ST-HL-AE-1744, Forwards Annotated Revs to FSAR Sections 1.3,3.1,3.6,3.8, 3.9,3.12,5.4 & 6.2.Revs Reflect That Reactor Coolant Loop Ruptures & Associated Dynamic Effects No Longer Included in Design Bases & Will Be Incorporated in Future FSAR Amend (2024)

{{#Wiki_filter:-The Light -MEhMf flouston Lighting & Power P O. Box 1700 11ouston, Texas 77001 (713) 228-9211 September 15 1986 ST-HL-AE-174d File No.: G9.10/Cl.1 Mr. Vincent S. Noonan, Project Director PWR Project Directorate #5 U. S. Nuclear Regulatory Commission Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Annotated FSAR Changes Concerning the Rule Change to GDCJ

References:

1) HL&P Letter to NRC, J. H. Goldberg to H. R. Denton, September 28, 1983, SI-HL-AE-1010

2) HMP Letter to NRC, J. H. Goldberg to H. R. Denton, July 17, 1984, ST-HL-AE-1096

3) HMP Letter to NRC, J. H. Goldberg to H, R. Denton, March 1, 1985, ST-HL-AE 1200

4) HMP Letter to NRC, J. H. Goldberg to H. R. Denton, August 19, 1985, ST-HL-AE-1326

5) NRC Letter to HL&P, T. M. Novach to J. H. Coldberg, June 26, 1985, ST-AE-HL-90645

6) HMP Letter to NRC, M. R. Wisenburg to V. S. Hoonan, March 7,1986, ST.HL-AE-1618

7) NRC Letter to HMP, N. P. Kadarabi to J. H. Goldberg, May 8, 1986, ST-AE-HL-90886

Dear Mr. Noonan:

The Houston Lighting & Power Company (HMP) via References (1) through (2) provided technical justification in support of our request for a partial exemption to the General Design Criteria (CDC-4) regarding tha treatment of Reactor Coolant System (RCS) pipe breaks inside containment at the South Texas Project Units 1 and 2. Reference (3) was a pro forma rraquest that the Construction Permits for the South Texas Project Units 1 and 2 (~:PPR-128 and CPPR 129 respectively) be amended to authorize a partial exception from GDC-4 to permit the elimination of postulated circumferencial ar.d longitudinal breaks in the RCS main loop piping and the acscriated dynamic effects from consideration in the structural design basis of the South Texas Project.Reference (4) modified the Construction Fermit amendment request such that the duration of the exemption requested would date from the day of the issuance of the amended Construction Permit until Startup following the second

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E609180307 G60915 C#V 1 Ll/NRC/br- PDR ADGCK 05000473 ( \A PDR

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ST-HL-AE-1744 [File No.: G9.10/C1.1 :Page 2 ;refueling outage as opposed to the end of plant life. This modified request was reluctantly submitted as s' result of Reference (5). Reference (6) !I summarized from References (1) thru (4) the specific requirements that would be eliminated based on NRC approval of our request and informed you that since 'we 'were confident that prior to Unit 1 fuel loading either the Commission -would issue a final rule modifying the GDC-4 requirements for protection ;against the dynamic effects of postulated pipe ruptures or ett exengtien

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request would be approved, we were taking action to remove the pipe whip l restraints on the RCS main loop and cross-over piping, ,By letter dated May 8,1986, (Reference 7) the NRC informed HL&P thet i since the Commission had approved the rule change to CDC-4, N2C action on car- ;previous exemption reques't had become unnecessary and indicated that the next action by the NRC staff would be the evaluation of the design changes ,implemented at the STP after receipt of the FSAR changes frok HL&P, l Attached are the annotated FSAR chan6es to Secti6ns 1.3, 3.1, 3.6, 3.B.3.9, 3.12, 5.4, 6.2 and various NRC question recycnses in the STP FSAR. These e changes identify that reactor coolant loop (RCL) ruptures and the associated !dynamic effects are no longer included in the design _ bases and-will be -included in a future FSAR amendment._ HIAP requests an expedited review of the ;changes and a meeting to facilitate the review as soon as possible.If you should _have any questions on this matter, please centact-Mr. M. E. Powell at (713) 993-1328. ,Very t 14yours,

 . f \ ,

e i

 , )M M.R(.Vi nburg M -

Manager, 1" ear Lice sir.g MEP/yd

 '( , . e Attachneut: Annotsted FSAR Changes Concerning Compliance to GDC-4 (FSAR Sections 1.3, 3,1, 3.6, 3.8, 3.9, 3.12, 5 4, 6.2, and Responses to NRG questiens 210.20N, 220.27N, 220.29N, 22.02, 221.1,

. 480.04N,- 480.05h, 460.0CN, 480.0*/M, 480.08N, 480.0SN and 44G.12N) i? (4

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I I; a .I-{ r.1/NR&l/ba ti s_ .

7.R Ifouston Lighting & Power Company ST HL AE-1744 File No.: G9.10/C1.1 PaSe 3 cc:Hugh L. Thompson, Jr., Director Brian E. Bervi'ck,.isquire Division of PWR Licensing - A Assistant Attorney General for Office of Nuclear Reactor Regulation the State of Texas U.S. Nuclear ' Regulatory Commission P.O. Box 12548, Capitol Station Vashington, DC 20555 Austin, TX 78711 Robert D. Martin Lanny A. Sinkin Regional Administrator, Region IV Christic Instituto Nuclear Regulatory Commission. 1324 North Capitol Street 611 Rysn Plaza Drive, Suite 1000 Washington, D.C. 20002 Arlington, TX 76011 Oreste R. Pirfo, Esquire N. - Prasad Xadambi, Project Manat,er Hearing Attorney U.S. Naclear Regulatory Commission Office of the Executive Legal Director 7920 Norfolk Avenue U.S. Nuclecr Regulatory Commission Bethesda, MD 20814 Washington, DC 20555 Cirudo E. Johnson Charles Bechhcefer, Esquire Sen'lcr Resident Inspector /STP Chairman, Atomic Safety &c/o U.S. Nuclear Regulatory Licencing Bsard Consission U.S. Nuclear Regulatory Cosaission P.O. Box 910 Washington, DC 20555~Bay City. TX 77414 Dr. James C. Lamb, III M.D. Schwarz, Jr. , Esquire 313 Woodhaven Road Baker & Botts Chapel Hill, NC 27514 One Shell Plaza Houston, TX 77002 Judge Frederick J. Shon Atomic Safety and Licensing Board J.R. Newman, Esquire U.S. Nuclear Regulatory Concission Newman & Holtzinger, P.C. Washington, DC ,20555 1515 L Strest, N.W.Washington, DC 20036 Citizens'for Squitable Utilities, Inc, c/o Ms. Peggy Buchorn !Director, Office or Inspection Route 1, Box 1684 and Enforcezent Brazoria, TX 77422 U.S. Nuclear' Regulatory Ccamission Washington, DC 20555 Docketing & Service Section Office of the Secretary T.V. Shockley/R.L. Range U.S. Nuclear Regulatory Commission Central Power & Light Company Washington, DC 20555 P.O. Box 2121 (3 Copies)Corpus Christi, TX 78403 Advisory Committee on Reactor Safeguards H.L. Peterson/G. Pokorny U.S. Nuclear Regulatory Commission City of Austin 1717 H Street P.O. Box 1088 Washington, DC 20355 i Austin, TX 78757 J.B Pcston/A, vonRosenber.g City Public Service Buard P.O. Box 1771 San. Antonio, TX 78296 Ll/NRC/ba Revi. sed 5/22/S6 b _

7ARI.].*t-2

 $KNIFIC/*T DPitfN dunn

• P&femntw Iree FW Ibecripefm of Owye Cold slurtdown axifficationn O mpters 3, 5, 6, 7, 9 Mxlificatiom to rwetor vennel hen! vent syntaa to provide a safety-grade letdwn.psth to the litr. Accoubtor vent valve sodificattom. CVt3 aryli-fIcatian to provide safety-grade boratim system awl other auxiificationr. 45 ricod protectim Section 3. 4 Categpry I,oht tem are prtterrei qrtirut_finoding f by ta tertight doers.

J^uu ast;-i.' .Pipe break criteria Sectim 3. 6 (cmrew n re .pi' alh so.r..irrun

 ~ ~

seet er a.a* 4 wo ) 'of the rem) tEint1EiiIt The tiens ofpresent design Branch Tecletcal bania pipe brd")fB Pbeitim 3-1[dtid, ry_ #thearoc /gE#T We interpreestirwtif the repairtswrtts of X)-1;46.fAs irrlicatsvi in Section 3.6.2.1.1 avl Table 3.6.1-3, incenediate taesk focations are limit:wl to tie sense determfrvul breaks awl tentfn,11 esis, i.e., no arhi-y_ tr:rry i.tenedtate n biens ' '. ^2 ' " . ,_^ ' arv poesisted.These pipe breeks are tased in the snelysin c,f ==Ju-vu^-4 pressurization,(y 4 jet inedneysamt as! pipe whip effect=, r

 %heryntes swlysis 5ctim 3. 7. 3 14dtip1m sp!ctre beve been used to twSxe hve us sv.zia in iaase '

cases. This che is foised on current scate of the art and availability of mqnster rtogram.Mr<iroomrotal Qn11f featte Section 3.10, 3.11 "Anriard p ogr e.Mr1 Ompter 4.0 Ounge fms 9 grid to 10 gridi.Thfnias centrol rods Sections 4.2, 4.*i Ownnes tre of either &IrKif or gC control rods tr* Thiniin reds.Fedice rtrivr of fitil-lergth Secticrs 4.3 Revised rtyndato y criteris have enA these enr.t el reds usserenne:y. ML/)g t 099 ftun 61 tn 57 ,y l ok Imi follivpekage 'Sectica 4.3 Centrol r:d D-bmk avl ndm1 average L.pwnen provide retum to y,[o pcwr capability fo* Innd folim etiatinatina the neel for part-Imgt'i control rods.gg JO k3 Eliminstion of part-length Secticns 3.9.4, 8.2 Pert-length emtrol rode elfr.timtel. 4 k 4 m e ttel itxis 4.3, 7.7

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ATTACHMENT l

 .o-ST-HL- AE- n elti l PAGE d.OF1Rf7 IBMRT ' . P,. TAsLE l 3-2 Page 1.3-C 4

i As tiiscussed in Reference 3.6-14 and parag aph 3.6.E.1.1.la, RCL ruptures and the associated dynamic effects are'not included in the design bases.4[i i4 Il-i il 7603N:02B3N/1

 - . . , . . . . - . . _ - - - . - . . _ _ , - - - . . , ~ - . . . - . . - .. .- . . . . . -. - - - - - - - - . -

t .l1 iTAlf21.}-2 (Owf meO S10ftFICMf DESIm 09NQS Nferences Lee ? SAP. _Iwacripcfm cf O'.mige n- nwizer pwtal Sections 5.1,5:4,7.4 theippd from 3 POPVs to 2 PCRVs. O.w.pd frte air operated PGtVs re- safety , -relief valves niated, eolencid-oct.ated ITF.'s, iRea:tw Vessel lieal Vent Systee Section 5.1 ,5.4 Mditim cf Pesetoe Veseel Hend Vmt Sate. F Rc. t. W b;" .3.A c. . a cGyss@ ff'_:', ;.c,Uut.h ply. _miswirit-pevvidenMWs,,h tg l

1. if Alen.4mpt pfpe sia ints z

 . w rt. i-w ; i; i Reactor elet p'pirig Section 5.4 'themal slee e: in reactor coolat 'Icnp immch :iozzles leve beer, deleted. l Cold dutdown r apability per MP Sect"on Appendix 5.4.A Maress Bl? RSB 5-1, aklition of safety-refatal ispd.111ty for M3 let&..

pwizer wntiris, stesse generator peeperstal relief and otler redi- }PSB 5-1

 - fications. F3 n,>

a ilydttyn tendirrrs Sectim 6.2.5 dienged to WestL.gwx mippifed I reca61wrs located iriside conteireeit.L.g _y Moration Systes Section 6.8, 7.3 EBS deleted.k r tor trip on turbine trip Sectim 7.') Oisne fma F-7 (10% pauer) to P-9 {50% ywr).interlock Nartow respe stessi ppnerator Section 7.2 mesuresents are mapematal for temperature effects on referenz fluid wit er level inessuttner.ts h g for SIP. g '04-4 Ini fce%siter flew signal for Section 7.2 low ferduater flow sips 1 for renetor trip has been deleted.tuctor trip o- m

 4 2 l &~4 3

h N Q a -o.E7 b,

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ATTACHMENT

 .- ST-HL.AE- 19 44 PAGE 4 OF d 1 INSEP.T TABLE 1.3-2 Page.l.3-9 ' As discussed in Reference 3.6-14 and paragraph 3.6.2.1.1.la, RCL ruptures and tiie issociated dynamic effects are not included in the design bases.

Therefore, RCL pipe restraints are no longer required.t i1 7603N:028SN/2 j

STP FSAR ATTACHMENT ST-HL-AE 1744 PAGE 5 OF la.O Review Report Status Section Cadek, F. F., S. Cerni, J. M. Hellman, W. J. Leech, U 4.2 and J. R. Reavis, " Fuel Rod Bowing," WCAP-8691 4.3 (Proprietary) and WCAP-8692 (Nonproprietary),December 1975.SNSERi 4.I

 )

, 1.6-2a Amendment 46

ATTACHMENT ST-HL E I'7Vl/

 ##9

• INSERT .

Page 1.6-2a Swamy, S. A., Lee, Y. S., Clark, H. F. and Holmes, R.A., B 3.6" Technical Bases.for Eliminating Large Primary Loop Pipe Ruptures as the Structural Design Basis for the South Texa's Project Units 1 & 2", WCAP-10559 (Proprietary) and WCAP-10560 (Non-Proprietary).7603N:0288N/3

~STP FSAR ATTACRMENT &+ .~ "17 ST H L-9 A,_f 0F?PAGE LIST OF TABLES (Continued) k CHAPTER 3 Table Title Page 3.6.1-2 Design Comparison to Position of NRC Branch Technical Positions ASB 3-1 3.6-29 3.6.2-1 High-Eaergy Pipe Break Initial Stress Summary 3.6-35 3.6.2-2 High-Energy Pipe Break Effects Analysis Results -0. u- 4 3 40 3.6.2-3 ru.iol.ica "rc;L L.; tir.; f;r :he LOC #3 *

 .c.;1,.i. uf ihm T.122 j C::L..; L^Cr is:6=*t 3.6.2-4 ~) Prix:.,.aus acconcary cos... Iui ..:ity Ranges-and-C-netrbreive S;;; Faiuia et-Bestgr. L ook '.~;ati:n in F.::: er f;;l-"' Loei 3.6-46 ][sterED 3.7-1 Damping Values 3.7-45 3.7-2 Method of Seismic Analysis Used for Category I Structures 3.7-46 3.7-3 Reactor Containment Building Natural Fre-quencies 3.7-47 3.7-4 Mechanical-Electrical Auxiliaries Building 34 Natural Freqencies 3.7-48 3.7-5 Diesel Generator Building Natural Frequencies 3.7-49

( 3.7-6 Fuel Handling Building Natural Frequencies 3.7-50 3.7-7 Damping Values Used for Seismic Analysis of NSSS Equipment 3.7-51 3.7.A-1 Final Analysis Cases, Soil / Structure Inter-action Studies 3.7.A-6 3.7.A-2 Calculated Dynamic Pressures on Tendon Galleries 3.7.A-7 3.7.A-3 Calculated Base Shear Forces on Buildings 3.7.A-8 3.8.1-1 Load Combinations for Containment Structure 3.8-91 3.8.1-2 Summary of In-Process Test Results, Cement 3.8-92 3.8.1-3A Summary of In-Process Test Results, Sieve Analysis and Fineness Modulus, Fine Aggregate (Sand) 3.8-93 3.8.1-3B Summary of In-Process Test Results Fine Aggregate (Sand) 3.8-94 3.8.1-3C Summary of In-Process Test Results, Sieve Analysis, Coarse Aggregates 3.8-95 3.8.1-3D Summary of In-Process Test Results, Aggregate No. 67 (3/4" Gravel) 3.8-96 3.8.1-3E Summary of In-Process Test Results, Aggregate No. 4 (1-1/2" Gravel) 3.8-97 3.8.1-4 Summary of In-Process Test Results, Water 3.8-98 3.8.1-5 Summary of In-Process Test Results Admixtures 3.8-99 3.8.1-6 Summary of In-Process Test Data, Concrete CP-244 Results 3.8-100 l1x Amendment 40

ATTACHMEN )ST HL AE-] 4 PAGE P OF / A)STP FSAR LIST OF F7 CURES (Continued)CHAPTEE 3 Figure Reference Number Number Title 3.6.2-2 h vi L.21:nd " e:M h .":!."x p S 3.6.2-3 Typic.1 U-Bar Restraint t- 3.6.2-4 Typical EAM Restraint 3.6.2 5 Imedown Heat Exchanger Compartment Pressure Prefile 3.6.2-6 latdown Heat Exchanger Compartment Temperaturc Profile 3.6-3 Single Pipe Penetration for High Energy Lines 3.6-4 Multiple Pipe renetration (Deleted) 3.6-5 Single Pipe Penetration for Moderats Energy Lines (Deleted) 3.6-6 All Austentic Stainless Steel and Nonferrous Piping and All ASME Section III Ferritic Steel Piping (Deleted)

3.6-7 Backing Strip Wald Prep. Details (Pipe) (Deleted) 3.6-8 open Butt Weld End Prep. (Non-Nuclear Ferritie Steel Pipe or Plate) (Deleted) 3.6-9 Mathematical Model of Structure (Deleted) 3.6-10 Mathematical Model for MS & FW Pipe Rupture Restraints (Deleted) 3.6-11 Typical Pipe Restraint (Yielding Member Type) (Deleted) 3.6 12 Typical Pipe Restraint (Crushable Pad Type) (Deleted) 3.6-13 Typical Pipe Whip Restraint Configuration (Delete 6) l 3.7-1 Horizontal SSE Design Response Spectra 3

3.7-2 Horizontal OBE Design Response Spectra I3.7-3 Vertical SSE Design Response C;ectra jl 3.7 4 Vertical OBE Design Response Spectra i3.7-5 Artificial Time Histories, Horizontal & 'Vertical SSE j3.7-6 Horizontal Response Spectra, SSE, 14 Damping~TC 3-35 Amendraent 54 ,

 . . -, ._ _ ._ _ _ __.._ _ __ _ . ___ ,. _ -. __ . _ _ _ _ __ _ . ~ . - - - _ . . . . . . -

ATTACFUAENT ST HL-AE- n4 STP FSAR PAGE 4 OF / 9 Equipment and facilities for fire detection, alarm an . ...quishment are provided to protect both plant and personnel from fir: r:::plosion., Fire Protection and Detection System reliability is ensure; - periodic tests and inspections, Administrative controls are used where applicable thrcu;hout the facility to minimize the probability and consequences of fires or explosions.The Fir'e Protection System is designed such that a failure of any component of the system:e Will not cause an accident resulting in significaat release of radioac-tivity to the environment.e Will not impair the ability of redundant equipment to safely shut down the reactor or limit the release of radioactivity to the environment in the event of a LOCA.For further discussion, see the following sections of the FSAR.Materials, Quality Control and Special Construction Techniques 3.8.1.6 Independence of Kedundant Safety-Related Systems 7.1.2.2 l 33 Independence of Redundant Systems 8.3.1.4 Fire Protection System 9.5.1 3.1.2.1.4 Criterion 4 - Environmental and Missile Design Bases: S truc-tures, systems and components important to safety shall be designed to accom-modate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated acci-dents, including LOCA's. These structures, systems and components shall be appropriately protected against dynamic effects, including the effects of mis-siles, pipe whipping and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. j{~*%-- l A/ EA9~3.1.2.1.4.1 Evaluation Against 33 systems, and components are design!d,C*3_tig

 - e immodate the loneffects 4 - Safety-related of and to be structures, compatible with the environmental ac Jty , a (including the pressure, tempera-ture, humidity and radiation conditions) associated with normal operation, maintenance, testing, and postulated accidents, including LOCAs. Protection criteria are presented in Sections 3.5 and 3.6 and the environmental condi- l 33 tions are described in Section 3.11.

These structures, systems, and components are appropriately protected against dynamic effects, includinc the effects of missiles, pipe whipping, and dis-charging fluids that may result from equipment failures and from events and conditions outside the nucitar power unit. Details of the design, environ-mental testing, and construe: ion of these s> tems, stru::ures and components are included in other secti:ns of the FSAR:I L N6Egi 2 3.1-4 Amendment 35

ATTACHMENT ST-HL AE./1dd PAGE /0 OF 'id 9 Section 3.1.2.3.4 Insert i However, the dynamic effects _ associated with postulated pipe ruptures of primary coolant loop piping in pressurized water reactors may be excluded from the design basis when analyses demonstrate the probability of rupturing such piping is extremely low under design basis conditions.Insert 2 The dynamic effects associated with postulated ruptures in the RCS main loop piping are shown to be of extremely low probability of occurring under design conditions and are not included in the design ~ basis.7603N:0288N/4

ATTACHMENT STP FSAR t ST PAGE HL-AE

 // OF f 7lH/a/

Water Le . Tiood)' Design 3.4 l33 Missile Prcn.: tion Criteria 3.5 Criteria for Protection Against Dynamic Effects Associated with Postulated Rupture of Piping 3.6 Design of Category I Structures 3.8

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Mechanical Systems and Components 3.9 33 Seismic Qualification of Seismic Category 3.10 I Instrumentation and Electrical Equipment Environmental Design of Mechanical and Electrical Equipment 3.11 Independence of Redundant Safety-Related Systems 7.1.2.2 Independence of Redundant Systems 8.3.1.4 Accident Analysis 15.0 l33 3.1.2.1.5 Criterion 5 Sharing of Structures, Systems or Components:

 -Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not s significantly impair their ability to perform their safety functions, inclu-ding, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

3.1.2.1.5.1 Evaluation Against Criterion 5 - The ultimate heat sink is the only shared safety-related system.- l33 For further discussion, see Section.9.2.5.3.1.2.2 Group II - Protection by Multiple Fission Product Barriers (Criteria 10-19).3.1.2.2.1 Criter' ion 10 - Reactor Design: The reactor core and associ-ated coolant, control, and protection systems shall be designed with appropri-ate margins to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.3.1.2.2.1.1 Evaluation Against Criterion 10 - The reactor core and asso-ciated coolant, control, and protection systems are designed with adequate margins to:

1. Preclude significant fuel damage during normal core operation and opera-tional transients (Condition 1)* or any transient conditions arising fro:r occurrences ef moderate frequency (Condition II)*.

k

• Defined by ANS: ': .. - 1973 NY 3.1-5 Amend c-.

STP FSAR ATTACHMENT ST HL AE-/W PAGE isLOF / ')Question 210.20N In order to assure that the pipe break criteria have been properly implement-cd, the Standard Review Plan requires the review of sketches showing the pnstulated rupture locations and of summaries of the data developed to select postula.ted break locations including, for each point, the calculated stress intensity, the calculated cumulative usage factor, and the calculated primary plus secondary stress range. The required sketches and tables for some high cnergy piping systems have not been provided at this time in the FSAR.Provide a schedule for submission of these data.pJst W

Response

Initial stress summaries regarding pipe break locations, stress levels, 53 cumulative usage factors are shown in Table 3. Sketches showing pipe breaklocationsareprovidedinFigure3.6.1-1g2-1. 4 Final design information, including as-built reconciliation, will be provided prior to fuel load.

 ~ ,- -

M_ outh l==.-Fr#3act TSTEWs--submitted =a=reFquess to G. K O-fer r g Men P to teneral_ Design Criterion'4' in order to delete postulation of Reactor

 ~ ~l Coolant Loop (RCL) pipe-breaks based upon the "Imak Before BreaE" analyses. ~

This has been justified .in WCAP-10560. (Refer s to NRC letters

 !T-HL-AE-l'D10. dated Septembers.28, 1983,.STJ1L- -1096, dated July 17, 1984, !.T-HL-AE-1200 fated _ March 1, 1985'and T*111.-AE-1326, (tgust 19,1985) .

l53 i Although the41RC has not yet reJs ed'to.sthe request, 'ths project is o ufficiently confident such. thEt the current' design is-proceeding on the *

 .s s ti'on that the exemption will be granted. Thus<,{RCL pipe breaks are not os ated and p d6 formation requested is not pattinent-to.STP for that cope. Howevgr, it should be noted that primary component supports have been esigned'to withstand the structural loads associated with non-sechanistig-neactor-Coolant-pipe breaks atthe -locations -described-in WCAP-8082. l53 1

l Vol. 2 Q&R 3.6-8N Amendment 53

 - . ATTACHMENT '

ST-HL AE- > l INSERT E AGE L6 df9 hd Pl lQ210.20N. l I

 .As discussed in Reference 3.6-14 and paragraph 3.6.2.1.1.la, RCL ruptures and the associated dynamic effects are not included in the design bases.

r 7603N:0288N/5 l

ATTACHMENT STP FSAR ST.HL-AE- lW4 PAGE 14 OF iaq Question 210,2)N C~ Provide a listing of those postulated pipe breaks where limited displacements have been used to reduce break areas.

Response

[efer to t last paragrap of the response Q 10.20N. Pipe breaks were q postulated in ecordanc ith WCAP-8082 for the esign of ppdnary components.Limited non-mech st pipe breaks were used in tla des of primary compor .-ent supports.For the calcuJ d ions of loo hydraulics for prima componen upport designs ,reduced break areas (150 in ) ereusedforthegeactorvesselinletand reactor vessel outlet br =h Limited displacements have not been used to reduce break areas for h 2..n af phervt piping on STP.C tl lVol. 2 Q6R 3.6-15N Amendment 50 t1 g . . . .. . . . . . . . . . . . _ . . - . . . . - . - . . . . . . . . .- . - . - .

r- -ATTACHMENT ST-HL-AE. I?yt/STP FSAR ._PAGE /5 OF /R '7 Question 210.29N Provide the loads, load combinations, and stress limits that were used in the design of pipe rupture restraints. Include a discussion of the de_ sign methods applicable to the auxiliary steel used to support the pipe rupture restraint.Provide assurance that the pipe rupture restraint and supporting structure cannot fail during a seismic event.

3. 6.o? . /, /,/a.

Refer to the=4est paragraph3f ef;.:.. ...you.. 6u Qu.. Lie. 21^.2^". RCL pipe breaks have been eliminated thereby eliminating the need for RCS loop restr-aints.

 ^- . Pipe whip restraints ': r^-- ;!.. ICL are designed as a combination of an energy-absorbing element (EAE) and a supporting (auxiliary) structure capable of transmitting the resistance load from the EAE to the main building structures (concrete walls, slabs, and steel structures). The EAE usually is either thin gauge cellular crushable material (energy-absorbing material, (EAM)) or stainless steel U-bars. The design limits for EAEs are specified in Section 3.6.2.3.4.1.2.

The supporting structures typically are structural steel frames designed to the loads, Icad combinations, and stress limits as specified in Section k 3.8.3.3 and Tables 3.8.3-2 and 3.8.4-2. Except for the main steam restraints inside the containment, the elastic working stress design method of Part I of the AISC specification 1969 (including supplements 1, 2 and 3) is used. The main steam line restraints inside the containment are designed using a non-linear method, with allowable ductilities per Section 3.5.3 and Table 3.5-13, where the ultimate strain is taken as 50 percent of ASTM specified minimum.Both the Operating Basis Earthquake (OBE) and the Safe Shutdown Earthquake (SSE) seismic events are specifically included in the loading combinations prescribed for the structural integrity of the pipe whip restraints. The restraints and their structures are treated as structural subsystems whose seismic response is determined from their frequency characteristics and the appropriate floor response spectra.l Vol. 2 Q&R 3.6-17N Amendment 50

F ATTACHMENT STP FSAR ST HL AE /7#l PAGE /6 OF lM 3.6 PROTECTION AGAINST THE DYNAMIC EFFECTS ASSOCIATED WITH THE POSTUIATED RUPTURE OF PIPING Pipe failure protection is provided in accordance with the requirements of 10CFR50, Appendix A, General Design Criterion (GDC) 4.In the event of a high- or moderate-energy pipe failure within the plant, ade-< quate protection is provided to ensure that the essential structures, systems, or components are not adversely impacted by the effects of postulated piping failure. Essential systems and components are those required to shut down the reactor and mitigate the consequences of the postulated piping failure.Appendix 3.6.B provides several examples of the evaluations made of the ef- 53 facts of postulated high energy pipe failures within the plant. The following sections provide the basis for selection of the pipe failures, the determination of the resultant effects, and details of the protection requirements.3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside Containment Table 3.6.1-1 provides a matrix of plant systems that indicates their classi-ficationi high energy, moderate-energy, essential, or nonessential. Selec-tion of pipe failure locations and evaluation of the consequences on nearby essential systems, components, and structures are presented in Section 3.6.2 40 and are in accordance with the requirements of 10CFR50, Appendix A, CDC 4 "r---! f.. A;;_;;.;;;;12 t 1::p '"),4 elections and evaluations are in accordance with the guidance of Nuclear Regulatory Commission (NRC) Branch('f Technical Positions (BTP) ASB3-1 and MEB 3-1,(12:1 ; ^ ^ 1. ' O . ... t'-

 ":fr- -- 1; n .ifr ^ - i__i L A ..le;;i r _.d .._1__ ".. ;f pip-breehe i TNSER7" 3.6.1.1 Design Bases. The following design bases relate to the eval-untion of the effects of the pipe failures determined in Section 3.6.2.

1. The selection of the failure type is based on whether the system is high-or moderate-energy during normal operating conditions of the system.

High-energy piping includes those systems or portions of systems in which the maximum normal operating temperature exceeds 200*F or the maximum normal operating pressure exceeds 275 psig.Piping systems or portions of systems pressurized above atmospheric pres-sure during normal plant conditions and not identified as high energy are considered moderate-energy.Piping systems that exceed 200*F or 275 psig for about 2 percent or less of the time the system is in operation or that experience high-energy pressures or temperatures for less than 1 percent of the plant operation time are considered moderate-energy.3.6-1 Amendment 53

ATTACHMENT

 /7W ST PAGEHL 17 AEOF /2 7 INSERT Page 3.6-1 .

The original design basis postulated pipe break locations'in the reactor coolant loop (RCL) are described in Reference 3.6-1. A detailed fracture mechanics evaluation, as described in Reference 3.6-14, demonstrates that the probability of rupturing the RCL piping is extremely low under design basis conditions. Therefore, postulated RCL ruptures and the following associated dynamic effects are not included in the design basis: missile generation, pipe whip, break reaction forces, jet impingement forces, decompression waves within the ruptured pipe, and pressurization in cavities, subcompartments and compartments. To retain high safety margins, the design bases for emergency core cooling systems, containment pressure boundary and equipment qualification are based on a non-mechanistic rupture of the RCL piping.l .l ll Ii 17603N:0288N/6 1

STP FSMR ATTACHMENT ST-HL-AE- / 79/PAGE /8 OF /49

2. The following ccsumpticna era unsd to datsrmina th2 thsrmodyntmic ctita in the piping system for the calculation of fluid reaction forces-{- g f:n ' "::::. C__1_..; ^,_::: 'EC:;;: )

a. For those portions of piping systems normally pressurized during operation at power, the thermodynamic state in the pipe and associ-ated reservoirs are those of full-(100-percent) power operation.

b. For those portions of piping systems only pressurized during other normal plant conditions (e.g., startup, hot standby, reactor cooldown), the thermodynamic state and associated operating condi-tion is determined as the mode giving the highest enthalpy.

r: ti: ".05, in:1_li... 11 1... LinumL ylpir.;, .L. om . - - ';-"- -_,_.. ie.. ..c.liti::: ..; __ 2 .. ie:ed i.. "cf:r: ;_ 1,

3. Moderate-energy pipe cracks are evaluated for spray wetting, flooding, and other environmental effects.

4. Where postulated, each longitudinal or circumferential break in high-energy fluid system piping or leakage crack in moderate-energy fluid system piping is considered separately as a single initiating event oc-curring during normal plant conditions.

5. Offsite power is assumed to be unavailable if a trip of the 40 turbine-generator system or trip of the reactor is a direct consequence of the postulated piping failure.

i

6. A single active component failure is assumed in systems used to mitigate the consequences of the postulated piping failure or to safely shut down the reactor, except as noted in paragraph 7 below. The single active component failure is assumed to occur in addition to the postulated piping failure and any direct consequences of the piping failure, such as unit trip and loss of offsite power (LOOP).

7. When the postulated piping failure occurs in one of two or more redun-dant trains of a dual-purpose, moderate-energy essential system, single failures of components in other trains are not assumed, because the sys-tem is designed to seismic Category I standards; powered from both offsite and onsite sources; and constructed, operated, and inspected to quality assurance, testing, and inservice inspection standards appropri-ate for nuclear safety systems.

Failures are not assumed in a system or component which is normally oper-ating at the time of break initiation and which also functions (without change in state) to mitigate the break event, provided the system is designed to seismic Category I requirements and is qualified for the en-vironment associated with the break event.1 3.6-2 Amendment 40'

5. .

ATTACHMENT

 -STP FSAR ST-HL AE- /7t/M PAGE /9 0Fl19

5. -Tec any postulated loss-of-coolant accident (LOCA), the structural ano leaktight integrity of the Containment is maintained.

6. The escape of steam, water, combustible or corrosive fluids, gases, and heat in the event of a pipe rupture will not preclude:

a. Subsequent access to any areas, as required, to cope with the postulated pipe rupture.

b. Habitability of the control room.

c. The ability of essential instrumentation, electric power sup-plies, components and controls to perform their safety func-tions to the' extent necessary to meet the criteria outlined in Section 3.6.1.1.

3.6.2 Determination Of Break Locations and Dynamic Effects Associated With The Postulated Rupture Of Piping This section describes the design bases for locating postulated breaks and 1 cracks in high- and moderate-energy piping systems inside and outside of the Containment; the methodology used to define the jet thrust reaction at the break location; the methodology used to define the jet impingement loading on adjacent essential structures, systems or components; pipe whip restraint:design; and the protective assembly design.40 3.6.2.1 Criteria Used To Define High/ Moderate-Energy Break / Crack(' ,i Locations and Configurations. :___,. .^.. .. ^^'. Nuclear Regulatory Commission (NRC) MEB 3-1 Ref. 3.6-3 is used as the basis of the' criteria for the postulation of high-energy pipe breaks. Specific moderate-energy pipe crack locations are not ascertained; and, therefore, they are assumed to occur as described in Section 3.6.2.1.2.A postulated high-energy pipe break is defined as a sudden, gross failure of the pressure boundary of a pipe either in the form of a complete circumferen-l tial severance (i.e., a guillotine break) or as a sudden longitudinal, uncon-trolled crack. For moderate-energy fluid systems, pipe failures are confined to postulation of controlled cracks in piping. The effects of these cracks in moderate energy fluid systems on the safety-related equipment are analyzed for flooding and wetting only. These cracks do not result in jet impingement or whipping of the cracked piping.Breaks as stated above are postulated in each pipe and branch run adjacent to a protective structure or compartment containing essential systems and components.Piping is considered adjacent to a protective structure or compartment con-taining essential systems and components required for safe shutdown if the distance between the piping and structure is insufficient to preclude impair-ment of the structure's integrity from the effects of a postulated piping failure, assuming that the piping is unrestrained.I tl 3.6-7 Amendment 40 I . -q._ _ _ _-- _

ATTACHMENT ST HL AE- 17#/PAGE 420 OF /d ')STP FSAR 3.6.2.1.1 High-Energy Break locations: With the exception of those por-tions of the piping identified in Section 3.6.2.1.1.5, breaks are postulated

• in high-energy piping at the following locations: 2

1. American Society of Mechanical Engineers (ASME) Boiler and Pressure '

Vessel; (B&PV) Code, Section III, Division 1 - Class 1 Piping.

 /M6ERT' a.[Pr A-di:: = h:::h 1:::ti::: --d -r i-* -

• i in th: RCL::: d:ri:"

the basis of stress and fatigue analysis. These postulated eak 40 loca s and the methods used to determine them are descri in Ref. 3.6- An analysis of each individual RCL confirm e break locations de d in Ref. 3.6-1. The stresses and c ulative usage factors resulting om seismic events are includ in the stresses and cumulative usage ctors which are discu d in Section 3.6.2.5 to verify the design bas break location n the RCL noted therein.At postulated circumferential br ocations, the piping is assumed to separate to allow double-en f unless structural restraints exist which physically limi he break ning area. As an example, j for the reactor coolant pe break at the ctor vessel nozzle, the pipe will be restra d, preventing the develo ent of a full 45 double-ended br . At other locations where a r ced break area is used pr ily due to structural steel or concrete straints, justif ion is provided in Section 3.6.2.5. Longitudi breaks ar saused to have an opening area equal to the flow area o he r i r--I b. Pipe breaks are postulated to occur at the following locations in ASME Code Section III Class 1 piping runs or branch runs outside the

RCL as follows:

1) At terminal ends of the piping, including:

a) Piping connected to structures, components, or anchors that act as essentially rigid restraints to piping trans- 40, lation and rotational motion due to static or dynamic l loading.i b) High/ moderate-energy boundary.such as piping runs which are maintained pressurized during normal plant conditions for only a portion of the run, i.e., up to the first nor-mally_crosed valve. The terminal end of such piping is the piping connection to the closed valve.c) Branch intersection points are considered a terminal end 40 for the branch line except where the branch and the main Q110.6 piping systems are modeled in the same piping stress analysis and the branch line is shown to have a significant effect on the main run behavior (i.e., the 50 nominal size of the branch line is at least one-half of that of the main or the ratio of the moment of inertia of Q110.6 main run pipe to the branch line is less than 10).

• 3.6 8 Amendment 50

ATTACHMENT ST-HL-AE iM4 PAG _E_fh OFI M INSERT Page 3.6-8 The original design basis postulated pipe break locations in the RCL are described in Reference 3.6-1. A detailed fracture mechanics evaluation, as described in Reference 3.6-14, demonstrates that the' probability of rupturing the RCL. piping is extremely low under design basis conditions. Therefore, postulated RCL ruptures and the following associated dynamic effects are not.included in the design basis: missile generation, pipe whip,. break reaction forces, jet impingement forces, decompression waves within the ruptured pipe and pressurization in cavities, subcompartments and compartments. _To retain high safety margins, the design bases for emergency core cooling systems, containment pressure boundary, and equipment qualification are based on a non-mechanistic rupture of the RCL piping.7603N:0288N/7

STP FSAR ATTACHMENT ST-HL AE igr/t PPGEJSOF Ig/;d b. Other Containment Fenetration Fining Containment penetration piping between the penetration flued head and containment isolation valves, up to and including the restraints that define the terminal ends for the run as stated in 6) below, may

 . be excluded from postulated breaks (i.e., may be treated as a break exclusion zone) when all of the following design requirements are met:

l

 . 1) AS;iE Ccde Section III Class 2 Fiping: if the following condi- - tions are not met, then requirements listed in Section 3.6.2.1.1.2 above apply, a) The maximum stress ranges as calculated by the sum of Equations (9) and (10) in ASME Section III, subarticle NC-3652, considering operational plant conditions (i.e.,

l f~ sustained loads, occasional loads, and thermal expansion i and an OBE event) do not exceed 0.8 (1.2 Sh

• IA)*

b) The maximum stress, as calculated by Equation (9) in sub* 40 article NC-3652 under the loadings resulting from a pos-

! tulated piping failure of fluid system piping beyond these portions of piping, does not exceed 1.85, except that, following a piping failure outside ContaTnsent, the pipe between the isolation valves and the first restraint is permitted higher stresses provided that a plastic hinge is 53 not formed and c;crability of the valves with such stress

( is assured in accordance with the requirements of Section 3.9.3.

2) Welded attachments, for pipe supports or other purposes, to these portions of piping are avoided except where detailed l

stress analyses or tests are performed to demonstrate that the

maximum stresses do not exceed the limits defined in 1) above.

i

3) The member of circumferential and longitudinal piping welds and

branch connections are minimized.

4) The length of these portions of piping is reduced to the mini-

' mum length practical.i

5) Fipe anchors or restraints (e.g., connections to containment j penetrations'and pipe whip restraints) are not welded directly to the outer surface of the piping (e.g., flued integrally

' forged pipe fittings may be used) except where all such welds are 100 percent volumetrically examinable as part of the l

Inservice Inspection Program (Section 6.6) and a detailed stress analysis is performed to demonstrate that the maximum

stresses do not exceed the limits defined in 1) above.

Exceptions to the 100 percent volumetric weld examinations 53 (e.g., due to access limitations) are documented in the ISI program.l Ib b3.6-11 Amendment 53 ii

 ---w-w,e,w-. - - - . - - --,w----.----.----.- _v=,--, *w .w~,m,w-

J ATTACHMENT ST HL-AE- /7 '

 /

STP FSAR PAGE AJOF r(' meet the requirements of ASME Code, Section III, Sub-article NE-1120, and are designed so that the maximum stress range does not exceed 0.4 (1.2 Sh

• 3A)*

2. - Through-wall leakage cracks are not required to be postulated in moderate-energy fluid system piping located in an area where a break in the high-energy fluid system is postulated, provided that such cracks do not result in environmental conditions more limiting than the high-energy pipe break.

3.' Subject to paragraph 4 below, through-wall leakage cracks are required to be postulated in ASME, B&PV Code, Section III, Division 1 - Class 2 or 3 piping at locations where the maximum stress range in the piping is greater than 0.4 (1.2 Sh + 8A )- 40 Q110.

4. Individual cracks are not required to be postulated at specific locations 09 determined by stress analyses when a review of the piping layout and plant arrangement drawings shows that the effects of through-wall leakage cracks are isolated or physically remote from structures, systems, and components required for safe shutdown.

5. Through wall leakage cracks are postuisted in non-seismic Category I piping at welded points where the effects might compromise essential equipment or structures.

To simplify analysis, cracks may be postulated to occur everywhere in moderate energy piping, regardless of the stress analysis results to determine the maximum damage from fluid spraying and flooding, with the consequent haz-

 '." ards or environmental conditions. Flooding effects are determined on the be-sis of 30 min operator time required to effect corrective actions. Further discussion of internal flooding effects is provided in Section 3.4.3 and $3 3.4.4.

50 Cracks in moderate energy ASME Code Class 1 piping are not postulated since Q210.36N there are no A!ME Class 1 moderate energy systems. All the ASME Class 1 Q110.10 piping systems are inside the containment Building and are high energy.3.6.2.1.3 Types of Breaks / Cracks Postulated:3.6.2.1.3.1 ASMESectionIII. Class 1RCLPiping-HighEnergy-4ho.No

m ._

 ;fbreaks/

discussed in R f. postulated [intheASMESectionIII, Class 1primaryRC *

 .6 1 C nuo 7444444)o# .S. d. 4. 4 4 la I 3.6.2.1.3.2 Piping Other than RCL Piping - High Energy - The following l types of breaks are postulated to occur at the locations determined in accor-l dance with Section 3.6.2.1.1.

40', 1. In piping whose nominal diameter is greater than or equal to 4 in., both lcircumferential and longitudinal breaks are postulated at each selected ibreak location unless eliminated by comparison of longitudinal and axial! stresses with the maximum stress as follows:

a. If the maximum stress range exceeds the limits specified in Sections f 3.6.2.1.1.1.b.2 and 3.'6.2.1.1.2.b but the circumferential stress j

(p

• range is at least 1.5' times the axial stress range, only a 49 longitudinal break is postulated.

3.6 13 ., Amendment 53,n.. . . . . . . . . ..

 ' s ATTACHMENT STP ESAR 7V4 ., ST HL PAGEdu/ AEdF IJU7 the forcing function. It should be noted that the rise time for ,the jet 50 .

thrust is no greater than one millisecond. For most applications, one of the Q21C j following situations exists:25N e Superheated or saturated steam.e Saturated or subcooled water.e Cold water (nonflashing). 40 Analytical methods for calculation of jet thrust for the above-described sit-untions are discussed in Refs. 3.6-5 and 3.6-6. 1. l i _ _ _ _ . . . . . .... j n

 " ' r: E; f_.._ 1.... f. _ .7CL L.._L i p il J 1.. "...i r 2.i.~.I.l.L. .T e.T" A ND For main feedwater, main steam, and reactor coolant surge line RELAP 4/5 is used to get the forcing function for the nonlinear time-history pipe whip load analysis. For other lines, Moody's thrust coefficient is used, as specified in Reference 3.6-6.

Nonlinear time-history pipe whip load analysis is a step-by-step det ination of piping / whip restraint transient response through time, explici y including both matarial (inelastic) and geometric (gap) nonlinear effects.The mathematical models are three-dimensional, lumped-mass models constructed from pipe elements, inelastic energy-absorbing elements, and energy-absorbing 49 device support structure mass and stiffness characteristics. This analysis is performed using Reference 3.6-11, which is based on direct integration of the 'lumped. mass model's equation of motion.Dynamic impact and potential rebound effects of the pipe whip problem are explicitly considered in the REIAP 4/5 computer code. Therefore, no additional dynamic amplification factor or rebound effect factor is applied to the nonlinear time-history results.The energy balance dynamic analysis method is limited to intermediate-size high-energy lines under 14 in. in diameter. Jet thrust load is taken as the maximum thrust load (with an amplification factor of 1.1) and applied throughout the pipe break event. Maxixmum restraint device deformation is computed for the energy principle. An appropriate dynamic load factor is then applied to the calculated restraint load for restraint devfce dqsign. :KO.L b ra nc.h ff E o (c.4 K -3.6.2.2.1.1 Time Functi Jet Thrust Force on Su- un d M E u RCL Piping - To determine thefthrust and reactive force loads to be applied to the RCL during the postulatedkW94, it is necessary to have a detailed description of the hydraulic transient. Hydraulic forcing functions are calculated for the ,_ _

 - RCLs as a result of a postulated b96A. g These forces result from the transient flow and pressure histories in the RCS. 40 The calculation is performed in two steps. The first step is to calculate the transient pressure, mass flowrates, and thermodynamic properties as a function ,

of time. The second step uses the results obtained from the hydraulic analysis, along with input of areas and direction coordinates, and calculates the time-history of forces at appropriate locations (e.g., elbows) in the RCLs. a The hydraulic model represents the behavior of the coolant fluid within the RCS. Key parameters calculated by the hydraulic model are pressure, mass 45 3.6 16 Amendment 50

ATTACHMEN ST-HL AE 17 STP FSAR PAGEc250F l 9

ensi ty. These are supplied to the thrust calculation, to-('

flowrate, a-gether with mt layout information, to determine the time-dependent loads exerted by ti.e fluid on the loops. In evaluating the hydraulic forcing func-tions during a postulated LOCA, the pressure and momentum flux terms are dom-inant. The inertia and gravitational terms are taken into account in the evaluation of the local fluid conditions in the hydraulic model.The blowdown hydraulic analysis is required to provide the basic information concerning the dynamic behavior of the reactor core environment for the loop forces reactor kinetics, and core cooling analysis. This requires the ability f5

 .to predict the flow, quality, and pressure of the fluid throughout the reactor system. The MULTIFLEX code (Ref. 3.6-7) was developed with a capability to provide this information.

The MULTIFLEX computer code calculates the' hydraulic transients within the entire primary coolant system. This hydraulic program censiders a coupled, fluid structure interaction by accounting for the deflection of the core sup-port barrel. The depressurization of the system is calculated using the method of characteristics applicable to transient flow of a hom*ogeneous fluid 40 in thermal equilibrium.The ability to treat multiple flow branches and a lar6e number of mesh points gives the MULTIFLEX code the flexibility required to represent the various flow passages within the primary RCS. The system geometry is represented by a network of one-dimensional flow passages. ,

 , The THRUST computer program (Ref. 3.6 8) was develeped to compute the trans-isnt (blowdown) hydraulic 1 cads resultiny, from a LOCA.

i De

 /

e :3.6 16a Amendment 49 r.. o . . . . .Y'

 ... AT TACPMENT 2

ST HL4E-IM/PAGE 46,0F !M The blowdown hydraulic loads on primsry i. cop compenents are r.<mputed from the

; - equations s a ,

(P - 14.7) +F = 144A .

 - g e A* 144'.

The symbols and ynits are as follows: ,F = Force (ibg ). ,0 4A = Aperture area (ft2 ). .P = System pressure (paia),s = Mass flowrate (1bs/a). ,P = Density (1ba/ft 8) i g = Cravitational constant 32.174 feriha/lb s* l 8 45 A, = Maes how area (ft')I In the model to compute forcing functions, the RCL system is represented by a model similar to that employed in the blowdown analysis. The entire loop layout is represented in a global coordinate system. Each node is fully described by: ,

1. Blowdown hydraulic information.

60 -

2. The orientation of the streamlines of the force nodes in the system, which includes flow arass, and projection coefficients along the three e axes of the global coordinate system.

Each node is modeled as a separate control volume with one or tro flow, apertures associated with it. Two apertures are used to simulate a chenge in! flow direction and area. Each force is divided into its x, y, and r compon- .ents using the projection coefficients. The force compenants are then summed

over the total number of apertures $n any one node te give a total a force, k .

total y force, and a total a force. These thrust forces serve 49 input to the pipiag/ restraint dynamic analysis.I The THRUST code (which uses MULTIFLEX resulta as input) calculates forces lexactly the same way as the (Raf. 3.6-5) STHRUST code (which woes SATAJi (Ref. 45 3.6-10] results as input). .3.6.2.2.1.2 , Dynamic Analysis _of the Reacter Coolant Loop Yipinst AIND_, _Equipment Supports #ftys-th!F4detretees - The dynamic analysis of the RCL for loadinas is described in Section 3.9.RCC lorwsk Pipe bra ^k 40 3.6.2.3 Dynamie Analysis Methods To verify Integrity and Operabilit_y. 'f 3.6.2.3.1 Dynamic Analysis Methods To Verify Integrity and Operability for Other then RCL The analytical methods of Refs. 3 6-5, 3.6-6, and 3.6-9 lare used to determine the jet impingement effects and loading effects ,i spplicable to components and systems resulting from postulated pipe breats and II e

n. . . . . . . . - . , . . . . - .,

 . . . ~ _ _ . . . _ - _ _ _ . _ _ _ _ _ . _ , _ _ . _ _ _ _ _ -

- = _ _ _ - -ATTA045 tan -3T HLAC. Igy STP FSAK d 0<cracks. Note that for khort periods of tiue, the pressure and enth:1p,v in certain systems will be higher than fu21 or ne real power operation i.e. ,102 parcant poseer. liceever, the full pcaer n' ode establiches the maxieue demands I of safsty systems in the event of a pestiilated pipe 'rup;;ure. Otter s

• s of normal operation have reduc 6d ne,eds for safety bystems to bring t.1e plant to a safe shutdown. Therefore, the ful'1 power operation mode is used, to sieterninc h the therinodynamics state in the piping system for the calculation of finid )

reaction forces, 40 3.6.2.3.2 Dynamic _. Analysis Methods To Verify Integrity and OperabCity for the RCLt 3.6.2.3.2.1 General - A LOCA is assuned to occur fc? a branch line Ercak dovn to the escond norusily open automatic isolation valve (Case II,11gtti 3 45 3.6.2-1) on outgoing lines ar}d down to and incigding the second ch6ck valvs ( '(Case III, Figure 3.t,.1-J) on incoming linea normally With flos. A pipe break l beypad the rentraint or secon.1 check yalva does r.ot result in an uncontro11cd 7 Joss of reactor coolant if either of the twa valves in th.e line closes.Accordiagly, both of the aut;matic isolation vsiver aru suitaby protected and k teatrais.ed as close to the valves as possible to that a, pipe break beyond the ra:itraint does not jeopardi:e the integrity and o'perahi'lity of the va!ves.40 i h rther, periodic testing of t'ie valves capabilf.ty to perform thair int 6nded 5 function te esaantiel. This etiterion takes credit fcr only one of the two e valves perfctming its intended firaction. For normally closed isolation or in-coming chack valvas (Cases ]t and IV, Figare 3.b.2-1), a LOCA is assemed to occur for pipe tresku on the reactor side of the valve.Branch lines connected to the. RCL are defined ab large strictly for t'ne pur- l pose of pire break criteria when they have an itisice diameter greattr chari +. l 45 ,in up to the largest connecting line. Rupture of t%st lines results in a rapid blowdown from the R6L, ec3 pro:ection is basically preytded by the accumulators ahd the . low-head safety injection (LHSI) pumpu.L0 Branch lines connected to the RCL are defit.ed as small ,for the purpose of pipe <

 'oreak analysic if they have an inside diameter equal to or less thar. 4 in.

This size is such that Durgency Core Cooling Systsm (ECOS) analyses, using realtrtic assumptions, show thot no fuel claddin6 damage is expected ,for 4 j 45 break area of up to 12.5 in.' corresponding so 4 in. Incide distwter y;ipin.n Pngineered safety featureo (ESFs) are provided for core cooling and boration, pressure reduction, and activity confinetient ir che event of a LOCA or steam u 40 or feedvater line break accident to ensure that the putilic is prcttered in acccrdance with 10CFRit>0 Guidelines.V These safety systems ake designed to ,provide protection for en RCS pipe p2ptt;re of a size up to and including a 45 double-ended severence of f.he RCS mtin loep.

 /A6SAf L

3.6-1p Amendment 45

(+ . _ . - - _ATTACHMENT

 .- ST.HL AE. I1% - - Ac2E,14 OF /Rg ,

INSERT Page 3.6-18 The original design basis postulated pipe break locations in the RCL are described in Reference 3.5-1. A detailed fracture mechanics evaluation, as described in Reference 3.6-14, dennnstrates that the probability of rupturing the RCL piping is extremely low under design basis conditions. Therefore, postulated RCL ruptures and tne associated dynamic effects are not included in the design basis. However, to retain high safety margins, 1L ki 7E03N:028SN/14 L

i ATTOCUMENT ST'P FSM ST-HL AE /7(/t/lPQGE M OFI Q _To assure the continued integrity of the basantial ecepcnents and the engi- I neered safety systems, consideration is given te the consequential effects of the pipe break itself to the extent that:

1. The minimum performance capabilities of the engineered safety systems are not reduced below that required to protect against the postulated break.

l 2.- The containment leaktightness is n:t det:reased below the design value if the break leads to a LOCA (1).

3. Propagation of damage is linited in type and/or degree to the extent that:

a. A pipe break which is not a LOCA or secae/feedwater lire break wil; j cot cause a LOCA or steam /feedwater line break, i-

b. An RC)(6ntgh pipe break will not caus6 a steam or feedwater system pipe break, and vice versa, in excess of small instrument or sample lines which are not required to function fellowing accidents.

L $2McH 3.6.2.3.2.2 Large RChiping - Prongetden of-damage-rettritingufra'n:the rupt-ure-efeRCL-is-perwittetl-te - ecw bee-emet-cet-exceed-ttre 1!ertyn-bac4-ftrr caleel4 ting-containean:::and-sebeempertner.0--presaure2 1^ p -hydraulic foceesc se seter--4nternale -eeset r f en-4 cads ,-pr4aary.cquipmt.ht7upport._,lo ads,--or--40 emergeney-core--cooling% eta priormancee tLarge branch line piping, es defined in Geetien 3.6.2.3.2.1, is restrained to meet the following criteria in addition to itena 1 thru 3 of Section I 3.6.2.3.2.1 for a pipe break resulting in a LOCA: '

l. Propagation of the break to the unaffected loops is prevented to ensure the delivery capacity of the accumulators and low head pumps.

2. Prcpagation of the break in the affected loop is permitted to occur but does not exceed 20 percent of the flow area of the line which initially ruptured. 'Ihe criterion is voluntarily applied oc as not to substantial- >

ly increase the severity of the LOCA.3.6.2.3.2.3 Small Branch Lines'- Should one of the stas 11 pressurized lines, as defined in Section 3.6.2.3.2.1, fail and result in a LOCA, the lpiping is restrained or arranged to meet the following criteria in additien to items 1 through 3 of Sectica 3.6.2.3.2.1:

1. Break propagation is limited to the affected icF; i.e., propagation to the other leg of the affected loop and to the other loeps is prevente.d.

Damage to the high-head safety injection (EHSI) lines connected to the other leg of the affected loop or to the other loops is prevented.(1) The containment is here defined as the containment structure liner and penetrations and the steam generator shell, the steam generator steam side instrumentation connections, the stean, f eedwater, blowdown, and steam generator drain pipes within the containment structure.3.6-19 Amndment 44 1

ATTACH.yENT l' 5 T-HL AQ. If r/c{f.d.CE,30OF /&Q STP FSAR .

 .2 . hopaFation of the break in the affected leg in permitted but inust be if25ted to a total. break area of 12.5 in.2 The exception ta fhis cese )

is when the initiating small break is a epid leg UH51 line. Furth6r propagatiot is r.ot pern1*:ted for this caso.

3. Pr.opagation of the break to a HMSI line connected to the affected Igg is prevented .1f the line breal results in a loss of core coolirg capability '

dom to a spilling injection line.40

 - /tbW .

3.6.,2}3.2.4 Design and Verification of Adequacy of RC1. Componetta and Sycrts -jThe esthnda_.dansrif>ed-belme-ace 22 d in-the-Weseittpc*.... -m,. __wer4Lication-ef-the-adequecy--e5 -prisery-AC4,-eeepenente--and-auppeseter--Thece -- .dethods af e used only to datermine jet impinge:st;nt loads on SCL .:omponents,,and av)Rortsar.:3arenotusedfordesignar.dcheckingofwallk,RCSbarrierpf j cab 1) strays, etc. O N *f' e T'.e desik basis postulated pipe rupture locations :for the RCL pp, ig are Jeterinined gsing the criteria given in Se:tien 3.6.2.1. Thenej design basis rupturss are ysed es the rupture la.ications for considerattenj 6f Set impinge-ment ef fects og prjaary equipment and supporta. / ,

 / I .

lA 'vnamic analysit sie veed to determine maxirum piping d'ikplactinents at spch i deaipt basis rupturet location. The eaxittua pipind d lacements a:e used to

 ! cmrpute the effective *upture f16w area at eaCh location. The flow aret und 40 rupture orientatien is bsn usnd to detersinc .th!/ jet flow pattern and to .

identify Any primary co*knents which are pote71 a1 targets for *Jtt irapinge'trant . N f i

 \ / i "the jet thrust at the peint of based on the fluid presaute and i temperatureconditionaoccurring%ptureirf uring full (100 percent) power operating t conditio:;t of the plant. At the poing!of rupture, the jet . force is equal and

oppc it2 to the j e'c thrurit. The f or g of the jet is, conservatively assumed t a l be ccnstant throughout. ttie jet fic dibqance. The sub-cooled jet is assumed to expand uniformly tt a half-angIe of 10 degrees, $ rom wh.ich chn area of the i jet on thu target anc tue fra fon of the h,,et interceptet, by the. t arget structuru (an be readily dete, Ined.

l The shape cf the target a ets the amount of agntum change in the jet and

 .thus affects the impingement force on the target. qThe targsit shape factor is degrees tlused awayr.o eteunt forax$d.gnt tar t the flow ',0

[

 .from the jet shaper which de not def1 l ) / A5 l- It he method med tn' compute the jet impingnment load on a arget is one of the l- lfo.11ovitg i \ t

1. is The evaludy:(tatic ed bry a; eff3ct plying ofajat impingem:nt etep load whoseen the targetisberuct magnitudo gi ren s:

 's F) = KjP,A,g M i

j/ ,ers:f F y- Jet 3 ---.impingement load, on carget.__4 3.6.20 Amendtent 45 L _ . - . _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - - _

TTACHMENT ST HL AE- sW PAGE st OF i ')Ih3ERT Page 3.6-20 eThe original desigf. basis postulated pipe break locations in the RCL are oescribed in Reference 3.6-1. The primar,y RCL components and supports design were based on the;e postulated break locations. A cletailed fracture raecnanics ;cvaluatioa, as described in Reference 3.6-14, demonstrates that the probabi'lity of ruptering the RCL piping is extremely low under design basis conditions. Therefore, postulated RCL ruptures and the following associated, dynamic effec'es are '.10t it.cludud in the design bas,is: missile generation, pipe whip, break ret:tior. forces, jet impingenent forces, decompression waves ,r within the ruptured pipe, and fressurizatipn in cavities, subcompartments and

corapartments. The dynwie effects fro

n ruptures in Class I branch lines and .!

J othat high energy piping are reviewed to verify that the effects are bounded ;by the curien; analyses.h fs Yi 1e fe

 /603N'028811/10 [ - ++ e---- .- <e..*---m . --. e-.,..-.-4.,,.<_--- _ -,,.-i,s. - - - -u..- -tme- e--'e- * - = -e* < - - - - - -- * ~ . - - *

• m

ETP FSAR ATTACHMENT! ST HL-AE I PAGE 3;lOF u lK = Dimen31culGs3 jot thrust co2ffici;nt bassd e,n inmeA

• fluid conditions in broken loop P,= Initial system pressure A

d

 = Calculated maximum break flow area = Fraction of jet interceptai by target '45 5= rget shape factor 4 - Discharge il , areas for limited flow ares circumferential breaks are obtainedfromrhetorcoolantanalysesperformedtodeIerminetheaxial y_

and lateral displacements of the broken ends as a fjdction of time; AaB is the maximum breag flow area occurris;g during 1;he transient, and is calculated as the total surface area through whihh the fluid must pass to emerge from the brokeit g pipe. Using geometricaI formulations, this surface area is determined to be a function /of the pipe separation (axial and transverse) and the d sensions of the 41pe (inside and outside )dianter). / I If a simplified static analysia,is rformed instead of a dynamic analysis, the above jet load (Fi), a multiplied by a dynamic load factor. l For an equivalent static analysis yf the target structure, the jet impingement force is multiplied by dynamic load factor of 1.2 to 2.0, 40 depending upon the time var 14nce of e jet load. This factor assumes 4 that the target can be reyr'esented as sentially a one-degree-of-freedoa

 ~/, system, and the impingement force is con rvatively applied as a step load.

The calculatoin of the dimensionless jet thrus coefficient and break flow area is di unsed in Section 3.6.2.5.

2. The dynamic ffect of jet impingement is evaluated a followin dime-dependent load to the target structure) pplying the 45

 = KPA RS 3 B w re the system pressure, P, is a function of time; the jet rust oefficient, K, is evaluated as a function of system prassure enthalpy; and the break flow area, A,, is a function of time.

3.6.2.3.3 Types of Pipe Wip Restraints:3.6.2.3.3.1 Pipe W ip Restraints ;J.... L Z i C r. n - To satisfy 40 varying requirements of available space, permissible pipe deflection, and equipment operability, the restraints are designed as a combination of an energy-absorbing element and a restraint structure suitable for the geometry required to pass the restraint load from the whipping pipe to the main building structure.3.6-20a Amendment 4

n. . . . . . . . . . .

ATTACHMENT ST-HL AE 17tM PAGE 3.3 OF laq STP FSAR The restraint structure is typically a structural steel frame.or truss and the energy-absorbing element is usually either stainless steel U-bars;ar energy-absorbing material as described-below:1,. Stainless Steel U-Bar

 ~

This type consists of one or more U-shaped, upset-threaded rods of stain-less srael looped s.round the pipe but not in contact with the pipe to al.Iow unimpeded pipe motion during seismic and thermal movement of the i

 . pipe. At rupture, the pipe moves against the U-bars, which absorb the kinetic energy of the pipe motion by yielding plastically. A typical ex-ample of a U-bar restraint is shown in Figure 3.6.2-3.

2. Energy Absorbing Material 40 This type of restraint consists of a crushable, stainless steel, inter-nelly honeycomb-shaped element designed to yield plastically under impact i of the whipping pipe. A design hot position gap is provided between the

pipe and the energy-absorbing material to allow unimpeded pipe motion

during seisuic and thermal pipe movements. A typical example of an

! energy-absorbing material restraint is shown in Figure 3.6.2-4. c l 3. 5-Way Restraint l A five way restraint is utilized to protect the main steam isolation valves (MSIVs) and main feedwater isolation valves in the event of a pos-l tulated pipe rupture outside the Containment. This restraint is designed so that postulated pipe breaks beyond the five-way restraint will not result in stresses greater than 1.8 5 being transmitted to the piping

betweentheisolationvalveandcontaInmentpenetrationorformationofa '

l plastic hinge between the isolation valve and the restraint.i

4. Containment Main Steam Line Restraints 53 The main steam line restraints inside containment are designed using nonlinear, inelastic methods with allowable ductilities given in Table 3.5-13. The anchorages to the internal structure are designed to the restraint backup structure using standard elastic design methods to ensure sufficient anchorage.

 - /IL52RI~

3.6.2.3.3.2 Restraints for RCL gfPir; re;treint;. free-and-locations-are l45 ,dieeuesed-ir.::;; tion-herla --Imeding-combinatie--r eM -screee 14=8 e=_ ara dia-ouseed-in-Section-h971w 3.6.2.3.4 Analytical Methods:3.6.2.3.4.1 Pipe Whip Restraints CO.; C.... ICL:..;.-eint:40 Q11

1. Location of Restraints 11

a. For purposes of determining pipe hinge length and thus locating the pipe whip restraints, the plastic moment of the pipe is determined in the followins; asnner:

M p

 - 1.1 s py S 3.6-21 Amende nt 53

l ATTACHMENT i

 . ST HL AE-]1yll l PAGE 34 0F IW) \ \

I INSERT Page 3,6-21 .'As discussed in Reference 3.6-14 and paragraph 3.6.2.1.1.'la, RCL ruptures and the associated dynamic effects are not included in the design bases. RCL pipe ,restraints are no longer required.

• i t

l il lI ,r ll I(t if ir 7603N:0288N/9 I

r.ATTACHMENT ST HL-AE- s7W/STP FSAR PAGE 35 OFIM

1) Stainless Steel U-Bars c- 0.5.C p where:

c"- ultimate uniform strain of stainless steel (strain at ultimate stress).

 - 40

2) Energy-Absorbing Material Q110.

11 c- 0.8 C p where:c-p maximum strain at uniform crushable strength. l49

e. A dynamic increase factor is used for steel which is designed to remain elastic. g 3.6.2.3.4.2 RCL Restraints of- fere- --imead u ^- 6 q r; of seassec-9 4it ng-eyetees-arew meef4ered in: M ixti = ciih un s.1 vr- ;t'-",

1:_f. -nd earthqua u .e .. ... .... _ _fga mf ..yyurt: 2n':::treia; in r? -- -

 ~ -----;.h. svuiluma ia.g ity of Mt:1;;;,;c.;;t: c.d FL":. 1- 'M;;e b' _;.1.u. end etr .. limit. e s e di. ....J 1.. . si m 3.^.1.

C. ,.3.6.2.4 Protective Assembly Design Criteria.3.6.2.4.1 Jet Impingement Barriers and Shields: Barriers and shields, which may be of either steel or concrete construction, are provided to protect essential equipment, including instrumentation, from the effects of jet impingement resulting from postulated pipe breaks. Barriers differ from shields in that they may also accept the impact of whipping pipes. Barriers and shields include walls, floors, and structures specifically designed to provide protection from postulated pipe breaks. Barrier and shield design is based on the methods of Ref. 3.6 5 Section 3.0, and the elastic-plastic methods for dynamic analysis included in Ref. 3.6 12. Design criteria and l49 loading combinations are in accordance with Sections 3.8.3 and 3.8.4. g 3.6.2.4.2 Auxiliary Guardpipes: The use of guardpipes has been min-inized by plant arrangement and routing of high energy piping. Where they are used, guardpipes are designed to withstand all dynamic and environmental effects of postulated breaks of the enclosed pipe. Auxiliary guardpipes are used only if inservice inspection requirements can be satisfied. Design criteria, loading combinations, and methods of analysis are similat to those for barriers and shields described in Section 3.6.2.4.1.3.6.2.5 Material To Be Submitted for the Operating License Reviev.40 3.6.2.5.1 Piping Systems Other than RCL: Pipe break locations are Q10.04 obtained in accordance with the criteria of Section 3.6.2.1, Figure 3.6.1-1 identifies the break locations in high-energy piping. The preliminary stress results utilized to determine the break types and locations 3.6-23 Amendment 49 m..

ATTACHMENT

 * . ST HL AE- /?(/y .. PAGE .14 0F laq s INSERT Page 3.6-23 As discussed in Reference 3.6-14 and paragraph 3.6.2.1.1.1'a, RCL ruptures and .

the associated dynamic effects are not included in the design bases. RCL. pipe restraints are no longer required. ,7603N:0288N/10

ATTACHMENT ST.HL AE 17</'/STP FSAR PAGE37 OF /M cre given in Table 3.6.2-1. The final summary stress analysis results )(as-built condition) will be provided upon their completion. Associated ctress nodes are shown in Figure 3.6.1-1. High-energy pipe break effects cnalysis for a selected portion of the plant are discussed in Appendix 3.6.B. l53 Appendix 3.6.B also references the appropriate sheet of applicable high-energy lines shown in Figure 3.6.1-1.Moderate-energy piping crack locations are defined in Section 3.6.2.1.2.4. .Evaluation of the effects of moderate-energy cracks is discussed in Section 53 3.4.3 and 3.4.4.The augmented inservice inspection plan is discussed in Section 6.6.Pipe whip restraints are designed in accordance with Section 3.6.2.3. Pipe whip restraint location and orientation for each high-energy break are shown in Figure 3.6.1-1. Barriers and shields are designed in accordance with the criteria of Section 3.6.2.4. Jet thrust and impingement forces were deter-cined in accordance with Ref. 3.6-5. Reaction forces for each pipe whip restraint are presented in Figure 3.6.1-1.3.6.2.5.2 Reactor Coolant Loop:INS EM h 2.5,2 2 :nf Figur: 'K 9M identify-the 1.{3!_ttion and orientations for the RCLs.J. 1 5u basis break loca- ,40 The pri ry and secondary stress intensity ranges and the fat e cumu-Q10*04 lative us e factors at the design break locations specifi in Ref.n in Table: 3.6.2-4 for a reference fatigu nalysis. The a

 -)

3.6-1 are g reference anal is has been prepared to be applicabl or many plants.[t uses seismic rella moments higher than thos sed in Ref. 3.6-1:Ln Ref. 3.6-1, one cation was at the limit, in the Reference anal-rsis the primary stres is equal to the lie of equation (9) in NB-365C lSection III of the ASME &PV Code) at a locations in the system.rherefore, the results of referen analysis may differ slightly from tef. 3.6-1, but the philosoph and nelusions of Ref. 3.6-6 are valid.

onsistent with Ref. 3.6-1, the are no other locations in the model ased in reference fatigue ana sis here the stress intensity ranges and/or usage factors excee he crite a of 2.4 S ,and 0.2, respectively.

Actual plant moments r STP are also give in Table 3.6.2-4 at the design basis break cations so that the refer nce fatigue analysis can be shown to be a icable for this plant. Sinc etual plant moments are shown to be no rester than those used in the refe ce analysis, it fol-lows that t stress intensity ranges and usage facta for STP are less than thos for comparable locations in the reference no . Thus, it is shown at there are no locations other than those identified in Rcf.5.6- where the stress intensity ranges and/or usage factors for STP ex-

ee the criteria of 2.4 S and 0.2, respectively. Thus, the applica-9 ~ af Ref. 3.6-1 to STP is verified -

r NO7 red 01 RED,,

2. gt.Fipe whip restraints ::ir"? riti th: . i.. "'"--are ;d -

6 i... .,.p . irrfin;::--ii.. 1v... .ud . u - - lisi:; - dirrrrrrf ir-

 - u m. . ...... ,

i 3.6-24 Amendment 53 t ,- .. . .. . . .

T - 1 ATTACHMENT ST HL-AE 17pf PAGE 38 OF 189[NSERT Page 3.6-24 The original design basis postulated pipe break locations in,the RCL are described in Reference 3.6-1. A detailed fracture mechanics evaluation, as described in Reference 3.6-14, demonstrates that the probability of rupturing the RCL piping is extremely low under design basis conditions. Therefore, post 0 lated RCL ruptures and the associated dynamic effects are not included in the design basis.i lt 7603N:0288N/ll

AfTACHMENT ST HL-AE /7/N STP FSAR PAGE39 OF /3,'1

 ;---. DE4571D

~

3. ...J 1. a sponse t.o Q21G.2Gii, air 'naz sul, 166.w - .. ..w su sie 3 k}kCfor- ib; ption to Gene Design Criterion M n order t elete 1)ostulation RCL pi reaks based on the " Leak fore reak" analyses.

herefore, jet ement loads associated with the ture of the main I CL piping ar q er considered in the plant de gn. owever* 53 y rimary c nent suppRtc have been designed to thstand A structural l oads ociated with nonhechanistic reactor olant pipe breaRs at the ~ ~

• 1 e... J.scritred in R.I...m.. 2 d 1. 40 Q10.

4 ' Design loading combinations and applicable criteria for ASME Class 1 com- 04 ponents and supports are provided in Section 3.9. Pipe rupture loads in-clude not only the jet thrust forces acting on the piping but also jet impingement loads on the primary equipment supports.C 3.6-25 - Juneadment 53

 , -~ a-- - - - - . . . . , - - - - , - - n--

.n. . . . . . . . . - .

AT STP FSAR ,SiTfruuea T_PAGE 4COF /gg

 '-HL h'U.~)il Section 3.6:

3.6-1 " Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop," WCAP-8082-P-A, (proprietary) and WCAP-8172-A (nonproprietary), Janur.ry 1975.3.6-2 "Subcompartment Pressure Analyses," BN-TOP-4, Rev.'1, Bachtel Power Corporation, October 1977.3.6-3 USNRC BTP MEB 3-1 Postulated Break and Leakage Locations in Fluid System Piping Outside Containment. Branch Technical Position attached to SRP 3.6.2, November 24, 1975.3.6-4 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, 1974 and 1975 Winter Addenda and other Addenda as appropriate.3.6-5 " Design for Pipe Break Effects," Bechtel Power Corporation, BN-TOP-2, Revision 2, May 1974.40 3.6-6

 ~

Moody, F. J. , " Fluid Reaction and Impingement Loads." Paper presented at the ASCE Spacialty Conference, Chicago, December 1973.

 ,. 3.6-7 '"MULTIFLEX, A FORTRAN-IV Computer Program for Analyzing ~ Thermal-Hydraulic Structure System Dynamics," WCAP-8708 (proprietary),. February 1976, and WCAP-8709 (nonpro-prietary), February 1976.

3.6-8 " Documentation of Selected Westinghouse Structural Analysis Computer Codes,"'WCAP-8252,~ Revision 1, May 1977.3.6-9 ANSI /ANS - 58.2, "American National Standard Design Basis for Protection of Nuclear Power Plants Against Effects of Postulated Pipe Rupture," December, 1980' .3.6-10 Bordelon, F.M., "A Comprehensive Space-Time Dependent Analysis of loss of Coolant (SATAN IV Digital-Code)"WCAP-7263, Proprietary (August 1971) and WCAP-7750, 45(Non-Proprietary (August 1971) 3.6-11 "PIPERUP" - Pipe Rupture Analysis Program, ME351, June 24, 1982 3.6-12 Biggs, J.M., Introduction to Structural Dynamics, McGraw-Hill Book Company, New York, 1964 i_. .rNSERI 152 3.6-26 Amen'dment 49 q,_ .. .-. , - . . - _ . _ - . . . - . - , - - . . . 4 - - --

ATTACHMEN ST HL-AE-M 4 INSERT PAGE V/0F JP7 Page 3.6-26 INSERT 1 3.6-13 NUREG/CR 2913 "Two Phase Jet Loads", dated January, 1983 INSERT 2 3.6-14 " Technical ' Bases for Eliminating Large Primary Loop P.ipe Ruptures as the Structural Design Basis for the South Texas Project", WCAP-10559, Proprietary (May 1984) and WCAP-10560, Non-Proprietary (May 1984)I! 7603N:0288N/21 i

TABLE 3.6.2-3

 's.NN POSTULATED BREAK LOCATIONS FOR THE LOCA ANALYSIS OF THE PRIMARY COOLANT LOOP

• i Location of Postulated Rupture g Break Opening Ares **

1. Reactor Vessel Inlet Guillotine Effecti rosn-Sectional Flow l Nozzle N Area of the Loop Pipe

2. Reactor Vessel Outlet Guillotine -'E ective Cross-Sectional Flow Nozzle Area of the Loop Pipe

3. Steam-Generator Inlet Guillotine / Cross-Sectional Flow Area of Nozzle N ,/ the Loop Pipe N/

4. Steam-Generator Outlet Guillotin( 'N Cross-Sectional Flow Area of Nozzle ,

x ~the Loop Pipe $m eh 5. Reactor Coolant Pump Gukilotine Cross-Sectional Flow Area of M Inlet Nozzle ,s the Loop Pipe $

6. Reactor Coolant Pump ,' Guillotine Cross-Sectional Flow Area of Outlet Nozzle / the Loop Pipe

7. 50' Elbow on the Intrados Longitudinal Cross-S'ectional Flow Area of f the Loop Pipe

8. LoopClosur[Widin Guillotine Cross-Sectional Flow Area of the Loop Pipe Crossovet Leg ,

IW

 '/ 'N , @%

g -I. 'N s o-m 8 \ 'muz U

• Refer to Figure 3.6.2-2 for location of postulated breaks in Reactor Coolant Loop. g.k"

[w a

 ** Less break opening area vill be used if justified by analysis, experiments, or considerations of physical restraints such as concrete walls or structural steel.

TABLE 3.6.2-3 (Continued)TULATED BREAK LOCATIONS FOR THE LOCA ANALYSIS OF THE PRIMA?J COOLANT LOOP,*1 I! Location of f Break Op6ning Ares **i Postulated Rupture -s g

9. Residual Heat Removal ( x Guillotine (Viewed Cross-Sectional Flow Area of Line/ Primary Coolant Loop from the RHR Line) the RHR Line 1

l Connection 's \\ h l1 10. Accumulator Line/ Primary Guillot Vleved Cross-Sectional Flow Area of Coolant Loop Connection from th ceumulator the Accumulator Line N '1 i 11. Pressurizer Surge Guillotine (Viewe'd from the Pressurizeh ,

 ) Cross-Sectional Flow Area of the Pressurizer Surge Line .i Line/ Primary Coolant Loop 40 Connection Surge Line) l b \ ,

f

 /

f ',

 \ $$D x . Mh.o5 3*

ts l

s

 ~ .c -4 b4 " ~3 $

• Refer to Figure 3.6.2-2 for location of postulated breaks in Reactor Coolant Loop.
• ** Less break opening area will be used if justified by analysis, experiments, or considerations of physical restraints such as concrete walls or structural steel.

1 ATTACHMENT STP FSAR ST HL AE-/')VV PAGE%/0F sy; TABLE 3.6.2-4 ,PRIMARY PLUS SECONDARY STRESS INTENSITY RANCES AND ,CUMULATIVE USAGE FACTORS AT DESIGN BREAK LOCATIONS /IN REACTOR COOLANT LOOP NReference Fatigue Loadings Used in f Analysis Reference Analysis / STP Loadings

 '\ j / - M fin-lb) M, (in-lb)

M (in-lb) Thermal M (in-lb) Tfiermal Cumulative OBE ppansion OBE Expansion Fa r ST(psi) 5 /S Moments Moments Moments Momen a ato N /

 \/ x 404 /  ,

413 \415 (Leter) \4 40 484

 ? \\ \ \ \ \ \, \ \ \

1. The loop closure weld, residual heat removal line connection, accumulator line connection, and surge line connection locations have not been included in this table since selection of these locations for postulated breaks is independent of detailed stress and fatigue analyses. Also, node numbers are defined in reference.

 ,. 2. Information to be provided at completion of RCL analysis.

3. SI = maximum primary plus secondary stress intensity range computed using equation 10 si NB-3553 of the ASME Code, Section III.

3.6-86 Amendment 53

s ATTACHMENT GE d e7 IGENERATOR VESSEL 43 8PUMP I5

 \ NOTES BREAK LOCATION 10 6 ,

i PLAN VIEW

 / :

T' erea sxnez \

 \t /i \ STEAM , \ GENERATOR g \ / Z

(.

 \

Te %4

 \ \

S O' REACTOR PRESSURE Ag_ PUMP i 4

 \ VESSEL \ . 5 j

l( , 1 9

ELEVATION

\

X

 \ \

SOUTH TEXAS PROJECT UNITS 1 & 2(~ Location of Postulated Breaks in Main Reactor Coolant Loop

 #igure 3.6.2-2 ' Amendment 40

STP FSAR ATTACHMENT ST.HL AE- Met /PAGE %OF #9 9 3.8.3.1.6 Interior Fill Slab: Tha interior fill slab is 24 in. thick and is placed on top of the foundation mat liner plate. This slab provides protection for the foundation mat liner from any. missiles generated in the -1 primary loop compartments and from the effects of temperatures induced by a C7 DBA. -Reinforcement is provided to resist temperature and shrinkage forces.3.8.3.1.7 Polar Crane: A polar crane consisting of a 417-ton (unit 1)/ 40 500-ton (unit 2) main hoist, and a 15-ton auxiliary hoist supported on twin bridge girders is provided inside the RCB for use during construction, mainte-nance, and repair operations.The crane moves on a~ circular rail, which in turn is supported on girders. l40 Brackets anchored on the cylindrical wall through the liner support these girders (see Figure 3.8.1-6). The polar crane is anchored to the rails with l40 mechanical guides to prevent its derailment when subjected to earthquake forces.Girders and brackets supporting the polar crane are designed to the same loading combinations as the crane itself. The crane is assumed to be loaded 40 with its maximum operating load of 352 tons under both OBE and SSE.3.8.3.1.8 Reactor Coolant System Component Supports: The support structures are of welded and/or bolted steel construction of-linear and plate types. These supports are tension and compression struts or beams and columns. The supports permit unrestrained thermal growth of the supported system but restrain vertical, lateral, and rotational movement resulting from seismic and pipe-break loadings. This is accomplished using pin-ended columns for vertical supports and girders, hydraulic snubbers, and tie rods for lateral supports.

 . {i Shiaming and grouting enable adjustment of all support elements during erec-tion to achieve correct fitup and alignment. Final setting of equipment is achieved by shimming and grouting at the building structure / support interface.

p 3.8.3.1.8.1 Reactor Vessel Supports - The reactor vessel supports con-sist of individual air-cooled, plate-type support pads as shown on Figure 3.8.3-1. One pad is placed under four of the vessel nozzles and is supported by an embedded plate-type structure which distributes loads to the primary shield wall. Two additional embedded plate type supports transfer lateral l 40 forces to the concrete.In addition to transferring loads from the vessel to the supporting structure, the pads also provide for the passage for cooling through the support to l 40 lprevent excessive primary shield wall concrete temperatures.I r The vertical upward force on the reactor vessel due to cavity pressurization is sisted,by (1) 'dweight o%e vessel and i r 2) the Ind{heJcavity restra H rovided b he, attachef primary coolant loo ping. T pressyr8htion forces acting on the vessel'are restricte Mo acceptable level

 /M _hy_ judicious design-ofH:he-flow geometry of-the primary. shield. wall annulus M he seal plates located at the upper reactor cavity are used r lini: :h: ; ': Ire d pr::: r:: 9 r-re:::c: _ 1 ~ .lii..g frea: ;im f m..;hl 40 k;h ;f ;he y.1-m y by pipir;;-r :11 r to provide shielding from neutron and gamma streaming.

3.8-51 Amendment 40

:_:: ==-==== :: -

All ACHMENT

 .. ST-HL AE 194/(

PAGE 49 0F /A?

 ~

INSERT Page 3.8-51 The original design basis postulated pipe break locations in the RCL are described in Reference 3.6-1. The primary RCL components and supports design' were based on these postulated break locations. A detailed fracture mechanics evaluation, as described in Reference 3.6-14, demonstrates that the probability of rupturing the RCL piping is extremely low under design basis conditions. Therefore, postulated RCL ruptures and the following associated dynamic effects are not included in the design basis: missile generation, pipe whip, break reaction forces, jet impingenent forces, decompression waves within the ruptured pipe,and pressurization in cavities, subcompartments and compartments. The dynamic effects from ruptures in Class 1 branch lines and other high energy piping are reviewed to verify that the effects are bounded by the current analyses.1 4

 ~

7603N:0288N/12

FSAR. ATTACHMINT ST-HL-AE 114/4 PAGE VJ OF Ie2'7 The blowdown analysis which determ hra the sdcqurcy of the reactor vessel ,eupports is comprehensive in that :it includes t^'::1, :L. . 10f 7.;:r -l

 <---E: _is the ef f ects of the inydraulic forces in the loop piping. ,

J m , l_: t: 2:;;; a. n , v 6im .L il;;11 ...;_L M at;;iatirer4 i ,a:al:ted pi r L'==ke ;h; 5;aisu uf ihe Vioe 1 auPP m = ill m - i=v re

 ,lif: .,7.- - .

3.8.3.1.8.2 Steam Generator - The vertical supports for the SG (see Figure 3.8.3-3) consist of four vertical columns bolted at top to the l40 vendor-supplied columns and at bottom to the floor slab. The lower lateral cupports consist of supports attached to the walls of each SG subcompartment cud bolted to the vendor-supplied beams. The upper lateral supports consist cf supports attached to the walls of each SG subcompartment and bolted to the vendor-supplied ring girder around the generator shell connected to hydraulic anubbers and supported by struts on the compartment walls. Loads are trans- l 40 ferred from the equipment to the ring girder by means of a number of bumper blocks between the girder and generator shell. ,3.8.3.1.8.3 Reactor Coolant Pump - The RCP vertical supports consist of three vertical columns (see Figure 3.8.3-4) bolted at top to the vendor sup- l40 plico columns and at bottom to the floor slab. The lateral supports consist of three supports attached to the compartment walls and bolted to the vendor-supplied tie-rod supports.3.8.3.1.8.4 Pressurizer - The pressurizer (see Figure 3.8.3-5) is sup- l 40 ported at its base by bolting the flange ring to the supporting floor slab.In addition, four lateral supports are provided which are attached to the ,compartment walls and bolted to the vendor-supplied supports which bear l40 cgainst the vessel lugs.3.8.3.2 Applicable Codes, Standards and Specificaticns.3.8.3.2.1 Codes, Specifications and Standards: The following codes, otandards, and specifications are used as a basis for the design, fabrication' 40 construction, testing, and surveillance of the Containment internal structure.Different issue dates of these documents may be used provided they meet the minimum requirements stated herein.

1. American Concrete Institute ACI 211.1-70 " Recommended Practice for Selecting Proportions for Normal Weight Concrete" ACI 214-65 " Recommended Practice for Evaluation of Compression Test Results of Field Concrete" 29 ACI 304-73 " Recommended Practice for Measuring, Mining, Transporting and Placing Concrete" ACI 305-72 " Recommended Practice for Hot-Weather Concreting" ACI 306-72 " Recommended Practice for Cold-Weather Concreting" ,

3.8-52 Amendment 40 W __ - . - - . - . - -

c

 / . Al TACHMEN /

v ST.HL AE 19 $STP FSAR PAGE @) OF i '7 The analysis results are then used to design the secondary shield wall uti-lizing the BSAP-POST OPTCON module. Concrete is assumed cracked whenever 40 tensile stresses are present.3.8.3.4.1.4 Other Concrete Internal Structures - Miscellaneous equip-ment, compartment slabs, and walls are analyzed using conventional beam / slab design assumptions and equations. 14adings for these structures consist of dead, live, seismic, pipe rupture, jet impingement, and subcompartment differ-ential pressures where applicable.40 l 3.8.3.4.1.5 Dynamic Analysis Procedures - Earthquake forces on the con- ;crate internal structures are determined by a dynamic analysis in accordance with the techniques described in Section 3.7. The dynamic loads thus deter-mined are then applied as static loads on the concrete structures, and a static analysis using the procedures described above is performed.The impact effect of the pipe rupture on the structural system is considered by either a conservative energy balance method or by an exact nonlinear time-history analysis.listed in Table 3.5-13. The structural system allowable ductility factors are h0. For impulse effects such as jet impingement forces, the structural system is allo'wed to respond inelastically with allowable ductility factors equal to the values listed in Table 3.5-13. .h0 49 7 Jet impingement loads due to na n coolant loop pipe br ks will be conserva-

 ,tivelycon(deredasastepfunct n whose magnitude is , tainedinaccordance}0 :

t with the methBds described in Sectio Q.6.2. Dynamic,lo'ad tors of 2.0 (or less if ,$tflitified) are being used to account for the dynamic na re of the load.3.8.3.4.2 Analysis of Steel Internal Structures: The steel internal structures are analyzed for all combinations of both service loads and nonservice loads as described in Table 3.8.3-2.

1. Static Analysis Procedures The steel internal structures are analyzed for static loads as appro-priate either by conventional methods which are well documented in appli-i cable textbooks, or by the Bechtel Structural Analysis Program (BSAP). 32 (See Appendix 3.8.A for a detailed description of the computer programs.)

2. Dynamic Analysis Procedures Modal response."epectra (MRS) analyses of the integrated floor systems were used for the analysis of seismic loads for design of beams and con- 32 l nections for the internal structural steel.

! 3. Dynamic effect of pipe rupture is discussed in Section 3.8.3.4.1.5.3.8.3.4.3 Design and Analysis Procedure for RCS Supports: The linear support systems for components for the SGs, RCPs, and pressurizers are h designed by elastic method of analysis. They are analyzed for and designed to 3.8-60 Amendment 49 r,. - _ _ _ , , . , - - - , - - _ , . . . , ,

ATTACHMENT STP FSAR ST-HL AE-17tN PAGE goof la'7

2. The small steam line break results in immediate reactor trip and ECCS l 41 l actuation.

3. A large shutdown margin, coupled with no feedback or decay heat, prevents heat generation during the transient.

4. The ECCS operates at a design capacity and repressurizes the RCS within a relatively short time.

3.9.1.1.8.3 Complete Loss of Flow - This accident involves a complete Ices of flow from full power resulting from simultaneous loss of power to all RCPs. The consequences of tit* incident are a reactor trip and turbine trip cn undervoltage followed by automatic opening of the Steam Dump System. For dscign purposes this transient is assumed to occur five times during the 40-year life of the plant.3.9.1.1.9 Faulted Conditions: The following primary system transients h:ve been considered faulted conditions. Each of the following accidents has baan evaluated for one occurrence; jeec brAnan

1. Reactor coolanthipe break O..r L15CR)

2. Large steam line break

3. FW line break

 '4 . RCP locked rotor

5. Control rod ejection

6. SC tube rupture 7.

Safe Shutdown_pTNSERT- - leep Earthquake brAM.h (SSE [) L,e op ES/ANCb 3.9.1.1.9.1 Re actor'CoolantsPipe Break "_:.,;rt0 cat - Following rupture of a reacter coolantlpipe resulting in a Leege loss of coolant, the primary cyatem pressure decreases, causing the primary system temperature to decrease.Because of the rapid blowdown of coolant from the system and the comparatively large heat capacity of the metal sections of the components, it is likely that th2 metal will still be at or near the operating temperature by the end of bicwdown. It is conservatively assumed that the ECCS is actuated to introduce water at a minimum temperature of 32*F into the RCS. The SI signal also l41 recults in reactor and turbine trips.3.9.1.1.9.2 Large Steam Line Break - This transient is based on a com-plate severance of the largest steam line. The following conservative assump-tions were made:

1. The reactor is initially in a hot, sero-power condition.

2. The large steam line break results in immediate reactor trip and in actu- l2 ation of the S1S. l41 .

3.9-12 Amendment 41

ATTACHMENT i f

 .- ST HL AE /7#f l ~* PAGE 6 /0F 1A/7 1 J

INSERT Page 3.9-12 The original design basis postulated pipe break locations in the RCL are described in Reference 3.6-1. The primary RCL components and support designs were based on these postulated break locations. A detailed fracture mechanics evalcation, as described in Reference 3.6-14, demonstrates that the probability of rupturing the RCL piping is extremely low ur. der design basis conditions. Therefore, postulated RCL ruptures and the following associated dynamic effects are not included in the design basis: missile generation, pipe whip, break reaction forces, jet impingement forces, decompression waves within the ruptured pipe, and pressurization in cavities, subcompartments and

 . compartments.

A I"l l -I l\7603N:0288N/15

ATTACHMENT STP FSAR ST-HL-AE PAGE M OFra / 7(t[ G 3.9.1.1.9.3 Feedwater Line Break - This accident involves a double-ended rupture of the main W piping from full power, resulting in the rapid blowdown of one SG and the termination of main W flow to the others. The blowdown is completed in approximately 43 seconds. Conditions were conservatively chosen to give the most severe primary side and secondary side transients. All AW flow that is delivered to the faulted SG exists at the break. The incident is 54 l5 terminated when the operator manually. terminates flow to the faulted loop.3.9.1.1.9.4 Reactor Coolant Pump Locked Rotor - This accident is based on the instantaneous seizure of an RCP with the plant operating at full power.The locked rotor can occur in any loop. Reactor trip occurs almost imme-diately, as the result of low coolant flow in the affected loop.3.9.1.1.9.5 Control Rod Election - This accident is based on the single most reactive control rod's being instantaneously ejected from the core. This reactivity-insertion in a particular reE lon of the core causes a severe pres-sure increase in the RCS so that the pressurizer safety valves lif t and also causes a more severe temperature transient in the loop associated with the affected region than in the other loops. For conservatism, the analysis is based on the reactivity insertion and does not include the mitigating effects (on the pressure transient) of coolant blowdown through the hole in the vessel head. vacated by the ejected rod.3.9.1.1.9.6 Steam Generator Tube Rupture - This accident postulates the double-ended rupture of an SG tube resulting in a decrease in pressurizer level and reactor coolant pressure. Reactor trip will occur due to the resulting SI signal. In addition, SI actuation automatically isolates the W l41 lines, by tripping all W pumps and closing the W isolation valves. When this accident occurs, some of the reactor coolant blows down into the affected SG, causing the shell side level to rise. The primary system pressure is reduced below the secondary safety valve setting. Subsequent recovery proce-dures call for isolation of the steam line leading from the affected SG. This y accident will result in a transient which is no more severe than that associ-ated with a reactor trip from full power. Therefore it requires no special treatment, insofar as fatigue evaluation is concerned, and no specific number of occurrences is postulated.3.9.1.1.9.7 Safe Shutdown Earthquake: The mechanical dynamic or static equivalent loads due to the vibratory motion of the SSE have been considered on a component basir. .3.9.1.1.10 Test Conditions: The following primary system transients l- under test conditions are discussed:l 1. Primary side hydrostatic test

2. Secondary side hydrostatic test 3.9.1.1.10.1 Primary Side Hydrostatic Test - The pressure tests include both shop and field hydrostatic tests which occur as a result of component or system testing. This hydro test has been performed at a water temperature which is compatible with reactor vessel material ductility requirements and a test pressure of 3,107 psig (1.25 times design pressure). In this test, the

($h6 3.9-13 Amendment 53 n, , . _. _ . . _ . _

STP FSAR ATTACHMENT ST HL AE-I7Vy PAGE43 OF /a7 RCS has been pressurized to 3,107 psig coincident with SG secondary side pres- ')sure of 0 psig. The RCS is designed for 10 cycles of these hydrostatic tests, which are performed prior to plant start-up. The number of cycles is indepen-dent of ether operating transients.Additional hydrostatic tests will be performed to meet the inservice inspec-tien requirements of ASME Section II. A total of four such tests is expected. l41 The increase in the fatigue usage factor caused by these tests is easily cov-ered by the conservative number (200) of primary side leakane tests that are considered for design.3.9.1.1.10.2 Secondary Side Hydrostatic Test - The secondary side of the SG is pressurized to 1.25 design pressure with a minimum vnter temperature of 120*F coincident with the primary side at 0 psig.For design purposes it is assumed that the SG will experience 10 cycles of this test. .These tests may be performed either prior to plant startup, or subsequently, following shutdown for major repairs, or both. The number of cycles is there-fore independent of other operating transients.3.9.1.2 Computer Programs Used in Analyses.41 3.9.1.2.1 NSSS Systems and Components: For the NSSS scope of study, the following Westinghouse Electric Corporation-developed computer programs have -been used in dynamic and static analyses to determine mechanical loads, I stresses, and deformations of scismic Category I components and equipment.These are described and verified in References 3.9-1Vand 3.9-1 .

1. WESTD - Static and dynamic analysis of redun ant ping systems

 -3

2. FIX - Time-history response of three-dimensional structures

3. displa ment WESDYN Piping systep p/E eress 5 "HE analysis SCENfrom-time

 /NcoRPourehistord mre 8Esoyd -

data - Thc56 c4PA b' lit

4. HRUST - Hydraulic loads on loop components from blowdown information

5. WESAN - Reactor coolant loop equipment support structures analysis and ,

evaluation

6. WECAN - Finite element structural analysis 5

7. ICES STRUDL-Il - aly is of RCS support structure 4110.

g, yuLTT~lEtBOP Jinear a heum.-hyd elastic rAvIkf ameNeocloet syshm dya.mes 1 3.9.1.2.2 Systems and Components: For the BOP secpe of study, the following public domain and/or Bechtel Power Corporation developed computer programs have been used. ,3.9.1.2.2.1 ME101 Program - ME101 is a finite element computer program which performs linear elastic analysis of piping systems using standard beam 41 '. . .theory techniques. The input data format is specifically designed for pipe ,stress engineering and the English system of units is used. ,3.9-14 Amendment 41 y . . . . . _ . . . .

 - ~

[~ATTACHMENT -STP FSAR ST HL-AE-(7W .PAGE Sr/ OF 129 l

 \

3.9.1.2.2.21 CE035-BASEPIATE II_ - A descripti'on of the program iz C {53 i

 )

C provided in Appendix 3.8.A. Documentation of the verification in maintaived in the Eechtel Information Services Library.3.9.1.2.2.22 CE413-VEIJ) - The VELD program is used to size fillet welds l53 !for connbetiens of vide flanges, tubes, pipes, angles, and channel. The i jprogram computes veld aires based on AISC, NF, B31.1 and minimum veld for 51 minimum heat transfer. The program is verified by hand calculations.

 .9.1.2.2.23 RELAP5/REPIPE - Thermal Hydraulic Transient Anal,y,,s_is,- l53 '

RELA /REPIPE is used for analysis of fluid transients in the piping sys.t.sm equations of conservation of mass, energy, and momentum are tolved in one dimension for steam and/or water flow. The effects of noncondensible gas on steare/ liquid flow are considered in the equations. REPIPE is the post proces-sor which gives the forcing function for use in ME101. The program verifier-tion report is on file with Bechtel Data Processing.53 3.9.1.2.2.24 ME150 FAPPS - FAPPS (Frame Analysis Program for Pipe Supports) is an inter-active computer program for the analysis and design of pipe supports. It optimizes sember sizes, welds, baseplates and embedsents ,based upon various user-specified design limitations. The program allows load combination by algebra'.c, absolute, or SRSS methods. The program has been .verified against Bechial Sta.idard Structural Analysis Program CE901 (STRUDL) and hand calculations.3.9.1.2.2.25 ME035 BASEPLATE - ME035 is a finite element-computer 53 I program for the analysis and design of baseplate. The program has important features like automatic mesh generation, availability of standard attachments, .multiple plate thicknesses, and different printout options. The program has C~J been verified against CDC Baseplate II (bechtel CE035).53 3.9.1.2.2.26 ME225 ANCHORPL. ATE ME225 is used to analyze and design ,interface anchors between non-seismic piping end seismically designed piping. 'Program has been verified by manual calculations.3.9.1.3 Experimental Stress /.nalysis. Experimental stress analysis '

 $0 method has not been used for any seismic Category 1 ASME B&PV Code, Section 0210 III wechanical system or equipment. ,'393 e 3.9.1.4 Considerations for the Evaluation of the Faulted Conditions.

nts and Supports in e- .9.13.1 Stress Criteria for Class 1 Com Nucle (arSteaseSupplySystemScope: The structural stress analyses performed l on the (NSSS Components and Components Supports () consider the loadings shown in Table 3.9-2.1. These loads result from thermal expansion, pressure, weight, OBE, SSE, W ' - i n i: L'N, and plant operational thermal and pres-aure transients. L g he m d p*p g st e.se k.

 .3.9.1.4.2 Analysis of the Reactor Coolant Loop and Supports: The loads used in the analysis of the reactor coolant loop piping are described in detail below: ,

3.9-19 Amendment 53

 ~

 " = ~ ~ . - -

SIP 73Ak 3 }U TACHMENT ,ST.HL. AE. t 79 t d PAGEJk70FJAr7- >

1. Presecre Pressure loading has been identified as either.mestrana design pressure )

or genyral operating piessure, depending upon its applicatien, Try uem-trane desich pressbre is used in centection with the longitudinal pree-sure stress and sinikam vall thicknes's calculeticus in cecordence with the ASME B6Py Code.The term " operating precsure" h.as teen used in connecticn with determina-t(on of the system deflectsons and suppgrt forces. The steady-state ,operating hydraulic fcr.c6s based on the systec initial pressure are applied as gene.rsi operating pressure leads to the reactor coolant loop model at change in ditettion er flow area.

2. Veight A deadveight analys1s has been performed to meet code requir! nts by he piping applying a 1.0g, load downward on the complete piping system.

in mas'1gned a distributsd mass er veight as a function of its , operties.Thic method providee e distributed loading to the piping syste as a function ci the weight of the pipe and contained flafd during r.ormal

 . operating condit$on.s.

3. Seismic ,

Tf.e forcing funce.icns for the reactot coolant loog reismic piping anal-yses have been der.tved froc dynamic response analyses of ,thG Ksactor ,Containment Building (RCB) subjected to scismig grosnd motion, Input is }in the form of floor re.tponse apectrum curves at varicas elevations with- .in the RC3. .For the OBE and SS7. ceismic analyscG, 2 and 4 percent erftical dumping, srespectively, ha se been uped in the reactor coolant ico;/sypports systen analysis. ,in the responss spectrum retbod cf ena11 sis, the total response loading +obtained from the seismic analysis coorists of two parrs! the inertja ;' response loading of the piping'systep and the differential anctor move-' ments loading. Tvo se.tc of ceismic aczents are required to perform an ASME Code analytiss ite first set includen only the m?ments resul;ing' from inertia effects, and these moments are urc6 it the resultant moment. .M,, value for Equa.tienc 9 and 13 of NB-3650. The.se;ond set includes the recents reculting from stirmic anchbr moticns and is used in Equatic.ns 19 and 11 of UB-3650. Diff erential anchor 8.novement is discussed in Section l3.7.

 \

4. Loss-of-Coolant Accident -- ks y_ gypANC s ,

Blowdown loads have be sn dev, eloped in the ir:- _-1 -1.1.. hen r.eactor coolant loops as a result of tra;rient flow and pressure fluctuations f ollowing a postulatedfpipe break W-: . d;'._ A.. .... .. 4...& 'p; . 'Structural consideration of dynamic effecti cf peetulated pipe break requires postulation .cf a finice number of break locations. Postulated a 3,pipe break locations are given fu Section 3,6, 3.9-20 Amendment 41 <

 . . . . - .=e 0' _ _ --. _ - -- .- . . _ _ - ._

I -ATTACHMENT -SIP PSAR ST-HL AE- r:r

 ,PAGE.56 OF I q L._. 2cq r f i " - l g ' me-histcry dynamic analyses have been l41 performed for these postuisted .reak cases. Hydraulic models have been ! < used to generate time-dependent hydraulic forcing functions used in the analysis of the reactor ec,elent loop for each break case. For a further description of the hydraulic forcing functions, refer to Section 3.6.

I

5. Transients Ths ASME M PV Code requires saticfection of certain requirements relative to operating transient conditions. Operating transients are tabulated in

 . Section 3.9.1.1.

The vertical thermal growth of the reactor pressure vessel (RPV) norzle centerlines has been considered in the thermal analysis to account for equipment nozzle displacements se an external movement.The hot moduli of clasticity, E, the crefficient of thernal expansion at the metal temperature,0 , the externt1 movements transmitted to the piping due to vessel growth, and the ter.perature rise above the ambient temperature,6T. defint the required input data to perform the flexibility analysis for therttal expansion, To provide the necessary high degree of integrity for the RCS, the tran-sient conditices eclected for fatigue evaluation are based on conserva-tive estimates of the magnitude and enticipated frcquency of occurrence of the tecperature and pressure transients resulting from various plant operation conditions.C, .3.9.1.4.3 Reactor Coolant Loop Models and Methods: The analytical meth-ods used in obtaining the solution consist of the transfer matrix method and stiffness matrix forraulation for the static struetural analysis, the response spectra c.ethod for seismic dynamic analysis, and time-history integration method for the 1.0CA dynamic analysis.The it.tegrr.ted reactor coolant / loop supports system model is the basic system model used to ccepute loadings on components .cceponent supports, and piping.The system model includes the stiffness and rass characteristics of the reac g tor coolant loop piping and components, the stiffness of supports.(the stiff-ne:s of auxiliary line piping vhich affcets the system.7'r'-; = e&*ts. The deficction solution of the entire system is obtained for the various 1cading cases from which the internal member forces and piping stresses are calculated.

1. Status The reac r coolant loop / supports system model, constructed for the WESTLY @ computer program, is represented by an ordered set of data which numerically describes the physical system. Figure 3.9-6 shows an 1sem.etric line schematic of this mathematical rodel. The SG and RCP vertical and lateral support members are described in Section 5.4.14.

(i

• 3.9-21 Amendment 41

STP FSAq I ATTACHME6f l ST-HL-AE./My

 .EAM MOfIQ _

The spatial geometric description of the reactor coolant loop mode 3 is based upon the reactor coolant loop piping layout and equipment drsvinF8 The node point coordinates and incremental length of the members are

 .)

determined from these drawings. Geometrical properties of the piping and elbows, along with the nodulus of elasticity, E, the cca:fficient of ther- -mal expension,0 , the average temperature change from enbfent terpera-ture. AT, and the weight per unit length, are specified for each e.lement.The primary equipment supports have been represented by stiffness matri-ces which define restraint characteristics of the supporta. Due to the symmetry of the static loadings, the RPV centerline has been represente:3 by a fixed boundary in the system methematical medel. The vertical ther-mal growth of the RPV nozzle centerline has been considered in the con- .struction of the model.The model is made up of a nucher of sections, each having an overall transfer relationship forsted from its group of elements. The linear elastic properties of the section have been used to defice the stF fness I matrix for the section. Using the transfer relationship for a sect 1on, the load required to suppress all deflections at the ends of the section 'arising frcm the thermal and boundary forces f6r the section have been obtained. These loads have been incorporated into the overall load vector.After all the sections have been defined in this matter, the everall stiffness matrix and associated load vecedr to surpress the.draflecticu of all the network oeints are determined. $y inverting the stiffness matriy, the flexibility matrix 1s determined. The fler.ibility 3. atrix is I multiplied by the negative of the load ver. tor to deter =ine t'he betwork 'point deflections due: to the therr.a1 ar.$ bour.dary force effects. Using the general transfer relationship, ths deflections and inteenal forces are then determined at all node points in the system.The static solutions for deadweight, therral, an'i gen ral prassure leading conditione are obtained by using the WESTDYT.e ecmputer program.The derivation of the hydraulic 1 cads foi the LOCA analysis of the loop is covere.4 in Section 3.6.2.J 2. Seismic The model used in the st'atic analysis has been nodified for the dynamle analysis by including the msss characteristics of the pipi.y and equip-ment. The effect of the equipment motion on the reactor coolant l 41 loop / supports systen is obtaf ned by modelir.g the mass and the stif fr:ess characteristics of the equipsient in the everall systes 7:odel.The SG has been typica.lly represented by four dicerete masses. The 2cwer mass is located at iche elevation of the lower rupport attachtrer: point.The second mass has been located at the SG upper support elevetion, the ,41 third mass has beer. located at the center of the cpper shell, und the -fourth teaca is located at the top of the &t'.tcm generator.The LCF has been typica'117 represented by a two-discre te-mass ciodel. The icwer mess is 3ocated at the intersection of the centerlines of the pump j 3.9-22 /dendmcat 41

 ,-m - -

v ,p

 ... _ , i S7P FCAR AlTACHMENT l w

ST HL5.5hE-PAGE OF neplf ur _ _suction ont discharge noznics. The upper masP. is located naar the canter of gravity of tna noto s.The RPV And coru intatnals have been t!' pics 11y repr'esented by approxi-mately 10 discrete masses. The masses are lumped at various locatioAs.along the lan,Tch d 'tht vessel and altng t' e length of the representation of the ente internali;.The conr,'onent lateral Suppcrts are it. active during plant heatop and { 41 cocidown and nosmal plant et'etating cenditions. However, e.hese

 . restrainte becorne actf.ve when the p.lant is at power and unMr the rapid inocions Cf the resetor coolant loop ompelunts' tbar 6ciur from the dynam-ic '.osdings, ctd ate represented by stiffnes;t matrices nndfor individual tension or cotyrcasion spring nambers in tt.e rlynamic model. The analyses have been performe( at th fril-pover condit,toa.

The ranponsc. spectra methort enployr. the lumped-n.1se technique, linear slaatic properties, and the 1,rinciple of pedal cuperposition. Tte floor 'respenre spectra hbe been applied sion'g both hori;ontal ar.as and the vertical exis sim'altareccsly.Fron the mathavatical descr.ipcirn cf the 93atem, the everall atiffness matrix, K, has be!n developed from the individual elawent stiffnesJ .matrices using the trat.sfer natriz. rettbd. After deleting the rcws and columns representing rigid restrainte, the stiffness n' atrix has been revised to obtain a reduced stiff ness mA~trix, KR. associated with riass d,*grees of freedom only. From tle nass matrix and the redcced sciffness C~,. matr'x, tLe nae. ural frequencies and the ncrmal godos are determined.

 "'he model part( ipetion factor riatrix is computed and combined with the appropriate: resp;nse spectra vAlue t; g!ve thi morial amplitude for each mode. The total medal acplitude L a been obesined by trking the square '

yoot of the attm .of the squares of the c?ntributions for eat.h direction.The modal emplitudes are the.n converted to ditplaceeents 16 the globa?coofdinate system and appifed to the corresponding mass point. Trom these datu, the forces, m ente, deflections, rotations, support reac-tions, and piping steersais .have been calculated f or all significant modess lThe total seismic response is computsd by combining the contributions of the significant mcder as described in Section 3.7. l41

3. Lear-of-Coolant Accident

 --- by /Nchdiq The mathematical podel used in the ststic antlyses has been modified' for the LOCA analysts b q:: " ^^ --- -- : rf th:::a a l:n: -,, ' '- ~" " M 1.L J '. loca z._ LJ Hiu dm 12:1rf:: f t 9-A h in th; .sf the mass pr.epe- 'mh characteristic $of theg:n,2ing t_a_ h t ien , tr:::: ch andYequipment.

QaiTrins h dynam u aegrees c' #-.-:? a-d 2::::' :n:::h :14 T L; *h ' " i" "d" 'I" ' A a4&hav eeisi -f 1. The natural fre quencies and eigenvectors are deter-mined from this M ^ - 2: q model.

 - r m, ka 3.9-23 - Amendment 41

S19 F3AR ATTACHMENT ST HL-AE- 19t}t- ) ,PAGE 59 0F /2 '7

 *The time-histnry hydraulic forces at the sede points have been ecebined-to ob".ain cas forces and r.onents acting at the correspoeding structural ' } '

i lenped-mass tiede points. .The dynamic structural solution for the full-power LOCA and steam line breer. L'as been obtained by using a modified-predictor-corrector-integra-tion technique and normal rede theory.When alements of the eystem can be represented as single-acting neebers(;ension or compression setbars), they have been considered as nonlinear i elements, Which are rep, resented mathematically by the combination of a gap, a spring, and a viscous damper. The force in this nonlinear element is treated as an extarnellj applied force in the.overall normal mode molution. Multiple nonlinekr elements can be applied at the same node.

 $f necessatey, i 3

The time-history solution has been perforned in subprogram FIXFM The input to this subprogrem consJsts of the natural frequencies, normal podes, applied forces, and nonlirear ele =ents. The natural frequenc.ies and normal modes for the modified reactor coolant loop dynamic model have .been deterained with the WESTDDCTprograe. To properly simulate the release of the etrain . energy in the pipe, the internal forces in the systen at the postuleted break location due to the initial steady-state .hydraulic forces, thetual forces, and weight forces are determined. The release of the strr.in energy is accounted for by applying the negative of these internal forces as a step function loading. The initial conditions 'are equal to zero because the solution is only for the transient problem (the dy~namic re9 pense of the system from the static equilibrium }position). The time-history displacement solution of all dynamic degrees ,of freedcm has'been obtained using subprogram FIXFMUand employing 4 'percent critical damping. %3 The LOCA displacements of the RFV have been applied in time-history form as input to the dynamic analysis of the reacter coolant loop. The LOCA analysis of the RPV includes all the forcer acting on the vessel, including 16ternale reactions, '-"'"and loop mechanical -! loads-. The RPV analysis is described in Section 3.9.1.4.6.I The tine-history displacement responre of the loop is used in computing isupport _ loads and in performing the stress evaluation of the reactor coolant loop piping. .The eupport loads have been computed by multiplying the support stiffness 41 matrix and the displacement vector at the support point. The support loads are used in the evaluation of the supports.

 -3 The time-history displacements of the FIXFM subprogran have been used as l input to the WESTDYN-2 subprogram to determine the internal forces. l41 deflections, and stresses at each end of the piping elements. For this calculation, the displacements are treated as imposed deflections on the reactor coolant loop masses. The results of this solution have been used in the piping stress evaluation. " .

3.9-24 A$nd=ent41 is

STP FSAR ATTACHMENT -ST.HL AE /7&/PAGE (p0OF ly/The resultant asyr.rutric m arnal pressure loads on the RCP and steam

generator resulting fr: a ostulated pipe rupture and pressure buildup in the loop compartments are applied dynamically to the reactor coolant loop model. This model is the same integrated RCL/ supports system model ,

used to compute loadings on the components, component supports and RC ]piping, as discussed above. The response of the entire system is j obtained for the various external pressure loading cases from which the ;internal member forces and piping stresses are calculated. The resultant ;equipment support loads and piping stresses resulting from the external !pressure loading are added to the support loads and piping stresses cal- '

 . culated using the loop LOCA hydraulic forces and V motion.

The break locations considerb for brAuck subcompartment Piff. kfA pressurization are 5 4 those postulated h RCLfi^C/ .r.;1,_1 , an discussed in Section 3.6 Q22.2 2.2 L f _ ..__:.5 i Th e; ef th::: ireaks h;; h:cr. identifi-? se , Q110.i nving th:::t:. nific nt effe a Jm RC /. w ii.. eje m . O.::: _ ,ctor coolant pump outlet nozzle break, steam genera et '

iozzle b . and the steam generator inlet elbow on the intrados.

The RPV inle zie and RPV outlet nozz aks are considered in the

 .4.6.)

(PV dynamic analys ce Section .Based on the design ~of C orts and pipe restraint system, the '

 .ise of single-end reak areas (equa cross-sectional area of th e pipe) is a e ervative upper bound for the RC nozzle and steam i generat utlet nozzle breaks. The steam generator inle ow split >re s also assumed to have a break area equal to the cross-se al. ,

t a nf the nine. == M e -ed i C,. ...

 ~

l The reactor coolant loop piping is evaluated in accordance with the -faulted condition criteria of ASME III, NB-3650 and Appendix F. The loads included in the evaluation result from the SSE inertia loading,

 - o.ou. e , yi.nute, =.^.peophydraulicforces,asymmetricsubcompart-i ment pressurization forces, rand reactor vessel motion. Individual 1 I

I loadings at critical stress ^ locations are combined and primary stress intensities are calculated for the combined load sets. The primary stress intensities at all locations are within the faulted condition l42 [stress limit.RC.Lbruc.hiyE. NCd Nd la"b ['N JET WR8,os4vINea f efe berAs.I i

4. Transients i

i j Operating transients in a nuclear power plant cause thermal and/or pres-i sure fluctuations in the reactor coolant fluid. The thermal transients l cause time-varying temperature distributions across the pipe wall. These temperature distributions resulting in pipe wall stresres any be further isubdivided in accordance-with the ASME B&PV Code into three parts, a

uniform, a linear, and a nonlinear part. The uniform part results in general expansion loads; the linear part causes a bending moment across the wall; and the nonlinear part causes a skin stress. ,

i lThe transiente as defined in Section 3.9.1.1 are used to define the fluctuations in plant parameters. A one-dimensional finite difference heat conduction program has been used to solve the thermal transient oroblem. The pipe has been represented by about 100 elements through the l4 l thickness of the pipe. The convective heat transfer coefficient employed ll I Amendment 41, 3.9-25 ip,, .. . .. . . . . . . , . - - . , . .

STP FSAR ATTACHMENT ST HL AE l'1L/ d PAGE /,/OF lsG in this program represents the time-varying heat transfer due to free and '4 forced convection. The outer surface is assumed to be adiabatic while the inner surface boundary experiences the temperature of the coolant .fluid. Fluctuations in the temperature of the coolant fluid produce a temperature distribution through the pipe wall thickness which varies with time. An arbitrary temperature distribution across the wall is shown on Figure 3.9-8.The average through-wall temperature T , is calculeted by integrating the temperature distribution across the wall. This integration is per-formed for all time steps so that T As determined as a function of time.TA(* ~g' The range of temperature between the largest and smallest value of T g is used in the flexibility analysis to generate the moment loadings caused by the associated temperature changes.The thermal moment about the mid-thickness of the wall caused by the temperature distribution through the wall is equal to:l M=E O (X - ) T(X,t)dX The equivalent thermal moment produced by the linear thermal gradient as shown on Figure 3.9-8 about the mid-wall thickness is equal to: .l H2 )g = EQ 12 4Equating g and M. the solution for T y as a function of time is:AT y =E O( ~The maximum nonlinear thermal gradient, T occurs on the inside surface andcanbedeterminedasthedifferencebekw,eentheactualmetaltempera-i ture on this surface and. half of the average linear thermal gradient plus the average temperature.ATy (t)

 ~

AT21(t) = T(0 t) - TA(* 2

5. Load Set Generation.

1 A load set is defined as a set of pressure loads, moment loads, through-vall thermal effects, and the axial thermal gradient at a given l41 location and time in each transient. The method of load act generation ,l 1s based on Reference 3.9-2. The through-wall thermal effects are func- i tions of time and can be subdivided into four parts: ;i

a. Averege temperature. AT , is the average tenperature through-wall of i the pipe which contributes to general expansion loads.

3.9-26 Amendment 41 (r y _ - - - . , . . . . . - . - .- - . . . ,

STP FSAR ATTACHMENT ST HL AE /7d PAGE /pdLOF lAr?

b. Radial linear thermal gradient. which contributes to the throughwall

g. '

bending noment , ATy .c.Radial nonlinear thermal gradient, AT,,, which contributes to a peak stress associated with shearing of the surface.The axial thermal gradient, defined by discontinuity temperature. l 41 d.T ~B' represents the difference in average temperature at the cross-sections A on each side of a discontinuity.Each transient is described by at least two load sets representing the maximun and minimum stress state during each transient. The construction of the load sets is accomplished by combining the following to yield the maximum (minimum) stress state during each transient.e AT g e AT I

• T - T GA A OB 3 e Moment loads due to TA e Pressure loads This procedure produces at least twice as many load set- as transients for each point.

 ].

As a result of the normal mode spectral technique employed in the Eight seismic load sets analysis, the load components cannot be given signed values.are used to represent all possible sign permutations of the seismic moments at each point, thus assuring that the most conservative combination of seismic loads is used in the stress evaluation.For all possible load set combinations, the imary-plus-secondary and peak stress intenrities, fatigue reduction factor , K , and cumulative usage fac-tors, U, have been calculated. The WESTDYN@pr$ gram has been used to perform this analysis in accordance with the ASME B&PV Code, Section llI, Subsection NB-3650. Since it is impossible to predict the order of occurrence of the transients over a 40-year life, it is assumed that the transients can occur in any sequence. This is a very conservative assumption.The combination of load sets yielding the highest alternating stress intensity The next most range has been used to calculate the incremental usage factor.severe combination is then determined and the incremental usage factor calcu-lated. Thgsprocedureisrepeateduntilallcombinationshavingallowable cycles <10 .are formed. The total cumulative usage factor at a point is the summation of the incremental usage factors.The static and 3.9.1.4.4 Primary Component Supports Models and Methods:dynamic structural analyses employ the matrix method and normal modeThe theory equip-for the solution of lumped-pa ameter, multimass structural models. (1) ment support structure models are dual purpose since they are required:{~ to quantitatively represent the elastic restraints which the supports impose Amendment 41 3.9-27 97

ATTACHMENT STP FSAR ST HL-AE /NL/PAGE /,5)F IM upon the loop, and (2) to evaluate the individual support member stresses due to the forces imposed upon the supports by the loop. l5 --f IQ110.1 A description of the supports is found in_Section 5.4.14. Detailed models have been devploped using beam elements and plate elements, where apolicable.The steam generator lower support is shown in Figure 3.9-13. The struts are ,represented by single-acting springs in the RCL analysis; the columns are modeled as individual double-acting springs. The stean generator upper sup-port is,shown in Figure 3.9-14 A model for the STRUDL (Reference 3.9-1) computer program (Figure 3.9-15) is constructed for the steam generator upper lateral. support ring girder. Structure geometry, topology, member releases, and concrete flexibilities are included to accurately represent the behavior of the support system. Rigid spokes, extending from a point on the steam generator vertical axis to points where loads are transferred to the ring girder, are included in the model. The steam generator upper support model is used to determine the spring constants used to represent the support in the 5 RCL model. Q110.1 The reactor coolant pump supports are shown in Figure 3.9-16. Single-acting springs represent the tie bars and double-acting springs represent the columns in the RCL model. The brackets of the compression and tension tie bars have slotted pin holes which make the. members single-acting only.A three-dimensional finite element model is used for the RPV support struc-ture. The WECAN (Reference 3.9-16) computer program is used' for the support analysis. )Th:::_.. m La . . . . i..; e , e te e; .u. . .w. Inh 6 ..o .. ins, .u" a L -uJ 'b L

1' 1:;; '- ., .:.;:1' r:11 . ..-. . 1;;;;;; 2: f:L 2 .. . .u .m... .um FAub For each operating condition, the loads (obt Lned from the RCL analysis) :

acting on the support structures are appropg tely combined. The adequacy of A 'l each member of the steam generator supported eactor coolant pump supportsp "l

 ' ' '; -r::: is verified by solving the ASME III Subsection NF stress l and interaction equations by means of hand calculations and the WESAN (Reference 3.9-1) computer program. The adequacy of the RPV support structure i

is verified using the WECAN computer program and comparing the resultant(stresses to the criteria given in ASME III Subsection NF.l 41 "l 3.9.1.4.5 Analysis of Primary Components: Equipment which serves as [f part of the pressure boundary in the reactor coolant loop includes the SGs, Ithe RCPs, the pressurizer, and the reactor vessel. This equipment is American jNuclear Society (ANS) Safety Class 1, and the pressure boundary meets the This equip-f requirements of the ASME B&PV Code, Section III, Subsection NB. The ment is evaluated for the leading combinations outlined in Table 3.9-2.1. I 5equipment is analyzed for: (1) the normal loads of dead weight, pressure, and lQ110.[thermal; (2) mechanical transients of OBE, SSE, and pipe ruptures, including 1 the effects of asymmetric subcompartment pressurization; and (3) pressure and temperature transfents outlined in Section 3.9.1.1.i 3The results of the reactor coolant loop analysis have been used to determine %,l the loads acting on the nozzles and the support / component interface locations.l jThese loads have been supplied for all loading conditions on an " umbrella" t l l Amendment 41 3.9-28

ATTACHMENT STP FSAR ?ST-HL PAGE G40F AE-lW{V 7 Joad basis; that is, on the basis of previous plant analyses, e set of loads has been determined which should be larger than those seen in any single plant :

' [L~ analysis. The umbrella loads represent a conservative means of allowing detailed component analysis prior to the completion of the systen analysis.

Upon completion of the system analysis, conformance has been demonstrated betweed the actual plant loads and the loads used in the analyses of the eco-ponents. Any deviations where the actual load is larger than che umbrella have been handled by individualized analysis. d Seismic analyses have been performed individus11y for the RCP, the pres- :surizer, and the SG. Detailed and complex dynamic models have been used for the dynamic analyses. The response spectra corresponding to the building elevation at the highest component / building attachment elevation have been used for the component analysis. Seismic analyses for the SG have been'per-formed using 2 percent damping for the OBE and 4 percent damping for the SSE. 'The analysis of the RCP for determination of loads on the motor, main flange.and pump internals has been performed using the damping for bolted steel ,structures; that is, 4 percent for the OBE and 7 percent for the SSE (2 per-cent for OBE and 4 percent for SSE is used in the system analysis). This 'damping is applicable to the RCP since the main flange, motor stand, and motor are all bolted assemblies (see Section 5.4). The RPV has been qualified by static stress analysis based on loads that have been derived from dynamic analysis.Reactor coolant pressure boundary (RCPB) components have been further quali-fied to ensure against unstable crack growth under faulted conditions by per-

 ,. forning detailed fracture analysis of the critical areas of this boundary.

1 Actuation.of the ECCS produces relatively high thermal stresses in the system.Regions of the pressure boundary which come into contact with ECCS water are given primary consideration. These regions include the reactor vessel belt line region, and the reactor vessel inlet nozzles. h I. Two methods of analysis have been used to evaluate thermal effects in the 4regions of interest. The first method is linear elastic fracture mechanics j (LEFM). The LEFM approach to the design against failure is basically a stress 4 intensity consideration in which criteria are established for fracture insta-l bility in the presence of a crack. Consequently, a basic assumption employed in LEFM is that a crack or crack line defect exists in the structure. The essence of the approach is to relate the stress field developed in the vi-cinity of the crack top to the applied stress on the structure, the material properties, and the size of defect necessary to cause failure.The elastic stress field at the crack tip in any cracked body can be described l by a single parameter designated as the stress intensity factor, K. The mag- !i nitude of K is a function of the geometry of the body containing the crack, ,the size and location of the crack, and the magnitude and distribution of the stress.The criterion for failure in the presence of a crack is that failure will -occur whenever the stress intensity factor exceeds some critical value. For ,the opening mode of loading (stresses perpendicular to the major plane of the

 ,,,_ crack), the stress intensity factor is designed as K and the critical stress

' 7 intensity factor is designated K . Commonly en11ed the fracture toughness. 'l {- K IC is an inherent material prope ty which is a function of ternerature a 1

3.9-29 Amendment 41

 .,, . .. _ _ . . . . . -r- . , _ . , , - _ _r - _ . _ _ . . . . _ _ . - . - , , , _ _ _ _ _ _ , ~ , - - ~ . . - - - _ _ . . . . _ . , ,

ATTACHMENT STP FSAR ST HL AE /7W l PAGE(gg OF /F1 i lstrain m e. Any combination of applied load, structural configuration, O' ***crack -geometry, and size which yields a stress intensity factor, KIC' I # ) I rial results in crack instability.The LEFM Analysis Methods in ASME XI, Appendix A and ASME III, Appendix G are l used to perforb the fracture evaluation of postulated flaws to establish that the vessel integrity is maintained. This LEFM Analysis is considered accurate 41 in the elastic range and conservative in the elastic-plastic range. ,Therefore, for faulted condition analvses, LEFM is considered applicable for !the evaluation of the reactor vessel inlet nozzle and belt line region.In addition, it has been well established that the crack propagation of existing flaws in a structure subjected toThus, cyclicthe.loading can be defined in principles of LEFM are also terms of fracture mechanics parameters.applicable to fatigue growth of a postulated flaw at the vessel inlet nozzle and belt line region. 41 An example of a faulted condition evaluation carried outThis according report to the discusses procedure discussed above is given in Reference 3.9-3.the evaluation proc postulated LOCA), and concludes that the integritv of the RCPB would be main-tained in the event of such an accident.The pressure boundary portion of RCS Class 1 valves has been designed and analyzed according to the requirements of NB-3500 of ASME B&PV Code, Section III.

 }

Valves in sample lines connected to the RCS are not considered ANS Safety Class 1 nor ASME Class 1. This is because the norvles where theThis line hole connects to the prinary system piping are orificed to a 15/64-in. hole.restricts the flow so that loss due to severance of one of these lines can be! made up by normal charging flow.3.9.1.4.6 Dynamic Analysis of Reactor Pressure Vessel g for PostulatedbrC,s.h We.h Loss of Coolant Accident:3.9.1.4.6.1 Introduction - This section presenus3 '__ the method of computing af a l-.: _uif:-!the reactor pressure vessel response to a postulated 5 SGW4.The structural analysis considers simultaneous application of the Q22.2 time-history loads g ehe reactor _ vessel resulting from the reactor coolant Q110.1 loop mechanical loads &+1nternal hydraulic pressure transientscal.a

 /Qny weeees,.

r ir.;

-i:c y...__cfr:ti - (f:r r:2_: 2 L. ..i 1,. .:.

 ' - ' " The vessel is restrained by reactor vessel support pads and shoes beneath four of the reactor vessel nozzles, and the reactor coolant i

loops with the primary supports of the steam generators and the reactor cool-ant pumps.

 '4=4* *h .i..n in :alled in M prie:ry =hi=1 A ==11 "f;: '*- '----- 2::: aks. An upper b reak ing area of the 1 nozzle pipe 41 reak eae essel and pi ,

a ren is det ned fr reak areas calculated us es have shown t1 at similar plant analyses. Detai s

 ' elative motions nozzles, ev ith a limi ed ~,

1 iP e break the ho cold leg reactor ves essel support loads an e ~l a, would give the ghest reacto I renk

 ^ m y1_: r--t:, prf ril,- J a er tM i.f!;:::: :f::sc*e env4 v M ._-. m~

Amendment 41 3.9-30

 + - ' - - - - >

ATTACHMENT STP FSAR ST HL-AE /7W PAGE 8 0FloV/

 ;----"-i _uuu. ny o vuo tu. . i... ;L... ut ans, sua mun == vere re.uus ._ d .f ~

s t loads are determined. For completeness, two breaks outside the d(' .,- wall, o ch there is no cavity pressurization, were also anal , specif-ically, the pu tiet nozzle pipo break and the steam ge or inlet nozzle pipe break vare cons . In summary, four loss olant accident condi-tions were analyzed.

1. Reactor vessel inlet nozzle rea

2. Reactor vessel ou nozzle pipe break

3. Reacto olant pump outlet nozzle pipe break Steam menerator inlet nozzle pipe break 3.9.1.4.6.2 Interface Information t., ...u m ... m . x ity;::::

fr:ti:n lead. . . pivvie.m ,,, 4.. 1..J._ .. L_ a l ... :h: _ .;1, .; . a..u .L.J m e - " -- E ?_'.? ?.?. All sehss input information was developed within 41 Westinghouse. This information includes: reactor internals properties, loop mechanical loads and loop stiffness, internal hydraulic pressure transients, and reactor support stiffnesses. These inputs allowed formulation of the mathematical models and performance of the analyses, as will be described.3.9.1.4.6.3 Loading Conditions - Following a postulated pipe rupture-se.

 ^- -::ter ------I::::le, the reactor vessel is excited by time-history 5 forces. As previously mentioned, these forces are the combined effect of Q22.2 ^i. _ 7:- _-_.. . 3 reactor coolant loop mechanical loads; '"r--- Q110.1 crity preeeneteet'-- '- ----- and TF) reactor internal hydraulic forces.

The reactor coolant loop mechanical forces are derived from the elastic analy-sis of the loop piping for the postulated break. This analysis is described lin Section 3.9.1.4.3. The reactions on the nozzles of e H the W;ir-h: 2 piping t hqrs are applied to the vessel in the RPV blowdown analysis. L-- RC L! N:::t:: :_. n, r.a 1:e;i .. f x .:: ..... fui;.L. pi;; irr:'r:: :h: . x::la 1 ,o les from the steam and water which is released into the reactor y 1 hro the annulus around the broken pipe. The reactor cavity pressurized i symmetkcallywithhigherpressureonthesideofthebrok pipe resulting

 .n horizontal forces applied to the reactor vessel. S 1 er vertical forces trising from prtssure on the bottom of the vessel a the vessel flanges are also applied to t reactor vessel. The analys of asymmetric LOCA loads to 609. The cavity pre _ re analysis is described in 41

onsistent with NUR Section 6.2.

The internals reaction forces op from asymmetric pressure distributions Lnside the reactor vessel. a ssel inlet nozzle break and pump outlet sozzle break, the depres ization va ath is through the broken loop inle t sozzle and into the on between the cot rrel and reactor vessel. This egion is called e downcomer annulus. The in waves propagate up, dow l

 . ind around th owncomer annulus and up through the fu In the case of an 1 PV outle ozzle break and steam generator inlet nozzle brea he wave 1 asses rough the RPV outlet nozzle directly into the upper inte s regioi t, g.; ae surizes the core, and enters the downcomer annulus from the bott f e iP---'. "" . a , f::::::tirt; 1. L .. . we covncomer .uuulu. 1:

3.9-31 'knendment 41 r.

 -r- -- - _.- . _ _ _ _ _ _ _ , , _ .

STP FSAR ATTACHMENT ST HL AE s 7N PAGE47 0F IG/7 mu ,..? si:S:::5 1:11:r liff. m. .. m. :::::: Luiin tally av o.;L.y et and outlet non-h2 cor rrel th or the inlet break. For bo he )i le breaks, apressurization waves continue theji agation by reflec" initial 1 ton a ranslat through the reactor vessel g'id,le-dr. but J=5 r g, ::;;i::tir :::: __: th; grc;;;;: c"er* e

• Q22.

2 Th2 reactor internals hydraulic pressure transients were calculated including Q110.tha assumption that the structural motion is coupled with the pressure tran- 1 cients. This phenomena has been referred to as hydroelastic coupling or fluid-structure interaction. The hydraulic analysis considers the fluid-structure interaction of the core barrel by accounting for the deflec-tions of constraining boundaries which are represented by masses and springs.The dynamic response of the ucit barrel in its beam bending mode responding to blowdown forces compensates irc internal pressure variation by increasing the volume of the more highly pressurized regions. The analytical methods used to d2velop the reactor internals hydraulics are described in WCAP-8708 (Reference 3.9-7).3.9.1.4.6.4 Reactor Vessel and Internals Modeling - The reactor vessel is restrained by two mechanisms: (1) the four attached reactor coolant loops with the steam generator and reactor coolant pump primary supports; and (2) fcur reactor vessel supports, two beneath reactor vessel inlet nozzles and two bzneath reactor vessel outlet nozzles. The reactor vessel supports are described in Section 5.4.14 and are shown in Figures 5.4-12 and 3.8.3-1. The cupport " shoe provides restraint in the horizontal directions for downward reactor vessel motion.The mathematical model of the RPV is a three-dimensional nonlinear finite element model which represents the dynamic characteristics of the reactor vessel and its internals in the six geometric degrees of freedom. The model 41 was developed using the WECAN computer code. The model consists of three concentric structural submodels connected by nonlinear impact elements and Q22.2 stiffness matrices. The first submodel, Figure 3.9-11 represents the reactor vessel shell and associated components. The reactor vessel is restrained by Q110.1 the four reactor vessel supports and by the attached primary coolant piping.E:ch reactor vessel support is modeled by a linear horizontal stiffness and a vertical nonlinear element with lift-off capability. The attached piping is

 -represented by a stiffness matrix.

The second submodel, Figure 3.9-12, represents the reactor core barrel, neu-tron panels, lower support plate, and secondary core support components. This cubmodel is physically located inside the first and is connected to it by a stiffness matrix at the internals support ledge. Core-barrel-to-vessel shell impact is represented by nonlinear elements at the core barrel flange, core barrel nozzle, and lower radial support locations.The third and innermost submodel, Figure 3.9-12A, represents the lower core support plate, guide tubes, support columns, upper core plate, and fuel. The third submodel is connected to the first and second by stiffness matrices and nonlinear elements.5 -3.9.1.4.6.5 Analytical Methods - The time-history effects of ti: 2 1 1:7 Q22.2 L

 ,pr: _si;;;dc- Irr' . internals loads and loop mechanical loads are combined nnd applied simultaneously to the appropriate nodes of the mathematical model Q110.

1 3.9-32 Amendment 44

 , .---,,--o-- - - - - . . - ,-.- -. _..-.--- --,.-, - - , - . , --

ATTACHMENT STP FSAR ST-HL AE /7tpf PAGE 43 0Flcy7 of the reactor vessel and internals. The analysis is performed by numerically C,- intergrating the differential equations of motion to obtain the transient response. The output of the analysis includes the displacements of the reac-tor vessel and the loads in the reactor vessel supports which are combined with other applicable faulted condition loads and subsequently used to calcu-late the stresses in the supports. Also, the reactor vessel displacements are applied as a time-history input to the dynamic reactor coolant loop bboudsen 5 analysis. The resulting loads and stresses in the piping components and sup- Q22.2 ports include both loop blowdown loads and reactor vessel displacements. Q110 Thus, the effect of vessel displacements upon loop response and the effect of ,1 y loop blowdown upon vessel displacement are both evaluated.GACL brme.h P'P&- r 3.9.1.4.6.6 Results of the Analysis - As escribed, the reactor vessel and internals were analyzed for dowr postulated break locations. Table 3.9-12 summarizes the displacements and rotations of and about a point representing the intersection of the canterline of the nozzle :tn i d .. 6 L .u .uih

 & 5-9 ___ p::tui...d .. ... and the vertical centerline of the reactor vessel. Positive vertical displacement is up and positive horizontal dis-placement is away_from and along the centerline of the vessel nozzle in the loop in which thep reak was_yostulated to occur.] O.::: di .m yl e. . . .. a _

_ N -ted-we6 3--e-oeneervative br9 -:nngtea fer L6 poscul.L a pipe 41 ruptures the vessel inle outlet noz and double-e d ruptures at the pump ou t nozzle SG inlet nozzle locat s. e areas are istimated prior rforming the analysis. Follow he reactor coolant system struct a sis, the relative motions the oken pipe ends are abtained om the react coolant loop blowdo analysis. actual break 3pe area is then verif to be less t the estimated ar used in the,

 - ._1,7 _ i; _c.d;;;;:: th:t th: 2n:.ly;i; i;. ce...... Li m.

C.:. The maximum loads induced in the vessel supports due to the postulated pipe 2-break are given in Table 3.9-13. These loads are per vessel support and * ' - "are applied at the vessel nozzle pad. M is n .::r::tir:17:::=:f th:num-horizontel--and-vertical-loads-occur-eimult:n::u;1y ;nd en th;;;;; sup

 ) though the time-history results show that these loads do n t-ectiiI"r simultaneous he same support. The peak vertical and h load >ccurs for a vessel n ozzle break. Note that th oads are conser-vative values since the brea n area for essel inlet and outlet aozzle break (as obtained from the dyn , oop analysis) is actually less 41 nhan the estimated upper bound area 4 sed to g te the applied loads. If o dditional analysis was per,fpmeds using the lower b opening area, the

] oads would be considez

abiy reduced. Furthermore, the pe vertical load anc j j eak horizont pas 5'do not occur on the same vessel support. largest vertic paas are produced on the supports beneath and opposite t oken l

ozzh. The largest horizontal loads are produced on the supports whic e' the _ _ rerpadi=L. i. th; irebu ov.d= uv. L .. JTenu dine .3.9.1.4.7 Stress Criteria for Class 1 Components and Component lSupports for BOP Scope of Supply: All Class 1 components and supports have ibeen designed and analyzed for the design, normal, upset, emergency and 53 faulted conditions as specified in the rules and requirements of the ASME B&PV Code, Section III. Stress criteria for Class 1 BOP valves and piping are outlined in Tables 3.9-5 and 3.9-7. Stress limits for Class 1 BOP component l supports are given in Table 3.9-7B.(L 3.9-33 Amendment 53 p, , - . ~ . . . _ . .

ATTACHMEN STP FSAR ST HL AE. /7 PAGE h1OF q 53 Tha Class 1 piping has been designed and analyzed for the design, normal, upset, emergency and faulted conditions in accordance with the requirements of

 )

NB-3600 of the ASME B&PV Code, Section III, 1974 Edition through Winter Addenda of 1975, NB-3658 of Summer Addenda of 1977, NB-3650 and NB-3680 of 41 Summer Addenda of 1979. When the stresses as determined by the methods given in NB-3630 exceed the limits thereof, the design can be accepted provided it mests the requirements of NB-3200. The rules of NB-3630 meet all the requirements of NB-3200. jR,C(,brANck pipe q 3.9.1.4.8 Evaluation of the Control Rod Drive Mechanisms - The Control J R:d Drive Mechanisms (CRDMs) are evaluated for the effects of postulated K breaks. A time-history analysis of 41

_:::: c::::1 n:: 12 lirit:f fie r l-----

tha CRDMs is performed for the vessel motion discussed in Section 3.9.1.4.6.5 A model of the CRDMs is formulated with gaps at the upper CRDM support modeled ca nonlinear elements. The CRDMs are represented by beam elements with lumped Q22.2 masses. The translation and rotation of the vessel head is applied to this q110 ocdel. The resulting loads and stresses are compared to allowables to verify ,g th9 adequacy of the system. The highest loads occur at the head adaptor, the 1ccation where the mechanisms penetrate the vessel head. The combined effect 41 including seismic loads is shown to be less than the allowable loads at this 1ccation.3.9.2 Dynamic Testing'and Analysis 3.9.2.1 Preoperational Vibration and Dynamic Effects Testing on Piping.Piping vibration tests will be performed during the initial test program to cesply with the r2 commendations of Regulatory Guide (RG) 1.68 and satisfy the rcquirements of ASME B&PV Code, Section III. )3.9.2.1.1 Nuclear Steam Supply System Scope: A preoperational piping vibrational and dynamics effects testing program will be conducted for the roactor coolant loop / supports system during startup functional testing of the plant. The purpose of these tests will be to conff.rm that the system has been cdequately designed and supported for vibration as required by Section III of the ASME Code, Paragraph NB-3622.3. The preoperat.ional piping vibration and dynamic effects test program for the primary coolant loop system (this includes the hot legs, cold legs, cross-over legs, reactor coolant pumps, oteam generators, and reactor vessel) at South Texas Units 1 and 2 is as follows: 43 a110

1. The primary coolant loop system as defined ab'ove will be instrumented .27 with accelerometers to measure the dynamic response of the system during c.

normal and transient operating conditions. In addition to normal steady l state operation, the test conditions will include steady state operation lwith various combinations of reactor coolant pumps in operation and tran-l sient conditions due to the starting and tripping of the reactor coolant lpumps.g l3.9-34 Amendment 53 :\ .l W -- - . -

STP FSAR ATTACHMENT ST-HL AE-/74/p PAGE yo OF M q

+ These members supply the radial and torsional constraint of the internals l f, at the bottom relative to the reactor vessel while permitting axial and lb radial. growth between the two. One would expect to see, on the bearing surfaces of the key and keyway, burnishing, buffing, or shadow marks which would indicate pressure loading and relative motion between the two parts. Some scoring of engaging surfaces is also possible and accept-able.

13. Gaps at baffle joints. Check for unacceptable gaps between baffle and l41 top fo mer and at baffle to baffle joints.

Upper Internals

1. Thermocouple conduits, clamps, and couplings.

2. Guide tube, support column, orifice plate, and thermocouple assembly locking devices.

I 3. Upper core plate alignment inserts. Examine for any shadow marks, bur-nishing, buffing, or scoring. Check for locking devices for integrity of'l lockwelds.1 4. Thermocouple conduit clamp welds.41

5. Guide tube enclosure welds and card welds.

Acceptance standards are the same as required in the shop by the. original l

 - design drawings and specifications.

During the hot functional test, the internals will be subjected to a total operating time at greater than normal full-flow conditions (four pumps operating) foy at least 240 hours. This provides a cyclic loading eof approximately 10 cycles on the main structural elements of the inter-nals. In addition, there will be some operating time with only one, two, and three pumps operating.Pre- and post-hot functional inspection results serve to confirm that the internals are well behaved. When no signs of abnormal wear and harmful vibrations are detected and no apparent structural changes take place, the four-loop core support structures are considered to be structurally adequate and sound for operation.3.9.2.5 Dynamic System Analysis of the Reactor Internals Under Faulted Conditions. The following events are considered in the faulted conditions l

 ** *E g h r A M C. b f8f

1. LOCA ( M rr 1^^ ' h^t 1^^ /uptures are considered.)

2. SSE 4'

Maximum stresses for SSE and LOCA are obtained and combined.Maximum stress intensities are compared to allowable stresses for the faulted 41( condition. Elastic analysis is used to obtain the response of the structure, and the stress analysis of each component is performed according to ASME 3.9-41 ' Amendment 41]e, .

ATTACHMENT STP FSAR ST HL-AE l?t/4 PAGE11 OFla q rev faulted conditions, stresses are above yield in C:de-approved tech. ige For thue cases only, some inelastic stress limits are -)a few locations.cpplied.41 The design rules.pf Subsection NG of the ASME B&PV Code, Section III, apply to those reactor internals components identified na core support ntructures.The criteria for acceptability in regard to mechanical integrity Thisanalyses implies are that adequate core cooling and core shutdown must be assured.that the deformation of the reactor internals must be sufficiently small so that the geometry remains substantially intact. Consequently, the limitations established for the. internals are concerned with the deflections and stability of the parts in addition to stress criteria to assure integrity of the ctmponents.For the critical internal structures, maximum allowable deflections, based on The basic opera-functional performance criteria, are listed in Table 3.9-9.tional or functional criterion to be met for the reactor internals is that the plant shall be shut down and cooled in an orderly fashion so that fuel-cladding temperature is kept within specified limits following a Design B: sis Accident.R2 actor Internals Analysis %(. bNCbPht DIE nternals is composed of two parts. The first The evaluation of the react al response of the reactor internals resulting from 41 part is the three-dimensfor 3.9.1.4.6.1. The J the fr--- N; 2_ . I '-- kconditions mentioned in Section reactor internals response is taken from the WECAN RPV and internals system 5 response as described in Section 3.9.1.4.6.4 for the RPV support ff -. of analysis.1., v ~ Q110.1

1. .n . 1._ rnW , di: x;4_i:... an;id:::U:

 .swee, The second part of this evaluation is the core-barrel shell response which consists of the various n = 0, 2, 3, etc. , ring mode response occurring in the horizontal plane. This second part, or ring mode evaluation, is inde-pendent of the loop forces: ' - m , n,......

M Rct bmnck P@e gREMJ Analysisofthereactorinternalsforblowdownloadsresultingfrom/e.J.eEAis based on the time-history response of the internals to simultaneously applied ,blowdown forcing functions. The forcing functions are defined at points in the system where changes in cross-section of direction of flow occur in such Thea way that differential loads are generated during the blowdown transient.dynamic mechanical analysis can employ the displacement method, lumped parameters, and stiffness matrix formulations, and assumes that all components behave in a linearly clastic manner.In addition, because of the complexity of the system and the components, it is necessary to use finite element stress analysis codes to provide more detailed information at various points.A blowdown digital computer program (Reference 3.9-7), which was developed for l 41 the purpose of calculating local fluid pressure, flow, and density transients that occur in pressurized water reactor coolant systems during a LOCA, is b applied to the subcooled, transition, and saturated two-phase blowdown regimes. This in contrast to programs such as WHAM (Reference 3.9-8) which g are applicable only to the subcooled region and which, due to their method of Amendment 41 3.9-42

STP FSAR ATTACHMENT -ST HL AEOF PAGE*/JL 11%/$ ,41 solution, could not be extended into the region in which large chang w u une sonic velocities and fluid densities take place. Multiflex is based on the f- method of characteristics wherein'the resulting set of ordinary differential i equations, obtained from the laws of conservation of mass, momentum, and energy, are solved numerically, using a fixed mesh in both space and time.l con ervation laws are employed, the code Although spatially.one-dimensio j sional system geometries by use of the can be. applied to describe threeSuch' piping networks may contain any number of equivalent piping networks.pipes or channels of various diameters, dead ends, branches (with up to six pipes connected to each branch), contractions, expansions, orifices, pumps, Sys and free surfaces (such as in the pressurizer).tion, contraction, and expansion, as well as some effects of the water / solid interaction, are considered.The blowdown code evaluates the pressure and velocity Each reactortransients forfor component a maximum which of 2,400 locations throughout the system.calculations are required is designated ar, an element and assigned an element number.Forces acting upon each of the elements are calculated, sunning the effects of:

1. The pressure differential across the element.

2. Flow stagnation on, and unrecovered orifice losses across, the element.

3. Friction losses along the element.

Input to the calculation code, in addition to the blowdown pressure and fvelocity transients, includes the effective area of each element on which the force acts due to the pressure differential across the element, a coefficient to account for flow stagnation and unrecovered orifice losses, and the total area of the element along which the shear forces act.

 ,41 The reactor internals analysis has been performed using the following assumptions:

e The analysis considers' the effect of hydroelasticity.e The reactor internals are represented by a multimass system connected with springs and dashpots simulating the elastic response and the viscous lThe modeling is conducted in such a way that i damping of the components. uniform masses are lumped into easily identifiab elastic elements are represented by springs.

The model described is considered to have a sufficient number of degrees

+of freedom to represent the most important modes of vibration in the jvertical direction. This model is conservative in the sense that further mass-spring resolution of the system would lead to further attenuation of f'the shock effects obtained with the present model.The pressure waves generated within the reactor are highly the In general, dependent more on the g.o location and nature of the postulated pipe failure.rapid the severance of the pipe, the more severe the imposed loadings on the( components. A 1-millisecond time is taken as the limiting case.Amendment 41 3.9-43

 -... ... - . ~ -- - . . .. . . ..... , .. .-...--n...

vv

 ,n.,.. , - .,.- . - . - .,, _ -----

ATTACHMENT STP- PSAn ST-HL W AEgFlM PAGE 73 br%NCg fife In the case o d an the hot leg break, the vertical hydraulic forces pro uceA' rarefaction w -)initial upward lift of the core. l Since the racetor hot leg nozzle into the interior of the upper f the barrel, core barre the. upper w;va has not reached the flow annulus on the outside T ohus, dynamic insta-b rral is subjected to an impulsive compressive In addition to wave.bility the (huckli possible responses of the barrel during hot leg blowdown.cb;v l hot leg nozzle.erra components as the fluid exits tie I

 .g -- BRANCH P PE l a reac-inthecaseoft]hecoldlegabreak,ararefactionwavepropagatesaonginlet nozzle of the(""

tsr inlet pipe, arriving first at the core barrel at theThe bankam loop .'h upper barrel is t cien radial impulse which changes as the raref action wave propagates l bot crcund Aftsr the coldthe legbarrel break, andthe down the outerhydraulic initial steady-state flow annulus lift forcesbetweenh vess (upward) decrease Theserapidly (within cause the reactoracore fewand milliseconds) and then increase in lower support structure downward direction.to move initially downward. d If a simultaneous seismic event with the intensityfor of each the SSE case isis postulate with the LOCA, the combined effect of the maximum stresses c:nsidered. In general, the loading imposed by the earthquake is small co pared to'the blowdown loading.is discussed in Section 3.7.3. h .

 )

A summary of the mechanical analysis is presented in the following l paragrap l41 s.R2ference 3.9-9 provides the basic methodology used in the reactor interna s blowdown analysis.V2rtical Excitation Model for Blowdown For the vertical excitation, the reactor internals are represented h lastic by a multimass system connected with springs and dashspots simulating Also incorporated in tthe ee 'rssponse and the viscous damping of the components. hich includes the 41 multimass system is a representation of the fuel assembly, win order to ob fuel assembly grids.internals response, the effects of internal damping, clearances l d in between l41 various internals, snubbing action caused by solid impact, and pre The oa s hold-down springs have been incorporated in the analyticaldmodel. into easily lmodeling is conducted in such a way that uniform masses a eprings.The apprcpriate dynamic differential equations for the multimass model describing the aforementioned phenomena are formulated and the results cbtained using a digital computer program (Reference 3.9-1) i which compu response f orcing functions.of the multimass model when excited by the time-i ously and independently to each of the masses in the system. functions of the program give the forces, oisplacements, and deflections )as Reactor inter- dgr time for all the reactor internals components (lumped masses . d nals response to both hot and cold legVpipe ruptures were analyze .br*4AXdi k

 ~

(' Amendment 41't, 3.9-44 1 ..._......___.-.._

ATTACHMENT l

+ STP FSAR ST HL-AE- 30W l' PAGE ft/OFl3M Transverse Excitation Model for Blowdown

(;'* - Various reactor internal components are subjected to transverse excitation during blowdown. The barrel, guide tubes, and upper support columns are ana-lyzed to determine their response to this excitation.

1. Cole Barrel 1 L .

 . btANC n Pif E. -- q For the hydraulic analysis of the pressure transients during hot legh blowdown, the maximum pressure drop across the barrel is a uniform radial compressive impulse.

The barrel is then analyzed for dynamic buckling using the following conservative assumptions: l

a. The effect of the fluid environment is neglected. ,

 )

i'

b. The shehl is tireated as simply supported, u bftP>C,h f f E.

1 During, cold leg blowdown, the upper barrel is subjected to a nonaxisym-metric expansion radial impulse which changes as the rarefaction wave

propagates both around the barrel and down the outer flow annulus between vessel and barrel.

a,, benNLA .PtPL The analysis of transverse barrel response to,, cold leg blowdown is per-formed as follows:

a. The core barrel is analyzed as a shell with two variable

 ' sections to model the core barrel flange and core barrel.

b. The barrel with the core and neutron shielding pads is analyzed l41 as a beam elastica 11y supported at the top and at the lower radial support, and the dynamic response is obtained.

2. Guide Tub _es hr@A.)Ch fIfC-The dynamic loads on ro dclustercontrolguidetubesaremoreseverefb ra LOCA caused by hot leg gupture than for an accident caused by cold leggrup-ture, since the cold leg break leads to much.ssaller changes in the transverse 4g coolant flow over the rod cluster control assembly guides. The guide tubes in closest proximity to the .y ..._2 outlet nozzle'/are wn the most severely loaded.

decrease with increasing The transverse guide tube forces during a blowd distance from the ruptured nozzle location. " FOR A H e7' 4 E G B R

• MC H P PE. BCEM.

A detailed structural analysis of the rod cluster control guide tubes is per-formed to establish the equivalent cross-section properties and elastic and' support conditions. An analytical model is verified by subjecting the control rod cluster guide tube to a concentrated force applied at the midpoint of the lower guide tube. In addition, the analytical model has been previously veri-fied through numerous dynamic and static tests performed on the 17 x 17 guide tube design.I

 ~

The response of the guide tubes to the transient loading from blowdown 41

 #' resulting from hot legVbreaks is found by representing the guide tube as an

! L ence.A b p,'pe.3.9-45 - . Amendment 41 g - .. _7 -....._--- - - - _ _ . --

ATTACHMENT STP FSAR ST HL-AE-PAGEr75 OFJP17t/4 )cquivalent single-degree-of-freedom system and assuming tne slope of the time-dependent load to be a step function with constant slope front end. )3 g er, g q q q}%gy, .pgg gmb Upper support c31umns located close to the bsukuunonozzi hot leg boonk will be cubjected to transverse loads due to cross flow. The loads applied to the c::1umns are computed with a method similar to the one used for the guide tubes, i.e. by taking into consideration the increase in flow across the col-umn during the accident. The columns are studied as beams with variable ccetions and the resulting stresses are obtained using the reduced section modulus and appropriate stress risers for the various sections.

i. {sg ts p{,; g g,oy,,In yyna1_s Ana p /s Maximum stresses due to the SSE (vertical and horizontal components) and a LOCA were obtained and combined. All core support structure components were fcund to be within acceptable stress and deflection limits for both hot leg cnd cold legvLOCAs occurring simultaneously with the SSE; the stresses and d2flectionsirhich would result following a faulted condition are less than those which <rould adversely affect the integrity of the core support struc-tures. For :he transverse excitation, it is shown that the barrel does not buckle durin; a hot legVbreak and that it meets the allowable stress limits during all s 3ecified tra'Estents. ' '

 - bm9eh P'PC Also, the natural and applied frequencies are such that resonance problems will not occur. I PM . )

icate that the relative displace-The results obtained fr linear analyses ment between the cor nents will close the gaps, nd consequently the struc-tures will 1 6 ..m on each other. Linear analysis 1 not provide informa-tion about the impact forces generated when components on each other; however, in some instances, linear approximations can and are applied prior to cnd after gap closure. The effects of the gaps that could exist between ves-cel and barrel, between fuel assemblies, and between fuel assemblies and baffle plates, were considered in the analysis using both linear approxi- l41 cations and nonlinear techniques. Both static and dynamic stress intensities are within acceptable limits.Even though control rod insertion is not required for plant shutdown, this cnalysis shows that most of the guide tubes will deform within the limits catablished to assure control rod insertion. For the guide tubes deflected l41 above the no-loss-of-function limit, it must be assumed that the rods will not drop. However, the core will still shut down due to the negative reactivity insertion in the form of core voiding. Shutdown will be aided by the great majority of rods that do drop. Seismic deflections of the guide tubes are generally negligible by comparison with the no-loss-of-function limit.3.9.2.6 Correlatiors of Reactor Internals Vibration Tests With the Analytical Results. AsstatedinSection3.9.2.4,itisnotconsideredneces-l41 cary to conduct instrumented tests of the STP RPV internals, as their adequacy #has been verified by use of the Sequoyah and Trojan results as well as by the )^1/7 scale model test results. References 3.9-5 and 3.9-10 describe predicted vibration behavior based on studies performed prior to the plant tests. These m,i 3.9-46 Amendment 41

7. ,. . - - -_ -_ . _ . , _ - - - - _ - .

STP FSAR ATTACHMENT ST44L PAGE ?GOFAE-taf /794 )

3. Spring preloads k ' 4. Coolsut flow forces (static) (

5. Temperature gradients

6. Differences in thermal expansion _,,

a. Due to temperature differences

b. Due to expansion of different materials

7. Interference between components

8. Vibration (mechanically or hydraulically induced)

9. All operational transients listed in Table 5.2-1

10. Pump overspeed

11. -Seismic loads [0BE and SSEl l 4g

12. Blowdownforces(duetocoldandhotleg[bresh)

The main objective of the analysis is to satisfy allowable stress limits, given in NB-3200 and NA Appendix F, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the func-( tioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also to limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials. Standard methods of determining strength of materials are used to establish the stresses and deflections of these components. The dynamic behavior of the reactivity control components has been studied using experimental test data and experience from operating reactors.3.9.4.3.2 Drive Rod Assembly: All postulated failures of the drive rod p0 assemblies either by fracture or uncoupling lead to a reduction in reactivity.If the drive rod assembly fractures at any elevation, that portion remaining coupled falls with, and is guided by, the RCCA. This always results in a reactivity decrease for control rods. l 30 3.9.4.3.3 Latch Assembly and Coil Stack Assembly: With respect to the CRDM system as a whole, critical clearances are present in the following areas:l

1. Latch assembly - diametral clearances l 2. Latch arm - drive rod clearances

3. Coil stack assembly - thermal clearances l( 4. Coil fit in coil housing i

i N3.9-69 Amendment 41 is i1

y. -,--,-m,...

 ,,,,,m.---w*y.,w..,,9 y w,. qm----w---.-y -g+g,.we.-,

ATTACHME ST-HL-AE 17 STP FSAR PAGE77OF '1 3.9.5.2 Design Loading Conditions. The design loading conditions that ,provide the basis for the design of the reactor internals are:

1. Fuel assembly weight

2. Fuel assembly spring forces

3. Internals weight

4. Control rod trip (equivalent static load)

5. Differential pressure

6. Spring preloads

7. Coolant flow forces (static)

8. Temperature gradients

9. Differences in thermal expansion

a. Due to temperature differences

b. Due to expansion of different materials

10. Interference between components

11. Vibration (mechanically or hydraulically induced)

12. One or more loops out of service

13. All operational transients listed in Table 3.9-8. ]2.

14. Pump overspeed

15. Seismic loads (OBE and SSE) , l 41 bhWd,k jolf 6

16. Blowdown forces (due to cold and hot leg break) g i The main objective of the design analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the functioning of the components.

The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also to limit the amplitude of the oscil-latory stress component in consideration of fatigue characteristics of the caterials. Both low- and high-cycle fatigue stresses are considered when the allowable amplitude of oscillation is established. Dynamic analysis on the

reactor internals has been provided in Section 3.9.2.

As part of the evaluation of design loading conditions, extensive testing and. inspectir.n have been performed from the initial selection of raw materials up hto nnd including component installation anu plant operation. Among these tests and inspections are those performed during component fabrication, plant construction, startup and checkout, and plant operation. .3.9-76

• Amendment 41

STP FSAR ATTACHMEN ST.HL AE r/PAGE1( OF q 3.9.5.2.1 Normal and Upset: The normal and upset loading conditions i that provide the basis for the design of the reactor internals are:(- Fuel and Reactor Internals weight.a) b) Fuel and Core Component Spring Forces including spring preloading forces.c) Differential Pressure and Coolant Flow Forces.d) Temperature Gradients.

a) Vibratory Loads including OBE seismic.

f) The normal and upset operational thermal transients listed in 3.9.1.1.6 and 3.9.1.1.7. Q110.5 g) Control Rod Trip (equivalent static load).h) Loads due to Loop (s) Out-of-Service.! i) Loss of Load / Pump Overspeed.3.9.5.2.2 Emergency Conditions: The emergency loading conditions that provide the basis for the design of the reactor internals are:a) Small Loss of Coolant Accident.b) Small Steam Line Break.c) Complete Loss of Flow.3.9.5.2.3 Faulted Conditions: The following faulted loading conditions are considered the most limiting and provide the basis for the design of the reactor internals are:kfuobtf Ob AW RCb a) The i ci,; i;;; c' C::M: i::if::: Qfgg .b) The Safe Shutdown Earthquake.3.9.5.3 Design Loadina Categories. The combination of design loadings fits into either the normal, upset, or faulted condition as defined in the' ASME B&PV Code, Section III. The allowable stress limits indicated in 4 Subsections NG-3222 (Normal Conditions), NG-3223 (Upset Conditions) NC-3224 2(Emergency Conditions) and Appendix F (Rules for Evaluating Faulted Condi-Q110 tions) are met.1 Internal Structures are analyzed to meet the intent of the ASME Code in 5l' accordance with Subsection NG, parsgraph NG.3311 (c). Stresses in the Core l^Support Structure induced by interaction with internal structures are analysed and shown to be in conformance with Core Support Code Limits. Design and Jconstruction for Core Support Structures meet Subsection NG in full. l3 1i l l .3.9-77 Amendment 41 I4

f~r ATTACHMENT <m-ST.HL AE /79t/STP FSAR PAGE 19 OF tg q REFERENCES JSection 3.9:3.9-1 M i , -~ . .. L'.), " Documentation of Selected Vestinghouse 5 Structural omputer Codes," WCAP-8252, Revision 1 (May 1977) A q 63

 . 3.9-2 " Sample Analysis of a Class 1 Euclear Piping System,"

prepared by ASME Working Group on Piping, ASME Publication, 1972.3.9-3 Bamford, W. H., and C. B. Buchalet, " Methods for Fracture Mechanics Analysis of Nuclear Reactor Vessels Under Severe Thermal Transients," WCAP-8510 (June 1976).3.9-4 " Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests" WCAP-8317-A (March 1974).3.9-5 "UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Functional Examinations," WCAP-851/ (March

 - 1975).

3.9-6 Trojan Final Safety Analysis Report, Appendix A-12, Docket No. 50-344.

 ) et al., "Multiflex - A Fortran-IV Computer 3.9-7 Takeuchi, K.,

Program for- Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708 and WCAP-8709 5 (non-proprietary) (February 1976).3.9-8 Fabic S., " Computer Program WHAM for Calculation of Pressure Velocity and Force Transient in Liquid Filled Piping Networks," Daiser Engineers Report No., 67-49-R l (November 1967).3.9-9 Bohm, G. J., and J. P. LaFaille, " Reactor Internals Response Under a Blowdown Accident," First International Confer-ence on Structural Mechanics in Reactor Technology, Berlin (September 20-24, 1971).O 3.9-10 B yd, C. N., W. Ciaramitaro, and N. R. Singleton, "Verifi-cation of Neutron Pad and 17 x 17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant,"l l WCAP-8780 (May 1976).3.9-11 Cooper, F. W., Jr., "17 x 17 Drive Line Components Tests -Phase 1B 11, 111, D-Loop-Drop and Deflection," WCAP-8446, Westinghouse Proprietary Class 2 and WCAP-8449 l-Westinghouse Nonproprietary (December 1974).

 ' " Neutron Shielding Pads," WCAP-7870 (May 1972).

3.9-12 Kraus, S.,

 .s 3.9-82 Amendment 41

STP FSAR ATTACHMENT GE khgr}TABLE 3.9-2.3 (Cont'd.)('DESIGN LOADING COMBINATIONS FOR ASME III CODE CLASS 1 COMPONENTS (BOP SCOPE OF SUPPLY)DEFINITION OF TERMS ,PD - Loadings associated with the design pressure. i iP0r - Loadings associated with operating pressures including, where applicable, any transient pressures associated with the loading ,conditions event under consideration.

 ~

DW - Loading associated with deadweight and liveweight.OBE - Inertial loadings associated with the OBE.l SAM (OBE) - Anchor point displacement loading associated with OBE earthquake. l BS - Single nonrepeated anchor movement (building settlement). i SSE - Inertial loading associated with the SSE. l lSAM (SSE) - Anchor point displaccrent loading associated with SSE. I41 i

 , RVC - Transient loadings associated with relief valve blowdown in a {

[, closed system. .RVO - Sustained loadings associated with relief valve in an open system.1 FV - Transient loadings associated with fast valve closure.TH - Loadings associated with thermal expansion. {LOCA - Loss of coolant accident - defined M -- " tasAme#=40GN= move 90 'a's M hose postulated accidents that result from the loss of reactor coolant, at a rate in excess of the capability of the 'reactor coolant make-up system, from breaks in the reactor coolant pressure boundary up to and including a break equivalent in size to the d ~t h d i rupture of the largestVpipe +6vthe reactor coolant W hmoch L c.euutt.Tt.h TD This condition includes the loads from the postulated pipe break itself and also any associated system transients or dynamic effects resulting from the postulated pipe break.HEB - Loadings associated with high-energy line pipe breaks (includes loadings from jet impingement, pipe motion, and pipe impact).DU - Loadings associated with other transient dynamic event classified as an upset condition.(

 ~

3.9-98 Amendment 41

ST HL AE-I7yt ATTACHMENT ]STP FSAR PAGE 3 i OF lar7 TABLE 3.9-8 (Continued)

SUMMARY

OF REACTOR COOLANT SYSTEM DESIGN TRANSIENTS Faulted Conditions

• pgr.h Occurences ,

1. Meen. reactor coolant pipe break 1

 .(bouwd LOCA) 1.arge steam break 1 i 2.

3. Feedwater line break -

1 1

4. Reactor coolant pump locked rotor 1 1

5. Control rod ejection

6. Steam generator tube rupture (included under upset condi-tions, reactor ,

trip from full power with safety injec-tion)Safe Shutdown Earthquake 1 7.Test Conditions ,

1. Primary side hydrostatic test 10

2. Secondary side hydrostatic test 10

 *In accordance with ASME B&PV Code Section 111, faulted conditions are not included in the fatigue evaluation.

6 yI 3.9-122 Amendaent 41

ATTACHMENT STP FSAR ST HL AE n44 PAGE 3SOF;O TABLE 3.9-12 NOEELE ELE.VATID'J MAXIMUM REACTOR VESSEL DISFLACEMENTS AT REAGW4-4EHHMM , CENTERLINE Maximum Horizontal Maximum Vertical Maximum 5

 ' Displacement Displacement Rotation (inches) (inches) (radians)

RG iulu ., m 1. . (Later) (Later) (Later)

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i 3.9-131 Amendment 41

r' i[4 Q~.A .

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i a STEAM OENERATOR ,,,g j RESTRA1RT p STE.8W LTRE lf eO 'i STEAM GENEitATOR O OF'PER SUPPORT REA/ TOR (QS a REACTORPOMP jPRESSURE N l VESSEL y COOLANT REACTOM sCOOLANT j g T

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T4Mt 3.121 (Cont {rwed, 1at9UL A10e' GMOE JERTRit etCULh10RT tutDE_ Mvis".ourj att's 3T4Ri3 fTs $1P ,1so. TITit Jtas ef3ntste _1 l 36 'Preoperationet Testing of tr&rdant On-Site 14.2 Rev 0 (1/G ) A 1.41 Electric teaser Systtes to verity A eper Load Crew Assiruncets 1.42 WitMrewn Control of Steintes Steel Cteddir.g of 5.2.3.3,7 Rev 0 ($f73) e 1.43 low sttoy Steet teapsewnts 5.3.1.4 1.44 Cetrol of tu Use of Sensittred stainless 3.5.5.1 Rev 0 (5/13) C 3ee wete 47 l33 4.5.2.4 8 Steet 5.7.3.4 5.3.1.4 1D.3.6.2 m

• N y 1.45 Reactor Ccotant Pressure Sourdery Leekage 5.2.5 6ev 0 (5/73') A ,

M Protection Systeam 11.5.2.9.8 Oe Tabte 7.1 1~No#E witwoCAtDs4 b 1,46 Protection Against Pipe WMp Inside Jeble-4:3-P r; O;'/ ,; / See Note 63 Contair e t .M-

 -M*t:f A&s4M 09 g(n >

7.1.2.6 Rev 0 (5/73) A gQ430.109H g 1.47 pypessed and ineywreble Status Indication for sucleer Power Plant safety Systems febte 7.1-1 l43 m[N '>1.5.4

 *i>$ 8.3.1.2.4 l36 N2 J

8.3.2.2.7

$ h y Figure 7.1 1 ,

4 a Rev 0 (5/73) C See Note 30

" 1,48 Design Limits ord Leading Cambinations 10.4.8.1.5 y for seismic Category i Fluid System Camponents Tabte 3.7-1 febte 3.9 2.5

 ~

ATTACHMENT n f I hFFSAR ST.HL-AE-i 7{1*f '7J PAGE P.50F -TABLE 3.12-1 (Cont'd.) J RECUIATORY GUIDE MATRIK le0TES $,i If a work activity and contract is for a two-conth period or less, an -43 audit is not necessary when a facility preaward audit has been conducted.The QA program for operations will conform to the requirements of RG 1.94.Revision 1, with the .same clarification: 5 ,55.* Refer to Sections 3.7.4,1 and 3.7.4.2 for the diccussion on iseistsic in-strumentation.

56. Refer to Section 5.2.*3.3.2 for Westingh,e,use alternate approse.h to 'RG 1.71. Also, refer to Section 10.3.6.2, for the BOP conformance to RG.

1.71.

57. STP alternate approach to RG 1.99 is discussed in Section 5.3.2.1.

58. STP alternate approach to RG 1.121 is discussed in Section 3.12.1. 33

59. Revision 0 is utilized during the construction phase for RS 1.58, Posi-tions C.5, C.6, C.7, C.8, and C.10 of Rev.1 are also utilized.

60. With respect to Section 3.1.2 of ANSI N45.2.3-1973, HIAP interprets the lighting level of 100 footeandles to be guidance. It is M1&P's normal .

practice that the lighting level for determining " metal clean" cf acces- .sible surfaces of piping and compenents is determined by the inspector. 'Typically he uses a standard two-cell flashlight supplemented by other '( lighting as he deems necessary.

61. See the response to Question 321.4 for the compliance with the Regulatory _

 #3 Guide. .

62. RG 1.1 as clarified by NUREC-75/007. pott.oW W 6 liSOANM NNJ 3a~

has be.cn ta NdretM

63. The -be.n L _ _ il,,3 2 . i. 1 ^ d AG 1.46i1; "- ' ;1- r _

: of NRC Branch Technical Position (BTP) MF.B 3-11 MRC BTP ASB 3-1[ N 49 ,'

anNM. Tables 3.6.1-2 and 3.6 1-3 provide a sunenary of the compliance with MEB 3-1 and ASB 3-1. ,, p ;r 64 The discussion of STP conformance to RG 1.97 Rev. 2 is presented in Table . i 7.5-1 and Appendix 78. As explained in Appendix 78, implementation of RC :1.97 requirements was integrated with the control room design review and 33 'was performed using the Westinghouse Owner's Group Emergency Response Cuidelines, and conforms with the intent of the RG. ,9

65. The QA program during operations will conform to the requirements of 43 .

Revision 2. l

 ?

. 66. The quality of DG fuel oil will be checked as identified in'Section

• 9.5.4.4. .9;

s I( !N Amendment 54 :3.12-32 tI t

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ATTACHMENT /STP 'SAR ST-HL-AE i#/s PAGE 760F(Q

 , LIST 0F FIGURES'(Contitmed)

Chapter 5 Figure Beforen:e liusber Title Number 5.4 10 Piepsur.ize.r 5.4-11 -Pressurizer Felief Tank 5.4 12 Reaator Vessel Supports 5.4-13 Stea.a Generator Supports ,5.4 14 Reactor Coolant Pump Support 5.4-15 Pressurizer Supports plarfD 5.4-16 7 4ee:: --- - __, "or >p E 'f.M.D 5.4 17 heeas .%:. . , bH.Ac C -Run-kertra*.Tirr p g.t.E T. C a 5.4-18 , Mot-4.*g-Pipe-Whip-Restraint

 . 5.4-19 %!.ET& h --A r'-1-!-Leg-friwsry-Stith -.a rip '?4 A.,E_ _,_._

5.4-20 Residual Mcat Removal Nap Performance Curves '/,9 ;I al l1 ds 1

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vi Amendrsent 49

ATTACHMENT 3;p pgg S1-HL-AC PAGE F70F/7yt[ry ra the spring-loaded safety vsCves, Ncte that eptpoint studies to date indicate that the pressure rise in.* four-loo,s plar.t fcr the design step 1 cad decrease of 10 percent from full pcmr is 11 nit.ed to 60 psi. The pressure rise is not l38 sufficient to actuate the PC Ms, and thus this design is conservative.138 The The pressurieer spray walves help te prevut actuation of the PORVs.spray rate is selected to prawnt the pressurizer pressure from reaching the operating setpoint cf the PCRVs follevity a 5ttp load reduction in power of ten percer.t of full Icw! vith te. ecto cont.roI 5.'4.13.4 Tests ani,7ps_pytiocr., A13 unfety and 'rclief valves are subjected to hydrestatie th$ts, otst Wkagt rteSts, operathnel te,sts, and inspections as required. For saf ety n.c reliefi vaives that are required to functien during a f aulted conatt%n, JCditional tests sie performed. These tests are descri6ed in Section 3.9.5.4.14 compocent Supp. orts g#h-5.4.16.1 Drafen 2 peu. Compe,nent surrorts allow v W tu 12y unrestrained lateral t. tert 3a1 sevraant of the loop during plant operation a d provide res-traint to ithe loops and .-omponente during m.ident wn61t10ns.3 The loading combinatioss med d.esign were.ss litLts are eisiussed $ n 5,vetsons .3.9,1.1 and 3,9.1.4.7. Gupport design .ounay4-hmgi; =--M (s in t. cordarce with the ASliE Code, Section .ITI, Subte.: tion NF., The design iLaintrins the integrity of the KCS bot.nda y for normal .std accident cen(itans, Raoulta of support. 30 striss evaluat. ion.s arti pretected in Section 3,9..i.4 4.

5. 4. I4.fr M riptyon. The support .strucc%ren are welded structural creal sectiono Mnear type ntractuees (tansf on and comprucadon struts, col- l38 tens, .tui hetas) . ara usec' creept for t.it reactor vessel auppef t.s, which are p). ate-t y;e e.tfucturt s,. A.tt.scimento to :he supported equiputre kre non-inte-pal type that are. . bolted tr.' c7 be ar apinst the component s. Tbe sur-

• p.cte+to-concrete .attschmento are wither es6edded ,tachor boli.e )r fabricated ensemblica.

The supports permit vit tually unrestrain6d thersial g.cowth of:.he supported! s/ stems bs.t rentet.in Pettied, layersI, and rotationql tovement ret.ulting from seiscic ar d pipe busir Iruditygs. Thia 18 atstCU;p11Bhed using spherical buchings ir.4 t.he coturnu fitt verticel support and 31rders, buu.per pedestals, byd.tanlic spuibers, and tie-rods tor leteral etpt,orr..i Because of manufacturing and t;onstrc :ticn tolerone.es, amp 1Le adjustment in tbs support structures owt W pr0r.'de.d ts entate pmper erectiott alignaent and! fit-ur.This is acecaplished by shinning or Srci. ting at; the supports-to-cuncreat interface axt by chiming at ,the .uppost;-to-*. quips.u t interface.i Fere_t,ogJr_essyr,e Vescel Suppttes for the resttor vessel (Figura %4-13) are individesi air-cooled see.tatp1At box structures benenth the va.iAn norti ns bolted% the prinary' eide14 us11 conerrte. Lech box struJ1urt ,.on81sts! a hnrizontal top plate that Yerciv.rs loado frrnt the reactor sessel shte, a hurizontal botto.n place sup. sorted by and t.ransf erring loads to the yrit,/.ty chield wall r.or: crete, and l! coomecting ve rtital pistes. The supports are vir-coo 14d to maintair, the! .euppurtitt, cf.n.et ete t.eepe,ature within r.ccept.tble levelt.t i3, b4; Armendent 38 L .

ATTACHMENT -

 .- ST HL AE I744 PAGE 18 OF in INSERT Page 5.4-42 ,

As discussed in Reference 3.6-14 and paragraph 3.6.2.1.1.la, RCL ruptures and the associated dynamic effects are not included in the design bases.7603N:0288N/16 a

ATTACHMENT ST HL AE /7t/t/STP FSAR PAGE f 9 OF t;to ,/Steam Generator

 -

• As shown on Figure 5.4-13, the SG supports consist of the following elements: h8

1. Vertical Support Four individual columns provide vertical support for each SG. These are bolted at the top co the SG and at the bottom to the concrete structure. 3g Spheric.1 ball bushings at the top and bottom of each column allow unre-strained lateral movement of the SG during heatup and cooldown. The column base design permits both horizontal and vertical adjustment of the CG for erection and adjustment of the system.

2. Lower Lateral Support Lateral support is provided at the generator tubesheet by fabricated steel girders and struts. These are bolted to the compartment walls and include bumpers that bear against the SG but permit unrestrained movement cf the SG during changes in system temperature. Stresses in the beam caused by wall displacements during con:partment pressurization and the building seismic evaluation are considered in the design.

3. Upper Lateral Support The upper lateral support of the SG is provided by a built-up ring plate girder at the operating deck. Two-way acting snubbers restrain sudden seismic

(or blowdown induced motion, but permit the normal thermal movement of the SG.Movement perpendicular to the thermal growth direction of the SG is prevented by struts.Reactor Coolant Pump Three individual columns, similar to those used for the SG, provide the ver-tical support for each pump. Lateral support for seismic and blowdown loading is provided by tension tie bars and compression struts. The pump supports are l 38 shown on Figure 5.4-14.Pressurizer The supports for the pressurizer, as shown in Figure 5.4-15, consist of:

 ). A steel ring plate between the pressurizer skirt and the supporting con-crate slab. The ring serves as a leveling and adjusting member for the pressurizar and may also be used as a template for positioning the con-crete anchor bolts.

2. The upper lateral support consists of struts cantilevered off the com-partment walls that bear against the " seismic lugs" provided on the pres-surizer.

ipe Restraint N -5.4-43 Amendment 38 t

STP FSAR ATTACHMENT ST HL AE n</q PAGE 900Fla q

 . Crossover Leg Restra(nt at each elbow of the reactor coolant pipe between the pump and the SG Krossover leg) is required to prevent excessive stresses on the system re hlting from postulated breaks in this pipe. The' support in- -

cludes pipe'tumpers Figure 5.4-16)xand steelwith straps members.supporting and steel thrust blocks -(as shown onAwhiprestrai provided (as shown on Figure 5.4-17) to prevent whipping of the crossover, les pipe following% postulated break at the SG /utlet nozzle. This /restraint attaches to the primary shield wa114nd extends horizontally to thai vertical run of th crossover N leg pip e/ / /l

2. Hot Leg A restraint is located at the 0*- elbow in the hot leg to prevent exces-

 ! sive displacement of the hot , leg following a postulated guillotine break at the SG inlet nozzle. This restraint consists of structural steel meb-l bars which transmit load 2'to the concrete structure, as shown on Figure ! 5.4-18. '/ N f Hot Leg and Cold Le \s j j q - Pipe restraints are provided in the primary shield wall at the reactor l c,oolant pipe as close to the elbow on the cold leg as possible, and on l

the hot' leg as close to the RPV outlet nozzle safe end as possible p (Figure 5.4-19). The function of these restraints is to limit the breat f area for guillotine breaks at the RPV safe end to approximately one i square foot, such that peak cavity pressure is reduced. Insulation or l30 cooling is provided such that the concrete temperatures are maintained-- ,*ithin acceptable-11mits. - - - - - - - - - - - - -5.4.14.3 Evaluation. A detailed evaluation ensures the design adequacy and structural integrity of the reactor coolant loop and the primary equipment I supports system. This detailed evaluation is made by comparing the analytical results with established criteris for acceptability. Structural analyses are performed to demonstrate design adequacy for safety and reliability of the plant in case of a large or small seismic disturbance and/or LOCA conditions.Loads which the system is expected to encounter often during its lifetime (thermal, weight, pressure) are applied and stresses are compared to allowable values as described in Section 3.9.1.4.7.The SSE and design basis LOCA resulting in a rapid depressurization of the system are required design conditions for public health and safety. The methods used for the analysis of the SSE and LOCA conditions are given in Section 3.9.1.4.The reactor vessel supports are not designed to provide restraint for vertical uplift movement resulting from seismic and pipe break loadings. However, PPV motion resulting from seismic and pipe break events is conservatively included in the RCS analyses described in Section 3.9.1.4 Thermal analyses are perforned for the RPV supports. Thermal growth of tk supports are included in the RCS analyses as thereni anchor movement. g 5.4-44 Amendment 38 s

 ^ ^ 's/31/713s.2 ST HL /?'/

PAGE 4i OF l M I

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T R.C. PUMP CROSS OVER PIPE arQ 'BUMPER \

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STEEL BUMPkR STEEL BUMPER-BLOCK SUPPORT BLOCK SUPPORT

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SOUTH kEXAS PROJECT UNITS 1 & 2 Crossover Leg Support Figure 6.4-18 Amendenent 38

ATTACHMENT ST.HL AE 1741 PAGEj$0F Ig7 h 0

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SOUTH TEXAS PROJECT i UNITS 1 & 2 l Crossover Leg Vertical Run Restraints Figure 5.4-17 Amendment 38.

ATTACHMENT

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ATTACHMENT i STP FSAR ST-HL-AE- nV( l PAGE 450F W/I Question 022.02 In the unlikely event of a pipe rupture inside a major component subcompart-ment, the initial blowdown transient would lead to pressure loadings on both the structure and the enclossd component (s). To assure the inte5rity of these design features, we request that you perform a subcompartment, multi-node pressure response analysis, and provide the following information:i (1) . Provide and justify the pipe break type, area and location for each !analysis. Specify whether the pipe break was postulated for the evalua-tion of the compartment structural design, component supports design or i both. .l (2) For each compartment, provide a table of blowdown mass flow rate and energy release rate as a function of time for the break which results in the maximum structural load, and for the break which was used for the component supports evaluation.(3) Provide a schematic drawing showing the compartment nodalization for the determination of maximum structural loads, and for the component supports evaluation. Provide sufficiently detailed plan and section drawings for several views, including principal dimensions, showing the arrangement of the compartment structure, major components, piping, and other major obstructions and vent areas to permit verification of the subcompartment nodalization and vent locations.l' (4) Describe the nodalization sensitivity study performed to determine the minimum number of volume nodes required to conservatively predict the maximum pressure load acting on the compartment structure. The nodal-ization sensitivity study should include consideration of spatial pres-sure variation circumferential1y, axially and radially within the com-partment. Describe and justify the nodalization sensitivity study performed for the major component supports evaluation, where transient forces and moments acting on the components are of concern.(5) Discuss the manner in which movable obstructions to vent flow (such as insulation, ducting, plugs, and seals) were treated. Provide analytical and experimental justification that vent areas will not be partially or completely plugged by displaced objects. Discuss how insulation for piping and components was considered in determining volumes and vent areas.(6) Graphically show the pressure (psia) and differential pressure (psi) responses as functions of time for each node. Discuss the basis for establishing the differential pressure on structures and components.

 -(7) For the compartment structural design pressure evaluation, provide the peak calculated differential pressure and time of peak pressure for each node. Discuss whether the design differential pressure is uniformly applied to the compartment structure or whether it is spatially varied.

if the design differential pressure varies depending on the proximity of p the pipe break location, discuss how the vent areas and flow coefficients \ )\

 }

Q&R 6.2-1 Amendment 49

STP FSAR ATTACHMENT ST-HL PAGE% AE- /7t/y<)OF:9 Question 022.02 (continued) were determined to essure that regions removed from the break location are conservatively desisned.(4) Provide the peak and transient loading on the major components used to 4establish the adequacy of supporte design. This should include the load c f (t)) and transient moments forcing (e.g., M functions (e.g., f _(t), Flt)) as 1los(o)v,ed*about l .a specific, identified coordinale(t),Msystel.(t),MProvide the projected area used to calculate these loads and identify the location of the area projections on plan and s'oction drawings in the selected coordinate system. This information should be presented.in such a manner that confirmatory evaluations of the loads and moments can be made. ,l 4

Response

The response to the various portions of this' question is identified by number in the left hand margin.

1. The spectrum of pipe breaks analyzed for each subcompartment is listed in Table 6.2.1.1-1. The break which results in the highest differential pressure across the walls of the respective compartment is designated as )

the Subcompartment Design Basis Accident (DBA). The high energy break used for the structural design evaluation of the RHR cubicle will be provided in an early amendment. These same breaks will be used for evaluation of primary system component support design, with the exception of within the steam generator compartment. Currently, a revised mass and' energy release for the equipment support design evaluation within the steam generator compartment is being generated and will be submitted in an amendment.

2. The mass and energy release rates used for evaluating maximum structural load and component supports design are provided in Table 6.2.1.1-1, 6.2.1.2-2, 6.2.1.4-5, and 6.2.1.4-6. The mass and energy release rate used for the RHR cubicle structural design evaluation and a revised mass and energy release for the steam generator compartment component supports design evaluation will be submitted in an amendment.

3. Schematic drawings which show compartment nodalization for determination e

of the maximum structural loads are illustrated on Figures '.2.1.0 L

 ' . .1M-2:, 6.2.1.2 -3, 6. 2.1.2-4, 6.2.1.2- 5, - 6.2.1. 2- 6, and 6. 2,1. 2 - 7. l Compartment nodalization for determination of maximum structural loads in the RHR cubicle will be provided in an amendment. Any nodalization changes from the current configuration will be reflected in an amendment.

4. STP utilized Bechtel code NE699 for the subcompartment pressure-temper-ature analysis, which is an NRC approved code. Refer to a letter from 53 R.L. Baer, Chief, Light Water Reactors, Branch No. 2, Division of Project j

 ).

Amendment 53 Q&R 6.2-2 l ll

ATTACHMENT STP FSAR ST HL AE. /7 )PAGE 9'1 OF q ,Response (Continued)

 \

4. Management to B.L. Imx, Bechtel Power Corporation, Febraury 1979, sub-ject: Evaluation of Bechtel Topical Report BN-TOP-4, "Subcompartment Pressure and Temperature' Analysis." Also NUREG-0609, " Asymmetric Blow- 53 down loads on PWR Primary System", page 13 states that sensitivity studies are not required for NRC approved codes. Therefore, sensitivity studies are not needed for STP subcompartment analysis.

I

5. , only normal flow paths between adjacent nodes based upon drawings of plant layout, equipment, and cable trays which could restrict flows were considered when flow areas were defined. Insulation breakage and plug-ging is assumed to conservatively yield. peak pressure differentials which !

are re:Ld for the civil / structural suppor't design. Insulation for piping l and for components were subtracted from node volumes and from junction flow areas.

6. Representative pressure responses as a function of time are shown on Figures T .i.i.L4 6.2.1.2-20, 6.2.1.2-23, 6.2.1.2-24, 6.2.1.2-25, 6.2.1.2-26, 6.2.1.2-27, 6.2.1.2-28, and 6.2.1.2-29. The differential pressure on structures was conservatively _ equal to gage pressure. The differential pressure on components is the computed pressure different between nodes.

7. The peak differential pressure in various modes and the time of these peak differential pressures are presented for each subcompartment in Tables W M 6.2.1.2-5, 6.2.1.2-9, 6.2.1.2-11, 6.2.1.2-13, and 6.2.1.2-19. Vent and flow coefficients were modeled to conservatively represent the actual compartment configuration. This assures a represen-Junction areas tative pressure distribution throughout the compartment. 1."

as well as flow coefficients are tabulated in Tables ' " ".6.2.1.2-6, 6.2.1.2-10, 6.2.1.2-11, 6.2.1.2-12, 6.2.1.2-14, 6.2.1.2-16, 6.2.1.2-18, and 6.2.1.2-20.

8. The transient loading on major components are given in Figures 0.2.1.2 1h 6.2.1.2-20, and 6.2.1.2-21. The projected area on the three mutually perpendicular directions are given on Tables 5.2.?_ ^ 4 6.2.1.2-7, and 6.2.1.2-8.

 ?

Q&R 6.2-3 . Amendment 53

ATTACHMENT STP FSAR ST HL 970F AE 17Vt/PAGE a q Question 222.1 Since pipe restraints are provided f or large primary system lines that penetrate the reactor cavity, limited of f set type breaks were analyzed for subcompartment pressure analyses. For breaks of this type, the break geometry may resemble an orifice in the brdcen pipe.Data f rom a number of investigators have demonstrated that f or two phase flow the mass flow rate per unit area for orifices is higher than for pipes. Justify that the SATAN-V methods are conservative for prediction of flow through orifices. Orifice and short nozzle flow data are found in:(1) NED0-13418, " Critical Flow of Saturated and Subcooled Water at High Pressure," Sozzi and Sutherland, July 1975; (2) " Blowdown Flow Rates of Initially Saturated Water," V. Simon, Topical Meeting on Water-Reactor Saf ety, Salt Lake City, Utah, March 1973; (3) " Choked Expansion of Subcooled Water and the I.H.E. Flow Model," R.L. Collins, Journal of Heat Transfer, May 1978; and (4) "The Marviken Full-Scale Critical Flow Tests Interim Report; Results f rom Test 7."Response .,sse s_s scTION 6 ' A h C ' ,O ' ' -

 . . , , _ _ - , - . . __ .., , ,,_. ,_ e...- ,, , -_- - ____-

.in ehgy release analyses have been reviewed with respect to curren !

!ata. e data present results which indicate that critical flow rough an orific s higher than critical flow through a pipe. The re its of this evaluat n show that SATAN-V results, in terms of press e , break flow, break ene , and yield conservative results in co rison to the applicable data.

IThe critical flow calcul on in SATAN-V is dep ent upon the fluid conditions at the break loca on. For subco d fluid conditions, a modification of the Zaloudek co elation i applied with a discharge coefficient of 1.0. The parameter whi determine subcooled liquid critical flow are the reservoir pres e and the saturation pressure co' responding to reservoir condit s. e Moody correlation, a thermodynamic eq uilibrium crit al' flow mo with a discharge coefficient of 1.0, is used for saturate and two phase fl d conditions. The Moody model is a function of st nation properties. (S Reference 1 for a more detailed description of hese break flow models). B h of these correlations are ind endent of break geometry.These critical ow models were compared to the subcooled an saturated data presente in References 2 (Sozzi and Sutherland) and 11 (. rvik en results). eferences 3 and 4 were not suitable for direct applic tion to the met s used in this analysis since they presented summaries o current resea and did not include suf ficiently detailed data. However, se eral of e studies referred to in these literature surveys were reviewed f a glicable data (References 5-8). Most of the references did not specif;,Q&R 6.2-31 Amendment 8, 10/22/79

ATTACHMENT STP FSAR ST HL-AE- IU/d PAGE 99 0F igg .

7 C'y Teservoir conditions, and it was therefore not possible to compare the Moody or Zaloudek equations to this data. Sozzi and Sutherland generated re esentative data f or nozzles of varying entrance geometry, lengt /and dia e r. The Marviken report used here obtained data for one not e only, In the t o-phase region, the Moody model underpredicted some of- e applicable data in the region of 0 percent to 0.4 percent quali y.References 6,7and8alsoprovidedtwophasedatapoints[orcompariso s purposes. By pplying a multiplier to the SATAN-V correlatidn, the data fell beneath the curve generated from this modified correlation. The multiplier decrekes linearly from a value of 1.4 at zero/ quality to 1.0 a t 0.4 percent quality Similarly, the Zalo ek model was nonconservative when compared to serval subcooleddatapoints.%dditionaldatafromReferences9and10were included in this comparison. AmultiplierontheKaloudekcorrelation, varying linearly f rom 1.4 \t zero degrees subcooJin'g to 1.0 at 50 F subcooling, provided a correlation which represented an upper bound to most of the data. Only a few of the Marviken datajibtained in the early portion of the transient did not fall neath the curve generated by the adjusted break flow model.However, closer inspection of the rvikpn results shows that, in general, the trend of the Marviken data with Kespect to subcooling is opposite to other data trends (e.g. , Sozzi and Sum _ erland). It is felt that the results obtained from the methods of/fi w in this facility were not indicative of the actual break flow /in t e very early portion of- the test k' due to the transient nonequilibriynl effee g. However, the modified correlation still provided a conservative prediction of nearly all the Marviken data, expecially those/obtained durlog the transient where the data are felt to be most meaningful.

 /

In order to determine the effect of these modified s correlations, the adjusted critical flow model was programmed into the SATAN-V code f or all break locations analyzed. These locations and break \ sizes are specified in Table 1. j

 / \

J 1Results indicate tha/t for all break locations, the analy\se,s performed using the adjusted critidal flow model compare favorably to the analyses which used the unmodiffed correlations. The largest variation in treak flow was less than 1 perdent. Figures 1 through 5 present plots of brhak flow transients ob,tained using SATAN-V with and without the adjustekcritical flow model.need for rea/Results for all was naly sis. There break nolocations change to clearly do not the peak indi ate release rat any for all breaks, and the transients shown in Figures 1 through 5 are virtua ly identical. Since a 40 percent margin was applied to the pressure diffeyentials calculated from these break flow transients, the resul s currdtly reported are sufficiently conservative under review of the current critical flow data. Of this 40 percent, 10 percent is directly attributable to the calculated mass and energy releases.i l Q&R 6.2-31a Amendment 8, 10/22/79 k

ATTACHMENT ST-HL AE- / 'IW/STP FSAR PAGE /000F l.fi_j YREFERENCES

1. " estinghouse Mass and Energy Release Data for Containment Design,"

W -8264,

2. .Sozzi, L. , and W. A. Sutherland, " Critical Flow of Saturated and Subcooled ater at High Pressure," NEDO-13418/

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3. Simon, V. , "B down Flow Rates of Initially Saturated Water," Topica L M'eeting on Wate hReacter Safety, (Salt Lake City, Utah, March 1973).

 ~ /,

4. Collins, R. L. , "Cho ed Expansion of Iubcooled Water and the I.H.E.

Flow Model," Journal o _ Heat Transfer, 100, (May 1978).N Moody, F. J . , " Maximum Flow Rate'bf a Single Component, Two-Phase 5.Mixture," Journal of Heat Transfer, (February 1965).

 /\

6. Starkman, E. S. , V. E. Schrock, K3 F. Neusen, and D. J . Maneely,

 " Expansion of a Very Low Quality T%-Phase Fluid Through a Convergent- ivergent Nozzle," ASME Jhrnal of Basic Engineering, (Juye 1964). j

7. Henry, R. E. , and H.' K. Fauske, "Two-Phas Critical Flow at Low Qualities," Nuclear Science and Engineering, 41, (1970). .

of Liquid-Vapor

8. Cruver, J. E. ,/and R. W. Moulton, " Critical F1 ).

Mixtures," A. 'I. Ch. E. Journal,13, (January 19

 . , " Flow of Subcooled Water Through Nozz s,"

9. Powell, A.' .

WAPD-PI(V)-90.1

 /

10. Schrock, V. E. , E. S. Starkman, and R. A. Brown, " Flashing low ournal of of Inipially Subcooled Water in Convergent-Divergent Nozzles," '

Heat Transfer, 99 (May 1977).

11. "The Marviken Full-Scale Critical Flow Tests Interim Report; Results from Test 7". -_

l Q&R 6.2-31b Amender;nt 8, 10/22/79

 - ~ _ __

ATTACHMENT STP FSAR ST HL AE /7M/PAGE /o/ OF laq Table 1-Break Location Break Area, Sq._In.SG Inlet Nozzle Elbow 777 SG Outlet Nozzle , 755 RCP Cold Leg Nozzle 594 Reactor Cavity Cold Le zzle 150 Reactor Cavity HWLeg Nozzle 150 30 6 9(7o(Q&R 6.2-31c Amendment 8, 10/22/79

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STP FSAR ATTACHMENT ST.HL AE 17V(/PAGElo10FJgr)Question 480.04N In Section 6.2.1.2.2.1 of the FSAR, it is stated that for the reactor cavity subcompartment analysis, the postulated pipe rupture occurs in the inspection toroid and only a small fraction of the blowdown enters the reactor cavity.Justify that only a small fraction of the break flow enters the cavity, i.e.,discuss how the break flow is prevented from entering the cavity, and why it is not appropriate to postulate a break inside the cavity. Provide appropriate plan and elevation drawings of the reactor cavity showing the inspection toroid, piping, pipe restraints, postulated break location and vent paths (flow area) to the reactor cavity and steam generator compartment, including the blevout panel for venting to the lower reactor cavity.

Response

 ,g,g g Section 6.2.1 3has been revised. L. p::;;rrp 5.2.1.2.2 f;r; 1tisenssion vf' dee4gn-feeeuresm c:ntsin-tha. distribution-ef blewdv.... Cec-Figuras.

6,4-idri und-6.2.1.2 2 fer-pha nd slevation-drawings-ef-the toruitly piping, yir. ...is.1Trtr ant!%a rr sct**-eevity..Addii.-iv..1 dwing- ==*= p*=id-d nd-r-separ=#- wr-1.iL. - deced . ene 14, 3A"5,:;T4ti.'Atrt272.I STP has app e (for'/ exemption from certain provisions bf) 4 (ST-HL-AE-1096 dated JulyJk198h -AE-1200 dated March 1, 198 Fwh ay affect the re use-Trovided for question. f l( Vol. 2 Q6R 6.2-5N Amendment 53 jl g ,, .- . - _ _ _ _ _ _ _ . - , _

STP FSAR ATTACHMENT ST.HL AE /N4 PAGE /03 OF 14 r)Question 480.05N The reactor cavity model described in Section 6.2.1.2.3.2 indicates that "one hundred and eighty-degree synnetry was assumed...". Explain and justify the use of the one hundred and eighty-degree model versus modeling the entire reactor cavity and inspection toroid.

Response

4 g Section 6.2.1 has been revised. 0,,__;..j ef :b;;;_;;; . .;,, 1. nu 1. . ...;

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i;v W .STP has app .for exemption from certain proyisio s af GDC 4 (ST-HL-AE-1096, dated Jul 7,1984 4-HL-AE-1200, dated Marfh 1,1985)'WLch may affect the resp ge provided for this cuesitan

 +

Vol. 2 Q&R 6.2-6N Amendment 53

STP FSAR l ATTACHMENT ST Hl. AE Old PAGE 109OF M).M Question 480.06N Concerning the blowout panel in the heating, ventilating, and air conditioning 'ducting leading from the loop compartment subpedestal space to the lower reactor cavity (i.e. , junction 110 in Table 6.2.1.1-4):

1. Justify the constant vent area of 4.05 square feet given for this vent path in Table 6.2.1.2-4.

2. Provide the dynamic analysis of the blowout panel that gives tha vent area as a function of time after the break.

3. Provide drawings showing details of the blowout panel and surrounding areas. ,

4. With regard to possible generation of missiles, describe the potential for damage to safety-related systems by the blowout panel during a loss-of-coolant accident within the reactor cavity / inspection toroid.

Response ,g,g l as been revised. %e-ftr1663 ..ei,c,r.ac. ..e 1 arrive da Section 6.2.1 ,m

 .wevised-eeetten . . The HVAC panels are in junction 108 of the revised analysis model. Theti .

are two panels, one on either side of the ducting supplying cooling dir

 < Yq m the reactor cavity cooling units. The total vent area is 3.5 square fee *..

N ,/ '

2. The panelss are assumed to relieve at 1 psi differentia 14ressure across the panels.NSince they are light panels they are assumed to provide full open area instantaneously when the differential pressure value is reached.

This occurred at 0.122 seconds into the analysis. -

3. The surrounding area issshown in Figure . -12 and drawings showing the blowout panel locations are,provid . the responso Q480.04N.

4. There are no safety-related equipment or components susceptible to damage by a missile which might be eyeated by the blowout panel during a 1DCA.

Safty-related equipment in e space with the HVAC panels are protected ,by concrete structures ocated at a.'substantially higher avaluation taking them out of the rea of potential impact. This will be verified as part of the ongging hazard analysis progra:n, N.

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Note: 'N STP has plied for exemption from certain provisions of GEC 4 (ST'-ML-AE-1096, dated ly 17, 1964, ST-r.L-AE-1200, dated March 1, 1985) which may affect the res nse provided for this question. N 't Vol. 2 QSR 6.2-7N Amendment 53

g. . . .. .........,---. ,.. ....., -.

ATTACHMENT STP FSAR ST.HL.AE-l?qtj l PAGEII0OFsoJ) gpestion 480.07N

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For the reactor cavity analysis, provide justification that vent areas will not be partially or complet61y plugged by displaced objects (e.g. ,insulation). Of particular cencern is the rationale for not considerin5 the blockage of the vent paths threugh the restricted clearance spaces around the primary piping nozzles in the reactor cavity / inspection toroid analysis. ,

 **" "'I . 0L . M . l Section 6.2.14 has been revised,t- i._1... ---. yl.. 1., L, 61 y1 l .tj:::. .. .. n w f= - - -- ' . t h a r::n: ;;;; 2. :r;;h :: , 304 eld ripius watt is

y. a assumed mi m..

to be col pipi netrations in the primary reduced 1 gging caused by displace objects,duceshouldbenotedthatthel nozzle k h ostulated in the toroid. Toj iqageofthespace j bet the nozzle'and the flow limiting spai plate would anderestimate er -" r: in-the::::ter es-'*y 7xemption from certain provisions of T-HL-AE-1096,f STP has appl dated Ju 7, 1984, E-1200, dated March 1, 198- hich may t the resp e provided for this que n.1 Vol. 2 Q&R 6.2-8N Amendment 53

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STP FSAR ST Hl. AE-tit /L ATTACHMEf4T]F PAGEfft OFif )Question 480.08N k In Table 6.2.1.2-1, it is stated ther the short term mass .and energy release rates " include a 104 margin not used" in the subcompartment analyais. Explain the origin of this 104 margin and justify why it is not to ba used in the subcompartment analysis. Present the mass and energy release rates actually used in the subcompartment analyses.Responne The 10 percent margin was applied to calculated results by the NSnS vendor and represents an additional degree. of censervatis:n deemed unnecessary in view of the other margins (e.g. volumes, vent areas) used in the calculations. 'fne mass energy release rates used in the subcompartment ane.lyser, are cbteined by multiplying the values in the table by a facter of 0.9091.(.bote: _--- 7 iSTP has applied for exempt 1 m certain previsions of CDC 4 (ST-HL-AE-1096 l dated July 17, 1984, -AE-12 tch 1, 1985) which may affect the response provid W or this question. __( [Vol. 2 Q&R 6.2-9N Amendment 53

 +, - - - - - - - - ,-,- + - , - e , 4

i iATTACHMENT STP FSAR~ ST-HL AE lqqt{PAGE llR OF igq guention 480.09N Provide the results of the nodal sensitivity studies performed for the steam generstor auhcompartment analysis referenced in Section 6.2.1.2.3.3. The j koncern results because of the gross nodal modeling, particularly in nodes 1 ;throu6h 8, which does not account fcr flow restrictions and variations around 1 piping and other obstructions.l 1

Response

Section 6.2.1 has been revised. The steam generator subcompartment analysis hss been changed to include a finer nodalization, specifically 33 nodes instead of 8 nodes used previously. This finer nodalization more carefully cecounts for flow restrictions presented by piping, platforms, supports, a cquipment and other structures in the steam generator compartments. The NRC approved COPDA computer program (Reference 6.2.1.2-2) was used to perform the subcompartment analyses. The modeling for the COPDA program requires that nedal boundaries be taken at significant flow restrictions. The addition of arbAtrary nodal beundaries would violate this requirement and could lead to arreneous results. This and the guidelines and recommendations of Section 3.2 of NUREG-0609 have been followed in the nodalization. In light of this and since COPDA is an NRC-approved program no sensitivity studies were performed.

• This is consistent with Section 3.2.1 of NUREG-0609.

Note: ,sf STP has applied for exemption from certain provisions of GDC 4 (ST-HL-AE-1096, j dated July 17, 1984, ST-HL.AE-1200, dated March 1, 1985) which may affect the response provided rne eht. nu..rean 1Vol. 2 Q&R 6.2-10N Amendment 53 l

ATTACHMENT STP FSG ST ML AE-t*i44 l PAGElfs OF 6,2'?ll :l jQuestion 4BO.12N The response to QO22.2, regarding the subcompartment analysis, is incomplete since it refers to a future amendment, Provide additional information to +complete the response.Response ,The response to QO22.2 has bsen revised.l( .?e If 3Note:STP has applied for exemptiseJrom tain provisions of CDC 4 (ST-NL-AE-1096, dated July 17, 1984, ST-HL- A ted March 1, 1985) which may affect the i response provided for th g question. N_ w W ' .I(Vol. 2 Q&R 6.2-13N -Amendment 53 ,9 F

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e 5 %-STP FS a YTACHMEN ~ , l ,ST+0AE-17 dQE JJ!# OF f j

 . 6.2 CONTAINENT SYSTEKE 6.2.1 Containmerit Functional Design 6.2.1.1 Containnent y ructun.

Design Masant The Containment design basis is to lisd t the

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6.2.1.1.1 release of radioactive asteriele, suksequen't tp postuieted accif ttus, Atch that resulting culculated offsite &ves are less than the ;guideLtw r9 1ces of 10CTE.100, In order to meet thtt requiesent, et deign (maritcu) Qnt.sitnant leakage rate has been defined in csuptnettom vitt perfornnu Jegwiptt?r.ts placed on other Engineered Safety Feittte (23F} syaters..The capability of the Containment structcre to maintsin Leedrigbs jltit% City and to provide a predictable envir.ontwne f2.r operation of EST yystacq tc en-sured by a comprehensive design, analysts, en.$ test (sg prograr th41 i W u h t consideration ef:

1. Peak Containment pressure end tangerstn'ca masacisted with thr Aest pwre postulated eccident coincidso3 whh tW SM s Ecutdown Eoche@ -

2. Maximum etternal pt'essura to shieb tche Ccntainrott say to mtgered np 4 result of inadvertent Conteirer. SystaW oferstieres tMt trenWily reduce Contaitxment internal prastwre bef. ort tuh8Ae M.ttt3pbrh btu sta-O <

6.2.1.1.1.1 accidents consideredPcstulated Ac.;tdepMeyjitg in deterWin 5% t at.uicaentp'Tc?tesp0Cr,TtM rat pr6WbtCJ p gesh yte.34ft (ed

 .. temperature), wobecapartment pesi pr.v.44.r% . tad cu nnaT vWFpy Ws s'rke <

rized in Table 6 2 1.1-1. TAe gie.(rn 9f trerte v-*i in tM Mygeshy Cora(- Cooling Eystem (ECCS) analysfr fcy fidsimma CotH bat.tV 4.',ebgel m 42n fit Sa- ,

 ~

fined in Section 6.2.1.5 and betica 15.6.5. iTel- tenoittsd +2rtqvittut $9 pipe break accidents, a discutifrio W .treak Ac M pok r. g$vtetuidMtm g<3 ' 2 W&ggy-Q l1 For Centainment structure and rnrre ptrtset?t pirts pse tJea SWyM t,. b Att assuced that each accider.t can @ rte *c.:drie5q W th $ ictk il0 of%fia y&.r@ f9 j and the most l'imiting single attivt is11st'ei p t w cec citb @ c usa a d w t d t6 ;cccur simultan6eusly or consecutiv6 7 ,For each of the categories of Coatstumenc issk plc.t&% . MrorpwQ9: pMk ,pressure, Containment external prerr;re, anj . Gens 1 Jet Wdev W4.@. r.be l Design Basis Accident (DBA) la #f1N3 La CM hur arMu d 60 vetM;La 4f l accidents postulated for each e.ase.% DEA. c46.Shne Mhmted urFS between calculated and de sign pressvre valws, ao4 bnb 2 W ?he vqit tw 7 containment are given in Ts.ble 6.2.1.1-8 AncitM h*(p lsestern coe given in Table 6.2.1.1-3. The DBA ce2tutetsv4 pie savt M thd seMtm ttht. sert qq$M ,calculated an'! design pressure valt:ss f e r vultu CGit topstfM4 er4 2=.fr a 7 are presented in Tables 4.2.I .2-2. 6.2.1:AD. GJ ,1.bb r. 2.1 7 W M.1.h13. ,! 6.2.1.2-15, 6.2.1.2-17 and 6 .1.L 2-10. (l 6.2.1.1.1.2 Mass and Entray Relesse' ~ TM Frartes aM anun9 ef %u>and energy release for the most severe of the 4Wifet.Ca Liste i 'd W.m ,(6.2.1.1-1 under the categories of Containt.setet p*41 ttM .tCt, trucMpumrA l6.2-1 muWAM i iL_ ~ ~ - - -

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 .n 'I ATTACHMENT , s }ST-BLAT.-l1tj$

PAGE NS Of f A q l ' :j..Jn5ERT PGge 6.2-1 ,As discussed in Mferecce 3.6-14 and peragraph 3.6.2.1.1.la, RCL ruptures and th associated dyisic eff6 cts are not included in the design bases.-5@c:xtpartmeot an.atytes are based on RQ. Aranch pipe breaks or secor.dery 1.ys.ts t pipe brc6ks. Containment pressure and teciperature design is based on nor<enechen5stic r!c,ubVe er43d 390totine tfL breaks.i ii Il ti lf bi il l-l lt 7603N:0288N/17

STP FSAR ST-Hl.-AE. Flyt ATTACHMENT PAGE h OF r Aq{qualification for equipment in the RCB were IDCA and MSI.B. A spectrum of

 ", break sires was considered in equipment qualification. The MSI.B provides the 49 highest RCE atmosphere f 3mperature; the IDCA provides the highest RCB atmo-sphere pressure. Combined MSLB/LOCA presuure and temperature profiles will be used for qualification of the equipment.

l 6.2.1.2 Containment Subcompartments.6.2.1.2.1 Design Bases: Subcompartments within the Containment, princi-pally the reactor cavity, the Steam Cenerator (SG) compartments, the pressur-12er compartment, the surge line compartment, the main steam line compartment, and the feedwater line compartment, are designed to withstand the transient differential pressures and jet impingement forces of a postulated pipe break.Venting of these chambers is employed to keep the differential pressures with in structural limits. In addition,:;_r. ....:- *' r P__.. ri g reaucee-

 - l49 -- :rw 241 -i=9" _ dct neither pipe whip nor forces transmit-ted through component supports threaten the integrity of the subcompartments of the Containment structure. p l'9 The spectrum of pipe breaks analyzed for each subcompartment is listed in Table 6.2.1.1-1. The characteristics of the h pipe ruptures ver k determined in accordance with the methods and criteria of Section 3.6.2.j

r. _ ;ael ely- h a *ka n- ~

• a r --! -
• D: p f:: th: ire.l. Id....ificu in Section 3. -ielded e Mulated break flow areas Ns an equivalent single-ended b k ow areas. These reduced brea reas result from consid-49

 ,ering the loop pi stiffness, primary equipse an rimary ;nt; equipment sup. i

{.ports,andres aints gned spe 111, w h pig displo.. i .. J.C,. . _..; J y :_1;t:f pip: n; " . The accident that results in the maximum differential pressure across the wal s of the respective compartment is desig-nated as the subcompartmentydsign, asis / d;..; '^*M . Calculated inte dif-ferential pressures are compared to the design pressure values used in the structural design of subcompartment walls and equipment to ensure that peak calculated values are less than design valuec. These design and calculated pressure differentials are presented under each subcoupartment section below. 49 6.2.1.2.2 Design Features:6.2.1.2.2.1 Reactor Cavity - The reactor cavity is a heavily reinforced concrete structure that performs the dual function of providing reactor vessel support and radiation shielding. It is describcd in Section 3.8.3.1 and is shown in the general arrangement drawings of Section 1.2. At the elevation of the primary piping nozzles, the reactor vessel is surrounded by an inspection 4 bi A/* A!- 49+g toroid,in di d-0.. yi r- r~tSY i

• phe.s posw.wrs.h su nc Rharc;e c;4bn;y oni d b r--sNsPEenen ronm$,V'd WP

 ..; ".w... din 5n.ii . 3.-0.0.1.e.1, . .. 1 71 t: i: 7:rcid:' rr = f th: n- q iel n les in the spection toroid to minimi e the blowd vn into the cavit o' rhus , a lativel small amount of mass and one from e break in the I toroid ent e cavity through the rastricted c r e spaces between the o seal plate ozzles and the opening between the el support and the concrete is a und the support. Venting from e re tor cavity is achievef through ( the door connecting the e ity to th oop compartment .. jubpedestalspacean b) two blowout panels n the heating, entilating and

{ y a _..oiciout-,,, i.ct L at.-a6*!? 0 JJ;;; Lh* 'aviU7 LD '""1"r "-i^~ "6.2 9 Amendment 49 3

ATTACHMENT-i ST HL- AE- 174tl9 PAGEfl70Fla

I'NSERT Page 6.2-9 INSERT 1 As discussed in Reference 3.6-14 and paragraph 3.6.2.1.1.la, RCL rupt'ures and

 ~the associated dynax.ic effects are not included in the design bases.

7603N:0288N/23

ATTACHMENT ST HbAE 19yQ PA Ella OF;;L?STP FSAR nugy...... .

 .... .r.jer p; ;i;; ef ;Le L1-.ov... zrom the cold leg rpyture is I ,I disch rged through primary piping penetrations into the SG compayttent and from t e to the Containment. The reactor cavity and inspecJton toroid subcompart nt is shown in Figure 6.2.1.2-1. The physical, a'rrangement of The the .igure 6.2.1.2-2.

seal piste, r torvessel,sealringetc.isshownin)ofthesectiongoing igure is in section with the plan,e-left half of thi ithout support, while the right half is in section with through the cold le the sectional plane gol through the hot leg. e blowdown takes place in the inspection gallery. ,3 The total free-volume of the torg,id ip ,177 ft . Eight primary pipe penetra-2 tiens from the toroid to the SGs h)v' a total combined vent area of 103.2 ft while a total flow. area of 14.J 1 'it2 s the inspecgion torus to the cavity The reactor cavity has a ne M ree volume 0,537 ft with the following two vent paths into the subppdt'stal region (a) a door of area 23.6 fts opening at 2 3.0 psi differential blowing out at 1 p,si,paa (b) two panels in the H74 duct an area of 13.5 ftdifferenti for the design.jche worst case (in which the HVAC plate attuated in the support nea e break blows out) is presented. Pipe restral ts are employed to limit e circumferential break flow area to less than 150 The resuJt of this analysis are used for the reactor cavity wall desig d to[ g W rata the vessel support loads.6,3,1, Q *3 6.2.1.2.2.2 ' ,g Generator comoertments - The SG subcompartments are ,34{showninFigure-ec_g) The SG and its supports have been described in Sec-tion 3.8.3.1 and the general arrangement of the SG and associated structural 49 arrangement is presented in Section 1.2. The general arrangement drawings ,presented in Section 1.2 have been used to define nodal boundaries of Figure ;6.2.1.2-3. The SG subcompartments consist of the entire free volume between the primary shield and ths secondary shield walls and frge El.19 ft to 83 f t.Each quadrant containing a SG has a volume of 41011.0 ft and has a vent area to the Containment of 571.3 ft2 at the top of the SG compartment. In addition to the above vent path, two r. ore vent paths vent the break nodes to the Con-tainment. These are (a) the 8 penetration paths that lead the hot and cold leg pipes to the reactor cavity with a vent area of 105 ft 2 and (b) the 6 HVAC vents between the SG compartments above El. 19 ft and subpedestal region below EL.16 f t with a total area of 146 f t2 SG compartments A and D, and B and C are directly connected together while A and B and C and D are connected via a passage which ranges from (El. 33' 6 1/2" and 19'0") to (E1. 26'2" an 19'0"),re ectively. The SG subcompartments are shown in Figure 6.2.1.2-3 (Sheet 3 of .N 6.2.1.2.2.3 Zressuriser Comoortment - The pressurirer subcompartment, shown in the general arrangement drawings of Section 1.2, consists of a verti-cal, rectangular, reinforced concrate structure surrounding the pressurizar which is supported at its base by a ateel skirt. The preasuriser subcompart-ment is shown on Figure 6.2.1.2-4.. 6.2.1.2.2.4 Surne Ling Subecmcertting - The surge line subcompartment consists of the area above the grating at El. 37 ft 3 in., the area below El.37 ft 31n., and the vestibule where the surge line penetrates the secondary shield wall. These subcompartments are shown in the general arrangement draw-ings of section 1.2 and on Figure 6.2 1.2-5.6.2-10 Amendment 52

ATTACHMENT STP FSAR ST-H L- AE- M PAGE liq OF i 2 '1 6.2.1.2.2.5 Main Steam Line and Feedwater Line Subcompartments - The sain steam line and feedwater line subcompartments ar6 located between the +[ secondary shield wall and the Containment wall where ruptures in these ;9 lines may occur. The general arrangement drawings in Section 1.2 show the lequipment and structures in these locations. The most confined spaces result-ing in maximum local pressures from either break are near the pipe penetra- ,tions to the outside of the Containment. Vent paths consist of a combination of series and parallel flow resistances joining major elevations of approxi-i mately one-half of the Containment. The subcompartments are shown on Figurs ,6,2.1.2-6.4.2.1.2.2.6 Regenerative Heat Exchanger Subcompartment - The regenera- ,tive heat exchanger subcompartment arrangement is shown in the general ar- !rangement drawings of Section 1.2 and on Figure 6.2.1.2-7. The nodal model !i net free volumes and vent areas are listed in Tables 6.2.1.2-15 and [6.2.1.2-16. The vent areas out of the regenerative heat exchanger subcom-partment consist of two openingt-the auxiliary feedwater pipe penetration 'opening and the wire mesh door. The subcompartment volumes and vent area are reduced to account for obstructions caused by equipment and insulation around piping and vessels. No blowout penels are used, thus the flow area is assumed !to be constant with respect to ;_me.6.2.1.2.2.7 Radioactive Pipe Chase compartment - The radionctive pipe ,chase subcompartment is shown on Figure 6.2.1.2-6. The nod &1 model not free volumes and vent areas are listed in Table 6.2.1.2-17 and 6.2.1.2-18. The ,vent area out of the subcompartment is a manvoy hols in the floor. The sub- l compartment volumes and vent areas are reduced to account for obstructions caused by equipment and insulation around piping and vessels. No blowout 49;panela are used, thus the flow area is assumed to be constant with respect to .(.' tLae.4.2.1.2.2.8 RHR Valve Room Subconnartment - The RHR valve room subcom- ,) partment is shown on Figure 6.2.1.2-9. The CVCS letdown line passes through both RHR LA and RHR 1E valve rooms. Because the valve rooms are identical, ,P only the valve room for RER LA is modeled and the results of the analysis are !

representative of both rooms. The nodal model not free volumes and vont areas j are listed in Tables 6.2.1.2-19 and 6.2.1.2-20. The vent area out of the

subco=partment is via a wire mesh door. The subcompartment volumes and vent areas are reduced to account for obstructions caused by equipment and insula-tion around piping and vessels. No blowout panels are used, thus the flow iarea is assumed to be constant with respect to time.

 ~

4 i 6.2.1.2.3 Design Evaluation:i 6.2.1.2.3.1 Ceneral - The subcompartment pressure transients were deter-mined using the COPDA computer code (Reference 6.2.1.2-1). The COPDA code employs a finite difference technique to solve the time dependent equations i for the conservation of mass, energy and momentum. This code and the assump-tions inherent to it are fully explained in Reference 4.2.1.2-2. Loss coeffi- i cients utilised were based on the formulations of References 6.2.1.2-3 and ,4.2.1.2-4. ,l l l

 $. l

! 6.2-11 Amendment 52 s>>l i

 . )

\> \

ATTACHMENT ST HL-AE- 1744 PAGE 12OOFlW7 STP FSAR 4Nodalization of each subcompartment was based on the physical arrangement of '

 )

the interconnected subcompartment and the structure, equipment, piping, venti-lation ducting, floor grating, and other physical obstructions to flow. By appropriate selection of node boundaries based on the physical arrangement, pressure differences within a node are minimized while pressure differences between podes are maximized.The Loss of Coolant Accident blowdown model used to calculate the short-term l49 mass and energy release rates for all primary system ruptures, including the surge line break and the preissurizer spray line break, is fully described in Reference 6.2.1.2-5. The mass and energy release data are presented in Table 6.2.1.2 1.The REIAP 5 code (Reference 6.2,1,2-6) was used to calculate the short term! blowdown of the main steam line and main feedvater line. The mass and energy release rates for these two lines sre provided in Table 6.2.1.2-1. Letdown 49 line break blowdcwn was calculsted using tuethodology of Reference 6.2.1.2-l and is given in Table 6.2.1,2 1. cg posrupru wE) 6.2.1.2.3.2 Reactor Cavity - We pip 3 6 bresh.f C ' ' ~_. , reactor cavity and inspection toroid.i'::i;;n- emhr ':r rs 1.1~7,--lu.7

 ^

1Mid my1m.. ..in 9 " A W -leg et er erf: :c ':: 4 J. L _ w _ _ .1 ,;; .s 7 p- & nin*= mee de s inne_d_ _ to an=nre- = hr= =1r fl eu -m ef - 1;;:: - 9 thir 8Amewnt.I The nodalization scheme selected for the model in shown in Figure 6.2.1.2-17 i The no ' boundaries were chosen wherever restrictions to flow ocodrred, The pode boun esinthetorusruninthecenterofthecoolantgpesandaddi-tional nodes dirq designated for the support ares adjacentjo'the break nodes.Nodesinthesuphrtareaweremodeledbecausethisvauldbouhdal.1 cases including the one ih which the IWAC plate welded to $he supports will be blot n ,out. Two cases, one . which the HVAC plate stayyIn place and the other in.which it blows out, have een studied. The casV in which the HVAC plate blowc out h e been shown to give e maximum calcu)4ted premeure across the shield ,wall. /in the analysis, the insulation on reactor vessel was conservatively as-sumed to remain intact. (uncrushed) ndXo be flush against the pressure vcs-sel. Ht. wever, the insulation ch exten below the bottoni of the vessel is 09 assumed to briesk loose into a 1 crushable aces.The insulation around th coolant pipe is cense tively assumed to plug roce sections of the wagon- eel restraint in the penect tion. Thus, with these j

 ' major assumptions, tb $0 node,122 junction incdel o is,uro 6.2.1.2-10 is l I simulated for the sq, in. break using COPD4.. The v ' um& of esch subcois- i psrtment 6s well s the initial conditions prior to the p tulated accidant faregivenonT e 6.2.1.2-2, The junction parameters 6 nam y vant areas, L/A's, head ss e.oeffients used in the calculatis>n. are give. in Able

6.2.1.2-3./ he t head loss is presented under two headings namel expansion '

losses a contraction losses.4 l %ei mogeneous equilibrium option has been used in the analysis. Th g flow 3ption is described in Reference 6.2.1.7-2. Tne re6ulting peak pressure J ,W 6.2-12 Amendment 49

 ~ . _ _ _ _ . _ _ _ . _ _ _ . . _ . _ . _ _ _ _ _ _ _ - . _ . _ . _ _ - _ _ . . _- _

ATTACHMENT STP FSAR ST.HL AE- INQ %PAGE131OFla 9 )i

1. y....ntea in Table 6.261.2-2. Tne complete pressure time profiles for al subcompartment nodes are shown in Figure 6.2.1.2-18. [

The subcompartment ures, when applied he projected areas of the sub compartments on the react .ssel, y} the force on the vessel. These force components at various eleY : impose a moment about a chosen axis.The axis system passing thr the co hot legs and the center line of the vessel has been sole to determine the nts for the vessel. Time histories of the hor ntal and vertical forces an ding moment imposed oa the vessel by t symmetric pressurization of the reactor ty are present -ed in Figure . .1.2-19. The force and moment coefficients for subcom-partment are given in Table 6.2.1.2-4. h esssoenese soe 6.2.1.2.3.3 Steam Generator Subcompartment SG sut88dy".7,ghng"[d ~st...u 71r preasure is determined by hemeeed=seco-breaks in the r r-trr -_ _12- ' E--

1f i:;;r. T': pip::;;rrte ;nf ..ke.;;;___i... 11 6 sue o dak .1 . to 1::: -' '~^' :i f e..; i.;;is . The $1owdown for decep reaks 90-inivt g)

 ";.... 1;l L...i 'rli- y a ' '- rplit, 00 __;.le;..... 1. 11_ii.0 .... 1.._ ', ou ,_;.1;; ;:ssh li_1..J ... 61. . ..... 1e1 i;::' (;; rill::1..; _suoi)

I 1;;;;.i1. 11 1 ;f:::: ;i.. 'eventi:1 i...i ':pli _.0.1); , ...d 4GP

 @*'-t n;;;1; 11 1;..J .... ci.. f;;;;tici i;;;t (;;-'11 a* h- r ':1) is presen-ted on Table 6.2.1.2-1. The noding of the SG compartments is shown on Figure 6.2.1.2-3. The node and junction disgram is shown on Figure 6 . 2 .1. 2 - 11. . - dre-. '- '" r---trr :2..., ... ". , _ E..,J;he flow parameters vers evaluated to account for all obstructions such as cabn trays supports and various small sized piping. The principal obstructions within the SG loop compartments are the SG 49 and reactor coolant pumps. The node and junction parameters for the SG loop compartments are given on Tables 6.2.1.2-5 and 6.2.1.2-6.

The flow from one node to the other was calculated using the hom*ogeneous equi-librium model option for the analysis. The peak pressures for each subcom- _partment are listed in Table 6.2.1.2-5.E Below is a summary of the break size and break node for each one of the 10 e :

1. Steam erator inlet nozzle split. 777 in n , bre node - 12

2. Steam Generato tiet nozzle split, 777 , break node - 15

3. Steam Generator outlet zie split, 7 ins , break node - 12 t

4. Steam Generator outlet nozzle t, 755 in n , break node - 14

5. Steam Generator outlet n sie split, 7 n 2 , break node - 2

6. Steam Generator ou et nozzle split, 755 in , 2 k node - 3

7. Steam Generat inlet nozzle split 777 in ,2break node 2 and 15

8. Reactor oolant? ump outlet split, 594 in2 . break node - 12

9. Reac or Coolant Pump outlet circumferential break (guillotine model), 594 inn , break node = 12 l 6.2-13 Amendment 52 l

ATTACHMENT ST HL AE IFl4 PAGE 1440F /29 STP FSNR rv . accam upnerator aucAen nozzle EuxAlum..;, 777 *2 , oisaK nodes - 12, 14f l J 50 percent blauda-- '..n The pressure differential given on Table 6.2.1.2-5 is generally evaluated with ^respect to node 41, the Containment volume, except where specified. 0....a t ,a -.. -- ..r,--.-a m, ,w<= ,.w3. .4--- .t --t-- d. a42 :: , ,,1g-w<.s --g 2,cc .--.<1_ -......... ow.- ----

f te ;he e;hmi L --a c====. The pressure time histories for all th==m== cases are presented in nodes close to the b'reak in Figure 6.2.1.2-20. The nodes considered here are 1, 2, 3, 12, 13, 14, 15, 21, 22, 23, 24, 30, 31, 32, and 33. These nodes are in the SG compartment in which the breaks occur.

49 hME ne' Force coefficients on the SG and RCP hegbeen evaluated to help. facilitate determination of forces and moments due to the pressures generated by the breaks. Force coefficients represent the projections of the SG and RCP on three mutually perpendicular planes selected for this purpose. These coeffi-cients have been presented in Table 6.2.1.2-7 and 6.2.1.2-8 for these two components.The forces and moments plot versus time for the SG and RCP have been presented on Figures 6.2.1.2-21 and 6.2.1.2-22 for cases 1, 7, 8, and 9.6.2.1.2.3.4 Pressurizer Subcompartment - The pressurizer subcompartment design pressure is established by a double-ended break in the pressurizer spray line at the_ side of the pressurizer. This break location is in the most restrictive location and results in the maximum pressure and equipment load. ;The noding of the pressurizer subcompartment is shown on Figure 6.2.1.2-4 with a node and junction diagram provided on Figyre 6.1.1.2-12. Node and junction parameters are provided in Tables 6.2.1.2-9 and 6.2.1.2-10. Plots of calcu-lated pressure are given on Figure 6.2.1.2-23, and calculated and design peak pressure are compared in Table 6.2.1.2-9. Mass and energy release rates are provided in Table 6.2.1.2-1.Surge Line Subcompartment - Surge line subcompartments are 49 6.2.1.2.3.5 shown on Figure 6.2.1.2-5, with a node and junction diagram given on Figure 6.2.1.2-13. The model shown is for breaks in the pressurizer skirt area and in the vestibule. Node and junction parameters are provided in Tables 6.2.1.2-11 and 6.2.1.2-12 Curves of calculated and design pressures are provided on Figure 6.2.1.2-24 and calculated and design pressures are compared in Table 6.2.1.2-11. Mass and anergy release rates are provided in Table 6.2.1.2-1.6.2.1.2.3.6 Main Steam and Feedwater Line Subcompartments - The main steam and.feedwater line subcompartments are shown on Figure 6.2.1.2-6, with a node and junction diagram given on Figure 6.2.1.2-14. A double-ended hain steam line rupture (8.1 ft ) was assumed to occur in either Node 1 or 3, with the peak pressure occurring for the break in Node 3.A double-ended rupture (2.837 ft2) of the main feedwater line was assumed to 49 occur in either Node 5 or 7, with the peak pressure occurring for the break in Node 7.Node and junction parameters utilized in the analyses are given in Tables 6.2.1.2 13 and 6.2.1.2-14. Plots of calculated pressures are given on Figures 6.2-14 Amendment 49

ATTACHMENT STP FSAR nuas 6.2.1.2-25 and 6.2.1.2-26, while calculated and design values are compartd in Table 6.2.1.2-13. Mass and energy release rates are provided in Table s 6.2.1.2-1. The mass and energy release rates are calculated using RELAP 5 I analysis.j 6.2.1.2.3.7 Regenerative Heat Exchanger Subcompartment - A double-ended j j rupture of the CVCS letdown line is the limiting break in the regenerative ' ,heat exchanger subcompartment. A node and junction diagram is given on F gure 6.2.1.2-15. The nodal model initial conditions, control volumes, vent areas land corresponding flow coefficients and inertial terms are given in Tables ,p 6.2.1.2-15 and 6.2.1.2-16. The calculated subcompartuent pressure response is j.J d shown on Figure 6.2.1.2-27. Calculated and design pressures are compared in ,Table 6.2.1.2-15. The blowdown rate for the CVCS letdown line break is calcu- ^lated using ANSI 58.2, Appendix E2, Methodology (Reference 6.2.1.2-7) and applying that to a one dimensional Henry-Fauske model for saturated liquid.Mass and energy release rates are shown in Table 6.2.1.2-1. Plant operation iis assumed to be in the heat-up mode. Thebreakisassumedtooccuragthe' inlet to the regenerative heat exchanger. Thg break area is 0.0884 ft for each end of the double-ended break (0.1768 ft total area). There are no significant restrictions to forward flow 2 but the reverse flow is restricted i by the CVCS letdown orifices (0.00166 ft ) located immediately downstream of the regenerative heat exchanger. In addition, the reservoir of reverse flow is limited since high energy fluid conditions extend only to the letdown heat 4 exchanger.49 6.2.1.2.3.8 Radioactive Pipe Chase Subcompartment - A double-ended rup-

 ' ture of the CVCS letdown line is the limiting break in the radioactive pipe

' chase subcompartment. A node and junction diagram is illustrated on Figure 6.2.1.2-16. The flow model initial conditions, control volumes, intercom-partment flow paths, and corresponding flow coefficients The calculatedand inertial subcompartmentterms are listed in Tables 6.2.1.2-17 and 6.2.1.2-18. The calculated and design pressure response is shown on Figure 6.2.1.2-28.pressures are compared in Table 6.2.1.2-17. The blowdown rate for the CVCS letdown line break is calculated using ANSI 58.2, Appendix E2 methodology and applying that to a one dimensional Henry-Fauske model for saturated liquid.Mass and energy release rates are given in Table 6.2.1.2-1. Plant operation f is assumed to be in the heat-up mode. The break is agsumed to occur at the Containment penetration. The 2 break area is 0.0884 ft for each end of double-ended break (0.1768 ft total area). Asignifgcantrestrictionto forward flow is the CVCS letdown orifices (0.00166 ft ) located immediately downstream of the regenerative heat exchanger. For reverse flow, the letdown !s heat exchanger reduces the line temperature to 115"F and a pressure reducing valve, immediately downstream of the letdown heat exchanger, reduces the line pressure to 300 psig, therefore, the reservoir of high energy fluid downstream of the break is limited.6.2.1.2.3.9 RHR Valve Room Subcompartment - A double-ended rupture of the CVCS letdown line is the limiting break in the RHR 1A and RHR 1B valve rooms. Because the valve rooms are identical, a break was postulated onlyAin jRHR 1A valve room. The results are representative for both valve rooms.node and junction diagram is shown on Figure 6.2.1.2-17. The nodal model initial conditions, control volumes, vent areas, and corresponding flow

 ,I coefficientsandinertialtermsarelptedinTable 6.2.1.2-19-and 6.2.1.2-20.

6.2-15 x Amendment 49 {h0

ATTACHMENT ST HL-AE-lW STP FSAR PAGE IMOFIM The calculated subcompartment pressure response is shown on Figure 6.2.1.2-29. )Calculated and design pressures are compared in Table 6.2.1.2-19. The blowdown rate for the CVCS letdown line break is calculated using ANSI 58.2 Appendix E2 methodology and applying that to a one dimensional Henry-Fauske

~

model for saturated liquid. Mass and energy release rates are given in Table 6.2.1.2-1. Plant operation is assumed to be in the heat-up mode. The break The break area is assumed tg occur at the penetration of the valve room wall. 2 is 0.0884 ft for each end of the double-ended break (0.1768 ft total area).A significapt restriction to forward flow are the CVCS letdown orifices (0.00166 ft ) located i:mnediately downstream of the regenerative heat ex-changer. For reverse flow, the letdown heat exchanger reduces the line tem-perature to 115 F and the pressure reducing valve, immediately downstream of the letdown heat exchanger, reduces the line pressure to 300 psig, thereby limiting the reservoir of high energy fluid downstream of the break.

 /Jon-mEcho05M. deude.-ended got}loNMC. .2.1.3 M::t and Energy Releate Analyses For Postulated loss-of-Cool-ant Ac_cidents. The Containment System receives mass and energy releases fol-lowingapoptugted}3ruptureoftheReactorCoolantSystem(RCS). These re leases 18ont'IUiYe ffrough blowdown and post-blowdown. The release rates are 4he calculated for pipe failure at three distinct locations: (1) hot leg, (2) pump suction, and (3) cold leg. Because of the pressure in the RCS before the postulated rupture, mass and energyVflow rapidly from the RCS to the Contain-ment. As the water exits from the ~upture, a portion of it Vflashes into steam due to the pressure and temperature in the Containment as compared to the pressure and temperature of the RCS The blowdownVreducep t he pressure in the RCS, _ g g kl _ I During the reflood phase, these brea have the following different character-istics. For a cold leg pipe break, all of the fluid which leaves the core must vent through a SG (SG) and become/ superheated. However, relative to breaks at other locations, the core flooding rate (and therefore the rate of fluid leaving the core) for cold leg breaks 6eQ1ow because all the core vent ~

paths include the resistance of the reactor coolant pump. For a hot leg pipe ;O gbreak, the vent path resistance ao[relatively low, whichyresults in a high core flooding rate, but the majority of the fluid which exits the coreybypasse Y

 )

aus the SGs in venting to the Containment. The pump suction breakfcombin'iiis the )effects of the relatively high core flooding rate, as in the hot leg break, and SG heat addition, as in the cold leg break. As a result, the pump suction gk creakfyield/ the highest energy flowrates during the post-blowdown period.The spectrum of breaks analyzed includes the largest cold and hot leg breaks, reactorinletandout{etrespectively,andarangeofpumpsuctionbreaksfrom Because of the phenomena of reflood, as discussed the largest to 3.0 ft .above, the pump suction break location de[tte limiting case with the N double-ended pump suction break being the most limiting. This conclusion is 1 supported by Westinghouse Nuclear Energy System studies of smaller hot leg l breaks, which have been shown on similar plants to b ss._ severe than the g double-ended hot leg. Cold leg breaks, however, mee lower both in the blowdown peak and in the reflood pressure rise. Thus, an analysis of smaller l pump suction breaks is representative of the spectrum of the break sizes. ,l iThe 1.oss-of-Coolant Accident (IDCA) analysis calculations model is typically g divided into three phases: (1) blowdown, which includes the period from acci-dent occurrence (when the reactor is at steady-state, full-power operation) to 6.2 16 Amendment 49< - ... ..n . . . _ . -

3 ATTACHMENT ST HL-AE l?4tl PAGE lA50FIA'7 STP FSAR TABLE 6.2.1.1-1 CONTAINMENT DESICN ACCIDENTS Containment Design Parameter Postulated Accidents Analyzed

 ' Loss-of-Coolant Accidents (LOCA)

Peak Pressure /Temperature DEPSG, Min. SI, Min. CHRS (LOCA-1)DEPSG, Max. SI, Min. CHRS (LOCA-2)DEHL, Max. SI, Min. CHRS (LOCA-3)DECL, Max. SI, Min. CHRS (LOCA-4) 49 0.6 gEPSG, Max. SI Min. CHRS (LOCA-5) 3 ft PSS, Max. SI, Min. CHRS (LOCA-6)

 ,S,e,condary System Breaks (MSLB) 1.4 ft DER, Min. CHRS, 102% Power 1.4 ft DER, MSIV Fails, 102% Power 1.4 ft DER, 2 MFIV Feils, 102% Power 49 0.86 ft Split, Min. CHRS, 102% Power 0.86 ft 22 SP lit, MFIV Fails, 102% Power 0.86 f Split, MSIV Fails, 102% Power 1.4 ft DER, Min CHRS, 70% Power 1.4 ft DER, MFIV Fails, 70% Power 1.4 ft gER,MSIVFails,70% Power 0.908 ft 2Split, Min. CHRS, 70% Power 0.908 ft Split, MFIV Fails, 70% Power 0.908 t Split, MSIV Falls, 70% Power ) 1.4 ft DER, Min. CHRS, 30% Power 1.4 ft DER, MFIV Fails, 30% Power 1.4 ft gER,MSIVFails,30% Power 0.942 ft Split, 2 Min. CHRS, 30% Power j 0.942 t Split, MFIV Fails, 30% Power 0.942 t Split, MSIV Fails, 30% Power 1.4 ft DER, Min, CHRS, 0% Power 1.4 ft DER, MFIV Fails, 0% Power l 1.4 ft DER, MSIV Fails, 0% Power l 0.4 ft SP lit, Min. CHRS 0% Power 2

0.4 ft Split, MFIV Fails, 0% Power 0.4 ft Split, MSIV Fails, 0% Power Subcompartment-Loop Compartment i Peak Pressure c.-- -- ,.vn1.,s w-

 , _ _ _ --- giu, m f_

Mit.- kgsTOItiT.ER 50#Id.:l-[d6 Se--- 0;.w. 6ut vu viii ..ez:1: limited - a c4*eumfer enugi osea,. (.plit 2;d;1) - Look i 49

 ', h hac.orno/a.k2 frtygcnon l nc St; n 0;r.creto. outlet inv. le liwiiod area eteeumfm M =':

micial bre9%i&the 6 dol;r -

 ':.. - iaa . u, 6.2-62 Amendment 49

ATTACHMEN STP FSAR ST-HL AE-I?PAGE la&OFlar)TABLE 6.2.1.1-1 (Continued)Containment Design Parameter Postulated Accidents Analyzed Subcomparatment _-r

 ' --- - -m m uu Peak Pressure (Continued) "- rt: C: :12.. - Fump uuulcu uv.ml; If-#*=d zre cir. .f;;;r.;i. L... k (;plit , :N1) -

E; ek A..a -;^4 ...}R:: t: C:rlant ";;p cutlwi uuttle limiuca

 ;;;. ci . . f.....mioj break 15ditiv'iue =vu 1)

L;. k .6ca - 594 in Reactor Cavity 2 reak in 150 in Leg u Inspee ion Toroid

 " 14 Pressurizer Subcompartment Spray Line Break on Side of Pressurizer Surge Line Subcompartments

• Surgeline Break in Pressurizer Skirt area l49 Surge Line Break in Vestibule Steam Line Subcompartment Double ended MSL Break at Containment Vall Feedwater Line Subcompartment Double Ended FWL Break at Containment Vall Miscellaneous High Energy Line Subcompartment Letdown Line Break in Regenerative HX Compartment CVCS Line Break in Pipe Chase Compartment RHR Valve Room subcompartment 4g External Pressure Inadvertent Spray Activation 6.2-62a ' . Amendment 49

l ATTACHMENT ST-HL-AE-i? 6

 . . - PAGE ta9 0F I2ry.

Chapter 6 Tables & Figures o Replace data in Table 6.2.1.2-1A thru F "To be provided later". This will be replaced with data from breaks in the Pressurizer Surge Line and the SI Accumulator Injection Line when the calculation is finalized.o Delete Tables 6.2.1.2-2 6.2.1.2-3 6.2.1.2-4 Revise Tables 6.2.1.2-5 "To be provided later" o6.2.1.2-6 6.2.1.2-7 6.2.1.2-8 o Delete Figures 6.2.1.2-1 Sheets 1-7 6.2.1.2-2 6.2.1.2-10 Sheets 1-4 6.2.1.2-18 Sheets 1-12 6.2.1.2-19 Sheets.1-12 Revise Figures 6.2.1.2-20 Sheets 1-91 "To be provided later" o6.2.1.2-21 Sheets 1-40 6.2.1.2-22 Sheets 1-32 Note: Figures will be revised at finalization of subcompartment analyses.l 7603N:0288N/18!}}