks of fast reactors were calculated. The results of broad-group calculation agreed well with the direct full-core Monte-Carlo calculation, including the effective multiplication factor, radial power distribution, total control rod worth and sodium void worth.

The applicability study and validation of TULIP code for full energy range spectrum

Chen Wenjie,Du Xianan,Wang Rong,Zheng Youqi,Wang Yongping,Wu Hongchun 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.12

NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi’an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutronmoderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system