Nuclear power plant seismic design—A review of selected topics

Nuclear Engineering and Design 79 (1984) 7-18 North-Holland, Amsterdam

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NUCLEAR POWER PLANT SEISMIC DESIGN--A REVIEW OF SELECTED TOPICS J.D. S T E V E N S O N

Senior Consultant, Stevenson & Associates, Cleueland, OH, USA R.P. K E N N E D Y

President, Structural Mechanics Associates, Newport Beach, CA, USA W.J. H A L L

Professor of Civil Engineering~ University of Illinois, Urbana, IL, USA Received 11 October 1983

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 seismicdesign 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.

1. Introduction

2. Background

The purpose of this report is to present a background review of the basis for the current seismic design criteria employed in the United Staes involving the so-called double earthquake approach to seismic design. The paper also provides observations as to other approaches that might be employed in handling the seismic design process. The report begins with a brief presentation of background material pertaining 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 observations pertaining to other approaches that might be employed in terms of selecting and carrying out the seismic design. Also presented is a brief discussion of some of the technical factors that would necessarily have to be considered in such a case.

One of the first major papers concerning seismic design of nuclear power plants was presented by Dr. G.W. Housner in the Proceedings of the Second World Conference on Earthquake Engineering held in Tokyo, Japan in 1960 [1]. In this paper he noted the following: "The special character of the earthquake safety problem of a nuclear power reactor is made clear by comparing with an ordinary coal-steam power generator. In the case of the coal-steam power generator, it is customary to classify the structures and equipment into two categories so far as earthquake-resistant design is concerned: a. Those components that are essential to the continuing operation of the power generator are designed to resist higher seismic factors than prescribed by the building ordinance. It is common practice in California to design these to resist a seismic loading of the order of 20 % g following the design procedures specified by California Building Codes.

0 0 2 9 - 5 4 9 3 / 8 4 / $ 0 3 . 0 0 © Elsevier Science Publishers B.V. ( N o r t h - H o l l a n d Physics P u b l i s h i n g Division)

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J.D. Stevenson et a L / Nuclearpowerplant seismic design

b. Those components of the installation that are convenient but are not absolutely essential to the operation are designed according to normal building code requirements which specify a seismic loading of the order of 10 % g." Housner goes on in this article to provide three classes of equipment. Class 1 includes those items designed by necessity to preclude a nuclear incident so that the probability of failure is essentially zero for the strongest probable earthquake (no percentage of gravity is given). Housner notes that Class 2 contains items which are essential to the operation of the power reactor but whose failure could not cause a nuclear incident. Such items might well be designed for 20% of g by California building code procedures. He also notes in a footnote that the actual strength of a structure depends not only on the code specifications, but also upon design procedures, allowable stresses, etc. Class 3 includes those structures that are convenient but not essential to the operation of the power generator. These structures might well be designed according to ordinary requirements of the building code. Later in this same paper, Housner goes on to say: "Because of the short history of recording strong ground motions, it is not possible to specify the strongest possible ground motion. In the United States various strong ground motions have been recorded and the strongest of these was recored at El Centro, California, during the earthquake of 18 May 1940. This record is usually taken to represent the strongest possible ground motion * in the highly seismic region of the United States (Zone 3). Ground motion one-half as intense is usually taken as the strongest probable ground motion in Zone 2; and the strongest probable ground motion in Zone 1 is taken to be one-half as intense as that in Zone 2 ** On this basis then, one can see that Dr. Housner was considering several levels of earthquake hazards for consideration in the design process. Although generally it appears that for any one class of equipment or structure one earthquake design level is noted as being applicable, for Class 1 structures safety factors of at least five are recommended, suggesting the need for an ample margin of safety. It is interesting to note that in same paper, Housner makes reference to the USAEC Handbook on Reactors and Earthquakes, with a note

* "The strongest possible sround motion may be significantly greater than this, perhaps two or three times as intense." ** Maximum zero period ground acceleration for El Centro, 1940 earthquake is equal to 0.33g.

that this handbook is now in the process of being published. The U.S. Atomic Energy Commission Handbook on Nuclear Reactors and Earthquakes [2], known informally as TID-7024, appeared in print in August of 1963. Although no statements of seismic design criteria precisely along the lines of those just noted in Housner's paper are found in the noted report, one does find towards the end of Chapter 1 in fig. 1.19 and fig, 1.21 a weighted average spectra composed of E1 Centro-1934, E1 Centro-1940, Olympia-1949 and Taft-1952 normalized to the average of the two components of the El Centro-1940 earthquake. These spectra, and in particular the acceleration spectra, are the ones that were routinely used to design nuclear power plants until the late 1960's and early 1970's. Concepts pertinent to employing spectra of different levels are discussed in TID-7024; for example, one portion of the discussion centers around three different spectra pertaining to site effects at different distances from a major earthquake source (fig. 1.25). The design procedures presented later in this same report are centered around the building code of California. The next major publication which addressed nuclear plant seismic design was authored by Housner and Hudson and appeared in 1966 [3]. In this paper, the authors clearly make a point about two earthquake design levels designated at MP and MC, referring respectively to Maximum Probable and Maximum Credible values, the Maximum Credible one being the larger. The authors go on to discuss the higher level of safety required in a nuclear power plant as compared to safety requirements in normal building design. They also discuss some factors related to the economics of the design problem and some pertinent observations about the required resistance of such structural systems. Although the authors identify two levels of earthquake design requirements, they do not explicitly identify the double earthquake requirement for nuclear power plant design. One can infer though that this is what they may have had in mind. This particular paper appeared at a time when other major advances in the development of seismic design criteria for nuclear power plants had already occurred, as will be observed in the material that follows. Another study of seismic design criteria for nuclear power reactors was performed by Dr. Newmark in 1964 *. Dr. Newmark was asked by the Atomic Energy * Newmark and Hall had undertaken some containment studies for Argonne National Laboratory on a research reactor in 1963.

J.D. Stevenson et al. / Nuclear power plant seismic design

Commission to look into the design criteria for a nuclear power plant being considered for the Bodega Bay, California site. Review of the existing records suggests that this design, at that time, was being considered for at least one large earthquake characterized by a predominant ground acceleration on the order of two-thirds gravity with high frequency acceleration spikes that could be considered to go up to l g or more, and a smaller earthquake similar to E1 Centro-1940. It should be noted that the Bodega Bay site was not used for a nuclear power plant. In all probability, had a plant been built, the design would have been required to accommodate differential ground motion (local faulting). To date, evidence suggest that no nuclear power plant in the world has been sited where seismic differential ground motion is expected. However, analyses have been performed which indicate that similar facilities can accommodate such surface movements in the range of one foot, and of course, all designs are made to accommodate normal static foundation differential settlement. The development of additional seismic design criteria for nuclear power plants was rather rapid beginning in 1964. It was noted in criteria review reports in 1965, in connection with the Dresden Nuclear Power Station Unit No. 2, that the two earthquake design approach was in place and employed for that particular station. The major paper by Dr. Newmark that summarized the status of earthquake resistant design in the United States and that formed the basis for the criteria used at that time and in subsequent years was presented at an International Atomic Energy Agency (IAEA) panel meeting on "Aseismic Design and Testing of Nuclear Facilities" held in Tokyo, Japan, in 1967. The draft of Newmark's paper which was dated 3 February 1967 [4] was circulated widely at that time to AEC and to key earthquake engineering designers and researchers around the United States. This paper, slightly modified, was published as a part of the proceedings of that conference [5] and clearly delineates the two earthquake approaches, namely, a so-called "design" earthquake and a "maximum credible" earthquake. The paper goes on to discuss the meaning of these terms and their use in design and design analyses. Newmark notes in this paper that there are only a few areas of the country where there is a relatively long period of recorded observations of strong motion earthquake intensities. He also notes that through correlation of the strong motion records and through use of qualitative reports of various effects one can gain some measure of the maximum intensities that have occurred in various regions. In other regions of the country where

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records are scarce, Newmark suggests that estimates of a similar nature can be inferred. Such considerations lead to what he defines as the design earthquake, one that can be expected to be experienced within a reasonable period of time. He goes on to indicate that it is more difficult to determine the maximum possible intensity of an earthquake and that the maximum or extreme earthquake might never have occurred in the recent past or probably has a not occurred within the period of recorded history. Newmark continues by suggesting that one must make some estimate of the maximum possible intensity of the earthquake hazard which the structure must survive, and he calls this earthquake the "maximum credible" earthquake. Interestingly, at that same conference in another paper Housner [6] treats many of the important problems that need to be considered in designing a nuclear power plant. However, Housner does not address the problem of the seismic hazard in other than general terms, he notes that a higher degree of safety and a more uniform factor of safety is required for nuclear power plants. Another publication resulting from the Tokyo Conference was issued by IAEA in 1968 [7]. This publication summarizes the findings of the conference an denotes under U.S. practice "Two intensities of design earthquakes are designated here as (a) design basis earthquakes and (b) maximum potential earthquake." In summary, at this point it seems clear from the foregoing that Housner and Hudson clearly played a formative role in putting forth conceptual thoughts about multiple level earthquake desing concepts, but that Newmark played a major role in implementing the two earthquake concept with regard to the design processes employed by the U.S. Atomic Energy Commission and subsequently by its successor the U.S. Nuclear Regulatory Commission. By way of furthrr history, the ne:~t paper that played a major role in earthquake design concepts was the paper presented by Newmark and Hall in Santiago, Chile, at the Fourth World Conference on Earthquake Engineering in 1968 [8]. This paper incorporated a formal approach for construction of response spectra, clearly delineated the double earthquake approach to design - - at that time the higher level earthquake was denoted as the design basis earthquake and the lower level earthquake was denoted as the operating basis earthquake. In addition, in 1971, Dr. Newmark presented a paper on seismic design criteria at the First international Conference on Structural Mechanics in Reactor Technology held in Berlin [9]. Although no reference is made therein to the double earthquake

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approach, many other problems of importance in seismic design were delineated, ranging from the latest response spectrum concepts and analysis approaches through handling of equipment design, relative motion effects, etc. Both of these papers reflected major developments in seismic design criteria. Thereafter, additional work by Newmark, Hall and Mohraz as well as Blume, Sharpe and Dalai led to major developments in response spectra that were incorporated as a part of the U.S. nuclear power plant seismic design process (USNRC Reg. Guide 1.60) [10]. The latest major criteria document prepared by Newmark and Hall appeared in 1978 and was developed specifically for use in the Systematic Evaluation Program (SEP) of the U.S. Nuclear Regulatory Commission [11]. Though this document discusses the dual level earthquake design criteria briefly, in fact, a decision was made to go ahead with the limited review studies of existing plants with one earthquake, namely, the upper level earthquake employed in some cases was quite site specific in nature. One must be exceptionally careful in employing site specific spectra since generally they are based on a limited data base, are often narrow banded (in frequency content), and may not reflect the effects of strong and longer duration shaking that can arise from distant, as well as possibly closer, larger earthquakes. It may be well to note that in these studies only a few selected systems and components were examined in detail.

3. Current U.S. criteria - - Federal Regulation 10CFRI00 Appendix A reactor site criteria

On of the most important criteria documents to appear in the United States toward the later part of the time frame just discussed, namely November 1971, is Appendix A to Title 10, Part 1, Code of Federal Regulations, Part 100 [12]. This document sets forth the seismic criteria applicable to the siting of reactors and discusses faultin and the selection of the double earthquake hazard to be used as part of the design process. Development of this document entailed several years of preparation, comment and discussion by individuals in governmental agencies, the nuclear industry, researchers, seismologists and earthquake engineers. The document also sets forth some very fundamental and basic criteria which will become evident later herein. Certain of the key sections of this document are noted below: Section 100.10: Factors to be considered when evaluating sites - - "... It is expected that reactors will reflect through their design, construction and operation an

extremely low probability for accidents that could result in release of significant quantities of radioactive fission products. In addition, the site location and the engineered features included as safeguards against the hazardous consequences of an accident, should one occur, should ensure a low risk of public exposure..." Appendix A: S e i s m i c a n d Geologic Siting Criteria f o r N u c l e a r P o w e r Plants - - I. P u r p o s e - - "General Design Criterion 2 of Appendix A to Part 50 of this chapter requires that nuclear power plant structures, systems. and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami and seiches without loss of capability to perform their safety functions. It is the purpose of these criteria to set forth the principal seismic and geologic considerations which guide the commission in its evaluation of the suitability of proposed sites for nuclear power plants and the suitability of the plant design bases established in consideration of the seismic and geologic characteristics of the proposed site." "These criteria are based on the limited geophysical and geological information available to date concerning faults and earthquake occurrence and effect. They will be revised when necessary when more complete information becomes available.'" Section III: Definitions, notes: "The Safe Shutdown Earthquake" * is that earthquake which is based upon an evaluation of the maximum earthquake potential considering the regional and local geology and seismology and specific characteristics of local subsurface material. It is that earthquake which produces the maximum vibratory ground motion for which certain structures, systems, and components are designed to remain functional. These structures, systems and components are those necessary to assure the following: (1) The integrity of the reactor coolant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) The capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposure of this part."

* The "Safe Shutdown Earthquake" until this time in Safety Analysis Reports also had been identified as the "Maximum Hypothetical Earthquake" or the "Design Basis Earthquake."

J.D. Steoenson et al. / Nuclear power plant seismic design

The "Operating Basis Earthquake" ** is that earthquake which, considering the regional and local geology and seismology and specific characteristics of local sub-surface material, could reasonably be expected to affect the plant site during the operating life of the plant [1]. It is that earthquake which produces the vibratory motion for which those features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional. In Section V entitled "Seismic and Geologic Design Bases", there is a lengthy discussion of the determination of the design basis for vibratory ground motion that includes a set of suggested procedures to be followed in determination of the Safe Shutdown Earthquake (SSE). Near the end, Section V states that the maximum vibratory accelerations of the Safety Shutdown Earthquake at the foundations of the nuclear plant shall be taken at least equal to 0.1g. Section V (1) "Determination of Operating Basis Earthquake," states the following: (2) "The Operating Basis Earthquake shall be specified by the applicant after considering the seismology and geology of the region surrounding the site. If vibratory ground motion exceeding that of the Operating Basis Earthquake occurs, shutdown of the nuclear power plant will be required. Prior to resuming operations, the licensee will be required to demonstrate to the Commission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public (3). The maximum vibratory ground acceleration of the Operating Basis Earthquake shall be at least one-half the maximum vibratory ground acceleration of the Safe Shutdown Earthquake." It is apparent that the Operating Basis Earthquake (OBE) has a triple definition. It is described as an earthquake: (1) for which the features necessary for continued operation of the plant are designed to remain functional; (2) to have a design value at least one-half that of the SSE; and (3) to be specified by the applicant. As is well known, ther are several plants, which for one reason or another, the OBE is either greater than or less than one-half of the SSE, but for the most part the OBE has been taken to be half that of the SSE. The nuclear safety concerns associated with the OBE ** The "Operational Basis Earthquake" until this time in Safety Analysis Reports had also been identified as the "Maximum Credible Earthquake" or the "Design basis Earthquake' 'when the "Maximum Hypothetical Earthquake" designation was used for the larger earthquake.

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in general have not been clearly identified in the Federal Regulation. For instance, from definition (1) in the Federal Regulation, it is clear that features necessary for continued operation (in the event of an earthquake up to the selected OBE level) must be designed to remain functional. However, there is no nuclear safety requirement that the plant continue to operate at or beyond the OBE or any other earthquake level. Presumably if the plant were designed to be shut down for inspection and safety evaluation in the event of an earthquake beyond a selected level, there should be no safety requirement to design for continued operation above that level, although the design must be made to ensure containment and safe shutdown. From the standpoint of the owner/operator, the option of selecting an OBE seismic design level considered necessary to protect his financial investment as indicated in definition (3) of the Federal Regulation should be provided. The design requirement to safely shutdown the plant up to the SSE level would seem sufficient to assure nuclear safety. The additional requirement for seismic design at the 0.5 SSE level has been established to provide high confidence of plant reliability at an earthquake level which could reasonably be expected to affect the plant site during the operating life of the plant. However, this goal could be better achieved by tying the OBE level to a specific recurrence interval rather than to 0.5 SSE. For instance, a 10 to 20% probability of occurrence in a 40 year design life would be consistent with recurrence intervals of 380 to 180 years, respectively. Thus, specifying a 300 year recurrence interval for the OBE would be conservative for an earthquake which could reasonably be expected to affect the plant site during the operating life of the plant. For low seismic areas, a 300 year recurrence interval earthquake would be substantially less than 0.5 SSE. The foregoing salient sections extracted from some six pages of the Federal Regulation for the most part provide the basis upon which current nuclear power plant design practices in the U.S. are based. Some of the engineering aspects are discussed in a later section herein. A document that may be of some interest in connection with development of future siting guidelines is that pertaining to a recent study conducted under the auspices of the National Academy of Sciences entitled "Earthquake Research for the Safet Siting of Critical Facilities" [13]. The report presents some interesting insight into problems connected with the siting and design of nuclear power plant facilities as well as other critical facilities.

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4. General seismic design considerations In general, two procedures are available for defining the earthquake hazard. In the first procedure, where there is an extensive history of earthquake activity, and where geologic and tectonic investigations are feasible, estimates can be made of the possible magnitude and the location of future earthquakes that may affect a site. In some instances, such earthquakes will occur along well-defined fault structures. From this information, one can make estimates of the earthquake motion intensity propagated to the site, taking into account the experimental and observational or measured data available. Sites with well-defined active or capable fault structures within a 100 mile radius of the site (regions typically identified as having high seismic risk) typically have SSE and OBE earthquake levels which are at least twice those in regions where no such structures have been identified. A second procedure for developing insight into earthquake hazard in a region is often used when occurrence of earthquakes is not generally associated with surface faulting (most nuclear power plant sites) or when insufficient data are available from records and observations. Under these conditions, relationships have been developed for correlating ground motions, generally maximum velocities or maximum accelerations, to a qualitative measure of the intensity of motion, as for example, in the U.S. that measure estimated from the "Modified Mercalli intensity scale." Because of the inherent subjective human judgemental role these relationships are not as readily based on mathematical determination as the relations for earthquake motion propagation associated with the first approach. However, there have been sufficient observations to permit useful probabilistic data bases to be obtained. On this basis then, the seismic design parameters associated with the SSE are normally assumed to have a return period on the order of 1000 to 2000 years and, in some cases, as much as 10000 years at a particular site [14,15]. It should be understood that this is not the SSE return period for the tectonic province as a whole which typically is composed of several tends of thousands of square miles. SSE return periods for such large areas typically are much shorter. In truth, estimates of return periods of such long duration, given the reliable data base of only a few hundred years, have a substantial degree of uncertainty. These return periods are several times longer than those associated with standard building codes, and thus the seismic design levels would be expected to be significantly higher also. The OBE, on the other hand, is generally believed to

be associated with a return period of the order of 300 to 1000 years at a particular site in regions without welldefined earthquake activity and 300 years or less in regions of well-defined active or capable fault structures. Many regional seismic hazard probabilistic studies have been made in recent years and have been used as a basis for comparison of building code provisions. In most cases these studies in the U.S. are based on the assumption of a probability exceeding 10 percent for a structure of 50 year life which leads to a return period of about 475 years [16]. However, it must be pointed out that seismic design for conventional engineered structures typically employs nonlinear response assumptions which reduce seismic design loads by a factor which ranges from 3 to 8 depending on the type of construction [171. In determining the acceleration levels and other earthquake effects associated with the SSE and to some extent the OBE, it is rather obvious that the history of recorded experience with earthquakes of this size is not that long; thus, estimates of this type are educated guesses at best. The original premise behind the selection of the two earthquake design levels was that of assuming that the SSE would control the design in nearly all aspects and that the OBE would serve as a separate check for certain systems, where continued operation was desired at a lower seismic design level. As it turned out many, if not most, of the load combinations were similar. With the load factors, damping, stress levels, and service limits employed after 1968 when the one-third increase in allowable stress for structures under the OBE loading was no longer accepted by the AEC/NRC, the OBE, rather than the SSE, controlled the design for many systems [18] updated as shown in Appendix A to this report. This was not the original intent when the double earthquake design approach was developed. The SSE was intended to be the foremost controlling one of the two seismic loadings for most cases and sites with the OBE providing additional operability assurance for probable seismic events. 5. Effective acceleration To date, peak values of the zero period ground acceleration measured by strong motion accelerometer instruments have been used generally to establish seismic design requirements. While peak instrumental ground acceleration is a relatively easily determined and convenient quantity, it is a poor measure of the damaging potential of earthquake motions [19,20]. As employed for nuclear plant design and review analyses, the term

J.D. Stevenson et al. / Nuclear power plant seismic design effective acceleration is associated with the significant part of the ground motion containing repetitive motion portions that possess strong energy content. Obviously duration of shaking as well as amplitude and frequency (time) characteristics are among the important parameters to be considered. These portions of the ground motion are of primary importance in evaluating the response and behavior of the structure or equipment elements, and therefore, are of importance in design and in assessment of damage potential. In this sense then, in accordance with the definition given above, the effective acceleration normally is not that value connected with the high spikes of instrumentally recorded highfrequency accelerations commonly found to occur close to the source of seismic energy release. This is especially true for structures on foundations of some size or weight. On the other hand, the effective acceleration would be expected to be very close to the peak instrumental acceleration for locations at significant distances from the source, zones where such high-frequencyacceleration peaks normally are not encountered. For design purposes, it is believed that the effective acceleration values should be used in the basic process or arriving at the anchor point for the design response spectra and not the highest instrumental peak associated with a set of records. The concept of effective acceleration, as defined recently by Drs. Newmark and Hall, may be stated in following manner: It is that acceleration which is most closely related to structural response and to damage potential of an earthquake. It differs from and is less than the peak free-field ground acceleration. It is a function of the size of the loaded area, the frequency content of the excitation, which in turn depends on the closeness to the source of the earthquake, and to the weight, embedment, damping, and stiffness of the structure and its foundation. The practice over the years has been to use a relatively broad-banded response spectrum for design such as that depicted by Regulatory Guide 1.60 [10]. Although this approach is still fairly sound, in the years to come there may well be a series of response spectra used in design of critical facilities, perhaps one for near-field seismic sources within 20 km, one for moderate distance sources (20 to 60 km), and one for distant sources (greater than 60 km up to several hundred kilometers or more away). Each of thes spectra may be controlling in different frequency regions. An enveloping type spectra and associated time-histories could be employed if it appeared convenient from a design viewpoint to use the inherent conservatism of the enveloping of seismic re-

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sponse from dissimilar and for the most part independent sources. The primary role of the distant earthquakes, of course, is that of affecting long period or low frequency items typically less than 1.0 Hz which in general are not present in nuclear power plant safety-related components. From the practical design standpoint, if a response spectrum analysis is to be employed, the spectral values should be representative of the actual inertial loads. Failure of structures and components to perform required safety functions usually means inelastic response has occurred in excess of the available ductility and strength of the structure or component. The details of the interrelationship between the spectral characteristics and the structural properties are becoming more evident as a result of the research being carried out by several groups in the U.S. at the present time (for example, see ref. [21]). Any reliance on significant nonlinear behavior requires detailed knowledge of the characteristics of the element resistance, both local and global, as well as the loadings. It also requires careful consideration of the margin of safety in case of accidental overloading. Thus, it is recommended that broad band response spectra be employed in design and that this response spectra be anchored to an effective (design) acceleration value representative of the applicable strong shaking to which the structure as a whole might be subjected. It is further recommended that consideration be given to defining inelastic response spectra [11] for use in rational seismic design which employ ductility requirements that limit damage in real structures and components to acceptable limits.

6. Alternative seismic design philosophies Before reaching any conclusion regarding current seismic design requirements, particularly as they pertain to the double earthquake approach to design of a particular structure or component, it would be well to review current Canadian and Japanese procedures which, in general, do not require such a double earthquake design procedure. 6.1. Canadian seismic design Nuclear power plants licensable in Canada must provide assurance against the release of potentially hazardous quantities of radioactive materials and must assure the integrity of structures and components of the nuclear power plant in the event of an earthquake.

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J, D. Stevenson et al. / Nuclear power plant seismic design

6.2. Canadian design earthquake levels To provide assurance of critical systems performing their safety-related functions in the event of an earthquake, selected safety-related systems in the nuclear power plant are designed to specified earthquake levels as follows: (1) Design Basis Earthquake (DBE) The Design Basis Earthquake for a plant is defined as an artificial representation of the combined effects in the free-field at the site of a set of possible earthquakes having a sufficiently low probability of exceedence during the life of the plant, expressed in the form of response spectra or a time-history. (2) Site Design Earthquake (SDE) The Site Design Earthquake for a plant is defined as the maximum predicted effect in the free-field at the site, having an occurrence rate of 0.01 per year, based on historical records of actual earthquakes applicable to the site, expressed in the form of response spectra or a time-history. The SDE shall have a peak ground motion acceleration not less than 0.03g. The DBE and SDE for the plant are specified by at least a peak horizontal ground acceleration value, as well as the design ground response spectrum or timehistory. The peak ground acceleration values for the DBE and SDE are determined from a study of site seismicity based on an examination of historical and instrumented earthquake records for the area, as well as the seismotectonics of the surrounding geological structure. The applicable design earthquake is treated as an extreme environmental load having a tow probability of exceedence. Only one design earthquake as applicable is assumed to occur during the life of the plant * 6.3. Canadian seismic classification Seismic classification of a sufficient number of structures and components is required to ensure that the function described below can be fulfilled in the event of a DBE at the site: (1) The reactor must be capable of being shut down and of being maintained in that state indefinitely. (2) It must be possible to remove decay heat from the fuel during this shutdown period. (Note: The primary coolant system boundary shall not fail in * Either an DBE or SDE is used in design of a particular structure or components but not both.

such a manner as to constitute a loss-of-coolant accident.) (3) The structures and systems outside the containment area are designed so that any radioactivity releases are within the limits permitted by the sitting criteria of the Atomic Energy Control Board. The seismic classification in Canada include the design earthquake level and a seismic category. Seismic classification similar to those found in Section 3.2 of the standard U.S. Safety Analysis Report lists for each project the systems and structures requiring seismic qualification. 6.4. Canadian seismic category The extent to which each structure and system shall remain operational is established by means of seismic categories for individual structures and components of each system. The seismic category defines the following two requirements of the component: (a) The detailed functional requirement, if any, of the component to meet the safety function, (b) The requirement to perform during, after, or during-and-after an earthquake. There are two basic categories: Category A: Systems which must retain their pressure integrity during and following an earthquake to ensure and maintain the safety-related system operation. Category B: Systems which must retain their pressure integrity a n d / o r function mechanically a n d / o r electrically, as applicable, during a n d / o r following an earthquake, to ensure and maintain the safety-related system operation. 6. 5. Japanese seismic design [25] The information contained herein was developed from the Japanese Regulatory Guide [25] as translated into English by Mitsubishi Atomic Power Co. 6.6. Japanese design earthquake levels The basic design earthquake ground motions are classified into S 1 and S2 depending upon their intensities: (1) For basic design earthquake ground motions, S 1 (hereafter referred to as "the maximum design earthquake"), reference is made to the earthquake among the recorded earthquakes that would have the greatest effect on the proposed site and sur-

J.D. Steoenson et al. / Nuclear power plant seismic design

rounding region and which may occur again in the same fashion, or among those earthquakes that might be induced by highly active faults in the near future. (2) For basic design earthquake ground motions, S 2 (hereafter referred to as "the extreme design earthquake"), reference is made to the earthquake among those earthquakes exceeding the maximum design earthquake that would have the greatest effect on the proposed site based on engineering judgment following a seismological review of past earthquakes, the nature of any active faults and the seismotectonic structure underlying the site and the surrounding region. For the earthquakes generating the basic design earthquake ground motions, both distant and near field epicentral distances shall be considered. In addition, an earthquake occurring nearby shall be considered for the basic design earthquake ground motions S 2. In determining the basic design earthquake ground motions, full consideration shall be given to the following items: (1) The magnitude, epicenter, hypocenter, aftershock area and volume, maximum intensity of earthquake ground motion (or estimated value), and resultant damage (including destruction rate of structures, overturning of tombstones, etc.) in earthquakes that have affected the site and the surrounding region in the past. (2) The statistical expectation of the intensity of past destructive earthquake ground motions. (3) The magnitude of the earthquake and the distance between the site and its center of energy release. (4) Past observation records for the general region as well as those for the site, including any results of bedrock property investigations. Pursuant to the above-listed items, the basic design earthquake motions shall be such that each of the following parameters can be evaluated realistically: (1) The maximum amplitude of earthquake ground motion. (2) The frequency characteristics of earthquake ground motion. (3) The duration of earthquake ground motion and the time-dependent variation of the amplitude envelope curves. 6. 7, Japanese seismic classification

The nuclear power reactor facilities shall be classified into the following categories, according to the effect

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on the environments of radioactivity which might be emitted after earthquake-induced damage: Class A: Facilities containing radioactive material or related directly to equipment containing radioactive material and whose loss of function might lead to the release of radioactive material to the atmosphere, facilities required to prevent the occurrence of such accidents, and facilities required to mitigate the consequences resulting from the spread of radioactive material in the event of an accident and whose effectiveness in mitigating such consequence is significant. Class As: Facilities contained within the reactor coolant pressure boundary except small diameter components whose breakage would result in mass flow which would be controlled by the ordinary make up water system. Facilities which make up the reactor containment boundary including the containment vessel and associated piping, valves, and containment closures. Facilities directly associated with: (a) Spent fuel storage. (b) Introduction of negative reactivity into the core to effect an emergency shutdown and to maintain the core in shutdown state. (c) Required for removal of decay heat in the event of a failure of reactor coolant pressure boundary. Class B: Same as Class A above but having relatively minor effectiveness. Class C: Facilities not classified as A or B. The facilities have the same degree of safety as ordinary industrial facilities. A single earthquake ($1) is used in Japan for design of all items designated Class A which would be categorized as Seismic Category I in the U.S. Some of the less critical components of the reactor coolant and engineered safeguards support systems, the radioactive waste system and spent fuel storage cooling system which would be designated Seismic Category I in the U.S. May be classified as Class B, which in general would receive a 0.5S 1 dynamic analysis or a 1.5C0 static analysis where CO is defined as the seismic coefficient for conventional design in Japan. In addition, some of the more critical Class A items such as the major portion of the reactor coolant system, containment system, spent fuel storage racks and storage pool and emergency shutdown system carry a designation Class As. In addition to the S 1, these items would be designed for an S 2 level earthquake which currently is defined as having an intensity equal to

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J.D. Stetrenson et al. / Nuclear power plant seismic design

1.5S 1. It should also be noted that in general it appears that there is no requirement in Japan that structures and components designed to the $2 earthquake level must consider the earthquake input on an elastic basis which is currently required in the U.S.

6.8. Seismic input

Nuclear power reactor facilities, in accordance with their classification, shall satisfy the following basic principles: (1) Each Class A facility shall be capable of withstanding the larger of the seismic forces determined by the maximum design earthquakes or by the static seismic forces defined in this section (S 1 level earthquake). In addition, each Class As facility shall maintain its safety function against the seismic forces determined by the extreme design earthquakes defined in this section (S 2 level earthquake) ** (2) Each Class B facility shall be capable of withstanding the static seismic forces defined in this section. An evaluation by dynamic analysis shall be made for those Class B facilities which may resonant (0.5S1 level earthquake for dynamic analysis). (3) Each Class C facility shall be capable of withstanding the static seismic forces determined using the static coefficient defined in this section. (4) In applying the above principles (1), (2) and (3), attention shall be given to the prevention of propagative failures among high ranked facilities following the rupture of low-ranked facilities. The design earthquake ground motions for aseismic design of nuclear power reactor facilities shall be derived from the earthquake motions at the free surface of the base stratum at the proposed site.

7. Conclusions

The seismic design requirements for nuclear power plants in the U.S. have remained essentially the same since 1975. These requirements were established with the publication of the Federal Regulation 10CFR100

** Components classified as As in Japan are designed to resist both the S1 and S: earthquake design levels. All other seismic classifications, A, B and C are designed for a single earthquake level.

Appendix A in 1971 and the pertinent implementing regulatory guides: 1.29 Seismic Design Classification-6/72, 1.60 Design Response Spectra for Seismic Design of Nuclear Power Plants 10/73, 1.61 Damping Values for Seismic Design of Nuclear Power Plants-10/73, 1.92 Combining Modal Responses and Spatial Components in Seismic Response Analysis-12/74. and the basic Standard Review Plan Sections 2.5, 3.2 and 3.7 in June 1975. There has been some effort in the U.S. to update nuclear plant seismic design requirements [26], but there has been little significant change in NRC criteria and methodology since 1975. Serious consideration should be given to updating the seismic design requirements, and the changes might well be based on considerations of the following type. We believe consideration should be given to decoupiing the OBE from the SSE. The OBE is intended to characterize an earthquake which could reasonably be expected to affect the plant site during the operating life of the plant. This goal can be reasonably and conservatively achieved by defining the OBE as the best-estimate 300 year recurrence interval earthquake unless the applicant desires to establish a longer recurrence interval ***. For low seismic regions, such an OBE would be substantially less than the 0.5 SSE. It is expected that for low seismic regions, this definition would result in OBE levels of about 0.25 to 0.3 times the SSE. If, based upon the above definition, the OBE was less than about 0.05g to 0.07g effective ground acceleration, the requirements for an OBE design could be eliminated on the basis that such a low earthquake level is not capable of damaging engineered heavy industrial facilities and power plants. However, in view of the nature of the critical facility, some recoverable margin of safety in resistance must be included in the design. Thus, keeping in mind that the safety systems are designed for a much larger SSE, the cost and effort associated with evaluating the plant for such a low OBE could well be considered to be an unwarranted engineering effort. In general, the no-damage provision coupled with the 300 year recurrence interval would eliminate the *** In Canada, Sweden and the United Kingdom, the equivalent of the OBE level earthquake is associated with a 100 year recurrence interval. In the Fed. Rep. Germany, the recurrence level associated with the currently defined OBE level earthquake is estimated at 1000 years.

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need for an OBE design in low seismic regions (that is, in general, regions where the SSE is equal to or less than about 0.2g). In addition, a 300 year recurrence interval OBE represents a rare event for which it is unnecessary to hold equipment and structures at normal stress levels, although significant overstressing should be avoided. More realistic elastic stress limits (near yield) would still permit predictable and reliable elastic seismic response, but not over-restrain the system so as to potentially cause failure during normal operation. In order to reduce the current tendency to over-restrain systems, very serious consideration should be given to permitting controlled and limited nonlinear and ductile response effects associated with the SSE. The Pressure Vessel Research Committee has recently formed a Steering Subcommittee on Piping Systems in response to a growing recognition that overrestraint of high temperature, high energy nuclear power plant piping caused by overly conservative seismic design requirements actually may be reducing nuclear power plant safety [27]. It is hoped this subcommittee may serve as a focal point in developing revised seismic design requirements as part of the ASME Code and N R C criteria development which would result in more rational and balanced designs. Thus, it is recommended that broad band response spectra be employed in design, and that they be anchored to an effective (design) acceleration value representative of the applicable strong shaking to which the structure as a whole might be subjected. It is further recommended that consideration be given in defining inelastic response spectra [11] for use in rational seismic design which employ ductility requirements which limit damage in real structures and components to acceptable limits.

ture and equipment reaction load as: 1.0D+I.OL+I.0E'.

In addition, the determination of the magnitude of the E load in the limit relative to the E ' load is dependent on the assumed damping factors. As an example, the maximum amplification factor for the 7% damping and the R.G. 1.60 ground response spectrum associated with the SSE for reinforced concrete is 2.59. The maximum amplification factor for 4% damping associated with the OBE is 3.20. This would result in the following relationship at maximum response, ignoring dead and live loads, of SSE to OBE before the SSE would govern design as: OBE =

A primary example of OBE dominance in design is found in safety class concrete structures other than containment. The current NRC Standard Review Plan 3.8.4, Other Category I Structures, specifies the loading combination for the design of concrete structures, including the OBE load, as: 1.4 D + l . 7 L + l . 9 E ,

where D = dead load, L = live load, E = OBE load, and E ' = SSE load. The same document identifies the loading condition which includes the SSE load in the absence of tempera-

1.0 = 0.426 SSE. 2.35

A second example of the OBE dominance occurs in the design of nuclear plant piping. In this case, the influence of the OBE relative to the SSE on design is even more pronounced. For ASME Class 1 Piping: OBE (Level A Service Limits) from eq. (9) [NB3652], San = 1.5 S m. SSE (Level D Service Limits) from eq. (9) [NB3656], San = 3.0 Sm.

Determine allowable stress Ss available to carry seismic load: OBE -- 0.9/2.4 = 0.375 SSE. The effect of damping amplification factors in accordance with NRC Regulatory Guides 1.61 and 1.60 is: 0.375 x

Appendix. Examples that illustrate why the OBE rather than the SSE governs design

1.0 1.9(3.20/2.59)

1

3.13 [ SSE (building 5%) 4.25 [ OBE (building 2%) ]

4.25[ SSE (piping 2%) ] × 5 - ~ OBE (piping 1%) = 0.23 SSE. For ASME Class 2 and 3 piping: OBE (occasional loads) from eq. (9) [NC3652.2] Sal 1 = 1 . 2 S h ,

SSE (Level D Service Limits) from eq. (9) [NC3655] Sai I = 3.0Sh,

OBE = 0.6/2.4 = 0.250 SSE, considering the effect of damping 3.13 4.25 OBE = 0.250 × ~ ×~ = 0.15 SSE. As a result of this comparison considering allowable

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J.D. Stevenson et aL / Nuclear power plant seismic design

stresses and damping effects, it can be seen that the OBE must be approximately one-quarter of the SSE for Class 1 piping and one-sixth of the SSE for Class 2 and 3 piping before the SSE will begin to control design of these elements using current U.S. design criteria.

Acknowledgment The authors acknowledge with gratitude the constructive review of the paper provided by Mr. S.L. McCabe, Research Assistant in Civil Engineering, University of Illinois.

References [1] G.W. Housner, Design of Nuclear Power Reactors Against Earthquakes, in: Proceedings Second World Conf. on Earthquake Engineering, IAEE, Tokyo, Japan, Vol. 1 (1960) pp. 133-149. [2] Nuclear Reactors and Earthquakes, Report TID-7024, U.S. Atomic Energy Commission, Washington, DC (1963). [3] G.W. Housner and D.E. Hudson, Earthquake research problems of nuclear power plants, Nucl. Engrg. Des. 3 (1966) 308-319. [4] N.M. Newmark, Selection of earthquake hazard for earthquake resistant design, Informal memorandum dated 3 February 1967, 22p. [5] N.M. Newmark, Design criteria for nuclear reactors subjected to earthquake hazards, in: Proceedings of IAEA Panel on Aseismic Design and Testing of Nuclear Facilities, held 12-16 June 1967, Tokyo, Japan, pp. 90-119 (incl. Appendix co-authored with W.J. Hall). [6] G.W. Housner, Seismic problems of nuclear power plants, in: Proceedings of the IAEA Panel on Aseismic Design and Testing of Nuclear Facilities, held 12-16 June 1967, Tokyo, Japan, pp. 53-56. [7] Aseismic design and testing of nuclear facilities, IAEA Technical Report Series No. 88, Vienna (1968). [8] N.M. Newmark and W.J. Hall, Seismic design criteria for nuclear reactor facilities, in: Proceedings Fourth World Conf. on Earthquake Engineering, IAEA, Santiago, Chile, Vol. II, Sect. B-4 (1969) pp. 37-50. [9] N.M. Newmark, Earthquake response analysis of reactor structures, Nucl. Engrg. Des. 20 (1972) 303-322. [10] Design Response Spectra for Seismic Design of Nuclear Power Plants, U.S. NRC Regulatory Guide 1.60, Revision 1 (December 1973). [11] N.M. Newmark and W.J. Hall, Development of criteria for seismic review of selected nuclear power plants, U.S. Nuclear Regulatory Commission Report N U R E G / C R 0098 (May 1978). [12] Reactor Site Criteria, Rules and Regulations, U.S, Nuclear Regulatory Commission, Appendix A, Part 100, Title 10,

Chapter 1, Code of Federal Regulations (April 30, 1975) pp. 100-1 to 100-6. [13] W.J. Hall et al, Earthquake research for safet siting of critical facilities, National Research Council/National Academy of Sciences, Washington, DC (1980) 49p. [14] Nuclear Structures and Materials Committee: Structural analysis and design of nuclear plant facilities, Manuals and Reports on Engineering Practice-No. 58, American Society of Civil Engineers (August 1980). [15] Natural and man-made hazards at power reactor sites--Guidelines for combining, ANSI/ANS-2.12, American National Standards Institute (1978). [16] N.C. Donovan, B.A. Bolt and R.V. Whitman, Development of expectancy MAPS and risk analysis, J. Struct. Div. ASCE 104 (St 8) (August 1978) 1170-1192. [17] Applied Technology Council: Tentative provisions for the development of seismic regulation for buildings, NBS Special Publication 510, NSF Publication 78-8 and ATC Publication ATC-3-06 (June 1978). [18] J.D. Stevenson, Rational determination of the operational basis earthquake and its impact on overall safety and cost of nuclear facilities, Nucl. Engrg. Des. 35 (1975) 327-333. [19] R.P. Kennedy, Peak acceleration as a measure of damage, Fourth International Seminar on Extreme-Load Design of Nuclear Power Facilities, Paris, France, August 1981. [20] W.J. Hall, Observations on some current issues pertaining to nuclear power plant seismic design, Nucl. Engrg. Des. 69 (1982) 365-378; see also: EERI Monograph entitled: Earthquake Spectra and Design, by Newmark and Hall (1982). [21] R.P. Kennedy, S.A. Short and N.M. Newmark, The response of a nuclear power plant to near-field moderate magnitude earthquake, SMiRT-6, Paris, France, 17-21 August 1981. [22] C.G. Duff and M. Singh, Design requirements, criteria and methods for seismic qualification of CANDU nuclear power plants, AECL-6691, Atomic Energy of Canada Ltd. (October 1979). [23] ASME Boiler and Pressure Vessel Code, Section Ill, 1980 Edition (Div. 1 and Div. 2, plus Appendices). [24] ASME Boiler and Pressure Vessel Code, Nuclear Code Cases. [25] Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, Japan Atomic Commission (September, 1978). [26] D.W. Coats, Recommended revisions to nuclear regulatory commission seismic design criteria, U.S. Nuclear Regulatory Commission Report NUREG/CR-1161 (May 1980). [27] Letter from S.J. Bush, Battelle Pacific Northwest Laboratories to Chairman Palladino, U.S. Nuclear Regulatory Commission, dated August 20, 1981. [28] L.J. O'Brien, J.R. Murphy and J.A. Lahoud, The correlation of peak ground acceleration amplitude with seismic intensity and other physical parameters, U.S. Nuclear Regulatory Commission Report NUREG-0143 (March 1977).