PVC、EPR和CSM材料辐照后热氧化降解的反应动力学分析

采用气相色谱定量测量了聚氯乙烯(PVC)、乙丙橡胶(EPR)和氯磺化聚乙烯(CSM)等线缆绝缘材料经辐照后在不同温度下热老化过程中的O2消耗量.对实验数据进行了反应动力学分析,发现线缆绝缘材料的热氧化降解反应为一级反应.同时获得了不同温度下的反应常数,并与根据Arrhenius 方程计算的理论结果进行了比较.     

气相色谱法研究核电站用电线电缆绝缘材料的热氧化降解

测量了聚乙烯（ＰＥ）、聚氯乙烯（ＰＶＣ）、乙丙橡胶（ＥＰＲ）和氯磺化聚乙烯（ＣＳＭ）等线缆绝缘材料经辐照后在热老化过程中的Ｏ２消耗量与ＣＯ２生成量，发现ＣＯ２生成量随Ｏ２消耗量呈线性增加。计算了氧化降解反应的表观活化能，探讨了氧化降解的反应机理。     

温敏NIPAAm在硅橡胶膜上共辐射接枝及接枝膜性质研究

采用共辐射方式将ＰＮＩＰＡＡｍ接枝到硅橡胶上,结果表明在实验条件下,随吸收剂量的增加（到４０ｋＧｙ）,接枝率增加;但随剂量率的增加（５０￣５００Ｇｙ／ｍｉｎ）,接枝率反而下降。此接枝反应可以在空气中进行,但有一个明显的诱导期。增大单体的浓度和加入少量无机酸也能提高接枝率。接枝后,膜表面也表现出温度敏感性,其ＬＣＳＴ接近３５℃（ＤＳＣ）,所以在不同温度下（低于或高于ＬＣＳＴ）,它在水中的溶胀程度有明     

单甲基肼还原Np（Ⅵ）：Ⅰ.反应动力学研究

用分光光度法研究了ｃ（ＣＨ３Ｎ２Ｈ３）、ｃ（ＨＮＯ３）、ｃ（ＵＯ２＋２）、ｃ（ＨＮＯ２）、离子强度和温度等因素对Ｎｐ（Ⅵ）还原速率的影响。求出了Ｎｐ（Ⅵ）还原反应级数和不同反应条件下的表观反应速率常数，建立了反应速率方程式。温度为２４℃时，还原反应的表观反应速率常数ｋ２＝６１．３ｍｉｎ－１。反应的活化能为（５９．６２±１．１２）ｋＪ·ｍｏｌ－１。提高单甲基肼浓度，降低ＨＮＯ３浓度和升高温度有利于加快Ｎｐ（Ⅵ）的还原；离子强度和ＵＯ２＋２浓度对反应速率无明显影响；当反应体系中有ＨＮＯ２存在时，Ｎｐ（Ⅵ）还原速率明显提高。     

C_(60)与甲苯的激光诱导反应

在激光辐照下,Ｃ６０与甲苯反应形成了一系列的富勒烯衍生物,对其中的可溶产物进行的高效液相色谱－质谱－紫外光谱（ＨＰＬＣ－ＭＳ－ＵＶ）分析研究表明,反应产物中包括Ｃ６０（Ｃ７Ｈ７）ｎ,Ｃ６０Ｈｎ,Ｃ６０（Ｃ７Ｈ７）ｎＯｍ等,它们的形成可能与甲苯在激光作用下形成的自由基有关,多级质谱（ＭＳ／ＭＳ）分析表明,这些衍生物是自由基链结在笼状Ｃ６０外围的结果。     

离子注入胞嘧啶核苷酸的释磷动力学研究

离子注入下胞嘧啶（ＣＭＰ）核苷酸释放无机磷具有快反应和慢反应两种方式。研究了温度对慢反应过程的影响，发现辐射诱发的不稳定磷酸酯释磷反应的特殊模式：在３５３Ｋ以下和３６８Ｋ条件下分别遵循一级反应和二级反应方式，给出了反应的支这方程和速率常数。     

辐照交联聚乙烯绝缘电力电缆的性能

研究了聚乙烯绝缘材料的辐射交联加工工艺。该材料可应用于电力电缆。它的各项性能均达到ＩＥＣ－５０２（１９８３）标准的要求。     

2—羟基乙基肼还原Np（Ⅵ）：Ⅰ.反应动力学的研究

用分光光度法研究了ＨＮＯ３介质中ｃ（ＨＯＣ２Ｈ４Ｎ２Ｈ３）、ｃ（Ｈ＋）、ｃ（ＵＯ２＋２）、ｃ（Ｆｅ３＋）、ｃ（ＨＮＯ２）、离子强度和温度等因素对还原Ｎｐ（Ⅵ）反应速率的影响。测定了不同条件下的表观速率常数，确定了Ｎｐ（Ⅵ）还原反应速率方程式。２５℃时，还原反应的表观速率常数ｋ２＝３９１ｍｉｎ－１；反应活化能为５６．６ｋＪ·ｍｏｌ－１。提高２－羟基乙基肼浓度、降低ＨＮＯ３浓度或升高温度，Ｎｐ（Ⅵ）的还原加快；增加ＵＯ２＋２浓度和离子强度，还原速率稍有降低；当ｃ（Ｆｅ３＋）≥１．０ｍｍｏｌ·Ｌ－１时，Ｆｅ３＋对Ｎｐ（Ⅵ）的还原有一定的加速作用。     

裂片核素在岩石中的迁移研究

用动态模型方程描述了裂片核素＾１３７Ｃｓ（实验中用＾１３４Ｃｓ模拟）随地下水流在花岗岩中的渗透迁移过程。由渗透曲线拟合求得模型参数－－综合扩散－弥散系数和吸附速率常数,并在自行设计的水浴恒温槽中,进行了＾１３４Ｃｓ在花夺片上的吸附动力学研究。测定不同ｐＨ值和不同温度下的吸附曲线,推算出相应条件下的吸附速率常数及不同ｐＨ值时的吸附能。     

~（99）Tc~m－d，l－HMPAO分解动力学及稳定方法的研究

报道了用国产药盒制备的￣（９９）Ｔｃ￣ｍ－ｄ，ｌ－ＨＭＰＡＯ（ＬＣ）分解动力学的研究。结果表明：ＬＣ分解反应为一级反应。在实验条件为２０℃、放射性浓度为１８５ＭＢｑ／ｍｌ、ｐＨ为７．０时，ＬＣ的分解速度常数ｋ为０．０１７±０．０７ｈ￣（－１）。ＬＣ溶液的稳定性与ｐＨ、放射性浓度及温度有关。在室温下，加入１ｍｍｏｌ／ｌ龙胆酸作稳定剂时，７４０ＭＢｑ／ｍｌ的ＬＣ能稳定３ｈ，放化纯度大于８０％。     

TiD_(1.92)热解析的动力学行为研究

采用热分析和理论计算的方法对TiD1.92粉末的解析行为进行了研究。计算的TiD1.92粉末有3类解析峰,均与TiD1.92解析中发生相变导致的氘扩散速率变化有关。球形颗粒模型的解析与平板颗粒模型的解析因其相变和扩散机制不同导致β→α相变解析峰表现出一定的差异。TiD1.92粉末的热分析结果与平板颗粒模型的计算结果完全一致,其中δ相和β相钛氘化物的解析由表面反应控制,β→α相变阶段的解析由氘在α相的扩散控制,实验获得氘在钛表面的解析活化能为152.8kJ/mol以及氘在α相中的扩散活化能为68.6kJ/mol。     

A linear equation of characteristic time profile of power in subcritical quasi-steady state

ABSTRACT

An equation of power in subcritical quasi-steady state has been derived based on one-point kinetics equations for the purpose of utilizing it for the development of timely reactivity estimation from complicated time profile of neutron count rate. It linearly relates power, P, to a new variable q, which is a function of time differential of the power.

It has been confirmed that the points (q, P) calculated by using one-point kinetics code, AGNES, are perfectly in a line described by the new equation and that the points (q, P) calculated from transient subcritical experiment data measured by using TRACY made a line with a slope indicated by the new equation.     

N36锆合金包壳堆内腐蚀模型研究

本文利用池边检查数据,基于阿累尼乌斯方程建立了N36锆合金包壳堆内腐蚀最佳估算模型。由于缺乏腐蚀转折前数据,N36锆合金包壳腐蚀转折前氧化膜厚度只是时间的函数,腐蚀转折后氧化膜厚度是包壳温度和时间的函数。通过在最佳估算腐蚀模型上添加工程因子,建立了不同加工工艺N36锆合金包壳腐蚀模型。N36锆合金包壳腐蚀包络模型在最小腐蚀转折点的基础上建立。模型验证结果表明,N36锆合金包壳腐蚀模型与验证数据符合较好,能够用于N36锆合金堆内腐蚀行为模拟。     

Thermal Fatigue Study on the Divertor Plate Materials

Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate.Taking tungsten as surface material,a small-mock-up divertor plate was made by hot isostatic press welding (HIP),A thermal cycling experiment for divertor mock-up was carried out in the vacuum,where a high-heat-flux electronic gun was used as the thermal source,A cyclic heat flux of 9MW/m^2 was loaded onto the mock-up,a heating duration of 20s was selcted,the cooling water flow rate was 80ml/s.After 1000 Cycles,the surface and the W/Cu joint of the mock-up did not show any damage,The SEM was used to analyze the microstructure of the welding joint,where no cracks were found also.     

Application of Polarity Correlation Method to Graphite Moderated Reactor

The polarity correlation method has been applied to an experiment for measuring the prompt mode neutron decay constant of a reactor with a long neutron lifetime. Measurements have been performed in the SHE VIII-l at critical and several subcritical steady states (0~ — 1$).

The analog signal of detector output is converted to a polarity signal having only one-bit of information on the amplitude of the detected signal. The polarity correlation function is obtained as the correlation function between two polarity signals. Attention has been paid in the design of the correlator to obtain ample stability in operation. Using a digital computer, the prompt mode neutron decay constant is determined with 2% experimental accuracy from the decay curve of the polarity correlation function. A series of the prompt mode neutron decay constants is fitted to an approximated expression of inhour equation to obtain βl. The value of each prompt mode neutron decay constant was found to be in good agreement with that measured by the pulsed neutron technique. The polarity correlation method is particularly applicable to critical or near critical state where a pulsed neutron source cannot be effectively introduced.

In Appendix, an estimation is made for the error in the measured polarity correlation function due to unbalanced mean setting level in the experiment.     

Neutron Behavior in Cluster-Type Fuel Lattices, (II)

A multigroup method of calculation is presented for describing the neutron behavior in a cluster-type fuel lattice. It solves the integral transport equation by a semi-analytical method proposed in a previous paper for calculating collision probabilities in the lattice of a clustered fuel element. Using only fundamental nuclear data, it gives space and energy dependent neutron flux. The method has been programmed for HITAC-5020F (computer code named CLUSTER-III).

The accuracy of the method has been tested by comparing the calculation with the experiment described in Part (I) of this paper. The lattices are 28-pin clusters of UO₂ or PuO²+UO² fuel pins, with heavy-or light-water moderators and with light-water coolant containing varying void ratios. The quantities studied are micro-parameters, reaction distributions in energy and space, thermal disadvantage factors and the multiplication factors. It is found that the calculated results are generally in good agreement with experiment, typically within 10% for micro-parameters and thermal disadvantage factor, and within 1% for the multiplication factor.     

温度及功率变化研究电子辐射GaAs电池的热淬灭效应

为进一步探究电子辐射GaInP/GaAs/Ge三结太阳电池GaAs中间电池的热淬灭效应,对该子电池进行了温度及功率变化光致发光谱测量,分析了发光峰强度及其峰位的变化。利用Arrhenius方程对发光强度随温度的变化进行拟合,得到辐射引入GaAs中间电池非辐射复合中心的热激活能（0.96 eV）。通过分析10~300 K温度范围内发光强度的热淬灭效应,发现发光强度与激发功率的关系从线性关系逐渐转变为二次方关系。利用此关系进一步验证了GaAs中间电池的主要缺陷是非辐射复合中心,探究了电子辐射GaAs中间电池热淬灭效应发生的机理。     

keff uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gas-cooled reactors(HTGRs).In particular,the contribution of nuclear data to the keff uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for keff uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive infor-mation,and the"sandwich"method was used to quantify the keff uncertainty.We also compared the keff uncertainties to other typical reactors.Our results show that 235U is the largest contributor to keff uncertainty for both the CZP and depletion conditions,while the contribution of 239Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of 28Si significantly contributes to the keff uncer-tainty owing to its specific fuel design.However,the keff uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12％owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty prop-agation and quantification study for small-sized HTGR.     

keff uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gas-cooled reactors(HTGRs).In particular,the contribution of nuclear data to the keff uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for keff uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive infor-mation,and the"sandwich"method was used to quantify the keff uncertainty.We also compared the keff uncertainties to other typical reactors.Our results show that 235U is the largest contributor to keff uncertainty for both the CZP and depletion conditions,while the contribution of 239Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of 28Si significantly contributes to the keff uncer-tainty owing to its specific fuel design.However,the keff uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12％owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty prop-agation and quantification study for small-sized HTGR.     

辐射法制备含偕胺肟和咪唑啉基团功能高分子膜的研究

本文描述了共辐照法丙烯腈在聚乙烯薄膜上的辐射接枝,该接枝体系的活化能由阿累尼乌斯公式求得为2.51 J·mol~(-1)。含偕胺肟和咪唑啉的两种功能高分子膜可以通过聚乙烯与丙烯腈的接枝共聚物分别与羟胺和乙二胺的化学反应制得。还对该两种膜在透过蒸发中的通量和分离系数作了比较。