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,j NUCLEAR REGULATORYCOMMisSION SECRETARIAT RECORD W ASHINGTON, D. C. 20555 g f
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MEMORANDUM FOR: Commissioner R. T. Kennedy I
THRU: p Lee Y. Gossick, Executive Director for Operations FROM: Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
SEISMIC EVALUATIONS OF FIVE NUCLEAR POWER PLANTS This is in response to your memorandum of March 14, 1979.
The seismic analysis methods for the five affected plants were reviewed in some detail, especially at the OL stage of review, and found to be acceptable. However, the staff in its review did not explore the spatial (intramodal) method of combination used in the dynamic analysis of system piping. The review was sufficient to disclose that acceptable methods were used in combining modal re-sponses, but we can find no indication in the agency records that the intramodal method of combination was described or questioned on any of the five plants. Records for other plants of Stone and Webster design that we have reviewed in recent years do contain descriptions of acceptable methods for these spatial response combinations. In addition, the enclosed describes the present state of verification of stress analysis methods.
A brief description of how our seismic design methods have evolved follows. In the early years of nuclear regulation, prior to 1967, there were no fomal regulations or guidance on seismic design methods. The state of the art of seismic design during this time was perhaps best described in a document entitled " Nuclear Reactors and Earthquakes" (TID-7024) issued in August 1963, by the U.S.
Atomic Energy Commission. The report reflected the practices employed in the design of government-owned reactors at that time.
Applicants for AEC licenses were made aware of the existence of such documents and instructed to employ them in design of their nuclear power plants. The methods used for seismic design in the period prior to 1967 were the so-called equivalent static methods.
In the equivalent static load method of analysis a single static force is applied at the center.of gravity of the structure or com-ponent. In using the method, the designers usually took the peak of the calculated dynamic response of the structure, multiplied __
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Commissioner R. T. Kennedy it by some factor between 1.5 and 2, and then calculated an ,
equivalent static force. This single force was intended to l represent the forces due to the inertia of the structure and i the amplification of those forces due to the dynamic nature i of the loading. Although this approach is suitable for systems of simple geometry, it was found to possibly underestimate the seismic response of complex systems in some cases and to over-estimate in others.
Starting about 1967, various experts in the field of seismic j design, most notably Dr. Nathan Newmark at the University of Illinois, published papers demonstrating that advanced dynamic j analysis techniques that were a technological spinoff from the aerospace industry could be applied to the seismic design of structures. The application of these advances in the state of :
the art to nuclear plant design was encouraged and supported by !
the AEC regulatory staff because they permitted better character- ~
ization of the actual response of nuclear power plant structures and systems to an earthquake. It is also important to note that the use of these more advanced dynamic analysis techniques in ,
design of complex structures like nuclear power plants was feasible by the late 1960s because of the increasing availability of computers with sufficient capacity and calculating speed.
When the staff began to require dynamic analysis in the design of structures and components for seismic loading in about 1967, the methods and practices employed by industry were based on the -
available technical literature and on what had evolved as accepted engineering practice in the field of dynamic analysis Inherent in theas it was dynamic appplied outside the nuclear industry.
analysis techniques was the recognition that actual structures and j systems would respond to an earthquake in several simultaneous modes of vibration. This meant that a mathematical method was 1
necessary for combining the spatial (intramodal) components of the
' seismic response at a given point in a structure or system to determine the total response. However, the regulatory staff guidance on acceptable techniques of dynamic analysis for use in i
license applications was limited to basic criteria such as earth-quake and accident loading combinations, allowable stress and defomation limits and damping values. These criteria were com-
' municated principally through the question and answer process used in the staff review of an application. The NRC records disclose
Commissioner R. T. Kennedy that no criteria were issued at the detailed level of analysis involving the combination of spatial response coaponents in piping or structures in these early years.
Beginning about 1967, consulting organizations were retained by the AEC regulatory staff to assist in the evaluation of seismic design criteria for most plants, including Maine Yankee, Surry, Fitzpatrick, and Beaver Valley. Expert and nationally recognized consultants were retained under contract with the AEC regulatory staff in lieu of hiring staff members with comparable expertise.
In the period 1970-1974 the staff was enlarged to include personnel with expertise in dynamic analysis, and a number of consultants were employed to assist in defining more specific requirements for seismic analysis. During this same period of time there was a great deal of activity in the engineering community in the develop-ment of techniques for dynamic analysis of nuclear power plants.
A number of studies were undertaken by engineers in both academic and industrial Circles to define the applicability and limitations of the analytical techniques that were coming into use, including the subjects of modal and spatial response combinations. From our regulatory point of view, this period culminated when the essence of these efforts was codified in NRC Regulatory Guide 1.92
" Components of Modes and Spatial Components in Seismic Response Analyses" first published in 1974 and revised in 1976. The guide is now in routine use in the licensing process and treats fully the method of response combinations of concern in the five affected plants.
l (W~- g Harold R. Denton, Director
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4 Office of Nuclear Reactor Regulation
Enclosure:
"Present State of Verification of Stress Analysis Methods" cc: Chairman Hendrie Commissioner Gilinsky Commissioner Bradford Commissioner Ahearne A. Kenneke, OPE L. Bickwit, OGC S. Chilk, SECY C. Kammerer, OCA J. Fouchard, OPn w -.g . , - - , - - - - - - . . - , - - - , - . - - - - - , . .
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- ENCLOSURE ]
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Present State of Verification of Stress f Analysis Methods*
Existinn detailed requirements contained in pertinent 'candard Review Plans and Reputatory Guides issued since the five plants were designed and approved have greatly reduced the chances that design errors of this type will take place. The Standard Review Plan sections and the Regulatory Guides which pertain to seismic analysis require a dynamic analysis, and provide for input time histories, ground response spectra, dampinp, modelling of structures, development of floor response spectra, and methods of combination of both spatial components and modal contri-butions. The Standard Review plan also requires that applicants verify "their dynamic analysis programs by comparison of results with those of other programs and with generally accepted solutions to benchmark problems. .
Had These current criteria are adeouate and do not require change.
they been in place at the time these five plants were reviewed, the error we are now concerned with would probably have been discovered.
To improve our confidence in computer results, the staff has for some time beenintheprocessofestablishi,ngpstandhdizedprogramforindependently evaluating and verifying the quality of computer programs used for dynamic and static structural analysis of nuclear piping systems and compcnents.
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 analyses
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as found in currently proposed or operating plants. Increased assurance of proper code verification will be provided by requesting applicants l to provide solutions' generated with their computer programs to these standardized benchmark problems, and comparing these responses with the benchmark solutions. Agreement or deviation of resu.ts I will provide an index of the adequacy and quality of an applicant's analysis methods.
This program will also provide the NRC with the capability to perfom independent calculations to verify applicants' dynamic analyses for particular designs.
The following paragraphs elaborate on the past and present staff efforts in the area of stress analysis code review and verification.
i 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 program to review these computer programs should be instituted that would have two goals: 9 l
- 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 1.ist of acceptable computer programs that would reduce the review effort in at least one area. -
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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 The requesting that they provide solutions to these problems.
acceptability of an applicant's computer program would be detennined by the similarity of the applicant's solutions and the benchmark solutions.
In October 1974, a work scope entitled, " Piping Benchmdrk Problems" was issued for assistance from a national laboratory in generating the benchmark solutions. This work scope described the requirements for such a program, and a preliminary list of problems suitable to be used as benchmarks.
The Brookhaven National Laboratory 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 wete to be used for generating the solutions were chosen and verified. Actual generation of benchmark solutions was begun in FY-1976. The computer program chosen for this effort was the program SAP-IV (Structural Analysis Program), developed at the University of California at Berkeley in ,
the early 1970's and widely available.
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Two reports detailing five benchmark problems and solutior, were .
published in December 1977 (BNL-NUREG-21241-RS and BNL-NUREG-23645),
and a draft 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, pumps and supports, subjected to earthquake motion. Addi-tional benchmark problems have since been developed which pertain to elastic structures involving gaps (a non-linear problem). Other problems are being developed which include newer techniques, 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 has required 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 Piping Bench-mark Program was designed to complement and expand these requirenents and provide additional verification. However, methods of analysis of
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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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- j iica'its have also provided descriptions and verifications nf their One such tonical
'T computer programs in the fom of. topical reports.
report was submitted in 1976 by the Westinghouse Electric Co. titled:
" Documentation of Selected Westinghouse Structural Analysis Computer Ccdes" (WCAP-8252). These programs and solutions were reviewed as thoroughly as possible without actually performing computer calculations, The bench-except for one program which involved a nonlinear analysis.
mark problem which the applicant submitted was reviewed under the Piping Benchmark Program by the BNL, by generating an independent solution to (This problem the same problem ed confiming the applicant's results.
will be incorporated in our standard list of benchmark problems.)
Duke Power Co. also submitted verification of its method for st analysis. The results by this applicant were also verified independently by BNL by running the same problems under the Piping Benchmark Program.
A final report on this method will be published in the near future.
Other analyses have been verified independently by the staff, and we L
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 I
evaluation project sponsored by the Amed Forces, and conducted by a The group called the Interagency Software Evaluation Group (ISEG).
The objective of the group
- NRC staff is represented on this group.
is to evaluate in depth the capabilities of some of the very large structural computer programs, such as ADINA, used nationwide.
STON E L WE DST ER r NGINE LRING CORI OR ATION 1-ENGINEERING DIVISION MEMORANDUM NO. can-79-I s
-ENGINUUt<1NG HUCHANICJ DIVISIONq Rey, 1 SUBJECT GENERAL PnOCcDUne POa Tilt sTatss D AT E April 28, 1979 ANALYSIS OF B31.1.0 BRANCH PIPING FROM RPwessel TO Distribution CC 1.0 PURPOSL This procedure provides a uniform approach for the design /cvaluation of branch piping that is consistent with USAS B31.1.0, 1967, through addenda to 1972 Code for pressure Piping. In delineating the various methods of analyzing branch piping, there is a latitude for independent judgement by the experienced stress analyst.
2.0 APPLICACILITY Branch lines are explicitly addressed in the B31.1 Code stating that branch lines should be considered by applying correction factors (stress intensification factors) at the branch connection. It does not specify when a branch line can be analyzed with the run pipe or when it can be treated as a separate, uncoupled system.
In view of the above, branch piping connected to run piping shall not be included in the run pipe scismic model, in general, if the ratio of the moment of inertia of the run pipe to the branch pipe is grector than 10 to 1 (with certain exceptions as noted below).
3.0 MODELING PROCFDURE OF BRANCH PIPE 3.1 Uhen the ratio of the moment of inertia of run pipe to branch pipe is equal to or less than 10 to 1:
3.1.1 The branch pipe should be modeled with the run pipe up to the first anchor on branch pipe (or up to and including the series of rigid constraints that effectively permits termination of the problem at some point remote from the pipe run). Piping outboard of the anchor (or series of constraints) should be analyzed by computer if the pipe is larger than 6" NPS and by manual methods if the pipe is 6" NPS and smaller.
3.2 When the ratio of the moment of inertia of run pipe to branch pipe is more than 10 to 1:
3.2.1 If the branch pipe is 6" NPS or smaller, the branch pipe should be decoupled from the run pipe and analyzed by the simplified manual method up to the first anchor (or up to and including the series of rigid constraints that effectively permits termination of the problem at some point remote from the pipe run).
If the branch pipe is larger than 6" NPS, the branch ,((
pipe may be decouplcd from the run pipe and evaluated in the same manner as specified in this paragraph, except for using a computer analysis in lieu of the manual method.
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3.2.2 Dranch piping that is decoupled from the run pipe j~ *[j should be analyzed with the inclusion of the thornal and scianic novenents of run pipe at the interucction of the run/ branch point.
3.2.3 llowever, there are two exceptions to decoupling branch pipes as delineated in 3.2.1 and 3.2.2 above:
3.2.3.1 If an anchor or rigid constraint on the branch pipe is located near the run pipe and significantly restrains the novenent of the run pipe, the branch run pipe should be included with the nodel of the run pipe, up to the anchor (or up to and including the series of rigid constraints that effectively permits termination of the problem at some point remote from the pipe run).
3.2.3.2 The branch pipe should be included in the mathematical nodel of the run pipe if more i preciso nagnitude of reactions are required
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at torninal points (i.e. equipnent, penetraticns ctc.) to deternino their (the reactions) acceptability.
4.0 VALVES IN LRANCII LIN1;S If the operational node of valvas located in branch piping causes a tonperature change in the branch pipe, the tenperature conditions must be considered in the branch pipe, regardless of the size of pipe or nothod of stress analysis used. This information should be obtained from the cognizant powcr engineer.
5.0 LOADS OM SUPPORTS /COMSTRAINTS Uhere applicable, reactions from a conputer analysis should be used for the design of supports. In the absence of conputer generated reactions, the supports should be designed in accordance with the standard support loads of PS-4.
6.0 In all of the above cases, appropriate S.A.R. seismic qualification criteria should be applied, where applicable.
If there is a seismic class change in the branch pipe, the scisnic analysis should include the piping outboard fron the seismic class change to the first anchor (or up to and including the scrics of rigid constraints that effectively pernits tcrnination of the prob 1cm at sono point renoto from the pipo run. The non-ceismic classified pipe from the first anchor (or from the serics of constraints) may be analyzed by conputer or nanual calculations, depending on the dianeter of pipc; typically, this is an instance where independent judgement must be exercised by the pipe strcas analyst to determino uhore the non-scionic portion of the systen should bu started.
1- 7.0' ATTACli!!!!!1T j
Typical configuration of series of rigid constraints that effectively porr.iitu ' termination of problen. Attachment 1 R. P. Wessel e n 9
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- I Eur,I!!EERING 11ECIINIICS DIVISIC'1 SUBJECT DATE April 7, 1979 STncSSInTrucIric7TIonPAbbons l NID FTRESSES FOR REDUCED OUTLET FROM nrecssel
- natacii Co:IntCTIour. (n31.1)
TO CC EnD TASn roncn !1rnanns 1 Purnone The purpose of this nerorendun is to provide the acceptable code criteria for the calculation of Stress Intensification Pactor (SIT) and stresses at the intersection of a run pipe and a reduced out-let branch connectign. Branch connections covered here include:
ANSI tees, unreinforced fabricated tees, rein #orced fc!ricated tees, ucidolets, nocholets, and branch connections per ficure D-1 of Appendix D, Sunner Addenda of 1973 to AMFI D31-1 Code, 1973.
These factors are applicable to P-Stress, 11upipe, and hand calcule-tiens.
- 2. Procedures .
2.1 The following SIP shoulel be applied on the branch pipe at the intersection of a run pipe and a reduced outlet branch connection:
a) SIP for branch' connections listed belop are provided in Attachnents 1, 1A and 1B:
- AUSI tecs (type A)
- unreinforced fal,ricated tees (type B)
- reinforced fabricated tees, ped thickness sare as run pipe thickness (type C)
- reinforced fabricated tens, pad thichness equals 1.5 tines run pipe thich (type D) b) SIP for tieldolets - Attachnent 2 c) SIP for sockolets - use the SIP for veldolets (per 2.1(b)) or use 1.3, whichever is greater.
The SIP = 1.3 is the UTAS D31.1.0 (1967) Code factor for fillet welds.
d) SIP for branch connections ner fig. D-1, 1973 Sunner Addenda A'ISI B31.1 Code 1973 (see Attachnent 3) , * 'here rn/nn d 0.5 ,
i = 1.5 (ng/Tr ) ! (rrI/E n) (Tb'/Tr) (r n/
- rp) 2.2 For the calculation of ntrens at the intersection of a run pine on1 e reduced outlet branch connection.
.+ It can bn denonstrated that the 'ollnuinn relotione.hin ejl1 aluavn nive connervative vnlues 'or the corrected branch stresses at reduced branch outlets:
F=S g 1 M 1D '
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= Branch stress fron SHoct'2 or 3 i = Ptress intensi fi cation " actor 'to be annlied
(#ron attachnents 1, 1h, 1B, f. 2 )
ig = Stress intensifiention " actor frno S!!ncy. 2 or 3 correspondinn to F, t o = Drn'nch calculation nine thi.ekness used in the n'inCI' 2 or 3 t s = The lessor of it, or tr tr = 'Inninal thickness o# the run nine In cases uhere'the value of 9 exceeds the allownh3e stress, the exa'ct ennrossion 'ron which the above relation .in derived shnuld he annlied. *his exact exnression is:
ll S=S (i) 2 # (3) 2 g o o (i") 2 + (a)2 7[ rt 2 s
where n = The rntin of torsionel nonent to bendino ronent
(?%/M 3 ) in the branch nine.
These values are extracted 'rnn thn s!!nr.1: 2 or 3 run.
r = mho nean radius n' the branch nire.
Bo = The section nodulun of the branch nino extracte.'
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