ML20076G181
| ML20076G181 | |
| Person / Time | |
|---|---|
| Issue date: | 08/30/1983 |
| From: | NRC |
| To: | |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR NUREG-1018, NUREG-1018-DRFT, NUDOCS 8308300780 | |
| Download: ML20076G181 (100) | |
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\\GR NUREG - IOld INTERIM REPORTL SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLANTS (USI A-46) 8308300780 830830 PDR NUREG PDR 1018 R
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a ABSTRACT The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform their intended safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years.
Therefore, the seismic qualificatiun of equipment in operating plants should be reassessed to determine whether requalification is necessary.
The objective of technical studies being performed under the Task Action Plan for Unresolved Safety Issue (USI) A-46 is to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of the seismic qualification of equipment at all operating plants, in lieu of requiring current qualification criteria which are applied to new plants.
This report summarizes the status of work accomplished on USI A-46 by the Nuclear Regulatory Commission staff and its contractors, Idaho National Engineering Laboratory (INEL), Southwest Research Institute (SWRI),
Brookhaven National Laboratory (BNL) and Lawrence Livermore National Laboratory (LLNL) that is applicable to USI A-46.
This assessment leads to the conclusion that the use of seismic experience data for equipment qualification provides the only reasonable alternative to current qualification criteria.
Consideration of seismic qualification by use of experience data was a specific task in USI A-46.
Several other A-46 tasks serve to support the use of an experience data base.
The status of continuing efforts to establish requirements for an experience data base is provided in this report.
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j TABLE OF CONTENTS Page ABSTRACT 11 1.
INTRODUCTION 1
1.1 Background
1 1.2 Staff Plan for Resolution of the Issue 2
2.
SUMMARY
OF TECHNICAL WORK ACCOMPLISHED 5
2.1 Identification of Seismic Risk Sensitive Systems and Equipment 5
2.1.1 Background 5
2.1.2 Summary of Task 5
2.1.3 Staff Position on Task 5
2.2 Assessment of Adequacy of Existing Seismic Qualification 6
2.2.1 Background 6
2.2.2 Status of Work as of June 1983 6
2.2.3 Staff Conclusions 12 2.3 Development and Assessment of In-Situ Testing to Assist in Qualification of Equipment 12 2.3.1 Background 12 2.3.2 Status of Work as of June 1983 13 2.3.2.1 Summary of Contractor Report "The Use of In-Situ Procedures for Seismic Equipment Qualification in Currently Operating Plants 13 2.3.2.2 Summary of Contractor Report " Preliminary Guidance and Acceptance Criteria for Application of Combined In-Situ and Analysis Procedures in Operating Plant Equipment Qualification" 20 2.3.2.3 Summary of Contractor Report " Summary of Work Performed to Date on Qualification Cost Estimate Task" 25 2.3.3 Staff Conclusions 26 iii i
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2.4 SEISMIC QUALIFICATION OF EQUIPMENT USING SEISf1IC EXPERIENCE DATA BASE 26 2.4.1 Background 26 2.4.2 Summary of LLNL Report " Correlation of Seismic Experience Data in Non-Nuclear Facilities with Seismic Equipment Qualification in Nuclear Plants" 30 2.4.3 Summary of EQE Report " Pilot Program Report-Program for the Development of an Alternative Approach to Seismic Equipment Qualification" 50 2.4.3.1 Methods Used in the Pilot Program 50 2.4.3.2 Conclusion and NRC Staff Comments 57 2.5 DEVELOPMENT OF METHODS TO GENERATE GENERIC FLOOR RESPONSE SPECTRA 65 2.5.1 Background 65 2.5.2 Summary of Work Completed 66 2.5.3 Staff Conclusion 69 3.
REFERENCES 72 4
APPENDICES A - Task Action Plan A-46 A-1 B - Related Topics Covered by the INEL Contractor's Report on In-Situ Testing B-1 iv j
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1.
INTRODUCTION
1.1 Background
General Design Criterion (GDC) 2 of 10 CFR Part 50, Appendix A for Nuclear Power Plants states that structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes, without a loss of capability to perform their safety functions. Appendix B,Section III of 10 CFR Part 50 also states that design control measures shall provide for verifying or checking the adequacy of design by the performance of a suitable testing program.
Furthermore, it requires that suitable qualification testing under the most adverse design conditions shall be included. These requirements point to the need for seismic qualification of safety related electrical and mechanical equipment in order to ensure structural integrity and functional capability during and after a seismic event.' Current criteria and methods of compliance are contained in Revision 2 to Standard Review Plan Section 3.10, " Seismic and Dynamic Qualification of Mechanical and Electrical Equipment," and Regulatory Guide 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants," which, with some exceptions, basically endorses IEEE Standard i
344-1975, "IEEE Recommended Practices for Seismic Qualification of' Class lE Equipment for Nuclear Power Generating Stations."
Based on the requirements and recommendations from these current criteria and methods, equipment is seismically qualified today by analysis and/or laboratory test. Analyses alone are acceptable only if the necessary functional capability of the equipment is assured by its structural integrity. Otherwise, some testing is required.
Seismic input motion to equipment is specified by required response spectra or by time histories.
When the test method is utilized, the equipment is mounted on a shake table and subjected to certain types of excitation corresponding to a test response spectrum which envelopes the required response spectra.
The equipment should be tested in the operating condition.
For equipment too large to fit on a shake table, a combined analysis and test procedure is adopted.
Since commercial nuclear power plants were first introduced, significant changes in seismic qualification criteria have occurred. The analytical and experimental methods used to qualify equipment have also changed. The margins of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform their intended safety functions Emay vary considerably,'and may not meet current seismic qualification criteria. Therefore, there is a recognized need to reassess the seismic qualification of equipment in operating plants to ensure its performance during and after a seismic event.
It was also recognized that it may-not be practical to qualify operating plant: equipment using current seismic qualification criteria and methods due
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v-to excessive plant down time, difficulties in shipping irradiated equipment tn a test laboratory and in acquiring identical old vintage equipment for laboratory testing.
In December 1980, the Nuclear Regulatory Commission designated " Seismic Qualification of Equipment in Operating Plants" as Unresolved Safety Issue (USI) A-46.
The objective of USI A-46 is to develop alternative seismic qualification methods and acceptance criteria that can be used to assess the capability of mechanical and electrical equipment in operating nuclear power plants to perform the intended safety functions.
1.2 Staff Plan for Resolution of the Issue A Task Action Plan (TAP) was developed for USI A-46 in the Spring of 1981.
The TAP is included as Appendix A.
Tasks selected for study were selected on the basis of their potential for providing reasonable alternatives to current requirements for seismic qualification.
It was recognized that a utility always has the option to re qualify equipment using procedures required for new plants. Only alternative procedures which provide :ame advantage over current requirements are likely to be used.
In addition, any alternative procedure must be sufficiently rigorous to provide a level of safety comparable to that achieved by current requirements.
A key element of the approach was to take advantage of experience gained by previous qualification tests and analysis and experience with actual seismic events.
Tasks selected for study were:
(1) Identification of seismic-sensitive systems and equipment; (2) Assessment of adequacy of existing seismic qualification; (3) Development and assessment of in-situ testing methods to assist in qualification of equipment; (4) Seismic qualification of equipment using seismic experience data; (5) Development of methods to generate generic floor response spectra.
As work progressed it became increasingly apparent that Task 4, " Seismic Qualification of Equipment Using Seismic Experience Data" was the most likely approach to develop a qualification method which is both economically attractive to the plant owners and acceptable from a public safety viewpoint.
Lawrence Livermore National Laboratory (LLNL), under contract to NRC, conducted a feasibility study (Ref. 1) which concluded that use of seismic experience data is feasible and can be as effective as current qualification methods. This study is discussed in more detail later in this report.
In addition, a utilities group, Seismic Qualification Utilities Group (SQUG) conducted a pilot program to independently demonstrate the feasibility of using seismic experience data.
Their report (Ref. 2) published in September 1982, was reviewed by the staff, and preliminary requirements for an acceptable data base were transmitted to the SQUG.
The staff is continuing to work closely with SQUG and their consultants to develop an acceptable
- procedure for using an experience data base. A more detailed discussion of this effort is presented in Section 2.4.1 of this report.
With the shift in emphasis of the program, Tasks 3 and 5, " Development and Assessment of In-situ Testing Methods to Assist in Qualification of Equipment" and " Development of Methods to Generate Generic Floor Response Spectra," play a strong supporting role. The emphasis on both tasks was focused to support use of an experience data base.
This is shown'in Figure
- 1. 2.1.
Task 2 " Assessment of the Adequacy of Existing Seismic Qualification," was an effort to develop methods to evaluate the acceptability of qualification by procedures used before current requirements were instituted.
For instance, a method was developed to assess results of a single axis test in terms of expected multiple axis response. A procedure was developed by Southwest Research Institute, but is of limited practical value in its present form because of the need to either know the fragility level or estimate the fragility of the equipment and know the required response spectra.
It may be useful in special cases.
Task 1, " Identification of Seismic Risk Sensitive Systems and Equipment,"
was an attempt to develop, on a generic basis, a minimum equipment list.
The study, performed by Brookhaven National Laboratory (BNL), was conducted on a hybrid model of a PWR plant and a hybrid model of a BWR plant using a seismic probabilistic risk assessment (PRA) model.
The contribution to risk of major systems and components was calculated and ordered by risk importance.
Although this study did provide some insight into the risk importance of systems and compcnents and demonstrated the effect of varying equipment fragility on overall risk, it is of limited usefulness in defining an equipment list at other plants.
The major conclusion of BNL was that they had demonstrated a methodology that could be applied on a plant specific basis to develop a risk based minimum equipment list.
For plants where an existing seismic PRA model is available, it may be feasible to evaluate the necessity to qualify specific systems or components oa the basis of risk contribution.
This task is described in more detail in Section 2.1.
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SUMMARY
OF TECHNICAL WORK ACCOMPLISHED In the remaining sections of this report, each of the tasks are described.
A summary of work done and major conclusions are presented.
Detailed discussions of certain tasks are then included as separate appendices.
The following sections summarize contractors results and conclusions of the various tasks.
Unless otherwise stated, they represent the contractors' viewpoint and recommendations.
The final staff positions on A-46 will be developed following completion of Task 4.
2.1 Identification of Seismic Risk Sensitive Systems and Equipment 2.1.1
Background
The objective of this task was to investigate possible methods of developing a generic minimum equipment list.
If a methodology could be developed to evaluate the risk importance of safety systems and equipment then equipment could be ordered by the contribution to risk.
Equipment whose failure resulted in a small change in risk could then be culled from the qualification list.
2.1.2 Sumnary of Task Brookhavg,gonalLaboratory(BNL)undercontracttotheNRCconducted a study to evaluate the seismic risk sensitivity of system and components in a PWR and a BWR.
Both plant models used were hybrids in that they are not representative of any existing plant.
The PWR model consisted of modified Surry Plant fault trees and event trees from the WASH-1400 study and used fragility data developed for the Zion plant.
The BWR model consisted of modified WASH-1400 Peach Bottom risk models and Cyster Creek' fragility data.
The intent of this study was initially to develop a generic risk ordered list of plant equipment which could be applied to specific plants with some additional guidelines to develop plant specific minimum equipment lists. However, BNL concluded and the staff agrees, that results of the study should not be used generically.
BNL's conclusion states that the study presents a methodology that can be applied on a plant specific basis to develop a risk ordered equipment list.
2.1.3 Staff Position on Task For plants with existing seismic PRA studies, the staff believes it may be possible in some cases to eliminate components from the seismic qualification program on the basis of low risk sensitivity.
If a utility should decide to conduct a PRA study using the methodology l
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developed by BNL, the staff would consider it to be an acceptable method subject to the analysis assumptions and inherent uncertainties.
The staff was unsuccessful in developing a generic minimum equipment list.
2.2 Assessment of Adecuacy of Existing Seismic Qualification 2.2.1
Background
This task involves a study by Southwest Research Institute (SWRI) to evaluate past and present methods to qualify mechanical and electrical equipment to withstand seismic events.
Conclusions have been documentcd in a contractor report titled " Correlation of Methodologies for Seismic Qualification Tests of Nuclear Power Equipment" (Ref. 4).
Some examples demonstrating the application of this approach are included in that report.
2.2.2 Status of Work as of June 1983 The concept of vibrational equivalence is a key factor in development of the correlation of methodologies for seismic qualification of equipment.
Vibrational equivalence forms the basis for a damage comparison between two different motions.
In the qualification of nuclear power plant equipment, a great variety of physical failure mechanisms may occur.
Therefore, the concept of vibration equivalence was generalized to include an arbitrary type of failure or malfunction, that can always be established by input vibrational conditions denoted as the fragility levels.
It is understood that the failure or malfunction may or may not impart permanent damage to the equipment.
The conceptual approach for applying vibrational equivalence to correlation of equipment qualification by test is shown in Figure 2.2-1.
The upper and lower halves of the diagram (conditions 1 and 2, respectively) each represent the independent establishment of a fragility, or threshold of failure level, in an equipment which is subject to a dynamic excitation at location x.
The effect of the response at location y is to actuate a failure mechanism which exists at that point in the equipment.
This arbitrary failure mechanism is dependent on the response amplitude at location y, and may also be dependent on time. Thus, the failure is indirectly dependent on the excitation amplitude, frequency, and time.
If the excitation is manipulated so that failure barely occurs, then the threshold of failure, or fragility function F f,t) is generated.
This function represents a surface, any point oh w(hich corresponds to failure of the equipment.
If more than one physical failure mechanism at more than one response point is present, then each possesses a failure surface, and the minimum value composite failure surface becomes of concern.
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central assumption of the vibration equivalence concept is then postulated:
the establishment of failure conditions (see Figure 2.2-1 for excitation conditions 1 and 2) is possible by various types of vibration excitations, and the corresponding amplitude, frequencies, and time durations constitute equivalent excitations.
Generally, the information on failure, or malfunction, is not required as part of an equipment qualification process.
On the other hand, functionality of an equipment at specified excitation levels is required for qualification.
Functionality and fragility are very much related--fragility is the upper limit of functionality.
Conversely, existing qualification data, which include excitation levels and functionality data, may be useful as a lower bound for fragility.
- Thus, since fragility data are necessary for a general application of the vibrational equivalence concept, use of such existing qualification data, where possible, is highly desirable to avoid the necessity of generating or collecting more precise fragility information for the great variety of equipment typically contained in a nuclear power plant.
The most general description of a fragility concept is shown in Figure 2.2-2 as a fragility surface. Inis surface can be represented as a function F f t
=M fragilityIdr(ac,e,)canbe(f,t),whereMintermsofth$a(mp,litudeoftheexcitation, f t), measured at the f
the response spectrum, power spectrum, or a variety of other parameters which may be used, or have been used in typical equipment qualification procedures. The true surface may be quite complex, but a simpler lower bound surface can be defined conservatively from existing qualification information which is acceptable for practical engineering purposes.
A convenient method of measuring the onset of failure is proposed by the contractor as the damage fragility ratio D
= M (f, t) 5,1 fr Mf (f,t) where M(f,t) is the value of the actual excitation function and Mf (f,t) is the value of the fragility function at the same conditions of frequency and time.
This is shown in Figure 2.2-3.
A damage fragility equivalence similar to that described in Figure 2.2-1 can then be stated as:
M(f,t) = M(f,,t,)
M (f,t) M (7,t )
f f 2 p This is the general basis for comparing various test motions.
The report then proceeded to define simple systems and complex systems.
A simple system is one whose fragility function is influenced by a single resonance, and therefore can be generated by a slowly swept sine or narrow band random excitation.
A complex system is one where _
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Due to the difficulties involved when considering complex systems, it is advantageous to develop approximations as required to reduce the system to a simple one.
A number of procedures have been developed in structural analysis to l
look at the combined effects of multiaxis and multimode response.
These procedures, such as Absolute Sum method, Square Root of the Sum of the Squares (SRSS) method, Double Sum method, Closely Spaced Modes method, Grouping method, Ten Percent method, Lin's method, and complete Quadratic Combination (CQC) method, are all generally based on modal or response spectrum analysis.
Any one of these methods will give an estimation of the combined maximum peak response of a complex system.
In developing a fragility surface for existing qualification data, it was recommended by the contractor that a correction factor, generated from resonance search data, be used to modify the level of qualification excitation in order to develop an approximate lower bound fragility function.
The next step is to establish a correlation between the approximate fragility function (namely existing qualification information) and the qualification corresponding to a different set of criteria, e.g.,
current criteria.
The possible pairs of comparison are listed in Figure 2.2-4.
In a specific application, some judgement must be used, the detail of which may vary with each case. Several examples which demonstrate the application of these methodologies are included in the contractor's report.
In summary, the results of a previous qualifications are used first to establish some form of an approximate or acceptable fragility function.
Then, the new criteria are compared to this acceptable fragility function to determine whether a greater or less severe test is implied.
If result shows a less severe test is implied by applying the new criteria, then it can be concluded that this equipment is still qualified to the new set of criteria.
In some cases, a more accurate fragility function may need to be established in order to provide a final determination of the comparison.
In these cases, the contractor suggested that it may be more practical to consider a complete new requalification.
It was also surmised by the contractor that much of the previously qualified equipment will be able to be requalified to new criteria by the analytical method developed. His belief is bas; d on the fact that many qualification tests prior to 1975 included sine wave and sine beat excitations of some form.
The comparison of relative damage severity indicated that such motions produce significantly more potential damage than do typical random motion simulations that have been more generally used after 1975. i l
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o 2.2.3 Staff Conclusions The technical basis and general methodology to correlate seismic qualification tests have been developed and demonstrated.
From what is shown in the contractor's report, this is a promising methodology.
However, in order for nuclear plant owners to be able to apply this methodology, more specific guidelines and acceptance criteria need to be generated.
Simpler guidelines and criteria are to be developed by SWRI by August 1983.
2.3 Development and Assessment of In-Situ Testing Methods to Assist In Qualification of Equipment 2.3.1 Background This task was selected for A-46, due to the potential that in-situ testing can be a promising tool in assisting the seismic qualification of equipment in operating plants.
The task is conducted by Idaho National Engineering Laboratory (INEL), and was started in early 1982.
The intent of this task is to investigate present in-situ testing methods and to evaluate the feasibility of using these methods to assist in re qualifying equipment, and to develop methods, guidelines and acceptance criteria for their use.
More specifically, the work scope for this task consisted of the following topics:
(1) Basic review of existing approaches to in-situ testing and identification of preliminary in-situ test methods for the qualification of equipment in plants which are currently licensed and operating.
(2) Review of approaches to laboratory testing and simulation of seismic events in the laboratory for qualification of equipment.
Limitations on the use of current guidance was also studied.
(3) Review of the analysis procedures fundamental to in-situ testing methods.
Review of use of subcomponent proof test and/or subcomponent fragility tests in the qualification process.
Review of the qualification requirements for anchors.
(4) Investigate techniques for assessing / monitoring the effects of chemical or metallurgic aging, mechanical fatigue, and wear during plant operation.
(5) Address adequacy, limitations and inherent shortcomings, and nonconserystisms of the various approaches above.
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(6) Development of guidelines and acceptance criteria for use of in-situ testing to support alternative methods of seismic quslification of safety related equipment.
(7) Define requirements for a test data base in support of seismic qualification of existing equipment in currently licensea operating plants.
(8) Develop cost estimate for alternate siesmic qualification methods.
(9) Verification and further development of combined in-situ and analysis methods suitable for equipment qualification.
Examine limitations and pitfalls of applying in-situ testing methods in determining dynamic characteristics and evaluating component mountings of structures which support, contain, or position safety-related equipment in operating plants.
Develop guidelines for minimum testing requirements and reporting requirements in qualification documentation.
2.3.2 Status of Work as of June 1983 Results of work on topics 1, 2, 3, 4, and 5 of Section 2.3.1 above are covered in an interim contractor report titled "The Use of In-Situ Procedures for Seismic Equipment Qualification in Currently Operating Plants (Ref. 5)" issued in December 1982. Another contractor progress report titled, " Preliminary Guidance and Acceptance Criteria for Application of Combined In-Situ and Analysis Procedures in Operating Plant Equipment Qualification" (Ref. 6) and issued in April 1983, provides the preliminary results on topics 6 and 9.
Finally, a separate draft contractor report titled, " Summary of Work Performed to Date on Qualification Cost Estimate Task" (Ref. 7) covers topic 8 and was issued also in April 1983. All three reports have been sent to the staff for review and comments.
Following is a summary of these tasks.
2.3.2.1 Summary of Contractor Report "The Use of In-Situ Procedures for Seismic Equipment Qualification in Currently Operating Plants" The goal of this study was to er. amine the most important uses of in-situ testing employed to assist in requalification of safety related equipment.
Theoretically, in-situ test procedures could be applied in the following three manners:
(1) Testing at full load level with equipment in place.
(2) Low load level testing with equipment in place.
(3) Periodic intermediate or low load level testing to support a continuing surveillance data base..
It is the conclusion of this study that among the three potential methods of in-situ test, only method 2 is normally practical and feasible. Method 1, which applies the dynamic load up to the Safe Shutdown Earthquake (SSE) level, has to satisfy certain conditions.
The required conditions are that:
(1) The motion applied to the equipment supporting structure should not excessively load the appurtenances, the components mounted thereon or in the vicinity and the equipment supporting structure itself.
(2) Sufficient access must exist in order to load the equipment mounting.
(3) No camage occurs to the local area where load is applied.
(4) No significant mechanical aging degradation has occurred during
. testing, such that component can be employed in service for its J
nominal useful lifetime.
These conditions severely limit the usefulness of full load level in-situ tests.
Valve operators are one equipment type that have been dynamically qualified in-situ by using a static load to perform an interference evaluation.
However, the potential for performing full load level in-situ testing is so limited that it is not considered further.
Method 3 above could, in principle, be useful for identifying aging degradation. However, the contractor concluded that for the types of equipment of interest in this program, no potential applications are apparent.
This is because changes significant to operability of safety related equipment (particularly in a seismic environment) cannot generally be detected by in-situ procedures.
The low load level in-situ tests are normally performed by applying hammer impact on equipment or supporting structures.
Portable electromagnetic or hydraulic shakers can also be applied to equipment or equipment supporting structures in place, in order to dynamically test them. The input force and output, normally acceleration, are recorded as loads are applied at various positions.
The recorded quantities are converted from time histories to a frequency representation by use of the Fourier transform.
Using the. frequency representation, transfer functions are calculated between points of input and output. These calculations are typically performed with minicomputers which are part of the modal analyzer system.
Software internal to these computers then identifies natural frequencies and mode l
shapes. The mode shapes encompass points on the structure where data was recorded.
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The contractor concludes from his study that in-situ testing will be useful in the following area related to equipment qualification:
- establishment of similarity between equipment with consideration of failure modes
- prediction of component specific Required Response Spectra (RRS) component mounting evaluation
- comparison of fundamental building frequency with equipment supporting structure frequency It was also concluded that in-situ testing will not be feasible and suitable for the following applications:
- to establish component / equipment seismic capacity
- to support a continuing surveilliance data base The applications of in-situ testing methods is further discussed below.
Other related topics covered by this contractor's report are described in Appendix B.
(1) Establishment of similarity between equipment with consideration of failure modes.
The most obvious application of in-situ testing method to seismic qualification of equipment in operating plants probably is to establish dynamic similarity between equipment.
As mentioned in Section 1.2, after reviewing the results to date of all the tasks of A-46, it was concluded by the NRC staff that seismic qualification using seismic experience data probably is the most likely approach to develop a qualification method which is both economically attractive to the plant owners and would be acceptable from a public safety viewpoint.
Two conditions will have to be established before the experience data base can be utilized to help qualifying equipment in operating plant.
They are:
(a) To establish that RRS of equipment in operating plant to be requalified is enveloped by the pertinent experience data base response spectra.
(b) To establish similarity between operating plant equipment to be requalified and equipment in the experience data base.
Condition (a) is addressed by No. 2 immediately following and also by Section 2.5.
Condition (b), the question of similarity between equipment, has been touched upon by No. 4 of the staff comments to SQUG Pilot Program Report (see Section 2.4).
The staff's position l
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on the definition of similarity was described as "...for equipment to be similar for the purpose of qLalifying an equipment item on the basis of experience data cr. another item, the safety function as well as the dynamic characteristics, should be similar.
This means that the experience data must include data on performance both during and after a seismic event.
Similarity parameters must include mass distribution, material, size, stiffness, configuration, restraints, and anchorage details...."
Similarity of dynamic characteristics can most effectively be addressed by conducting an in-situ test. Dynamic characteristics of an equipment consists of mode shapes, natural frequencies, mass distribution, and damping.
In-situ procedures identify the natural frequencies and mode shapes.
In certain cases the mass distribution can also be estimated (alternate methods for determining the mass distribution are proposed by the contractor in his report). A characterization of viscous damping is also possible by using in-situ tests which represent the damping that actually occurred during the test.
Since damping may depend on response level, the contractor proposed that values obtained from low level in-situ tests may not necessarily be valid and Regulatory Guide 1.61 (Ref. 8) is recommended for damping values.
The safety function aspect (operability and failure modes) of similarity is further discussed in paragraph 1 of Appendix B.
(2) Prediction of component specific RRS.
In order to seismically qualify a piece of equipment, it is first necessary to establish the specific RRS.
For equipment mounted on a floor, the response can be predicted by the floor response spectra. However, numerous safety-related components are mounted on or attached to the equipment supporting structures (such as electrical cabinets, racks, etc.), the RRS for these components will thus be different from the floor response spectra.
In situations like these, three methods are studied and proposed by the contractor to establish component specific RRS.
Each method will utilize in-situ testing to a different extent.
(a) The first approach is to develop a finite element computer model of the equipment supporting structure and the mounted equipment.
The analysis procedures involved here are those of the typical time history method.
In this process, (1) a synthetic time history is developed from a specific floor response, (2) the modes, frequencies, and modal participation factors are calculated from the model, (3) a time history analysis is performed on each significant mode, (4) the modes are algebraicly combined to determine total time histories, and (5) the time histories are converted to RRS for the t o e
t 4
1 components of interest.
It is felt by the contractor that this basic procedure is potentially unreliable because of system complexity and unreliability of boundary condition modeling.
Consequently, it can only be used if the equipment is already installed and in-situ procedures are used to verify the calculated modal parameters.
A major disadvantage of the approach is that it is relatively costly because of the cost associated with developing a finite element model.
An advantage is that if minor equipment modifications are made at a later date the model can be updated and a new set of RRS calculated.
(b) The second method to generate component specific RRS is an analysis method by utilizing modal parameters directly.
The process involves using the frequencies and mode shapes determined from in-situ procedures directly in constructing a i
numerical solution.
There is no need to develop a finite element model. As with the finite element approach, the j
response of individual modes is calculated and then superimposed for the total response.
The contractor offered several comments on using this method.
First, as the natural frequency increases it becomes more difficult for in-situ procedures to resolve the associated mode shapes.
For seismic analysis it is felt that higher modes, or modes with several antinodes will result in low or negligible modal participation 4
factors.
Consequently accurate calculation of. only the lower mode shapes will probably be necessary. The situation must be checked for every individual case.
The second comment concerns closely spaced modes.
The decomposition of the total i
frequency response into a modal frequency response function is one step in the development of the mode shapes.
Closely spaced mode shapes reduce the accuracy with which the modal frequency response functions are calculated from the experimental transfer functions..The existence of closely.
spaced significant modes could render the direct use of modal parameters infeasible.
It is anticipated that this situation will occur infrequently in which case the alternative of method "a" above can be used to determine RRS. A final comment is that the advantage of the direct use of modal parameters is that the modal parameters are relatively inexpensive to generate experimentally. Generation of modal parameters by the_ finite element method will require substantially more cost. Consequently, analysis procedures which use experimentally determined modal parameters are-recommended to be the prime candidate for predicting RRS in operating plants, i.
~:.
_m.. _ _ _... -. _,
(c) The third method involves direct response spectra transfer.
Procedures (a) and (b) discussed above employed variations of time history analysis where a synthetic time history is used to define the load.
Using these procedures an input response spectra can be transferred to an output location yielding an output response spectra.
Since the input is initially specified by a response spectra, the use of time history analysis in transfering response spectra is essentially artificial and the output response spectra is not uniquely defined by the input spectra.
Methods for transferring the input response spectra in a unique, more meaningful, and less costly way are preferrable.
Direct methods for response i
spectra transfer have been looked at by various investigators.
A direct method uses the input or floor response spectra in combination with the modal parameters and modal participation
{
factor to determine output response spectra.
The associated analytic procedures are algebraic.
Any direct method will eliminate the time history analysis portions of the transfer l
process.
In addition, by using mode shapes, frequencies and i
transfer functions determined from in-situ procedures the need for a finite element model can be eliminated, yielding a very cost effective method.
However, one distinct mode of dispersion, i.e., the feature by which the transferred response spectra become non-unique, seems to exist. This happens where the spectral frequency is near one of the structure frequencies, i.e., tuned conditions.
The acceptance of a method for direct response spectra transfer, in the contractor's opinion, awaits a firm resolution to predicting response at tuned conditions.
A final contractor's position on this matter is expected in August 1983 and may possibly be included in the final guideline for A-46.
(3) Component mounting evaluations.
Mounting inadequacy has been a major cause of retrofit and retest in qualification programs.
The current qualification process essentially qualifies mountings during shake table testing.
For operating plants several options are available.
Analysis procedures using data from in-situ testing can predict the maximum acceleration of equipment. Thus, the loads that mountings must transmit can be predicted.
It should be a straight forward process to assess existing designs.
The main distraction is the large number of mountings that exist.
Enveloping the maximum acceleration could be an approach to reducin'i this workload.
Examining mountings on a theoretical basis may not address some (perhaps the major) problems.
It is pointed out by the contractor that quality of instellation or use of problem prone designs may be a stronger influence on mounting adequacy than strength __
considerations.
To address these concerns, the contractor suggests a physical mounting review by practitioners experienced in' seismic 3
qualification testing as well as current mounting design practice would be an effective mounting evaluation measure.
This process would be enhanced if the reviewers were supplied with an equipment table identifying an enveloping acceleration, equipment weight, and a simple description of the mounting.
The plant walk-down would then screen mountings for those requiring in-depth review or retrofit.
The effectiveness of this process is that it screens out items which are clearly adequate and concentrates more costly review on questionable items.
(4) Comparison of fundamental building frequencies with equipment supporting structure frequencies.
The level of equipment supporting structure response during a seismic event can be related to the corresponding floor response spectra. The design floor response will generally contain a region with significantly amplified magnitude.
The center of this amplified region will generally lie between 2 and 10 hertz and coincides with the fundamental frequency of the building.
The motion of the equipment supporting structure is reckoned as a combination of its free vibration modes whose maximum values are determined from the floor response spectra. Generally the first mode has the largest modal participation factor and is the most important.
Knowing the first mode frequency and its modal participation factor, the maximum response is estimated readily from the floor response spectra.
Tuning of the equipment supporting structure and the building containing it occurs when a natural modal frequency of this equipment supporting structure coincides with the fundamental building modal frequency. As an example, cabinet frequencies between 5-15 hertz are typical so that tuning is possible.
In case tuning occurs, the floor response spectra may result in a response level 2-5 times the predicted non-tuned response. A complicating factor is that the lowest natural frequency of an equipment supporting structure depends on how it is attached to the floor as well as its physical properties.
For instance a welded mounting will result in a higher frequency than a mounting with a minimum number of bolts. Thus for operating plants uncertainties relating to equipment supporting structures include both physical properties and the mounting boundary condition.
Hence, equipment design environment will depend heavily on the relationship between the equipment supporting structure and building fundamental frequencies.
It is clear that most of the safety related systems were not intentionally designed to function in highly amplified dynamic environments (i.e., tuned conditions).,,
l The contractor suggests that systems which may be subject to these loads should be identified by in situ procedures.
Here an abbreviated process can be followed where all equipment supporting structure natural frequencies below 15 hertz are experimentally determined. Mode shape determination is not required.
A modal analysis crew should be able to check a number of cabinets in a single day so cost is not an overwhelming burden.
Where amplified equipment supporting structure response is identified, two options are recommended.
Regardless of the criteria applied to other equipment in operating plants, the contractor recommends that this equipment should be qualified vigorously.
The first option is to determine the design basis environment (or component specific RRS) and qualify equipment to that environment.
The second option is to modify the equipment supporting structure, depending upon which is appropriate.
That a lower response is assured should be verified by in-situ procedures.
2.3.2.2 Summary of Contractor Report " Preliminary Guidance and Acceptance Criteria for Application of Combined In-Situ and Analysis Procedures in Operating Plant Equipment Qualification" This preliminary contractor report covers the progress to date on Topics 6 and 9 defined in Section 2.3.1 of this report.
Twelve technical areas are identified by the contractor that require guidance and acceptance criteria.
The guidance in in-situ testing procedures and pitfalls will be addressed in a final version of this document, presently scheduled to be published in August 1983.
Following is a summary of the preliminary guidance and acceptance criteria in the twelve technical areas.
The contractor has been requested to continue his investigation and provide more definitive guidance by August 1983.
(1) Dynamic Parameters from Tests. Guidance is required on the number and position of nodal points for mode shape description.
Node points are to be located at all significant masses, and there should be no less than four node points between local maximums and minimums of all significant modes.
(2) Analytically Determined Dynamic Parameters.
Guidance relating to analytically determined equipment supporting structure models is that these models are to be verified by comparing computed and experimentally determined natural frequencies.
The analytic and experimental frequencies must correlate to a reasoncble tolerance -
say 10%, for frequencies in the range of interest.. _.
(3) Analysis Methods.
The time history analysis method is currently accepted and the same guidance can be applied to operating plant application.
There are no currently accepted direct transfer methods.
Stochastically/ statistically based methods look very promising.
These methods will be reviewed to determine if their use can be justified.
Providing that the methods are justifiable, corresponding acceptance criteria will be developed.
(4) Functional Similarity and Functional Requirements.
Experience data must identify the functional requirements tested during the real seismic events.
Claims that these functional requirements were tested must be supported by documentation of the methods used to establish these functional requirements.
Acceptance criteria in these areas will be developed in the follow-on effort and available in August 1983.
(5) Experience Data Floor Response Spectra (FRS).
Currently accepted guidance is to estimate floor spectra by analysis followed by peak broadening (Regulatory Guide 1.122, Ref. 9).
This procedure is applicable to experience data minus the peak broadening requirement.
Guidance on the use of the ground response spectra (or some fraction of it) to provide a conservative estimate of the FRS has not been developed.
Specific guidance and acceptance criteria will be provided in the follow-on effort. The criteria will be directed at equipment near the ground level.
See Figure 2.3-1.
(6) Damping. Guidance relating to damping is that experience data for equipment in supporting structures be estimated based on a range of damping ratio values (2%, 4%, 6%, and 10%).
Guidance for correlating operating nuclear power plant equipment supporting structures damping with experience data damping has not been developed yet. Guidance will be forthcoming in the follow-on effort.
(7) Modal Participation Factor (MPF).
Proposed guidance is to determine the mass matrix ([M]) from physical characteristics of the system and calculate MPF according to the following equation:
$$ [M][1] = MPF$
Additional guidance will be developed allowing the calculation of MPFs from dynamic parameters determined from in-situ testing. -.
W) = Fundamental building frequency Floor response spectra 1
l
-l" Ground response spectra i
e I
O jd j
Region in whicn the floor response spectra may not l
envelope the ground response I
spectra l
~' ig[_
N 1
Frequency - hz Figure 2.3-1 Comparison of floor and ground response spectra.
l l
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(8) Fundamental Frequency Determination.
The lowest equipment supporting structure frequency is acceptable if the transfer function in the low frequency range is determined from data maintaining a coharence of 0.8 or greater.
(9) Margin. Additional margin against uncertainties in equipment supporting structure frequencies is not required if peak broadening (Regulatory Guide 1.122) is accounted for in predicting RRS (i.e.,
if time histories used for analysis are consistent with peak broadened FRS).
The RRS can also be predicted using a floor time history without peak broadening.
In this case, two analyses are performed.
In the first analysis, a RRS is calculated using the modal parameters from in-situ testing.
In the second analysis any natural frequencies in the vicinity of floor peaks are modified by 15% in the direction of the peaks. The remaining modal parameters are unchanged and another RRS is calculated. A final composite RRS is developed by enveloping the two response spectra.
(10) Equipment Supporting Structure Linearity.
Equipment supporting structure attached to the floor using bolt attachments must be secured such that installation preloads are not reduced by greater than 90% during the SSE environment.
(11) Enveloping Criteria. As with. current criteria, the Test Response Spectra (TRS) for rigid equipment must envelop the RRS at the Zero Period Acceleration (ZPA).
Envelopment at lower frequencies is not essential.
For equipment supporting structures, envelopment is required only at frequencies greater than the equipment supporting structure fundamental frequency (with 15% margin).
See Figure 2.3-2.
If justification can be provided that equipment is not specifically sensitive to low frequency inputs (i.e., so that the input does not have to be rich in low frequency content to perform a qualification test), then envelopment can be restricted to the remaining frequency range.
(12) Component Mounting Structural Integrity.
Loads on component mounting can be calculated using dynamic parameters developed from in-situ procedures. An acceptable maximum acceleration is calculated using the peak broadened FRS, the modal parameters, and methods from Regulatory Guide 1.92.
The mass is taken as the sum of the component and mounting fixture masses. Assurance must be provided that at least 80% of the component will move as a rigid body during the dynamic loads.
i d -- _.
Wgg = fundamental frequency of NPP building W g = NPP equipment supporting structure fundamental frequency
~
experience data equipment supporting W
ID =tructure'tundamental s
_a i
frequency
=
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N RB ID IN Frequency - bz Figure 2.3-2 Comparison of Envelopment.
1 2.3.2.3 Summary of Contractor Report " Summary of Work Performed to Date on Qualification Cost Estimate Task The objective of this task was to estimate costs associated with the steps of implementation of alternative seismic qualification methods as depicted in Figure 2.3-3.
A table of estimated costs is given in the contractor report. These costs will be used in development a regulatory analysis to support proposed requirements developed by the staff.
Assumptions used to develop the cost estimates are described below.
Eouipment List The equipment list was obtained by modifying the list offered in the report " Survey of Methods for Seismic Qualification on Nuclear Plant Equipment Components (Ref. 10)."
The modifications resulted from a comparison of the list with two complete lists of safety-related equipment for two new plants -- one PWR, one BWR.
Analysis The " Analysis" cost estimates were based on experience in estimating analysis jobs and on reviews of such analyses performed during staff audits of new plants during licensing reviews.
Equipment which has no estimate for analysis is not suitable for qualification by analysis.
Test and Analysis The numbers under " Test and Analysis" represent the cost to determine equipment / support dynamic characteristics via in-situ testing.
These numbers were based on an attachment to the contractor's report. "In-Situ Structural Characterization Test Cost Estimates." Cost of labor, travel of personnel, transportation of test equipment are included in the estimates.
Replacement
" Replacement" is the cost incurred to replace equipment with qualified equipment.
This includes purchase of the equipment with qualification documentation and installation.
It does not' include freight charges.
Estimates are primarily based on " Process Plant Construction Estimating-Standards," by Richardson Engineering Services, Inc.
Two editions of the standard were used, one dated 1975 and the other 1931.
Estimates taken from the 1975 edition were increased by 30% to account for inflation.
Two components on the list (MSIV & CRDM) were not covered by the standard.
Estimates for these two were obtained by contact with equipment vendors.
Qualification documentation was assumed to cost 150% of the cost of the unqualified components for all but three of the components -- small instrument valves, transducers, and relays.
These components are l,
I 1
i produced in large quantities and required in large quantities in typical j
plants. Their qualification documentation is assumed to be less costly
-- 50% of the cost of the unqualified component.
i Comparison The " Comparison" estimate is the cost of comparing dynamic and functional characteristics between equipment in plant and that in the data base.
The estimate is based on the assumption that necessary data is readily available.
Therefore, no costs resulting from analysis or in-situ testing have been included.
The tables in the contractor's report are not reproduced here since they are of preliminary nature.
The final version of the estimation will be included in the regulatory analysis supporting staff implementation requirements.
2.3.3 Staff Conclusions As mentioned in Section 1.2 of this report, as work progressed it became increasingly apparent that Task 4, " Seismic Qualification of Equipment Using Seismic Experience Data" was the most likely approach to develop a qualification method which is both economically attractive to the plant owners and acceptable from a public safety viewpoint.
The application of experience data to qualify equipment in operating plants involves the confirmation of two items.
First, the experience response spectra should envelope the nuclear equipment RRS for frequencies within the range of interest; second, similarity between the equipment in the nuclear plant and equipment in the experience data base has to be established. The study conducted by the contractor indicated that in-situ testing can assist to provide information for both items.
The staff agrees with this conclusion arrived by the contractor.
- However, the staff believes that more specific guidelines and acceptance criteria for conducting the in-situ testing (such as in the areas of test setup, data acquisition and data processing) are needed in order to achieve meaningful and valid information from the in-situ testing.
INEL is presently working on this topic and results are expected in September 1983.
2.4 Seismic Qualification of Equipment Using Seismic Experience Data Base 2.4.1 Backgrcund It is well known that numerous non-nuclear power plants and industrial facilities containing equipment similar to those in nuclear power plants underwent major earthquakes in various parts of the world.
It is also recognized that during the course of qualifying safety related equipment for licensing nuclear plants in the last decade or so, numerous equipment -.
FIGURE 2.3-3 Al'lERNATIVE SEISMIC OllAllFICATION PROCElX1RE FOR USE Willl USI A-% RESllLT De,,c/op nes czn - sifu regast;&cotten Te rf/Ana /ysis or wo/ Re9 ar;re d Generic Specf ra ),
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- 4. Co s f c :far.ainn is metth Or enofysis a >Jy
TABLE 2.4-1 SEISMIC CUALIFICAIION UTILIT/ GECUP l'ESSS EALTIMCRE GAS a ELEcTR!c CCMPArn 30STCN EDISCN CCidPAW CCttCW ALTri EDISCN CCPPMY CCNSCLIIN:s EDISCN COvPANY CCNSt. PES PCha CovPAtu Deir:0!T EDISCN CCPPAt#
DUKE PCb 5 CCMPANY FLcRIDA :Cha CCR. cRATIcN c
R NUCLEAR CCRPOPATICN NEERASKA PuaLIc PCwa DISTRICT NCRTFSN STATES PCWS CGvPMY PENNSYLVANIA Pchs & LIGHT CCe.oANY Roc =2iu GAS & A ecTalc CCPPMY ScuTr-EN CALIFORNIA EDISCN CCtdPAfU ICLEDO EDISc4 CCtdPANY YMxEE ATcMIc ELECTRIC CovPAw l
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i items were tested for seismic capability on shake tables in laboratories. Therefore, there is a wealth of information regarding seismic experience that potentially can be utilized as an alternative to formal qualification of equipment in operating plants.
To use this information the data must be collected and organized and guidelines and criteria developed. Two independent efforts to develop a seismic experience data base were initiated.
The SQUG (Table 2.4-1) conducted a pilot program, " Program for Development of an Alternative Approach to Seismic Equipment Qualification." The pilot program has been completed by their contractor, EQE Incorporated.
Results cf this pilot program were recorded in a two-volume report issued in September 1982 (Ref. 2).
The second effort was one initiated by the NRC staff, with Lawrence Livermore National Laboratory (LLNL) as the contractor.
A draft report
" Correlation of Seismic Experience Data in Non-nuclear Facilities with Seismic Equipment Qualification Nuclear Plants" (Ref. 1) was published in November 1982.
The results of both studies confirmed the feasibility of utilizing non-nuclear seismic experience data to qualify equipment in operating nuclear power plants.
Staff comments on the pilot program report were principally an assessment of what further data collection efforts were needed and suggested guide-lines for acceptability of an experience data base.
The staff's assessment is that use of experience data provides the only viable alternative to current qualification criteria.
Several of the other A-46 tasks will directly support the use of an experience data base.
To seek NRC management endorsement of the concept of using an experience data base before committing additional resources, the SQUG requested a meeting with NRC management in March'1983. At that meeting, the SQUG proposed a Senior Seismic Review and Advisory Panel (SSRAP) to provide consulting services and expert opinion on the use of experience data.
NRR management endorsed the formation of an expert panel at the meeting.
The staff subsequently met with the SQUG and agreed on SSRAP membership and the protocol for use of the panel.
The staff is continuing to work closely with the SQUG and the SSRAP to develop an acceptable approach to using seismic experience data.
In the coming sections the two studies mentioned above will be described first, followed by staff comments on the SQUG pilot program..,.
i l
2.4.2 Summary of LLNL Report " Correlation of Seismic Experience Data in Non-nuclear Facilities with Seismic Equipment Qualification in Nuclear Plants" The study was completed by LLNL and a report issued in November 1982.
This study was intended to answer the question:
Is it feasible to use experience data on the performance of equipment in non-nuclear facilities during earthquakes in addressing issues concerning the seismic qualification of equipment in operating nuclear power plants located in the eastern United States?
The study shows that the answer to this posed question is affirmative.
LLNL's general approach to the feasibility determination is based on the conservative assumption that if experience data can be shown to be equivalent to current seismic equipment qualification requirements, then it is feasible to use experience data.
The basic approach was to develop an overall summary statement evaluating seismic experience data and current requirements, as embodied in twelve different NRC Standard Review Plan sections, Regulatory Guides and national standards.
A comparison of the two summary statements provides the basis for the feasibility determination.
In LLNL's approach, thirty categories (issues) of possible seismic equipment qualification requirements are identified.
That is, seismic equipment qualification standards might be (but presently are not) formulated in terms of requirements and criteria that addresses each of the thirty issues.
Each of the thirty issues was ranked and a minimum set identified.
Table 2.4-2 lists the thirty issues and a brief description of each issue.
The twelve " current requirements" documents which are considered most important in terms of seismic equipment qualification for new plants are listed in Table 2.4-3.
LLNL's evaluation was performed by first reviewing the twelve current requirements in each of the thirty categories in Table 2.4-2, followed by an overall evaluation of these requirements.
The evaluation was performed by ranking the current requirements in the thirty categories using the following numerical weights:
- Adequate - 3:
This is the highest ranking.
It is used to show that the current requirements are judged to adequately address the particular issue.
Adequately means that "the issue is addressed as well as is needed." It should not be interpreted as " ideally" or " perfectly" or that it " addresses the issue as perfectly as can be conceived."
P
L, Table 2.4-2 Categories of Possible Seismic EQ Requirements Category of possible seismic E0 recuirement Brief description of cateoory Physical attributes 1.
Sampling For equipment items qualified by testing, only a limited number of the items installed in a plant is tested.
2.
Simil arity The EQ for one item of equipment is sometimes extended to similar but different items.
3.
Mounting simulation The mounting and orientation used in the qualification of equipment may be different from those of installed equipment.
4.
Peripheral attachments Peripheral items such as electrical cables, small control piping, large piping, and so forth are often attached to the major item of equipment.
5.
Dummy components Equipment is sometimes qualified by testing with a dumy item substituted for the actual item.
For example, an electrical cabinet might be qualified with a dummy component substituted'for a relay.
Seismic loads
- 6. Generic loads Generic loads (loads that envelop all the required design loads for a particular category of equipment) are sometimes defined.
- 7. Enveloping load assumption It is of ten assumed that if an item Sf equipment is qualified for load Ln i
then it is also qualified for-loed I
L, where L1 is greater than L2
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- 8. Required design load Here the question is whether the.k required design load and parameters adequately reflect EQ issues and concerns.
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Categories of Possible Seismic EQ Requirements
- 9. Margin Here the question is whether there is sufficient margin in the capacity of the equipment.
Here the question is whether tolerances
- 10. Tolerances are specified for the required qualification load.
- 11. Single vs. multi-axis testing Here the question is what number of independent test excitation axes are required.
A number of issues are related to the
- 12. Wave form waveform of the test motion imparted to f
equipment.
Strength / capacity
- 14. Fragility This category. addresses whither or not the EQ requirements address' the strength. of equipment, and if so, how.
- 15. Failures This category addresses failures that occur during qualification testing.
- 16. Functional requirements This category addresses the functional performance of the equipment before, during, and after qualification testing.
- 17. Critical parameters This category addresses the parameters that are most important to the survivability or functionality of equipment.
- 18. Degradation under test Here the question is whether the qualification testing has been so severe.that the capacity of the equipment to perform as required in the future can be questioned.
- 19. Response This category addresses the observed response of the equipment during qua'lification testing.
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F Table 2.4 2 ( Cont. ).
, Categories of Possible Seismic EQ Requirements
- 20. Unexpected results Unexpected results include f ailures at unexpectedly low levels, unusual response patterns, and behavior that is inconsistent with predictions.
Seismic and other loads 21. Load combination This category. relates to appropriate combinations of loads such as seismic, thermal, and pressure.
- 22. Load sequencing Load sequencing-is a variant of load combination.
Miscell aneous
- 23. Errors This category includes design, qualification, construction, mounting, and maintenance errors.
- 24. Maintenance This dategory includes consideration of how n6rmal' (rather than erroneous) maintenance might affect the qualification status of equipment.
t
- 25. Mounting adequacy This category addresses the adequacy of the equipment mounting.
- 26. Post Earthquake This category addresses the issue of assessing EQ subsequent to an earthquake.
- 27. Value/ impact This category addresses the benefit of seismic EQ in risk reduction (value) versus the cost of such requirements (impact).
- 28. EQ by analysis This category addresses the issue of performing EQ by analysis rather than testing.
- 29. EQ by testing and analysis This category addresses the issue of perfonning EQ by a combination of I
testing and analysis.
l 30.
In-situ testing This category addresses the issue of the possible role of in-situ testing in I
EQ.
33 -
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TABLE 2.4-3 Documents Most Important for Seismic Equipment Qualification o
U.S. Nuclear Regulatory Commission, Standard Review Plan, Section 3.10 " Seismic and Dynamic Qualification of Meenanical and Electrical Equipment," NUREG-0800, Rev. 2, July 1981.
o U.S. Nuclear Regulatory Commission, Regulatory Guides:
o 1.40
" Qualification Tests of Continuous-Duty Motors Installed Inside the Containment of ' dater-Cooled Nuclear Power Plants,"
March 16,1973.
o 1,73
" Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants,"
January 1974.
1.100
" Seismic Qualification of Electric Equipment for o
Nuclear Power Plants," Rev.1, August 1977.
o 1.148
" Functional Specification for Active Valve Ass 5mblies in Systems Important to Safety in Nuclear Power Plants," March 1981.
o IEEE Standard for Type Tests of Continuous Duty Class 1E Motors for Nuclear Power Generating 5tations, ANSI N41.9-1976, IEEE Std.
334-19/4.
o IEEE Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating 5tations, ANSI /IEEE Std. 344-1975.
o IEEE Standard for Qualification of Safety-Related Valve Actuators, IEEE Std. 382-1950.
l o
IEEE Standard Soismic Testing of Relays, IEEE Std. 501-1978.
o IEEE Standard for 0ualifying Class 1E Motor Control Centers for Nuclear Power Generating 5tatious, IEEE Std. 649-1980.
o Sel f-Ocerated and Power-Operate.d Safety-Related Valves Functional Specification Standaro, AN5I NZ/8.1-19/5.
l n.....
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e TABLE 2.4-4 So-ary of reasibility tvaluation.
)_
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SRP 3.10 I
R.G. 1.40 l
R.G. 1.13 l
R.G. 1.100 l
R.G. l.143 IEEE 5td. 334-1914l lEEE 5td. 344-1915 l
l.
Sampling Sampling is
- A
- prototype unit
- A
- prototype unit' lA
- prototype unit
- Impliclt acceptance of acceptable.
lto be tested under to be tested under lto be tested under I sampliag. at least for Sample slae is lmost adverse
- most adverse imost adverse l
- cases where fragility not defined.
l design conditions. I design condlilons. l design conditlans.
l testing is performed.
l l
l l
l l
l l
1(stension of (Q by test 2.
..lel & !!)
N 1
l l
l lto slallar equipment is l
l l
l l
l allowed using a combl.
I I
l l
1 I
laation of test and l
l l
1 l analysis.
i I
I I
1 3.
Mnunting
)The flature l
l l
l lIhe equipment shall be s imulation l design should l
l l
l mounted in a manner simulate the l
lthat simulates the
- actual service l
l Intended service leounting.
I l
l
[ mounting.
i l
I
=
4.
peripheral lHajor peripheral l l
l l
lIhe effects of 8
c'.tachments l attachments are l
l l
l peripheral attachments l addressed.
l l
lsuest be considered.
I l
l l
1 I
I I
l 5.
Dummy l Dummy specimens l l
l l
l lUse of dunay specimens compor.ents lare allowed to. l l
l l
l lIs allowed.
=
lsimilate the l
l l
l l
l l mass effects and l l
l l
l l
l dynamic coupling l l
l
(
l l
,to the supports.
l l
l l
ll l
6.
Generic l
l l
l leads l
I I
l l
1 l
l l
1 l
l l
l l
1 l
1 7.
Enveloping
'het clear l
l lThe assumption l
l Ilhe assumption is made.
load juhether the l
l lls made.
l l
l assumption lassumption l
l l
l l
g" lis made.
l
.l l
l l
1 l
l l
1 l
l 8.
Itequired l
l l
l l
l design load l l
l l
l l
l l
j.
l l
l 1
i
- A blant ladicates that no requirement was found.
e
~
TABLE 2.4-4 (Cont.). Summary of f easthility Evaluation.
First Seven Sources 1
I l
l
"~~~
l I
IF*E. ~,te. 3a4-1975 Category l
SNP 3.10 l
R.G. l.40 1
R.G. l.73 l
R.G. 1.100 i
R.G. 1.148 llEEE 5td. M4-1974 :
l l
l I
I I
3.
mrgin N rgins are l
l l Margins are I
l llos margins are specl-required but l
l Irequired but i
I j fled for the response Inot specified.
I l
lnot specified.
I l
1 spectrum at the l
j mounting point of the 1
I l
l equipoent.
I I
- 10. lolerances i l
I
~
lMultiasts testing is ll. Slagle vs. livo simultaneous I
multlants lanes of input l
l l
l l suggested. Single ants j
~
testing lare generally l
l l
l l testing is alloueJ if i
trequired, k
I l
l construative, or if the General proce-l l
l responses in the ases ldures are i
I I
l are independent.
I I specified.
l l
l I
l l
ca l
- 12. Idave form lihe character-(
l l
Requirements for slau-listics of the l
l l
lating earthquake are e
' required input I
l l
)given. Specific should be l
l l
requirements for proof
, specified by l
l l
l ltesting are specified, response l
l l
l Ilspectrum or time' I l
l l
l l history methods.
l l
l l
l l
-l I
i l
- 11. fatigue 15tructural l
l lrerformance must I
lihe requirement is lintegrity and l
l lhe assured during l l
[five 08[s plus art 55[.
l operability l
l land af ter an 55f l
l g
must be assumed l l
fpreceded by l
l l
lunder an SSE l
l Iseveral c6ts.
l l
l l preceded by l
l l
l 1;
l
,several 08ts, I
l l
l l
l 1
l l
l
. fragility testing
- 14. fragility l
l reconomerufed, but not l
I l
l required, for equipacnt l
l l
l l
) to be used in a nua6er I
l l
l l
lof appllcations.
l l
l l
l l
l r
t
O TABLE 2.4-4 (cont.). Su ary of reasibliity Evaluation.
First Seven Sources Category l
SRP 3.10 it.G. 1.40 1
R.C. 1.73 l
R.G. 1.100 R.G. 1.148 11(([ 5td. 334-1974 IffE Std. 344-1975 i
- 15. Failures l
l l
l l
l l
Functional 10perationality l l General, Indirect,, Reference is made, 5elsmic loput is for devices (relays, require.
Ishould be l
l treferences to
- to AN51 N278.l.
l assumed to occur
- motors, sensors), it is ments lvertfledduring l l
lf unctionality are l1975.
julth motor stand-assumed that the land /or after l
[given.
l lstill, starting. lselsmic Input can be ltesting.
l l
l l running,or limposed while simul-l l
l l
lccasting down.
llating normal operative l
ll l
l l
l
-land sensing perform-i l
l l
l l
lance.
l i
l l
l l
1
- 11. Critical 1
15cee parameters are parameters
- suggested as possibly (critical and are e
lrecoasmended for identl-d l
IfIcation.
i
- 18. Degradation l
l l
lUpon completion under test l
l l
lof the test, the ll l
l l
l l
[ motor shall be I;
l l
l l
l l dismantled and l
l l
l l
l l Inspected.
l ll
-1 I
l l
=
l l
I l
- 19. Response l
l l
l l
ll l Monitoring is required, 1
l l
l l
l lbut specific require-l l
lments are not given.
l l
l 1
l l
1 I
i
- 20. Unespected l
l l
l l
l l Analysis alght be used i
results l
l l
l l
l lto esplain unearccted l
l l
l l
l ibehavior during a test.
l l
l 1
- 21. Load
'It is not clear l l
l l
l I; normal operating loads combination l what combine-l l
l l
l lwhich adversely af fect Jtions ase l
l l
l l
[ function must be lacceptable.
l l
l l
l lccelned with seismic i
l l
l l
l l
l loads.
I l
l 1
1 I
I
a TABLE 2.4-4 Icont.). Su ary or reasibility tuluatloa.
First Seven Sources l
i I
Sar 3.10 l
n.G. 1.40 t
n.s. 1.73 a.c. 1.100 l n.c. 1.14e l ltEE 5td. 334-1974 litt 5td. 344-1975 I
I l
- 22. Load l Load sequencing l l
Itoad sequencing l
l l
sequencing lIs to follow l
l Ils ladirectly i
I I
IIEEE Std. 323-l l
l addressed.
l l
l Il9M.
l i
I I
i l
1 I
I I
- 23. Errors l
l l
l l
l 1
l 1
1 I
I I
- 24. Malatenance I l
l
- 25. Mounting lacquirements I
i l
l l
lThe mounting method adequacy los mounting i
l l
l l
lshall be the same as ladequacy are l
l l
l l
Ithat reco. mended for lgiven ulth I
l l
l l
l active service, trespect to l
l l
l l
1 l testing and/or l
l 1
l l
l lanalysis I
(
l I
l l
lassessments.
l l
1 I
i e
i i
i l
I
- 26. Post
(
l earthquake l
l l
l l
l l
l l
1 l
l i
i I
l l
l l
- 27. value/
i I
I l
l l
laput l
1 I
i i
l I
I I
I l
l l
1 l
l 1
l l
l
- 28. EQ by lEQ by testing l
l tQ by other l
1EQ by analysis is not analysis lIs preferred.
i l
Itesting is l
l l generally gece W aded l
llePilCitly l
l luf th3ut test eacht l
l laccepted by l
l luhere structural l
l l
littE Std. 344-l I
Itategrity alone can l
11975.
l lensure equipment l
l lfunct*on.
i I
I J
l I
l i
e TABLE 2.4-4 (cont.). 5 nary or reasintlity tvaluatloa.
First Seven Sources I
l l
l l
l L
' ~~~
Category l
SAP 3.30 l
R.G. l.40 l
R.G. 1.73 l
R.G. 3.100 I
a.s. I.Its ll(Et 5td. 334 1974 ( Ifft 5td. 344 1975 l
I l
l 1
I I
I l
l l
l (Q by combined testing
- 29. EQ by
~ l l
l l
l and analysis is accept-testing and l
l l
'able, but only vaguely defined.
analysis l
l
- 30. In. situ lIn site testing l l
l l
l In situ testing can testing is not required. l l
l H
l lbe a part of fQ by but it is l
l l
l l combined testing allowed.
l l
l l
land analysis.
I i
1 I
l D
TABLE 2.4-4 (coat.l. Suasaary of feasibility evaination.
Sources 8-12 and Other Data ll il l
I l
15 core on l l
l l
l l current l
l5 core on IEEE 5td.
life Std.
l It[E 5td.
l An51 N218.l-l ANSI S16.41-l require-l leaperience l 382-1960 501-1918 l 649-1980 l
1915 l
1981 l ment.*
l Esperience data ll data
- l l
l l
l 1
I l
l I
l.
Sampling 14 procedure (A minimum of lAt least one l
l Testing of at l 3 li5everal units are ll 6
l suggested for lthree specimens l device must lie l lsust one saagile. l lconnonly escited at l
lsesecting the lls required.
l tested, but not l lls acceptable.
l lance by an earthquake. l Itest units is l
lone motor l
l l
lIherefore,esperience l lgiven in App. A. l l control center. l l
l l data are potentially l l
l l
l l
l l rich in sampling, l
l l
l l
I l
1 I
I 2.
Stellarity 151milarity is IExtension of l General gulde-l lGuldelines are l 3 l Equipment among non-l 6 l addressed la lquallfled relays lIlnes are given l Igiven to entend l l nuclear f acilities is l l terms of generic lLo relays not lto estead the l
l qualification l
lusually quite similar. l l groups of valve l tested is l qualification lof value l
l A casual comparison l
l actuators from l allowed.
lof motor control l assemblies to j
lalso indicates that l
'lwhich test units l l centers to other l$1milar units.
]
lthe equipment is also l lare drawn.
l l units.
l l
lquite stellar to l
l l
l l
l l
lthat in nuclear l
l l
l l
l
[ facilities'.
l I
I 8
3.
Mounting lihe valve actu-Ilhe relay must lThe motor l
l l 4
- Esperience data ll 6 A
simulation lator is reeutred lbe mounted as it.l control center l l
l jreflect the true l
O lto be mounted to lnormally would lsust be mounted l l
l l mounting conditions.
l e
lthe shaker table lin service.
las it would be l l
l Ilherefore, mounting is l las it would be l
lIn a plant.
l l
l lnot an issue for such l
- attached to the l l
l l
l l data.
l valve.
l l
l l
l l
l
.l l
l l
l l
1 l
l 1
I l
I L
4.
peripheral Elec trical.
I lAntlCipated l
l Electrical.
l 4 lThe Credibilty of l 6 at tachment s lhydraullc or l
l additional l
lhydr'aullc. or.
l l effects from l
l pneumatic l
l weight and l
l pneumatic l
l peripheral attach-l l connections must l lesternal connec-l l connections ll lments is not an issue l lbe attached.
l Illons shall be l lst.all be j for espertence data.
l l
l lslaulated.
l required.
[
t l
i l
1 l
l l
l l
1 I
l S.
Dussey l
1 l
i I
l 4 lDuniny specimens do not l 6 components l l
l l
l l represent an issue for l l
l l
l leaperience data.
l l
l I
~l
~
6.
Generic l Generic loads l Fragility lCeneric load [1] l l
lNot I Not loads lfor valve actu-testing is l techniques are l l required.
l l required.
lators are regulred for l allowed for l
l l
l lestablished for ' relays; there-Igroups of equip-l l
l l
lmost plants.
(fore gencric l ment.
l l
l l
l l
l loads are l
l l
l l
l l
-lessentially l
l l
l l
l l
Irequired.
l l
l l
l l
l l
l l
l l
l l
- 5ee Table Bl.
TABLE 2.4-4(cont.). Susaary of feasibility evaluatio...
Sources 8-12 and Other Data I
I I
l i
15 core on i I
l l
l l
l Icurrent l
l5 core on l IEEE 5td.
l IEEE Std.
l IEEE 5td.
ANSI N278.1-l ANSI B16.41-l require-l leaperience 1 382-1980 l 501-1978 I 649-1980 l
1975 l
1988
- ment.*
i Emperience data Idate+
1 I
I i
l 1
I I
i 1
l i
7.
Enveloping l Enveloping is l
l l
l 1
2 lEspertence data could l 2
lugd lprobably estab. l l
l l
l l provide an indication I assug tlen lIlshed through l
I l
l cf equipment perform-1 ance at loads that l
l generic loads.
l l
l l
,l envelop required loads l l
l l
l l
l l
l l
l l
Ifor Eq.
l l
1 l
l t
t l
l l
I
- a. 8equired I lhe required I
l l
l 6
l Although loads esperi-l 3
design l design load l
l l
l Jenced are realistic, l
load lmay be deficient.l l
l l
l hthe adequate reflec-l l
l l
l l
ltion of such loads to l
l l
l l
lareas of concern la l
l l
l l
lEQ of nuclear plant i
l l
l l
l l
l equipment day be l
l l
l ll l
l l lacking.
l l
l l
l l
t i
I I
I g.
Margin
'IMargins are (fragility llMargins are l'
I I
3 15cee evaluations l 6 A
l Included in the Itesting includes lspecified 14 l
l l
Ilndicate that some non-l generic leads.
Ithe concept of liable 1.
l l
l inuclear facilities l
e l margins.
l l
l l
lhave esperienced l
l l
l l
l lselsmic loadings l
l 1
l l
l l
lIn escess of design l
l
- l l
l l
1 loadings in nuclear l
l l
l l
l facilities.
l 1
l l
l l
l l
I I
I I
I l
l l
l l
1 I
I i
- 10. Tolerances l
l Tolerances are [
l l
lNot l
lNot l
lspecified for l
l l
l required. l l required.
l l Instrumentation,l 1
l l
l l
l l
l l
l l
l l
1 l
t i
I I
l l
l 1
1 I
I I
II. Single vs.
l$lantal testine lTrlantal testing l l
l 6
lEsperience data la i
6 l general ccasist of I
multlaats l required.
lIs desired. but l l
l l
l lblastal testing I l
l l
three-dimensional l
lis acceptable.
l l
encitation.
]
i l
l I
- 12. Wave fors l Requirements are lTuo multi-l l
l g
flaputs In esperience l 6 l consistent with lf requency, l
l l
l data can be either l
llEEE Std. 344-l standard l
l l
l narrow banded if the I 11914.
Iresponse spectra l l
l lequipment is mounted l
l lare specirled l
l l
los a structure or l
l jfer goallflCa.
l l
l piping system, or l
l ILlon of relays. l l
l l
l broad banded if l
l l
l le l
l lmur.te4 on the l
l l
l l
l jfenedclea.
l 1
I I
I L
i
)
l
TABLE 2.4-4 l coat.l. Sumary ef faasibliliy evalution.
Sourcss 8-12 sad othtr Esta l
l l
l l-l5 core on l l
l l
l l
l current l
15 core on l IEEE 5td.
l IEEE Std.
l IEEE Std.
ANSI N278.1 l ANSI 816.41-l require. l lesperience i
1 382-1980 1 501-1978 l 649-1980 l
1975 l
1981 Iment.*
l Esperience data l data' l
I l
l 1
l l
1 1
1
- 13. Fatigue
.08E and SSE Five 00E plus an l
3
, tow-cycle fatigue may I 3
itesting are 155E testing are l
l be revealed by esperl, I l required. Each frequired. Mini l lence data.
l Itest must be laum duration is l l
l l
l l
l15-s minimum.
Il5-s per test l
l l
l l
l l
I l
l l
l l
l
- 14. Fragility l
l Fragility l
l l
l 3 lPresent Indications l 3 l testing is l
I l
l lfrom a limited review l I
l l
lof esperience data l
l required for l
i l
l relays.
l l
l l
lsuggest that few or l
Ino f ailures of equip. l
.I l
l ll l
iment will be observed. ll l
l l
l l
l l
I i
i I
3 lFallure information l
3
- 15. Fallures IDetermination of I I
l l
lwhat consitutes I
l l
l lmay be limited.
l I failure for l
l l
l l
l 1
Irelays is given. l l
l l
ll l
l l
l
- 16. Functional
.Ivalve actuators IRelays must be IMotor contr'ol l Valve assemblies IFunctional I 9 Esperience data on l 6 requirements lsust t,e func-l tested in the l center opera-lsust be operable l requirements are l Ithe functionality of l e
ltional before.
liransillon from ltional capa-lduring and after Igiven for valve i lequipment may be l
lduring. and lnonoperating Imust be lthe' test.
l assemblies.
l lrelatively scarce.
3ro lafter testing.
lto operating
! demonstrated.
l l
l l
I l condition.
l l
1 1
1 I
l I
l l
- 17. Critical l
l
'I I
8 l
1 I
I 1
15 tace few or no.
1 0 parameters l l
l Ifallures have been l
l l observed, it is l
l l
ll lunlik.ely that experl. l l
l l
1 l
l lence data will reveal l i
l l
l l
l l
lcritical parameters.
l l
l l
l l
lThe most important l
l
-l l
l jfallures observed I
l l
l 1
I l
lhave biten f ailures of l l
l l
l l
1 Imountings or attach-l l
1 l
l l
lments.
l I
I l
l l
I I
I I
i
- 18. Degradation l l
l l
lintpection of 0
l Degradation is l
3 under test l
l l valve assemblies
- l 19enerally not an l
l ishall be l
llssue for esperience l l
l
[
lperformed before l l data.
l l
I I
l and after l
I l
l testing.
l l
l l
l l
l l
l l
l
'See Table 81.
Table 2.4-4 l cont.). 5 ary of feasibility evaivation.
Sources 8 12 and Other Data l
l I
~1 15 core on i I
l l
l l
l current ll l5 care on l IEEE Std.
I IEEE Std.
I IEEE Std.
I ' ANSI N278.l.
! require. l l,esperience 1 382-1980 1 501-1918 1 649-1900 1975 l
1981 l ment.*
l Emperience data Idata*
I L
I I
l i
- 19. Response l
l l
l l
lNot l
lNot l
l
[ required.
trequired.
i l
I
- 20. Unespected l,
l, 1
I l
l lNot l
l Not results l
l l
l l
Required. I
, required.
l l
l I
- 21. Load l
15elsmic testing I
l l
l 6
INormal operating loads l 4 combination l lof relays can be l l
1 1
1104d5 are espected to l l
l performed under l
,be present already l
lprevJiling li l uhen an earthquake l ambient condi.
occurs.
I l
ltions of the testl
,l i
l llaboratory.
l l
l l
I I
I I
I
. 22. Lead lA standard load l l Sequencing of I
llA sequence of l
6 Equipment in operating
, 4 sequencing l sequence is 1:
lpreaging and l testing is I
l plants can be espected Il Irequired.
I lselsmic testing lspecified for i
to have experienced l
t e
lls specified.
lvalveassemblies.l normal environments, i
l I
l lltransients, and I
9
~a l
l l
1 l
l lIn-situ vibration.
l I
I I
.I I
I O
l Equipment la plants 2
,l l
l lrresumably has been l
- 23. Errors l
l l
l l
l installed with a more l il l
l l
l l
lor less typical set l
I l
l ll l
ll lof errors.
l I
l I
I
- 24. Malatenance
.Malatenance to belMalatenance can l Modifications i
Ilf maintenance,
0 Emperience data should 2
l performed during lbe performed lduring testing l lor adjustments l
lbe valuable in assess-l lthe test must be laf ter a given lshall be evalu-I lare required l
lIng if, and hou. rain-l lspecifled.
l fragility test. lated to deter-l lduring testing. l ltenance affects seismic l l
lmine their i
lacceptance of I
l performance.
l l
leffect on the l
lthe test r.;Jst be l
I lEQ.
l evaluated.
l 1
I I
. 25. Mounting lThe valve actv-l Recommended IMounting must be l Ilhe valve 9
jFellure of mountings 6
adequacy later must be l mounting hardwarelby velding or l assembly must be l l appears to be the l
1 ishake table as l
lselsmic testing.
, supported as l
l single most important i l mounted to the lmust be used.
Ibolting for required to l
,fallure; therefore.
l lit would be l
l 1per'mit testing i l eaperience data can be l
~
\\
l mounted to a l
l lIn accordance lespected to provide l
l valve.
l l
l lulth the luseful Information l
l l
l ll l standard.
tregarding mounting l
l l
l l
l sde t;u acy.
l l
.l l
1 1
4 3
e c
nn d
oe e
t r
er 1
r e 'a i
5 9
I u
I I
opt t q cs a oe 5ed N. R ll l
lll
- llIlllll Ill ll lI_
-c t
r s
el a
dn ee a
ob o cu o
nl f
ll e
t ut nf - r.
et diui eo ibb l
s eet oy r
e odk as c
a d
d,
.l ssil aes inswean st e u.
ruovl li y i nmo spu ne t
esen a
sit oce paa ab r
bsai oeq k
i d
o cit pe hi t aga m yt f ot pcirdi peet esbget anfn ol aet a c nanl e
skja
,e nbr de i cadd u
eaheaa c
eabr di a
msd n
ll b
cl sep eaar f aeel a.
pe css n
uue s
,l e
t q s pt eusk c
o o
cbave ldot uee nh on r oeay nss d n t eu i
l nil r
et y eohsul eatli eneoe l
ueuntt e
mrllif ssqn i
sg.
sai tdda avt oil p
ipat a s e aho rt yus u
r ov t einll s
enmn r a t
esl on t eeeo sds ii E
u er a et rrt paa r a espvrgs
- rcc qnuorhaoau senhe h es ao t
tonoaa E aqntIdf eb El at m It eh mai ItIfff Il llIl l
I llI llI
'Il
~
d t e e
nr*
ir ei.
rut 2
u 1
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Moderately Adequate - 2:
This is the next highest ranking.
Marginally Adequate - 1:
This is a poor ranking.
The issue is addressed, but not very satisfactorily.
- Inadequate - 0:
This is the worst ranking. The issue is either not addressed at all or, if it is addressed, it is addressed poorly.
- Ranking not required.
This ranking usually occurs when an issue does not have to be addressed, but is included for completeness.
Next, the use of experience data was also evaluated for each of the thirty categories.
The same ranking as above was used.
These rankings were then weighted according to importance, and the two sums (current requiements and experience data) compared to arrive at a feasibility judgement. The results of the evaluation is summarized in Table 2.4-4.
Table 2.4-4 shows that when the current requirements in existing NRC and national standards were evaluated against the common set of thirty issues, they were estimated to score 91 out of 156 overall, or about 60%.
Experience data was estimated to score 97 out of 156 overall, or also about 60%.
Since the current requirements and experience data score about the same (60%), this led to LLNL's conclusion that it was feasible to use experience data on seismic equipment qualification-issues.
Besides the feasibility study, LLNL's report also addressed recommended guidelines for the use of experience data.
For all the categories considered to be the most important (those given an importance ranking of 3), guidelines were developed. The categories considered are:
- Sampling
- Similarity
- Required design load Margin
- Single vs. multi-axis testing
- Wave form
- Fragility
- Failures
- Functional requirements
- Mounting adequacy j
- t
The guidelines, as taken directly from the LLNL report, are combined under the five headings as follows:
Samplina (1) Experience data should be gathered on all non-nuclear facilities that have experienced (a) a significant earthquake, or (b) failures of any kind or either temporary or permanent loss of functional capability. We anticipate that 10 to 50 facilities will fall into this class.
If fewer than ten facilities, three significant earthquakes, or all facilities that have experienced some kind of mechanical, structural or i
functional failure are included in the data base; then we do I
not recommend that the NRC accept experience data as fully as we have otherwise recommended.
(2) The numbers of each type and size of affected equipment should be obtained for each facility in (1).
If fewer than three items of each type and size of interest are found, then a justification must be provided to extend the experience data.
I Similarity (3) The issue of the similarity of equipment in non-nuclear facilities to equipment in nuclear facilities must be addressed. However, exact similarity need not be established.
Rather, what is required is reasonable assurance that the equipment in non-nuclear facilities
- is of the same type and basic design
- was manufactured by the same manufacturers in the same period as the equipment of interest in nuclear facilities.
Required Design Loao, Wave Form, and Dimensionality (4) The approximate location of each item of equipment in non-nuclear facilities must be established in order to obtain a " rough" idea of the type of earthquake motion it experienced.
" Rough" means that dynamic modeling or analysis is not required. Two categories are suggested:
a.
Dimensionality. Was the earthquake motion affecting the equipment predominantly one, two, or three-dimensional in nature?
l b.
Wave form. Was the earthquake motion affecting to the equipment:
- random like an earthquake (as for equipment in the foundation or free-field) i
- random because of superposition of a number of narrow-band pass motions,'each with a different center frequency (as for horizontal motions on equipment in the lower elevations of a structure)
- sinusoidally random, that is, essentially a single-band pass motion (as for horizontal motions on equipment in the higher elevations of a structure).
Criteria are difficult to establish in this area, because in some respects they are dependent on the motions expected for the equipment of interest in nuclear facilities. However, if the experience data indicate significant two-or three-dimensionality of motion and sinusoidally random motion with a mix of center frequencies, then the experience data are acceptable.
Margin (5) The facilities in (1) should be selected in order of decreasing severity (for example, peak acce}eration) of earthquake, that is, the most severe earthquake first. A reasonable assurance of margin for plants in the eastern U.S.
is provided if the experience data are obtainea from earthquakes with a peak acceleration greater than the SSE peak acceleration for the nuclear plants of interest and the duration is greater than 10 seconds.
However, inevitably questions will arise about the more detailed aspects of the motion affecting the equipment in non-nuclear facilities (for example, in-structure response spectra) and its relation to similar motions in nuclear facilities.
Th'e authors believe that the above requiement for acceleration and duration provides reasonable assurance on the issue of margin, and nothing further is recommended.
If, however, the NRC decides that more needs to be done on the margin issue, then three steps are recommended:
)
l 4
4.
I a.
As a first step, realistic analyses can be performed on the non-nuclear facilities.
For example, a comparison of realistic non-nuclear and nuclear design in-structure spectra, as in Ref. 2, may establish the required confidence in margin.
b.
If a. is not chosen or if it does not indicate margin is present, then the following may be an acceptable alternative.
Realistic, best-estimate analyses, with uncertainties explicitly characterized, as in Ref. 11, should be performed on both the'non-nuclear (for the earthquake that occurred) and nuclear (for design earthquakes) facilities.
The median of the two results should be used as a measure of whether or not adequate margin exists.
For example, median in-structure spectra from the two analyses can be compared.
As part of either a. or b. above, margin is assured if, for c.
example, margin exists at the frequencies of interest but not at some other frequencies in the spectra.
Fragility, Failures, Functional Requirements, and Mounting Adequacy (6) A vigorous effort to seek out failures or incipient failures in experience data is required.
In addition to mechanical or structural distress or failure, incipient or actual functional failures should also be sought. This effort includes examination of plant system logs and interviews with plant operators or other personnel present during the earthquake.
The six guidelines above are concerned with experience data obtained from non-nuclear facilities.
The next three guidelines are concerned with actions recommended for nuclear facilities.
Functional or Other Failures (7) Nuclear plant equipment should be examined very closely for any and all failures revealed in (6).
For example, experience data suggest that mounting failure is the single most important cause of failure of equipment.
All nuclear equipment of interest should be examined for adequacy of mounting or attachment.
(8) The NRC should develop a detailed and definitive check list to aid in a " walk-down" of equipment of interest in nuclear plants.
Such a walk-down should then be performed in each operating nuclear power plant where there is concern about the seismic adequacy of equipment. The items and procedures in the check list should be drawn from three sources:
l l
l,
a.
Information gathered from the collection of experience data; b.
Information gathered from laboratories experienced in seismic equipment qualification testing; c.
Recognized experts who have performed walk-downs in the past.
(9) A limited amount of shake table testing should be performed on equipment obtained from operating nuclear power plants to confirm the perceived strength of equipment. This testing should satisfy the following:
a.
The test objective is to obtain the " capacity" of each equipment item tested.
Capacity includes:
- incipient or actual " structural" failure
- degradation of or loss of function
- identification of failure modes and key parameters related to failure or capacity
- anomolous behavior An example of such testing can be found in Ref. 12.
b.
The equipment should be tested while functioning or in such a manner that capability of function is assured.
c.
The equipment need not be artificially aged or subjected to loads or environments other than seismic.
d.
The equipment should be tested as is.
That is, it should not be modified, adjusted, disassembled and tested separately, etc., after it is selected for removal or removed from the plant.
The testing should be limited in the number of categories of e.
equipment tested but comprehensive in addressing each operating plant and category of equipment.
For example, one item of each category of equipment should be obtained from each category of equipment, and the same test program executed for each.
f.
The number of categories of equipment should be limited.
The selection of the category of equipment to be tested should be based on importance, estimated vulnerability, (that is, choose a category that is believed to be relatively weak rather than strong) and diversity of equipment type.
For example, these objectives may be satisfied if the testing is limited to:
I I
l I
l
- 125 v vital bus (electrical equipment)
- motor operated valves (mechanical equipment) g.
The above requirements may lead to testing on the order of 100 items of equipment, depending on the number of plants involved. As an alternative to 100 tests on only two categories of equipment, as outlined above, a minimum of five tests on 20 or so categories would be acceptable.
2.4.3 Summary of EQE Report " Pilot Program Report-Program for the Development of an Alternative Approach to Seismic Equipment Qualification Numerous non-nuclear power plants and industrial facilities containing equipment similar to those in nuclear power plants have experienced major earthquakes in various parts of the world.
A sample of this experience is shown in Table 2.4-5.
The SQUG with help from EQE, initiated a pilot program to evaluate the potential for using experience data as the basis for qualification.
Stated goals of the pilot program were:
1.
To develop a historical data base on the performance of equipment in power plants during and after strong earthquakes.
2.
To show that much of the equipment in those plants is similar to equipment found in nuclear power plants.
3.
To determine whether data from actual earthquakes are sufficient to conclude that seismic qualification by conventional methods is not necessary for certain classes of equipment.
4.
To develop a methodology for using earthquakes data to evaluate the necessity for seismic qualification of specific items of equipment by conventional methods.
2.4.3.1 Methods Used in the Pilot Program Two types of facilities were addressed in the pilot program:
nuclear power plants and non-nuclear power facilities that have experienced strong earthquakes (also referred to as data base plants by the SQUG).
.The steps involved in collecting data from the da'.a base plants and the nuclear power plants and in comparing the data are shown in Figure 2.4-1.
Before walkdowns of the data base plants were conducted, available records of the seismic event at each site were collected.
These data included ground motion traces recorded near the plant sites.
Facilities that had experienced significant ground motion and that also appeared to contain equipment appropriate to the investigation were selected for visits and walkdowns. _
l TABLE 2o4-5 SELECTED MAJOR EARTHOUAKES THAT HAVE AFFECTED POWER AND INDUSTRIAL FACILITIES (Yanev, 1981) i l
l Recorded Estimated Peak Number Approxi-Ground of Power mate Accelera-Ground Plant
. Richter tion Motion Units Earthquake and Location Year Magnitude (g)
Records Affected
- 1. Eureka, CA 1980 7.0 0.15+
8 3
i
- 2. Imperial Valley, CA 1979 6.6 0.81+
50 4
'.4 0.40 100+
10+
- 3. Miyagi-Ken-oki, Japan 1978 7
- 4. Friuli, It.aly 1976 6.5 0.30+
30+
7
- 5. Eureka, CA 1975 5.5 0.35 Several 3
- 6. Point Mugu, CA 1973 5.9 0.09 10+
4
- 7. Managua, Nicaragua 1972/3 6.2 0.60 4+
3
- 8. San Fernando, CA 1971 6.5 1.25 60+
20 +
- 9. Caracas, Venezuela 1967 6.5 Several*
- 10. Seattle, Washington 1965 6.5 0.08 3
Several
- 11. Alaska 1964 8.4 7
- 12. Niigata, Japan 1964 7.5 0.18+
Several Several
- 13. Chile 1960 8.5-None Several
- 14. Kern County, CA 1952 7.7 0.13 5+
1
- 15. Long Beach, CA 1933 6.3 0.15+
Several 5
+ Indicates equal to or greater than the number shown.
- Actual number not determined.
I
~.
Figure 2.4-1 ElED USED IN PILOT SR0Y NUCLEAR FGER RJNTS DATA BASE PUWTS REvtEW TYPE OF EQUIPMENT REVIEW RECORDS CN FACILITIES Wi!CH HAVE EXPERIENCED EARTHQUAKES 4
v SELECT REPRESE. TATIVE Pl. ANTS SELECT REPRESENTATIVE V
AND ECUIPMENT AND PERFCRM PLANTS & PERFORM WALKDCMfS WALKDCWNS v
v SELECT PLANTS & ECUIPMENT SELECT Pts,tCS AND EQUIRENT FOR DETAILED SN PLING FOR DETAILED SAH.ING 1
Y COLLECT ECUIPMENT DATA COLLECT EQUIPMENT DATA AND AND FLOOR RESPONSE SPECTRA FLOOR RESPCNSE SPECTRA COMPARE EQUIPMENT DATA AND RESPONSE SPECTPA 4
Dcic. dine IF EQUIP!ENT REQUIRES DETAILED QUALIFICATION 9.
4 a
Preliminary and final walkdowns were conducted at both the nuclear power plants and the non-nuclear facilities.
Preliminary walkdowns at the nuclear power plants were used to identify types of commonly encountered safety-related equipment.
Preliminary walkdowns at the non-nuclear facilities were used to record the locations of types of equipment that are similar to nuclear power plant equipment.
Following the walkdowns, particular classes of equipment were selected to be the focus for the remainder of the pilot program.
Final walkdowns were used for collection of detailed data, including conducting in-situ dynamic testing.
Low excitation level in-situ testing was conducted on approximately 200 pieces of equipment in the data base and nuclear power plants to determine approximate primary response frequencies and mode shapes.
This permitted estimates tv be made of equipment response to floor motion.
Seven classes of equipment were selected for detailed study (see Table 2.4-6).
Each class was reviewed to determine similarities between equipment in the two types of power plants.
The following characteristics were examined to establish similarity:
primary structural and functional characteristics; dimensions and name plate data; and ranges of dynamic-response frequency. The response frequencies found during the in-situ testing were compared to determine whether the equipment in the data base plants and the nuclear plants could be expected to have similar dynamic response properties.
It was noted by the SQUG that most of the equipment of interest in the data base plants is located at grade, in basements, or in the first two floors of the structure (up to the turbine decks).
In addition, most of the data base structures are relatively stiff, many are either light concrete structures with shear walls or braced steel-frame structures.
Therefore, the SQUG concluded that no large amplification of ground motion by the structure was expected for the locations of most of the equipment of interest.
Free-field ground spectra were used as conservative estimates of the floor-response spectra for the data base structures that were not analyzed.
Thus, amplification of the data base floor response spectra was conservatively excluded.
The floor response spectra required for the nuclear power plants were obtained from the operating utility. Wherever spectra were unavailable for a specific item, amplified floor spectra were assumed on the basis of nearby spectra.
The data base floor response spectra and the nuclear equipment required response spectra obtained as above are then compared to assure that floor response spectra of the data base envelope those of the nuclear equipment.,
l
TABLE 2.4-6 EQUIROIT SELECTION EQJIRENT SELECTED:
J ftTOR CONTROL CENTERS 480 VOLT SWITOtiEAR 2,4 To 4KV SWITC-iGEAR MOTOR-OPERATED VALVES 4
AIR-OPERATED VALVES HORIZONTAL RFPS VERTICAL PUMPS SEVEN NUCLEAR PCHER PLMTS VISITED - THREE SA %ttu FOR EQUIPMENT DATA COLLECTIONi
~
DESIGN EASIS PLANT SSE DRESDEN 3 0.2G CALVERT CLIFFS 1 0,15G PILGRIM 0,15G
- - =-
- =
j
ITEM:
480 Volt Hotor Control Center Cabinets 400 Volt Motor Control Center 39-3 IVA-6VA, P3A & P4A (Eight Units)
PLANT:
Sylmar Converter Station Dresden Nuclear Plant, Unit 3 i
HANUFACTURER:
General Electric 7700 Line Series, 1970 General Electric 7700 Line Series, 1971 LOCATION:
Sylmar Converter Station basement, facing Reactor Dutiding Clevation 570, facing northeast and southwest east (grade is at Elevation 517.5)
)
FUNCTION /SYSTEH: Control of pumps and valves for rectifier Control of various Class I mechanical cooling systems systems l<
CABINET:
Each cabinet is four cubicles wide; the Cabinet is six cubicles wide. The cabinet specific arrangement of starter units varies contains starter units in cubicles of i
from cabinet to cabinet; they are otherwise various sizes.
very similar.
1 m
COMPONENTS:
A typical starter unit consists of a General A typical starter unit consists of a Electric CR-106 magnetic contactor, a c1'r-General Electric CR-106 or CR-105 magnetic cult breaker switch, a control transformer, contactor, a circuit breaker switch, on-off pushbuttons and a terminal block.
a control transformer, on-off pushbuttons,'
and a terminal block.
l ANCHORAGE:
The bottom channe) of the cabinet is tack The bottom channel is tack welded to I
welded to a baseplate embedded in the con-an embedded baseplate, two welds at the crete floor. At least one cabinet was base of each stack of cubicles, front inadequately anchored at the time of the and back.
4 earthquake and slid a few inches.
I Ai PLICABLE The records taken at Pacolma Dam are shown The calculated floor spectra for the
RESPONSE
scaled to 40% of the measured amplitudes as Reactor Building, Elevation 589 are shown.
SPECTRA:
a conservative estimate of the ground mo-Spectra at Elevation 570 were not generated.
tion at Sylmar.
i EQUIPHENT The HCCs were in operation at the time of the earthquake. No damage to either cabloct i
STATUS DURING or components was reported. One cabinet slid a few inches due to lack of floor anchorage.
I' AND FOLLOWING Tile EARTHQUAKE:
Source: References 1, la, 20, 25, 26, 27, 30 and 32 (Appendix A).
i l.
Table 2.'4-7 Comparison of equipment data.
a
TABLE 2.4-8 SLft1ARY OF DATA BASE PL4fTS
& EARTliCUAKES ESTIMATED EARiliCUAlES FACILITY PGA Sm FERNmDO
- 1. SYLMR CONVERTER STATION 0.50 - 0.75 1971
- 2. VALLEY STEAM PLANT 0.l40
- 3. BUREMK PcwER PLANT 0.35
- 4. GLENDALE PcwsR PLm T 0.30
- 5. PASADENA PcwER PLANT 0.20
- 6. RINALDI RECEIVING 0.50 "
~
- 7. VINCENT SUBSTATION 0.20
- 8. SAUGUS SUBSTATICN.
0.39 P] INT MAGU
- 9. OROND BEACH PLmT 0.20 19/3
- 10. SmTA CLARA SUBSTATIcN 0.10 S#frA BARBARA
- 11. GOLETA SUBSTATIch 0.28 1978
- 12. ELLWOOD PEAIGR PLMT 0.30 - 0.l;0 IMRIAL VAf ' P(
13 & Ce<TRo STEM PLmT 0.51 1979
- 14. fvavmx GEon-wAL PLMT 0.20 - 0.30 l
EASED m:
%m=D PEAK GROUND Arm SMTIm - AT PLANT SITE brATED NEAR STRONG ltITtm RECmDS... _ -. -...
_ _ ~. _ _
The performance of data base equipment during past earthquake was evaluated and conclusions regarding the seismic resistance capability of similar nuclear equipment were reached. A typical comparison is shown in Table 2.4-7.
For the purpose of the pilot program non-nuclear power plants and other facilities in southern California where significant earthquakes have occurred was chosen for the study.
Table 2.4-8 shows the four earthquakes in southern California that were reviewed in detail in this program. -The facilities that contained the largest number of equipment items of interest and were reviewed in detail are the Sylmar converter station, Valley steam plant, Burbank power plant, Glendale power plant, Pasadena power plant, and El Centro steam plant.
Seven nuclear power plants were visited, and three were selected for equipment data collection, they are Dresden Unit 3, Calvert Cliffs Unit 1 and Pilgrim.
These plants were selected so that the equipment reviewed for the project would be a representative sample of a variety of nuclear plant characteristics, including reactor type and vintage.
Only equipment required for safe shutdown was considered.
2.4.3.2 Conclusion and NRC Staff Comments The goals of this pilot program were evaluated by the SQUG against the results obtained from the study. Table 2.4-9 listed the goals, findings and conclusion as seen by the SQUG.
Finally, the SQUG arrived at the following overall summary:
The SQUG has shown that the structural integrity of anchored power plant equipment and component is not compromised in strong earthquakes of up to 0.50G peak ground acceleration.
- The SQUG has shown that typically power plant equipment operability is not compromised in strong earthquakes with peak ground acceleration of about 0.20G to 0.30G.
1 While the staff is in general agreement with the SQUG on the first overall summary, the staff however has reservation on the second overall summary.
The NRC staff has completed the review of the pilot program report, and has concluded that it is feasible to accept experience data as a basis for seismic qualification. The staff realizes that the SQUG pilot program was intended to be a feasibility study and reviewed it in that context.
Therefore, the' comments are generally an assessment of what further work should be done to provide an acceptable experience data base.
The comments, which were sent to the SQUG in December 1982, were organized in eight. subject areas as follows:
i TABLE 2.4-9 MAJOR CONCLUSIONS OF SOUG GOAL 1 GOAL:
DEVELCF A HISTCRICAL DATA EASE CN THE PERFCFPNCE OF ECUIFfdEN IN CCNVEITICNAL POWER PL4(TS DURING AND AFTER STRCNG EARTHCUAKES.
FINDINGS:
SEVERAL PChER FLANTS AND OTHER INCUSTRIAL FACILITIES FAVE EXPERIENCED STRCNG EARTHGUAKES EXC:AING m: enz: e l:8 n SAFE-SHUTCChN EART.-iGUAKES RECUIRED FOR THE CESIGN OF MCST U.S.A. NUCLEAR ?LANTS.
THE FLMTS RESPCNDED WELL TO THE EARTHCUAKNS AND USUALLY CCNTINUED TO CFERATE CR hERE SACK CN LINE SHCRTLY Ari:rt TriE EARTriCUAKES.
lANiY CF THE FAC:LITIES hERE IN CFE:ATICN AT TriE TIME OF THE EART{UAKES; THUS TriEIR ECUIPENT WAS SUEJELitu TO NOF?AL CFECATING LCADS IN ADDITION TO THE SEISMIC LCADS FRCM TriE EARTHCUAKES.
WITH A FEW MINCR EXCE:TICNS, TriE ECUI?tdSiT CCNTAINED. IN TriE FChER FACILITIES WAS. UNDMAGED AND WAS PJNCTICNAL Ah cx WE EARTHCUAKES.
IHE ECUIPMENT WAS NOT KNCWN TO EE t<DIFIED BECAUSE CF TriE EARTriCUAKES.
~
$UFFICIENT CATA EXIST TO ESTIt' ATE TFE SPEtieA EXFERIENCED BY TriE FL4GS MD TdIR E UIRiEfT.
TERE IS A LARGE, AVAILG.E CATA 3ASE, CNLY A FCRTION CF WICH HAS SW A IN TrUS STUDY, CF PChER FWT EQJIPMENT TPAT PAS EEEN SUE.;E ico TO STT<NG EARTHGUAKES.
j CCNCLUSICN i
IHERE IS A LARGE SC0Y CF AVAILAELE IATA CN TriE PERFCWANCE CF FCHER FLANT EQUIR1NT IN STRCNG EARTHCUAKES, INCLUDING SOTH l
tEC'ANICAL MiD ELMICAL EQUIRdENT. IdANY CCNVENTIO W. PCNER R.NES AND INEUSTRIAL FACILITIES FAVE EXPERI2 ICED E4iThCUAfSS THAT SU&JECTED TdIR EQJIFtENT TO SEISMIC EW.IROeENTS E2.9L TO CR EXCEEDING SEIS4IC LCADS ASSCCIAicu WITH SA-' 3UTICNN EARIHQUAJES REQJIRED FOR TE DESIG4 0F MCST WCLEAR FCAR R.AKTS..
w-y.-
me-
TABLE 2.4-9 (CONTINUEJ)
GOAL 2 GDAL:
SHOW Tr%T MUCH OF TriE ECUI? MENT ITNESTIGAi 9, hHICH HAS EXPERIENCED STRCNG EARTrCUAKES, IS SIMILAR TO EQUIPMS6 FCUtG IN NUCLEAR PCWER FLAT 65.
FINDINGS:
A Frrf MAJCR ECUIPMENT MMNUFACTURERS SUPPLY M)CH CF TriE EGUIRENT FCR 50Tri CCfNENTICNAL AND i:UCLEAR PChER FLNES.
IHEPE IS LITTLE CESERVAELE DIFFERENCE BETh~cEN TriE EASURED DYNAMIC RESFCNSE FRECUENCIES OF ECUIP"Eli IN NUCLEAR PO69 FLN65 MO THOSE IN CCiNENTICNAL PLAl65.
IHERE ARE NO GENERIC DIFFERENCES OTr=~R TrW1 AGE ScMN EGJIPMENT FCOO IN CChvF.liTICNAL AND NUCLEAR PCWER PLR65.
CCNCLUSIONS:
CERTAIN T(FES C.: MEGANICAL MO ELECTRICAL ECUIFENT FCUND IN NUCLEAR FChER FLNiTS ARE VERY SIMILAR IN CCNFIGURATICN, FLNCTICN, MANUFACTURER, AND MCCEL TO TriE TYPES FCUte IN
\\
CCNVENTICNAL PLANTS. FUCH OF THE EQJIPMENT IN NLCLEAR PCWER
?tRf75 MO CCtNENTICNAL PCbER PLANTS IS TriE SAME.
O e
W 9 -.
TABLE 2.4-9 (CONTINUED)
COL 3 GCAL:
CETERitNE '+: ETHER ACTUAL EARTHGUAKE DATA ARE SUF:ICIENT TO CCNCLLE THAT SEISMIC GUALIFICATION OF CERTAIN CLASSES CF EQUIPMENT BY CONENTIONAL METrCDS IS NOT NECESSARY.
FINDitES:
EXCLUDING SCME UNANCHCRED ECUIPMENT AND CNE AIR-CPERAscu VALVE, NO FAILURES WERE REFCRico IN #1Y OF THE SEEN TYPES CF ECUIPMENT ADDRESSED IN THIS STUDY.
WITH THE FOSSIELE EXCc. i10N OF ELECTRICAL R: AYS, THERE J
IS NO EVIDENCE OF MAL:UNCTICN CF THE REVIEnED EGUIP"ENT DURING THE EARTHCUAKES.
IHE ESTIMAsco GRCCPO-RESPCNSE SPECTRA FRO 4 SEVF.5AL CALIFCRNIA EARTriCUAKES #tD TriE CCtrcNTICNAL PCHER FLMTS Arec.i=.o BY TriEM ENVELCFE THE FLCCR-RESPONSE SPECTRA FCR l
THE SAFE-Sm' EARTrGUAXES REQUIRED FCR NUCLEAR PChER PUNTS IN TIE PJNGES CF MCST EQJIPMENT RESFCNSE FREQ.ENCIES.
CCtWENTIOut PL#1TS Tr%T 'eERE SUSJEt.:cu TO :JRTHCUAKES l
WITH PEAK GRCUND ACCELEPATICf4 CF ABCUT 0,3CG CR LCWE=,
GENEPALLY CCNTINUED TO CPERAic TriRCUG'-iCUT THE EARTHQUAKES.
C[NCLUSICN:
l SEISMIC CUALIFICATICN C<F tu.CL:JR EGUIFMENT EY CCNVENTICNAL METriCDS DCES NOT APPEAR _TO EE.NECESSARY FCR TriE CUSSES OF EGUIPMENT EVALUAicu FCR MCST LEWLS OF SAFE-5-3JTT4W4 EARTHCUAKES.
60 -
-,,m..
-,--__y 7,
-,e--g.. -.
,,,.p,-
.y
--w,---
,y,_y-g--
w,-
,i
-.-.a
,m--
3.-w.-
,7 y
TABLE 2.4-9 (CONTINUED)
GOAL u
~
GOAL:
DEVELCF A PETriODOLOGY FCR THE USE OF ACTUAL EMTriCUAKE DATA TO Cti FMINE bhETHER SEISMIC CUALIFICATION CF SPECIFIC ITEMS OF ECUIPPENT BY CONVENTICNAL tETKDS IS tlECESSARY.
FINDINGS:
IHE SEISMIC FERFCrTANCE OF THE REVIEWED ECUIFFEN~r AFFEARS TO EE INDE?ENCEHT FRCM #1Y OF THE FOLLCWING FACTCRS:
AGE OF ECUIFPENT YEARS cF SERVICE 1"ANUFACTURER AND tCCEL FCUNTING CCNFIGUPATI'CN DYAAMIC PROPERTIES IHE PETHODCLOGY USED IN THE ?! LOT FRCGRAM TO EVALUATE CLASSES OF EQUIFtfNT WCULD EE EGUALLY AFFLICABLE TO SPECIFIC ITEMS OF EQUIWENT.
CCNCLUSICN:
THE FItoT F.RCGRM HAS DENCNSTRAieu THE PETHCDCLOGY, THERE IS
- 1 ASUNDANCE CF DATA THAT CAN EE USED TO IDENTIFY SPECIFIC ITEMS OF ECUIWENT THAT CO NOT RECUIRE ACDITICNAL SEISMIC CUALIFICATICN, l
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(1) Extent and Organization of the Experimental Data Base t
Five of the six data base plants included in the pilot study were
' subjected to the 1971 San Fernando earthquake.
The sixth plant, the El 4
Centro Steam Plant, was subjected to the 1979 Imperial Valley event which was of relatively short duration and originated from a source very close to the site.
It is widely recognized that certain earthquakes are j
more damaging to certain types of structures and that duration time and travel path are important parameters.
The staff believes that the data base should be extended to include additional events. The additional events should include to the extent possible, a wide range of durations, amplitudes, and frequency contents.
The data base should be organized such that all equipment in a "similar" group is catalogued along with a record of the individual event time histories and a corresponding response spectra which can be applied to members of this group.
A minimum of three separate and distinct earthquake histories should be used to construct a data base spectrum.
i (2) Adequacy of Seismic Input For the two earthquakes considered in the pilot program, characterization of the shaking motion of the non-nuclear sites is adequate. The staff believes it is reasonable to assume that the seismic environment experienced by equipment located on the first floor or in the basement can be characterized by the estimated free-field motion at the building foundation, e.g., equipment located at-higher levels in the plant can be assumed to have been subjected to the estimated free-field motion at the foundation except.in cases where sharp, short duration peaks in the time history may have been filtered and " damped out" by SSI effects as may be the case for the El Centro steam plant.
If an acceptable dynamic analysis 'is performed on a data l
base plant to calculate response spectra at equipment attachment points, then these amplified floor spectra may be used as the experience data input.
When comparing experience data. spectra to the required response-spectra l
at a nuclear plant, the nuclear plant seismic analysis and development j
j of floor spectra must meet current regulatory requirements properly-accounting for.the hazard and the spectrum shape or be approved-by the i
i staff for use-in qualifying equipment on some other defined basis.
Recent information indicates that eastern earthquakes may exhibit significant frequency' content above 15 hz. --The staff is continuing to evaluate this phenomenon.
In the event the staff determines that I
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a seismic design criteria should be modified, to account for the high frequencies, a change in criteria would be proposed.
(3) Adequacy of Equipment Anchorages Equipment qualification by comparison with experience data assumes the adequacy of equipment anchorages.
Deficiencies in the anchorages of electrical equipment were identified in the SEP phase II review and resulted in the issuance of IE Information Notice 80-21.
The staff believes that, regardless of the qualification procedure used, equipme7t anchorage is a key element and adequate anchorage must be proven.
This means adequate strength and no separation of anchor points during loading.
(4) Definition of Similarity An acceptable definition of similarity needs to be developed. The staff believes that for equipment to be similar for the purpose of qualifying an equipment item on the basis of experience data on another item, the safety function as well as the dynamic characteristic, should be similar. This means that the experience data must include data on performance both during and after a seismic event.
Similarity parameters must include mass distribution, material, size, stiffness, configuration, restraints, and anchorage details.
For active mechanical equipment, such as pumps and valves, where long-term operability is needed after an earthquake, the possibility of incipient damage to similar equipment in data base plants should be investigated.
This could be done by reviewing plant maintenance seco ds supplemented by detailed inspection of the in-situ data base compotents as required.
(5) In-Situ Testing Comparisons of frequencies between non-nuclear and nuclear equipment in the pilot program were based on low-level in-situ testing.
The staff believes that verification of the validity of low-level testing is required to show that there are not significant differences in response between low level tests and excitation levels associated with significant earthquakes.
The staff is concerned that differences in anchorages may significantly affect equipment responses. Verification of anchorage integrity and similarity.must be addressed. Anchorages are addressed further in item 3 above.
(6) Equipment Operability The primary design criteria of nuclear plant safety equipment during seismic events are (1) to operate on demand and (2) to have no spurious status change. The experience data implies that electrical switchgear and motor control center equipment loacted in non nuclear plants opened during actual seismic events and required subsequent operator action to r
e be re-closed.
EQE stated during a briefing to the SQUG and the NRC staff that data base plant equipment actuated spuriously during the seismic event. The data base should be more explicit regarding equipment operation during a seismic event, i.e., did it operate as intended or did it actuate spuriously?
To use the experience data as a basis for qualification of electrical equipment in nuclear plants, the required functions must be compared and/or detailed information on the state of the data base equipment must be known for comparison with the required function of the nuclear plant equipment.
Even though many of the electrical equipment components in fossile fuel plants are similar to those in nuclear power plants, safety systems in nuclear plants are considerably more complex and less accessible due to radiation hazard. Therefore, the reset of circuit breakers, and relays, and the sequencing of 6he safety systems may not be possible in a timely fashion.
Operability following an earthquake should also be considered.
For instance, equipment in the residual heat removal system must operate on a long-term basis following an earthquake event:
the data bTse should include records which verify that equipment that appeared to be functional immediately following the earthquake was not partially damaged or approaching a " damage" threshold.
(7) Degradation of Seismic Performance Due to Aging Electrical components that include age-sensitivie elements, such as motor insulation, are likely to degrade and influence seismic performance. The conclusion drawn in the EQE report that there is no observed degradation of seismic performance related to aging is based on limited data. More definitive data on aging, including available test results, should be reviewed to verify the conclusion.
(8) Margin and Fragility The question of seismic margin has been of concern for many years.
Qualitatively it has been answered by acknowledging that some margin exists in the design procedure, and by the use of an enveloping response spectrum in addition to the " inherent strength" of structures and equipment. The fragility of equipment cannot be determined unless the failure level is known.
Very little actual fragility data exists. Most fragility curves are based on estimated failures levels either~from consensus of expert opinion or by extrapolation of test data.
Even through some failure tests have been conducted, the amount of test data available is too sparse to be statistically significant.
was to develop a set of " generic floor response spectra" which can be utilized by the utilities for the purpose of qualifying equipment.
The task of developing generic floor response spectra was undertaken by Brookhaven National Laboratory (BNL).
The task now is essentially complete. A draft report (Ref. 14) was issued in April 1983.
Following is a summary of this contractor report.
2.5.2 Summary of Work Completed The development of a generic floor response spectra starts with the concept that there is a degre* of boundedness to the structural responses. The report follows this concept and shows that the response can be bounded within a useful range.
The general approach was to study the effects on the dynamic characteristics of each of the elements in the chain of events that goes between the applied loads and the responses. This includes the seismic loads, the soils and the structures.
Two actual structural models, one BWR and one PWR, were used in the study.
For the BWR model (Model 3), a Mark I containment structure is modeled as a single stick, as shown in Figure 2.5-1.
"or the PWR model (Model 4), the system is modeled as three separate structures on a common foundation.
Three stick models are used to represent the shield structure, the steel containment and the internal structure. Figure 2.5-2 shows this PWR model.
A free-field earthquake response spectra from the El Centro earthquake was used to generate horizontal earthquake time histories.
Vertical spectra were not developed in this program.
The peak acceleration of this input time history was scaled to a 1 g level as a normalization procedure to study the response.
In reporting the proposed generic response spectra, the peak values were normalized to a more realistic time history peak of 0.1 g.
The excitation was applied through the soil and into the various structures to produce responses in equipment at each level. An entire range of soil conditions was used with -ach structure, from soft soil (with a shear wave velocity of 800 ft/sec.) to solid rock (shear wave velocity of infinity) in seven steps.
For both the BWR and PWR models, stiffness properties were varied, with the same mass, to extend the fundamental base structure natural frequency from 2 cps to 36 cps.
This resulted in fundamental mode coupled natural frequencies as low as 0.86 cps and as high as 30 cps.
From all of these models of soils and structures, floor response spectra were generated at each floor level.
The proposed spectra were reported for the top level of a generic structure, based on an earthquake time history with a peak acceleration of 0.1 g.
Reduction factors are applied to the peak accelerations to account for the site specific time history maximum acceleration.
A second factor was obtained which recognizes a reduced level of acceleration for equipment located at lower elevations..
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Figure 2.5-3 is the maximum generic floor response spectra which were deduced from this study.
The curves apply to the top of the structure, which is the point of maximum acceleration.
They were normalized from an earthquake time history with a peak acceleration of 0.1 g.
These spectra are for five different classes of soils (shear wave velocity from 800 ft/sec to infinity). As shown in the Figure, curves A through E are associated with interaction frequencies (a natural frequency calculation obtained by taking the square root of the ratio of soil stiffness to an equivalent mass of the soil and structure) of 2 cps through greater than 50 cps, or from soft soil through solid rock, respectively.
Figure 2.5-4 shows the reduced peak acceleration values that apply to the accelerations in the response spectra at different floor levels.
This figure corresponds to soil condition of solid rock (Case E) which has a maximum peak acceleration of 7.2g at the top level for a 0.lg earthquake. The peak was calculated to be 6.0g for 0.lg earthquake.
This was increased by 20 percent to 7.2g because only one earthquake time history was used for the horizontal specta.
As the shear wave velocity of the soil decreases (softer soil), the maximum floor response acceleration decreases. The peak acceleration at the top level of a structure on soft soil was taken to be 5.0 g.
This is 30 percent less than the peak floor response acceleration of 7.2 g at the same elevation for a solid rock soil.
In summary, this report established a procedure for the generation of the horizontal generic floor response spectra to any operating plant.
The procedure allows a utility to use as much or as little information as is available.
The conservatisms of the spectra generated increases if little seismic data, is available. Generic spectra in the vertical direction were not developed in this program.
Due to the conservatism accumulated by this approach every step along the way, the NRC staff believes that a conservative vertical generic floor spectra can be reasonably estimated by taking two thirds of the values of generic floor spectra in the horizontal direction.
2.5.3 Staff Conclusions A Required Response Spectra (RRS) is needed whether analysis, test or experience data is used for the qualification.
If equipment is attached to the floor, the floor response spectra will be the RRS.
If equipment is attached to a supporting structure, the RRS at the equipment attachment point can be generated by a variety of ways (see Section 2.3) from the floor respcnse spectra. _
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By using the methodology described in this section, the floor response spectra can conceivably be generated with reasonable conservatism without having to go through the rigorous time history and finite element analyses normally required.
However, the staff ~ believes that this approach will have its limitations, and these limitations should be spelled out clearly.
3.
REFERENCES
~
1.
" Correlation of Seismic Experience Data in Non-Nuclear Facilities with-
~
Seismic Equipment Qualification in Nuclear Plants (A-46)," by P. D. Smith and R. G. Dong.
LLNL draft report, November 1982.
2.
" Program for the Development of an Alternative Approach to Seismic Equipment Qualification," by P. I. Yanev, S. W. Swan, EQE report (2 Volumes) prepared for SQUG, September 1982.
~
3.
" Identification of Seismically Risk Sensitive Systems and Components in-Nuclear Power Plants," by A. Azarm, J. Boccio, P. Farahzard, BNL draft report, November 1982 (revised April 1983).
4.
" Correlation of Methodologies for Seismic Qualification Tests of Nuclear.
Plant Equipment," by D. D. Kana and D. J. Pomerening, SWRI report'no.
SWRI-6582-002, Contract NRC-04-81-185, Task 2 Summary Report, June 1, 1983.
5.
"The Use of In-Situ Procedures for Seismic Qualification in Currently Operating Plants," by S. Sadik and B. W. Dixon. INEL interim report, -
December 1982.
~
6.
" Preliminary Guidance and Acceptance Criteria for Application of Combined In-Situ and Analysis Procedures in Operating Plant Equipment Qualification," INEL preliminary report, April 1983.
7.
" Summary of Work Performed to Date on Qualification Cost Estimating Task," INEL preliminary report, April 1983.
8.
U. S. Nuclear Regulatory Commission Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants." October 1973.
9.
U. S. Nuclear Regulatory Commission Regulatory Guide 1.122, " Development, of Floor Design Respense Spectra for Seismic Design of Floor-Supported Equipment or Components," Revision 1, 1978.
10.
" Survey of Methods for Seismic Qualification of Nuclear Plant Equipment r.nd Components," by D. D. Kana, J. C. Simonis, D. J. Pomerening.
SWRI report no. SWRI-6582-001-01, Contract NRC-04-81-185, Task 1 Summary Report Part I, October 1982.
~..
Y
-<,1.-
11.
" Seismic Safety Margins Research Program, Phase I Final Report-SMACS,"
by J. J. Johnson, G. L. Goudreau, S. E. Bumpus, O. R. Maslenikov, LLNL report, July 1981.
12.
" Operating Function Tests of the PWR Type RHR Pump for Engineered Safety System Under Simulated Strong Ground Excitation," by T. Uga, K. Shiraki, T. Homma, H. Inazuka, N. Nakagima, Japan Atomic Energy Research Institute, JAERI-M8354, August 1979.
3, 13.
" Seismic and Dynamic Qualification of Safety Related Electrical and Mechanical Equipment in Operating Nuclear Power Plants," by J. Curreri, C. Costantino, M. Reich, BNL draft report, April 1983.
14.
"A Study of the Effect of Aging on the Operating of Switching Devices,"
by S. P. Carfagno, G. H. Herberlein, Jr., IEEE Transactions on Power Apparatus and Systems, Volume PAS-99, No. 6, November / December 1980.
i T
APPENDIX A TASK ACTION PLAN SEISMIC QUALIFICATION OF EQUIPf1ENT IN OPERATING PLANTS (TASK A-46)
Division of Safety Technology (DST)
Generic Issues Branch (GIB)
Task Manager:
T. Y. Chang, GIB Lead Supervisor:
Karl Kniel, Chief, GIB, DST NRR Principal Reviewer:
David Reiff
'quipment Qualification Branch Division of Engineering John Knox Systematic Evaluation Program Branch Division of Licensing Pef-Ying Chen Systematic Evaluation Program Branch Division of Licensing Frank Skopec Radiological Assessment Branch Division of Systems Integration
' Applicability:
All Light Water Operating Reactors Projected Completion Date:
December 1984 l
A/1
l 1.
DESCRIPTION OF PROBLEM There is a recognized need to demonstrate the functional capability of safety-related nuclear plant equipment subjected to a seismic event. The General Design Criteria (GDC) for Nuclear Power Plants states that structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes, with ut a loss of capability to perform their safety functions (10 CFR Part 50, Appendix A, Criterion 2). Also the GDC states design control measures shall provide for verifying or checking the adequacy of design by the performance of a suitable testing program. Suitable qualification testing under the most adverse design conditions shall be included (10 CFR Part 50, Appendix B, Section III), Guidance on compliance with these provisions of 10 CFR Part 50 is contained in Revision 2 to Standard Review Plan Section 3.10. " Seismic and Dynamic Qualification of Mechanical and Electrical Equipment."
Today, equipment is seismically qualified by analysis and/or testing. Analyses alone are acceptable only if the necessary functional operability of the equipment is assured by its structural integrity alone.
If not, some testing is required.
Seismic input motion to equipment is specified by a required response spectrum or time history.
Equipment that is small enough is subjected to a test response spectrum which envelopes the required response spectrum.
The equipment should be tested in the operating condition.
For equipment too large to fit on a test table, a combined analysis and test procedure is adopted.
Since corrinercial nuclear power plants were first introduced, significant changes in seismic qualification criteria have evolved.
Also, the analytical and experimental methods used to qualify equipment have changed. Therefore, the seismic resistance of existing equipment installed in operating nuclear plants may vary considerably.
Operating plant equipment may not meet the current seismic qualification criteria. The seismic qualification of equipment in operating plants may have " be reassessed to ensure its performance during and after a seismic ' nt.
The objective of this Unresolved Safety Issue (USI) is to develop seismic qualification methods and acceptance criteria that can be used to assess the capability of mechanical and electrical equipment in onerating nuclear power plants to perform their intended safety function during and after a seismic event.
Technical work in support of USI A-46 will be provided by technical assistance contracts managed by the Office of Nuclear Reactor Regulation (NRR), from ongoing research programs in the seismic area, and possibly from the Systematic Evaluation Program (SEP).
A/2
The Equipment Qualification Branch (EQB) of NRR is supporting a program which includes (1) a risk sensitivity study of safety system components, which will form the basis for development of a minimum equipment list; (2) cost / benefit analyses of seismic qualification of equipment on the minimum equipment list; and (3) development of guidelines for generation of generic floor response spectra.
The Office of Nuclear Regulatory Research (RES) is supporting a research program which, in part, is an historical survey of methods used for seismic qualification of nuclear plant equipment and components and a comparison with current criteria.
The Generic Issues Branch (GIB) of NRR is supporting a program for correlation of seismic response of equipment in non-nuclear facil-ities to the qualification of nuclear plant equipment, and a program for the development of insitu test methods and the collection and correlation of test data from both laboratory tests and insitu tests.
2.
PLAN FOR PROBLEM RESOLUTION A.
Approach A minimum list of equipment to be qualified will be developed from a risk study conducted under contract to Brookhaven National Laboratory (BNL). The reliability of components in systems which perform important safety functions will be varied and the effect on risk computed. A sensitivity analysis will be used to identify equipment where changes in reliability result in large incremental risks, allowing a cost / benefit analysis to be made.
Only those components whose failure significantly affects safety functions will be included on the minimum equipment list.
Mechanical and electrical components on the minimum equipment list will still be too numerous to consider on an individual basis.
Generic groups of these components will be developed according to function and simila;ity of methods to be used for seismic qualifi-cation.
A review of past and present criteria and methods used to i
structurally and operationally qualify the various categories of I
equipment is being conducted. Both analytical and test methods are being considered. The conservatisms, disavantages, deficiencies, and anomalies of the methods will be determined. This review is part of a research program sponsored by RES and being performed at the Southwest Research Institute (SWRI). Activities of this research program in support of A-46 are an evaluation of past and present analytical and test methods of seismically qualifying operability of safety-related equipment and correlation of these methods with current criteria.
A/3
^
~
l The SEP of the Systematic Evaluation Program Branch (SEPB) complements this USI program.
In Phase I of SEP Topic III-6, a sampling of existing seismic design documents from five older plants was reviewed, and a limited amount of reevaluation was also made.
Some structrual retrofitting to ensure proper equipment anchoring was recommended for the five plants. Safety-related systems and components were reviewed in selected plants to some extent for operability. Some systems and components were found to require additional seismic evaluation. SEP plant owners have initiated a generic program to tabulate the equipment present in the SEP plants.
If appropriate information can be developed in time, it will be reviewed and incorporated for consideration in developing USI A-46 resolution.
An effort has been initiated by the Seismic Qualification Utilities Gr up (SQUG) to survey mechanical and electrical equipment installed in non-nuclear plants built in high seismic areas. Non-nuclear power plants and many industrial facilities contain mechanical and electrical equipment similar in design and function to equipment used in nuclear power plants. A number of these non-nuclear power plants and industrial facilities have been subjected to seismic events. Experience with equipment in thue plants and facilities can be useful in determining the seismic and dynamic response of comparable equipment in nuclear power plants. One task of USI A-46 is to monitor that survey, and if it is determined that the resulting information is useful, it will be integrated into the development of seismic qualification guidelines.
A program has been initiated for development of insitu test methods to assist in qualifying equipment in operating plants.
In that program, a review and sumary of existing c:etjods for performing insitu testing will be made.
In addition, operability and failure for various types of equipment will be defined, and a data base of laboratory test and insitu test information will be developed.
Information on insitu and laboratory tests will be used in development of guidelines for qualification of equipment in operating plants.
Following completion of A-46 technical worr. and development of proposed regulatory requirements, a value/ impact analysis will be performed to determine the cost effectiveness of implementing the proposed requirements. The A-46 technical findings, the implementation recomendations and the value/ impact analysis will be submitted for review by the Comittee to Review Generic Requirements 4
(CRGR). Following CRGR review, the proposed requirements with supporting documents will be issued for public comment prior to final approval by the NRC staff and CRGR.
A/4
i l
r B.
Tasks Task 1.
Develop Minimum List of Equipment to be Qualified It must be ensured that (1) modification of safety-related equipment provides substantial additional protection which is required for the public health and safety, and (2) equipment considered for upgrading be those that contribute most to risk.
Subtask 1(a).
Perform Sensitivity Analysis Using a list of systems essential to reactor shutdown and prevention cf radioactive release, a sensitivity analysis will be performed using previously developed computer codes. The result is expected to be a list of equipment whose changes in reliability result in large effects on public risk.
Subtask 1(b).
Perform Cost / Benefit Analysis of Seismic Upgrading of Equipment Using the list of equipment developed in Subtask 1(a), cost will be estimated to upgrade the equipment. Benefit to the public will be estimated.
Task 2.
Survey and Evaluation of Equipment Seismic Qualification Methods This task involves a study sponsored by RES to evaluate past and present methods to qualify mechanical and electrical equipment to withstand seismic events. The structural adequacy of equipment subjected to seismic events is also being reviewed by SEPB.
If this information can be developed by SEPB in time, it will be reviewed and incorporated into this task.
Subtask 2(a).
Evaluation of Methods Used to Seismically Qualify Equipment Past and current analytical and test methods used to qualify equipment will be cataloged, compared and evaluated.
The contractor's developed equipment list will be used in this subtask.
Subtask 2(b). Comparison to Present NRC Requirements for Equipment Qualification Methods to qualify equipment in operating plants will be compared to present requirements.
Important differences will be determined amd acceptability of qualification method will be recommended.
A/E
i Task 3.
Develop Methods of InSitu Testing to Assist in Qualification of Equipment This task will involve surveying existing methods for performing insitu tests which may be used to assist in qualification of nuclear plant equipment. Also, analytical methods which would be used in conjunction with those insitu tests will be reviewed and summarized.
The effects of component aging will be considered.
The final part of this task will be an effort to improve insitu testing for use in seismic equipment qualification.
Subtask 3(a). Develop Preliminary InSitu Test Methods Operability and failure of various types of equipment will be defined in the first part of this subtask.
Existing methods for performing insitu tests will be surveyed.
Equipment will be categorized according to which test procedures are appropriate.
Limitations, shortcomings and nonconservatisms associated with the methods will be identified.
Subtask 3(b).
Improve and Verify InSitu Methods The limitations identified in Subtask 3(a) previously will be studied and recommendations made for improvement and verification of test methods.
Subtask 3(c). Develop Requirement and Acceptance Criteria for Insitu Test Method Requirements and acceptance criteria for using insitu testing method in conjunction with experience data to qualify equipment will be developed.
Subtask 3(d). Prepare Program Report A formal report, in NUREG format, will be written covering the results of Task 3.
Task 4.
The Seismic Qualification of Equipment Using Non-Nuclear Plant Inservice Dynamic Response Information A program has been developed by the SQUG to survey equipment in non-nuclear plants which has been subjected to seismic events. The eouipment to be surveyed is similar to equipment installed in operating nuclear plants. The seismic events which the equipment survived were, in some instances, significant. The SQUG program will be closely monitored as part of this task and the results will be studied for possible use in development of qualification require-ments. Other sources of information pertinent to. response, damage and operability of equipment in non-nuclear facilities subjected to l
seismic events will be reviewed to determine if non-nuclear equip-i ment experience can be used to predict equipment fragilities.
If it A/6
is possible to predict equipment fragilities from non-nuclear equipment surveys, then methods will be developed for the use of seismic experience in non-nuclear facilities in developing guide-lines for equipment qualification in nuclear plants.
Subtask 4(a).
Feasibility Study To assess the feasibility of using data on equipment from non-nuclear plants which have been subjected to strong earthquakes, a significant amount of data will be assembled from known sources and from the SQUG program.
It will be determined if a correlation exists or can be developed between structural and functional survival of equipment in non-nuclear plants and nuclear plants. To assist in assessing the feasibility, expert consultants will be provided by the contractor to review subtask results.
4 Subtask 4(b). Develop Guidelines for Application of Experience Data Guidelines for the use of the experience data collected previously will be developed and recommendations will be made for criteria to be incorporated into the proposed guidelines on equipment qualification.
Task 5.
Guidelines and Criteria for Development of Generic Floor Response Spectra The feasibility of seismically qualifying equipment using a set of generic floor response spectra will be investigated in this task.
Guidelines for developing these response spectra will be developed.
Subtask 5(a). Feasibility' Investigation The feasibility of seismically qualifying equipment by using a set of generic floor response spectra will be investigated. These response spectra will be derived by considering specific earthquakes zones in accordance with Uniform Building Codes, specific site geological conditions, specific plant installation configurations, or a combination of all of the above.
Subtask 5(b). Recommend Guidelines and procedures to Develop Generic Floor Response Spectra Guidelines and procedures to develop generic floor response spectra will be recommended.
A/7
~
Task 6.
Establish Guidelines, Alternative Methods, and Acceptance Criteria for Seismic Qualification of Equipment in Operating Plants Subtask 6(a). Develop Guidelines for Assessing Adequacy of Existing Seismic Qualification, Define Alternative Methods and Acceptance Criteria l
From the conclusions reached during the continue ~d performance of research programs on equipment qualification, the SEP on seismic qualification and this task action plan, a set of explicit guide-lines will be written to assess the adequacy of equiprrent seismic qualification methods. Both structural and functional qualification requirements will be considered.
If previously used qualification metiiods are found to be inadequate, alternative methods for requalifying equipment will be defined. Acceptance criteria for the alternative methods will also be developed.
Subtask 6(b). NUREG Final Report In this task a final NUREG report will be written to summarize program' accomplishment, conclusions, and recommendations. The justification for each guideline will be stated and limitations will be given. The NUREG report will be issued for public comment prior to final issuance.
Subtask 6(c). Licensing Changes In addition to providing technical bases for the reccmmended guidelines and criteria, proposed changes to Standard Review Plans and/or Regulatory Guides or issuance of a generic letter will be recommended if needed, and issued for public comment prior to implementation.
C.
END PRODUCT On Task 6 of this study, proposed guidelines _ and criteria for requalification of equipment in operating plants will be developed.
A NUREG report will be written summarizing the work performed, the conclusions reached, and recommendations regarding methods of requalifying equipment.. Guidelines for the qualification of equip-ment in operating plants will be presented in detail. Also the logic behind these guidelines will be given.
If a generic letter or changes to Standard Review Plans and/or Regulatory Guides are needed they will be prepared.
A value/ impact analysis will be conducted and submitted to the CRGR along with the NUREG report and proposed regulatory requirement documents. Following CRGR review and public comment, the final regulatory position on implementation will be developed and again reviewed by CRGR prior to final issuance.
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3.
JUSTIFICATION FOR CONTINUED OPERATION Although many operating plants were designed before the development of current licensing criteria, the design rules and procedures incorporated inherent conservatisms. These include:
(1) the margins between allowable stresses and ultimate strength of engineering materials; (2) the methods used for combining loads; (3) the inherent ductility of materials; and (4) the seismic resist-ance of nonstructural elements which are not normally considered in design calculations.
An expanding data base of observations at large industrial facilities that have experienced strong ground motion suggests that these facilities have significant seismic resistance capabilities.
From that data, it can be inferred that the inherent seismic resistance of engineered structures and equipment is usually greater than is generally assumed. When even the most modest attention is paid in design to providing lateral load carrying paths, significant capability results. Most nuclear power plants have been designed using more rigorous techniques; therefore, it is reasonable to expect high inherent margins.
Furthermore, Office of Inspection and Enforcement Information Notice
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No. 80-21, entitled, " Anchorage and Support of Safety-Related Electrical Equipment" was sent to all operating plants from~ the NRC on May 16, 1980. This Information Notice informed licensees of potential safety deficiencies in the design of safety-related electrical equipment supports in the SEP plants.
They were requested to review the information for possible applicability to their facilities.
Because of the above cited reasons and the continued staff review of seismic issues, it is concluded that operating plants can continue to operate without endangering the health and safety of the public pending resolution of this USI.
4.
PROGRAM SCHEDULE AND EFFORT The following schedule has been established for the tasks.
Completion Date Task 1 6/83 Task 2 8/83 Task 3 9/83 Task 4 12/83 Task 5 6/83 Task 6 4/84 A/9
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The level of f;RC effort to complete A-46 is summarized below in staff years:
FY82 FY83 FY84 GIB/ DST 1.0 1.0 1.0 MSEB/RES 1.0 0.1 0.1 EQB/DE 1.0 0.3 0.5 SEPB/DL 0.1 0.1 ORAB/DL 0.1 0.1 PSB/OSI 0.05 0.05 RAB/DSI 0.05 0.05 5.
TECHNICAL ORGANIZATIONS INVOLVED A.
Generic Issues Branch, Division of Safety Technology The Generic Isuses Branch has the overall responsibility for the performance of this USI program.
(1) Task 3 The GIB will establish a plan evaluating methods for insitu and laboratory qualification of equipment in operating plants.
This will be done through a technical assistance program with Idaho National Engineering Laboratory (INEL) to study methods of requalifying equipment installed in operating plants. The GIB will manage the performance of this technical assistance program and the publication of a final study report.
(2) Task 4 The GIB will develop a program plan to review and correlate available information on the inservice response of non-nuclear plant equipment that has been subjected to seismic or severe dynamic events.
This will be accomplished by close cooperation between an in-place SQUG program which is collecting data on equipment in non-nuclear plants which have been subjected to earthquakes, and the technical assistance program with Lawrence Livermore National Laboratory (LLNL) and INEL. The GIB will coordinate these programs and manage the performance ~of the technical assistance program.
(3) Task 6 The GIB, in conjunction with EQB, will establish the appropriate guidelines and acceptance criteria for the seismic qualification of equipment in operating plants.
The technical bases of these guidelines will be documented in a final NUREG report. This report A/10
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will also summarize the work performed in this USI program and the conclu;!;
reached.
B.
Mechanical and Structural Engineering Branch, Division of Engineering Technology, Office of Nuclear Regulatory Research (1) Task 2 The Mechanical and Structural Engineering Branch has a contract entitled, " Seismic Qualification of Nuclear Plant Mechanical and Electrical Equipment," with SWRI.
This research program will be
)
coordinated with the USI program.
The research program will survey existing knowledge and develop a basis and the methodology for evaluating conservatisms, limitations, and anomalies related to current and past methods used to qualify equipment.
C.
Systematic Evaluation Program Branch, Division of Licensing (1) Task 2 The Systematic Evaluation Program Branch (SEPB) is conducting a program to review and evaluate the seismic design criteria and the ability of safety related mechanical and electrical equipment to retain their structural integrity during and after a seismic event.
The functional operability of the equipment is not being considered.
This SEP branch program will complement the USI study.
If appropriate information can be generated in time, it will be integrated into the USI program.
D.
Equipment Qualification Branch, Division of Engineering (1) Task 1 EQB is developing a program to (a) identify equipment that contributes most to risk during and after a seismic event, and (b) perform a cost / benefit analysis to establish the extent to which safety-related equipment needs to be upgraded.
This will be accomplished by a technical assistance program with BNL.
(2) Task 5 EQB is developing a program to investigate the feasibility of seismically qualifying equipment by employing a set of generic enveloping response spectra. This will be done through a technical assistance program with BNL.
t (3) Task 6 i
The EQB, in conjunction with GIB, will establish the appropriate guidelines and acceptance criteria for the seismic qualification of l
equipment in operating plants.
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6.
TECHNICAL ASSISTANCE Technical assistance funding is as follows:
FY-82 FY-83 Task 1
$108K
$ 75K BNL (T.A. Contract by NRR/EQB)
Task 2 SWRI (Funded by RES)
Task 3
$125K
$150K INEL (T.A. Contract NRR/GIB)
Task 4
$ 75K
$ 20K LLNL (T.A. Contract NRR/GIB)
Task 5
$ 99K
$ 15K BNL (T.A. Contract NRR/EQB)
Task 6 7.
INTERACTIONS WITH OUTSIDE ORGANIZATIONS In Task 4 of this program, a program to review and correlate available information on the inservice response of non-nuclear plant equipment that has been subjected to severe seismic or dynamic events and to define data base and establish guidelines for use of non-nuclear experience data, will be developed with technical assistance from LLNL and INEL. A concurrent program is being sponsored by SQUG to collect data on equipment in non-nuclear plants which have been subjected to earthquakes. The owner's group program will be closely monitored by GIB so that data from that program can be used in the LLNL/INEL program.
As this task progresses, it is anticipated that meetings and information exchange with industry groups such as the Atomic Industrial Forum and the Electric Power Research Institute will take place.
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8.
ASSISTANCE REQUIREMENTS FROM OTHER NRC 0FFICES Requirements for assistance from NRC Offices other than input.
through RES _ sponsored work at SWRI discussed in Task 2 are not anticipated at this time.
9.
P0TENTIAL PROBLEMS None expected at this time.
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RESOURCE REQUIREMENTS
SUMMARY
FY 83 FY 84 Contract Dollars for Technical Assistance in Thousands 240X 50K NRR Manpower in Person Years DST GIB 1.0 1.0 SPEB RRAB DSI RSB ICSB CSB ASB PSB 0.08 0.08 CPB AEB.
ETSB RAB 0.05 0.05 i
DE MEB SGEB GSB EHEB MTEB CHEB EQB 0.3 0.5 DHFS HFEB OLB LQB PTRB DL SEPB 0.1 0.1.
ORAB 0.1 0.1 RES Manpower in Person Years RES 0.1 0.1 l
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-A/14
APPENDIX B Related Topics Covered by the INEL Contractor's Report on In-Situ Testing Even though the contractor report is concerned mainly with how to utilize in-situ testing to assist in performing seismic qualification of equipment, other related topics were studied by the contractor. Among them are the following.
(1) Operability and Failure Modes:
In order to develop methods to utilize experience data to qualify equipment, the contractor suggested that a systematic treatement of operability is necessary. The failure modes which result in inoperability, from the contractor's viewpoint, are an essential ingredient to these methods. The contractor first defined inoperability and its causes and then identified all possible failure modes that may cause inoperability during an earthquake.
Inoperability is defined as any action or interaction of component parts or interfaces which prevent a component from performing an active operation or maintaining a state continuously.
Inoperability can result from:
inability to monitor the control condition
- inability to change states when so directed inability to maintain the current state when no state change is directed It is suggested by the contractor that inoperability during earthquake occurs through the following modes:
structural integrity - stress limits are exceeded, permanent defonnation occurs, flaw initiation or extension occurs.
operability loss due to temporary or permanent reconfiguration -
vibratory elastic motion results in a state change or prevents a state change from occurring.
structural interference - excessive relative motion results in a tolerance mismatch, nonstructural changes in state-peizoelectric effects, effects of dynamics on contact resistance, and others. Anywhere a fundamental nonstructural response is affected by vibration or stress.
The contractor then proposed that similarity between two equipment designs can be defined as similarity in potential failure modes. The basic premise involves two pieces of non-identical equipment having a common critical 8/1 l
failure mode. The first piece has been qualification proof tested and its controlling design features are either identical or inherently more fragile than the equipment in question.
In that case qualifying the first amounts to qualifying the other to the same environment. The procedure below is suggested by the contractor to establish seismic capacity based on similarity.
Specify operability requirements, take into account whether equipment is required to operate and/or maintain a continuous state during earthquakes. If there are no requirements during the earthquake than certain failure modes will be eliminated and qualification is simplied.
- Identify the design features /subcomponents which affect operability.
The procedure will be impractical if there are too many.
- Identify similar pieces of equipment, i.e., equipment with nominally the same or less seismic capacity in the potential failure model(s).
Some form of design evaluation / comparison will be required in making this assessment. Equipment used for comparison must be of known seismic capacity.
It is the staff's belief that in-situ testing will be a valuable tool to establish dynamic similarity between equipment through the comparison of the dynamic characteristics (mode shapes, natural frequencies, damping, size, shape, weight, etc.).
(2) Environmental Aging Consideration:
The environmental history of a piece of equipment can produce changes in properties and dimensions which affect its seismic capacity. Addressing the total environmental qualification of equipment in operating plants is f rpractical. An approach based' on the interaction of aging with seismic capacity is adopted by the contractor. Such an approach suggests that since some aging mechanisms will not affect seismic capacity these cases need not be considered in seismic qualification.
The use of in-situ testing in evaluating the effects of aging on seismic qualification has been considered by the contractor, however, no well developed technologies were identified. Consequently, aging has been examined in a broader context where:
- The consequences of aging degradation are examined. This allows the relationship between dynamic qualification and aging degradation to be organized in a fashion which more clearly demonstrates the interaction.
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- Alternate criteria based on failure mode and similarity analysis.
This provides both an organized aging assessment procedure and a method for using test data from "similar" equipment.
- Equipment without specific operability requirements during seismic events have been identified as less vulnerable to aging.
The effect of aging on seismic capacity is illustrated in Figure B-1.
A systematic basis for evaluating aging degradation is provided by the failure mode analysis and the procedures embodied in Figure B-1.
This method as proposed by the contractor is as follows. First, a determination of any aging effects produced by the design basis environments should be conducted.
This involves listing all vulnerable materials and examining environmental data for each. Presently, such data is only available for some materials.
Those components demonstrating no environmental aging require no furthee examination. For components containing materials affected by the design environments the aging mechanisms are defined and categorized by the contractor as follows.
Category I aging: This includes all aging mechanisms which modify the dynamic response. The changes in dynamic response can affect all four failure modes defined earlier. Each failure mode must be examined in light of the anticipated degradation.
If it cannot be established that no significant change in seismic capacity occurs then the critical failure modes should be established. A similar system with a known aged seismic capacity may provide data on which to base the aged seismic capacity. Adversely affected items should be qualified to current criteria.
- Catetory II aging: This is any aging mechanism which could affect the operability of safety equipment when combined with the predicted seismic loads.
It is assumed that the dynamic response has not been affected. This is a type of aging mechanism which impacts only the nonstructural effects.
It need only be examined if a known aging effect exists in a component. Again, seismic capacity can be inferred from tests on similar equipment. However, the requirements on similarity are somewhat more stringent in this case. Any loss of seismic capacity will be due to degradation combined with local strcctural dynamics. Thus, similarity requires that both be simulated.
Category III aging: The mechanisms of this category are those identified which have no effect on seismic qualification (Ref.14 ').
For a typical component many mechanisms would fall in this category.
r B/3
C7 The application of the above approach would probably be most economical if 4
conducted in stages. The contractor proposed that initially all equipment would have a cursory examination for (a) no aging, (b) some aging, though with no effect on seismic capacity, (c) aging with a potential effect on seismic capacity, or (d) too complex to determine easily. For situations where further consideration is warranted the steps are similar to those as described in paragraph (1) of this appendix. The failure modes are used to establish similarity, and data from similar equipment is transferred to the equipment in question. The important factor is that much equipment will exhibit no significant seismic aging interaction of concern and thus screening can narrow the field effectively without overlooking substantial aging degradation.
d
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Environmental I
Aging Yes Seismic Capacity Unaffected
/
Dynamic Dynamic
Response
Response Affected Unaffected i
l Operability Affected b
namic Non-structural Degradation Effecting Seismic Capacity
/
N I
Load Load Magnitudes, Seismic Path Frequency Environment Content a Contributor T
Structural Operation During Normal Integrity Environment l
Affected e--
Structural e
Interference Not a Concern of Dynamic Qualification: In Province of Routine in-service Surveifiance Reconfiguration Figure B-1 Effect of aging on seismic capacity
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