CARR铱源试验靶件设计和试验验证

为配合中国先进研究堆(CARR)铱源辐照生产项目,设计制造了铱源试验靶件,对试验靶件的设计参数、结构尺寸进行了介绍。在堆外使用专门的传热装置模拟铱源靶件的外部和内部传热工况,测量了用于模拟辐照罐壁面温度和样品温度的传热装置的壁面温度和内部温度,结果验证了热工分析方法是合适的。入堆试验靶件由含有铱片样品的辐照罐和等量发热的模拟罐组成。堆内试验获得的数据综合验证了试验靶件物理热工的分析结果,这个结果可对CARR铱源辐照生产安全评审提供依据且偏于安全。     

秦山核电厂甩负荷试验

本文简要地叙述了秦山核电厂甩负荷试验的目的、方法和结果,同时也给出了核电厂主要参数的瞬时变化。该试验结果表明:秦山核电厂的综合联锁性能正常,试验结果与验收准则一致。     

ACME试验台架典型小破口工况试验及数值分析

对我国首个大型非能动堆芯冷却系统整体试验台架（ACME）中的典型小破口事故进行了试验及数值分析。分析结果表明：在ACME上开展的典型小破口试验,其事故序列及试验现象符合预期;RELAP5数值分析的主要结果能较好地反映试验现象,与试验结果吻合良好;堆芯棒束区相间摩擦模型的选用对堆芯坍塌液位的计算有较大影响,在不同阶段选用不同的模型可使计算结果更好地与试验值相匹配。     

西安脉冲堆带核调试试验

介绍了西安脉冲堆带核调试试验内容，给出了稳态堆芯、脉冲堆芯及72小时额定功率运行试验堆芯的反应性和中子注量率等参数的测量结果。测量结果证明，西安脉冲堆性能参数达到了设计指标，并为该堆的运行提供了必不可少的运行参数。     

泰山核电二期工程安全壳结构整体性试验

本文结合秦山核电二期工程IRX安全壳结构整体性试验，介绍了其测试原理和方法，验证标准以及试验结果与分析。     

套管型燃料组件动态特性研究

采用试验结合数值分析的方法对套管型燃料组件在不同安装边界条件下的动态特性进行研究。首先采用有限元分析方法对燃料组件的固有频率进行理论预估,计算结果可为动态特性试验参数的设置及模态参数的辨识提供参考;试验以敲击法为主,单点激励,多点拾振,通过对试验数据的多方法拟合,获得燃料组件的固有频率、振型等参数。试验结果与数值分析结果吻合较好。     

CEFR堆芯燃料组件抗震试验方法研究

为了验证中国实验快堆（CEFR）堆芯燃料组件的抗震性能,保证地震下结构完整性和气密性,必须研究制定兼具代表性和包络性的堆芯组件抗震试验方法。本文基于俄罗斯组件耐振试验方案分析,结合国内试验规范和堆芯实际约束条件,提出了一套新的组件抗震试验方法,并通过分析计算论证新方法的合理性。结果表明：新方法的试验结果是保守的,可保证在相同地震输入下单组件应力、冲击响应基本能包络处于堆芯组件群中的组件响应,新方法要求单根组件分别在刚性台架和柔性台架上依次完成抗震试验。本文结果对快堆堆芯组件的抗震试验研究具有重要指导意义。     

先进堆非能动余热排出系统综合试验研究

介绍了先进堆非能动余热排出系统综合试验研究的试验装置和冷热芯位差阈值研究结果、稳态试验研究结果、瞬态特性分析结果,以及MISAP2.0程序改进、验证结果。试验研究结果可为先进压水堆核电站非能动余热排出系统原型设计(系统布置、设备容量和系统启动方式等)提供试验依据,并为舰船核动力装置非能动余热排出系统的研究与设计提供可参考的试验数据,开发的具有自主知识产权的MISAP2.0程序为我国自行设计先进堆非能动余热排出系统提供了必要的设计手段。     

秦山核电二期工程反应堆堆内构件模型流致振动试验研究

秦山核电二期工程是我国自行设计的第一个600MW级核电站,有必要进行反应堆堆内构件流致振动试验研究.本文介绍了按相似准则设计的实堆15的堆内构件试验模型,进行流致振动试验采用的试验方法,完成的试验内容以及试验数据的分析和处理.测得了吊篮结构在冷却剂流动冲刷下的脉动压力和各种响应参数.试验结果可用于秦山核电二期工程安全评审,并提供了吊篮流致振动响应计算的载荷谱和实堆振动监测、故障诊断的参考样本.     

1E级安全壳内用电动机鉴定的核安全审查

对压水堆核电厂1E级安全壳内电动机的鉴定过程和鉴定文件进行审查,要求对鉴定试验结果与标准法规的符合性以及与安全壳内环境要求的一致性做出评价。首先介绍1E级安全壳内用电动机的老化试验、设计基准事故试验等主要鉴定试验,对随后审查过程中遇到的典型问题进行分析,并将IEEE 334与RCC-E两套标准进行对比探讨。     

除碘器非破坏性检验装置和除碘器的泄漏检验

本文介绍了用氟里昂-112作指示剂的活性炭除碘器非破坏性检验装置及其某些部件的性能;选取检验条件的原则和具体的检验条件;在所选定的检验条件下,对两种结构的折叠式除碘器样机的检验结果。调试和检验表明,该装置设备简单;工艺参数便于控制;可在5min 之内完成一台除碘器的泄漏(泄漏率<0.01%)检验,可在30min 内对不同结构和加工质量的除碘器的综合性能作出评价。     

Prediction of air and steam leak rate through cracked reinforced concrete panels

The leak rate prediction of air and steam through a cracked concrete wall is an extremely important issue in assessing the safety of a nuclear reactor containment building. Such a problem requires a multidisciplinary approach involving both the non-linear analysis of the structure, and the thermodynamics aspects related to the flow of a gas through a conduit. In the present paper, some of the available leak rate evaluation formulae are reviewed, and an application to the prediction of the leak rate of either dry air or air+steam mixture through a cracked concrete panel is presented. Finally, in order to validate the numerical procedure herein adopted and to give some indication on the relative merit of the different leak rate formulae considered, the results of the numerical application are compared against leak rate values measured during an experimental test carried out at the ISMES laboratory.     

Durability and field condition study of seal of SAVY-4000 storage container

The conditions of use for rubber O-rings are at least as important as their physical properties in their effect on the quality of a seal. Under normal use conditions, O-rings may be subject to wear and environmental contaminants such as hair and dirt. This study examines how these factors impact the leak tightness of a nuclear material storage container and the likelihood of the inadvertent release of radioactive material. The durability lifetime of an O-ring was explored by opening and closing four SAVY-4000 1 quart containers 100 times and periodically performing helium leak testing, though no significant change in leak rate was observed. This study also explored how the accumulation of dust or hair on the O-ring surface would affect the leak rate of the containers. A single hair crossing the seal, or a sufficient amount of particulate matter would compromise the seal, but after cleaning, the seal was re-established.     

泳池式轻水反应堆内电磁线圈可控温辐照装置设计

本文设计了在泳池式轻水反应堆（简称泳池堆）内在线测量电磁线圈电性能的可控温辐照装置。采用MCNP程序进行中子物理计算，对泳池堆、线圈骨架的结构尺寸与物质组分进行了精细全尺寸模拟，得出辐照装置的发热功率和中子注量率。通过初步估算，使用ANSYS CFX进行了数值模拟，得出辐照装置内线圈在堆运行时的温度，并提出温度控制的方法。辐照装置采用铝材加工制造，并进行了垂直度测试、气压测试、检漏测试。增加了绝缘设计，将辐照装置与泳池堆之间进行绝缘。在线圈处预埋铠装热电偶，对线圈温度进行实时监测。在泳池堆内对电磁线圈进行辐照试验，结果表明，本文设计的辐照装置能满足电磁线圈在泳池堆孔道内进行辐照试验的要求，并可对电磁线圈进行实时温度控制。     

LBB泄漏率计算与热力学非平衡效应影响评估

裂纹泄漏率计算是破前漏（LBB）在核电站管道和设备上应用的基础。在Fauske模型基础上,整个裂纹内流体流动假设为等焓过程且充分考虑摩擦效应对裂纹临界泄漏率的影响,利用Mathcad计算得到了管道裂纹两相泄漏率,与已有文献中实验数据进行对比,将其发展成为可准确计算裂纹泄漏率的计算机程序。同时根据两相流动不平衡理论,对模型进行热力学不平衡参数影响修正。结果表明：随裂纹长径比(L/D)增大,两相泄漏率减小;随裂纹入口滞止压力增大,两相泄漏率增大;裂纹入口流体过冷度增大,两相泄漏率增大,数学模型计算结果与实验结果趋势一致,但忽略热力学非平衡效应,数学模型计算得到的临界流量小于实验流量。对于热力学不平衡参数修正后模型,模型计算得到的结果均与实验数据符合很好,故由修正后模型编制的Mathcad程序可完成裂纹泄漏率的准确计算,为LBB在核电站管道上的应用提供基础。     

Evaluation of methods for leak detection in reactor primary systems

No currently available, single leak-detection method combines optimal leakage detection sensitivity, leak-locating ability, and leakage measurement accuracy. Technology is available to improve leak detection capability at specific sites by use of acoustic monitoring. However, current acoustic monitoring techniques provide no source discrimination (e.g., to distinguish between leaks from pipe cracks and valves) and no leak-rate information (a small leak may saturate the system).Seven cracks, including three field-induced IGSCC specimens and two thermal-fatique cracks, have been installed in a laboratory acoustic leak detection facility. The IGSCC specimens produce stronger acoustic signals than the thermal-fatigue cracks at equivalent leak rates. Despite significant differences in crack geometry, the acoustic signals from the three IGSCC specimens, tested at the same leak rate, are virtually identical in the frequency range from 300 to 400 kHz. Thus, the quantitative correlations between the acoustic signals and leak rate in the 300–400 kHz band are very similar for the IGSCC specimens. Also, acoustic background data have been acquired during a hot functional test at the Watts Bar PWR. With these data, it is now possible to estimate the sensitivity of acoustic leak detection techniques.     

Explosive welding method for manufacturing ITER-grade 316L(N)/CuCrZr hollow structural member

In this study, a new explosive welding method provided an effective way for manufacturing ITER-grade 316L(N)/CuCrZr hollow structural member. The welding parameters (stand-off distance and explosion rate) were calculated respectively using equivalent frontal collision wave model and effective energy model. The welded samples were subject to two step heat treatment cycles (solution annealing and aging). Optical microscopy (OM) and scanning electron microscopy (SEM) were utilized to analyze the microstructure of bonding interface. The mechanical properties of the welded samples were evaluated through microhardness test and tensile test. Moreover, the sealing property of the welded specimens was measured through helium leak test.Microstructural analysis showed that the welded sample using effective energy model had an ideal wavy interface. The results of microhardness test revealed an increase in hardness for both sides near to the bonding interface. And the hardening phenomenon of interface region disappeared after the solution annealing. SEM observation indicated that the samples with the post heat treatments exhibited a ductile fracture with dimple features after tensile test. After the specimens undergo aging strengthening, there was an obvious increase in the strength for all specimens. The helium leak test results have proven that the welded specimens are soundness.     

Technical note Leak detection in nuclear piping outside containment

The results of an experimental investigation performed at Wyle Laboratories to evaluate various methods for detecting small leaks in high energy piping system is described. These experiments were designed to support the leak-before-break methodology currently being employed by the United States nuclear industry. This methodology requires that:

1. (1) the flow rate through a hypothetical leak be accurately predicted, and

2. (2) the lower limits of detecting the flow rate through such a leak be established.

The research described in this work was designed to establish experimentally this limit of detectability.The experiments performed covered a range of leakage flow rates between 0.1 and 5 gpm (378–18925 cm³/min through small penetrations in a 6-in. diameter carbon steel pipe. Insulation, typical of the types found in nuclear power plants, covered the test pipe.The key observation made is that leak rates down to 0.1 gpm (378 cm³/min) are easily detectable.