A quantitative methodology for reactor vessel pressurized thermal shock decision making

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Consideration of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS.A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation.Once the specific event sequences of concern are identified, detailed thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. This paper addresses key aspects of the thermal-hydraulic and fracture mechanics analyses of the reactor vessel. The effects of incomplete mixing of safety injection flow in the primary cold leg and vessel downcomer and the application of warm prestressing are emphasized. The results of these analyses are being used to define further modifications in vessel and plant system design and to operating procedures.Previous design considerations that have evolved as a result of reactor vessel integrity evaluations are mentioned. These include the development of realistic design analysis tools and selection of plant system modifications. Modifications that are being developed or are under consideration are also mentioned. These include vessel fluence reductions, additional modifications to operating procedures, increased use of probabilistic event sequence and fracture mechanics analysis methods, enhanced material fracture toughness, and reductions in the severity or frequency of occurrence of dominant reactor vessel PTS transients.     

Implication of PSA uncertainties on risk-informed decision making

Nuclear power plants risk-informed policy is introduced in order to improve safety decision making and regulatory efficiency. The corresponding regulatory guides define the acceptable risk measures and their changes resulting from the modifications in the licensed design of the nuclear power plant. The risk measures used in the acceptance guidelines are the core damage frequency and large early release frequency.The risk measures and their corresponding changes are assessed by the Probabilistic Safety Assessment. The uncertainties of Probabilistic Safety Assessment should be appropriately addressed in the context of the decision making, considering their implication on the obtained results. The Probabilistic Safety Assessment uncertainties include epistemic uncertainties resulting from parameter, model, and completeness uncertainties.The paper presents the obtained results from the uncertainty analysis of the Probabilistic Safety Assessment of the reference nuclear power plant and their implication on risk-informed decision making. The paper focuses particularly on parameter and model uncertainties. The analysed modification is extension of the test interval of the emergency diesel generators. The core damage frequency is the used risk measure in the analysis.The need for the appropriate consideration of the uncertainties in the Probabilistic Safety Assessment in order to adequately support the risk-informed decision making is identified. The deficiency of usage of percentile measures is identified and acknowledged. The need for the adaptation of the risk-informed decision-making principles considering new nuclear power plants is recognized.     

PRA for decision making? — Some issues of the reliability data

The development of probabilistic safety analysis methods offers the decision maker a potentially powerful tool to assist in the decision making process. In practice there is considerable discussion about the validity of PRA methodology and issues of uncertainty in the PRA results are raised as a possible barrier to its usefulness. In this paper some of the issues related to the use of reliability data in PRA are discussed and it is noted that while such data is uncertain a proper appreciation of the situation can still lead to useful results being produced.     

Seismic isolation for nuclear power plants: Technical and non-technical aspects in decision making

Technical aspects of seismic isolation systems show merit for their use in nuclear power plants. Less quantifiable non-technical aspects must be evaluated in the decision to employ a seismic isolation system.First, non-technical aspects are discussed. An historical and applications perspective is given, and it is suggested that the number of applications of seismic isolation systems is correlated with the amount of research activity in this area. For nuclear plants, it is suggested that application of seismic isolation systems is in part related to standardized plant designs in high seismic regions. Also, for nuclear plants, it is suggested that direct capital cost, enhanced seismic safety, regulatory licensing and unknown locations of nearby active faults are all factors which can weigh in favor and/or not in favor for seismic isolation application.Second, technical aspects are discussed. The technical results show that seismic isolation reduces building response, and reduces floor response spectra/equipment response. These results combine in application to reduce seismic risk and thus enhance safety for nuclear plants.     

Onboard assessment of XRF spectra using genetic algorithms for decision making on an autonomous underwater vehicle

In order to optimise use of the limited resources (time, power) of an autonomous underwater vehicle (AUV) with a miniaturised X-ray fluorescence (XRF) spectrometer on board to carry out in situ autonomous chemical mapping of the surface of sediments with desired resolution, a genetic algorithm for rapid curve fitting is reported in this paper. This method quickly converges and provides an accurate in situ assessment of metals present, which helps the control system of the AUV to decide on future sampling locations. More thorough analysis of the available data could be performed once the AUV has returned to the base (laboratory).     

全局决策树方法在核电站人员可靠性分析中的应用研究

以核电厂人员可靠性分析中的全局决策树方法及其应用研究为主要内容。重点介绍全局决策树方法的基本框架、背景和实施步骤。结合秦山二期核电厂的人员可靠性现场访谈形成影响因子和品质等级描述,并以SLOCA和SGTR为例进行全局决策树方法分析。该方法以图形化决策树的形式,综合考虑了各个影响因子对人误概率的影响。分析表明该方法能够适应人误分析过程中出现的不确定性,具有较高的可靠性和实用性。     

The decision making method of task arrangement based on analytic hierarchy process for nuclear safety in radiation field

Unstructured and multi-objective task arrangement in radiation field is a complicating problem decision-makers are most concerned about. Proper task arrangement is helpful to improve efficiency of engineering and reduce hazard. The goal of this work is to discuss some criteria of task, build task hierarchy structure model, and then propose a task analytic method in radiation field based on analytic hierarchy process (AHP). The proposed method is useful for selection of schemes to reduce hazard and increase efficiency of task. In addition, we use this proposed method solve a hypothetical case, which is utilized to explain how the method works and to prove the method effective.     

Application of fuzzy decision making method to the evaluation of spent fuel storage options

In this paper, a fuzzy decision making method is proposed which is based on the fuzzy set theory and the hierarchical structure analysis. The method employs two key concepts: linguistic variables and fuzzy numbers. The linguistic variables and fuzzy numbers are used to represent the decision-maker's subjective assessments for the decision criteria and the decision alternatives versus the decision criteria. The fuzzy mean operator is used to aggregate the decision-maker's subjective assessments and the total integral value method is used to rank the decision alternatives. As a case study, the method is applied to evaluating the spent fuel storage options in Korea.     

Analyzing the decision making process of certifying digital control systems of nuclear power plants

Safety-critical computing systems need regulators’ approval before operation. Such a permit issue process is called “certification”. Digital instrumentation and Control (I&C) certification in the nuclear domain has always been problematic and lengthy. Thus, the certification efficiency has always been a crucial concern to the applicant whose business depends on the regulatory decision. However, to our knowledge, there is little basic research on this topic. This study presents a Regulatory Decision-Making Model aiming at analyzing the characteristics and efficiency influence factors in a generic certification process. This model is developed from a dynamic operational perspective by viewing the certification process as an evidence–confidence conversion process. The proposed model is then applied to previous nuclear digital I&C certification experiences to successfully explain why some cases were successful and some were troublesome. Lessons learned from these cases provide invaluable insights regarding to the regulatory review activity. Furthermore, to utilize the insights obtained from the model, a prototype of a computer-aided licensing support system has been developed to speed up review evidence preparation and manipulation; thus, regulatory review efficiency can be further improved.     

一种针对多机组事故的操作干预水平应急决策辅助系统

操作干预水平(OIL)是指对应于某一特定公众防护行动通用干预水平的易于测量的辐射量,用于在核事故期间直接与监测结果比较给出应急防护行动建议。福岛核事故后,我国核应急准备和响应向着多机组方向发展,本研究提出了一种针对多机组事故优化的OIL应急决策辅助系统,将每个机组发生的事故后果进行单独计算和评价,最后将它们对环境辐射后果的贡献进行迭加并代入原先计算公式中,即得到多机组事故下的OIL值。在优化方法的基础上,系统实现了单机组和多机组事故情景下的OIL查询、计算、修正以及后续的应急辅助决策功能。目前,该系统已应用于秦山、田湾、宁德、阳江等核电基地,其模式和功能在实际应用中得到了不断验证和完善,为核事故应急提供了重要的辅助决策手段和工具。     

A new method for on-line surveillance of nuclear power reactors based on decision theory

In this paper a new method for the automatic monitoring of reactor operational states based upon decision theory is introduced. The observation space is represented by spectral density functions estimated in discontinuous time intervals. After an empirical mathematical modelling of the process and its hierarchical structuring, a combined identification, estimation and detection algorithm is described. The classification of each subsystem is given via a sequential hypothesis test. Furthermore, the possibility of an adaptive learning or training is demonstrated. After a logical, system specific association of the resulting pattern, a diagnostic prognosis about the system state is made which contains integral information about a time interval extending in the past. The implementation of this method is illustrated by an example and the preliminary results are presented.     

Use of uranium hexafluoride for making 1,2-difluorotrichloroethane (freon-122A)

All-Union Refrigeration Engineering Research Institute and Chemical Research Institute, Nizhegorod University. Translated from Atomnaya Énergiya, Vol. 72, No. 2, pp. 192–195, February, 1992.     

A study on the calculation of stopping power and CSDA Range for incident positrons

In this work, we have given a new method to calculate stopping power for incident positrons in a number of different targets. Our method are based on the Gümü?’s method for incident electrons. We have calculated the collisional and radiative stopping power and then found the total stopping power of targets. For collisional stopping power, we used Gümü?’s method for electrons but adapted it to the positrons. Tsai’s method has been used to calculate the radiative stopping power and after the founding the total stopping power with these two values, we used the Continuous Slowing Down Approximation to obtain the range values. Calculation was done for Al, Si, Ag, LiF, CaF₂ and H₂O targets and results were compared with the ICRU 37 report and Penelope 2005 code and found that agreement is less than 10%.     

A simple model for quantifying the degree of layer-by-layer growth in low energy ion deposition of thin films

Layer-by-layer growth of thin films can be promoted by using low energy ion deposition (LEID) techniques. The basic process affecting the growth are often quite diverse, but often the ion impact induced inter layer mass transfer processes due to adatom insertion to lower step edges or pile-ups to step edges above dominate. In this paper we propose a simple phenomenological model which describes the growth of thin films in LEID under these conditions. The model makes possible to distinguish the dominant growth, the detection of the transition from the 3D growth to 2D growth, and it can be used to quantify the degree of layer-by-layer growth. The model contains only two parameters, which can be phenomenologically related to the properties of the bombarding ion beam.     

A fuzzy decision tree method for fault classification in the steam generator of a pressurized water reactor

This paper extends a method previously introduced by the authors for building a transparent fault classification algorithm by combining the fuzzy clustering, fuzzy logic and decision trees techniques. The baseline method transforms an opaque, fuzzy clustering-based classification model into a fuzzy logic inference model based on linguistic rules which can be represented by a decision tree formalism. The classification model thereby obtained is transparent in that it allows direct interpretation and inspection of the model. An extension in the procedure for the development of the fuzzy logic inference model is introduced to allow the treatment of more complicated cases, e.g. splitted and overlapping clusters. The corresponding computational tool developed relies on a number of parameters which can be tuned by the user to optimally compromise the level of transparency of the classification process and its efficiency. A numerical application is presented with regards to the fault classification in the Steam Generator of a Pressurized Water Reactor.     

应用γ能谱法找金研究

本文介绍应用γ能谱法在山东某矿区找金的可行性研究。 从实测结果可知,放射性元素在不同岩性中的含量以蚀变带最高,混合花岗岩次之,石英(矿)脉中最低。在横穿各条矿脉的测量剖面上,都出现了与矿脉相对应的Th、K及Tc(放射性元素总含量)低值异常,三者各自的峰背比值分别为0.5—1.0、0.3—0.7和0.3—0.7之间,都能反映矿脉的存在。     

放射性碘连续监测系统判断阈和探测限的确定

结合中国辐射防护研究院研制的PIG连续监测系统中对放射性碘的采样、测量和数据分析方法,按ISO规定的判断阈和探测限的计算原则确定了PIG连续监测系统放射性碘监测部分的判断阈和探测限.     

A preliminary design study for the HYPER system

In order to transmute the long-lived radioactive nuclides such as transuranics (TRU), Tc-99, and I-129 in LWR spent fuel, a preliminary conceptual design study has been performed for an accelerator driven subcritical reactor system, called HYPER (HYbrid Power Extraction Reactor). The core has a hybrid neutron energy spectrum which includes fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU are loaded into the HYPER core in a TRU–Zr metal form because a metal type fuel has very good compatibility with the pyro-chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb–Bi is chosen as a primary coolant because Pb–Bi can provide a good spallation target and produce a very hard neutron energy spectrum. As results, the HYPER system does not need any independent spallation target system. 9Cr–2WVTa is used as a window material because this advanced ferritic/martensitic steel is known to have a good performance in the highly corrosive and radiative environment. The support ratios of the HYPER system are about 4–5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4–5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is facing with.