ML19345G011
| ML19345G011 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 02/17/1981 |
| From: | Knight J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17150A134 | List: |
| References | |
| NUDOCS 8102190722 | |
| Download: ML19345G011 (31) | |
Text
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02/17/81
,'O UNITED STATES OF AMERICA fiUCLEAR REGULATORY C0!! MISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of
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PUBLIC SERVICE COMPANY OF
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Docket Nos. 50-443 NEW HAMPSHIRE, ET AL.
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50-444 (Seabrook Station, Units 1 and 2)
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TESTI'40NY OF JAMES P. KNIGHT, Q.1.
Please state your name and describe your present position.
A.l.
My nace is James P. Knight.
I am employed as. Assistant Director for Structures and Components Engineering, Office of Nuclear Reactor Regulation, U.S.- Nuclear Regulatory Cornission, Washington, D. ~ C.
20555.
I am responsible for the review and evaluation of design criteria for structures systems, and mechanical components, the dynamic analyses and testing of safety related structures, systems and components and the criteria for protection against the dynamic effects associated with natural environmental loads and postulated failures of fluid systems for nuclear facilities.
I am responsible for the conduct of operations in i
four branches: the Geosciences Branch, the Mechanical Engineering Branch, i
the Structural Engineering Branch, and the Hydrologic and Geotechnical Engineering Branch.
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Q.2.
Please describe your educational background and previous positions 4
held.
l A.2.
I received a B.S. degree in Mechanical Engineering from Northeastern University in 1957. Since that time, I have completed the equivalent of approximately 35 semester hours at the graduate level in structural dynamics, nuclear engineering and fracture mechanius at the Massachusetts Institute of Technology, Lehigh University and the George Washington University.
From June 1957 to September 1959 I served as a commissioned officer l
with the U. S. Army Corps of Engineers.
From September 1959 to October 1963 I was employed by the Special Products Division of the American flachine & Four, dry Campany, Alexandria, Virginia.
In the latter period of this experience, I had full responsibility for design concept, material selection and analytical review for critical components of high speed spin test equipment, re-entry simulation syst' ems i
and spin stabilization test systems for manned are un-manned spacecraft, i
j In October-1953, I joined the Reactor Radiations Division at the
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National Bureau of Standards.
During this period, I was responsible for the mechanical and structural design, testing and certification for normal loading, seismic loading and postulated accident loading of the NBSR core elements, control rod drive mechanisms and high level radiation handling equipment; I was also fully responsible for the analytical review and experimental certification of the NBSR reactor vessel and a variety of experimental equipment to the requirements of the ASME Boiler and Pressure Vessel Code.
In early 1967, I was appointed Chief of the Engineering Services Section responsible for all mechanical and electrical engineering.
design services for aoth the NBSR facility and experimenta1' equipment-
i development.
Following receipt of the NBSR operating license, I was appointed Vice-Chairman of the NBSR Hazards Committee responsible for review of tne mechanical and structural hazards for all experiments proposed for insertion in the NBSR.
In September 1968, I joined the U. S. Atomic Energy Commission and have remained with this organization through the transition to the U. S.
Nuclear Regulatory Commission.
From 1968 until 1970 I was a member of the Structures and Components Branch.
In 1970 I became a member of the newly organized Mechanical Engineering Branch and in 1972 I was made Chief of the Mechanical Engineering Branch.
In 1976 I was appointed to my present position.
During this time, I have participated in the review and evaluation of over fif ty construction permit and operating license applications and participated in the review and planning activities for Government and industry sponsored programs such as the Heavy-Section Steel Technology Program, the Seismic Safety Margins Progrem, the development of the B31.7 14uclear Power Piping Code and the ASME Nuclev Component Code.
I have served as a member of numerous industry code and standards writing bodies including:
the ASME Section III Subgroup on Pressure Relief, the ASME Section III Working Group for Design of Valves, the ASME Section III Working Group for Design of Pumps, ANSI B16 Subcommittee N - Steel Valves and ANSI B16 Subcommittee H - Valve Operability.
Q.3.
Please explain the seismic design process for nuclear power plants.
A.3.
Seismic design of nuclear power plants requires interaction between two principal endeavors:
(1) definition of seismic hazard, in terms of intensity ar.d characteristics of shaking, and (2) design of structures, systems and components to resist the defined seismic shaking.
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As to the first endeavor, the definition of seismic risk involves consideration of the geologic features of the plant site, observed and recorded ground motions related to these geologic features, and observed and recorded structural response to earthquakes.
The information available l
from historic records, measurements recorded in more recent years, insights that can be gained from various types of analyses and damage assessment following earthquakes must be synthesized to arrive at the best engineering methodology for seismic design of nuclear power plants.
I l
Dr. Reiter's testimony discusses the various avenues that can be taken j
to make a determination of the appropriate seismic ground motion for a given 1
l nuclear power plant site.
The seismic ground motion thus defined cannot, however, be viewed as an end product.
The art of earthquake engineering l
as applied to a NPP has as its central purpose the development of structures, i
systems and components that have the capability to sastain a very broad range of ground motions and yet have design features that offer minimum r
conflict with safety requirements for normal operation.
There are, as in virtually every human endeavor, a number of trade-offs that must be made to achieve the best possible end product.
The necessity for trade-offs is present whether the project involved is a large apartment, building, a school, a hospital, a dam, or a nuclear power plant.
However, the need for a good balance between the seismic input level chosen and the functional requirements of the end product (that is the structures, systems and components i
i as an integrated package) is probably nowhere more necessary that in a nuclear power plant because of t.ha added safety gained by systems that are able to response to thermal and fluid flow induced motions without the incumberance l
of excessive restraint to piping and equipment as a rescli. of seismic design.
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- As to the second principal endeavor, design of structures systems and components to resist designed seismic shaking, the seismic input, once defined, is used in a mathematical process to deterrine how the structure woulc vibrate in response to the seismic shaking.
In order to perform this operation, the structures 3re characterized in a mathematical model by means of the mass of the major parts (floors, walls, domes, etc.) and the stiffnew of the connections between these parts. The stiffness is usually characterized as a spring, and we therefore commonly speak of a spring-mass model.
Through the use of proven and common principles of applied mechanics and mathematics, the response of each of the major portions of the structure, as well as the response of the structure at the mounting location of safety-related systems and components, can be defined for design purposes.
Throughout this process, the characterization of very complex structures by fundamental characteristics, such as mass and stiffness, requires icealizatiorr of the various structural parts.
Because of this, a principal part of the engineering practices involved is the use of techniques which yield a conservatise estimation of the various physical quantities being represented.
In the analytical process these physical quantities interact in comolex ways.
Li order to achieve overall conservatism, it is standard engineering practice to establish a conservative quantity at each stage in the analytical process. The results obtained are therefore recognized as very conservative, but prudent, until such time as a more complete understanding of the interation between the various quantities can be obtained.
The design of the various structural parts is then based upon the results of the design analyses. There is a common misconception that the design of the structural elements is such that the capacity of those structural elements just meets the requirements called for by the analyses.
In fact, much of
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l the structural design is controlled by the size of standard structural 1
members such as reinforcing rods and beams, and construction requirements l
j such as access to make large concrete pours.
In addition, engineering codes specify " code minimum strength" for materials.
These code minimum strengtns are in turn specified by the applicant when the materials are 1
ordered; any material found to be under that strength is rejectec. The result is that the material supplier, in order to assure that he stands no risk of having costly material returned, provides material of considerably i
higher strength.
These higher strengths are born out by the mill test reports for steel and concrete cylinder tests.
l Q.4.
Please describe the interface between seismology and structura' j
engineering in the design of nur. lear power plants.
2 A.4.
It is a rather broad interfate.
It is difficult, even within the disciplines, to define exactly where the line should be drawn, The seismologist makes that transition between the study of the natural processes that lead to ground motion in an earthquake and tbc Effect of that ground motion on structures.
The structural engineer on the other hand makes the transition between the response of the structures and the influence of that structural response on the ground motion.
It should also be reccgnized that massiva structures such as found at a nuclear power plant can have a marked effect on the ground motion actually occurring at the structural foundations.
In other words you have to come at it from both sides.
You have to consider the effect of structures on the site, where they are located, as well as the effect of the site on the structures. The seismologist and the
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engineer must work together to determine an appropriate ground motion, 2
Initially the seismologist comes to the engitteer with descriptions of j
l ground motion that are appropriate for the site considering geologic and seismologic factors alone. There are a number of questions that still a
must be asked before one can determine what is the single conservative
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characterization that will be used for design.
For instanct, depending 4
on the nature of the structure and tha nature of the components in that structure, it might be desirable to consider amplification in a certain range of frequencies.
Other ranges of frequencies may be of no consequence.
It may be desirable to consider other amplifications, depending on the purpose of the structure, and on the sensitivities of the particular structural design.
All of these things are a matter ta be worked out between these two disciplines.
Q.5.
In what form are the seismologist's descriptions of ground motion transmitted to the engineer for design of nuclear power plants?
A.5.
The seismologist provides a design response spectrum. There are two features of the response spectrum that determine the seismic loads that must be considered in the design of nuclear power plant structures and equipment; the shape of the spectrum (provided by Regulatory Guide 1.60 in most instances) and the so called anchor point or high frequency "g" value. This "g" value (0.259 for Seabrook) has no significance in and of itself. That quantity, when associa'.2d with and used to anchor a specific l
l spectral shape then takes on some meaning.
There is often great emphasis perhaps, mistakenly placed on the "g" l
value, which is just a single point on the design spectrum. A possible l
reason for the overemphasis is the similarity of this quantity (acceleration t
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r expressed in "g's" with) the accelerations obtained from seismographs and often reported as a measure of earthquake severity. There is a direct relationship between the zero period acceleration of an earthquake spectrun, and the acceleration values recorded on seismograph records but it is the significance in terms of structural response tha is meaningful for nuclear power plant design and that relationship must be developed in light of experience with performance of well designed structures that have undergone i
seismic loading.
The spectral shape which is used for design of nuclear power plants (Reg. Guide 1.60) is made up of a suite of earthquakes. No single earthquake l
would produce a spectrum which would have those characteristics.
Spectra l
derived from any of the individual earthquake time histories, may have some very sharp peaks (high-acceleration values) but such spectra could be deficient in certain of the frequency ranges of interest (i.e., frequencies at which nuclear power plant structures would respond). We know from experience tnat those high peaks are not of engineering significance because we can inspect structures that nave been exposed to very similar earthquakes (i.e., earthquakes with random high peaks and significant high l
frequency energy) and find tnat there is no damage although analyses usinn the high peak values predict significant failures.
So clearly, those i
peak values have no meaning.
As to what does have meaning, we're into an area of judgment and experience as to what peaks to ignore and what high frequencies should be filtered before we get results that begin to correlate i
with experience.
It is that body of experience that is built into the development of Reg. Guide 1.60, and into the staff methodologies.
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Q.6.
Please summarize the engineering conservatisms included in the seismic design of the Seabrook NPP.
A.6.
These conservatisms are briefly itemized as follows:
1.
Conservatisms associated with the characterization of the design event.
- a. Wide band ground response spectra with conservative amplification factors is employed as a characterization of seismic input as noted in Answer 5 above.
1 The ground response spectra used as input are smoothed, and broad banded (i.e., high energy throughout the range of frequencies at which NPP structures will respond, see Regulatory Guide 1.60).
The spectra for a real earthquake are jagged in nature, producing less response in certain frequency (or period) ranges of the spectra than in adji. cent frequency (or period) ranges. The spectral amplification factors are determined from considerations of the spectra for a set of real earthquakes.
In the case of the development of R. G.1.60, the amplification factors at each fequency were based on consideration c,f about an 84 percent confidence level that the respc,r.,e at a particular frequency would not be exceeded.
b.
Enveloping synthetic time histories.
In the calculation of seismic responses for the design of structures, l
systems, equipment and componants, synthetic earthquake t te histories are developed as input to the mathematical mode': of the structures. These synthetic time histories are required by the NRC i
staff to provide energies equal to or greater than the energies j
' presented by the ground response spectra. The practicalities of l
.ne process by which the artificial time histories are developed inevitably lead to a significant increase ir; actual input over that characterized by the ground response spectra. This occurs because the time necessary to get the synthetic time history spectra any closer than ten or fifteen percent above h ground response spectra j
becomes extensive in what is essentially a trial and error process.
2.
Conservatisms associated with the methodologies for seismic analysis, design, and qualification.
i a.
Structures, systems and components for Nuclear Power Plants are required by NRC rules to ha dasinned to undergo earthquakes up to the Operating Basic Earthquaxe with a larger margin than is required for the Safe Shutdown Earthquake. The OBE design requirement typically controls the strength of the ttructural elenents comprising the plant. Thus these structural elements have capacity significantly above that just required to meet the loading from the SSE.
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!l b.
Three input ci vonents.
l For desior purposes the earthquake is characterized by three componente motion; two horizontal (one North-South and one Ea'.
3>st) and one vertical The three components are considered to occur simultaneously, that is peak notion in both horizontal directions and the vertical direction is imposed on the structures system and components at the same -
instant whereas this occurrence in any actual earthquake is extremely unlikely.
c.
Loading combinations.
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The loading combinations employed for the Seabrook Station j
consider other loadings (e.g., dead weight, live loads, pressure loads, etc.) in addition to the seismic loadings.
Seismic loading is only a part of the total loading and in fact, other loadings other than seismic govern design of the majority of the structural elements in the plant.
d.
Effect of inelcstic behavior.
she allowable design limits for the Seabrook Station result in very little if any ine?astic. action (i.e., stresses beyond the yield point of the materials).
In reality, well engineered structures, components and systems are capable of sustaining loads which are beyond those which would bring the materials of construction to tne elastic limit without sustaining damage.
For small excursions into the inelastic range, seismic inertial loads are reduced as a function of the amount of inelastic action in comparison with those calculated elastically.
This phenomenon can have the effect of reducing accelerations of elastically calculated responses by as much as 1/3 without loss of structural function.
e.
Use of maximum response spectra for multiple supported systers.
Where the system has multiple supports, the maximum response spectra are generally applied to all support points so that conservative seismic loads are generated for design purposes.
f.
System Redundancy.
I Even identically designed redundant systems may not always experience similar seismic excitation due to different counting locations, with different structural filtering effects. -Thus a single loss of redundancy may. not mean a loss of function for
- d the system. This provides additional assurance that a needed
- t function will safely withstand a seismic event.
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g.
Test input response spectra for qualification of electrical and mechanical equipment.
The test input spectrum for seismic qualification of electrical I
and mechanical equipment.at the Seabrook Station is required by the !4RC Staff to be equal to r greater than the calculated in-structure response spectru, at the mounting location of the i
specific item undemoing tes*
As in the development of artificial time histo,;es di. cussed in 1.b. of this answer above, the practicalities of the process by which the test spectra are dmloped lead to significant conservatisr' in comparison to the in c ucture spectrum that represents the required level of
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motion.
h.
Multi-axis testing.
The test input motions must be applied to the vertical and the horizontal axes simultaneously unless decoupling of responses along two directions is justifiable.
A number of OBE tests (usually five) are performed orior to testing at the SSE level.
The number of OBE tests is conservatively i
selected to represent the upper bound for c plant site.
This provides an additional margin in the consideration of cyclic loading effects.
i 3.
Conservatisms resulting from actual (vs. design) structural and mechanical resistance.
a.
Allowable stress limits.
Engineering codes specify " code minimum strength" for materials, as noted in Answer 3.
The:.e codes minimum strengths are in turn specified by the applicant when the materials are ordered, any material found to be under that strength is rejected. The result is that the material supplie provides material of higher strength.
t Also, margins exist between allowaole stresses and ultimate strengths.
b.
28 day concrete strength.
Designs are usually based upon the 28 day design strength of concrete.
Concrete continues to gain strength with increasing time beyond
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28 days. Additionally, the strength at 28 days often exceeds that called for design strength.
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Static strength vs. dynamic resistance.
Code material strengths are based upon statis load tests.
Since dynamic loads contain a limited amount of energy and are applied at a faster rate, the margin between stress limits and failure for dynamic loads is greater than that for static loads.
d.
Standard size structural members and pipes.
The design of the structural elements is such that, as noted in Answer 3, their capacities usually exceed the requirements called for by the analyses.
Much of the actual structural design is controlled by the availability of standard structural members such as beams and piping sections, so that larger sizes than are needed are often used.
e.
Redundancy in indeterminite structures and components allows for redistribution of loads.
4 In the seismic analysis for Seabrook each structural element must meet allowable design limits.
However, from the standcoint of function, majre structures and components can tolerate much deformation, and typically failure of nunerous structural members.
This deformation and loss of structure.1 members can be sustained because of redundancy, (i.e., more than one path available to carry loads) which allows for redistribution of loads formerly carried by failed members, g.
Minor attachments absorb energy.
Nonstructural elements which are not considered to carry any loads in design, do absorb energy through inelastic behavior or collapse
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during a siesmic event.
h.
Nuclear quality assurance (QA) program.
The nuclear QA procedures are more stringent than most found through-out the construction industry. This provides additioral safety for nuclear plants beyond that considered acceptable for nost non-nuclear facilities designed using many of the same practices as used for nuclear plants.
These conservatisms are difficult to quantify; however, the extent of these structural and mechanical conservatisms for plants designed using current standards, such as Seabrook Station, he been estimated by studies -
made by Newmark, and Cornell; -1/
a median factor of safety for structures, 1/
"On the Seismic Reliability of Nuclear Power Plants," C. A. Cornell and N. M. Newmark, May 1978.
equipment and piping has been estimated to be within the range of 4 to 8.
Additional studies currently being performed for the NRC staff by Battelle Columbus Laboritories and based on detailed analyses of nuclear plant components are providing data that confirm this level of seismic margin for pumps, valves and piping such as being designed for the Seabrook Station.
Q.7.
What would be the response of the Seabrook nuclear power plant to a Modified Mercalli Intensity VIII or greater earthquake at the site?
A.7.
Given an intensity VIII earthquake at the site, the Seabrook Nuclear Power Plant could be expected to suffer little if any detectable distress. Although regulatory requirements are designed to assure that safety related structures, systems and components remain functional and these requirements do not address those portions of the plant necessary for continued power generation, it is extremely likely that continued operation of the plant would be possible, albeit under the requirements of Appendix A to 10 C.F.R. Part 50, the plant would have to be shutdown nd examined should any earthquake exceeding tna OBE -2/
occur. The expectation that an intensity VIII earthquake at the Seabrook Site would not cause significant damage to those portions of the plant designed to remain functional at the SSE level (Regulatory Guide 1.60 Spectra Anchor at l
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0.25 g) stems from two essential factors.
i 1.
The design procedures as discussed above typically yield an actual capacity well in excess of the design goals.
t 2/ See Appendix A to 10 C.F.R. Part 100, s III(d).
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2.
A structural element or item of equipment that just barely made the design goal would still be operating in the elastic range, i.e., no i
permanent deformation would occur.
The conclusion that the remainder of the plant, that is those structures, systems and components that are not specifically designed to resist seismic 1
loading would also suffer no significant damage stems from observation of the remarkable seismic resistance demonstrated by well designed industrial facilities the world over.
These same observations add high confidence to the predictions of seismic resistance made for seisnically designed facilities.
The high pressure piping systems process control instruments and components, pumps, valves, heat exchangers, pneumatic and hydraulic systens that make up a nuclear power plant have their total analogy in chemical process plants pipelines and fossil pcwer pit 'ts that have safely withstood very large eartnquakes a'though little if any specific seismic design was required.
l The design practice and procedures necessary to assure reliable function i
over the lifetime of these chemical processing plants pipelines and fossil power plants bring aoout that very balance between strength and flexibility I
l that diso provide a very necessary element of seismic resistance and i
j provide us with a base of experience that we have used to assure that seismic design requirements for nuclear power plants are adequate.
Should an earthquake significantly larger than the Safe Shutdown Earthquake o: cur at the sit 0, there is tvery reason to expect that the i
systems necessary to shutdown and cooldown the reactor would still remain l
l functional.
As noted in Answer 6 above large margins of conservatism in the as built plant result from the seismic design process. These margins assure that the energies of an earthquake significantly greater than the SSE would still not render the Seabrook Station unsafe.
In studies performed by and for the NRC Staff, a minimum margin of 1.5 to 2 for each element cf the structures systems and components have been demonstrated to result from the design practice and design procedures employed for nuclear power plants.
Above these minimum margins up to factors between 3 and 8 some degradation and loss of system eventually might be expected but the primary system function couid be expected to continue.
The availability of these large factors of safety would depend on the structure, systems or components being fabricated, installed and maintained in the condition anticipated by the design analysis.
- However, given the unusual attention (unusual that is with respect to other types of construction) focused on these matters in nuclear power plant design, tnere is every reason to expect that the majority of these large margins would indeed be available.
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