Evaluation methodology for seismic base isolation of nuclear equipments

An evaluation methodology for seismic base isolated nuclear equipments is proposed. The evaluation can be classified into two steps. In the first step, the seismic functional failure probability during the lifetime of equipment without base isolation devices is quantified in order to decide the applicability of the base isolated structure. The second step is comparative and calculates the ratio of the seismic failure frequency of the equipment without base isolation devices to that with them in order to evaluate the effectiveness of the base isolated structure. The sample evaluation considers the case of high voltage type emergency transformer with ceramic tubes.     

Assessment methodology for confidence in safety margin for large break loss of coolant accident sequences

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis (BEPU) is increasingly being used for deterministic calculation in NPPs. The PSA methodology integrates information about the postulated accident, plant design, operating practices, component reliability and human behavior. The deterministic and probabilistic methodologies are combined by analyzing the accident sequences within design basis in the event trees of a postulated initiating event (PIE) by BEPU. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event.     

Recent advances in nuclear power: A review

The rise in oil prices and the increased concern about environmental protection from CO₂ emissions have promoted the attention to the use of nuclear power as a viable energy source for power generation. This review presents the recent advances in the field of nuclear power and addresses the aspects of nuclear economics, safety, nuclear reactor design and spent fuel processing and waste management.     

Nuclear power plant seismic design—A review of selected topics

This paper presents a background review of the basis for the current seismic design criteria employed in the United States with particular attention given to the so-called double earthquake approach to seismic design. This paper also provides details of approaches used in other countries, namely Canada and Japan, which at least in part do not use the two earthquake concept in design.The paper begins with a brief presentation of background material relative to the approach employed in the seismic design of nuclear plants in the 1960's along with comments on the development of the current procedures. The next section contains a brief discussion of the criteria contained in Appendix A of 10CFR100 which today largely governs the seismic design of nuclear power plants in the U.S. The last section discusses effective versus instrument acceleration in design and observations pertaining to other approaches that might be employed in terms of selecting and carrying out the seismic design.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 3. Evaluation of seismic safety margin

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), had conducted a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate an actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, was used for this test. The test model and the results of pressure and leak tests are described in Part 1. Test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load–deformation relationship are described in Part 2. Part 3 reports the seismic design safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 will report simulation analysis results by a stick model with lumped masses.     

A comparison of background seismic risks and the incremental seismic risk due to nuclear power plants

The seismic risk for the continental United States, in terms of the expected annual number of deaths and severe injuries, and the expected property damage, is evaluated in this work. Probabilistic models and correlations are developed and used in the evaluations of the risks, accounting for such important variables as the variability of property values, damage factors and so on. In addition, the incremental seismic risk due to the presence of nuclear power plants is evaluated utilizing results and methods available in the literature. The results show that the incremental risk is generally very small compared to the background seismic risk, even if a very high probability for core melt is postulated.     

Approximate formulas to calculate the seismic response of light attachments to buildings

Simple approximate formulas are proposed to compute the maximum response of equipment or any other light secondary system attached to buildings subjected to earthquake ground motions. The formulas are derived on the basis of a modified version of the conventional response spectrum method and the consideration of building and attachment as one unit. Notwithstanding, they are expressed in terms of the independent dynamic properties of the two components and ordinates from the response spectrum of a specified ground motion. Secondary systems with multiple degrees of freedom attached to one or two arbitrary points of a supporting multistory structure may be considered. As presented, however, the formulas are restricted to cases in which the independent primary and secondary systems are linear elastic with classical modes of vibration, and the masses of the secondary system are small in comparison with those of the primary one. Their accuracy is verified by means of a comparative study with time-history solutions. In this comparative study, the approximate formulas yield an average error of about 4% and a maximum of about 22%.     

Preliminary perspectives gained from individual plant examination of external events (IPEEE) seismic and fire submittal review

As a result of the US Nuclear Regulatory Commission (USNRC) initiated Individual Plant Examination of External Events (IPEEE) program, every operating nuclear power reactor in the United States has performed an assessment of severe accident due to external events. This paper provides a summary of the preliminary insights gained through the review of 24 IPEEE submittals.     

Vertical soil-structure interaction of nuclear power plants subjected to seismic excitation

The vertical soil-structure interaction problem is investigated by coupling an N-mass lumped mass structure to a two-dimensional elastic half space. This problem is formulated as an integral equation of the Volterra type. Numerical results are obtained by iteration for an idealized threemass two-mode model of a nuclear power plant containment structure. The effects of interaction are evaluated by comparing free-field acceleration spectrum response curves with similar curves determined from foundation motion.     

Comparative methods for analysis of piping systems subjected to seismic motion

The dynamic analysis of a three-dimensional piping system of a nuclear power plant is conveniently performed through a finite element method. When the modal analysis is used, only the first few modes of vibration are computed for practical purposes. In this paper is proposed a method of residues which evaluates the neglected modes and combines them with the first calculated modes to estimate the total seismic response of the piping. This methods emphasizes the importance of the selected modes. When the approach is made through a time history input function, this latter is usually characterized by a combination of several recorded accelerograms, e.g. El Centro, San Francisco and Taft. The response of a particular piping has been evaluated by means of these two methods: the use of the modal approach will be strongly recommended due to its inherent advantage of economy and also computation time and reliability.     

Non-linear seismic response of base-isolated liquid storage tanks to bi-directional excitation

Seismic response of the liquid storage tanks isolated by lead-rubber bearings is investigated for bi-directional earthquake excitation (i.e. two horizontal components). The biaxial force-deformation behaviour of the bearings is considered as bi-linear modelled by coupled non-linear differential equations. The continuous liquid mass of the tank is modelled as lumped masses known as convective mass, impulsive mass and rigid mass. The corresponding stiffness associated with these lumped masses has been worked out depending upon the properties of the tank wall and liquid mass. Since the force-deformation behaviour of the bearings is non-linear, as a result, the seismic response is obtained by the Newmark's step-by-step method. The seismic responses of two types of the isolated tanks (i.e. slender and broad) are investigated under several recorded earthquake ground to study the effects of bi-directional interaction. Further, a parametric study is also carried out to study the effects of important system parameters on the effectiveness of seismic isolation for liquid storage tanks. The various important parameters considered are: (i) the period of isolation, (ii) the damping of isolation bearings and (iii) the yield strength level of the bearings. It has been observed that the seismic response of isolated tank is found to be insensitive to interaction effect of the bearing forces. Further, there exists an optimum value of isolation damping for which the base shear in the tank attains the minimum value. Therefore, increasing the bearing damping beyond a certain value may decrease the bearing and sloshing displacements but it may increase the base shear.     

Recent applications of nuclear microprobe techniques to microelectronics

Applications of nuclear microprobe techniques to microelectronics, in particular, two- or three-dimensional analysis of integrated circuit structures using micro RBS (Rutherford Backscattering) and IBIC (Ion Beam Induced Current) measurements are discussed. SEU (Single Event Upset) mapping and IBIC measurements in SRAMs (Static Random Access Memories) and DRAMs (Dynamic Random Access Memories) are reviewed together with charge carrier collection simulation and transient IBIC measurements. Importance of applications of microprobes to new microelectronic structures such as SOI (Silicon-on-Insulator) devices for future hand-held information systems are also discussed. Possible applications of SEU measurement as alternative experimental procedures to cosmic ray neutron strikes are discribed.     

Recent contribution of JET to the ITER physics

In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value δB_T = 0.08% up to δB_T = 1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular, β_N values above the ‘no-wall limit’ (β_N ～ 3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg.     

Recent and future upgrades to the DIII-D tokamak

Research on the DIII-D tokamak focuses on support for next-generation devices such as ITER by providing physics solutions to key issues and advancing the fundamental understanding of fusion plasmas. To support this goal, the DIII-D facility is planning a number of upgrades that will allow improved plasma heating, control, and diagnostic measurement capabilities. The neutral beam system has recently added an eighth ion source and one of the beamlines is currently being rebuilt to allow injection of 5 MW of off-axis power at an angle of up to 16.5° from the horizontal. The electron cyclotron heating (ECH) system is adding two additional gyrotrons and is using new launchers that can be aimed poloidally in real-time by an improved plasma control system. The fast wave heating system is being upgraded to allow two of the three launchers to inject up to 2 MW each in future experiments. Several diagnostics are being added or upgraded to more thoroughly study fluctuations, fast ions, heat flux to the walls, plasma flows, rotation, and details of the plasma density and temperature profiles.     

New Zealand seismic base isolation concepts and their application to nuclear engineering

This paper describes recent New Zealand experience with seismic base isolation, and in particular, the distinguishing feature of the New Zealand system - that of using mechanical energy dissipators to control response.There are presently 22 structures in New Zealand which use base isolation concepts for seismic protection. Nuclear power plants are not used in New Zealand, but the principles adopted for the above buildings and bridge structures are appropriate for nuclear structures and, as shown by use of hypothetical examples, significant benefits can be achieved for these structures also. The most important is a marked reduction for the in-structure acceleration spectra with consequential implications for the enhanced safety of secondary structures and equipment.The paper also reviews the experimental and analytical studies undertaken to validate the technique and addresses the issues of reliability and cost.     

A cost-effective methodology to internalize nuclear safety in nuclear reactor conceptual design

A new methodology to perform nuclear reactor design, balancing safety and economics at the conceptual engineering stage, is presented in this work. The goal of this integral methodology is to take into account safety aspects in an optimization design process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behavior during accidents (safety performance indicators), are synthesized on Design Maps. These maps allow one to compare the safety indicator with limits, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimization process, by means of additional rules to the neutronic, thermal-hydraulic, and mechanical calculations.An application of the methodology, implemented in Integrated Reactor Evaluation Program 3 (IREP3) code, to optimize safety systems of CAREM prototype is presented. It consists in balancing the designs of the Emergency Injection System (EIS), the Residual Heat Removal System (RHRS), the primary circuit water inventory and the containment height, to cope with loss of coolant and loss of heat sink (LOHS) accidental sequences, taking into account cost and reactor performance.This methodology turns out to be promising to internalize cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels.     

Development of the methodology and approaches to validate safety and accident management

The article compares the development of the methodology and approaches to validate the nuclear power plant safety and accident management in Russia and advanced industrial countries. It demonstrates that the development of methods of safety validation is dialectically related to the accumulation of the knowledge base on processes and events during NPP normal operation, transients and emergencies, including severe accidents. The article describes the Russian severe accident research program (1987–1996), the implementation of which allowed Russia to reach the world level of the safety validation efforts, presents future high-priority study areas. Problems related to possible approaches to the methodological accident management development are discussed.