祁漫塔格地区黑山铀矿点矿化特征及控矿因素

在祁漫塔格地区黑山铀矿点成矿特征研究基础上,总结该区铀矿控矿因素为变质片岩混合岩、花岗岩体边缘过渡带、晚期岩脉、断裂构造和热液蚀变联合控矿;提出了该区铀矿找矿标志:围岩标志、构造标志、蚀变标志和放射性标志;指出在祁漫塔格地区寻找碎裂蚀变岩型铀矿的重点区域为祁漫塔格主脊断裂两侧的花岗岩边缘过渡带。     

大桥坞火山岩型铀矿床地质特征及勘查方法评述

通过新一轮找矿工作(2006～2008年)的勘查实践及研究认为:(1)大桥坞火山岩型铀矿床赋矿主岩为火山角砾岩,具两种类型的角砾岩.一种为火山通道相隐爆角砾岩,另一种为火山口相受层间构造控制的透镜状角砾岩:NW向组断裂构造是主控矿构造,矿床各矿带铀矿体的总体走向主要为NW向.(2)普查阶段勘查工程的布设应注重灵活机动原则,随地质认识的加深和变化及时调整工程布置方案.     

喀什凹陷北缘侏罗系沉积特征及其含铀性

喀什凹陷处于塔里木盆地西部边缘,受到塔里木板块、西昆仑和南天山造山带的影响,不但其地质结构和构造层具有自身特点,而且控制着中新生代沉积特征与分布。本文从沉积学的角度分析喀什凹陷北缘侏罗系岩性-岩相特征,进而分析其含铀性,认为喀什凹陷北缘侏罗系具备有利于铀成矿的岩性-岩相条件,即具备泥-砂-泥结构,但由于后期构造运动、沉积、压实等共同地质作用使侏罗系成岩程度较高,岩石较为致密,但是南天山有丰富的铀源,局部地段侏罗系中铀含量较高,而且地表铀有迁出现象,致使侏罗系的铀重新分配,可能进入地层深部,表明侏罗系碎屑岩有富集铀的能力。     

塔里木盆地中新生代地质构造演化与砂岩型铀成矿作用关系探讨

在综合分析塔里木盆地中新生代地质构造演化的基础上,结合铀矿石同位素年龄,划分了盆地砂岩型铀成矿作用的期次。通过对含矿目的层和后生改造成矿过程的构造控制作用研究,认为最后一次大规模喜马拉雅造山运动之前(N2前)的构造作用对砂岩型铀成矿有利,之后主要是对早期形成的矿体有所改造。对构造控矿的成矿模式进行了研究,认为在构造变形较弱部位主要形成库捷尔太式铀矿床、在构造变形强烈部位主要产出萨瓦布其式铀矿床、而在原生红色或杂色层位(K～N)中主要形成巴什布拉克式铀矿床。     

Inherently safe pool-type reactor as a generator of low-grade heat for district heating, air conditioning and salt water desalination

The society has a heavy demand for low-grade heat to satisfy its various needs. Different factors govern the expediency of applying nuclear reactors for these purposes. The required capacity of heat sources varies in a very wide range. In a majority of cases heat sources have to be located in the immediate vicinity of the users, therefore, nuclear reactors to be used for heat generation must feature enhanced safety. Pool-type reactors can be successfully used for producing low-grade heat. Owing to their design they feature a very high safety level. The absence of positive pressure excludes the possibility of a sudden rupture of reactor tank (vessel) or a fast loss of coolant. The availability of a large amount of water in the tank ensures long-term accumulation of residual heat. The adopted integral layout of equipment, as well as natural circulation of primary coolant improve reactor reliability and safety even further. Negative temperature coefficients of reactivity provide for reactor self-protection against reactivity accidents. Pool-type reactors can be used in newly established heat supply systems and can be built in the operation systems as well, which allows to reduce fossil fuel consumption by 80–90% depending on local conditions. Pool-type reactor heat can be used for desalinating salt water and for cooling water in absorption refrigerating machines with subsequent utilization of cold water for air conditioning, cooling of special premises, and the like. Pool-type reactors can also generate electric power to their in-house needs as well as household power requirements of a neighboring town.     

Preoperational vibration test of a primary heat transport system of the Canadian deuterium uranium reactor

During preoperational tests of Wolsong-2 nuclear power plant (NPP) which is a Canadian deuterium uranium (CANDU) reactor, vibration measurements were made on the primary heat transport (PHT) system. These measurements were evaluated by spectral methods to determine modal displacements and the modal stresses induced in the piping. The main aim of the measurement program is to confirm that the structural systems and components are adequately dimensioned for the operational vibration loads during the design life of the reactor. Structural analysis of the PHT system was performed using computer code to determine modal displacements and modal stresses analytically. The measurement results were compared with the analytically calculated resonance frequencies and modal values. A reasonable correspondence of the test and analytical results was achieved. The stresses evaluated on the PHT system were below the endurance limit for the material with a margin of safety.     

Kinetic parameters of a low enriched uranium fuelled material test research reactor at end-of-life

The kinetic parameters at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA’s 10 MW benchmark reactor. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time and effective delayed neutron fraction. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that in comparison with the beginning-of-life values, at end-of-life, the neutron flux increased throughout the core, the prompt neutron generation time increased by 3.68% while the effective delayed neutron fraction decreased by 0.35%.     

Anomalies and other concerns related to the critical heat flux

This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes or bundle subassemblies. They include the following:

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Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated channels. This phenomenon has been observed at high flows in both water and Freon.

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Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30-90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment controlled and the deposition controlled region and causes problems for prediction methods of the form CHF = f(X).

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Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, and diminishes the effects of flow regime/heat transfer regime transitions responsible for some of the unusual CHF trends, and results in a more gradually decreasing CHF vs. X curve.

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Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a temperature excursion) to a positive slope.

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Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the axial temperature distribution of a heated tube experiencing CHF when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical.

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CHF fluid-to-fluid modelling: differences in CHF trends at certain conditions between refrigerants and water at equivalent conditions.

The mechanisms responsible for these trends and the implications for bundle geometries are discussed.Concerns regarding the reported uncertainty of predicted CHF values and the range of application of CHF prediction methods are also discussed.     

Thermodynamic properties of uranium dioxide: A study of the experimental enthalpy and specific heat

The enthalpy content and specific heat of condensed uranium dioxide are discussed and analysed in this report. In particular the high temperature form of the specific heat curve, and the contributions to it, are considered. The excess specific heat is isolated, and the results compared with theoretical predictions.