Advanced modularity design for the MIT pebble bed reactor

The future of all reactors will depend on whether they can be economically built and operated. One of the major impediments to new nuclear construction is the capital cost due in large part to the length of construction time and complexity of the plant. Pebble bed reactors offer the opportunity to reduce the complexity of the plant because the number of safety systems required is significantly reduced due to the inherent safety of the technology. However, because of its small size, the capital cost per kilowatt is likely to be large if traditional construction approaches are followed. This strongly suggests the need for innovative construction concepts to reduce the construction time and cost. MIT has proposed a modularity approach in which the plant is pre-built in space-frame type modules which are built in factories. These space frames would contain all the equipment contained in a given volume. Once equipment in the space frame is installed, the space frame would then be shipped to the site and assembled “lego-style.” Studies presently underway have demonstrated the feasibility of the concept. Thermal stress analysis has been performed and an integrated design with the space frames has been developed. It is expected that this modularity approach will significantly shorten construction time and expense. This paper proposes a concept for further development, not a final design for the entire plant.     

Ultimate strength of reinforce concrete shear walls under multi-axes seismic loads

“The model test on multi-axes loading on RC shear walls” had been carried out as for the 10-year project aiming at comprehension of the earthquake response behavior of three-dimensional (3D) reinforced concrete (RC) shear walls under the 3D of multi-axes earthquake loading condition. The motivation of the project building-up is that the current seismic design of nuclear power plant building is carried out by applying one-dimensional (1D) dynamic earthquake load to an analytical building model in each direction independently whereas actual earthquake jolts the building in the three directions simultaneously. Therefore, there were opinions requesting some testing confirm whether or not the current seismic design methodology is reliable for the input motions exceeding the design earthquake ground motion moreover for the input motions of the 3D. The project had completed with various valuable outcomes that can reply to the opinions. Moreover, the outcomes will play an important role in evaluating seismic margins of important structures in a nuclear power plant. In this paper, based on the published documents relating to this test project, the author describes a review of the whole testing and summarizes the major outcomes extracted by the test project.     

Important viewpoints proposed for a safety approach of HTGR reactors in Europe: Final results of the EC-funded HTR-L project

The inherent safety features of modular High Temperature Reactors (HTRs) make events leading to severe core damage highly unlikely and constitute the main differentiating aspects compared to LWRs. Furthermore, while a known and stable regulatory environment has long been established for Light Water Reactors (LWRs), different ways of thinking may help to develop a more appropriate licensing process for HTR-based power plants.The HTR-L project funded by the European Commission in the 5th Framework Programme was dedicated to the definition of a common and coherent European safety approach and the identification of the main licensing issues for the licensing framework of the modular HTRs. Several topics were developed during the course of this project.Due to the characteristics of the HTR design, it has been necessary to define specific defence-in-depth requirements which have then been evaluated for implementation in the safety approach. Safety-related functions appropriate for the HTR design have also had to be identified and listed.On one hand, the different possible solicitations of the fuel particles constituted the starting point for the identification of the accidental conditions (by means of the Master Logic Diagrams methodology); these accidental conditions were classified and the most appropriate methods to consider ultra low probability severe accidents were examined.On the other hand, the elements constituting the source term and, in particular, the requirements for the confinement of radioactive products and the conditions required to prevent the need for a “conventional” containment structure have been discussed.In the definition of the safety approach, attention has been paid to the need to maintain the potentially interesting economic perspectives of HTR reactors. Key issues to be addressed in the licensing process of the HTRs have also been identified. An innovative systems, structures and components classification method has been developed and rules that will govern equipment qualification proposed.     

Design tools for computer-generated display of information to operators

More and more computers are being used to process and display information to operators who control nuclear power plants. Implementation of computer-generated displays in power plant control rooms represents a considerable design challenge for industry designers. Over the last several years, the Electric Power Research Institute has conducted research aimed at providing industry designers tools to meet this new design challenge. These tools provide guidance in defining more “intelligent” information for plant control and in developing effective displays to communicate this information to the operators.     

Methodologies for rapid evaluation of seismic demand levels in nuclear power plant structures

A methodology for rapid assessment of both acceleration spectral peak and “zero period acceleration” (ZPA) values for virtually any major structure in a nuclear power plant is presented. The methodology is based on spectral peak and ZPA amplification factors, developed from regression analyses of an analytical database. The developed amplification factors are applied to the plant's design ground spectrum to obtain amplified response parameters. A practical application of the methodology is presented.This paper also presents a methodology for calculating acceleration response spectrum curves at any number of desired damping ratios directly from a single known damping ratio spectrum. The methodology presented is particularly useful and directly applicable to older vintage nuclear power plant facilities (i.e. such as those affected by USI A-46). The methodology is based on principles of random vibration theory. The methodology has been implemented in a computer program (SPECGEN). SPECGEN results are compared with results obtained from time history analyses.     

The DEEPSSI project, design, testing and modeling of steam injectors

The DEEPSSI project is a steam injector research programme. Among thermal-hydraulic passive systems, the steam injectors (also called “condensing ejectors” or “steam jet pumps”) are very interesting apparatus with very specific characteristics (high velocity, very low pressure). The envisaged reactor application is the Steam Generator Emergency FeedWater System (EFWS) of Pressurised Water Reactors (PWRs). The heart of this project is the development and the testing of an innovative steam injector design. Three experimental facilities are involved: CLAUDIA in France, IETI in Italy and IMP-PAN in Poland. In these facilities, different design options have been tested and some significant improvements of the initial design have been obtained.In addition to the experimental studies, the development of a steam injector computational model has been undertaken in order to model industrial systems based on steam injectors. The one-dimensional module of the system code CATHARE2 has been chosen to be the basis of this model. The first results obtained have confirmed the capabilities of CATHARE2 to describe the steam injector thermal-hydraulics.     

Methods and requirements for seismic design in the United Kingdom

The United Kingdom is in an area of low but significant seismicity compared with the more active areas of the world where there are major active faults or tectonic plate boundaries. This paper presents the methods and requirements that are adopted to consider the extreme load in the design of nuclear facilities. In the United Kingdom, detailed procedures for demonstrating seismic adequacy are not specified by the nuclear licensing authority and as such the methods described in this paper are based on precedents arising from recent licensing applications. In presenting the method and requirements, the paper discusses the applicability of simplified methods for seismic qualification for both “new” and “existing” facilities. The paper concludes that simplified methods are applied to a significant extent for demonstrating the adequacy of existing plant. However, for new plant these methods have been limited in some cases to the evaluation of design loads and to the qualification of items where the required degree of assurance is less than that associated with formal qualification and for supporting studies which do not directly affect design. It is expected that as the body of experience in earthquake engineering develops in the United Kingdom, there will be a greater tendency to adopt more simplified procedures with a greater degree of confidence.     

An earthquake isolation system for steam generators in nuclear power plants

This paper describes an experimental/analytical study of the effectiveness of base isolation and damped interaction between a model of a steam generator and its primary housing structure in a nuclear power plant subjected to earthquake ground motion. The design of the test generator model, its connection to the primary structure by yielding elements and the influence of such yielding restrainers on the response of the generator are included. Details of an optimal design problem for selection of the “best” combination of isolation and energy absorption devices are presented and their effectiveness demonstrated.     

Applicability of structural wall test results to seismic design of nuclear facilities

A review of tests on earthquake-resistant reinforced concrete structural walls is presented. Laboratory tests of isolated walls and construction joints are discussed. Where appropriate, design recommendations are given. The review indicates only few experimental data are available for short walls which are directly applicable to nuclear power plant design. In particular, tests of short rectangular walls subjected to load reversals are needed. Tests are also needed to determine the damping and frequency characteristics of cracked short walls. Analytical and experimental results should be correlated so that the hysteretic response observed in tests can be realistically related to the analytical response “demand” of nuclear power plant structures.     

The development of a diagnostic and monitoring system for vibrating pipe branches

This paper deals with a diagnostic and monitoring system for assessing the integrity of pipe branches, during the operation of the nuclear power plant. This system have been developed under the concept of “easy to use without any sophisticated analysis” and “portable”. The accuracy of the diagnosis is based on the model optimization subsystem, which automatically modifies the numerical vibration model so as to fit its natural frequency to the actual natural frequency. The information obtained by this system may be reflected to a maintenance program of the plant to assure more reliable operation of the plant.     

A critical reappraisal of nuclear power plant safety against accidental aircraft impact

The overall problem of nuclear power plant safety against an accidental aircraft impact is discussed in relation with its structural analysis and design. Associated risks, such as fire, which is a potential source of damage for buildings and other structures, are not considered.The paper is divided in two parts. In part I different approaches used for determining the reaction-time curve are discussed. The influence on the results of target motions is examined next. It is shown that for the evaluation of structural response an aircraft-structure interaction analysis is usually an unnecessary refinement, “mean” reaction-time and impact area-time curves being sufficient to define the excitation. Preliminary results for oblique impact are also given. Since the conditional probability of a normal impact is very small, the consideration of oblique impact may become acceptable in future design criteria.In part II, available solutions for the resulting structural dynamic problem are reviewed. The feasibility of resorting to a static analysis is also discussed. Present practices to evaluate floor response spectra are reviewed next. The short-comings of the “deterministic” approach are pointed out. It is proposed to define the excitation as a mean plus a fluctuating force. The latter is treated as a nonstationary random process and the problem solved by numerical integration in the time domain. Although such solutions get prohibitively expensive when the number of degrees of freedom becomes large, results obtained for simple models may help to clarify which are the important variables of the problem.     

Nuclear power development in market conditions with use of multi-purpose modular fast reactors SVBR-75/100

This article presents an innovative nuclear power technology, based on the use of modular type fast-neutron reactors SVBR-75/100 having heavy liquid-metal coolant, i.e. eutectic lead–bismuth alloy, which was mastered in Russia for the nuclear submarines’ reactors. Reactor SVBR-75/100 possesses inherent self-protection and passive safety properties that allow excluding of many safety systems necessary for traditional type reactors. Use of this nuclear power technology makes it possible to eliminate conflicting requirements among safety needs and economic factors, which is particularly found in traditional reactors, to increase considerably the investment attractiveness of nuclear power based on the use of fast-neutron reactors for the near future, when the cost of natural uranium is low and to assure development of nuclear power in market conditions. On the basis of the factory-fabricated “standard” reactor modules, it is possible to construct the nuclear power plants with different power and purposes. Without changing the design, it is possible for reactor SVBR-75/100 to use different kinds of fuel and operate in different fuel cycles with meeting the safety requirements.     

Structural integrity of lmfbrs including leak before break

The German Basis Safety Concept is an approach which allows the possibility of catastrophic failures to be excluded. It was developed in Germany to render the probabilistic approach unnecessary for safety cases relating to nuclear power plants. The process of evaluation started in 1972, and in 1979 the Basis Safety Concept was officially published and thus became a legal requirement for LWR plants. With appropriate modifications in regard of the particular features of LMFBR, this concept has also been applied to SNR 300. The “Structural Integrity Demonstration Concept” of SNR 300 is based on five principles:

• - Principle of quality by design and fabrication

• - Principle of multiple examination

• - Principle of worst case consideration

• - Principle of operating surveillance and documentation

• - Principle of verification and continuous development.

The same principles are taken over for SNR 2.The specific requirements on the components relevant to safety have to be defined at an early stage so that the components can be designed appropriately to the feasibility of the measures required by the concept.     

Inherently safe containments for nuclear power plants

Nuclear energy cannot be avoided in the near future. To regain public acceptance the safety of nuclear power plants has to be increased. Consequently, feasibility studies have been carried out for a containment proposal for future pressurized water reactors which will keep people unharmed even in the case of severe nuclear accidents under the assumption “all that can go wrong, will go wrong”. The main features of the design concept are a core melt cooling and retention device, a passively acting cooling system to remove the decay heat and a double-wall containment which is able to withstand high static and dynamic internal pressures due to hydrogen detonation. Internal structures are designed to resist extreme loadings resulting from various accident scenarios including in-vessel steam explosion and vessel failure under high system pressure.     

Basic risk analyses for high-temperature reactors

The methodology of PSA/PRA is available for the HTR and has already been applied to various plant concepts. The results are predictive and generic in nature; the analyses have to struggle with less detailed technical information (paper design instead of real operated plants) and little experience from plant practice. The overall degree of uncertainty is similar to studies for LWRs mainly because operating experience can be transferred to some extent and the physical phenomena are much easier to describe. Therefore, the topology of design and beyond-design accidents has been established.For medium-sized HTRs (e.g. HTR-500) of current design failure of active systems for decay heat removal, resulting in core heatup, clearly dominates the risk and leads to the largest releases of radioactive nuclides into the environment. For small-sized HTRs (e.g. HTR-Module) temperature-induced releases from the fuel are insignificantly low for all types of accident; plate-out activities on the steam generator surfaces remobilized in the course of water ingress accidents can be regarded as the main contribution to the comparatively small source term.The largest releases are so low for all HTR concepts that early health effects can be ruled out in any case, including no evacuation. For small HTR plants even late cancer effects need practically not to be expected.A comparison with licensed released values has shown that the applicable current requirements are met by all HTR concepts examined. However, small HTRs especially offer an additional potential for compliance with more stringent safety requirements, “taking the fear out of hypothetical accidents”, by limiting maximum releases. Incidentally, the classically defined “risk” to the population from both plants is generally very low.     

Experimental investigation into the seismic behavior of nuclear power plant shear wall structures

Nuclear power plant structures are designed to resist large earthquakes. However, as new data are obtained on earthquake activity throughout the United States, plant design earthquake levels have increased. The U.S. Nuclear Regulatory Commission is sponsoring an analytical-experimental research program to obtain information on the structural response of nuclear power plant shear wall structures subjected to earthquake motions within and beyond their design basis. Using different size scale models constructed with microconcrete and prototypical concrete this research has demonstrated consistent results for measured values of stiffness at load levels within the design basis. Furthermore, the values are well below the theoretical stiffnesses calculated from an uncracked cross-section strength-of-materials approach. Current program emphasis is to assess the credibility of previous experimental work by beginning to resolve the ‘stiffness difference’ issue.     

Japanese regulatory practices on licensing, inspection, operational supervision and related standards

Japanese view on the safety of nuclear power plants is based on the concept that the primary responsibility for securing safety lies on electric power companies, installers of reactors.Under this concept, the Ministry of International Trade and Industry (MITI), in the course of designing and construction, has been performed an examination of the basic design and the detailed design of nuclear power plants, and in each stage of construction, a pre-operational inspection process. In addition, MITI, in operating stage, has been made throughgoing investigations on the causes of troubles and incidents as well as accidents that may affect operation, forcing utilities to take measures to prevent recurrence, and implementing safety regulation based on the “preventive maintenance” including elaborate checkings and overhaulings at the periodical inspections conducted for a period of three to four months after every 12-month operation cycle under the laws and regulations.This paper discusses the current status of nuclear power development in Japan, safety regulatory systems, views on safety and future prospects of securing safety.     

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.