ML19282D011
| ML19282D011 | |
| Person / Time | |
|---|---|
| Issue date: | 03/21/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19282D009 | List: |
| References | |
| NUDOCS 7905100035 | |
| Download: ML19282D011 (52) | |
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Statement of Harold R. Denton, Director Office of fiuclear Reactor Regulation U. S. fluclear Regulatory Commission Before the Subcommittee on Energy and Water Development Committee on Appropriations U. S. House of Representatives March 21,1979 s
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Mr. Chairman, and members of the Committee, I appreciate this opportunity to discuss our decision last week to order the shutdown of five nuclear power reactors.
With your permission, I would like to include copies of the Orders to Show Cause, along with other supplemental background information, in the record of this proceeding.
In short, we found that a computer code that had been utilized by the Stone and Webster Engineering Corporation in the seismic design of safety-related piping for these five reactors produced results significantly different from results recently calculated with their current code.
On determining that the capability to shut down the plants safely in the event of an earthquake could not be assured on the basis of the results of a partial reanalysis of the Beaver Valley Power Station, and having no reason to expect that the other four units would turn out any differently, I recommended that operation of the affected plants be suspended.
I consulted with the Commissioners before issuing the order and they concurred in that action.
So that you may better understand the basis for our actions, and their safety significance, I would like to briefly describe the events leading up to my decision and the staff's views on the seis-mic capabilities required in nuclear power plants.
2-In October 1978, Duquesne Light Company, the licensee of the Beaver Valley plant, was informed by Stone and Webster that there were over-stresses in pipe supports associated with Safety Injection System piping.
At the time Beaver Valley was shut down because of main transformer dif ficulties.
Stone and Webster was reanalyzing stresses in connection with a system modification required by the NRC staff.
During this reanalysis effort, the S&W engineers also came across information that had been provided to them by Westinghouse n May 1978 that showed soma check valves in these lines were actually heavier than assumed in the earlier analysis.
Some'ime during this reanalysis, either in connection with the planned modifications or in reexamining the effect of the increased valve weights, S&W found some instances of overstresses in two lines.
They In corrected this by adding a snubber and modifying one support.
doing the analysis related to making this correction, S&W used at A new one called NUPIPE predicted least two computer programs.
much higher stresses than the one used during a 1974 as-built check of these lines.
That older code is called PIPESTRESS.
On October 26, 1978 the licensee orally notified our Office of Inspec-tion and Enforcement about the design error which required correction.
c As the Office of Inspection and Enforcement delved into the matter, they became concerned about these large differences in results between the NJPIPE and PIPESTRESS codes even though both codes showed the orioinal design error would be corrected with the modification to be r:ade by the licensee.
On December 6,1978 the licensee submitted additional information in a Licensee Event Report.
In February, the Office of Nuclear Reactor Regulation initiated a technical effort to assist Inspection and Enforcement in resolving the matter.
At this time it did not appear to be a concern other than an error in application of the codes.
Phone discussions with the licensee and S&W to identify specific dif ferences were not effective since without the actual computer runs to look at there was a communications problem.
During a meeting held on March 8,1979 to discuss these n1+.ters, the Beaver Valley licensee informed the staff that the difference in predicted piping stresr,es between the two computer codes was attributable to the fact that the PIPESTRESS code uses an algebraic summation of the loads calculated separately for the horizontal and the vertical component of earth-quake motion.
A detailed chronology of events is given in Enclosure 1.
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The algebraic technique used for. combining the intramodal responses is not conservative for response spectrum modal an. lyses because time dependency phasing is not considered.
Using the algebraic summation technique could in some cases result in a zero intramodal component The techniques in calculating the overall response to an earthquake.
required by the staff for combining intramodal responses that do not include phasing of the dynamic components of an earthquake are either the square root of the sum of the squares or the absolute summation of the response components.
The analytical treatment of load combinations becomes significant because horizontal earthquake motions can produce piping movement in both the horizontal and vertical direction and the vertical earth-quake motions can also produce piping movement in both horizontal and vertical directions.
For some designs the calculated piping stresses may differ significantly depending on the load summation techniques used in each rode of response.
A more detailed discussion of our seismic design requirements, and the margins of safety in a properly designed system is given in Enclosure 2.
Based on the three systems tha,t had been reanalyzed by the newer code on Beaver Valley at the time of the March 8,1979 meeting, stresses over allowable values were expected to be found primarily in piping supports although significant increases in piping stresses had been Based on.the information received on March 8 and 9,1979 observ ed.
_i-I decided to send reviewers to Stone and Webster's Boston office to determine the extent of this problem on Beaver Valley 1 and other potentially affected plants.
Concurrently, the nuclear system suppliers fc-the plants (Westinghouse, General Electric and Combus-tion Engineering) were contacted to determine if their codes for pipe stress analyses during that period used the algebraic approach.
In following the course of the reanalysis at the S&W offices over the weekend of March 10,11 and ' 2, it became apparent that a number of piping systems had calculated stresses over the allowable value for the design basis earthquake.
Also, for a few of these systems the more probable operating basis earthquake resulted in stresses above the allowable value.
In addition, the structural integrity and functionability of pumps, valves and other essential equipment could be affected.
The eastern United States is generally believed to be a region of low seismicity, when it is compared with the western part of the country.
It is not, however, without significant histori-cal seismic activity.
The recurrance interval of the operating basis earthquak9 for these facilities is on the order of 200-400 years.
Detailed information on the historical record of earthquakes for the four plant sites is given in Enclosure 3.
Although results were still incomplete by Monday morning, analyses of a significant fraction of the affected piping system indicated that high stresses were cal-culated in a number of systems important to safety.
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Because the overstressing of pip ~ing and supports was predicted even for earthquakes which could reasonably be expected to occur during the lifetimes of these facilities, the problem took on considerable safety significance.
Some of the systems identified as having over-stressed conditions under earthquake loadings were part of the reactor coolant pressure boundar), whose failure could cause a loss of coolant accident.
In addition, systens which would be needed to shut the plant down safely in the event of a loss of coolant accident were also affected.
Thus an earthquake, of not extremely low likelihood, would have the potential both for causing an accident and.for pre-venting safety systems, designed to cope with that accident, from operating.
A secondary concern was whether or not systems needed to provide adequate long term cooling for the plant in the event of an earthquake without a LOCA could be assured.
It was this
" common mode" effect that gave me greatest concern for the safety of continued operation of these plants.
Concurrent with our Beaver Valley review, we reviewed our records to determine whether or not other facilities had used these same analysis techniques.
Based on the review of their records and in-formation provided by Stone and Webster we concluded that four other facilities used the same techniques.
The four facilities are Maine Yankee, Fitzpatrick and Surry Units 1 and 2.
The systems which were analyzed using these techniques perform sub-stantially the same safety functions as those in question at Beaver Vall ey.
The staff concluded the potential for serious adverse effects in the event of an earthquake was sufficiently widespread that the basic defense in depth provided by redundant safety systems may be compromised.
I subsequently concluded that the public health and safety required that the affected facilities be placed in a cold shutdown condition pending further order of the Comission.
The Orders provide that within 20 days each licensee must respond with respect to:
(1 ) why the licensee should not reanalyze the facility piping systems for seismic loads on the piping system and any other affected safety systems using an appropriate piping analysis computer code which does not conbine loads algebraically, (2 ) why the 1icenses should not make any modifications to the facility piping systems indicated by the reanalysis, and (3) why facility operation should not continue to be suspended until completion of the reanalysis and any r equired modifi-cations.
All of tne plants are now in a cold shutdown condition.
(Surry Unit I
2 was already in an extended outage for steam generator replacement).
The staff met with Stone and Webster and a number of the utilities involved on March 16th and 17th to receive an up-to-date status on actions being taken by the utilities and Stone and Webster.
Actions taken or being taken are:
Identification of all safety related systems that have been analyzed with a piping comput r code involving a program deficiency is nearing completion.
Computer input are being checked to assure that all reanalyzed piping will reflect the as-built condition at each plant.
Piping analyse. are being rerun and piping and supports ex-ceeding allowable stresses will be identified.
We understand that for those systems which exceed allowable stresses certain refinements in calculational techniques may be proposed to justify, in some cases, not modifying the facility design.
Whether such refinements will be acceptable to justify interim or long-term operation is yet to be determined.
In addition, the staff is presently involved in a series of independent actions to assure that the re-analysis of piping systems is conducted in an acceptable manner.
These actions are as follows:
Piping stress computer' codes to be used for reanalysis of the
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piping will be run with NRC established benchmark problems.
A description of current methods used by tha staff to indepen-dently verify such computer codes is given in Enclosure 4.
An independent audit of selected piping runs will be conducted by NRC staff consultants to verify the piping stress reanalysis.
The Union of Concerned Scientists has charged that NRC knew of the deficiency in the Stone and Webster stress analysis method for five years and did nothing about it.
To the best of our knowledge the charge is untrue.
On the contrary, the staff did not learn of the algebraic summation of intramodal stress components in the Stone and Webster PIPESTRESS code until March,1979.
The UCS has supplied no documents to support their charge, but they apparently derive the five year period from the 1974 date of first publicat'on of Regulatory Guide 1.92.
The Guide describes acceptable methods for combining modal responses and spatial components for three dimensional earthquake analysis. While not directly applicable to the analyses done earlier by Stone and Webster for the plants in question, the deficient Stone and Webster method is not acceptable under the requirements of Regulatory Guide 1.92.
But contrary to the implications of the UCS charge, the method used by Stone and
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Webster was equally unacceptable prior to the issuance of Regulatory Guide 1.92.
The staff and its consultants in the period from 1967 to 1974 are agreed in their recollection that the proven and generally accepted engineering practice for the combination of intramodal stress components has always been either the absolute or the SRSS method.
That is, in their judgement, an algebraic summation of intramodal responses in connection with a spectrum type of dynamic analysis would have been recognized, either in that time period or today, as erroneous.
This completes my prepared testimony.
I would be glad to answer any questions.
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CHRONOLOGY TABLE PIPE STRESS ANALYSIS ISSUE Note:
This chronology represents information available to the NRC as of March 17, 1979.
It does not necessarily fully or accurately reflect the actual sequence of events which occurred prior to the March 8, 1979 r.eeting.
10/2/78 Stone and Webster notified the Beaver Valley Unit 1 Station Manager of an error discovered in the original hand calculated stress analysis of some safety injection lines.
The error was discovered while evaluating the impact of correcting the weight en 14 safety injection system check valves. Since this was, technically, a deviation from the Final Safety Analysis Report, it was to be referred to the Station Safety Committee.
To that end, the Station Superintendent asked for more specific information on the error *.
10/13/78 At a meeting held at the Beaver Valley site between Duquesne Light Company and Stone and Webster repre-sentatives, additional information on the error was provided but more specifics were requested by DLC*.
10/23/78 Stone and Webster provided DLC more infonnation.
The Station Superintendent asked for additional clarification and was told Stoae and Webster personnel would be at the site the naxt week *.
10/26/78 During a site visit, Stone and Webster informed DLC that one safety injection line would actually be significantly overstressed.
DLC then made a prompt telephone notification to Region I of the Office of Inspection and Enforcement *.
- Tnese entries provided from memory by Duquesne Light Company representative on March 17, 1979.
~ 10/26/78 Prompt report LER 78-053/0lP to NRC Region I via telecon from I. 'quesne Light Company.
Reported information received from Stone and Webster that hand calculation errors resulted in ' stress levels above ANSIB 31.1, 1967 but only in one case of six flow paths.
10/27/78 Daily Report by Region I to I&E headquarters included as a reportable occurrence - inadequate piping supports during review of safety injection pipe stress analysis by the A/E (S&W), several points on the 6-inch ana smaller piping were found to be inadequately supported. In the event of safety injection system-operation during a DBE, 5 points could exceed the code allowable stress.
A design change for safety injection piping supports will be accomplished prior to unit startup in mid-November.
10/27/78 Written interim LER submitted by Duquesne Light Company.
DLC characterized the errors reported by Stone and Webster as resulting from a hand calculation method of analysis.
10/31-11/3/78 IE Inspection 50-334/78 Region I followup on 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report.
Inspector raised a number of questions including: What assurance can be given to show that the calculational error applies only to the six points in question? To only the Safety Injection system? To only the Beaver Valley facility?
11/9/78 Second interim LER submitted by Duquesne Light Company indicates that the original report was erroneous. The line stresses were thought to have been hand calculated only, when in fact they were subsequently computer calculated and found acceptable.
DLC also indicated that a full report on the situation was in preparation by Stone and Webster.
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11/14-17/78 IE Inspection 50-334/78 Region I inspectors followup but no information available onsite.
11/16/78 Region I Daily Report indicated a rereview by A/E found that the previously reported condition was erroneous and that no inadequately supported piping existed, a full report of the situation is being prepared by the A/E and a followup to the LER will be submitted by the Licensee to NRC.
11/30/78 Followup calls to site by the IE inspector attempting to seek additional information.
12/01/78 Followup calls to site by the IE inspector attempting to seek additional information.
12/04/78 Followup calls to site by the IE inspector attempting to seek additional information.
12/05/78 Followup calls to site by the IE inspector attempting to seek additional information.
12/06/78 LER 78-53/0lT-0 was submitted to NRC by licensee.
Conclusion was that " corrective action has been reviewed, approved and satisfactorily completed".
The report based on information supplied by Stone and Webster attributes the pipe overstress to diff-erences between stresses analyzed by PSTRESS code and those done by the chart method.
It mentions differences between PSTRESS and NUPIPE codes in force summation but does not elaborate on them.
It concludes that PSTRESS used methods acceptable for Beaver Valley Unit 1 generation plants.
It states that Reg. Guide 1.92 issued in December 1974 established for facilities docketed after April 1975 more conservative techniques for intra-modal combinations of generalized loadings. The report states that analysis showed that only one safety injection system pipe required modification -
the addition of one snubber and the redesign of one support. The attachment to this LER provided additional historical information as follows:
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Duquesne Light Company reported in an attachment to the December 6, 1978 LER 78-53/01T-0 that to generate data needed for installation of a net positive suction head modification to the Beaver Valley Unit i safety injection system, they (Stone and Webster) decided to " code in" the six inch SI lines into a currently used computer program (NUPIPE).
DLC indicated original design used the PSTRESS code.
No results of an analysis at this stage were reported by DLC to NRC.
Subsequent to the above activity the attachment states the Beaver Valley Power Station was notified by a vendor that check valves in SI system were actually heavier than used in design at construction stage.
This increased weight was used as input to the above NUPIPE model and found not to " affect" the piping design.
The Architect Engineer (Stone and Webster) also concluded that the hanger designs need not be changed as a result of using the correct (heavier) weight for these valves. However errors were said to have been discovered in the hand calcu-lation method. It was determined that piping analysis showed local overstress at several anchors but no overstress in "the pipe" alone.
Per atthchment to LER 78-53/0lT-0, a more thorough evaluation was initiated to determine if "any other annulus piping" originally designed by the chart (hand calculation) method was overstressed.
Per attachment to LER 78-53/01T-0, licensee found that SI lines had been "as-built" reviewed in 1974 and that two of ' e six lines had been (at that time) coded into PS1RESS (not just hand calculation method).
The PSTRESS code was re-run using the correct valve weights and resulted in acceptable pipe stresses.
Also per attachment to LER-78-53/0lT-0, licensee states "The models run in PSTRESS and NUPIPE are geometrically similar; however, the mass distribution and support stiffness are different.
Further, the method of force summation (intra-modal) is different.
NUPIPE utilizes more conservative techniques for intra-modal combinations of generalized loadings.
These newer techniques arose following establishment of Beaver Valley Unit No. I design criteria.
In December, 1974, the USNRC published Regulatory Guide 1.92, applicable to facilities docketed after April,1975, which required the use of the more conservative combinations.
The PSTRESS methods used were accepted dynamic analysis techniques for Beaver Valley Unit 1 generation plants, and is the basis for all computerized Category I pipe stress analysis performed".
(It is NRC understanding that results were unsatisfactory on two of three lines, but snubber and support modifications on one line reduced the overstress on the second line such that no modifications on that line were necessary.)
The pre December 6,1978 review of annulus seismic piping was limited to lines that had been previously analyzed using the hand calculation method (2-1/2 inch to 6 inch lines).
103 lines were identified, 55 were reviewed and found acceptable.
Licensee noted that PSTRESS results were still available for 48 of the 103 lines from the 1974 as built review and were " acceptable".
Licensee notes its Engineering Department is " continuing a review of the architect-engineer findings".
12/11/78 Followup calls to site by the IE inspector to seek additional information.
Region I IE inspector telephoned NRR Licensing Project Manager to obtain a contact for informal discussion of technical questions.
12/12/78 Region I Daily Report - Further review of in-containment SI system piping supports identified one line requiring support modification, attributed to an error in original design calculations.
12/14/78 Regional inspector was telephoned by NRR individual who was designated as contact.
Preliminary technical discussion was held about potential problems.
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12/18-20/78 IE Inspection 50-334/78 Region i followup on 12/6 LER.
During this insp2ction, the inspector reviewed the detailed report submitted to the licensee by A/E and discussed the resu~.ts of that review with representatives of the licensee and A/E.
12/22/78 Region I inspector discussed with NRR individuals via telephone questions he had as a result of discussions he had with S&W on 12/18-20/78.
The NRC indivisivals involved determined that there was a possible problem.
1/18/79 Legion I mailed to IE Headquarters a memorandum requesting that information be forwarded to NRR for review. The memo defined concerns t-include:
1.
Reconciliation of the differing as ' ysis results to assure that the design methods used are neither incorrect nor unconservative.
2.
The need f or further licensee review of piping potentially affected by any incorrect or nonconservative calculation.
1/23/79 The IE Inspector provided copy of the 01/18/79 memorandum to Licensing Project Manager.
About 2/2/79 Discussion between IE inspector and NRR project manager determined that a formal transfer of lead responsibility between I&E and NRR had not been made of the 01/18/79 memorandum to NRR.
2/2/79 A formal request for DOR's Engineering Branch support (TAC form) was prepared by the project manager.
2/5/79 IE inspector was informed by IE:HQ that telephone discussion had established that NRR was working on the problem and that a formal transfer of lead to NRR would be made.
3/i/79 During a conference call to DLC and S&W, a. computer run was requested for 00R review.
Since S&W corporate policy was not to provide such proprietary data, a meeting was set up for S&W to bring in a computer run for DDR review at Bethesda.
3/8/79 A technical mee*ing was held between DLC, S&W, and the NRC staff to discuss and review the PIPESTRESS and NUPIPE codes. The NRC approached the review with the belief that the two codes were acceptable and that some modeling or input problem created the results in question.
It was revealed that the PIPESTRESS code used an algebraic summation of seismic loads which in the absence of a detailed time history analysis, gave unconservative results in the seisnic stresses.
Management was immediately informed and a management level meeting arranged with DLC and S&W.
3/8/79 A management level meeting was held with DLC and S&W to arrange for immediate revie6 of the Beaver Valley pipe stress analyses.
Commitments were requested of S&W to identify the systems and plants involved, the inadequacies expected and tne reanalysis to confirm safe operation.
No definitive information was available at that time.
DLC was requested to ' 've its plant safety committee review the situation.
3/9/79 Numerous staff meetings were held at Bethesda to scope the problem with respect to the effects if a seismic event were to occur. Telecons were made to S&W on the schedule of commitments for further information on Beaver Valley.
The other utilities identified by S&W as having plants with the same problem were notified. These plants and utilities were:
Fitzpatrick, Power Authority of the State of New York; Maine Yankee, Maine Yankee Atomic Power Company; Surry 1 and 2, Virginia Electric and Power Company.
The Chairman was advised. Three staff members were sent to Boston to provide inmediate review and analysis of results.
DLC sent eight people to Boston to assist in expediting the review.
3/9/79 In view of the problems and with the Offsite Safety Review Committee concurrence, the Beaver Valley Unit 1 was placed in hot standby for the weekend by DLC to await further analyses from S&W.
3/10/79 Staff meetings continued as pieces of information were fed back from Boston. The I&E Duty Officers were advised of actions.
The NSSS vendors for the plants were contacted to assure no cther codes for
pipe stress during that period used the same algebr$1c approach.
A D0R Assistar.t Director was sent to Boston to provide management review and coordination.
S&W's computer was dedicated full time to these stress calculations and extended work hours for data reduction was instituted for S&W staff.
NRC options were explored and draft materials developed to support appropriate action based on the technical results becoming available on Beaver Valley.
3/11/79 Early S&W reanalysis results on Beaver Valley runs indicated problems with pipes as well (originally thought only supports).
Licensees' top management was contacted to assure action underway by all plants to identify inadequacies and obtain reanalyses of stresses in all affected safety systems.
3/12/79 Additional information from DDR staff in Boston confirmed pipe stresses above allowable and unaccept-able.
Arrangements were made to brief the Commission on this matter. All the licensees were notified of a pending decision.
3/13/79 In view of the safety significance of this matter as discussed above, the Director of the Office of Nuclear Reactor Regulation proposed to the Commission that the public health and safety requires that the present suspension of operation of the facility should be continued:
(1) until such time as the piping systems for all safety systems have been reanalyzed for earthquake events to demonstrate conformance with General Design Criterion No. 2 using a piping analysis computer code which does not contain the error discussed above, and (2) if such reanalysis indicates that there are cunponents which deviate from applicable ASME Code requirements, until such deviations are rectified.
The Commission concurred in the NRR Director's decision.
Prior to the NRC final decision to order the plants shutdown, the Beaver Valley Offsite Safety Review Committee recommended the facility be placed in cold shutdown based on the data and analysis recieved from S&W.
The DLC ordered the plant shutdown.
- 3/14/79 The licensees confirmed by telecon that the Orders were received and provided times when each facility would be in cold shutdown.
All facilities will be at or below 200 F by 10:40 p.m. on March 15, 1979 in conformance with the Order.
Subsequently all affected licensees were notified by telephone that the Orders were executed and that a copy would be transmitted by facsimile.
3/16-17/79 Meetings were held with Stone and Webster with the Utilities to discuss acceptable methods of analysis for interim and long term fixes of the piping and supports.
General Description of Seismic Desian Reauirements The seismic design of nuc. ear power-plants involves two principal consi-1 derations:
(1) the definition of the seismic effects to be designed against, in terms of intensity and characteristics of shaking, and (2) the design of structures, systems and components to resist the defined seismic shaking.
The definition of seismic risk involves consideration of the geologic features of the plant site, observed ground motions related to these geologic features, and observed structural response to earthquakes.
The information available from historic records, measurements recorded in more recent years, insights that can be gained from analyses, and damage assessments following earthquakes have been synthesized to arrive at the engineering methods used to define the seismic hazards for nuclear power plants, dams and other public structures.
The seismic input, once defined, is used in a mathematical process to determine how the structure would vibrate in response to the seismic shaking. Through this process very complex natural phenomena and the responses of complex structures and equipment are idealized so that common principles of applied mechanics and mathematics can be employed to determine the response of each of the major portions of the structures and equipment.
To compensate for these idealizations, the engineering practice involved in seismic design for nuclear power plants generally incorporates
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e conservative design considerations at various stages in the analytical This crocess proceeds from the specification of a design earth-process.
quake that is severe enough to exceed any likely possibility of occurrence during the lifetime of the facility, though a combination of conservative assumptions at many points in the design.
The final design resulting from compounding of the individual conservatisms of the various steps is therefore judged to be conservative.
Once the maximum peak acceleration level is chosen for the design earthquake, conservatisms are provided in the definition of other seismic input parameters, such as the definition of wide band response spectra and the use of synthetic artificial time histories which envelop the response spectra.
Conservatisms are also provided in the seismic analysis and design for structures, systems and components in the following areas:
(1)
Elastic-dynamic analyses are performed using conservatively low damping values in either a time-history or a response spectrum method I
of analysis.
i (2)
Soil-structure interaction which can have a significant effect on reducing seismic response is conservativel; considered.
The behavior t of the subgrade area during seismic excitation is represented in a soil-structure interaction model.
As an example of such conserva-l tism, in both Beaver Valley 1 and Surry 1 & 2, the subgrade' area was l
a modeled using a half-space lumped mass and spring system with damping values limited to no more than 5 to 10% af critical damping; whereas, soil-structure interaction studies have shown that actual subgrade camping can be as high as 30 to 50% sf critical damping.
(3) Three input components of an earthquake (two horizontal and one vertical) are considered with both of the horizontal components con-sidered to be of the same intensity.
(4)
For cases where piping materials are subjected to small excursions into the inelastic range the dynamic response is reduced as a function of the amount of inelastic action.
This can be represented by a ductility factor which is 1.0 for purely elastic behavior and increases with increasing inelastic behavior. A ductility of up to 1.5 can be assumed for vital piping. This would have the effect of reducing accelerations of elastically calculated response spectra by as much as 1/3.
In the design of structues and equipment, it is convenient to assure that all elements of. the structure or equipment are designed to stress levels well below the actual strength of the materials so g
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tnat any permanent deformation,is small.
This approach eliminates tLe need for complex and costiy inelastic analyses.
Actually, from the standpoint of functionability, major structures and components in nuclear plants, as well as other industrial equipment, can toltcate This deformation and usually even the failure of some structural members.
deformation and loss of structural members crn be sustained because of redundancy, i.e., there is more than one path available to carry loads, and load sharing results so that the load formerly carried by a failed member is redistributed to other members.
(5)
Stress limits, whether elastic or inelastic, are based upon material behavior under static loading conditions.
Since dynamic loads contain a limited amount of energy, the margin (between the stress limits and failure) under dyna'..ic loading is greater than under static loading,if elastically calculated peak response is compared with the stress limits and strain-rate effects are neglected.
(6) The design of the structural elements is such that the capacity usually exceeds 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 those prascribed by the analyses are often used.
(7)
Engineering codes specify " code minimum strength" for materials.
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These code minimum strengths are in turn specified by the applicant wher, the materials are ordered; ary material found to be under that strength is rejected.
The result is that the material supplier, in order to assure that he -tands no risk of having costly material returned, provides material of higher strength than specified.
Additional conservatisms for equipment and piping can be found in the following areas:
'; hen the floor response spectra are developed for the design of (1) components located at different locations in the structure.
the peaks in the individual floor response spectra are broadened in order to yield responses that account for uncertainties.
(2) Where the system has multiple supports, the staff require', that the maximum response spectra be applied to all support points to account for uncertainty in the seismic loads.
(3)
In calculating the seismic loads for these components, the damping values are applied several times (first, to major structures, then to the intermediate structures and finally to the equipment itself).
The inultiple application of these low damping values compounds the conservatism in the seismic response for which the equipment is designeo or testec.
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6 (4)
Even identically designed redundant systems may not always experience identical seismic excitation b'ecuase they are mounted in different locations in the structure, with different structural filtering effects.
Thus the loss of one system may not mean a loss of function if the other redundant system remains intact.
(5)
For simplicity Jf calculating, assumptions of rigid boundaries are made in many places (e.g. at nozzles, restraints and snubbers).
Consideration of the actual flexibility of the boundaries would tend to reduce the calculated loads.
Tr.e end result of applying these conservatisms is that structures and cc ponents have seismic capability in excess of the established design goal.
There is ncrmally no motivation to go back and assess the true as-built strength of various structures, systems and components, because the costs of reanalysis and time lost would swamp any reduction in building size or equipment capabilities that might be gained.
The specific information necassary to quantify these conservatisms in the licensing process is therefore not usually developed.
By way of cu.:parison, hospitals, schools, major apartment complexes, large structures that house many people, and essential facilities that have to be designed to resist exterisive loss of life are designed to
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criteria that are, for the same earthquake exposure in terms of acceleration, from 8 to 20 times less conservative than those applied to a nuclear power plant, when the total design process is con;idered.
EASTERN U.S. SEISMICITY The eastern United States is generally believed to be a region of low seismicity, when it is compared with the western part of the country.
It is not, however, without significant historical seismic activity.
For plant sites in the eastern United States relatively little information is available on magnitudes (i.e., few instrumental determinat'cns er.ist) for the larger historical earthquakes, and we must rely on their descrio-tions in historical recorts of earthquakes effects (i.e., intensity).
In determining the Safe Shutdown Earthquake (SSE) in the eastern U.S.,
the staff is required by Appendix A to 10 CFR Part 100 to identify the most severe earthquakes associated with tectonic structures and tectonic provinces in the region surrounding the site.
Normal staff practice with regard to tectonic provinces is tc accept the largest historical earthquake associated with the tectonic province as the "most severe earthquake." To assure appropriate conservatism in light of the short nistorical record of earthquake cctivity, the staff has required that the geographical area encompassed by a tectonic province be large for sites in eastern North America.
By doing this, the earthquake data sample has a much greater likelihood of including earthquakes largt
i 4
2 enough 50 that they have a low probability of occurring in the site vicinity during the lifetime of the facility and are thus consistent with the desired conservatism for a Safe Shutdown Earthquake.
The attached chart indicates are estimate of the horizontal acceleration which is not expected to be exceeded (90 percent chance of not being exceeded) during a fif ty year period in the eastern United States.
Fifty years is roughly equal to the operating lifetime of a nuclear power plant.
A comparison of this chart with the design operating basis earthquake (OBE) levels determined in accord with 10 CFR Part 100 Appendix A indicates that tne OBE generally has a lower probability of being exceeded than the
~
chart shews; this is expected.
A section of "A Catalog of Southeastern United States Earthquakes 1754 through 1974," by G. A. Bollinger is also attached.
This material shows the seismicity in the region of the Surry plant. A chart is also provided showir; the relative location of significant earthqut kes in the surround-ing area.
A listing of earthquakes which indicate the seismicity of the Fitzpatrick site is provided.
Two charts illustrating seismicity in the region of the Beaver Valley site are also provided.
Finally, two charts illustrating historical seismicity in the vicinity of the Maine Yankee plant are attached.
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The peak accelerstions derived for the Safe Shutdown Earthquake (SSE) and the Coerating Sasis Earthquake (0BE) at the four sites are listed below.
PLANT SSE OBE "aine Yankee C.10;
.05g Eeaver Valley C.12Eg
.059 Fitz;atrick C.15;
.CSg Stery
- .159
.079 Tnese accelerations are approximately equivalent to those which current staff practice would associate with Modified Mercalli Intensity VII earthquakes of various strengths for the SSE values, and about MM VI for the OBE values.
Based upon our knowledge of the regional geology and nistorical seismicity associated with the sites and previous estimates of earthquake recurrence intervais for other sites in the eastern U.S., it is our initial judgment
-3
-4 that the risk of exceeding Intensity VII is of the order of 10 to 10 per year at these locations.
Because of the relatively lower design accel-eration and somewhat higher local seismicity, the Maine Yankee site would appear to be at the higher end of this risk of exceedence range.
The chance of exceeding the GBE levels is widely es imated to be on the order of 5 tiries tnat of the SSE level.
.P
- n ci;;.a:e of recurr:r.ce intervals of ;eak
" !s a? ;;roscr*-
It is r.c:.;;cd directly acca'eratica in dif ferent parts of the U. 5.
in ? RC st.if f deter:f r.aticns of.eismic design re'quirc cnts.
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I Prelirninary Map of Horizontal Acceleration (Expressed As Percent Of Gravity) in Rock With 90 Percent Probability Of Not Seing Exceeded in 50 Yecrs
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A I-Catalog of Southeastarn U..ited States Ear:hquakes I8 1754 through 1974 L
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l by G. A. 5cti'eger I Scierces De;artment cf Geologic:
Ver;irda Pcf ytechnic InstiNte and Sta:e University Slac'ssburg, Vir; inia 2.:061
- 975
LIST OF VIRGi!;!A EARTHCUAKES In tensity-felt i.res L ocality Time Latitsce ' lor';itude
-.fagnituce ISq. Mil Re.'erences Year Cate 1774 Feb.21 Petersburg 1400 37.3 77.4 Vil 58.000 3.;S.25 ceo. 22
'.% ttia msbu rg P '.t 37.3 76.7 IV 3.28
'775
- ta r.16 Nelsen County 14 C01500 37.7 7S.8 V
25 Aug.30 Nelsen County
- 00 37.7 78.8 25 1729 Nov.19 Fe de rick sburg CEDO 33.3 77.5 3.33.25 1791 Jan.13 Nelsen C:unty 0400 37.7 78.5 IV 25 Ja n.15 Richmond 05CO 27.5 77.4 3.38
- 795 F o.11 Aicnmenc Frece cksburg 0C0 38 77 5 3.03
'301 F e o.10 Ly nch burg.
- C0 37.4 79 2 (me tect ?)
3.33 a g. 23 Rient end. Lynchburg?
0500 37.5 77.4 V
3.05.25.60 302 u
.507 A r.20 Va.. N C.
CaC0 V
3.25.25 1512 Ce. 02 Richmond C330 37.S 77.4 lV 3.03 Apr. 22 Richmond C200 37.6 23.4 V
3.25.3S
- S16 Cec.31 Norfolk 1300 36.8 76.3 Ill 4.25.33
- S S Aug.09 Richmond 2100 37.6 77.4 ti.fil 3.25.38 Aug.10 Richmond 1200 37.6 77.4 3.25.38
- S:8
..t a r. 09 Virginia 2:00 V
218.000 3.17.37.25.38
- 833 Aug.27 Central Va.
0500 (Near Richmond?)
V 52.000 3.25.38.13.21 346 Oc:.18 Lexington 21CO 25
'350 C::.17 c amvi:te 37.4 78.4 IV.V 3.25.;S
'352 Apr. 29 V a..N.C.. T e nn.
1300 36.6 31.6 VI 162.C;3 3.25.38.17.37
- tay 03 Abingdon (7) 0300 36.7 82.0 3.38 Sept.18 Abingdon (?)
0300 36.7 S2.0 3.38 Nov. 02 Virginia 1835 VI 12.000 3.25.38
- 853 Jan.30 Woodstock 38.8 78.5 3 38
'tay 02 va..W.va. Chio C920 33.5?
79.5?
V 72.000 3.25 3S
J I
Intensity-felt Ares Y tar Date L ccality Tirne L a titude L ongitude
..fa;1itude (Sc..'.11.1 b'erences 1654 Nov. 22 To:ev.e:1 County 16CC 37.1 '
21.5 Ill 3.25.38 1255
- eb. 02 Vi gi-;a 0300 V
3.000 3.25.28 1856
.an.16 "S nche ster C200 39.3 78.2 IV
' Af:er.n ock s ?)
3.25.38
- ar. 21
- eisen County C300 37.7 78.8 lit ?
3.25.38 1859
' tar. 22 Tar ewell Coun:y 37.2 31.5 3.28
- 861 3.ug. 31 V:r; inia 0522 (N.C. e eent?)
VI 300.000 3.17.37.:S 1872
- ?ar. 01 Cha: nam 7'A 35.8 "S
fil ?
3.05.38
. re C4 Che::er f.e d
- :;- 3C0 37.6 77.4
!Il 9.000 2.25.32.00 1973 Cet. 03 5.rkecae C7:5 37.2 78.2 IV 5.000
- 3. 0 5.33.G0 1875
- ?ar. 10 G ec:.:an: C:unty 1200 37.7 78 Ill 3.25.3S C ec. 22 R 'c h t-.c nc 2345 37.6 77.4 VII 50.0C0 3.11.17 (Af:ersnceks?)
25.38.48 C ec. 23 Gecc.":sno Ceunty P.'A 37.7 78 (Af tershocks ?)
3.38 C ee. 26 F:wna:sn C unty 1200 37.6 77.9 3.38 1876 Jan.02 P:v.ha:an Ocunty 2130 37.6 77.9 111 3.25 Cee. 21
'A'y theville 1030 26.9 81.1 3.38.25.60 Dec.22 Ches:erfield 2345 37.4 77.5 (Aftershecks) 3.38 1878 Jan. 02 L: visa 1900 37.8 77.8 111 ?
3.25.38.60 iS32 A:r. 02 Ne.v *.'a r k e t 28.6 78.6 Duncer-3.28.50.25 s:Orm ?)
- S84
- tar. 29 Ac: mac County 2C00 37.7 75.7 3.38.25
- 585 Jan.C2 V a. a n d *.1 d.
2116 39.2 77.5 V
3.5C0 3.28.25.60 (Af:ershocks)
Fec.02
'.*/y th e ville 0710 36.9 81.1 IV 3.38.25.60 Oct. 09 Nelson County 2335 37.7 78.8 V
25.000 3.17.38.25.00 (Aftershocks)
- SS6 Sept.01 Charlet:esville 22 28.1 78.5 3.38.60
I intensi:y.
felt Ares
'est Cott L ocality Tin:o Latitude Longitude
'..'e;nitude (Sq. ?.fi.)
Refere. ces i35 Sept.3
./ythev.:le 2400 25.9 81.1 3.03.50 Sept. 24
'.*!/ th e ville 2156 25.9 81.1 (Af tersn :ks) 2.50
- 37
'.*a r 03 Pulaski 1:18 27.1 80.7 VI 20.000 3.17.38.25.50 (Afterstecks)
?.tay 31 Giles C:unty 1353
._ 37.3 20.7 Vill 280.000 3.11.1.s.17.25 (Aftershocks)
.uce :3 Reancke 23012400 37.3 79.9 V
9.500 3.05.2S 37 3 20.7 IV 7 05 te:t. 3 Pe :t rg 3.25.32 Se:t. 4
'/.'y th e v ill e C6 0 26.9 81.1 Oct. 21
'/ly hevHle 2220 26.9 81.1 V
23.CCO 2.17.23
'tev. 27 A s'.:a -d 1556
~ 7.7 77.5 IV V 3.25.38
~;e:. ' E Ash!ard
- 245 37.7 77.5 V
10.000 3.17.38. 5 53 Fe:.C5 Faiask. //ythe.. le
- 5CO 27 St VI 34.JC0 3.25.38 (Af tershecks ?)
Fec.CS Cublin 2100 37.2 80.6 25.33 Nov. 25 Pulaski>//y theville 1500 37 81 V
65,000 3.25.38 353 F e o.13 Fulaski-Wy theville 0430 37 81 V
115.000 3.17.25.32.60 (Aftershocks)
.. tar. 03 Norfolk 36.3 76.3 3.38 32
- iay 17 Fearisburg 23C0 37.3 80.7 V
3.28.50 i:5 A;.29 Seef ero C:ty 37.4 79.5 3.35.60 107 F eo.10 Scottsville 1930 37.8 78.5 Ill 3.33.50 Fec.11 Arvonia C322 37.7 78.3 VI 5.600 48.60 308 Aug. 23 Fewhatan Cour.ty 0430 37.5 77.9 V
1.500 3.17.23.40.60 (Foreshock 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> earlier; Aftershock Aug. 23, 20:00) 909 Apr. 2 Winchester 02:5 VI 2.500
- t0 Feb.C8 Shenand;ah Vaffey 0900 33.7 78.7 IV 1.100 3.38.60
( Af tersheck s)
Yav 03 Ar.cnia 1610 37.7 78.4 V
4.000 3.17.38
.'n te r~sity.
Felt.4res Yest Cate locality T7me Latitude Lor';itude
!.tegnituce (Se Mi.)
References ill F e o.10 02n n:le C522 26.6 79.4 IV 3.33.29
-12 Aug.07 Arvc. ia 2000 37.7 714 IV 400 3.23.20.29 n
3.38 37 82
!!7 Apr.19 S'N Va 3.38 213 Aor.09 Luray 13C8 38.7 7 ?.4 Acr.09 Lurav 21C9.
38.7 78.4 VI 71.000 3.17.23.35 A ct. " O Luray C200 23.7 78.4 (Aftershecks) 3 3.08 a r.16 Luray C240 23.7 78.4 o
apr. ! 9
'icrf eik Suf fo:k 1155 25.3 76.3 til 1.400 3.23.29 i'i Ce:t. 5 N Va.
2146 23.8 78.2 VI 2.:7.23.25.29 22
.uy 24 L. ray P 'A 05.7 78.4 IV 2.23.29 35 6 32.3 VI 2.17.22.50 3:1 J.!v 15
- 'endora Aug. 37
.c.v Canton 0120 37.8 7 S.4 VI 2.500 3.17.28.60 223 Dec. 31 Carke County 2400 39.2 78 V
(Af tershocks) 3.28.60 39.2 78 IV.V 3.60 224 Jan. 31 Carke County Dec. 25 Roanoke 2400 37.3 79.9 V
3.38
- 225 vav 15-Scuta Richmond 2030-37.3 77.5 (Aftershocks) 3.28 0030 May 16 July 14 Richmond 1620 37.6 77.4 IV 3.38
'277 June 10 Augusta County 0216 38 79 V
2.900 3.17.28 52:
Oc:. 20 R:c. mend C645
.17.5 77.5 IV 3.100 3.28.52.50
'223 Dec.26 Chart ottesville 2156 28.1 78.S VI 1,000 3.17.52.:8 3.38.52
'920 Sect.15 Richmo.s C240 37.5 77.5 3.38.52 1931 Oct.05 New Canton 2215 37.7 78.3 1932 Jan.04 Suckingham County 2305 37.5 78.6 V
800 3.38.52 Dec. 25 Petersburg PM 37.3 77.4 (Seismic 7) 3.28.52
- 232 Jan. 26 F-iersburg 2200 37.3 77.4 ill
( Af:ersnocks) 3.23.52 July 23
- ew Canton
'C00 37.7 73.3 111 3.23.52 2.22.52
'?;4 1er. 02 Fe e'sburg 2105 37.3 77.4 3.23.52 125-Fe:.10' Fe:e sourg I S45.
37.3 77.4
's:S Apr. 09 Charlot:esville 0742 28.1 78.5 111 3.28.52 55
la tensity.
Fett Arez u
Ca te L ocality Time Latitude L ongitude
. fa7nitude (Ca..'.fi.)
References 17 Feb.02 A termarie County 2026 37.7 78 6 til.IV 5.500 3.38.52 400 3.38.52 40
- tar. 25 Ediebur;, Toms 2100 33.8 78.5 V
iAf:ershocks)
Brock..'.'codstock 2228.0301 2215 37.6 78.6 Ill 1.500 3,08.52
- 2 Cct. C6 acckingum Ceunty 3.38.52 1543 37.5 78.5 45 Cct.10
'lew Canton 3.38.52 1500
~~
37.6 78.5 Cet.12 Ci!!wyn 3,38,52 Cct. 29 Di!!wYn New Canton 2129 37.6 78.5 3.38.52
-5
'tay 24 A:berr-ar!e County 1440 28 78.5
- 3 Jan.".4 EucWegham C:unty 2145.2220 27.5 78.6 IV 1.7C0 3.33.52
( Af:ershock s) 2300.0100 3.38.52
' Mar. 26 CNrro::ervi:le 1848 28.1 78.5 ai May 08 Pov.haran Ceunty C601 37.6 77.9 V
2.700 3.38.52 40.41 1630 6.8 83.0 II.!!!
Sept.*6 Lee Caunty 0430 36.8 83.0 IV V 700 3.38.40.41.52 Sept.17 Lee Ccunty 0245 37.7 78.4 V
900 3.38.52 s50 Nov. 25 Suckir gham County 3.38.52 0200 37.6 77.4
'?51 Mar. 09 Richmond 252 Se pt.10 Charlottesville 2215 38.1 78.5 IV.
400 3.38.52 (Afrershocks) 3.38.52 0300 37.7 78.2 IV 253 Feo.07 Gocchiano County 1530 36.6 82.2 IV 1.800 3.38.52 E55 Jan.06 Va..Tenn.
Jan.17 Prince Edward Ccunty 0737 37.3 78.5 IV.V 2.200 3.38.52 253 Apr. 23 Va AV.Va. border 2058:41 37.5 80.5 V1 3.000 3.17.38.52 1817 37.4 80.7 IV 500 3.38.52 July 07 Giles County 1220 37.4 30.7 IV GOO 3.38.52 Aug. 21 Giles County 1840 37.4 79.3 IV (Attershock 3
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> later)
'360 Sept.4 Boonsboro C540: 26.8 37.3 79.9 IV 3.800 52 553 Jan.17 Salem 2238 36.7 31.0 V
1.300 3.52
( Af:e shecks)
Cct. 28 Ga:nx C619:02 37.6 78.0 V '3.7 23.C00 3.4.52
'956
- t ay 31 Powhatan County C538: 15 37.3 S0.8 IV;4.1 3.200 2.3.4.52 is3 ' Mar. C8 - Narrcws 2344:39 28.0 77.9 IVi3.5 5.500 24.52
'is9 Dec.11 Louisa CCC6:27.1 33.1 77.4 Iv!3.4 1.000 6
( Af tershock)
- 371 Sept.12 Frederick sburg
- 6C0; 0.9 37.5 77.7 V:3.4 2.300 1
272 Se:t 05 T c.' r:- o r-d Sect. 5.194 7. Se:t. S.1045)
(Pess:b;c Af:ers5:cks:
56
Intensity-Fett Area Yest Cate L ocality T7mr Lati: ace Lcngiruce stagnitude
(:q..tfi.)
Refe*ceces
- 373 a t. 09 Chesterik!d Ccunty 2311 37.3 77.7 IV 2700 45 c
- 974
' tar. 23 Faucier County 02:6 23.3 23.9 77.3
- 2. 5 45
- tay 30 G;1es County 2128:37.2 37.4 30.4 3.8 45
'4cv. 7 Goochland County 2131 37.7 78.2 2.4 8
Q
P
'The listing of earthquakes which follows is obtaineo fecm the " w Haven preliminary Safety Analysis Report.
It is cons. ered o e a:plicable to the Fitzpatrick site since the "ew riaven site is ess 6 miles distant.
TABLE 2.5-3 EARTl!QUAEES TELI_AIJ!EHJIAVE11 Epi-Dato Lat.
Long.
Contral Ita gni-Dist. to Site Intensity fr____119 _J.L1 J1RlllMll_L'! illL_
DI)
InunniL 19de_ jite (mil A".a.
Dawn r""""
EDmAIll 1638 Jun 11 20 47.65 70.17 IX 412.5 4.4 1661 Feb 10 12 45.50 73.00 VII 213.0 3.5 1663 rob 5 1730 47.60 70.10 X
412.6 5.3 1665 rob 24 47.80 70.00 VIII 425.3 1.3 ir27 Ilov 9 2240 L
42.00 70.60 VII 290.6 3.0 1732 sep 16 1600 45.50 73.60 VIII 192.3 4.6 1737 Doc to 40.80 74.00 VII 219.3 3.4 1755 Itov to 0412 L
42.70 70.30 VIII 306.9 3.9 r
1791 tiay 16 0800 L
41.50 72.50
-VII 236.7 3.3 1791 Doc 6 23 47.40 70.50 VIII 389.5 3.5 1811 Duc 16 0200 L
36.6 89.6 XII 849.5 IV 5.7 1816 Sup 9 45.50: 73.60 VII 192.3 3.6 1840 Jar 16 2000 43.00 75.00
-VI 73.3 4.0 1840 trov 11 39.80 75.20 VII 260.3 3.2 18 53 1:a r 12 0700 43.70 75.50 VI 42.7 ltR 4.7 1855 rob 6 2330 42.00 74.00 VI 155.0 3.0 10 V. Oct 23 2015 43.20 78.60 VI 117.0 3.4 1066 Oct 17 1113 47.50 70.10
-IX 400.1 4.4 1861 Jul 12 2100 L
45.40 75.40 VII 139.4 IV 4.1 rett report at Oswego; :teu llavon is about 10 mt naaror opicentor.
1867 Doc 18 0300 44.65 75.15 VI 98.6 IV
-IV 3.6 rult report at Oswego.
about 10 mi moro distant 1 of 4
e liYSELG PSAR i
TABLE 2.5-3 (Cont'd.)
Epi-Dato Lat.
Long.
Central Magni-Dist. to site Intensity Yr llo DA 11F.littif.C_L1 (liL H3)
Iniin11L7_. t u d ri Sit <r (mil A11 1"""
C""""
Etnith from epicentar then site.
1670 Oct 20 1630 47.40 70.50 IX 389.5
-III 4.5 rett report at osvego; llow
!!aven is about 10 mt nearer the uptcentur.
1871 Oct 9 0904 L
39.70 75.50 VII 264.2 3.1 1877 Itov 4 0156 L
45.20 73.90 VI 167.9 lir 2.8 Report from oswego, about 10 mi more distaat from opicuntur than site.
l 1804 Aug to 1907 40.60 74.00 VII 231.1.
IIR 3.4 1006 Aug 31 2151 L
32.9 80.0
-X 757.2 t(R 4.0 1893 llov 27 1650 45.50 73.30 VII 202.8 liR 3.6 1697 tiar 23 1007 L
45.5 73.6 VII 192.3 IIR 3.6 1897 Ita/ 27 2216 L
44.50 73.50 VI 155.6
-IV 2.9 rolt report at Oswego; tieu llavon i s about 10 mt nearer opicun.ur.
1903 Dec 25 1230 44.70 75.50 V
92.8 IIR 2.7 1913 Apr 29 002857 44.87 73.33 VI 107.0 IV 3.5 rolt report at Oswego; lluu Haven is about 10 mt nearor epicentor.
1914 Feb to 1831 46.00 75.00 VII 104.9 V,-VI 3.7 rolt reports from rutton, 12 nites su of site, and from Laasing 7 ni u of site.
- Iuu Haven is nearer epicentur.
1922 Dec 8 1624 44.35 75.12 V
83.7 IIR 2.8 1924 Jul 15 0010 45.70 76.50
-VI 153.2 3.0 1924 Sep 30 085230 47.60 69.70
-VIII 5.5m 426.9 IIR 3.3 g
1925 Itar 1 021920 47.60 70.10 IX 412.6 V
IV 4.3 rett report at llev liaven.
liYS E1G PSAR TABLE 2.5-3 (Cont'd.)
Epi-Date Lat.
Long.
Central MagnL-Dist. to Site Intensity lt p_.,_D A }(glglitLG_lJ_ (13 )
DI)
Initn).LtL tudq Site (niil E B"""
C""""
RCFL3IM
'327 Jun 1 1223 40.30 74.00 VII 249.3 liR 3.2 928 liar 18 1525 44.50 74.30
-VI 121.5 lik 3.3
'.929 Aug 12 112448 42.87 78.35 VIII 111.4 IV 5.4 rett rebort atosuego[-
I mi nearer ep about contor than sito.
i 1929 trov 18 1532 44.
56.
X 1014.2
<II 3.2
- 931 Jan 8 001336.5 47.63 70.17
- 5. 4tt 411.6 3.0
'.931 Apr 20 1954 43.40 73.70 VII 130.3
!!R lir 4.2
'.934 Apr 15 025813 44.67 73.80
-VI 148.4 IIR 3.0 !
(
1935 tiov 1 060340 46.78 79.07 VII 264.3
-V IV 3.1 rett report at Oswego; sitt and Osuego about equal dLs-tance from epicanter.
1935 llov 2 143158 47.23. 78.17 5.4H 273.9 liR 3.8 i
L 1938 Aug 23 050455 40.25 74.25 4. 8 11 246.6 3.1 g
I 1938 trov la 221906 44.75 75.25
-v 101.7 2.5 250.4 r
III 3.2 rolt report at Syracuse; 5.4gLE Syracuse and site about
'940 Doc 20 072726 43.80 71.30 VII equal distance from I
epicenter.
i 1940 Dec 24 134344 43.80 71.30 VII 5.4m 250.4 IV III 3.2 rett report at Syracusei g
bEC Syracuse and site about equal distance from epicenter.
1942 Itay 20 121922.8 45.77 74.67 4.41t 176.8 3.1 g
1944 Jan 22 215509.1 45.03 76.78 4.31t 163.6 3.1 g
1944 Sep 5 043045 44.97 74.90 VIII 123.7
-V V
5.3 rolt report at Pulaski, approximately 10 mt nearor epicenter than sito.
tiYSE10 ?SAR TABLE 2.5-3 (Cont'd.)
Epi-Date Lat.
Long.
Central
)'agni-Dist, to Site Intensity f r_"n 9.1 jlP:!P!1T.C_.L".
11D,_
(u)
Irtut[ titty _,
tudt, lite (nll A w a.
Band C====
Rg!'tidJ 1944 Sep 5 085106 44.98 74.90 4.611 12'. 3 3.9 1944 Sep 9 232448 44.98 74.90 4.111 124.3 IIR 3.2 L
19 (;. Oct 31 084225 44.98 74.90 4.011 124.3 3.1 L
19'22 Aug 25 0007 43.00 74.50 V
96.4 IIR 2.5 19',a Jan 31 1230 42.89 77.28 IV 64.0 llR 2.2 195* Apr 27 021408 43.10 79.20 4. 1 11 147.9 3.0 1958 Itay 14 174121 46.97 76.55 5. 4 11 240.8 4.0 L
195d Jul 22 014640 43.00 79.50 4. 3 11 164.1 3.1 L
'1963 Oct 15 135953 45.18 77.59 4.511 196.4 3.1 s
L 1964 rob 13 194642 40.40 78.20 5. 2 11 233.9 3.7 L
1966 Jan 1 132338 42.00 78.20 VI 106.6 lir 3.5 1967 Jun 13 190854 42.90 78.20 VI 103.6 tlR tir 3.5 1971 Itay 23 062427 43.02' 74.54 3.7m 90.8 2.5 bLg 1911 Itay 23 092959 43.94 74.55 3.6m 92.7 2.3 1975 tiov 3 205455.9 43.89 74.64 3.9m 87.4 2.9 bl.g
' 1975 trov 11 205455 43.91 74.64 3.9m 87.8 2.9 bl.g E0TJ.S t
- L = Local time IIR - tiot reported as felt in the area of the site.
tif - tiot felt.
aalntensity documented at localities near the site.
Reports compiled in Appendix 2.5-F.
adastte intensity observed from isoseismal maps compiled in Appendix 2.5-F.
==dastto intensity dottvud using: Isito : Io + 3.7 - 0011 ( km) - 2.7 Log.. ( km)
(Guraa and tiuttli, 1976).
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Present State of Verification of Stress Analysis Methods Existina detailed requirements contained in pertinent Standard Review Plans and Reculatory Guides issued since the five plants were designed and aporoved have greatly reduced the chances that design errors of this tyce will take place.
The Standard Review Plan sections and the Regulatory Guices which pertain to seismic analysis require a dynamic analysis, and provide for input time histories, ground response spectra, damoinc, modelling of structures, develocment of floor response spectra, and methods of combination of both spatial components and modal contri-butions. The Standard Review Plan also requires that app.icants verify their dynamic analysis programs by comparison of results with those of other programs and with generally accepted solutions to benchmark problems.
These current criteria are adequate and do not recuire change.
Had they been in place at the time these five plants were reviewed, the error ae are now concerned with would probably have been discovered.
To improve our confidence in computer results, the staff has for some time been in the process of establishiag a standardized program for independently evaluating and verifying the quality of computer p;ograms used for dynamic and static structural ar. clysis of nuclear piping systems and components.
This program consists mainly in the definition and solution of a set of standardized benchmark problems involving the analysis of a set of structures of progressively increasing complexity, representing typical piping system analys
2 as fcund in currently proposed or operatino plants.
Increased assurance of proper code verification will be provided by requesting applicants to provide solutions generr.ted with their computer programs to these standardized benchmark problems, and comparing these responses with the benchmark solutions.
Agreement or deviation of results will provide an index of the adecuacy and quality of an applicant's analysis methods.
This program will also orovide the NRC with the capability to perform inuependent calculations to verify applicants' dynamic analyses for particular desions.
The following paragraphs elaborate on the past and present staff efforts in the area of stress analysis code review and verification.
In 1973, the staff realized that there was a proliferation of computer programs for stress analysis, all of which would be required to be examined in the process of licensing reviews.
Due to the substantial number of plants under review at that time, it was decided that a generic procram to review these computer programs should be instituted that would have two goals:
1.
To provide independent in-depth verification of the capabilities of the programs claimed by the applicants in the SARs; and 2.
To provide the staff with a list of acceptable computer prog s tb-t would reduce the review effort in at least one area.
.. In February 1974, in outline for a validation program was developed proposing that computer programs be evaluated and verified by means of benchmark problems and solutions. These benchmark problems were to be developed independently by the staff, and submitted to applicants recuesting that they orovide solutions to these problems.
The acceptability of an applicant's computer program would be detennined by the similarity of the acclicant's solutions and the benchmark solutions.
In October 1974, a work scope entitled, "Pipinq 5enchmark Problems" was issued for assistance from a national laboratory in generating the benchmark solutions. This work scope described the requiremants for such a program, and a preliminary list of problems suitable to be used as benchmarks.
The Brookhaven National Labot _ tory in Upton, New York, was chosen to provide the required solution.
In Fiscal Year 1975, the actual benchmark problems were selected by the NRC staff and BNL personnel, and computer programs that were to be used for generating the solutions were chosen and verified.
Actual generation of benchmark solutions was begun in FY-1976. The comruter program chosen for this effort was the program SAP-IV (Structural Anrlysis Program), developed at the University of California at Berkeley in the early 1970's and widely available.
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r-Two reports detailing five benchmark problems and solutions were published in December 1977 (BNL-NUREG-21241-RS and BNL-NUREG-23645),
and a draf t request for information became available in January of 1978. The benchmark problems in these reports pertain to linear elastic structures and range from a simple structure under static loading to a two-loop primary piping system compiling a reactor vessel, steam generators, puros and supports, sucjected to earthquake motion.
Addi-ticnal benchrark problems have since bean developed which pertain to elastic structures involving gaps (a non-linear problem). Other problems are being developed which include newer technicues, such as multiple support excitation, and preliminary efforts have been made in developing benchmarks for inelastic piping analysis.
In the course of licensing reviews, the NRC staff ?.as r~
ed descrip-tion and verification of structural programs since the early 1970's, and formalized these requirements in te Standard Review Plan published in 1975, (Section 3.9.1).
Applicants submitted verification solutions which were based on simple benchmark problems only.
The Pipino Bench-mark Program was designed to complement and expand these requirerents and provide additional verification. However, methods of analysis of nuclear power plants for structural response under seismic and other loading conditions, which were the basis for these computer programs and were used in the. design of early power plants (1968), have been presented in the open literature since the late 1960's.
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Acolicants have also provided descriptions and verifications of their ccmputer programs in the form of topical reports.
One such tooical reonrt was submitted in 1976 by the Westinghouse Electric Co. titled:
"Cocumentation of Selected Westinghouse Structural Analysis Computer Ccdes" (WCAP-8252).
These programs and solutions were reviewed as thoroughly as possible without actually performing computer calculations, except for one program which involved a nonlinear analysis.
The bench-mark problem which the applicant submitted was reviewed under the Piping Benchnark Program by the BNL, by generating an independent solution to the same problem ad confirming the applicant's results.
(This problem will be incorporated in our standard list of benchmark problems.)
Duke Power Co. also submitted verification of its method for structural ar.alysis.
The results by this applicant were also verified independently by SNL by running the same problems under the Piping Benchmark Program.
.- final report on this method will be published in the near future.
Other analyses have been verified independently by the staff, and we are presently performing an evaluation and verification of the design techniques of certain component support members.
Related to the Benchmark Program is a much more general computer program ev'luation oroject sponsored by the Armed Forces, and conducted by a group called the Interagency Software Evaluation Group (ISEG).
The NRC' staff is represented on this aroup. The objective of the group is to evaluate in de>th the capabilities of some of the very large
. structural computer programs, such as ADINA, used nationwide.