小角中子散射原位热力耦合加载装置

热力耦合近工况条件下材料微观结构的原位实验研究,对于深入理解材料服役性能演化机制十分重要,可给出样品微观上的纳米结构尺度分布。为充分发挥小角中子散射统计性好、取样体积大可开展原位实验等优势,本文基于中国先进研究堆小角中子散射谱仪,设计并研制了一台高温和拉力同时加载的原位实验装置,并实现了高温高压下原位测量材料的纳米尺度形貌变化。实验测试结果表明,装置最大载荷可达20 kN,最高温度800℃,控温精度优于±1℃。利用该装置对镍基单晶高温合金样品进行了原位小角中子散射测试,发现温度拉力条件下样品内部纳米结构的明显变化,表明基于该装置可开展热力耦合加载下的原位小角中子散射实验。该装置及其相应实验方法,可用于核电不锈钢等多种高温结构材料的原位加载实验研究,提供微观结构演化数据。     

低水平光激中子探测装置

光激中子由~(226)Ra γ源发射的γ量子轰击氘靶产生,由~3He正比管阵列组成的高灵敏中子—中子甄别探测器记录,给出计算氘靶内氘含量的信息。该装置主要用于含氘部件氘含量的无损测定。     

LaNi5-xAlxDy的中子衍射研究

分别测量了LaNi4.25Al0.75和LaNi4.75Al0.25两种合金样品在未吸氘及不同含氘量的氘化物样品的中子衍射谱,用Rietveld程序拟合出它们的晶体结构,并观察到了一些过去未曾注意的现象.LaNi4.25-Al0.75在吸氘过程中,氘原子优先占据的原子坐标为(0.17,0.32,0.45).     

低温制备气相色谱法分离氢同位素

介绍了采用低温制备气相色谱法以Al2O3装填分离柱进行氢同位素分离的装置。研究了氢同位素分离色谱柱(Al2O3分离柱)的制备和操作步骤，并对其分离效果进行了讨论。实验结果表明：经分离、净化后，氘纯度可达99．9％以上。该装置具有结构简单、操作费用低、一次性投资少等优点，在那些对氢同位素纯度要求高，用量小的部门或实验室有良好的应用前景。该装置不仅可用于氢氘分离，还可用于氘氚分离。     

LaNi5-xAlxDy的中子衍射研究

分别测量了LaNi4.25Al0.75和LaNi4.75Al0.25两种合金样品在未吸氘及不同含氘量的氘化物样品的中子衍射谱，用Rietveld程序拟合出它们的晶体结构，并观察到了一些过去未曾注意的现象。LaNi4.25Al0.75在吸氘过程中，氘原子优先占据的原子坐标为（0．17，0．32，0．45）。     

中国先进研究堆冷中子源氘系统充氢调试

正中国先进研究堆(CARR)冷中子源装置采用单相液氘作为冷中子慢化剂,液氘慢化剂通过堆内装置热虹吸回路封闭式自然循环和氦气夹层内低温氦气(约19 K)的冷却维持在液态状态。CARR冷中子源氘系统充氢调试的主要目的是采用氢替代氘进行首次氢装载、运行考验和氢卸载试验,验证氢装载/卸载操作程序;通过氦气制冷系统中氘气体分析仪测量氦气中的氢含量以进一步检查冷包及热虹吸回路的密封性。在充氢之前不开堆情况下启动冷源系统备用模式(SO)运行,采用含有200ppm氘气的混合氦     

高流速下钯上氢氘的排代研究

在钯氘化物上进行了高流速下氢氘的排代实验。结果表明，氘的排代效率随排代时间的增大而增加；与低压、低流速相比，高流速下排代压力的增加部分弥补了流速增大对氢氘交换速率的影响，因此仍可以得到较高的氘回收率。     

Ti-Mo储氢合金氘化物的中子衍射研究

Ti-Mo储氢合金合金是一种重要的氚靶材料,但其吸氢后的晶体结构变化还未见报道。由于氢的原子序系数很小,X射线衍射方法不易观察。而氢(氘、氚)的相干中子散射长度很大,中子对     

Ti—Mo合金氘化物热解吸过程研究

采用非零初压热解吸法研究了不同钼含量的钛钼合金TiMo_x(x=0.03、0.13、0.25、0.50、1.00,Mo与Ti原子数之比)氘化物的热解吸动力学,测试了氘解吸量与解吸时间的关系,应用反应速率分析方法得到了其热解吸速率常数k_d和热解吸表观活化能E_d;并与氕化物的热解吸动力学行为进行了比较。结果显示,x=0.03时,合金氕化物E_d小于氘合金化物E_d,与钛放氢动力学同位素效应一致;x=0.13、0.25时,氕化物E_d大于氘化物E_d;x=0.50、1.00时,氕化物与氘化物的E_d差别不大。通过初始解吸时合金中氕、氘含量的比较,结合室温下合金吸氕、氘量及物相结构,对合金放氢动力学同位素效应的本质进行了探讨。     

51 100kV低能纳秒脉冲化系统的研究

在低能氘-氘γ核聚变的研究中，为和正在研制的低能核聚变装置在能量上衔接起来，根据中子核数据测量的要求，提出研制出低能量的纳秒脉冲中子发生器装置。     

LaNi5－xAlxDy（x＝0.1，0.25，0.3，0.45，0.5，0.75）微观结构中子衍射研究

在室温下分别测量了一系列LaNi5－xAlx（x＝0.1,0.25,0.3,0.45,0.5,0.75）合金及其氘化物的中子衍射谱,利用FullProf程序对LaNi5－xAlx合金的衍射数据进行了精修,确定了它们的空间群为P6/mmm,得到了合金的晶胞参数、晶胞体积以及原子的占位,发现Al仅替代了部分处于3g位置的Ni,随着Al含量的增加,晶胞体积的变化量与Al替代量之间基本呈线性关系。对LaNi5－xAlxDy衍射数据的精修确定了氘化物的空间群仍为P6/mmm,分析了充氘后的相成分及晶胞参数、晶胞体积变化,得到了氘原子在晶胞中的占位和分布。     

Development of Pt-Mo Alloy Thermocouples for Incore Temperature Measurement in HTGRs

The Pt-Mo alloy thermocouples with element wires of Pt-5%Mo and Pt-0.1%Mo were developed for incore temperature measurements at a maximum temperature of 1,200°C for a long term and of 1,350°C for a short term in VHTRs. The thermocouples were tested at high temperature conditions of 1,000-1,400°C in an outpile electric furnace and in an incore capsule in the JMTR. The total thermal neutron fluence irradiated was 8.1×10²⁰ n/cm². The changes in their electromotive forces due to those exposures to high temperatures and radiation in the reactor core were evaluated to be less than ±2%. This meets the specifications that is required for the incore temperature measurements in VHTRs. Hence, the developed thermocouples can be used also in the incore temperature measurement in advanced type HTGRs and in incore irradiation experiments in the HTTR, in future.     

Study of pressure effects on ocean in-situ detection using laser-induced breakdown spectroscopy

Laser-induced breakdown spectroscopy (LIBS) has attracted extensive attention as a new technique for in-situ marine application. In this work, the influence of deep-sea high pressure environment on LIBS signals was investigated by using a compact LIBS-sea system developed by Ocean University of China for the in-situ chemical analysis of seawater. The results from the field measurements show that the liquid pressure has a significant effect on the LIBS signals. Higher peak intensity and larger line broadening were obtained as the pressure increases. By comparing the variations of the temperature and salinity with the LIBS signals, a weak correlation between them can be observed. Under high pressure conditions, the optimal laser energy was higher than that in air environment. When the laser energy exceeded 17 mJ, the effect of laser energy on the signal intensity weakened. The signal intensity decreases gradually at larger delays. The obtained results verified the feasibility of the LIBS technique for the deep-sea in-situ detection, and we hope this technology can contribute to surveying more deep-sea environments such as the hydrothermal vent regions.     

In-core coolant flow monitoring of pressurized water reactors using temperature and neutron noise

Noise measurements were performed at the Loss-of-Fluid-Test (LOFT) and Sequoyah-1 pressurized water reactors (PWRs) in order to investigate the possibility of inferring in-core coolant velocities from cross-power spectral density (CPSD) phases of core-exit thermocouple and in-core neutron detector signals. These noise measurements were used to investigate the effects of inlet coolant temperature, core flow, reactor power, and random heat transfer fluctuations on the noise-inferred coolant velocities. The effect on the inferred velocities of varying in-core neutron detector and core-exit thermocouple locations was also investigated. Theoretical models of temperature noise were developed, and the results were used to interpret the experimental measurements.Results of these studies indicate that the neutron detector/thermocouple phase is useful for monitoring core flow in PWRs. Our results show that the interpretation of the phase between these signals depends on the source of temperature noise, the response times and locations of the sensors, and the neutron dynamics of the reactor. At Sequoyah-1 we found that the in-core neutron detector/core-exit thermocouple phase can be used to infer in-core coolant velocities, provided that the measurements are corrected for the thermocouple response time.     

Irradiation experiments related to reactor safety in particular pressure vessel steels

The dependence of neutron induced embrittlement of reactor pressure vessel steels on irradiation temperature and neutron exposure was investigated for steels with different copper content. A pronounced increase of the ductile to brittle transition temperature shift with decreasing irradiation temperature was found and quantitatively determined. The influence of the neutron energy spectrum and flux density on the embrittlement was not significant.Rigs for irradiating assemblies of fracture mechanics specimens (CT and WOL) up to 100 mm thickness and also for irradiation experiments under cyclic loading were developed. Irradiation experiments with these rigs are in progress.Creep experiments on canning tubes under different load conditions (uniaxial load and biaxial load under internal and external overpressure) as well as an irradiation device for investigating defective PWR fuel rods are briefly reported.     

Tests for “cold fusion” in the Pd-D2 and Ti-D2 systems at 40–380 MPa and −196-27°C

Experiments have been conducted on the Pd-D₂ and Ti-D₂ systems at 40–380 MPa and −196-27°C to investigate the possibility that “cold fusion” occurs in palladium and titanium deuterides generated by reaction with high-pressure D₂ gas. The experiments were performed using a 4.8 mm i.d. stainless steel pressure vessel that can be operated routinely at pressures as high as 400 MPa. In experiments completed to date, reactions between high-purity Pd or Ti and D₂ were monitored with: (1) an array of three BF₃ neutron detectors, (2) an internal, type-K sample thermocouple, and (3) an internal, type-K reference thermocouple located approximately 10 cm above the sample thermocouple. Using a²⁵²Cf source, the efficiency of the BF₃ detector array was determined to be approximately 6%. During experimentation, the three neutron detectors were immersed in a water bath thermostated at 27°C. The neutron count rate, D₂ pressure, sample and reference thermocouple readings, and bath temperature were recorded continuously at time intervals ranging from 6 seconds to 10 minutes. Experimental results obtained so far range from negative to potentially significant. No sustained heat production has been observed in any experiment. Thermal pulses that persist briefly after pressurizing Pd with D₂ gas are attributable to small amounts of chemical heat released when Pd and D₂ react to form palladium deuteride. No sustained neutron flux above background was observed in any Pd-D₂ experiment. On the other hand, in a Ti-D₂ experiment just completed, potentially significant results were obtained. During this experiment, there was a period of 5 consecutive hours when count rates rose to approximately 60 counts/hour above the average background rate. This detector count rate corresponds nominally to 1000 neutrons/hour emitted from the Ti-D₂ sample. However, due to several deficiencies in our neutron detection methods and equipment, we cannot demonstrate conclusively that our experimental data are valid. Consequently, we are upgrading our neutron detection equipment in preparation for a second, improved Ti-D₂ experiment.     

高氚增殖比包层的设计及热工水力分析

本文设计了一种高氚增殖比包层(HBRB),该包层采用多孔U-10Zr合金作为中子倍增剂,Li4SiO4球床作为增殖剂,低活化马氏体(RAFM)钢作为结构材料。在详细研究包层加工工艺、流量分配、中子性能等问题的基础上,完成了包层内部详细结构设计。利用中子学软件分析计算了包层的氚增殖比(TBR)和热沉积分布,并根据计算结果对包层进行热力耦合分析。结果表明:包层TBR较高,且核性能稳定;冷却剂的流量分配情况和压降合理;包层内各组件冷却充分,温度和结构材料热应力不超过限值。     

中国先进研究堆中子残余应力谱仪校准

为校准中国先进研究堆（CARR）中子残余应力谱仪,获得一维位置灵敏探测器（PSD）的灵敏度系数、位置坐标与衍射角的几何关系及零点等参数,设计了校准实验和软件,并将获得的校准参数用于标准Fe粉前4个衍射峰的数据预处理,校正衍射峰位与计算衍射角的差值符合残余应力谱仪的误差要求。结果表明,该工作能准确获得校准参数,并可用于CARR中子残余应力谱仪的数据预处理。     

Consolidation of scientific and technological expertise to assess the reliability of reactor pressure vessel embrittlement prediction in particular for the arctic area plant (COBRA)

The evaluation and prognosis of reactor pressure vessel (RPV) material embrittlement in WWERs and the allowable period of their safe operation are performed on the basis of impact test results of irradiated surveillance specimens. The main problem concerns the irradiation conditions (irradiation temperature, neutron flux and neutron spectrum) of the surveillance specimens that have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV wall condition. In particular, the key issue is the possible difference between the irradiation temperature of the surveillance specimens and the actual RPV wall temperature. It is recognized that the direct measurement of the irradiation temperature by thermocouples during reactor operation is the only way to obtain reliable information. In addition, the neutron field's parameters in the surveillance specimens location have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level.The COBRA project, which started in August 2000 and had a duration of 3 years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contained state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The selected reactor for the experiment was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings, the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5 ± 4 °C. Uncertainties were evaluated also with experimental work carried out in the WWRSZ research reactor and by finite element modelling of surveillance capsules. The results obtained show that there is not need to perform temperature correction when surveillance data of irradiated specimens are used for embrittlement assessment of WWER-440(213) reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals 2.7 × 10¹² cm⁻² s⁻¹ with E > 0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.     

医院中子照射器物理启动

医院中子照射器是专门用于硼中子俘获治疗的核装置。在堆芯相对两侧,设有热中子束流和超热中子束流用于治疗,另外,在热中子束流内引出1条热中子束流用于病人血硼浓度测量。本文介绍其物理启动的6个实验,实验结果表明：满功率最大运行时间为12 h,最终后备反应性为4.2 mk,满功率运行时各工艺房间辐射水平满足设计辐射分区要求,4.2 mk反应性释放实验证明医院中子照射器具有固有安全特性。