Uncertainties in fatigue life prediction and a rational definition of safety factors

To cope with uncerainties in mechanical and structural design, enigineers exercise their judgement through the use of safety factors based on service experience and laboratory data on relevant design parameters. Using the problem of fatigue life prediction as a vehicle, the relationship between the size of a safety factor and the associated risk and cost-benefit estimates of the engineering judgement based on new technical information, is demonstrated. The subtle influence of the choice of a distribution function for a given set of data is exhibited by comparing the gaussian with the three-parameter Weibull fits of a set of fatigue life data on 6061-T6 aluminum. A system of ranking the importance of different sources of uncertainties based on an analysis of service data is proposed along with an example to “refine” the system using up-to-date laboratory and field measurements. The concept of a rational definition of safety factors as a tool for engineers who design under uncertainty is discussed.     

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

This paper presents a method for evaluating “response factors” of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design response to actual response. This method has the following characteristic features: (1) the components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components.This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups.     

Dancoff factors with partial absorption in cluster geometry and their effects on unit cell multiplication

In the present work, Dancoff factors for perfectly (Black) and partially (Grey) absorbing fuel rods are calculated by the collision probability method, in cluster cells with square outer boundaries. Comparisons are made with the equivalent cylindricalized cell, considering specular and white boundary conditions for the square and cylindrical cases, respectively, using the WIMSD code. The results show the expected asymptotic behaviour of the solution with increasing cell sizes. Effective multiplication factors are also computed and the differences reported for the cases using the perfect and partial absorption assumptions.     

Experimental and numerical methods related to safety factors in the PEC fast nuclear reactor

In this paper the author summarizes the activity of structural analysis related to the safety of the PEC fast nuclear reactor. There are two principal aspects of safety concerning problems of structures: the localized incident and the hypothetical core disruptive accident (HCDA).With regard to the first point, the phenomenon is dependent on the hydrodynnamic and structural behaviour of the fuel elements. With regard to the HCDA, it is necessary that the reactor vessel is able to absorb the explosive energy, whereas the plug must not sustain movements such as to alter the overall seal of the installation.Given the complexity of the phenomena, therefore, it was considered necessary first of all to carry out numerous experimental tests on both full-size and reduced scale models. The experimental tests on the individual hexcan, on the group of seven hexcans and on the vessel were carried out at the EURATOM centre of Ispra, in the context of a collaboration agreement between ENEA and EURATOM.Some of the results of these tests are presented in this paper, together with relevant comparisions with the numerical values.     

Important factors affecting fault detection coverage in probabilistic safety assessment of digital instrumentation and control systems

As digital instrumentation and control (I&C) systems are gradually introduced into nuclear power plants (NPPs), concerns about the I&C systems’ reliability and safety are growing. Fault detection coverage is one of the most critical factors in the probabilistic safety assessment (PSA) of digital I&C systems. To correctly estimate the fault detection coverage, it is first necessary to identify important factors affecting it. From experimental results found in the literature and the authors’ experience in fault injection experiments on digital systems, four system-related factors and four fault-related factors are identified as important factors affecting the fault detection coverage. A fault injection experiment is performed to demonstrate the dependency of fault detection coverage on some of the identified important factors. The implications of the experimental results on the estimation of fault detection coverage for the PSA of digital I&C systems are also explained. The set of four system-related factors and four fault-related factors is expected to provide a framework for systematically comparing and analyzing various fault injection experiments and the resultant estimations on fault detection coverage of digital I&C systems in NPPs.     

Extension of Load Follow Capability of a PWR Reactor by Optimal Control

The problem of extending that part of the fuel life cycle during which a reactor is capable of sustaining load-follow operation is formulated as an optimal control problem. A two-node model representation of pressurized water reactor dynamics is used, leading to a set of non-linear ordinary differential equations. Differential Dynamic Programming is used to solve directly the resulting nonlinear optimization problem and obtain the trajectories of soluble boron concentration and control rod insertion. Results of computations performed for a reference reactor are presented, showing how the optimal control policy stretches the capability of the reactor to follow an average daily load curve towards the end of the fuel life cycle.     

Mode-C从基负荷到负荷跟踪的过渡策略研究

Mode-C在基负荷运行时与Mode-G类似,而在负荷跟踪运行时与MSHIM类似,综合了Mode-G和MSHIM的优点。但为满足反应性控制需求,Mode-C在循环末期由基负荷向负荷跟踪过渡时容易出现轴向偏移等难以控制的问题。在负荷变化过程中,控制棒抽插、中子注量再分布和氙反馈等影响反应性和轴向偏移控制的各因素之间是紧密耦合的,难以采用将各影响因素分离后单独计算分析的方法。为此,以某双环路压水堆的典型日负荷跟踪为例,通过对临时调硼、提前调硼和基负荷深插K棒组3种策略进行直接模拟计算研究。结果表明:只有基负荷深插K棒组策略可快速平稳过渡到负荷跟踪工况,且该策略还有循环寿期长、过渡时调硼量少及无需修改参考轴向功率差等优点。     

Examination of oil paintings by using a PIXE/PIGME combination

The applicability of the simultaneous use of particle induced X-ray and gamma-ray emission (PIXE/PIGME) for nondestructive analysis of art paintings has been investigated. By proper choice of detection geometry and proton energy the PIXE/PIGME combination yields almost covering elemental analysis of inorganic pigments even through the surface varnish of the painting. Accordingly the method is very suitable for example for a rapid detection of pallette errrors.     

Reduced scale PWR passive safety system designing by genetic algorithms

This paper presents the concept of “Design by Genetic Algorithms (DbyGA)”, applied to a new reduced scale system problem. The design problem of a passive thermal-hydraulic safety system, considering dimensional and operational constraints, has been solved. Taking into account the passive safety characteristics of the last nuclear reactor generation, a PWR core under natural circulation is used in order to demonstrate the methodology applicability. The results revealed that some solutions (reduced scale system DbyGA) are capable of reproducing, both accurately and simultaneously, much of the physical phenomena that occur in real scale and operating conditions. However, some aspects, revealed by studies of cases, pointed important possibilities to DbyGA methodological performance improvement.     

The spirit of safety: oriental safety culture

Failure of a large system causes disasters. However, after an accident, the causes are frequently attributed to human error when the operators do not survive the accident. It might be difficult to prove that the real cause of the accident is human error. Process decision program chart (PDPC) would be a useful tool in indicating the causes of an accident since it can clearly show that if the operator made the correct choice, the safety of the system could be maintained. The case of the incident of the nuclear reactor at Mihama Unit 2 is indicated by PDPC in which the sequence of events and the operations are indicated in this paper together with the safe operation. One can easily understand the cause of the incident and the way to avoid it. Also, PDPC for the Three Mile Island (TMI) accident is shown. Initially, in order to prevent an accident, mental training and safety culture is most important. The oriental safety culture based on Zentoism, a school of Buddhism is discussed.     

Precise nitrogen depth profiling by high-resolution RBS in combination with angle-resolved XPS

Nitrogen depth profiling in a high-k gate stack structure, SiON/HfO₂/SiON/Si(0 0 1) was performed by high-resolution Rutherford backscattering spectroscopy (HRBS) in combination with angle-resolved X-ray photoelectron spectroscopy (AR-XPS). The nitrogen depth profile is determined so that both the HRBS spectrum and the angular dependence of the XPS yield are reproduced. The obtained nitrogen profile is compared with the result of high-resolution elastic recoil detection (ERD) which is the most reliable technique for depth profiling of light elements. The agreement between the result of the present combination analysis and that of high-resolution ERD is fairly good, showing that the present combination analysis is a promising method for the analysis of light elements.     

Summary of NET plant probabilistic safety approach and results by means of ENEA fusion plant safety assessment (EFPSA)

This paper summarizes the probabilistic safety assessment for the main accident scenarios associated with failures originating in the In-Vessel Plant Area of the Next European Torus (NET). The assessment refers to the Basic Performance Phase of operation under normal running and conditioning. For the corresponding accident sequences, the values of the annual expected frequency and the seriousness of consequences expressed as early dose to the Most Exposed Individual (MEI) of the public are listed.     

300MW和1000MW核电站安全壳氢燃烧负载估算

美国联邦法规10CFR§50.34(f)规定：当100％燃料包壳金属-水反应产生的氢气释放进入安全壳,并且伴随着氢气燃烧时,安全壳必须能维持其完整性。本文在保守地假设氢气在安全壳内等体积、绝热、完全燃烧的基础上,根据热平衡方程和理想气体方程推导出事故后安全壳内氢气燃烧引起的压力负载估算方法,并用此方法分别对国内新设计的300MW和1000MW核电站估算了100％燃料包壳金属-水反应产生的氢气在安全壳内燃烧引起的压力负载。计算结果表明两者的安全壳设计均能满足相关法规要求。     

Determination of view factors by contour integral technique

This paper presents the application of contour integral technique to derive the diffuse radiation view factor expressions (analytical) for elements of nuclear reactor fuel bundle. The cases considered are: (i) view factor between two cylindrical rods of equal diameter and finite length, (ii) view factor between two cylindrical rods with interference by another rod and (iii) view factor between a cylindrical rod and a non-concentric cylindrical enclosure. The contour integral method is significantly more accurate than the area-integration method. The view factor results based on the analytical expressions derived for these finite length geometries are compared with that of the exact expressions available in the literature for infinite length. It is observed that the use of infinite length approximations in finite length cases can lead to significant errors.     

一种复合火灾探测器

介绍了一种由电离室和温度传感器组合成的复合型火灾探测器,这种探测器内置单片微机,在相应的软件中含有火灾特征参数。最后,给出了复合型火灾探测器的试验结果。     

Addressing the challenges posed by advanced reactor passive safety system performance assessment

Passive safety features play an essential role in the development of nuclear technology and within advanced water cooled reactor designs. The assessment of the reliability of such systems in the frame of plant safety and risk studies is still an open issue. This complexity stems from a variety of open points coming out from the efforts conducted so far to address the topic and concern, for instance, the amount of uncertainties affecting the system performance evaluation, including the uncertainties related to the thermal-hydraulic (T-H) codes, as well as the integration within an accident sequence in combination with active systems and human actions. These concerns should be addressed and conveniently worked out, since it is the major goal of the international community (e.g. IAEA) to strive to harmonize the different proposed approaches and to reach a common consensus, in order to add credit to the underlying models and the eventual out coming reliability figures. The main key points that may influence the reliability analysis are presented and discussed and a viable path towards the implementation of the research efforts is delineated, with focus on T-H passive systems.     

利用人外周血淋巴细胞染色体畸变估算离体模拟局部照射剂量

多数电离辐射事故均为局部照射。对于局部照射剂量估算,国际原子能机构(IAEA)推荐采用Dolphin's 模型,该模型需推算出照射部位染色体畸变率,再代入离体均匀照射情况下建立的剂量效应曲线来估算局部照射剂量。准确推算照射部位染色体畸变率对于估算剂量十分重要,选用合适的剂量效应曲线对估算剂量也同样重要,而对于局部照射,关于不同剂量率剂量效应曲线对估算结果影响的报道还十分有限。基于此本研究利用人外周血淋巴细胞染色体畸变估算离体模拟局部照射剂量,分析不同剂量率剂量效应曲线对估算结果的影响。利用⁶⁰Co γ射线离体照射人外周血样品(样本A和样本B),剂量分别为1 Gy和5 Gy。将照射血与未照射血按25%和75%比例混合以模拟局部照射,分析混合血样中淋巴细胞的双着丝粒染色体(dicentric chromosome,dic)加着丝粒环(r),利用Dolphin's模型估算局部照射dic+r率,并用剂量率不同的两种剂量效应曲线估算局部照射剂量。结果显示,大部分混合血样dic+r分布不符合泊松分布,为过离散分布。利用与实际照射剂量率一致的剂量效应曲线估算的样品受照剂量大多与实际照射剂量比较接近,相对偏差在10%以内,但两样本的1 Gy 25%组的估算受照剂量与实际照射剂量偏差较大。利用与实际照射剂量率不一致的剂量效应曲线估算的样品受照剂量与实际照射剂量相对偏差均超过10%。结果表明dic+r分析用于估算离体模拟局部照射的剂量有可行性,采用剂量率和照射剂量率一致的剂量效应曲线估算的结果更为准确。