Correction to Sheese et al. (2004).

Reports an error in the original article by B. Sheese et al (Health Psychology, 2004[Sep], Vol 23 [5], 457-464). On page 459, Table 1 contained errors in the reported values. The corrected table is provided. (The following abstract of this article originally appeared in record 2004-18051-003.) Research has shown that writing about emotional topics can positively influence physical and mental health. The current study tested the efficacy of an e-mail-based writing treatment and shows how such an implementation can aid in the search for moderators. Participants (N = 546) were randomly assigned to either a long- or short-interval traumatic writing condition or to a nonemotional writing control condition. In contrast to previous disclosure research, participants received and submitted their writing responses via e-mail. Health outcomes were assessed weekly for 5 weeks after treatment and were reported at the conclusion of the study. Results supported the effectiveness of an e-mail-based writing treatment in producing positive health outcomes and successfully identified several moderators of the writing treatment effect. The moderators implicated varied depending on the nature of the health outcome assessment. (PsycINFO Database Record (c) 2010 APA, all rights reserved)     

CAP1400控制棒提升极限分析方法研究

控制棒提升极限用于限定控制棒组棒位和可溶硼浓度的范围,以防止慢化剂温度系数突破限值。CAP1400核电厂采用机械补偿运行策略,使控制棒及硼浓度运行范围大为扩展,同时功能独立的M棒组和AO棒组同时插入堆芯使得插棒情况更为复杂,因此与传统核电厂相比,CAP1400核电厂的控制棒提升极限更难界定。本文建立了适用于CAP1400核电厂的控制棒提升极限分析方法,并给出计算结果。本文提出的方法合理地解决了复杂的控制棒运行情况给提升极限造成的影响,并充分地利用了电厂实测数据对提升极限进行修正。基于本文方法得到的提升极限精确且具备一定的保守性,所以便于电厂实际运行时使用。     

Reply to Dunnette's "Comments on Abrahams and Alf's 'Pratfalls in moderator research'."

Discusses M. Dunnette's (see PA, Vol. 49:Issue 1) comments on moderator techniques and shows that they are only partly correct because of their total reliance on improvements in hit rate. Instances are discussed in which it would seem hazardous to rely solely on empirical results in evaluating a moderator technique, while dismissing conceptual arguments based on methodological flaws. Limitations of his revision of quadrant analysis are discussed. Dunnette's further suggestion for completely removing the bias is shown to be theoretically impossible. Finally, the value of his 2-stage strategy is reduced considerably as a result of incomplete use of data but may, in some cases, improve prediction. (PsycINFO Database Record (c) 2010 APA, all rights reserved)     

一种适用于脉冲中子探测的RPC探测器研制

研制了一个阻性板内表面含Gd的RPC探测器用于脉冲中子探测研究,并成功测得了D-D脉冲中子发生器产生的中子信号。使用长硼计数管探测器作为参照,用相对效率法测得RPC探测器前、后端慢化体厚度分别为6 cm和4 cm时,对脉冲D-D中子的探测效率最高,约为0.1%。     

Unsteady two-dimensional multiphysical simulation of a pressure tube model expanded to contact with the outer concentric tube

For the blind calculation of the International Collaborative Standard Problem (ICSP) experiment on heavy water reactor moderator subcooling requirements, the COMSOL Multiphysics code is used to simulate plastic deformation of a pressure tube (PT) as a result of the interaction of stress and temperature. It is shown that the thermal stress model of COMSOL is compatible to simulate the multiple heat transfers (including the radiation heat transfer and heat conduction) and stress strain in the simplified two-dimensional problem. The benchmark test result for radiation heat transfer is in good agreement with the analytical solution for the concentric configuration of PT and calandria tube (CT). Since the original strain model of COMSOL only considers an elastic deformation with thermal expansion coefficient, the PT/CT contact cannot be predicted in the ICSP. Therefore, the plastic deformation model by the Shewfelt and Godin, widely used in the fuel channel analysis of CANadian Deuterium Uranium (CANDU) reactor, is implemented to the strain equation of COMSOL. The heat-up of PT, the strain rate, and the contact time of the PT/CT are calculated with the boundary conditions (BCs) given for blind calculation of the ICSP experiment.

The result shows a sudden expansion of the inner concentric PT within a few milliseconds. This unsteady simulation should be helpful for the conceptual design of experiment as well as for the understanding of multiphysics inside the fuel channels of the CANDU reactor.     

严重事故管理导则入口条件研究

严重事故管理导则的入口是从电厂应急运行规程(EOP)向严重事故管理导则(SAMG)转换的条件,也是严重事故缓解行动的重要依据。本文选取失去四级电源导致的典型高压熔堆序列以及大破口失水事故(LLOCA)导致的典型低压熔堆序列,根据严重事故堆芯剧烈氧化机理,得出了燃料温度、氢气产生速率及产氢量、入口集管过冷度以及慢化剂液位的关系。结果表明入口集管过冷度小于0且持续十几分钟,同时慢化剂系统的状态指示慢化剂液位低于6 900mm,可以作为严重事故管理的入口条件。最后,阐述了目前电厂EOP向SAMG转换的机制,并提出了改进的意见。     

Thermal neutron cross-section libraries for aromatic hydrocarbons

Solid phases of aromatic hydrocarbons, such as benzene, toluene, mesitylene and a 3:2 mixture by volume of mesitylene and toluene, were studied as potential moderator materials for a cold neutron source. Existing information on the (lattice) translational and rotational modes of the different molecular species was used to produce generalized frequency spectra; the latter included the internal vibrational modes which in turn involved the analysis of the weights of the different modes.Cross-section libraries were generated in ENDF and ACE formats for hydrogen bounded in those materials at several temperatures, and were used in Monte Carlo calculations to analyze their neutron production compared with standard cryogenic materials like liquid hydrogen and solid methane, the best moderators in terms of cold neutron production. In particular, cross-section libraries were generated at 20 K, which is a typical operating temperature for the majority of the existing cold neutron sources.It was found that those aromatic hydrocarbons produce neutron spectra which are slightly warmer than that of solid methane while presenting a high resistance to radiation, conforming in this way a new and advantageous alternative to traditional moderator materials.     

Anisotropic Thermal Expansion of Uranium-Refractory Metal-Carbides: UWC2 and UMoC2

The anisotropic diffusion coefficient has been calculated in a cylindrical cell with use made of the integral transport theory. The previous method⁽¹⁾ of calculating the diffusion coefficient requires much computer time to evaluate the generalized first-flight collision probabilities between two mesh points for a square cell. To circumvent this drawback, we introduce new calculation methods for determining the anisotropic diffusion coefficient in a cylindrical cell. Two methods are proposed–one uses the diffusion theory in the outermost moderator region of a cell; the other adopts the integral transport theory in that region as well. For the latter method, a new boundary condition is introduced for the cell surface, which, in the present problem, supersedes the usual isotropic return boundary condition. Using these two methods, the anisotropic diffusion coefficient can be evaluated with very short computer time. Moreover, an analytic expression is obtained for the special case where a cell is composed of a fuel and a moderator.     

Pulsed Neutron Intensity from Rectangular Shaped Light Water Moderator with Fast-Neutron Reflector

With a view to enhancing the thermal-neutron intensity obtained from a pulsed neutron source, an experimental study has been made to determine the optimum size of a rectangular shaped light water moderator provided with fast neutron reflector of beryllium oxide or graphite, and decoupled thermal-neutronically by means of Cd sheet. The optimum dimensions for the moderator are derived for the neutron emission surface and the thickness, for the cases in which the neutron-producing target is placed beneath the moderator (“wing geometry”) or immediately behind the moderator (“slab geometry”). The major conclusions drawn from the experimental results are as follows. The presence of the Cd decoupler inserted between the moderator and reflector prevent the enhancement of thermal-neutron emission time gained by the provision of reflector. with a graphite reflector about 14 cm thick, (a) the optimum area of emission surface would be 25x25 cm² for wing geometry and still larger for slab geometry, and (b) the optimum moderator thickness would be 5.5 cm for slab geometry and 8.5 cm for wing geometry. It is thus concluded that a higher neutron emission intensity can be obtained with slab than with wing geometry provided that a large emission surface can be adopted for the moderator.     

Total Neutron Cross Section in Heavy Ice at 77 and 15 K for Neutron Energies between 0.3 and 100 meV

A 10MW High Temperature Gas Cooled Reactor (HTR-10) designed by the Institute ofNuclear Energy Technology (INET) of Tsinghua University is being constructed now. The steam generator (SG) of the HTR-10 is one of the most important facilities for reactor safety. In order to investigate the thermal-hydraulic performance of the SG, a full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 in detail. The test assembly of the SGTM-10 simulates practical thermal and structural parameters of the HTR-10. The SGTM-10 consisted of three separated loops, primary-helium loop, secondary-water loop, and third-cooling water loop. There are two parallel tubes arranged in the test assembly. The main experimental equipment is shown in this paper. Analysis shows that for once-through steam generator simulation experiment, the electric-heated simulation method could not match practical operating condition. The results may not reflect true phenomena. The main results of experiments, for example effects of the outlet pressure, effects of the heating power, effects of the inlet sub-cooling are described. Experiments indicated, when the heat load of the HTR-10 is more than 30% the SG will be stable.