ML20039A136
| ML20039A136 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/10/1981 |
| From: | Chaudhary S, Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20039A127 | List: |
| References | |
| REF-SSINS-6820 50-289-M81-226, IEB-79-02, IEB-79-2, NUDOCS 8112160264 | |
| Download: ML20039A136 (20) | |
Text
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region 1
' Report No. M81-226 Docket No. 50-289 License No. DPR-50 Priority
. Category C
Licensee:
Metropolitan Edison Co.
P. O. Box 480 Middletown, Pennsylvania 17057 Facility Name: Three Mile Island Generating Station, Unit 1 Meeting at:
King of Prussia, Pennsylvania Meeting conducted: October 6, 1981 AR//d/<f"/
Reporting Inspector e.re S. K. Chaudhary, Reactor Inspector dite signed Approved by:
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L. E. Tripp, Chief, Material and date signed Processes Section, EIB Meeting Summary:
Technical Meeting on October 6, 1981 (Report 50-289/M81-226)
Areas Discussed: A meeting was held at the request of the licensee to discuss the technical justification for the licensee's actions pursuant to IEB:79-02.
Results:
The licensee presented.a summary of the design approach used in the resolution of IEB:79-02 requirements including.a detailed discussion of the factors of safety in the loaa carrying capability of pipe supports utilizing concrete expansion anchor bolts for seismic load conditions.
8112160264 811211)-
DR ADOCK 05000
' DETAILS 1.
Meeting Attendees.
Met-Ed/GPU-Nuclear D. K. Croneberger, Director, Engineering and Design A. P. Rochino, Manager, Engineering Mechanics E. G. Wallace, Manager, PWR Licensing Gi'.bert/ Commonwealth C. Chen, Section Manager, Speciality Structures K. E. Nodland, Asst. Project Manager, TMI - Continuing Services R. M. Rogers, Project Manager, TMI - Continuing Services U.S. Nuclear Regulatory Commission R. C. Haynes, Regional Administrator, Region I (part-time)
T. T. Martin, Director, DETI R. R. Keimig, Chief, Projects Branch #2, DRPI L. E. Tripp, Chief, Material & Processes Section, EIB, DRPI S. K. Chaudhary, Reactor Inspector, M & PS, EIB, DETI J. Fair, Senior Mce.hanical Engineer, IE:HQ E. Gallagher, Senior Structural Engineer, IE:HQ 2.
Meeting Scope On October 6, 1981, a meeting was convened at the request of the licensee at the NRC Region I offices at King of Prussia, Pennsylvania.
The purpose of the meeting was to discuss technical justification for the licensee's actions pursuant to NRC Bulletin IEB:79-02.
3.
Areas Discussed The licensee and licensee contractor discussed the approach and rationale utilized in responding to IEB:79-02. A copy of the documents used in this presentation are attached. The presentation and resultant discuss-ions are summarized below.
a.
The plant FSAR requires a factor of safety of more than one fe-concrete anchors (maintain system functional), and the Bulletiri requires conformance to the FSAR.
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3 b.
Current practice (RG 1.60) allows 2.0% and 7.5% damping factors in piping and concrete respectively. The licensee has conserva-tively used only 0.5% in piping and 5.0% in concrete, which results in higher predicted piping support loads. However, they acknow--
ledged that RG 1.60 amplification had not been used for developing the response spectra.
c.
Typically, the supports with grouted base plates will not experience up-lift (tension load). The piping systems have been analyzed for operability by omitting these supports in the analyses. However, in response to a NRC question as to up-lift (tension) loads in any support, the licensee expressed inability to provide definitive information as the attendees were not familiar with the procedure of analysis used by pipe stress analysts, and stated that it will be confirmed later.
d.
.Because the code piping allowable stresses for the safe shutdown earthquake (SSE) are higher, if stresses due to the operational basis earthquake (0BE) are within the code allowable limits for OBE, they automatically meet SSE limits.
e.
In response to a NRC question as to the number of supports with expansion anchor bolts, the licensee stated that there were more than 828 such supports. They also stated that each individual support had been analyzed _for a factor of safety, and the concrete expansion anchor bolt test results did not indicate any need for expanding the data base. A 95% confidence level that less than 5 percent defective anchors are installed in any one seismic Category I system, was achieved as required by the Bulletin.
f.
The licensee stated that the appearance of an excessive number of inadequate supports reported on LERs were due to the same LERs covering the Bulletins 79-02 and 79-14, whereas the final report covers only IEB 79-02.
g.
The anchor bolt strengths were evaluated using 3000 psi concrete whereas most locations had actual concrete strengths of at least J
5000 psi.
h.
The licensee stated that supports will be upgraded to provide factors of safety of at least greater than 2 against OBE and greater than 1 against SSE before the restart.
4.
NRC Position-The NRC position as summarized by Mr. T. T. Martin was as follows:
v
4 a.
NRC staff present at this meeting disagreed with the licensee with respect to required factors of safety (i.e., the previous general licensing basis addressed in the FSAR, which required maintaining systems functional, and IE Bulletin.79-02 specified factors of safety were r.ot inconsistent).
b.
The question of testing grouted base plates was unresolved, c.
-SSE sh'ould have been used in developing loads to determine the factor of_ safety for supports.
d.
Reaion I would be contacting the. licensee subsequent to this meeting to resolve these concerns.
Enclosure:
Copy of Material used in Licensee Presentation t
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W" TMI-1 79-02 PROGRAM APPROACH INTRODUCTION:
e IE BULLETIN 79-02 + REVISIONS e
TMI-1 FSAR BASIS PROGRAM APPROACH:
1.
INSPECTION PROGRAM A.
OBJECTIVES B.
ITEMS INSPECTED C.
ACCEPTANCE CRITERIA II.
TEST PROGRAM A.
TENSIAL IEST LOAD B.
ACCEPTANCE CRITERIA Ill.
LOAD COMBINATIONS IV.
ANALYTICAL METHODS A.
PRYING FORCES B.
FACTOR OF SAFETY O-e
V.
SCREENING CRITERIA A.
SHORT TERM MODIFICATIONS B.
LONG TERM MODIFICATIONS C.
NO MODIFICATIONS Yl.
RESULTS A.
SHORT TERM MODIFIC'ATIONS B.
LONG TERM MODIFICATIONS C.
NO MODIFICATIONS VII.
ESTIMATEDSRiETYP.ARGINS I
A. ORIGINAL S'AFETY MARGIN CURRENT CRITERII, ', -
B.
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IE Bulletin No. 79-02 (Revision 2)
Date: November 8, 1979 Page 2 of 8 Action to be Taken by Licensees and Permit Holders:
This Bulletin addresses those pipe support base plates that use concrete expansion R1 ancho~r bolts ir. Seismic Category I systems as defined by Regulatory Guide 1.29,
" Seismic Design Classification" Revision 1, dated August 1973 or as defined in the applicable FSAR.
For older plants where.Scismic Category I requirements did R1 Jnot exist at the time of licensing it cust be shown that piping supports for R1 safety related systems,.as.de. fined _in_the. Final Safety Analysis Report, meet R1
'd s'ign requirements.
R1 The revision is not intended to penalize licensees who have already completed some R1 of the Bulletin requirements.
In those instances in which a licensee has com-R1 pleted action on a specific item and the Bulletin revision provices more conser-R1 vative guidance, the licensee should explain the adequacy of the action alreacy R1 performed.
It should be reiterated that the purpose of the Bulletin actions R1 are to assure operability of Seismic Category I piping systems in the event of a R1 seismic event.
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FihAL SAFETV ANALYSIS REPORT L
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' Safety Analysis Report follows the established guidelines published by the Divisi6n of Reactor Licensing, "A Guide for Organt:ation and Centents of Safety Analysis Reports" dated June 30. 1966 and the applicable portions of the USAEC Rules and Regulations 10 CFR, Part 50, " Licensing of Production and Utilization Facilities."
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Power piping vill comply with 3 31.1.0 - 1967.
Nuclear piping is being de-signed in a: crdance vi:h 3 31.1.0 - 1967 but is being fabricated, :ested and inspected to 3 31.7 - Feb 1963 ilasue:1 for Trial and C:n.ent) by 31, 32, and 33 classification.
'";ca he issuing Of 3 31.7 - Oraft, which cecurred after r neipt of the construction pe.~it, 3 31.1.0 was i=plimented by specifying s.
Nuclear Piping for fabrica:ica, tes.ing, an:1 inspectica o 3 31.7 - Feb.1968-
,u, Draft vr.ich replaced and improved on the 3 31.1.0 Huclear C de Cases.
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" Nuclear" Piping in the 3 31.7 - 1968 Oraft code is interpreted to be piping that ner-C 'y ::::ains a ridicactite substance.
Nuclear valres vere spec'.fied to be tested and inspected to 3 31.7 - 1968 Draft.
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.SEIS:ECITY g 7 0s-rzer s s u me The seismicity analysis indicated that Pennsylvania is relatively inactive seismically, based upon 200 years of historical data and 20 years of instru= ental data.. Earcquakes in the greater Pennsylvania area, which have or =ight have affected the site, were studied and their intensity at the site was determined by attenuatica of the earthquake with distance.
Earthquakes which have affected *he site were studied in tvo categories,
those within a 50-cile radius of One site, and these beyond this radius.
Nearly all the earthquakes censidered were felt ever very li=ited areas,
which are generally elliptical in shape, and aligned with the general structural trend of the area.
The high attenuation of these earthquakes indicates that their foci =ust have been close to the earth's surface.
An estinate of the maxi =u= expected intensity of an earthquake was predicated on the assu=ptica that the activity which would affect the 4{ site would criginate along the bcrder fault of *he Triassic Lculand, five to six miles north of de site. The highest intensity earthquake to occur on this fault has been =cdified Mercali */I.
'"he intensity of such an earthquake at :he site would be 7, based upon the rapid atten-Ccnsider,-7)J
' untion of similar earthquases in the area and along te fault.
ation was given; hcuever, to the future occurrence of an earcquake at the a greater depth in the fault, with a conservative assu=ptica that resulting intensity at the site could apprcach the epicentral intensity, j and not be rapidly attenuated.
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. e conservative esti= ate of the =axi=u= ear 2 quake intensity to be expected at the site is a icv intensity 7I.
Using relatienships pub-
,}ished in :;uclear Rese crs and Earthauakes " O 7 24, "nited States A c=ic Energy Cc==issien, this intens:. y corres;cnns Oc 2 g cund accelera:ica D of 0.0hg.
~he design is censervatively based en a basic grcun:1 =ctica d of 0.06g =ax1=u=.
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p 2.6.2 ACCELERATIC:( RESPC:ISE SPECTRA e objective of this study was to establish an acceleration response il spectra correspending to a possible ear tquake of icv intensity 7I (Modified Mercali), as established by the seis=icity analysis.
Figure 3 -24 indicates the average s=coth response spectra derived frc= the
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ground =cticns of the 1957 Golden Gate Park, San Francisco ear:nqusae together with the revised acceleratica spectra reflecting de greater response at lever frequencies cased upon the 19ho Il Centro spectra.
The resulting acceleratien respense spectra (Figure 2-2L) vere developed as described in Section 2.7.1
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5 1.2.1.2 i3 Components and systems classified as Class I have been designed in accordance with the following criteria:
Primary steady state stresses, when co$bined with the seismic stress resulting from the response to a ground acceleration of 0.06g acting horizontally and 0.0hg acting vertically and occurring simultaneously has been maintained within the allowable working stress limits accepted as good practice and, where applicable, set forth in the appropriate design stands, e.g., ASME Boiler and Pressure Vessel Code, USAS B31.1 Code for Pressure Piping.
Primary steady state stress, and corresponding strains, when co=bined g31tr with the seismic stress resulting from the response to a ground horizontally and 0.08c acting vertically acceleration of 0.12g actin $
and occurring simultaneously, has been limited so that the function of the component, system, or structure is not impaired as to prevent a safe and orderly shutdown of the plant.
Stresses resulting from the simultaneous occurrence of the maximum ee.rthquake and the loss-of-coolant accident shall be limited to permit a safe shutdown of the plant. Refer to Section h.
For piping stress criteria refer also to j g ection 5.4.h.
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Tae folleving da= ping factors have been applied to the maxi =um earth-quake in the seismic design of ce=ponents and structures:
Percent of Cemeonent or Structure Critical De=;ing 1.
Reactor Building 2.0 2.
Concrete SuPc.or: Structures Inside the Reactor Building 2.0 3.
Assemblies and Structures (a) Bolted or Riveted 2.5 (b) Welded 1.0
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Other Concrete Structures above ground 5.0 M
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Lu= ped-Mass Analysis The system is represen cd by a series of concentrated = asses.
First, the space coordinates ire established f:: the systes, and :: ordinates ef = ass points are deter =inel.
Using s stati: analysis, flexibill:7 =atri:es corresponding : these = ass ;cin s sre ce=puted.
Hext, the equati:ns of motion are vri ten in = atrix f:r=.
Responses of =cdes vi-h frequencies 10 percent within each ::ner, are sdded sbsolutely.
Then the square ::c -of-the-su=-of-the-squares ep;roaca is aken for -he res; of the =cdes.
Since the absolute addition cf the ci se = odes =eth0d vas incor;cra ed into -he computer progra= after certain per;icas of the pipe sys:c=s vere run,1 eq, general check of these syste=s were =ade.
The effect of absolute additi:n 1 increases the res; case less than 12 percent.
Hence for conservative reasca, the responses of these syste=3 vi h absolute addi:1:n are =ul-iplied by s j
factor of 1.12 to secount for the effe :.
A =cre ec=plete description of the analytical =etheds e=;1:yed is contained in Gilber: Associates, :::,,
Topical Reper: No. 1729, entitled "Oyna=1: Analysis of Vital Piping Systems Subjected to Seissi: Moti:n." The application of the 1=;11tication curves cited in this re;crt, is explained in Professer J. M. Siggs and J. M. acesse:'s r
i paper entitled " Seismic Analysis of Iqui;=en: Mounted :n a Massive Structure.
In the paper by Professcrs 31ggs and Rcesset, the single degree-of-freedc= =cdel was not used to ec=;sre with the =ulti-= ass syste= directly.
- stead, it was used to ec=;are vi:h one of the nor 21 : odes of :ne :ulti-casa system.
This is the standard procedure in the =ethod of scr=al = ode.
Iven with I
the time history 2;;rcach the sa e ec=parisca is used, if the nor=al =cde
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method is e=;1oyed.
An approxt=ste ec=parisen of our ricor response :urves with these derived from ti=e history ne: hod (13, IL), indicated that the
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method of Professors 31ggs and Reesse is no less conservative than the j time history =ethod.
Eovever, this fact was recently verifie1 by inalyzing h the lu= ped = ass =cdel of the reactor building interior concrete as sncvn s in Figure 5 47 for:
Ep l's El Centro normalized tL=e history method a.
A wax iAt b.
Golden Gate nor=ali:ed time history =ethod C"W
// The resulting uns=cothed ground res; casa spectra are shown in Figure 5 h8 and co= pared with the Three Mile :sland ground response spectru=.
- is interesting to note tha: the Golden Gate spectra has a low res;cnse s the j
period of the funda ental = ode of the building (0.09 sec).
This in turn results in the relative large difference in -he ficor response spectra as shown in Figures 5 L9, 5-50, 5-51, 5-52, and 5-53 The dotted curve shcws the ficor res;0nse using Biggs ind Roesset's =e: hod with the uns=cothed Golden Gate res;cnse s;e:;ru= ss sn:vn in Figure 5-LS.
Syste= and c =;onen; masses less th:= ten percen Of the structure = ass vere not included in the d
dynamic analysis of the structure.
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Q'l In addition to the earthquake response for the pipe systes, the =odels described shove vere used to deter =ine forces and =c=ents with restiting stresses for any trsnsient or per=anen; displace =ents whi:h were induced
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INSPECTION PROGRAt1 A.
0JBECTIVES VERIFY DESIGN) REQUIREMENT SATISFIED (BO 1.
CONFIGURATION 2.
VERIFY INSTALLATION REQUIREt1ENTS SATISFIED (I.E. EMBEDMENT)
B.
ITEMS INSPECTED SELFHEAD)DRI(LING SHELL TYPE ANCHORS (FFG. By LTT PH NED-WERE.THE ANCHOR BOLTS USED IN I M HE FOLLOWING WERE ITEMS INSPECTED (DER SAMPLING TECHNIQUES DESCRIBED IN BULLETIN):
- 1...IZE OF ANCHOR AND BOLT / PLATE CONFIGURATION S
2.
PLUG DEPTH 3.
ANCHOR PROTPUSION OR RECESSION IN RELATION TO THE CONCRETE 4.
THREAD ENGAGEMENT OF BOLT OR STUD 5.
PLATE HOLE SIZE VS ANCHOR DIAMETER 6.
SHELL TO PLATE GAP 7.
SHELL LENGTH.
C.
ACCEPTANCE CRITERIA BASED ON INGPECTION PAGAMETER COMPLIANCE OR NON-COMPLIANCE (DEVIATIONS 1 ANCHORS WERE SUBJECTED TO TEST LOADINGS IN ITEM ll.
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TEST PROGRAM A.
TENSILE IEST LOAD:
1.
NO DEVIATIONS PER INSPECTION PROGRAM.
A.
IENSILE TEST TO A MINIMUM OF 20% OF THE MANUFACTURERS ULTIMATE LOAD.
2.
DEVIATION OCCURRED PER.INPECTION PROGRAM.
A.
IENSILE TEST TO A. MINIMUM 0F 40% OF THE MANUFACTURERS ULTIMATE LOAD.
B.
ACCEPTANCE CRITERIA l.
BOLT ACCEPTED THE TEST LOAD (20% OR 40%)
0F MANUFACTURERS ULTIMATE).
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III.
LOAD COMBINATIONS
-SUPPORT LOADS WERE GENERATED AS AN OUTPUT OF A DYNAMIC
-PIPING ANALYSIS CONSIDERING THE GOVERNING LOAD COMBINA-TIONS AS FOLLOWS.
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DEADWEIGHT + THERMAL + DBE* SEISMIC + OCCASIONAL MECHANICAL LOADS = IOTAL DESIGN LOAD.
- OBE - OPERATING BASIS EARTHQUAKE = DESIGN BkSIS EARTHQUAKE (DBE) AS DEFINED IN FSAR.
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IV.
ANALYTICAL METil0DS A.
PRYING FORCES (BASED ON "AS INSTALLED" CONDITIONS)
WERE CONSIDERED IN THE FACTOR OF SAFETY DETERMINATION BELOW.
B.
FACTOR OF SAFETY THE FACTOR OF SAFETY (F.S.) WAS DETERMINED USING THE FOLLOWING SHEAR-TENSION INTERACTION EQUATION.
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'TO = ULTIMATE TENSION CAPACITY OF THE ANCHOR VA = SHEAR FORCE ON ANCHOR VO = ULTIMATE SHEAR CAPACITY OF THE ANCHOR.
- REDUCED TO 40% OF M NUFACTURERS CAPACITY IF MINOR A
DEVIATIONS EXISTED.
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V.
SCREENING CRITERIA A.
PRIOR TO RESTART ANCHOR BOLTS WITH SAFETY FACTORS (S.F.)< 2 WILL BE REDESIGNED AND REPAIRED '(SHORT TERM MODS),
B.
AFTER RESTART ANCHOR BOLTS WITH S.F.'S BETWEEN 2 AND 5 WILL BE REDESIGNED AND REPAIRED (LONG TERM MODS).
C.
ANCHOR BOLTS WITH S.F.'S 25 REQUIRE NO ADDITIONAL' EVALUATION (NO MODS).
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VI.
' RESULTS A.
PRIOR TO RESTART ANCHOR BOLTS WITH SAFETY FACTORS (S.F.)< 2 WILL BE REDESIGNED AND REPAIRED (SHORT TERM MODS).
- 1. - 479 CASES OR 17.9% OF TOTAL.
B.
AFTER RESTART ANCHOR BOLTS WITH S.F.'S BETWEEN 2 AND 5 WILL BE REDESIGNED AND REPAIRED (LONG TERM MODS).
- 2. - 185. CASES OR 6.9% OF TOTAL.
C.
ANCHOR BOLTS WITH S.F.'S2 5 REQUIRE NO ADDITIONAL EVALUATION (NO MODS).
- 3. - 2012 CASES OR 75.2% OF TOTAL.
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VII.
ESTIMATED SAFETY MARGIN AFTER 79-02 PROGRAM COMPLETION A.
ORIGINAL SAFETY MARGIN (AGAINST FSAR CRITERIA) 1.
S. F.=
5 (0BE) & S.F. 2.5(SSE) 2 s
B.
AGAINST CURRENT CRITERIA (USING NRC REGULATORY GUIDE 1.61 1.
S.F. -
5 (0BE) & S.F. - 4 (SSE)
C.
CONSIDERING ACTUAL CONCRETE STRENGTH AND CURRENT CRITERIA S.F.*- 6.5 (OBE) & S.F.4$ 5.2 (SSE) e o
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-VIII.
ADDITIONAL C0flSIDERATI0f1S A.
CURRENT CRITERIA EFFECT MARGINS NpMERICAL REE VS.AS BUILT tFFECTS 1.
REG. GUIDE 1.29 0
0 0
t 20% ON OBE 2.
SEISMIC INPUT
& SSE 3.
REG. GUIDE 1.61 1.25 OBE = SSE a
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60% ON SSE OAD 7% STRUCTURAL 2% PIPING DAMPING ONLY
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AS BUILT EFFECTS 1.
CONCRETE STRENGTH (DESIGN VS. ACTUAL) 2.
STEEL STRENGTH (Fy MINIMUM uSED vS. Fy ACTUAL) 3.
STEEL STRAIN (ELASTIC VS. PLASTIC) 4.
PIPING ANALYSIS (RIGID VS. FLEXIBLE SUPPORTS) s S
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