Three-dimensional Core Design of a Super Fast Reactor with a High Power Density

LIANGZHI CAO,YOSHIAKI OKA,YUKI ISHIWATARI,SATOSHI IKEJIRI,HAITAO JU 한국원자력학회 2010 Nuclear Engineering and Technology Vol.42 No.1

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/cm3. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.

Scattering Correction for Image Reconstruction in Flash Radiography

LIANGZHI CAO,Mengqi Wang,Hongchun Wu,Zhouyu Liu,Yuxiong Cheng,Hongbo Zhang 한국원자력학회 2013 Nuclear Engineering and Technology Vol.45 No.4

Scattered photons cause blurring and distortions in flash radiography, reducing the accuracy of image reconstruction significantly. The effect of the scattered photons is taken into account and an iterative deduction of the scattered photons is proposed to amend the scattering effect for image restoration. In order to deduct the scattering contribution, the flux of scattered photons is estimated as the sum of two components. The single scattered component is calculated accurately together with the uncollided flux along the characteristic ray, while the multiple scattered component is evaluated using correction coefficients pre-obtained from Monte Carlo simulations.The arbitrary geometry pretreatment and ray tracing are carried out based on the customization of AutoCAD. With the above model, an Iterative Procedure for image restORation code, IPOR, is developed. Numerical results demonstrate that the IPOR code is much more accurate than the direct reconstruction solution without scattering correction and it has a very high computational efficiency.

SCATTERING CORRECTION FOR IMAGE RECONSTRUCTION IN FLASH RADIOGRAPHY

Cao, Liangzhi,Wang, Mengqi,Wu, Hongchun,Liu, Zhouyu,Cheng, Yuxiong,Zhang, Hongbo Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.4

Scattered photons cause blurring and distortions in flash radiography, reducing the accuracy of image reconstruction significantly. The effect of the scattered photons is taken into account and an iterative deduction of the scattered photons is proposed to amend the scattering effect for image restoration. In order to deduct the scattering contribution, the flux of scattered photons is estimated as the sum of two components. The single scattered component is calculated accurately together with the uncollided flux along the characteristic ray, while the multiple scattered component is evaluated using correction coefficients pre-obtained from Monte Carlo simulations.The arbitrary geometry pretreatment and ray tracing are carried out based on the customization of AutoCAD. With the above model, an Iterative Procedure for image restORation code, IPOR, is developed. Numerical results demonstrate that the IPOR code is much more accurate than the direct reconstruction solution without scattering correction and it has a very high computational efficiency.

THREE-DIMENSIONAL CORE DESIGN OF A SUPER FAST REACTOR WITH A HIGH POWER DENSITY

Cao, Liangzhi,Oka, Yoshiaki,Ishiwatari, Yuki,Ikejiri, Satoshi,Ju, Haitao Korean Nuclear Society 2010 Nuclear Engineering and Technology Vol.42 No.1

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/$cm^3$. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.

Resonance Elastic Scattering and Interference Effects Treatments in Subgroup Method

Li, Yunzhao,He, Qingming,Cao, Liangzhi,Wu, Hongchun,Zu, Tiejun Korean Nuclear Society 2016 Nuclear Engineering and Technology Vol.48 No.2

Based on the resonance integral (RI) tables produced by the NJOY program, the conventional subgroup method usually ignores both the resonance elastic scattering and the resonance interference effects. In this paper, on one hand, to correct the resonance elastic scattering effect, RI tables are regenerated by using the Monte Carlo code, OpenMC, which employs the Doppler broadening rejection correction method for the resonance elastic scattering. On the other hand, a fast resonance interference factor method is proposed to efficiently handle the resonance interference effect. Encouraging conclusions have been indicated by the numerical results. (1) For a hot full power pressurized water reactor fuel pin-cell, an error of about +200 percent mille could be introduced by neglecting the resonance elastic scattering effect. By contrast, the approach employed in this paper can eliminate the error. (2) The fast resonance interference factor method possesses higher precision and higher efficiency than the conventional Bondarenko iteration method. Correspondingly, if the fast resonance interference factor method proposed in this paper is employed, the $k_{inf}$ can be improved by ~100 percent mille with a speedup of about 4.56.

The JFNK method for the PWR's transient simulation considering neutronics, thermal hydraulics and mechanics

He, Qingming,Zhang, Yijun,Liu, Zhouyu,Cao, Liangzhi,Wu, Hongchun Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.2

A new task of using the Jacobian-Free-Newton-Krylov (JFNK) method for the PWR core transient simulations involving neutronics, thermal hydraulics and mechanics is conducted. For the transient scenario of PWR, normally the Picard iteration of the coupled coarse-mesh nodal equations and parallel channel TH equations is performed to get the transient solution. In order to solve the coupled equations faster and more stable, the Newton Krylov (NK) method based on the explicit matrix was studied. However, the NK method is hard to be extended to the cases with more physics phenomenon coupled, thus the JFNK based iteration scheme is developed for the nodal method and parallel-channel TH method. The local gap conductance is sensitive to the gap width and will influence the temperature distribution in the fuel rod significantly. To further consider the local gap conductance during the transient scenario, a 1D mechanics model is coupled into the JFNK scheme to account for the fuel thermal expansion effect. To improve the efficiency, the physics-based precondition and scaling technique are developed for the JFNK iteration. Numerical tests show good convergence behavior of the iterations and demonstrate the influence of the fuel thermal expansion effect during the rod ejection problems.

Homogenized cross-section generation for pebble-bed type high-temperature gas-cooled reactor using NECP-MCX

Qin Shuai,Li Yunzhao,He Qingming,Cao Liangzhi,Wang Yongping,Wu Yuxuan,Wu Hongchun 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.9

In the two-step analysis of Pebble-Bed type High-Temperature Gas-Cooled Reactor (PB-HTGR), the lattice physics calculation for the generation of homogenized cross-sections is based on the fuel pebble. However, the randomly-dispersed fuel particles in the fuel pebble introduce double heterogeneity and randomness. Compared to the deterministic method, the Monte Carlo method which is flexible in geometry modeling provides a high-fidelity treatment. Therefore, the Monte Carlo code NECP-MCX is extended in this study to perform the lattice physics calculation of the PB-HTGR. Firstly, the capability for the simulation of randomly-dispersed media, using the explicit modeling approach, is developed in NECP-MCX. Secondly, the capability for the generation of the homogenized cross-section is also developed in NECP-MCX. Finally, simplified PB-HTGR problems are calculated by a two-step neutronics analysis tool based on Monte Carlo homogenization. For the pebble beds mixed by fuel pebble and graphite pebble, the bias is less than 100 pcm when compared to the high-fidelity model, and the bias is increased to 269 pcm for pebble bed mixed by depleted fuel pebble. Numerical results show that the Monte Carlo lattice physics calculation for the two-step analysis of PB-HTGR is feasible.

Calculation of thermal neutron scattering data of MgF2 and its effect on beam shaping assembly for BNCT

Hu Jiaqi,Qiao Zhaopeng,Fan Lunhe,Tang Yongqiang,Cao Liangzhi,Zu Tiejun,He Qingming,Li Zhifeng,Wang Sheng 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.4

MgF2 as a moderator material has been extensively used in the beam shaping assembly (BSA) that plays an important role in the boron neutron capture therapy (BNCT). Regarded as important for applications, the thermal neutron scattering data of MgF2 were calculated, based on the phonon expansion model. The structural properties of MgF2 were researched by the VASP code based on the ab-initio methods. The PHONOPY code was employed to calculate the phonon density of states. Furthermore, the NJOY code was used to calculate the thermal neutron scattering data of MgF2. The calculated inelastic cross sections plus absorption cross sections are in agreement with the available experimental data. The neutron transport in the BSA has been simulated by using a hybrid Monte-Carlo-Deterministic code NECP-MCX. The results indicated that compared with the calculation of the free gas model, the thermal neutron flux and epithermal neutron flux at the BSA exit port calculated by using the thermal neutron scattering data of MgF2 were reduced by 27.7% and 8.2%, respectively.