ML20027A444
| ML20027A444 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 11/15/1978 |
| From: | Knight J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20027A437 | List: |
| References | |
| NUDOCS 7812040338 | |
| Download: ML20027A444 (56) | |
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TESTIMONY OF JAMES P. KNIGHT
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In tne Matter of
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Docket Nos. 50-275 0.L.
PACIFIC GAS AND ELECTRIC COMPANY
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50-323 0.L.
(Diablo Canyon Nuclear Power Plant,
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Units Nos. 1 and 2 My name is James P. Knight.
I am Assistant Director for Engineering in the Division of Systems Safety, Office of Nuclear Reactor Regulation, of the U.S. Nuclear Regulatory Commission.
The attached testimony, antitled
" Staff Testimony Concerning Seismic Design of the Diablo Canyon Nuclear Power Plant," was prepared under my supervision with the aid of the follow-ing individuals:
Dr. P. T. Kuo, Dr. P. Y. Chen and Mr. E. J. Sullivan, Jr.
This testimony is in partial response to Invervenor's Contentions numbers 3, 5, 4, 6 and 7.
My professional qualifications and those of the above cosponsors of this testimony are attached.
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STAFF TESTIMONY CONCERNING SEISMIC DESIGN OF THE l
DIABLO CANYON NUCLEAR GENERATING STATION Introduction In this testimony we will briefly describe the seismic design process, with emphasis on uncertainties and compensating conservatisms employed in the design of structural and mechanical elements.
We will then relate this background to the staff review of the Diablo Canyon Nuclear Generating Station (DCNGS).
General Seismic design of nuclear power plants requires interaction between two principal endeavors:
(1) definition of seismic risk, in terms of intensity and characteristics of shaking, and (2) design of structures, systems and components to resist the defined seismic shaking.
The definition of seismic risk involves consideration of tiie 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 frem historic records, measure-ments 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 engineering methodology that will yield i
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, optimum design parameters.
Dr. Newmark's testimony discusses the applica-tion of this process to the determination of appropriate seismic risks for the DCNGS; a summary of the criteria invoked for DCNGS is given under the heading " Ground Response Spectra For Use In Design Data" in this testimony.
The seismic input, once defined, is used in a mathematical process to determine how the structure would vibrate in response to the seismic shaking.
In order to perform this operation, the structures are charac-terized in a mathematical model by means of the mass of the major parts l
(floors, walls, domes, etc.) and the stiffness of the connections between these parts. The stiffness is usually characterized as a spring, and we therefore commonly speak of a spring-mass model.
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ihrough 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.
l Throughout this process, the characterization of very complex structures by fundamental characteristics, such as mass and stiffness, requires idealization of the various structural parts.
Because of this, a principal part of the engineering practices involved is the use of techniques which yield a conservative estimation of the various physical quantities being represented.
In the analytical process these physical quantities interact l
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In order to achieve everall 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 interaction between the various quantities is 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 elements just meets the requirements called for by the analyses.
In fact, much of the structural design is c,ntrelied b tire sfre of standard stettetural members such as reinforcing rods and beams, and construction requirements such as access to make large concrete pours.
In addition, engineering codes specify " code minimum strength" for materials.
These code 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 supplier, in order to assure that he stands no risk of having costly material returned, provides material of considerably higher strength.
These higher strengths are born out by the mill test reports for steel and concrete cylinder tests.
There is normally no motivation to go back and assess the true strength of various structures, systems and components, because the costs of reanalysis and time lost swamps any reduction in size or equipment capabilities that may be gained.
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In the design of structures and equipment, it is convenient in typical engineering analyses to assure that all elements of the structure or i
equipment remain elastic or nearly so, i.e., stresses below the yield point of the material so that any permanent deformation is very small.
One of the principal reasons for this is that the maintenance of elasticity negates,the need for complex interaction analyses to determine margin to failure.
From the standpoint of function, major structures and components in nuclear plants, as well as in other commercial applications, can tolerate much inelastic deformation and typically loss of numerous structural members.
This deformation and loss of structural members can be sustained because of redundancy, i.e., more than one path available to carry loads and load sharing or redistribution, i.e., the load formally carried by a j
failed member is redistributed to other members.
The end result of the conservatisms employed in the analyses followed by the conservatisms resulting from standard design practices is a structure with a seismic capability well in excess of the established design goal.
This is the reason _that the record is replete with cases where well-engineered structures, even those for which no specific seismic design standard was invoked, have withstood major earthquakes while remaining fully functional.
The testimony above spoke of the numerous conservatisms accruing as a result of the use of standard structures, shapes, sizes and materials.
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very analogous phenomena occurs in the testing of the equipment and compo-nents.
In order to assure fully representative testing with respect to both direction and characterization of vibratory input, a given piece of equipment is subjected to a large number of individual tests, any one of which often equals or exceeds the most likely vibration to be seen by the.
equipment in any actual earthquake.
The number of tests typically range from 10 to 50 before a program for an individual piece of equipment is completed.
In this way the question of aftershock or marginal performance of prototype equipment that may not be fully characteristic of installed equipment is adequately addressed.
Clearly, the history of vibratory loading established during the test program exceeds even the most pessi-mistic view of possible effects of aftershock loading.
Any concern that some fatal flaw that may hinge on a subtlety in fabrication or installation l
may not be discerned by a single shaking has to be put aside.
In addition to the number of tests employed, the magnitude of tests, once again, due to the practicalities of designing tests equipment to meet myriad test requirements, always exceeds that required (already conservatively defined by virtue of the structural analyses).
l The Significance of the Operatino Basis Earthouake as a Safety Parameter In the design of structures, systems and componen a, the operating basis earthquake (OBE) is treated as a fully expected occurrence during the lifetime of the plant.
For this reason the design methodology employed to assess the effect of OBE loads is identical to that employed for any cther l
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. aspect of normal operation of the plant; given the occurrence of an earthquake, up to and including the level selected for the OBE, the owner of a plant is expected to continue operation without intervention of regulatory authorities.
The resulting design is usually such that the effect of the OBE in terms of stresses in reactor and auxiliary systems is 1.ess than the cumulative effect of some normal plant operations such as startup and cooldown.
The significance of the OBE with respect to the health and safety of the public has to be measured with respect to continued operation of the plant when and if such an earthquake occurred.
- Clearly, the plant designed to shut down safely in the event of a safe shutdown earthquake (SSE) has an ever increasing range of capacity for safe shutdown for all those earthquake magnitudes from the SSE level down to the OBE.
The central question with respect to the OBE then is not whether the public health and safety would be adversely affected should the OBE occur, but whether continued operation of the plant throughout its lifetime could be accepted without reevaluation of the plant's systems, structures and components should the OBE occur.
Aoplication of the OBE to the DCNGS l
As discussed in the testimony of Dr. Stepp, the OBE level of 0.2g designated l
l for the DCNGS defines the earthquake that could reasonably be expected to affect the plant during the operating life of the plant.
The seismic instrumentation network at the DCNGS (the most extensive in the country, perhaps the world) will provide a specific characterization of any i
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. earthquake that might occur both in terms of free-field motion and struc-tural responses.
If a seismic event greater than that specified as the OBE did occur or if the structural responses were greater than those predicted for the OBE (or excessive under any seismic event), a detailed e nluation of the plant structures, systems and components would be required to assess damage that would be unacceptable for lifetime operation.
DIABLO CANYON REVIEW Backoround The Diablo Canyon plant was originally designed to maintain safe shutdown capability for a 0.4g earthquake.
Subsequent to discovery of the Hosgri fault, the applicant assessed the ability of the plant to withstand a 0.5g earthquake.
The capability to withstand this approximately 20 percent increase in seismic loads was found to exist with very little or no modif'-
cation to the plant and equipment, a fact not at all surprising given the numerous conservatisms employed in seismic design as practiced today.
When it became necessary to assess the Diablo Canyon plant for a seismic event originating on the Hosgri fault, it was recognized that simple application of current seismic design criteria could not be an appropriate basis for assessing the acceptability of a plant that was on the verge of completion when a major change in required seismic resistance was defined.
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. The same economic principles that dictated the use of relatively simple analytical techniques, coupled with very conservative acceptance criteria for initial ~ design purposes, now dictated detailed analyses and use of actual material strength data as well as evaluation of failure modes and minimum requirements to maintain functionality.
No significant structural modifications were required for the containment building or the auxiliary building.
These two buildings were the focus of the seismic design effort in the original plant design.
The turbine building, although not originally considered as a seismic Category I building, was found to have sufficient seismic resistance to maintain required design function in the lower (0.4g and 0.5g) earthquake levels but required extensive modification for the Hosgri event.
The intake structure and safety-related tanks have been the subject of intensive analyses to assure stability against sliding and tripping, and extensive modifications have been found necessary for the tanks.
The major reactor components and other major fluid system components were found to require little or no modification; however, a very large number of piping supports had to be modified, numerous pieces of electric and electronic equipment were found to require modification after testing,' and several component support structures were found to require additional bracing.
An interesting phenomenon occurs when these modifications are made.
Just as the engineering practices used in the original design l
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s provided large measures of conservatisms, so did the sal::e practices increase conservatisms in the modifications.
Thus, upgrading to meet a new seiscic level has in fact resulted in design capabilities in most of the modified systems exceeding current (Hosgri) seismic design requirements.
Review of Structures, Systems and Components There are some areas (a very few in comparison to the myriad items considered) where criteria different than those presently accepted by the staff have been employed in the final evaluation of the Diablo Canyon Nuclear Generating Station.
In the following portion of our testimony, we discuss each of these areas and our bases for acceptance. We will also describe the characteristics of the safety-related structures, systems and components that are pertinent to discerning the appropriateness of analyses and tests utilized to assure that these items will remain functional under severe seismic loading.
And finally, we will describe each of the appropriate analytical techniques and test programs that were employed for the safety-related structures, systems and components at the Diablo Canyon Nuclear Generating Station.
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The format and much of the content of the following sections of this testimony parallal Supplement Nos. 7 and 8 to the Safety Evaluation Report l
for the Diablo Canyon Nuclear Generating Station.
Greater detail concerning the subjects addressed can be found in Supplement Nos. 7 and 8.
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. We believe that the widespread reevaluation and detailed staff review portrayed by this summary provide ample basis for assuring that the struc-tures, systems and components important to safety and the Diablo Canyon Nuclear Generating Station will retain function and perform as required under the extreme seismic loading identified for this facility.
Ground Response Spectra for Use In Design Two sets of ground response spectra were involved in the reevaluation of the DCNGS for the Hosgri earthquake.
One set had been-developed by our consultant, Dr. N. M. Newmark, and the other had been developed by the applicant's consultant, Dr. J. A. Blume.
Our evaluation of the criteria discutcad above was provided in Supplement No. 5 to the Safety Evaluation Report.
This was done by reference to Dr. N. M. Newmark's report, in Appendix C to Supplement No. 5, which l
described the bases for the spectra derived by Dr. Newmark.
The November 1976 staff report to the Advisory Committee on Reactor Safeguards discussed how the applicant's consultant (Dr. J. A. Blume) had derived his spectra, compared the Blume spectra with the Newmark spectra, and indicated that the Newmark spectra would be limiting in most cases.
The Blume spectra and the Newmark spectra were both used by performing separate calculations and using the more controlling result.
For horizontal analyses, the response quantities were computed for four earthquake inputs,
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. namely, the Blume and the Hewmark Hosgri event inputs, each in the north-south and east-west directions.
For vertical analyses, the applicant found that the Newmark Hosgri input was the governing one and, therefore, l
it was used in the reevaluation.
Both the Blume and Newmark spectra have been reduced for large structures to account for building size effects.
This reduction is associated with nonsynchronized ground motion waves under the structures, an' where it is d
used, allowance has been made for torsion caused by such waves.
It has frequently been observed that structures with frequencies higher than about two Hertz on large foundations appear to respond with less intensity to earthquakes than do smaller structures and, more specifically, than does free-field instrumentation.
Several researchers, e.g., Yamahara in 1970, N. Ambraseys in 1975, and R. Scanlan in 1976 (References 1 through 3) have attempted to provide a rational explanation for this observed behavior.
These references give, in general, a relationship between the average acceleration over the area of the foundations as a function of the relative foundation width compared to the predominant wave length of earthquake input motion.
l In Supplement No. 5 of the Safety Evaluation Report for DCNGS, the NRC staff stated that the following reduction in horizontal response spectra t
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, for DCNGS structures should be allowed to realistically account for the above described observation:
Af = A, R
where Af = reduced acceleration for foundation A,= a;celeration for free field R = 1 - 5t but not less than 0.67, where R = response spectra reduction factor T = seismic wave transit time (in seconds) providing a measure of the extent of acceleration averaging effect applicable to the foundation.
This reduction factor "R" for DCNGS structures is based on studies of the spectrum amplification factors obtained from the recorded Pacoima Dam response spectra.
The lower limit on R (i.e., 0.67) is purposely kept high for adequa'e conservatism in the application of this concept to DCNGS structures because the concept is based on a small amount of data.
More extensive discussion of the technical bases for the above described concept is provided in Dr. Newmark's report in Appendix C to Supplement
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, No. 5 of the Safety Evaluation Report for the Diablo Canyon Nuclear Generating Station, Unit Nos. 1 and 2.
Associated with translational deformation due to earthquake wave motion, there are torsions and tiltings of a building foundation.
These torsional effects are accounted for by assuming an eccentricity of horizontal seismic force of seven percent of the width of the structures if added on a square-root-of-the-sum-of-the-squares basis, or five percent if added on an absolute sum basis.
The tilting effects are judged minimal and neglected in the analysis.
In accordance with the usual practice, vertical ground motions were taken as two-thirds of the horizontal ground motions.
No reduction for the effect of large foundations was used for vertical motions.
Methods of Analysis for Structures t
The analysis methods used for structures were common to accepted seismic design practice with the exception of the items listed below:
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o Average values of material properties, from tests of the actual j
materials installed, were used to determine allowable stress levels instead of using code-specified minimum material properties.
In the original analysis, code-specified minimum values were used in accordance with standard practice.
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. In our review we have found the use of actual material strengths acceptable since some margin remains.
For concrete, the appropriate average 28-day
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test strength was used.
Since concrete continues to gain strength with age after 28 days, the installed concrete will be stronger.
For steel, average mill test strength was used.
Since the steel is ductile and the structures are designed to remain below yield (with two possible exceptions as discussed above), margin remains.
o A modified procedure was used for smoothing the raw floor response spectra.
For the reevaluation, smoothing was done by free-hand averaging of floor response spectra except at the peaks where it was widened by 15 percent on the low frequency side and 5 percent on the high frequency side without reduction of the peaks.
In the original analysis the peaks were broadened by 10 percent on both sides after being lowered by 10 percent.
i The purpose of widening the peaks is to account for possible variations in the predicted structural frequencies.
Our current criteria indicate widening by 15 percent on both sides of the peaks.
However, since actual material strengths are being used in the reevaluation, the calculated structural stiffness is closer to the maximum stiffness than usual, indicating a lesser need for peak broadening on the high frequency side.
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. Containment Shell Analysis The reactor containment shell is a cylindrical, reinforced concrete struc-ture which consists of a 142-foot high cylinder, topped with a hemispherical dome.
The cylinder wall is 3 feet 8 inches thick, and the dome is 2 feet 6 inches thick.
The inside diameter is 140 feet.
The base is a circular slab 153 feet in diameter and 14 feet 6 inches thick, with the reactor cavity near the center.
The inside of the shell is lined with welded steel plate which forms a leaktight membrane.
The liner is 3/8-inch thick on the wall and dome and 1/4-inch thick on the base slab.
An uisymmetric model was used to compute the transistional response of the shell due to the horizontal component of the ground motion.
Since the center of mass and the center of rigidity coincide, the translational analysis did not yield any inherent torsional modes or responses in the shell.
Peak responses and acceleration time-histories were determined for various nodes using the time-history modal superposition method.
In accordance with the general methods discussed above, the torsional response due to horizontal inputs was determined using a time-history modal superposition method.
Two lumped mass models were used assu. ming an accidental eccentricity of five percent and seven percent, respectively, of the structural dimension in the direction perpendicular to the applied iced.
The torsional response from the two lumped mass models was ccmbined with the corresponding horizontal translational responses from the i
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, axisymmetric model.
The applicant has demonstrated the equivalency of the axisymmetric model and the lumped mass model by comparing the natural frequencies and participation factors of the two models.
Vertical acceleration and displacement responses were determined by a time-history, modal superposition analysis of the axisymmetric finite element model.
The maximum shell forces and moments were determined by a response spectrum analysis of the same axisymmetric model.
The appropriate responses from the horizontal and vertical components of the ground motion were combined at the mid-thickness of the shell using the square-root-of-the-sum-of-the-squares rule.
Raw floor response spectra for both the Blume and Newmark inputs were l
computed from the acceleration time-history responses at various nodes of l
the containment shell.
l To develop the horizontal floor response spectra, the translational spectra were combined with the torsional spectra.
The rc:ulting spectra were then i
combined on a square-root-of-the-sum-of-the-squares basis with the horizontal l
component due to the vertical input to yield the design spectra.
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~. horizontal floor response spectra are applicable in any horizontal direction due to the symmetry of the structure.
The vertical floor response spectra were genereted based on a dynamic vertical analysis of the exterior structure. These raw floor spectra were calculated for the various damping values appropriate to different equipment and were smoothed in accordance with the general methods discussed above.
These floor response spectra were then used in the reevaluation of components, pipes and electrical conduits attached to and supported by the c.ontainment shell.
Containment Interior Structures Analysis The containment interior structures consist of two concentric concrete walls with outside radii of 53 feet and 17 feet, respectively. The inner cylindrical wall varies in thickness from 3 feet to 8.5 feet; it supports the reactor vessel.
The outer cylindrical wall houses the steam generators, the reactor coolant pumps, the pressurizer, and other equipment.
Concrete slabs provide horizontal diaphragm stiffening action at elevations 114 feet and 140 feet.
A polar gantry crane is mounted on top of the outer cylinder wall.
The interior and exterior structures are supported by a common heavy concrete base slab of 14.5 feet thick for most part of the slab except the reactor cavity area where the slab is nine feet thick.
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- An equivalent axisymmetric model of the interior structures was used in the analysis.
The model was assumed fixed at the base.
Because some of the shear walls are not symmetrical about the structural axis, the axisymmetric shell elements were established on the basis of the equivalent cross-sectional properties of the structure.
Small irregularities in the foundation slab w ee neglected.
The equivalent mass density used for each element reflects the mass of attached mechanical equipment and all other associated masses, as well as the concrete mass density.
A time-history modal superposition analysis of the axisymmetric model was used to compute the translational response of the structure due to the horizontal component of the ground motion.
Because the center of mass and the center of rigidity of the model coincide, the translational analysis does not yield any inherent torsional modes or responses in the structure.
For the torsional response due to horizontal inputs, a time-history, modal l
superposition analysis of two lumped-mass models was performed.
- Again, the lumped masses represented the accumulated mass of the concrete struc-ture, the steel framing, the equipment, the reactor and the ~other masses.
l The center of rigidity and the center of mass of the interior structures 1
are approximately coincident ~for the east-west ground motion, but are separated by 1 foot between elevations 114 feet and 140 feet, and by 2.5 feet between elevations 91 feet and 114 feet for the north-south 1
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Because the eccentricities for the north-south ground motions are the larger, the torsional response to that motion was calculated and used for both orthogonal horizontal directions.
Vertical responses were determined by a time-history, modal superposition, analysis of the lumped-mass model. A coupled model incorporating both the concrete interior structure and the steel annulus structure was used for the analysis.
At our request, the applicant established the equivalency of the axisymmetric model and the lumped-mass model by comparing the dynamic characteristics of the two models.
The difference was insignificant.
Raw horizontal and torsional response spectra for both the Blume and Newmark inputs were generated from the acceleration time-histories at various elevations along the center line of the interior structure.
These response spectra were used for both the north-south and the east-west directions due to the assumed axisymmetry of the structure.
Equivalent horizontal response spectra at the location of interest were developed using the method described previously.
The raw vertical floor response spectra were generated based on a dynamic vertical analysis of the interior structure.
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, The above raw floor response spectra were calculated for various values of damping appropriate for systems and equipment and were smoothed according to the procedures discussed above.
The smoothed horizontal and vertical floor response spectra were used for the reevaluation of systems and components.
Auxiliary Building The auxiliary building is a reinforced concrete, shear wall box-type structure except for the fuel handling area, which is a steel structure.
The shear walls are generally 3 feet thick, with a minimum thickness of 2 feet.
Slabs are generally 2 feet thick.
The walls of the spent fuel pools are a minimum of 6 feet thick except for local areas around the fuel transfer tubes.
The foundation slab under the spent fuel pits has a minimum thickness of 5 feet.
The sides and bottoms of the spent fuel pools are lined with stainless steel plate, 1/4-inch thick on the bottoms and 1/8-inch thick on the sides.
The main floor levels in the auxiliary building are at elevations 60, 73, 85, 100, 115'and 140 feet.
Elevations 60 feet and 73 feet are below ground level, which is at elevetion 85 feet except for the east side of the building, where ground level is at elevation 115 feet.
The only connections between the auxiliary building and other structures are the fuel transfer tube and miscellaneous piping.
The fuel transfer
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-s., tube is fitted with expansion bellows which allow relative movement between the auxiliary building, the containment shell, and the containment internal structure.
The design of the expansion bellows considers the maximum axial and lateral relative deflection that could occur during the Hosgri event.
The auxiliary building was analyzed for the three components of the Hosgri event ground motion using a modal superposition time-history procedure.
Peak responses were calculated, and acceleration time histories were determined for various nodes for use in determining raw floor response spectra.
Three different mathematical models of tre auxiliary building were used for analyses:
one in the north-south direction, one in the east-west direction, and one in the vertical direction.
The models were assumed fixed at the elevation 85 feet (i.e., ground motion input was applied into the structure at this elevation).
l All degrees of freedom have been defined at the center of mass.
Part of the structure between elevations 60 feet and 85 feet is below grade and was not lumped as a separate mass but was assumed to be part of the foundation soil mass.
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- To account for the embedment effect of the upward-sloping grade above elevation 85 feet, a set of equivalent spring stiffnesses was determined based on the elastic half-space theory.
These stiffnesses were then increased to account for the effect of a sloping surface on the east side of the building. At our request the applicant performed a parametric study by varying the stiffness values.
The applicant demonstrated that the variation of the stiffness values has little effect on the structural responses.
Locations of the centers of mass and rigidities were calculated for each level to consider the torsional mode of vibration in the analysis.
The actual (geometric) eccentricities combined with five percent assumed accidental eccentricities have been included in the model for the north-south ground motion.
For east-west analysis, the auxiliary building was considered symmetrical and, therefore, only accidental eccentricities l
were included in the snodel.
To establish whether the five percent or the seven percent assumed accidental eccentricity gave the higher responses, the results from the i
two methods discussed previously were computed.
The applicant indicated that the absolute sum of the horizontal response and the torsional response due to the five percent assumed eccentricity in every case gave higher values than the square-root-of-the-sum-of-the-squares combination
... of the horizontal response with the torsional response due to the seven percent assumed eccentricity.
For vertical analysis the model was based on the assumption that the floor slabs were rigid (i.e., have fundamental frequencies higher than 33 Hertz).
This assumption holds for all auxiliary building floor slabs except the floor of the control room.
The control room floor was modeled by finite elements.
The wall and column supports of the slab were assumed to be rigid in the vertical direction; rotational degrees of Yreedom were modeled by springs whose rotational stiffness was equal to that of the walls.
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The smoothed horizontal, torsional and vertical floor response spectra for damping values appropriate to systems and equipment were developed from raw spectra determined from the acceleration time histories at various nodes.
The spectra were defined at the actual (geometric) centers of mass in the building.
The horizontal and torsional spectra given are for the l
controlling case of five percent accidental eccentricity.
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l No torsional spectra were given for the roof of the steel structure over the fuel-handling area (elevation 188 feet).
Torsional effects were 1
accounted for by increasing the north-south and east-west horizontal response at this level by 10 percent, thus yielding significantly higher l
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. torsional response (double) than would be required under any building code.
Intake Structure Analysis The intake structure is a reinforced concrete shear wall building 4
constructed with 3,000 pounds per square inch miafmum specified strength concrete.
Except for the auxiliary saltwater conduits, watertight doors and ventilation, it was not originally classified as seismic Category I.
However, it was analyzed to assure protection of the auxiliary saltwater system in the event of an earthquake and it has been evaluated for the Hosgri event.
The structure is approximately 205 feet long and 100 feet wide.
The long dimension corresponds to the north-south direction, as assumed in the analysis, and is parallel to the seaward face of the structure.
The intake structure is surrounded by rock backfilled on three sides, while the fourth (western) side is exposed to the Pacific Ocean.
The top deck of the structure has a maximum elevation of +17.5 feet.
A small concrete ventilation tower extends to an elevation of +31.0 feet.
The structure is supported by a concrete mat foundation at an elevation of
-31.5 feet.
The top level of the structure consists of a concrete slab la inches thick.
At elevation -2.1 feet, the pump deck floor supports the four main circulating water pumps and the four seismic Category I auxiliary
, saltwater pumps.
Water for these pumps is brought in through the exposed western (ocean) side of the structure.
The structure is symmetric about a vertical plane in the east-west direction through its centerline.
The ocean side of the structure at the level of -2.1 feet has no floor diaphragm and is connected to the land side of the structure by a thick, vertical north-south shear wall running from elevation -7.7 feet to the top-deck level.
At the top-deck level, this wall is connected, through an 18-inch thick horizontal diaphragr.: with numerous openings, to the remaining 18-inch and 24-inch thick slabs over the pump area.
The reismic analysis of the intake structure was carried out by initially separating the structure into two basic parts:
(1) the pump deck base, consisting of the massive landside portion of the structure, from elevation
-31.5 feet to the -2.1-foot pump deck level; and (2) the remainder of the structural system.
The pump deck base was considered to be the ground for the analysis of the remainder of the structure.
Three three-dimensional mathematical models were developed to model the structure:
an east-west model, a north-south model, and a vertical model.
Each model is structured to represent the vibrational response character-istics of the structural system due to sei::aic input in the corresponding direction.
All three models use typical finite element methods suitable for the structural system.
For all models, the floor slabs and most vertical walls were modeled as flat plate elements to include both
. membrane (in plane) and bending (out-of plane) behavior.
Some thick chear walls near the symmetry plane of the structure in the east-west direction have been modeled as three-dimensional solid elements.
Provision for accidental torsion was taken into account by assuming a torque equivalent to the total base shear multiplied by five percent of the long dimension of the structure.
Only those modes with frequencies of oscillation less than or equal to 33 Hertz have been considered significant for response computations.
The fundamental frequencies of the system in the north-south, east-west and vertical directions are 10.4 Hertz, 25.6 Hertz and 16.0 Hertz, respectively.
Shear stresses in the walls due to the east-west earthquake are within the allowable 120 pounds per square inch of the concrete alone.
Based on the low magnitude of the stresses in the elements of the structure, there is no danger of structural damage from east-west earthquake forces.
The applicant found the reinforced concrete seaward piers to be considerably overstressed, leaving two principal options to short of modifying the piers, namely, to take credit for structural ductility or to demonstrate that an assumed failure of the piers would not disable the auxiliary seawater system.
s,. s' The applicant performed the additional structural analyses using two different techniques to determine the ductility required for these piers.
The energy reserve technique resulted in a ductility ratio of 1.13, within the limit of 1.3 which is set forth in the structural specification.
A more conservative technique based on moment curvature relationship of structural members resulted in a ductility ratio of 1.52.
Since the calculated ductility results are reasonably close to the conservatively established ductility limit criterion in the structural specification, we conclude that these piers will remain structurally sound during or after a Hosgri event.
The maximum damage that is likely to occur would be some spalling during or after a Hosgri event.
The spalling could cause pieces of concrete to fall into the pump bays of l
the intake structure.
In order for these pieces to reach the auxiliary saltwater pumps, they would have to penetrate one bar rack and a set of l
traveling screens and travel approximately 70 feet into the auxiliary l
saltwater pump bays. We consider this adequate protection against concrete pieces of sufficient size to damage the pumps travelling the above route and reaching the pump intake bells (an additional upward _ motion of 10 feet i
or more).
In addition, there would be insufficient material resulting from spalling to block water flow through the intake structure traveling screens.
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- s. Turbine Buf1 ding Analysis The turbine building was not originally designated a seismic Category I structure.
However, it contains some seismic Category I equipment, principally the emergency diesel generators, associated vital switchgear, and the component cooling water heat exchanger.
Early in the operating license review, in response to our request for assurance that this equipment would be protected from earthquakcs, the cpplicant adopted the approach of performing dynamic seismic analyses for the entire structure as would be done for a Category I structure.
The same approach has been followed in the seismic reevaluation.
It is worth noting that, since this structure was not originally designated seismic Category I, it initially met design earthquake criteria rather than double design earthquake criteria.
This explains, at least in part, why the turbine building now appears to require more extensive modification than the other major structures.
The turbine building is a combined steel frame and concrete structure in which a combination of vertical X-bracing and reinforced concrete shear walls provides lateral force resistance.
Unit No. 1 is a long rectangular building approximately 400 feet long in the north-south (longitudinal) direction, and 140 feet wide in the east-west (transverse) direction.
Unit No. 2 is similar.
The building consists of four working floor levels, approximately at elevations 140 feet,119 feet,104 feet and
'9
, 85 feet, with grade being located at elevation 85 feet.
The massive reinforced concrete turbine pedestal, which supports the turbine generator, is located in the center of the building.
This pedestal has been structurally isolated from the floors at elevations 140 feet, 119 feet and 104 feet, but it does share a common foundation mat with the building.
Two 115-ton capacity overhead cranes will operate at elevation 180 feet.
The roof is supported by trusses spanning approximately 138 feet These are connected by moment-resisting cornections to 40-inch deep welded plate columns to fone rigid bents in the 4est west direction.
The various bents are tied together at the roof level with the lower-chord bracing system, and at elevation 140 feet by the floor framing and concrete diaphragm slab.
The welded plate columns of each bent extend from approximately elevation 85 feet up to the roof trusses.
The north-south, east-west, and vertical analyses of the turbine building were decoupled to facilitate the analysis of the structurally complex l
building.
i The models used for analyses were assumed fixed at elevation 85 feet.
The modal responses were combined in each direction using the square-root-of-the-sum-of-the-squares method.
Seismic forces in the east-west direction are resisted primarily by the diaphragm slab at elevation 140 feet, and by i
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The concrete shear walls run from grade to the diaphragm slab at elevation 140 feet and are 24 inches and 20 inches thick, respectively.
In addition to the shear walls, east-west lateral force-resisting elements are in the form of a shear wall from grade to elevation 104 feet, with vertical X-bracing running from elevation 104 feet to the elevation of the truss lower-chord bracing system.
The interior framing supporting the floors at elevations 104 feet,119 feet and 140 feet is supported on columns running from the foundation to the structural steel floor framing at elevation 140 feet.
These columns are frared to the floor beams and girders with essentially simple connections und are, consequently, assumed to offer no resistance to seismic forces.
Time-history analysis was used to generate floor response spectra.
Response
spectra analyses were used to cocpute structural responses.
The analysis in the east-west direction was done using the SAP IV computer program.
A three-dimensional model was necessary because the primary structural resisting system and the mass of the building are unsymmetrical.
The turbine pedestal was not included in the east-west turbine building model because it is not attached to the building at any other point other than the foundation.
A north-south model was generated and the modes and mode shapes and frequencies determined using the TABS computer program.
This
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. information was used as input to the MATRAN computer program to develop the acceleration time history at various floor elevations.
The vertical time-history analysis was performed.
The model consisting of floor beams, girders, columns and equipment weights was analyzed using the SAP IV computer program.
With regard to the computer codes mentioned above, we have required that the applicant provide verification for those codes which either have been modified by the applicant or which are not generally accepted and verified codes in the public domain.
As a result of the seismic analyses and evaluations, several major modifications to the existing structural system were found to be necessary.
Descriptions of each of the more significant modifications are listed below:
(1) New exterior concrete buttresses with horizontal diaphragms and a new interior concrete shear wall below elevation 119 feet will be constructed.
(2) The existing north-south and east west concrete shear walls below elevation 104 feet were strengthened.
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(3) The existing floor gratings at elevations 104 feet and 119 feet have been substantially replaced by steel checkered plate.
(4) Floor framing members at elevations 104 feet and 119 feet have been strengthened.
(5) New concrete shear walls will be constructed along the east and west sides of the building between elevations 104 feet and 140 feet.
(6) New vertical steel bracing has been added along the east and west sides of the building between elevations 140 feet and the roof.
(7) New horizontal steel bracing has been added in the lower chord of the roof trusses of the building.
(8) The connections of the vertical and horizontal bracing member joints have been strengthened.
(9) The horizontal crane rail support has been strengthened.
(10) The exterior plate girder columns have been strengthened.
Based on our review of the applicant's analyses and our assessment of the existing turbine building structure, we conclude that these modifications,
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, in conjunction with necessary crane load restrictions discussed below, will enable all safety-related equipment in the turbine building to remain functional during and after the Hosgri event.
At our request, the applicant has performed an analysis to assess whether there is enough clearance between the turbine pedestal and the adjacent structural elements for a Hosgri event to prevent pounding.
The analysis indicated that a larger clearance was necessary.
The applicant has been required to take necessary remedial actions (by removing concrete to enlarge the gap and stabilizing the turbine pedestal with anchors).
The applicant found that the turbine building end bents (structural steel members forming the end of the building) would be stressed beyond the yield point if the cranes were located at the end of their travel coincident with a Hosgri event.
As an interim alternative, the applicant has chosen to impose administritive controls to prevent the cranes from being in that location until such time as a definitive analysis of crane' safety had been reviewed and approved by the staff.
We have reviewed the turbine building with the cranes restricted under the proposed administrative controls and we conclude that this operational restriction will prevent any stressing of the end bents beyond the yield point, thus assuring that safety-related equipment in th: turbine buildig would not be jeopardized by excessive deformation of the turbine building structure.
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$ Buried Pipe Analysis The buried diesel fuel oil pipes connect the fuel oil tanks to the diesel generator systems inside the turbine building.
The buried auxiliary saltwater pipes are anchored to the circulating water intake conduits at 40-foot intervals.
In order to assure the structural integrity of the conduits during earthquake motions, they were reviewed for the same i
criteria as the auxiliary saltwater piping and the diesel fuel oil piping.
The portion of a buried pipe far from the ends, and free of any external support other than the surrounding soil, was assumed to move with the
. ground under the propagation of seismic shear and compressional waves.
With this assumption the stresses in the pipe were computed as the products of soil strains and the modulus of elasticity of the pipe material.
For the circulating water conduits, the concrete was assumed to crack and the rebars were assumed to carry the entire tension load.
The axial and bending stresses due to propagation of.a shear wave were calculated in the following manner:
(1) The stresses of the pipe due to the horizontal motion were combined with the stresses in the same direction due to the vertical seismic component, which was assumed equal to two-thirds of the horizontal component, by the square-root-of-the-sum-of-the-squares method.
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s (2) The resulting stress was added to the stress due to the internal pressure.
The maximum stresses induced in the auxiliary saltwater and diesel fuel oil piping were calculated as 20,000 pounds and 36,000 pounds per square inch, respectively.
The maximum stress in the circulating water conduit reinforcing steel was determined as 28,000 pounds per square inch.
These maximum stresses were in all cases less than or equal to the specified minimum yield strengths for the material of construction of 36,000 pounds per square inch.
Isolation sleeves or flexible couplings are used where the pipes enter the buildings to accor.modate relative displacement between the soil and the buildings.
Outdoor Tanks Analyses The applicant has reevaluated two types of outdoor tanks, namely, the outdoor water storage tanks which are resting on the ground, and the fuel oil tanks which are buried in the ground.
The outdoor water storage tanks are cylindrical structures originally fabricated of welded steel plates and anchored to a concrete foundation.
The tanks consist of a dome with a radius of 40 feet and a cylinder of 40-foot diameter and 52.5 feet in height.
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' For the analysir if the water storage tanks, the free-field Blume and Newmark horizontal response spectra at four percent critical damping were used.
The criteria employed in the seismic evaluation of the outdoor' water storage tanks for the Hosgri event were similar to those for the major structures discussed earlier.
A lumped mass-spring model fixed at the base was initially used for analysis.
We did not accept that analysis because the tanks had not been analyzed as a thin shell, which would include ovaling effects in the tank walls.
At our request the applicant has performed another analysis using an axisymmetrical finite element model for the stress analysis.
The hydrodynamic pressure exerted in the tank wall was determined following the procedures recommended in Veletsos and Yang, 1976 (Reference 7 of Appendix D to Supplement No. 7).
The paper suggested the procedures for computing the hydrodynamic effects in rigid tanks in terms of impulsive and convective pressures.
The applicant also compared the results from this approach with those obtained by the approach described in Chapter 6 of the Atomic Energy Commission publication TID 7024 (Reference 8, Appendix 0 of Supplement No. 7).
The applicant has decided to add concrete shells, typically 8 inches thick, surrounding the steel tanks; this will strengthen the tanks against j
ovaling.
In addition, the applicant is removing all fill beneath the tanks down to rock and replacing that-fill with concrete and adding s.
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19 rock anchors around the tanks, thus assuring the stability of the tanks under seismic loading.
There are two fuel oil tanks buried approximately 7 feet below the ground surface.
These tanks are 10 feet in diameter and 36 feet long.
They are resting on compacted soil.
For the analysis of fuel oil tanks, the applicant has used the free-field Newmark horizontal res :snse spectra at five percent of critical damping and the corresponding time-history acceleration coupled with deconvolution analysis to generate the inputs to a finite element model.
A finite element analysis was performed for the fuel oil tanks using the FLUSH program.
This program is based on the vertical shear wave propagation theory and has been used in many seismic analyses and is generally accepted by the engineering profession for use as applied here.
Cranes The applicant has found that some cranes can withstand the loads from a l
Hosgri event only if restrictions on the loads to be carried by these cranes are imposed:
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. (1) The polar gantry crane in the containment structure.
The applicant's analysis indicated that the polar gantry crane could safely withstand c Hosgri event in its parked position, but would be subject to tipping in its unlocked position and under operation.
The polar gantry cranes inside containment will be parked and seismically locked for all modes of plant operation except mode 5 (cold shutdown) and mode 6 (refueling).
In this condition, the analyses indicate no overstressing or instability for a Hosgri event.
For the unlocked condition, the analyses to date have not conclusively demonstrated stability or lack of overstress.
Further analyses are underway.
The applicant has provided an analysis of the consequences of assumed damage to safety-related equipment due to overturning of l
the crane in order to demonstrate acceptability.
This analysis takes l
l credit for the plant being shut down and cooled down.
We would have I
to review such an analysis in detail before accepting it and there has not been time to perform such a review.
Accordingly, we will require that the applicant modify the crane to remain intact and stable, both loaded and m acc cd. during a Hosgri event or, alter-nately, demonstrate,n :p"c 3.e our satisfaction that assumed crane i.
failure would not result in unacceptable damage to essential equipment.
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'l We will require that one of the above conditions be met prior to plant operation.
(2) The cask handling crane in the fuel handling building.
The fuel handling building crane is rated for 125 tons on the main hook (for_ handling spent fuel shipping casks) and for 15 tons on thc auxiliary hook.
The analyses to date have indicated no overstressing or instability for a Hosgri event with loads up to 15 tons on either hook.
Analyses for larger loads on the main book are underway.
The applicant proposes to restrict the loads to 15 tons or less until the crane can be shown safe for larger loads.
We will impose such restrict' ions as a condition of the operating license until this has been demonstrated to our satisfaction.
(3) The two turbine building cranes.
1 The turbine building cranes are rated for 115 tons.
Analyses to date have indicated that the cranes themselves are adequate for 100 tons i
with a Hosgri event.
However, the analyses to date have only demon-l strated the capabilities of the supporting building columns for a crane load of 15 tons during a Hosgri event.
Further analyses are underway.
The applicant proposed restrictions against lifting loads greater than 15 tons over safety-related equipment until it has been l
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We will require as a condition of the operating license a somewhat more severe restriction--that the crane (rather than only the load) not be over safety-related equipment when lifting loads in excess of 15 tons until it is demonstrated to our satisfaction that the crane is capable of carrying greater loads during a Hosgri event. We will also require the use of mechanical stops (which can be removed when appropriate) to enforce the restriction.
In addition to the above restrictions, we will also place restrictions on the turbine building cranes in relation to turbine missile risks (Section 3.2.1 of Supplement No. 8 to the Safety Evaluation Report) and in relation to the turbine building end bent analysis discussed earlier in this testimony.
Restriction on the maximum loads to be handled, and on travel of the cranes when loaded, is therefore required for the cask handling crane and the turbine building cranes.
Similarly, a safety analysis considering tipping of the containment gantry crane during refueling is also required.
Methods of Analysis for Mechanical System and Come nenus The methods used in the Hosgri event reevaluation were common to presently accepted seismic analyses with the exception of the items listed below.
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o In some cases, where material test data were available, actual material properties were used in lieu of code specified minimum properties to establish allowable stress limits to justify structural integrity.
Allowable stress values were established using the bases prescribed by Appendix III of Section III of the ASME Code so that the factors of safety used in the code are preserved (for example, if normal practice called for use of two-thirds of the code specified minimum yield stress to be used as a limit, the DCNGS review was based on two-thirds of the actual material yield stress).
For this reason, we consider the use of actual material properties acceptable for the reevaluation.
o The responses to Hosgri earthquake loads or the double design earth-quake loads (whichever was more limiting) were combined with the response due to normal operation and the response due to postulated loss of-coolant accident loads using the absolute summation method for response combination.
This is a conservative procedure which results in the reactor coolant system being designed for loads well in excess of those calculated for a seismic event alone without a pipe break.
Even though the assumed seismic event is not expected to cause a pipe break in a seismically designed piping system, these loads are combined for design purposes to produce extra margin.
A further conservative element is the requirement that the
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. peak system responses (which occur at random times and last only milliseconds) be combined on an absolute sum basis as if all responses occurred at exactly the same moment.
o A one quarter scale model structural test was performed on the reactor vessel shoe and pad system to determine the load-carrying capacity of the assembly rather than simply using code allowable stresses.
The allowable load was limited to 80 percent of the ultimate load obtained from the test.
This follows the practice permitted by Appendices II and F of Section III of the ASME Code.
o In the reevaluation, low amplitude shock or vibration testing of systems and components as they are actually installed (in situ testing) has been performed to experimentally validate the natural frequencies, mode shapes and damping values used in seismic analysis.
This was done for selected components and supports such as tanks, heat exchangers, valves, piping systems and supports.
Where significant differences were found, the analyses were revised to correspond to knowledge gained in the tests.
Evaluation of Methods for Mechanical Systems and Cocoonents As noted in the earlier portions of this testimony, several conservatisms are inherent in the usual procedures for the seismic design of nuclear i
- power plant mechanical systems and components.
The general methods used by the applicant in this reevaluation as described above contain few variations frcm the usual procedures as defined by our usual criteria.
One is the higher damping value allowed for the reactor coolant loop analysis.
As discussed above, this is based on tests and is normally acceptable for Westinghouse reactor coolant systems provided that similarity with the system that was tested is demonstrated.
Another is the use of actual material test strengths.
As discussed above, the code safety factors have been retained in using these material test data to establish allowable stress levels.
The in situ testing program represents a further improvement, relative to the normal case, in our knowledge of the plant's seismic capabilities.
In our review we have found that in the individual steps where there are variations from the usual procedures, those individual steps have remained conservative and have retained adequate safety margins.
In the remainder of the analysis, the usual conservative elements apply.
Floor Resoonse Soectra Floor response spectra characterize the seismic inputs to mechanical systems and equipment at various points in the structures.
They were develeped in the structural analysis as discussed earlier in this testimony.
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" We have reviewed the methods used by the applicant to obtain from the floor response spectra the appropriate response spectra to be used in the analysis of particular piping systems and components.
The floor response spectra generated from the Blume and Newmark ground motion inputs were enveloped to obtain the actual floor response spectra used in the analyses.
The spcctral accelerations were obtained from the enveloped rotational floor response spectra added to the accelerations from the enveloped horizontal floor response spectra.
As required by the routing of a piping system, linear interpolation was used to generate the spectra for supports or anchors located between floors.
The different response spectra at the corresponding support or anchor elevations were then enveloped to obtain the appropriate response spectra for the analysis.
At the criteria implementation review meetings, we reviewed these procedures as used in actual problems and found that these methods constitute a conservative manner of utilizing the calculated floor motions for the design of mechanical systems and components.
Resoonse Combinations l
The response spectrum modal superposition method used by the applicant to combine the responses from horizontal and vertical earthquake components was the =cthod developed and in general use before the first issuance of Regulatory Guide 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis," December 1974.
The method used by the
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, applicant provides for absolute summation of the response due to one horizontal component of excitation and the response due to the vertical component of excitation for each piping system frequency.
All modal responses at various frequencies were then combined by the square-root-of-the-sum-of-the-squares rule to obtain the total response.
This process was then repeated for the other horizontal component and the vertical earthquake component and the controlling value selected for design.
One of the procedures contained in Regulatory Guide 1.92 (Regulatory Guide 1.92 method) consists generally of adding responses due to the two horizontal components of excitation and the one vertical component of excitation by the square-root-of-the-sum-of-the-squares rule -for each mode (frequency), calculating each of the three components independently.
Responses for the various modes are then combined by the square-root-of-the-sum-of-the-squares rule to obtain the total response.
Appropriate adjustments are made for modes that are closely spaced in frequency.
The applicant's method and the Regulatory Guide 1.92 method contain different types of conservative elements.
The applicant's method consists of using the absolute sum of the responses to a horizontal excitation and a vertical excitation.
The Regulatory Guide 1.92 method consists of combining the responses to both horizontal components and the vertical component at the same time.
As a result of the different approaches the Regulatory Guide 1.92 method gives more conservative results in some le
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Thus, neither method is universally more conservative than the other.
Reactor Coolant System Main Looos The applicant has performed detailed structural analyses for the reactor coolcnt system for the loads induced by a loss-of-coolant accident resulting frcm postulated pipe ruptures.
Included in the analyses were postulated pipe breaks that produced the most limiting loads on (1) the reactor coolant piping systems including nozzles, (2) the reactor vessel, vessel internals and vessel supports, (3) the steam generators and supports, (4) the pressurizer and supports, and (5) the reactor coolant pumps and supports.
The analyses included the effect of modifications to the plant which reduce the severity of the postulated reactor vessel nozzle break by the addition of pipe displacement restraints in each primary shield wall pipe annulus.
All loads acting on the system were included in the analyses.
Among these are reaction loads from the pipe rupture, asymmetric loads due to decompres-sion waves acting on the reactor internals (for the reactor vessel nozzle break) and the external asymmetric subcompartment pressure loads.
The dynamic model for seismic analysis of the reactor coolant system included the four reactor coolant loops, their respective steam generators and reactor coolant pumps and their supports, the reactor vessel and supports,
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, and the r. actor vessel internals.
The influence of the main steam lines was included by applying the effective stiffness at each of the four steam nozzles.
The response spectrum model superposition method was used for the analysis.
Hosgri response spectra and a damping value of 4 percent were used in the evaluations.
The reactor coolant loops (unbroken portions) were analyzed for the combined loads.
The loads on the components such as the piping, the reactor pressure vessel including its internals and the control rod drive mechanisms, the reactor coolant pumps and the steam generators and their supports were evaluated and the resulting stresses found to be acceptable.
In performing these analyses, the applicant evaluated the adequacy of the Diablo Canyon reactor coolant system considering several load combination assumptions as listed below:
(1) The seismic event occurring alone (normal + seismic).
(2) The loss-of-coolant accident occurring alone (normal + LOCA).
(3) The seismic event and loss-of-coolant accident occurring simultaneously with the peak loads combined by the square-root-of-the-sum-of-the-squares rule (normal + [(seismic)2 + (LOCA)2 1/2),
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. (4) The seismic and loss-of-coolant accident events occurring simultaneously with the peak loads combined by absolute summation (normal + seismic + LOCA).
At our request, as indicated by item (4) above, this included absolute summation of normal loads, peak seismic loads (whichever was more limiting between the double design earthquake loads and the Hosgri event loads),
and peak pipe break loads.
This was the most conservative of the load combinations considered.
The applicant's evaluation of the effects of the most conservative combination indicated that all components and supports of the primary coolant system are capable of withstanding the effects of the simultaneous occurrence of normal operating loads, peak seismic loads, and peak loads from a postulated loss-of-coolant accident.
Reactor Coolant System Branch Piping This section discusses branch piping connected to the reactor coolant system main loops.
The piping under discussion runs from the reactor coolant loops to the first piping anchor since the branch piping must be analyzed to the first anchor.
It often extends beyond the reactor coolant system boundary, which is defined by isolation valves rather than piping anchors.
The reactor coolant loop branch piping was examined for the faulted condition loaaing combination using the Hosgri event and a postulated loss-of-coolant s.
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accident occurring simultaneously with the peak loads combined by absolute summation.
Dynamic time history analyses were performed for five branch lines including the pressurizer surge line.
The systems chosen have dynamic restraints close to the loop connection and therefore are likely to be the most highly stressed branch lines.
The seismic analyses were performed using the analysis method the applicant has been using for all the piping systems.
The results of the combined Hosgri plus loss-of-coolant accident analysis indicated that the piping stresses were well within the allowab'.es and the supports would be adequate for the combined loads.
Sucoort Stiffness The applicant had assumed rigid piping system supports rather than modeling the actual stiffness of the supports.
Modeling the actual stiffness should provide more accurate results.
We requested a sensitivity study to verify the acceptability of the rigid support assumption.
This study consisted of performing Hosgri event seismic analyses of several representa-tive piping systems using actual support stiffnesses.
The results indicated l
l that both the piping stresses and support loads increased at some locations and decreased at other locations.
There were no cases where the increased piping stresses exceeded the allowable stresses.
As discussed below, the increased support loads were also found acceptable.
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. With regard to piping supports, the applicant has evaluated all the piping and supports for spectra associated with the design earthquake, the double design earthquake and the Hosgri event.
Many systems were evaluated twice for the Hosgri event, first using estimated spectra and then again using the final calculated spectra.
During the Hosgri event evaluation supports were added to control the piping for the greater earthquake input.
The addition of these supports resulted in a redistribution of loading during the Hosgri event.
The resulting effect was an overall reduction in seismic loading for certain supports that had been previously qualified and installed to withstand higher seismic loads.
All the supports which had increased loads in this study were qualified for these loads by either qualification for the increased load from a previous analysis, support member stress analysis or by a comparison with the ultimate load capacities from the support manufacturer.
The acceptability of the supports for all the increased loads that were obtained in the sensitivity study has been demonstrated.
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In the applicant's judgment sufficient margin to accommodate seismic anchor motion stresses was available between the thermal expansion stresses and the secondary stress allowables.
At our request the piping systems which would experience the largest seismic anchor motions were analyzed.
This consisted of eight piping 1inas,.The applicant.used.Hosgri event seismic anchor displacements, although current code rules require only an evaluation for anchor displacements caused by an operating basis earthquake (for this plant the design earthquake).
The Hosgri event anchor movements are more limiting than those which would be obtained by using the design earthquake.
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the in:lusica of the stresses from the Hosgri seismic anchor movements, the total piping stresses in these eight lines were still conservatively within the allowable stresses specified in the appropriate piping codes.
With regard to pipe support stresses due to seismic anchor movements, we requested that, where the support loads would clearly produce primary stresses in the supports, such as supports consisting of a member in pure l
tension, the Hosgri event seismic anchor movement loads be included in the faulted plant condition load combination.
This is a more conservative s.
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Using faulted condition allowable stresses such supports were found to be acceptable with seismic anchor movement loads included.
The evaluations required by the current code rules were also performed and the resulting stresses were determined to be acceptable.
Seismic Qualification of Eoulpment The majority of the safety-related electrical instrumentation and control equipment was qualified by testing.
The balance was qualified by analysis or a combination of test and analysis.
The equipment was previously qualified to the level of the double design earthquake or higher.
Where the original qualification did not envelope the required seismic inputs to equipment for the Hosgri event, the applicant has requalified the equipment for the Hosgri required response spectra.
In the requalification process the applicant has, at our request, committed to employ seismic qualification methods that conform to the vioratory input methods of current criteria (Regulatory Guide 1.100, Revision 1,
" Seismic Qualification of Electrical Equipment for Nuclear Power Plants,"
l August 1977, and IEEE Standard 344-1975, "IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations").
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As used here, the term single frequency means testing at a number of different frequencies, one frequency at a time.
Similarly, single axis testing means excitation in one direction at a time.
The current staff position is that multi-frequency, multi-axis testing should generally be performed except in cases where the characteristics of the required input motion indicate that the motion is dominated by one frequency and the anticipated response of the equipment is adequately represented by one mode, or that the single frequency input has sufficient intensity and duration to excite all modes to the required amplitudes such that the test response spectra envelopes the corresponding required response spectra of all the modes.
We have established a seismic qualification review team.
The team has been auditing qualification programs for operating license applications since 1975 in order to assess the adequacy of the vibratory input employed j
in seismic qualifica; ion testing particularly for equipment qualified to earlier standards.
The team visited the Diablo Canyon plant in 1977, inspected selected vital mechanical and electrical equipment as installed, and identified concerns about the adeque y of original seismic qualific-$n methods for some of I
l the items inspected.
For those items where the team expressed concern, the applicant has included the appropriate equipment among the items to be l
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, requalified by testing in accordance with IEEE Standard 344-1975.
This eliminated the team's concern.
The seismic qualification review team also visited Westinghouse Electric Corporation in 1976 to audit qualification methods for equipment supplied as part of the nuclear steam supply system at Diablo Canyon and other plants.
Although no final evaluation has yet been issued, all the equipment supplied as part of the nuclear steam supply system at Diablo Canyon has been reviewed and found acceptable.
Conclusion The staff review of tae seismic design of the DCNGS has been the most extensive we have ever undertaken.
This review has extended from the basic input criteria employed through the details of myriad analyses to the implementation in final design.
Our goal throughout the review has been to assure that demonstrably conservative practices were followed at each level of design.
We believe that this goal has been fulfilled in all aspects of the DCNGS reevaluation, including confirmatory analyses and tests, design of modifications, and the establishment of operating restrictions where necessary.
It is our conclusion, therefore, that the structures, systems and components necessary at the DCNGS to assure the health and safety of the public will remain functional under the loading that would result from any seismic event of severity up to and including that specified for the Hosgri event.
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