Contribution of external neutron sources in excess neutron generation for SCNES

Self-Consistent Nuclear Energy System simultaneously meets four requirements: energy generation, fuel breeding, burning of radionuclides and system safety. The key element of the system is excess neutron generation. Analysis of various neutron sources is done with respect to transmutation requirements. Impact of neutron source on energy system performance is analyzed in terms of excess neutron cost. Special emphasis is made on Fusion Neutron Source.     

Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.     

On the calculation of flux in slightly subcritical reactors with external neutron sources

The behaviour of the flux inside a subcritical reactor in the presence of external neutron sources is examined. It is shown, in particular, that the flux can be approximated by the flux resulting from eigenvalue calculation as the reactor approaches its critical state. A method based on the perturbation technique is described, allowing an estimation of spatial effects on the flux by the local sources.     

用综合核方法求解中子输运临界问题的误差分析

基于中子积分输运方程的综合核近似方法,具有准确、快速的特点,其计算精度和收敛性与求积组的选取密切相关.文章简要介绍了求解中子输运临界问题的综合核方法,采用数值方法分析了综合核近似的计算误差和收敛性,并提出了新的求积组来提高综合核方法的计算精度.应用综合核方法计算了均匀平板介质中各向同性和线性各向异性散射的单群、双群中子临界问题,并与离散纵标法S32结果和文献结果进行了比较.计算结果表明采用合适的求积组,综合核方法在低阶时能够得到较高精度的结果.     

Effects of neutron spectrum and external neutron source on neutron multiplication parameters in accelerator-driven system

The neutron multiplication parameters: neutron multiplication M, subcritical multiplication factor k_s, external source efficiency φ^*, play an important role for numerical assessment and reactor power evaluation of an accelerator-driven system (ADS). Those parameters can be evaluated by using the measured reaction rate distribution in the subcritical system. In this study, the experimental verification of this methodology is performed in various ADS cores; with high-energy (100 MeV) proton–tungsten source in hard and soft neutron spectra cores and 14 MeV D–T neutron source in soft spectrum core. The comparison between measured and calculated multiplication parameters reveals a maximum relative difference in the range of 6.6–13.7% that is attributed to the calculation nuclear libraries uncertainty and accuracy for energies higher than 20 MeV and also dependent on the reaction rate distribution position and count rates. The effects of different core neutron spectra and external neutron sources on the neutron multiplication parameters are discussed.     

Performance of the CENDL-3.2 and other major neutron data libraries for criticality calculations

Nuclear data are the cornerstones of reactor physics and shielding calculations.Recently,China released CENDL-3.2 in 2020,and the US released ENDF/B-Ⅷ.0 in 2018.Therefore,it is necessary to comprehensively evaluate the criticality computing performance of these newly released evaluated nuclear libraries.In this study,we used the NJOY2016 code to generate ACE format libraries based on the latest neutron data libraries（including CENDL-3.2,JEFF3.3,ENDF/B-Ⅷ.0,and JENDL4.0）.The MCNP code was used to ...     

Standardization of neutron sources in the graphite prism of a reactor

A method of standardizing neutron sources is described in which a comparison is made in a graphite prism of the strength of the source and the neutron loss produced by an absorber placed in the neutron field. The experimental apparatus and the measurements carried out by the author in 1951 are described; Ra--Be and Ra--Be sources were calibrated with an accuracy of ± 3%.     

Solution of one-speed neutron transport equation for strongly anisotropic scattering by TN approximation: Slab criticality problem

In this study, a recently proposed version of Chebyshev polynomial approximation which was used in spectrum and criticality calculations by one-speed neutron transport equation for slabs with isotropic scattering is further developed to slab criticality problems for strongly anisotropic scattering. Backward–forward-isotropic model is employed for the scattering kernel which is a combination of linearly anisotropic and strongly backward–forward kernels. Further to that, the common approaches of using the same functional form for scattering and fission kernels or embedding fission kernel into the scattering kernel even in strongly anisotropic scattering is questioned for T_N approximation via taking an isotropic fission kernel in the transport equation. As a starting point, eigenvalue spectrum of one-speed neutron transport equation for a multiplying slab with different degrees of anisotropy in scattering and for different cross-section parameters is obtained using Chebyshev method. Later on, the spectra obtained for different degree of anisotropies and cross-section parameters are made use of in criticality problem of bare homogeneous slab with strongly anisotropic scattering. Calculated critical thicknesses by Chebysev method are almost in complete agreement with literature data except for some limiting cases. More importantly, it is observed that using a different kernel (isotropic) for fission rather than assuming it equal to the scattering kernel which is a more realistic physical approach yields in deviations in critical sizes in comparison with the values presented in literature. This separate kernel approach also eliminates the slow convergency and/or non-convergent behavior of high-order approximations arising from unphysical eigenspectrum calculations.     

APCP分析程序使中子活化实验条件最佳化

反应堆中子活化分析(NAA)是近年来迅速发展的核分析技术,由于该方法取样量少、快速、准确、灵敏度高、具有多元素同时分析等优点,适合于生物样品中微量元素的分析研究。实验中,如果感兴趣的元素很多,往往要全面考虑到各个待测元素的放射性核素,选用合适的中子通量和辐照时间、衰变时间、计数时间。在这种情况下,正确合理选择最佳化实验条件是十分重要的。传统的方法是通过反复实验尝试和误差方法来取得,这将变得非常费时费     

MCNP在临界计算中的影响因素分析

简单说明计算所用的实验资料,在此基础上,介绍了采用MCNP计算MOX燃料实验堆有效增殖系数的过程。说明以不同的随机数序列重复计算问题,得到的计算结果总存在一定的涨落,并随着总迭代次数的增加逐渐减小;紧接着在统计涨落足够小的基础上,讨论了数据库的差异对计算结果的影响;对于含中子热化问题,指出采用S(α,β)热散射处理是得到合适的有效增殖系数计算值的必须条件。     

Questions concerning the determination of the effective external neutron irradiation dose

The ambient equivalent dose and effective dose are calculated using the ROBOT-Monte Carlo code. The absence of conservativity in the ICRU operational quantities with respect to the effective dose is discussed. It is shown that the factors W_R need to be revised. 4 figures, 14 references. Central Physicotechnical Institute of the Ministry of Defense of the Russian Federation. Translated from Atomnaya énergiya, Vol. 87, No. 4, pp. 297–302, October, 1999.