国内运行核电厂经验反馈在新建电厂应用的研究

考虑到运行核电厂的经验反馈对新建同类型核电厂的借鉴意义,列举了几项前一阶段运行核电厂提出的重要修改申请,并对修改中涉及到的各种改进方案加以介绍,同时对其在新建核电厂中的适用性进行了探讨。     

Experience with inspection qualifications for austentitic piping

The performance demonstration initiative (PDI) was formed in 1991 to address implementation of appendix VIII to Section XI of the ASME Boiler and Pressure Vessel Code. All US utilities and three foreign utilities are participating. Appendix VIII differs from previous Code approaches in that it does not specify a particular approach. It does require that the capabilities of the personnel, procedures, and equipment be demonstrated. Appendix VIII describes, in detail, the demonstration requirements and acceptance criteria for ultrasonic examinations. Piping demonstrations have been performed for 390 examiners since April 1994. The database of demonstration results includes 10 000 detection and length sizing data points and 5000 depth sizing data points, which are available for analysis. The performance of these candidates provides insight into the difficulties of the inspection process for austenitic piping. The length and depth sizing accuracy along with the detection rate as a function of false call rate will be presented. The results of recent investigations will also be described.     

Evaluation of piping damage in German nuclear power plants

Information relating to piping damage in safety-relevant systems in German nuclear power plants with light water reactors (both pressurized water reactors (PWRs) and boiling water reactors (BWRs)) were analyzed with respect to the modes and the causes of damages. In general, the total range of observed piping damage is low. The incidents (82) in plants with PWRs affected mainly pipes with small diameters. Almost all damaged piping showed wall-penetrating cracks combined with leakages, which revealed the damage. Initial cracks at piping with larger diameters were discovered in isolated cases during in-service inspections. With regard to the incidents (71) in plants with BWRs, piping with small as well as large diameters was affected to different degrees. Wall-penetrating cracks combined with leakages were detected at piping with small diameters. For large-diameters pipes, cracks were indicated during in-service inspections and supplementary examinations. The results of the incident evaluations confirm the conservativeness of the safety concept chosen for the design of German nuclear power plants with light water reactors.     

A proposal for an aseismic design method of equipment and piping for NPPs in a low seismicity area

Recently a regulatory code for an aseismic design of high-pressure gas facilities became effective by the order of the Ministry of International Trade and Industry (MITI) in Japan. This order includes details of the aseismic design of vessels whose “factor of importance” are relatively lower than Class A (Class I) items in nuclear power plants.The author develops his idea on an aseismic design method of equipment and piping of nuclear power plants in a Low Seismicity Area (LSA) based on his experience of the new code for petro-chemical industries and oil refinaries pertaining to high pressure gas facilities mentioned above.The definition of LSA is usually the area whose maximum intensity has never exceeded MMI VI or VII. However, there are two types of LSA, one is really such a low seismicity area, and the other type is the area which has the possibility of stronger earthquake occurrence than those mentioned above, even though it is low. One of the typical examples is the area subjected to “New Madrid Earthquake-1812”. The author develops his concept along these two lines.He briefly describes the new code for high-pressure gas facilities in Japan. This code describes the design methodology of both types of aseismic design analysis, that is, rather sophisticated dynamic methods for facilities whose potential hazard is as high as those in a nuclear power plant, such as liquified chlorine gas storage, and simplified dynamic and static methods for most of the equipment and vessels in those plants. One of the features of this code is the use of design formulae and charts to simplify their design procedure as well as the set of specific computer codes by the MITI. These computer codes are prepared by the MITI or approved by the MITI for providing equivalent capability to the practice designated in the MITI order.The author's philosophy for the code of equipment and pipings in LSA is that they must be as simple as possible, and most of the analytical work for the design should be eliminated, or at least limit the use of simplified methods, such as the static seismic coefficient method or the modified seismic coefficient method with a simplified response spectrum. The use of general design criteria or a guideline of structural details may be better than a sophisticated design analysis as a result.     

Some aspects on the optimum repository plug design with respect to the escape of gas

BfS is in the progress of developing a closure concept for the repository for low and intermediate level radioactive waste in Morsleben (ERAM). In the course of this work, the optimal design of the plug is currently being evaluated with respect to gas escape and the exchange of potentially contaminated brine through the plug. For the sealing to behave well in the long term, it is important that the gas formation processes do not disrupt the plug or enhance the radionuclide release, e.g. by means of excessive pressure build-up. The object has been to study different scenarios for gas and brine transport for two alternative plug concepts, by using the multi-phase flow model TOUGH. Rock convergence due to creep has been included in the modelling. The results of the calculations indicate that the closure concepts restrict he exchange of brine and allow escape of gas; an excavation-damaged zone around tunnels is a potential pathway for gas and brine, and the effect of the rock convergence is small. The results also indicate that a very dense plug results in excessive pressurisation of the repository, whereas a permeable plug results in an increased exchange of brine.     

Design of nuclear facilities other than NPPs in relation to external events

The safety of a nuclear installation requires in general that it is sited, designed, constructed and operated to protect individuals, society and the environment against an uncontrolled release of radioactivity. External events, both natural and human-induced, play a major role in challenging the plant defense. Therefore, appropriate design provisions are needed to assure an adequate safety margin in case of such events.In recent years, the development of design criteria, design methodologies and assessment approaches for external events received major emphasis for nuclear power plants. Other nuclear installations, however, received less attention even though their radioactive inventory may be quite significant (sometimes comparable with the NPPs, like in the case of some research reactors or fuel re-processing plants). Also the risks for radiological (and chemical) contamination is often rather high (as in the case of the fuel re-processing plants) and their location may be very close to densely populated areas.There is a lack of generally accepted international standards in this field and the direct application of general safety principles for nuclear installations is not straightforward.The IAEA addressed this issue in the past several years. Some of the results have been collected in an IAEA Technical Document (TECDOC).This paper is intended to address some of the main issues that have been identified and discussed in the referenced document.     

核电站反应堆保护机柜失电缺省值分析研究

为了降低反应堆保护机柜(RPC)失电引入的安全风险,红沿河核电站开展了针对RPC失电的缺省值分析工作,论文在简要介绍红沿河核电站数字化仪控系统(DCS)平台的基础上,对RPC失电相关的缺省值分析范围进行了界定,通过实例对其分析原则进行了介绍,对其实现方式及应用进行了说明。该研究对提升DCS本身的可靠性、电站的安全水平和可用性有重要意义。     

Cracking in stabilized austenitic stainless steel piping of German boiling water reactors—characteristic features and root cause

Cracks have been found in the welds of piping systems made from stabilized austenitic stainless steels in German boiling water reactors (BWR). In the course of the intensive failure analysis metallographic examinations, microstructural investigations by electron microscopy, corrosion experiments and welding tests have been performed. The results show that cracking under the given medium conditions is due to intergranular stress corrosion cracking (IGSCC) in those parts of the heat affected zone (HAZ) which are overheated during welding and where solution of titanium carbides and subsequent precipitation of chromium carbides and depletion of chromium along the affected grain boundaries could occur.     

Dosimetric characteristic of VVR-M reactor with respect to tritium

Translated from Atomnaya Énergiya, Vol. 67, No. 2, pp. 111–113, August, 1989.     

A CFD approach to the atmospheric dispersion of radionuclides in the vicinity of NPPs

Most studies of atmospheric dispersion of radionuclides released from Nuclear Power Plants (NPPs) are based on Gaussian plume models or on the use of a convection–diffusion equation. Such methods, which do not involve solving the flow problem, are useful in the atmospheric mesoscale, of the order of 2–2000 km from the NPP. However, they do not account for the turbulence generated by the interaction of the wind with obstacles and with the released material stream, which are the dominant factors in the local scale, of the order of 0–2 km from the source of emission. Here, the authors advocate the use of computational fluid dynamics (CFD) to study the dispersion problem. The physical model comprises the Navier–Stokes equations, a convection–diffusion energy equation, and transport equations for the radionuclides. The paper details the stabilized finite element formulation used, stressing its connection with the variational multiscale/large eddy simulation approach. Adaptive techniques combining error estimation and remeshing are also employed. The method is implemented on a Beowulf parallel computing system using domain decomposition and the message passing interface (MPI). Controlled emissions from a chimney and release from severe accidents have been simulated, showing the importance of the local phenomena on the dispersion problem.     

Experimental strategy of fuel performance testing with respect to PCI

Basic metallurgical investigations have revealed that stress corrosion cracking of Zircaloy tubes requires critical stress and iodine concentration for a minimum of time. Based on this observation KWU has structured its experimental strategy on the RSST approach. This means the evaluation of a defect-free power Range below a PCI defect threshold and a defect-free LHGR Step (naturally beyond the threshold), a limited Speed of power increase if both limits are exceeded, and a minimum Time for any mechanism to become effective.KWU has initiated a large ramp test program at HFR, Petten, the results of which are backed by the participation in international ramp test programs at Studsvik. The first target was the determination of the failure threshold as a function of burnup for the different fuel rod designs. Then, the allowable safe speed for passing the defect threshold was investigated. The defect-free range was confirmed by power reactor experiments on a broad statistical basis with rods of original length. In forthcoming experiments in Petten the verification of safe steps shall be a point of main priority. The permissible time above the thresholds may be controlled purely by the crack nucleation. In the Studsvik Demo Ramp II Program this is a point of special consideration.Detailed PIE results show that these performance limits can well be interpreted by observable phenomena like grain growth, fission product redistribution, fission gas release etc.     

Economic competitiveness of the new generation of NPPs with NP-500 units in Russia

An economic analysis of NPPs with the new design of the average capacity unit (NP-500) developed in Russia is carried out. The design is characterized by the application of passive safety systems, and a double protective containment, that allows a decrease in the probability of a severe accident by 2 or 3 orders of magnitude in comparison with present VVER units, operated at modern NPPs. The NP-500 unit has a capacity of 635 MWe; it is more compact, and therefore it has a smaller specific consumption of materials and a smaller number of regular staff. Licensing of the design on the basis of international practice is now nearing completion. The published data on costs by the Joint Parallel Nuclear Alternatives Study (JPNAS), executed for the US-Russia Joint Commission on Economic and Technological Cooperation, as well as recent forecasts by different authors for the period 1995–2010 on power consumption, specific investments in thermal power plants and the costs of organic fuels for different regions of Russia are used. An original methodology for analysis of equilibrium prices of energy (marginal costs), competition for developing power technologies and accounting for a factor of inertia in power systems is described. The total self-sufficiency of Russia with nuclear fuel for the forecasting period and the necessity of adequate development of the organic power infrastructure are taken into account. The individual rates of development of the different competitive energy technologies and appropriate returns on capital are determined. The competitiveness of NPPs with NP-500 units in different regions of Russia is analyzed. The attractiveness of investment in this objective is emphasized.     

Evaluation of damping values of a piping system with regard to damping values used in design codes

To investigate the damping behavior of an in-situ structure a piping system was subjected to seismic-like loads during the SHAM test series at the HDR test facility in Kahl, FRG. Tests were run using random excitation to identify natural frequencies, mode shapes and damping values for different restraint configurations with MDOF system identification methods. The damping values were achieved concerning the stiffness of the respective restraint configuration. For one restraint configuration tests with earthquake-like excitation were evaluated by fitted parameter calculations for different load levels. This study served to determine the damping values as a function of excitation and stress levels. Both sets of evaluations cover the spectrum of restraint configurations used in practice, trace the load increase, and provide possibilities for comparing the results with pipe damping values used in current design practice.     

Sensitivity of reactivity insertion limits with respect to safety parameters in a typical MTR

The sensitivity of various safety parameters, affecting the reactivity insertion limits imposed by clad melting temperature for a typical pool type research reactor, have been investigated in this work. The analysis was done for low enriched uranium (LEU) core with scram disabled conditions. The temperature coefficients of fuel and coolant, void/density coefficient and β_eff were individually varied and the reactor behavior for different ramp reactivity transients was studied. In this work ramp reactivity insertions from 1.6 to 2 $/0.5 s were selected and peak power, maximum fuel, clad and coolant temperatures were determined. Results show that peak power decreases with an increase in the Doppler coefficient of reactivity. However, it rises with an increase in the reactivity insertion. Core remains insensitive to the coolant temperature coefficient of reactivity for ramps in the range of 1.6–1.9/0.5 s. Peak power decreases with an increase in the void coefficient of reactivity (0.1 $/%void to 0.8 $/%void). With a decrease in the void coefficient of reactivity, the maximum fuel and clad temperatures show a non-linear rise. Power and temperature peaks in the transient are sensitive to the values of β_eff. Finally, it can be concluded that LEU is a safe core due to its smaller β_eff, larger Doppler coefficient and void coefficient of reactivity. It is inferred through this work that reactivity insertion limits of LEU core are quite insensitive to β_eff, the Doppler coefficient and the coolant temperature coefficient of reactivity. They are highly sensitive to the change of the void coefficient of reactivity in the core.     

Safety study on HIC containing waste resin with respect to hydrogen release

To explore the behavior of radiolytically produced hydrogen release from the waste resin stored in a high integrated container(HIC), and the mechanism of hydrogen diffusion in a near-surface disposal facility, both experimental studies and numerical simulations were performed through an accelerated irradiation test and simulated disposal, respectively. Results indicated that,100 years after disposal, the highest hydrogen concentration appeared in the cell where the HICs were placed. The volume fraction for different scenarios postulated in the numerical simulation was 2.64% for Scenario 1, 2.28% for Scenario 2, and 3.965% for Scenario 3, all of which are lower than the hydrogen explosion limit of 4.1%. The results indicated that the simulated HIC disposal scheme is safe.