Automatic Plotting of Isoflux Contours with On-Line Printer

The effect of neutron streaming upon the neutron multiplication factor in a fast critical assembly FCA-VI is estimated by considering the anisotropy of the diffusion coefficient. In this paper, the Benoist formula is adopted to obtain the diffusion coefficient. In his original expression, many effects such as those of neutron angular distribution, multiple collision (correlation of different neutron passes) and finiteness of lattice system are included. In order to estimate these effects, the generalized first-flight collision probability method is adopted as in the previous paper, and an improvement is brought over the previous method so as to calculate the effective diffusion coefficient in a practical slab lattice cell which is asymmetric about the center of the cell.

Using the method of calculation described above, the anisotropic effect of neutron streaming in the lattices TA, TB, TC and TD used in the FCA VI-1 assembly is estimated. The effect of anisotropy of the diffusion coefficient upon the multiplication factor for the TA, TB, TC and TD lattices is found to be ?0.156, ?0.181, ?0.242 and ?0.330% δk/k, respectively.

Usually, the effects of neutron distribution, multiple collision and anisotropic scattering are neglected in evaluating the diffusion coefficient when using the Benoist formula. Among these factors, the effect of multiple collision of neutrons has the largest effect upon the diffusion coefficient and upon the neutron multiplication factor. For the TA lattice, 30% of the overall effect of anisotropy is attributable to multiple collision.     

Current Status of the Fabrication of Li4SiO4 and Beryllium Pebbles for CN HCCB TBM in SWIP

Tritium breeder and neutron multiplier as functional materials play an important role not only in ITER test blanket module (TBM) but also in fusion reactor. The paper describes the status of the fabrication of the two materials in Southwestern Institute of Physics (SWIP). Li4SiO4 pebbles were fabricated by melt-spraying method. Most of the pebbles with the diameter of 1.0 mm are well spherically shaped. The properties of the pebbles have been investigated. The results show that the pebbles produced by this method have a high density of 93% TD (theoretical density). It was also found that the open/closed porosity will be decreased after thermal treatment, but the average crush load will be increased to 7 N. The rotating electrode process (REP) has been adopted to produce beryllium pebble for impurity control and mass production. The pebbles with the diameter of 1.0 mm were produced by REP. The beryllium pebbles produced by REP look almost perfectly spherical with a very smooth surface and a high density of 98% TD. The test results indicate that REP method has excellent prospects for the fabrication of beryllium pebbles and the attractiveness of their properties.     

Comparative analyses of coated and composite UN fuel – Monte Carlo based full core LWR study

Longer continuous operation of a nuclear reactor leads to higher availability of nuclear power plant, entailing economic gain. In this context the option being explored, recently, is the use of higher density fuels as compared to current fuels i.e. UO₂. Uranium mono nitride (UN) fuel is one of the options being explored in this regard. UN fuel has been used in nuclear industry for a long time with fast reactor option. More recently, studies have targeted its use in Light Water Reactor (LWR) environment. The main problem with using UN fuel in LWR is its potential reaction with water which produces hydrogen. Researchers have proposed different approaches to overcome this problem.One option is the use of coatings around UN fuel pellet to avoid the direct contact of water with fuel. The second option is use of a secondary phase (10 vol. percent) like ZrO₂ which can make an oxide layer in case of contact with water and protect the main phase, Uranium mono nitride (80 vol. percent). Remaining 10 vol. percent is considered to be consumed by porosity. This study aims at comparison of neutron physics behavior of both these options in LWR environment. The upshot of the study is to quantify the impact of adding layers or secondary phase with respect to pure/complete UN fuel. To study these effects, two theoretical densities i.e. 95% and 80% for UN fuel are chosen for analyses. To avoid the problem of C-14 production from N-14, fuels studied are considered to be having 100% N-15.A validated model of Benchmark for Evaluation and Validation of Reactor Studies (BEAVRS) is used to perform all the analyses. Integral neutron physics parameters like neutron energy spectra, Radial Peaking Factors, Axial Peaking Factor, Doppler coefficient of reactivity, Isothermal Coefficient of reactivity and Temperature defect, Control Rod worth and excess reactivity for whole core are compared at Beginning of Cycle (BoC). Burnup obtained by different fuel option is also compared. Consistent with pin-cell based earlier findings, this full 3D analyses with UN based fuel shows noticeable spectral hardening leading to decrease in the value of control rod worth and less negative Doppler coefficient of reactivity while power peaking factors remain mostly unchanged. The economic advantage of switching to UN based fuels is expected when UN fuel above 80% TD is used. Approximately 19% increase in Equivalent Full Power Days (EFPDs) is witnessed by using 95% TD UN based fuels.     

A study of transmutation of minor-actinide in a thermal neutron field of the advanced neutron source

Since Minor Actinides (MAs) have a large cross section in a thermal energy region, a thermal neutron field has possibility to provide high relative transmutation rate. The transmutation of MA in a thermal neutron field was investigated in this study, focusing on relative transmutation rate, transmutation rate of weight and neutron economy in a thermal neutron field in comparison with those in a fast neutron field. The followings are the major results: (1) a thermal neutron field, especially a well-thermalized neutron field, provides high relative transmutation rate of MA. (2) However, the transmutation rate of weight of MA is limited since the region with high flux is restricted even in the ANS. (3) The accumulation of ²⁴⁶Cm slows the relative transmutation rate of MA in a thermal neutron field. However, it can be compensated by providing a neutron field with high flux in keV region. (4) The neutron economy of the transmutation of MA is not so bad in a thermal neutron field, and is rather good when the neutron flux is high.     

反应堆物理实验中的源倍增法研究

给出了反应堆物理实验中临界测量和次临界度测量通常所采用的源倍增方法研究。首先从有源的扩散理论出发,导出了与以前不同的源倍增方法的公式。源倍增方法测量的参数实际是次临界系统在外源作用下的有源次临界中子倍增因子ks,而不是在这之前的中子有效倍增因子keff,然后研究了实验装置的临界质量,研究了ks与外源位置和能谱的关系,证明了导出的源倍增方法的理论是正确的。该方法可像过去那样用于反应堆物理实验中的临界外推测量,但不能用于次临界度测量。解决了长期困扰人们有关源倍增方法测量的参数问题。最后讨论了ks和keff的差别和关系以及对临界外推测量和核临界安全的影响。     

Investigation of ortho↔para hydrogen conversion by collisions with neutrons

Liquid hydrogen (LH) is constituted by para hydrogen (pH) and ortho hydrogen (oH) molecules. We analyzed the neutron scattering cross sections for pH→oH, pH→pH, oH→pH, and oH→oH reactions in LH. Results show that incident neutron energy lower than 14.7 meV is not sufficient to achieve pH→oH conversion. As energy increases, the cross section of pH→oH reaction increases sharply. If incident neutron energy is larger than 30 meV, branching fraction of pH molecules being converted into oH molecules upon inelastic scattering with the neutrons approaches 100%. As for the opposite process, branching fraction of oH→pH conversion is 10%–20% when incident neutron energy is lower than 30 meV. This conversion fraction oscillates with increasing incident neutron energy, eventually stabilizing at 30%. Based on the cross sections of the four reaction channels, we calculated the corresponding conversion rates based on a reactor cold neutron source model. There is 0.12% pH molecules of LH moderator would be converted to oH for one month operation . If neutron flux increased by 2 orders of magnitude, more than 10% pH would be transformed into oH, which means relative conversion rate of pH→oH induced by neutron inelastic scattering is not negligible.     

高保真中子输运计算的多级加速理论及应用

为解决高保真中子输运计算耗时严重的问题,本文提出了多级加速理论。其中,针对迭代求解过程,采用迭代加速的思路,即通过等价的低分辨率系统加速以减少迭代次数;针对瞬态求解,采用时间步加速方法,通过建立多级预估校正系统,实现在大时间步长下开展准确的高保真中子输运计算。最终引入不同分辨率系统的概念,将时间步加速方法与迭代加速方法整合形成一套完整的多级加速理论,并将其应用到精细化中子物理计算程序HNET中。采用典型瞬态基准题验证HNET程序加速效果。数值结果表明：多级加速理论能够在保证高保真中子输运计算精度的同时,极大地提升计算效率。     

Experimental studies of the absolute total delayed neutron yields from neutron induced fission of U in the energy range 1–5 MEV

Measurements of the total delayed neutron yield from fast neutron induced fission of ²³⁸U have been made using the experimental method based on the periodic irradiation of the fissionable sample by neutrons from a suitable nuclear reaction. The preliminary results on the energy dependence of the total delayed neutron yield are presented. According to the comparison of the experimental data with our prediction based on correlation properties of delayed neutron characteristics, it is concluded that the value of the total delayed neutron yield near the threshold of the (n,f) reaction is not a constant.     

Influences of Neutron Source Spectrum and Thermal Neutron Scattering Law Data on the MCNPX Simulation of a Cyclotron-Based Neutron Field for Boron Neutron Capture Therapy

In our previous study, the simulation of a cyclotron-based neutron field for boron neutron capture therapy (BNCT) using a (p,n) spallation source with the MCNPX code was validated through measurements of the neutron energy spectrum behind the moderator assembly and the thermal neutron distribution in an acrylic phantom using reaction rates of ¹⁹⁸Au. These validations showed that the simulation generally well reproduced the measurements. However, some discrepancies between the measurements and the calculation remained for clinical trials. In this paper, we investigated the influences of neutron source spectrum and thermal neutron scattering law data in the simulation to resolve those discrepancies. We also compared measured and calculated neutron doses behind the moderator assembly with results obtained using a tissue equivalent proportional counter. We clarified that the neutron source spectrum calculated using the LA150 data led to the overestimation of high-energy neutrons in a phantom, but this overestimation did not significantly affect the neutron dose distribution in a phantom, because a dominant part of the absorbed dose is due to neutrons of energies below 1MeV. The study of the influence of neutron scattering law data in a phantom also indicated that the use of selected S(α,β) data led to an improvement in the simulation of thermal neutron behavior.     

食管癌放射治疗计划的剂量学比较

比较食管癌患者应用螺旋断层(Tomo Helical,TH)、径照断层(Tomo Direct,TD)、容积旋转调强(Volumetric modulated arc therapy,VMAT)和固定野动态调强(Intensity modulated radiation therapy,IMRT)4种放疗计划的剂量学差异。选取18例食管癌患者,利用Pinnacle9.2计划系统设计单弧360°VMAT放疗计划和5野IMRT放疗计划。利用Tomo HDTM2.0.7计划系统设计TH放疗计划和5野TD放疗计划。利用剂量体积直方图(Dose volume histogram,DVH)统计靶区剂量参数、适形性指数(Conformity index,CI)和均匀性指数(Heterogeneity index,HI),肺、心脏、脊髓剂量体积参数,出束时间和治疗跳数。TH计划靶区适形性和靶区均匀性略优于TD计划,VMAT计划靶区适形性和靶区均匀性略优于IMRT计划,且前两种计划明显优于后两种;TH计划和VMAT计划肺V20Gy、V30Gy,心脏V30Gy、V40Gy分别优于TD计划和IMRT计划;但是TD计划肺V5Gy具有其他计划都不具有的优势。TH计划优于TD计划优于VMAT计划优于IMRT计划。但如果考虑性价比,本研究认为对于食管癌VMAT计划是首选;如果考虑放射治疗计划的质量,TH计划是首选;但如果靶区体积比较大,肺的低剂量无法达到临床要求时,可以考虑用TD计划解决这一难题。     

Point Scattered Function (PScF) for fast neutron radiography

Fast neutron radiography opened up a new range of possibilities to image extremely dense objects. The removal of the scattering effect is one of the most challenging problems in neutron imaging. Neutron scattering in fast neutron radiography did not receive much attention compared with X-ray and thermal neutron radiography. The purpose of this work is to investigate the behavior of the Point Scattered Function (PScF) as applied in fast neutron radiography.The PScF was calculated using MCNP as a spatial distribution of scattered neutrons over the detector surface for one emitting source element. Armament and explosives materials, namely, Rifle steel, brass, aluminum and trinitrotoluene (TNT) were simulated. Effect of various sample thickness and sample-to-detector distance were considered. Simulated sample geometries included a slab with varying thickness, a sphere with varying radii, and a cylinder with varying base radii. Different neutron sources, namely, Cf-252, DT as well as DD neutron sources were considered. Neutron beams with zero degree divergence angle; and beams with varying angles related to the normal to the source plane were simulated.Curve fitting of the obtained PScF, in the form of Gaussian function, were given to be used in future work using image restoration codes. Analytical representation of the height as well as the Full Width at Half Maximum (FWHM) of the obtained Gaussian functions eliminates the need to calculate the PScF for sample parameters that were not investigated in this study.     

微型中子源反应堆中子谱参数测量

以Au、Zr和Fe为活化探测器,采用裸探测器法测量中国原子能科学研究院微型中子源反应堆的中子谱参数f、α、f_F和φ_th。内辐照座的α、f和f_F分别为－0.007±0.003、20.8±0.4、5.5±0.2。该方法对φ_th的测量结果与4πβ-γ符合法的一致,相对偏差小于2%。与SLOWPOKE相比,微堆有较高的α、f_F值。与已有测量数据的比较表明,微堆中子谱在很长一个时期内是稳定的,利用微堆作为中子源的k₀法中子活化分析不需中子注量率监测器,且比较器一经照射和测量后,可用于其后较长时间内所有分析的计算标准。     

核临界安全中的源倍增法研究

文章对核临界安全研究中通常采用的现场测量技术———源倍增法进行研究。从有源扩散理论出发,导出了与keff不同的有源次临界中子有效增殖因子ks的表达式,并在次临界系统上进行了验证研究。验证实验研究证实了所导出的ks 的正确性。源倍增法测量的参数实际上是次临界系统在外源作用下的有源次临界中子有效增殖因子ks,而不是以往的中子有效增殖因子keff,这就解决了长期困扰人们的有关源倍增法测量的参数问题。文章讨论了ks 与keff间的差别和关系以及它们对核临界安全的影响。     

Neutron fluctuations induced by absorber vibrations: the QMC reactor experiment—I

The dependence of rod vibration induced, neutron density fluctuations on the static neutron gradients was investigated experimentally at the QMC research reactor. By appropriate fuel loading arrangements the rod was made to vibrate (a) within a flat flux and (b) a flux gradient. The neutron density was fluctuating with the frequency of the vibrating absorber only when the static flux gradient in the vicinity of the absorber was not zero. The double frequency effect was not observed.     

100～500keV能区~9Be(d,n)~(10)B反应中子产额及角分布测量

金属铍具有较稳定的理化性质,同时也易与入射的轻核粒子发生反应产生中子,在小型中子发生器的研制中具有其他靶材所不具有的优势。9Be(d,n)10 B反应的中子产额和中子角分布数据,尤其是低能区的数据,对小型中子发生器的设计非常重要。本文在中国原子能科学研究院的600kV高压倍加器上测量了100～500keV能区9Be(d,n)10B反应的中子产额和中子角分布。     

0.144~1.2MeV单能中子参考辐射场的建立

本工作利用2×1.7MV串列加速器建立了0.144、0.250、0.565和1.2MeV单能中子参考辐射场。所需能量的中子通过⁷Li(p,n)⁷Be和³H(p,n)³He反应产生。利用TARGET程序计算了中子注量谱，利用自行设计的反复充气式反冲质子正比计数器绝对测量了上述能量点的中子注量，合成标准不确定度≤2.0%。2001年参加了由国际电离辐射咨询委员会第三分部组织的单能快中子注量测量国际比对，取得了较为满意的结果。     

中子输运的随机理论

运用概率理论,考虑t时刻n个相空间点(r,u_iΩ_i)单位体元中分别出现N_i(i=1,2,…,n)个中子的概率P_N(r,t,uΩ),提出一个新的中子输运的随机理论,导出概率母函数Fn的非线性积分微分方程组。在某些近似下,n=1概率分布一阶矩方程恰好是中子平均数玻尔兹曼方程。将各向同性散射的单速中子随机理论应用于点堆模型。在一个超临界系统中,当t→∞时,出现有限个中子的概率为零,P_N=0(0＜N＜∞),即系统内或没有中子,或有无限多中子。给出了母函数的近似解,导出了母函数概率分布各阶矩的近似方程及解式。标准差公式表明,当初始中子数起伏ξ0较大,初始中子平均数N0不够多,或中子源强Q很弱时,对于0＜λ＜1的增殖系统,中子数的起伏很大,应予以重视。     

Neutron spectrum change with thermal moderator temperature in a compact electron accelerator-driven neutron source and its effects on spectroscopic neutron transmission imaging

Recently, Compact Accelerator-driven Neutron Sources (CANSs) are attracting attention. In CANSs, a simple thermal neutron moderator such as polyethylene is often used from the viewpoints of cost, simplicity and maintainability. In most cases, the temperature of such a moderator has not been controlled although it is natural that the moderator temperature and the neutron spectrum will change with accelerator-operation. Thus, we simultaneously measured neutron spectra and the temperature of a polyethylene moderator at the Hokkaido University Neutron Source (HUNS) driven by a compact electron accelerator to observe the effect of any temperature change on the reliability of spectroscopic transmission measurement. The ratio of the neutron effective temperature and the moderator temperature was constant in HUNS case, although both increased by 4–5 K within one hour after the start of accelerator-operation. This indicated that the neutron effective temperature was well estimated by the moderator temperature. The effect of the temperature change can be easily avoided by excluding data collection before the moderator warms up. These results suggested that the monitoring of moderator temperature is recommended in compact electron accelerator-driven neutron sources with a thermal neutron moderator to guarantee reliability of spectroscopic transmission measurement without sacrifices of cost, simplicity and maintainability.     

Count-loss effect in subcriticality measurement by pulsed neutron source method, (II) proposal for utilization of neutron detection system operated in current mode

It was pointed out in the previous paper that the neutron decay constant determined by the pulsed neutron source method that employs the neutron detection system operated in the pulse mode is expected to be biased owing to the count-loss effect even when the intensity of pulsed neutron source is not high. To avoid this difficulty, by paying attention to the current mode that is inherently free from the count-loss process, the pulsed neutron source method with neutron detection system operated in the current mode was proposed. Using this method, not only the neutron decay constant but also the absolute value of subcriticality are obtained when a proper time constant of neutron detection system is selected.     

<Superscript>239</Superscript>Pu prompt fission neutron spectra

Calculations of ²³⁹Pu(n, F) prompt fission neutron spectra have been performed for neutron energy up to 20 MeV. The exclusive spectra of pre-fission neutron reactions (n, xnf) were calculated on the basis of the Hauser-Feshbach model simultaneously with the cross sections of (n, F) and (n, 2n) reactions. The spectra of neutrons emitted by fission fragments were approximated by a sum of two Watt distributions. The components of the prompt fission neutron spectra due to pre-fission neutrons are manifested in the prompt fission neutron spectra and the average neutron energy. A correlation is established between this effect in the contribution of emissive fission (n, xnf) in the fission cross-section of ²³⁹Pu(n, F) and ²³⁵U(n, F). It is shown that the ²³⁹Pu(n, F) prompt fission neutron spectra used in applied calculations do not correspond to the experimental differential data and the systematic regularities in the spectra and their average energy found for the most carefully studied nuclei ^235,238U and ²³²Th. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 119–124, August, 2007.