A heuristic model for risk and cost impacts of plant outage maintenance schedule

Cost and risk are two major competing criteria in maintenance optimization problems. If a plant is forced to shutdown because of accident or fear of accident happening, beside loss of revenue, it causes damage to the credibility and reputation of the business operation.     

Availability of a probabilistic cost estimation for the price effect of Cu powder and bentonite on an HLW disposal cost in Korea

This study aims to demonstrate the availability of a probabilistic cost estimation related to the price effects of Cu powder and bentonite. From a sensitivity analysis of those materials’ prices on the overall disposal costs, it was found that Cu powder was a more dominant cost driver than that of bentonite among the material costs to dispose of 52,000 tU of spent fuels by the deterministic cost estimation method even though the used volume of Cu powder will be smaller than that of bentonite, whereas those conclusions can be changed by a probabilistic cost estimation method. Namely, its conclusion depends on a decision maker's personal opinion because of the resultant uncertainties. The disposal cost includes too many uncertainties due to the long construction and operational durations of a repository. Therefore a probabilistic cost estimation can be useful to provide the information related to an uncertainty.     

The neutronic potential of nuclear power for long-term radioactivity risk reduction

An analysis to evaluate the comparative neutronic transmutation potential of different nuclear power system (standard or advanced fission reactors and accelerator driven hybrids) is presented. The analysis is based on an evaluation of neutronic constraints for the reduction of both long-lived fission product toxicity and fuel waste toxicity integrated over the life of the nuclide families, taking into account the overall neutron balance of the systems being considered.     

Basic factors to forecast maintenance cost and failure processes for nuclear power plants

Two types of maintenance interventions are usually administered at nuclear power plants: planned and corrective. The cost incurred includes the labor (manpower) cost, cost for new parts, or emergency order of expensive items. At the plant management level there is a budgeted amount of money to be spent every year for such operations. It is very important to have a good forecast for this cost since unexpected events can trigger it to a very high level. In this research we present a statistical factor model to forecast the maintenance cost for the incoming month. One of the factors is the expected number of unplanned (due to failure) maintenance interventions. We introduce a Bayesian model for the failure rate of the equipment, which is input to the cost forecasting model. The importance of equipment reliability and prediction in the commercial nuclear power plant is presented along with applicable governmental and industry organization requirements. A detailed statistical analysis is performed on a set of maintenance cost and failure data gathered at the South Texas Project Nuclear Operating Company (STPNOC) in Bay City, Texas, USA.     

压水堆核燃料组件价格形成及对核发电成本的影响初探

从分析核燃料组件价格的形成入手,讨论了核燃料组件价格的形成特点及其对核发电成本的影响。     

Radiation removal of synthetic estrogens in aqueous solution: influence of reduction or oxidation system and toxicity test

Estrogens as a kind of steroidal sex hormone are widely used in humans, especially quinestrol(QS),dienestrol(DS) and norethindrone(NET, 19-nor-17-alphaethinltestoster-one), which cannot be completely degraded after application. Steroidal estrogens at low concentration pulling into environment can disturb the normal biological function of wide life and thus lead to great threat to humans. So it is important to explore its degradation mechanism and its behavior in the environment. In this study, we investigated the oxidation or reduction system under gamma irradiation for reducing estrogenic activity in the aqueous solutions as well as degradation kinetics, its by-products and yield of transformation by different analytical methods such as GC–MS and HPLC. Gamma irradiation could effectively degrade estrogens in aqueous solution. The degradation reaction of estrogens could be depicted by first-order reaction kinetics. The total organic carbon of solution decreased with an increasing absorbed dose with the order: quinestrol [ norethindrone [ dienestrol. The toxicity of the three estrogens was declined after irradiation. Mono- and quadric-hydroxylated intermediates as well as organic acids were formed after gamma irradiation.     

Decreasing snubber inservice inspection costs through snubber reduction and improved test limits

Snubber inservice inspection (ISI) requirements, along with a history of snubber malfunctions, has made inspection and maintenance of snubbers a significant part of a nuclear power plant's ISI budget. These expenses can be minimized through snubber reduction and the use of improved test limits for snubber functional testing. This paper presents a snubber overview and reviews snubber ISI requirements. Examples are given of the high cost that maintaining a snubber in an operating nuclear plant represents.Snubber reduction refers to reducing a plant's snubber population by eliminating snubbers shown not to be required to restrain piping for design basis dynamic loadings, and by replacing snubbers with other types of restraints, such as rigid struts. Snubber reduction is discussed in terms of what makes removing snubbers practical along with approaches to, and results of recently implemented snubber reduction programs.Improved or increased test limits for snubber functional testing are discussed along with an approach to, and results of an Electric Power Research Institute sponsored program to develop improved limits that would not significantly affect piping response. Improved piping acceptance criteria can be used to justify the use of increased test limits provided by snubber manufacturers. An additional use is to justify the operability of piping on which faulty snubbers were found.     

Characteristics of test and maintenance human errors leading to unplanned reactor trips in nuclear power plants

The test and maintenance (T&M) human errors involved in unplanned reactor trip events in Korean nuclear power plants were analyzed according to James Reason's basic error types, and the characteristics of the T&M human errors by error type were delineated by the distinctive nature of major contributing factors, error modes, and the predictivity of possible errors. Human errors due to a planning failure where a work procedure is provided are dominated by the activities during low-power states or startup operations, and human errors due to a planning failure where a work procedure does not exist are dominated by corrective maintenance activities during full-power states. Human errors during execution of a planned work sequence show conspicuous error patterns; four error modes such as ‘wrong object’, ‘omission’, ‘too little’, and ‘wrong action’ appeared to be dominant. In view of a human error predictivity, human errors due to a planning failure is deemed to be very difficult to identify in advance, while human errors during execution are sufficiently predictable by using human error prediction or human reliability analysis methods with adequate resources.     

压水堆核电厂换料物理启动试验优化研究和应用实践

核电厂设计和运行相关核安全法规、导则要求核电厂换料后必须进行物理启动试验。随着堆芯换料设计日趋成熟,试验程序和试验方法得到充分检验。为提升运行经济性,各核电厂设计和运行人员不同程度地开展了换料后物理启动试验优化的研究与实施。本文基于压水堆核电厂监管要求和核电厂运行要求分析,针对物理启动试验优化提出了定性评价、物理分析和试验验证的系统性论证方法,并以秦山核电厂320 MWe机组为例,进行了完善的研究与可行性论证。实施物理启动试验优化后,核电厂换料大修时间大幅缩短,相比以往可提前约2天进入满功率运行,显著提高了核电厂运行负荷因子,提升了运行经济性。     

A study of risk reduction for some mitigation features for PWRs

The risk reduction attainable with mitigation features in a large-dry pressurized water nuclear reactor (PWR) is evaluated. The calculations are made in a probabilistic risk analysis framework, and they are based on Zion Probabilistic Safety Study (ZPSS). Some of the modifications made to this study are also taken into account.The mitigation designs considered consist of features for simultaneously controlling late containment overpressure, containment basemat penetration, and hydrogen burning. The individual mitigation features include: a passive containment heat removal system (PCHRS), a filtered-vented containment system (FVCS), a core ladle, and controlled hydrogen burning. Emphasis is placed on comparison of PCHRS and FVCS design options. The results include calculations of the sensitivity to several failure mode probabilities and to the probability of core meltdowns with containment bypass.     

Simulation and test of quench in CICC: The SMES experience

This paper summarizes the work done as part of the U.S. SMES program to simulate quench evolution on the 200 kA CICC developed by the Bechtel Team. As a large-scale CICC with a central tube, this work has led to a number of results applicable to other conductors sharing similarities with the SMES-CICC. The paper presents the evolution in computational models, since 1987 to date, and describes QUIPS, a test intended to validate these computer models. The paper concludes with observations on the directions in the field as perceived by the author.     

核燃料循环成本与核电的竞争力

国际市场核燃料价格变幻莫测,对我国核电的成本和发展产生影响,本文提出了控制整个核燃料循环成本的设想,以提升中国核电的竞争力,促进核燃料产业和核电产业的共同发展。     

A decade''s experience in chemistry maintenance of once through steam generator of fast breeder test reactor

The steam generator of fast breeder test reactor (FBTR) at Kalpakkam (India) is a once through Steam Generator (OTSG) which requires the feed water at high purity level. Therefore, for maintaining feed water chemistry, all volatile treatment (AVT) is adopted along with a full-flow deep bed Condensate Polishing Unit (CPU) in the steam water system. Operational difficulties such as premature termination of operation cycle of the CPU, enhanced impurity pickup resulting in increased load for CPU, early silica breakthrough, etc. were observed on occasions. This paper describes the modifications carried out in the steam water circuit to overcome these problems. A decade's experience in operating the CPU and maintaining the feed water quality is also discussed.     

The relationship between notched bar impact bending test, robertson test and drop-weight test

The temperature of the Robertson-Test referred to as crack-arrest-temperature determines the change of the fracture mode from brittle to tough. It was found out that for 22 NiMoCr 37 (similar to A 508 Cl 2) and St 37 steels this temperature was corresponding with the upper shelf temperature of the impact test. This is also the temperature at which there cannot be initiated any cleavage fracture (fracture transition characteristic temperature) in the modified drop-weigh-test, and it is corresponding with the FTE-point for large wall thickness in the FAD by Pellini which is also 65–70 K above NDT. Thick walled components as well can be considered at this temperature to be absolutely safe in terms of brittle fracture, if possibly generated brittle areas of base or HAZ material will remain localized in small isolated volumina. The latter requirement has to be verified by adequate manufacturing.     

Particle swarm evolutionary computation-based framework for optimizing the risk and cost of low-demand systems of nuclear power plants

In this paper, an adapted multi-objective multi-swarm co-evolutionary particle swarm optimization (PSO) framework is developed to simultaneously optimize the risk and cost of low-demand systems of nuclear power plants (NPPs). In the built framework, multi-swarm co-evolutionary strategy is introduced to handle the fitness assignment puzzle of multi-objective optimization problems. Besides, to deal with the mixed-integer problem of the decision variables vector, a sub-interval covering-based nearest boundary method is also adopted. To illustrate the effectiveness and efficiencies of the proposed method, a typical high-pressurized injection system (HPIS) is analyzed. The results indicate that, compared with the classic non-dominated sorting genetic algorithm (NSGA)-II approach, the proposed method is more simple and easier to be convergent, besides, of which the Pareto front is better distributed.     

基于最小化发电成本的聚变驱动次临界堆参数优化分析

利用自主发展的聚变驱动次临界堆 (FDS)系统分析软件SYSCODE ,通过遗传算法求解以最小化发电成本为目标函数 ,以聚变功率、中子壁负载、安全因子和包层平均热功率密度限等为约束条件的最优化模型 ,得到FDS的包层能量增益因子 (Qb) ,等离子体位形参数 :环径比 (A)、拉长比 (κ)和三角变形因子 (δ) ,工程与物理参数 :规一化比压 (βN)、轴上纵场 (BT)与等离子体电流 (IP)等的最优化设计 ,并详细分析了发电成本对最优设计点的敏感性。分析分两种情况 :一是聚变功率为定值 (设为 1 5 0MW ) ;二是包层平均热功率密度为定值 (分别设为 1 5、5 0、1 0 0和 5 0 0MW /m3)。     

The EPR: A clear step forward in dose reduction and radiation protection

For the European Pressurized Water Reactor (EPR) a large effort was made to improve the plant design with respect to radiation protection using the experience gained during the design of former generations of pressurized water reactor (PWR) in France and Germany, and their current operation. Keeping the radiation exposure of personnel to an acceptable level is one of the main objectives of the EPR design. Both the individual and the collective doses are considered.Internationally comparable limits based on recommendations of the International Commission on Radiological Protection (ICRP) have been established for individual doses. These limits describe the framework within which the individual dose shall be kept as low as possible, applying the principles:

• Justification: No practice involving exposures to radiation should be adopted unless it produces sufficient benefit to the exposed individuals or to society to offset the radiation detriment it causes.

• Optimization: In relation to any particular source within a practice, the magnitude of individual doses, the number of people exposed, and the likelihood of incurring exposures where these are not certain to be received should all be kept as low as reasonably achievable (ALARA), economic and social factors being taken into account.

• Limitation: The exposure of individuals resulting from the combination of all the relevant practices is subject to dose limits.

The paper describes the design provision and measures introduced in the plant design to achieve the above described goals.They are in essence:

• measures to avoid or to reduce sources of radiation;

• layout aspects;

• provisions made in the component design with respect to ease of operation and maintenance management;

• improved possibilities of decontamination;

• use of operating experience for design improvements.

The radiation protection layout principles compiled on the basis of safe operating experience gained from the existing pressurized water reactors in France and Germany are used to develop an improved plant design with respect to radiation protection aspects and dose optimization.Summary: The European Pressurized Water Reactor is an evolutionary third-generation pressurized water reactor with a rating in the 1600 MWe class. Its development was started in 1992 by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, now AREVA NP. Being the product of intense bilateral cooperation the EPR combines the technological accomplishments of the world's two leading PWR product lines—the French N4 reactors in operation at Chooz and Civeaux and the Konvoi reactors in operation at Neckarwestheim, Emsland and Isar in Germany. From the very start, development of the EPR was focused on improving plant safety and economics even further and also a large effort was made to improve the plant design with respect to radiation protection. Keeping the doses received by operating and maintenance personnel to a level far below the limiting values was one of the main objectives of the EPR design. Both the individual and the collective doses are considered in this article.     

The materials test station: A fast-spectrum irradiation facility

The United States Department of Energy is developing technologies needed to reduce the quantity of high-level nuclear waste bound for deep geologic disposal. Central to this mission is the development of high burn-up fuel with significant inclusion of plutonium and minor actinides. Different fuel forms (e.g., nitrides, oxides, and metal matrix) and composition are under study. The success of these cannot be judged until they have been irradiated and tested in a prototypic fast neutron spectrum environment. In 2005, the US Congress authorized funding for the design of the materials test station (MTS) to perform candidate fuels and materials irradiations in a neutron spectrum similar to a fast reactor spectrum. The MTS will use a 1-MW proton beam to generate neutrons through spallation reactions. The peak neutron flux in the irradiation region will exceed 1.2 × 10¹⁹ n m⁻² s⁻¹ and the fast neutron fluence will reach 2 × 10²⁶ n m⁻² per year of operation. Site preparation and test station fabrication are expected to take four years.