ML20082Q862

Ahroud Safety Assessment

ML20082Q862
Person / Time
Site:	Vermont Yankee, Duane Arnold  File:NorthStar Vermont Yankee icon.png
Issue date:	04/18/1995
From:	Casillas J, Electona E, Ranganath S GENERAL ELECTRIC CO.
To:
Shared Package
ML20082Q851	List: ML20082Q862 ML20082Q876
References
GENE-523-A018-0, GENE-523-A018-0295, GENE-523-A18, GENE-523-A18-295, NUDOCS 9505010125
Download: ML20082Q862 (36)

Text

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Duane Amold ac.d Vermont Yankee Shroud Safety Assessment Apdl 18,1995 b

Prepared By:,_

• Electona, Eng)nect En6 ncering & 1.icensa Conetting Services i

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. oss L. Camillas, Ptiacipal Engineer enc neenns & 1.icensing Consching Services i

Approved By. ~h

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Sampath Rangasth, PhD, Proja:ts Mar.Eger En0 neering & Licensira Consubg Savices i

GE Nm&ar Energy San Jose, CA i

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P4 GE Nuclear Energy,_

_ GEM 6623-A018 o295 DISCLAIMER OF WARRANTIES AND LIMF0ATION OF LIABILITIES 1

This repott was prepned by the organiation named beltrw as an account of work sponsored by the Electdc Power Roscarclilmtitute, Inc. (EPRI). Neither EPRI, any manber of EPRI, the org uszation natned below, nor any peram actirig on behalfof any of thetn:

(a)

Makes any warranty or represemation whatsoever, empress or implied. 0) with respect to the use of any infonnadon, apparaius, method, process, or similar'itea disclosed in this report, including warranties of merchantabuity and fitness for a particular purpose, (ii) that such um does not inninge on or intefere with privately owned dahts, inchiding any party's intellectual proputy rights, or (iiQ that this repo<t is suitable to any particular user's circumgance, or (b)

Assumes responsibility for any damages or other liability whatsoever (meluding any consequcatial damages, even if EPRI, th organintion named below'or any representatin thereof has been advised of the possibility of such damages) rendting Bom your selection or use of this repmt or any infamation, appatatus, nwthod, pmcess, or aishilar item disclosed in this mport.

Organization that prepared this report GE Neetear Euergy Revision 0 ki a

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GENE-623 A018 0296 G E N uc!cer Engy__,, _,.,__,,_,__,., _,

Tab! efCoutcau

1. Introduction 1

2. 5kruud Condition Impxt on Plaut Ogrations 2

14

3. Shroud Condition hepact ce AccWents,

4. Summary and Condusions.,

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GENE-523.A0184295 L INTRODUCTION Generic 14tter (GL) 9443 requires BWR uhlities to provide a plant spec Ec safety assessrxnt supporting continued operation pending a complete inspection of their core shrouds. In addition, h aho requires that hceraces provide (i) a schedule for shroud inspection, (ii) drawirigs or shroud cordigurations, aod (iii) history of prior inspections.

Additionally, thE GL also requires licensees to sutnoit, as part of the plant specinc safety analysis, details of plant conditions (e.g. caibon contect, water chernistry, material type and form i.e., plate or forgig), ufsty ana!ysis considerms Mim Steam Line Break (MSLB) and Recirculation Line Dseak (RLB), and assesrecat of t!e plant respodse i.e., Control Rod i

Insertion and ECCS injection. This report prosh the safety analysis response to GL9443 for the Duane Arnold and the Vamont Yaakee plarus.

SpeciBrally, the rcport includes:

0)

Evaluation on the impact of postidated shroud through wall thickuess ciacking on nomul pix.it operations and timiting abnormat operational occuriexes (Section 2).

1 (B)

Safety enluation of the shroud assunsng through wall thickness cracking on Design Basis Ausdenta. SpardeaWy, it provides information on the regome to a postulated main stearn line and recirculation use break and an assessment of the operahdity of the plant safety features (Sectico 3).

(iii)

Overau condusion wacercung plarn safety.

The analyses and evahmtions describcd here are plmt unique and were based on specaSc

-l BWR/4 characteristics for both the Duane Arnold and Vermont Yankee plants.

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GENE 523 A018-o2s4

2. SHROUD CONDITION IMPACE ON PLANT OPERATIONS This report contains the results of the plant specdic evaluation on the impact of core shroud cracka. His evahisthat conwrvatively assumes that the applied loads are sufEcient to cause shroud sepasation along any one particular wckl location. Figures 2-1 A and 2,-1B show the weld locatiods, alon8 with their desigannons for DA and VY plantss The actuat elevation of each weld, referenced to vessel zero, is, given im Table 2-1. In subsedion 2.1 are the resuhs of the shroud separation on the core characteristics during plant normal operations. In abssetion 2.2 are the calculated magnitudes of akroud displacement'during plam nostnal opoistions.

In subsection 2.3 is the evaluttion of the impact of postulated shroud displacement on abnormal operational ocean ence calculaisons i

2.1 Norwal Plant.Qpt.aligna if h is postidoted t!v,t the shroud may be sumcie aly craciced at any of the horizontal weld locations, such that an upwat d load may cause the upper portion of the shroud to lift, an anomalous core characteristic resulting from the flow through the gap will be detected. This ammaly will be the resuh of reduced moderation in the core due to either increased coolant temparatures or reduced coolant tiow. An micreased coolant temperatum will be the readt of flow escaping to the outside shroud region thicu6 shroud separation at locations above the h

fbel top guide (e.g. M1, H2, and possibly H3), where two phase coolant is presed. For rample, considering a 360 degree one quarter inch gap, the calculated leakage flow and restnting thermal power loss is as shown in Table 2-2.

This impact on cora power is significant compared against the normal instrumentation power uncertainty of 1 to 2%. A decreased coolant flow will be the resh of flow escaping to the outside shroud region through shroud separation at locations below the fuel top guide (e.g. ID through H7), where subcooled coolant is presom. Por example, cons.daing a 360 degree one quarter "mch gap below the core support plate (e 3. H6B ami H7), the cQmtsted leakage flow and resulting themal pouar less is as shown in Table 2-3. This impact on core power is also sig"=*

compared against the noemal histrumentation power uncatainty of 1 to 2% These magnitudo Redelen 0 2

Wn Vb-so uc~20.,i inacc n wnw A o < x rna hv. uiuo m oot P.07 APR 88 '93 OG:46ft1 GE KEM Nr GENE 423 Aols4295 GE Nuch Emn of power anomalies will be detected and w)J1 lead the reactor operators to perform a norrnal plan +. shutdown. Also prescm wiu be other signiscaru abnorrnal core monitoring indications, such as lower measured cose iguppet plate prea. aire Merence (DP) vs. core Sow, higher recircuktion flow temperatures vs core power, and anatic local power measurements.

Analogous sittarious have previously been observed in BWRs. In 1984, a plant began startup with shrood head bolts impropnly engagd readting in bypass fkyw paths simular to those that would result from ttuough-wall cracking of the shroud. A similar l

stuation also occuned at a Merent phnt in 1991 In both casos, anomalies such as those

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described above were detected and the operators shut the plant down.

Table 2-1: Hothcotal Wekt Identification P.ganettst.d Varssent Yankeg Waldtil (4rtgijg.gjtyss, EggatloaAwam YeAsel 74rp.[{g)

NasMiZero fini i

HI 382.61 388.0d H2 347.02 354.75 H3 344.78 352.75 H4 103.24 256.84 H5 218 '14 184.56 H6A*

176 00 N/A H6B*

I71.94 131.56 H7' 101.88 108.50

• These welds are also krevn a4 M6, R7, and HS for VY.

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At 19 '55 06 5,FM CC NIJ.W ETMr-P.11 GE Nuc: sat Emrgy, GENE 623-A0184295 Table 24: Ifpper Shrend Leakage

.Q3_DJ!3 W Data Shauud Gap, lock or 025 Ruid Densky,4/fe 12 '.

13.1 Shroud DP. pal 70 4.2 Gap Mow, Mlb/hr 2..I 2.1 Gap How. % of Rated 4?

4.3 Entha8py Chsage, Blu/lb 44 3.7 Fewee Ch1nge,% of Rated

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-3.7 Table 2-3: Lower Sbsvud leakage It& itis),a W Data Slussd Gsp, inch 0.2$

0.25 51ond Density, th/ft*

4$ 9 46.1,

Shrosd DF, psi 10.I 22.9.

Cap Flow,Miha.r B1 8.3 Gap flow, % of Rated it,6 17.3 Pawer Ch aage, % of Rated

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Measured DP Less,pai

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2.2 HomJiL0picdima.Shrsild Salarsn9st With res[cca to a vertical shroud displac&ncut, the two key DP locations are the shroud head and the shroud support. The sh'oud head DP applies an upward led on wcld locations above the core utpport plate. % enod support D7 appliam an upward lead on weld locations bdow the com support plate. Also, the mass of the shroW components above each of the varkms we'dn will in0uence the magnine of the vertical dbpl=maar at each weld. The various component weights applicable ki the B'WR/4 plants are listed in Table 2-4.

The use of these componen weights muhs in the diftsent weight at each weld locanon u Sev/slon 0

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GENE 623.A018-0205 shown in Table 2-5.

Under notreal operating cordtions, at 100% power and maxbnum allowable uute flow, the DP across the shroud head is 7.0 pm and 4.2 psi, for DA and VY respectively. The conv=oonding DP acrim tae ahmud support is 30.1 pai nad 22.9 pai, (br DA and VY, respectiveh fri Table 7 6 are listed the caleutated shroud separation paramasers.

The DP required to catae sepantion is given, as well as de estimated core Bow which yields this DP The vertical s@arat' n is calculated for twu conditions, kleotified as equilibrium and e

mounmim sepa' rations hi Table 2-7 The equilibraura separation is derived such that the abroud leakade is sufficient to redxe the shroud DP to match the weight of the shroud components above the weki. The maximum separation is alicul?ted considermg a sudden, dynamic separation such that the uppe: ab:oud pottion ova. shoots the equilibrium sepamtion. The expresion used accout,ts for both the ohmud nies sad the Icalcage Sow through the orack, but conservatively Is;nores both the hydrodyrumic mau and the Md ding on the shrnud.

I Table 2 4: Key Shruud Weigbts

[ht91 DH8_1 Ytt134$1 Y M DBi AntM ArapM Yulide Yukee C1!Bittst,g 3:2 hpf2mtc..f Et2 Submeme(

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Shrd 11d & Sepaintors 65.2 58.7*

_ 44.9

.40.4*

Shrd Studs & Guide Rods 10.7 96 12.2 11.0 Core Spray 1.1

O 1.3 1.2 Top GJide 6.7 60 6.0 5.4 Stroud 54.5 49 1 83.8 75.4 Co i Nts 12.s 11.3 19.3 17.4

• The separatnr3 arn conwwively usuisJ to tu mienwged.

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<}gg.623.A018 0296 Tatde 2-5: Hoc r.ctitsJ Weld Weights D.yAnthracti ImmettI*=kee Etid1%

faewammt s m oonant Ed8bl.Udp.C Welebt(king}

HI 71.0 57 i H2 44 9 683 10 92.0 75.0 H4 M Tr 103.7 H5 i16.0 125h H6A 124 i N/A IEB in 7 145.6 H7 151.0 167.5 Inhie 2 6: 5hread Sepasaltom Farameters Efid MSEsttil n D.LGn.Drts H.5snatafan warem L191191 DLEd 14.afA3.!e.g DP. mai

% WRatsd HI 3.8 74 2.1 71 H2 4.2 77 2.5 78 H3 52 86 3.2 87 H4 3.6 90 4.4 103 HS 6.6 97 5.4 113 H6A 7.0

'00 N/A N/A H6B 8.4 53 6.5 53 H7 92 55 7.3 57 Revislan 0 5

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OENE 523.A014429s Table 2-7: Shauud Separation Results Egld Lacalhts AA,,Marknwa I)A Eguj[ijtai!3g DA Mauiqmm Demi IraatA@sJndiss smaintloadshet HI

'I O 2P 4.9*

H2

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H3 7.0 09 2.0 H4 70 07 1A HS 70 02 0.4 II6A 70 0

0 H6B 30.1 0.s *

• 0.5 *

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H7 30.1

0. 5 "

0.5"

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DERA fira(4.VasJachfa senma h taches HI 4.2 2.5

• 6.0*

H2 4.1 1.s*

4.0*

H3 4.2 09 1,9 H4 42 o

0 H5 42 te o

IM 22.9 0.5"

.0.5" H7 22.9

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' Separatio at Hi or H2 &es not afbt core georneuy.

• • This servica liraited by d4 cicuar.ce hetwean the cose aspport plate and the top edge of the costrol rod guide tubes.

'ne vertical separadoa for sa H1 weld loution will not be obstructed by any othcr vessel coniponents. Howe. var, Mmn itecifivesce n espected for the other weld locations by the Cos e Sprmy piping penetuting to the imier shroud region. His interfarence is not strong because the pipe couplimg allows some dh.plact: ment. For purposes o(this svabation, it is conservslively assusued that the obstruction dos out, afect the amoudt of separatica. For weld locations below the top guide (H3 to H6A fw DA) and (H3 to'H5 fbr VY), proper Revidon 0 9

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GGE-523 Ao1a4296 alignment of t!w cote is ssautedif the separation is less than 14 560 and 11.625 inches far DA and VY, respectively. As the top guide would need to 118 these distances to lose comacuith the top of the (ud chsuu:ett. These distances are the result of the applicable BWR/4 acometry, the fud chatvids enend 10 inch beyoral tlw top gmds, and the bottom of the top guide grid is 13.560 and 10.625 inchsa bdow the top portion foi DA and VY respectively.. For weld locations below the core support plate (H68 and H7), the separation is limited to approximately one halfis.h due to the 'mteifomace of the control rod, guide tubes with the I

core support plate. The soide tubes are stiged by the weight of the fbEl support casting and the fuel in limiting the tauxiatum Ett As abown above, the taitialindicatens vdll occur as a function of shroud DP and will be appment at cute Auws as low as 53% otrated. For ensmple, separshon at H6B weld will be apparent when the core tow' exceeds approxiwtely 53% of rated. Whereas s9paration at H6A weld for DA and H5 weld for VY wiU not be apparartt even at cost Sows of 100% of rated as the DP to cause separation is greater than the one calculated Ibr 100% dow. Also, once the conditions for sepamtion exist, a gap will develop such that significant flow will escape to the outer shroud region The cdculsted displamments shown above demonstrate that in the event of shrced separation due to a 360 degree througkwall shroud cock durhys rennd operation, the plant operatots will be ab!c to detoot the abnormal conditions and proceed with a noanal almt.devn.

2.3

&ntidgend_QperaborssLQrdal!Itflita Assuming there are no indications of sheoud leakage during nonnat operation, this section discusses the posible impact on anticipMai operational everns of the shroud condition. For this evalestion it is assuned ths: the du uud is sufEclently ernoked as anny of the borisordet wcld locations, mich that an increased upwstd load may cause the upper portion of the shroud to lift. Two types of est.ns are rerwwed, fmet those whleh are considered limitlag events for the BWR/4 plaat, and then those which ingse highest loedi on the shroud. He limiting events may not be altected s!gdkantly by the condition of the shroud, however since Revis.'on 0 10

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. 0ENE-623.A0184335 these events deteradne te mininun margh to fia:1 tbrmal limits (e.g. MCPR dnd LHGR) and vossol prc3Sure lindts, they have the greatest potential to impact safety lunits The highest throud Icad events may not lead to hmiting con.1itions, howmor thaty may dcicemine the unanimum whicud &placeaent and consequeraly have the grestost potential t s#ect the shroud fi:nctions.

A total of eight unique limiting anticipated cperstonal events were evaluated for the BWR/4 type plant. These events are norutally calculaed (Reference 2-1) to time if the opetaling liinits are impacted for plant changes, such as tnp setpoints and core chaiacteristics.

These events are the Foedwater Controller Failure (FWCF), the Tuttilne Trip witl%t Bypass (r1NBF), the Oonerator Lead Rejection wnhout Dypov (LRNDP), the Puel M Error (FLE), Control stod Withdrawal Error ()tWE), the Inndvertett Wigh Pressupe Coolant Inje6rion (HPCI), the f.oss of Feedwater Heating (LFWH), and the Main Steasn Isolation Valve closure with High Flux,Scratn (MSIW). The 6rst three (FWCF, TmBP, sad.

LRNBP) and the last (MSIVF) of those events are chractedzed by a rapid pressde increase, resulting in a core overpower cdulition. % ewd does t ot reatilt in an apprecip increase la core Bow or steam flow through the steam separators. Iberefore, no increase in shroud loads is predicted and shroud separation a not espected (eg. ainee nu separation' exists prlor i

to the event mad load is rwx increased during the evwd, the shroud is not expected ta lhil durhqg the event). Thus the results cf the r.utant analyses remain unchanAed aM no impact on safety fuuits exists The FLE and RWE evems are very localized,:and do obt result in significantly diferent shroud conditions stan during nocital operation. 'Iherefore,' no impact on safety limits exists for these events. The IFWH and HPCI events are charact'erized by a slow rise in core powa due to a temperature decreaso in the core inlet Bow. The resulting j

higher power resuhs in a small lacrease in siirotd load, wnich u sy lead to verticaliscparetloo.

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1 Howeva, as discusmi in Lbc previous acctum, this separation resuhs is a loss of dare power, thus the resultant thennai overpower wW be limited and bounded by the caw where the i

shroud does not separaf a. Tterofone, no impet on ufay linJts exist fbr these events Re vis k n 0 11 I -

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Three anticipated operatiorul everus are ioennfini which result in lughest shroud I

loads. These are the pressure regulator failure open, recirculation flow 008t1101 failure -

increasing to maximum 80w, spd inadvertent actustion of the Automatic DeprWestion System (ADS) Ihose eveuts are dixussed in detail in dw following sub sections.

2.3.1 Plt.mur,Brlphim Enduac.Opsa I

This postubted event evolves a failure in tic preuvre controls such thatlthe hubina controt valves and the tuibine bypass valves are opened as far as the maxit.w$ combined steam Bow limit allcws. For DA and VY. the bypus capacity.la 24% and t td.M of rated steam Scw regectively, and the addidohal tuitAne wnnel valve capacity is 5% of tated steam flow, Therefore, the maximum steam Sow duric.g the svue for DA is abod 129% of rated, and for VY in shont 215% of rated. The irx:rease in sir;ud loada due to this event is bounded by the Main Steam Line Break, and shroud liA h branded by the vabes listed in hon 3.1.

Because the fud remains properly aligrwi, core gownstry is maintained and succeinihl scram assured. The leads on the weld locations below the cara support plate 'are only increased by about 5% and do not readt in much diffbut consequences than at normal oper

. The consequences of this swat most the applicable Bocasing critais.

232 Racirenladon.Bowfontrol.Esilvrs This postulated event irmives a recirculation control faihns that utses at recirculation loops to increue to maxiemim 0ow in this type of case..(the upN pressure will change from a part-load condidon to the higWmaxhnum systesa dow capabilitj condition over a time period or shout 30 seconds The inerensed liAiris dbrees corrash to the maximurn allowable core Sow of aboet t02.5% of norsant maximneythus the lacrease in loads is approaimately 5%. However, shroud separation at die upper welds wl1 decrease crverpowers becounc the botter cool.us ma lions one power inorkse.

Alpo, shmud separasion at the middle and lower welds will Istst the event as cors power will carespond to l

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the core now w&h successfAy enters the core and increases reactivity N consequences of tids evoet meet the s1.p!! cable licensing criteria.

2.3.3 IntdrukatAEhatjen.d&DA t

Inadvateee actuatsoa of the ADS valves is am4fwt postulated event that increases load on the shroud. b madmurs steans Sow and the depressurization rate are bounded by main steamline break, causirs a short-term increase in steam Sow of 50% of rated steam flow (comervatiw.ly based on 4 ADS valves) for the ItWR/4 plant. The ' crease in the shroud DP m

resubg frein the opening of the ADS ' valves would occur over a period of about one second, sparading the efk-t of the clange in fond. The increase 'm shroud loads due to this event is bounded by the Main Steam Line Break, mi shroud lin is bounded by the vahies listed in section 3.1 Because the fuel rentains property aligned, core geometry is maintained and successful scram assured, and then: fore, tbv consequences of this event mest the applicshie licensing critoria.

2.4 REFERENCES

2-1 "GESTAk 11, General Ekctric StaWatd Apphcation for Reactor Fuer, NEDE-24011-P-A. (GE PROPRIETARY)

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' APP.-20-95 WED 9:38 VANKEE ATOM 10 X 6732 FAX NO. 0405941537 P 02 ypt is 'Y.; - 00:52Pl1 CL NtX.tFM EN.R.Y P.19 N

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3. SHROUD CONDrUON IMPACT ON DESIGN BASIS ACCIDENTS I

Two Design Basis Accidents.(DBAs) are ovahinted: a Main Stearn Line Break

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l (MSLB) and n Racirculation Line Break (RLB). M M51.B imide primary contalisment is the worst case bewzse it resuits in tt.s most sevc:c dcpsessurization. During h event, the l

reactor is rapidly depres<urimd as a remit of a pocoland double-ende'd break of the largest steamline. Tims a maxinwrn DP develops scoss the shroud as fhid Sow is drap from the core region toward the bresk. The MSLB impts: the largest lifting noeds on the diroud head and shroud support, and has the greater poteasid to defeat the shroud:funchoca The RLB de uut impose large peessure drops on the shroud, and in fact the shroud pressun: drop deenasas from its initial value. Howmor, this break reeutts in maximum final temperatures, and consequendy challenges the Emer6 ncy Core CoolMg System (NCCS) flawtions to a greater degroe. Additiondy, the RLB imposes lateral forces on the shroud.

l I

h potential for shroud displactment is incrtased if a seismic event is considered coincident vdth the DBA.

For this enluation the accident is considered withput seisatic efBxAs Gast, and time the addett effect of tin seistrde loads from a safe shutdown' earthquake (SSE)is exananed I

3.1 Main.SigyLLiggggk 1

i f

For this evaluation, the MSIR was cdculated for the BWR/4 plant using the TRACG modet. This model is a best-cstinute coupt.ter pogram for the analysis of Bdling Water Reectors (DWRs). IRACO is based on a nMJti-dunansaial two&dd model for Wie rutor thennal hydrouk:s and a thros.nucidonal reuun kinctka modd. M swoodd mcxW used fbr the thennalhydradica solvts the conservwinn nyi Anv 86r ream morneceum and enery abr both the gas and liquid phases. W ahts mal. hydist6c modd is a muki<hrnemional forntil%r the vesed conponent and a one-dianeadonal fonnalation for all other components. M conservation I

equations are closed through an edensive set of basic tnodels consisting ofegnstitutive so Telations rm shest ant heat tuesfer at the gastquid irterfacg as wd as at the waL 'his structure i

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is based on a modubir approar.h. The thermShydreAc model contains a sd of basic components.

such as pipes, valves, im hel clumels, and vcssa A.klitionally, TRACO camfaina a control syscon modet capable ofstruulatmg the nmp BWR corsoi symacs sud as the and wata le*>el contro8es. Reactor simuhti.ms are pionned ley uuuuuding s'usudal us$g the basic cornponents as buil&g blocks Any a mts of those components may be combined. The numbe i

of coniponcola, their interaction, as well as the deiait in endi WM are spectied thmush code input. TNerefoit. TRACG has the espatilhy of accuratety simulatics most BWR' phe Additional detads on the TRACG model are provided in Reference 3-1, and infonnation on the modef qu15fsationis docunwted in Rdercras 3-2.

i

& TRACG uols! prepared 1tr tis BWTU4 pkmt corsists of a Macar WWE d, WIdeh is dividad irdo eixtem seal levels a4xl four ro4ni sings TW nodaliration was seleced based on TRACG quali6 cation rond:s, as well as certain reticeamts necessary to shccinately pimitem the perfom ance of the BWR/4 p! mt to a MSI,B cvent.

I 1

The initial cor.ditions used for the MSLB calculation correspond to the nicst limiting I

rated core power and flow opctnting condition. Sevetal event characteristics.) wisch are nonnally consuivativdy ignored or simpitfled, were factored into this aluld Most of thes.e ehnetaristics are Ailitated by the TRACC mcdci capebilitics, and were specifically deterniined for the BWR/4 plant i

The irstial conditions and chare.cterisucs used ter the TRACG MSL DA and VY are listed in Table 3-1.

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Tabis M Nhroud safety Anahsis Conditions.

j Eevinis[alC93.ggit31 pyurgegag]d $ asis YY Bads Core Power I658 M Wit 1593 MWTh Core Flow 49 0 Mik1r 48.0 MRVbr Vasd Steam Flow 717 Wh'hr 6.43 MlWhr Dorme Preswre 1040 psia 1024 pain Turbine Preimre 9015 paa 962.5 pela

[

Feedwater Teotperatine 199 Deg F 346 DegP-Shroud Head DP l

'/.0 psi 4.2paa i

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Shroud Support DP 30.I psi 22.9 pd

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Norrnal WaterLevel 53o" above vessel zero 510' aboveEsel sem Steam LireDiameter 20 in 18la' Rocarcuhtion Line DianwNr 2 in 23 in l

Vessd Stears Line safe End Are f.767 sq ft 1.418 84 A l

Steam Line Flow Limiter Ama OA43 sq fl 0.416 sq 8 1

M51V Clomre, t' ne to fhH closese 5.5 we 5.5 sec I

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• 1te rwulta of this cchuladon ase a mwanuun shroud hsad DP of 15.5 psi br DA, an 10 *i psi for VV. Shome in Fidures 31 and 3 2 m de time histodes of these Ms. This calculation shawa the liling load ta last leu thu ths e woonds.

I The cdcufstej nM-="=

shroud support DP is 40A psi for DA, and 31.1 for VV. 'the rassiting vert the various weld elevaticia for these loads are given in Table 3-2.

As expeh,ted, these separations are higher than ibr normal opastion, enwpt for the lower w i

the acparadon is limited by the control rod guide tubes I.

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Figure 3-1 UA MSLB Shroud Head Prmure Diseience l

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As berme, some interference is expectix! ibi :he weld locations H2 thr H6A fbr DA and H2 through H5 for W by the forc Spray paping penetrating to the region. This 'mteafemce is not strong bcome the pipe coupling aDown some d 12 cement.

For pwposes of this evaluation, it is ennwvativdy assumed that tho affect the attount ofseparation. For mid location below the top snide (H3 to Hi and (H3 to H5 for VY), proper alisenant of the core is assured if the separation is less thaa 14.56 and 11.625 inchos for DA and VY respectively, as the top guide would need to lift l

thew distences to lese contact with the top of the h:ei channels. For.the BW10 4 plant the mamanum lie of tie top guide is calculated to be 9 5 inctes for DA, and 10.1 *nWsea fb and thus core alignmem in aaswod.

His staxnmsm htt is calculated using a dynamic expression for the shrotal motion, the inputs for ux, calculation are t!as: TRACO calettated boundary cordrions for the idet and outht flow 6om tie diroud. This calculatio n accounts for both the hydsod nsmic aiass and the fMd drag on the shroud. For.' weld locations below 3

the core support ptate (H6B and H7), the scparation is limited to appr**ly o ne halfinch due to the interfererice of the control rod g'alde tuh 111e guide tubes are as by the wedgle of the ibal support Castmg and tite thd in haating the maximum lift to oMy one half imh Iberefore, th'c lia impwt on the cora imeanals for a MSLB is the same as for normal operation.

A seismic event coincident with a MSLB may resuit is additional ventical and/or latersi displacement, of the upper shroud portion which incomes detached at a weld locados, with respect to the lower shtotd sMtion. Additiona'Jy, aller a short duration lift, onco the upper shroud portion agaea rests on the towar skovd portion, the lateral seismic loads apply a shroud tipping anonmot.

Calculations performed to simuiste the pwible shroud displacement skse to the sdsmic loods, for the lining portion of tbs event, resilt la less than 1.0 iwh additiomal vertical dispiscement. Vns vertical contribution to the owrall dispbcerneut is conservative sloce the

nagnitude of any displacemont is pri:natily deterntned by the upward thermal hydrimlic loads.

Aity vatical ditplacument h ordy temporary, a the upper sluoud portion la expectei to retum i

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' less than 1.0 inch lateral (e.g. Itorizontal) displacemett m

This displacacientit perrnenent as the upper shroud may not be peifectly =Iigned when it retums to =W on These additional diglacementa, from thosc tabulated in Table 3-2, are the s esult of the motion of tbo lower abroud portion during the sek.nse oeut.

For t$e pottion of the ' event when the upper ahroud rests on ! he lowe t

shroud, the lateral scismic loads apply a tipping necwm on the upper shroad. The irgal.

se seismic tipping moinent is experienced at the He.A and H5 wdd locations.; How restoring momeca of the shroud weigit is gi eater, no tippma or rotation will occur, The added displacement calculsted for itv: senade event does not rb in a add &nal reactor components heirg affected.

I 2

1 Table A2 M51.9 Shmud Separation i

.EtM DAMaxi_ar.s_a DAJKaxisias yy Maximum IX.Esakstum LKW.itA DIaRd Sena#3tiOA isthet RF4 haat'-.,jggggg HI 15.8 15.2*

10.1 r

1 5.4

• H2 15.8 13,3-10.1 1 4.3*

H3 15.8 9.5 10.1 i

to.1 H4 15 8 83 10.1 16 H5 15.8 63 10.1 3.6 i

H6A 15.8 5.5 N/A N/A H6B 40.4 0.S**

31.1 0.5" H7 40.4 0.5"*

31.1 Oh"

• Separation at H1 or H2 does not affem core geoantry.

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" This separation limited by the clearance bulwcon the com mpport plate and the the controlred guide tubes

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For the RI.B, the dt&rential preuuic neross the shroud decreases th a tlas init.ial value as the core flow is reduced due to the break, ispwnr 1 fhrees are reduced, sad' thus there J

is no sigri6 cant th eat to core shroud insegrity Any initial shrrud separaupn alona a 1

particular wcld location will be limited to a tig!n cracir, sice any signideant sepsfa to the accident, would be detected duricg noimal operation as discussed,in sectionj2.1.

Lateral forces far wcld locations in the begfrukag of a MB are lairge ac c forces of short duration, less ilso ten adllisecoeds, followed by smaller blowdown ibroes!for several seconds. Hortacotai nxWion is not expceted because of the resiatance atthe i star crack sunface to honzontal motion wiinout liAing. If somcium liAing occurs prior to thelwM* it will be detected during nonnd operation as dismssed in section 2.1. Tipping 0.e.l rotation) is not expected from the acoustic loading as it is of very short duratiod. The adoustic load cafuttated for tv shroud is cotsarvatively applied to the rotation.

The % RLB Blowdown force was calcolated couwering bods, tile DA and VY plant charactedstics. The resulting lateral simar forear and tipping noments, as a ibaction of ahroud elOn, are shown in Figures 3 3 through 3-6 for both plarn This ELB Blowdown calcdlation was performed by scaling the resJts of a spedal RIB TRACG cakulattom to the specillo DA and i

VY dimesions. This special caleelsfian rersexated a typical Iet pumji BWR pigat and the outer region of the reactor vcssel was subdivided &om the standard 15 aa$tions into approrinmiety 21c sectioca.

This detad was tweded to accurately cakaalate h transient pressure distribadon around the shroud dosing the nasty portion of the RLB. Tis scallag proces considers the racirculation nmAn area md its location whh respect to the inhroud, the vessel and shtoud dinKsulons, and thejet mmip geoinary. Based on these results, the limitias tipping moment is for the IMA and M5 locanons for DA sad VY respee'tively, as{shown Tabics 3 3 and 3-4. Any tipping for weld locatioca towar than H6A am1 H5 res@vely is fltnited by trie guide tube interference, and Ibr highar weld locations the momedt is much lower. Also included on Tables 3-3 and 3 4 la the restodog maament resulting haar the shroud t

weight. A coolparuwm of the DLB tippr* g momet 4 salon the sostaring'amoement shows that Swlsten 0 21 i

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tbc reatustrig moniatt is rauch larga, and in the absence of upward fierces, will fnaintain th shroud in phe

,% y upward hcs nie the resuk of the core Gow, which wtil decrease as a result of abe RLB to a very inw mlue in fosa thuu r.me second 'Iherefore, shd tippinals very udkely because the uppet welds apenance a staall RLB moment, and tho lowe welds have large restoring montent. However, muy postuinted tipping at the limiting E 6A and H5 welds for DA and VY ienpoctively is rest icted W thejet pump riser brke dearakce with the shroud to appronnatdy one moh. The diiradou of this tipping is less than 1 secodd and by rewwting moment the saroud raums to its originai posidan. 'Iberefers, shroud displacement occura, the RLB imults are unchanged.

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Figure 34 VY RLB Shroud Moment vs Elevation l

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• Lateral mention 11adred by theis pung rim braces i

" The tamal rotations! movent:nn of these we!d locations is limitad by the vertical displacemaa of the tore scpport plate againd the control rod guide t For the RIE sinttikaneous with a seignic event eddirland vertical and let'eral wiR exist. As diwunaed ahose, the forces in the txxy region resultag frorn atmost instantaneots downwani pull on the shinud alwl wiE prevent, v dirplaccmoru akeg a wekilocation. The lateral seismis loada, combine Revisfon 0 27 8

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APR-19-95 TUE 13:55 YANKEE ATOMIC X 6732 inX N0, 0405941537 P.11 P,33 ppa m 95 m:53PM GE TE U F D G 0Y GE Nuclear Energy _,,,__,,

GENE 423.A015.o295 bimvdown loads may result in a snaill mociantvf yping This tipping will be licited by the t

jet pump siser braces clearance with the diroud Howewr, the restoring moy of the ahmud weght will prevent permar. cot displaceowna The morantary displacenuat calculated for the postulated RLB event insects the jet pump risar braces, however the results of a RIB renusin unchangd. Current Losa,of Coolart i

Accident analyses are uoofected, and limitics calmt.tted results are applicabic.

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thm.5Afcty.ltitts This section a4heseca the impact of an tedetected cracic as it a5ects the key safety fadori of control rod inaestability, coolable gametry, ECCS pesformance, and SILS effectiveness. The primary factor in this detconication is the expecded movenient of the shroud durirg the poeru!ged evem The basis for the shroud roovement is as dmmaated in sections 2,3.1 and 3.2.

I Control rod insenab2ity will be asural is trie fuel remalna properiy arrarged in the corn, and the auide tubos and shroud remain aEgned No pennanent misalisnment; occurs for either the MSLB or the RLB. withect sounc ch ts, and tin:s the control rod motaba will not be afected. With scismic effects, the ataxinnan mislignment is only 1.0 inch at thaj top guide or the core support plate. The control rod tuotion wiB ret be a5ected by this small initaligrcmit. Control Rod Insertion tests conducted by GE show that control r not a&ccted up to at least a displacement of 4.8 i:iches at the top guide, or 1.5 inches at the core support plate. For thcsc BWR/4 evaluadous, suge the top guide remamm in ccdact with the fbd ehmuu.ls, for all cooditicim, the fuel asentilles will ramala propedy arranged in the core. Additione!!y. a uknie naciBetary station wi t assure tl+nt the enntrri rod inution in not t

irnpeded, as shroud alignment is periodically achieved. Any knpact on' Scram thries is not critical because the reactor per is primarih reduced by the void formation la the cpre.

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A coolable ge.ometry wi!! be mahMahid if the. fuel and vessel internal he normal Sow of coolant to the foot l'or weet pipe bicek locations above the Top of Active Fuel (TAV), short and long :erm cockng in accoruplished by SCC 8 is e muywhere in the reactor venet in a flow amount equal to the neamins rate of the core Por vessel pipe break locations below the TAF, abon terra cooling a accomplised by ECCS the shroud, over the kd ud elsewhere. l.ong terna cooling is anenmplished by maintaining I

the levet inside the shroud, to tNejet pump suction (evel, by ECCS irisction 'tha efore for a MSLB event, ECCS injectico into the vessc1 is not precluded due to a degrah condition. Also, for a RLR event, ECCS iqiestien insle the shroud is not produd ud due to a dogmbd shroud condition. The af5setM Crire Spray peins dwing.:a postuh ted MSLB abroud lift does not innpact the consequences of a MSI.B. as the Ami remialns covei ed, and the Core Spray is still able to provide coolant into the vnsel. For the RLB, the Co re Sprayin essential to core coolms, however, for this event the Coos Spray water delivery ha xion is not affected by use to one ar)d a half inch tippite of the shroud because of the mechanical flexibilityin the core spray firms.

6 Proper ECCS performance is achieved if the ECCS coolant is available when and whore needed. Tbs BCCS providing coolut for this plant type cocaists EfCore Sp rays (CS),

Low Pressure Coobut bdection-(LPU), nr.d EUgb pn:ssurv Care I4ioction (HPC D systems.

'Ita CS and LPCIinject through cpper shroud pemrations and jet purnps. The H lato tiie outer shroud repon through the feedwater Ene. The impact of a postufale failure on the consequences ofM5LB selinsed to same BCCS injection 'degradstic

e. Thisis the rer.nlt of CS line intareranze by a liAins shroudHowever, the arnount ofICC5 flow required under a MSLB cvtst is mhsinM sad thne no inpct in ovenallBCCS pd efbrmance exists The impact of a posrutsted shtowd faiture on a R1R is liralted so some additional ECCS flow needed to reaintain a two thirds water inct in the corn, Tids in tho rundt at coo! ant leakage through lower welds cracks.

this lealutge is minimal compared ty a single-purnp Eccs capacityibe the RIR pump or the CS pump, and the normal overflow throu thejet pumps, and thr:rcKa e no impact in overmil ECCS potfotmasce exist 4 Revklon 0 29

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... _. _. SENE-a23-Ao14-0295 l

FKoctivetna of the SLCS is niaitualned if iniedion and cnxing of borog is possible.

4 The SLCS in used to shut down the reactor if unuel rods are mot adoquatcly iderted (for a RLD, the natme of die acdtem ruay-limh the elkenvuness of SLCS by draining the boroa firorn the core). The SLCS injects thro. ugh lower planan lines fbr these, plaa separation above TAF, shroud ancka and dae!acement wiu not. prevent SIES from performing its intended function for cither the RLB or MSLB. For a shfoud sepasation below TAF, ahmud cracks and disptscerneat do not affbet SIES performanos for the R9LB, and SLCS afectrvence for the RLB is also not anscted as some boros may be drained through the vessel breakw;th or without shroud cracks.

3A Sa(sy Assessmen.guanterr

\\

%e applied loads to tim shroud are low, such that even with sigalacant er present the shroud design margins are maintainect Also, the safety consequeq af shroud fhilure fbr t

the BWR/4 plaat are not pgn%at.

'th: postolated shroud failuro during ooanal operc. tion win be readily detaeed through normal musirenwnts of'prant perfbnmece. The detected soomah willlead to a nxmal plant shutdown 1

The postutsted almoud fhilure during alawnnal are accideot event cWM a will not lead to greater chaknee to both transicci sduy limita, and accident consequences. l I

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ResisAm 0 30

APR-19-95 TUE 13:57 YANKCE (EPM 10 X 8732 FAX NO, 0405941537 P.14

p. tu 'W iz9:32PM GC fic. EAR DEWS P.3G g

GE Mar Ence r/-- -

GEffE-5 n A01'5420e u

REFrAENCES l

I I

31 "TRACG Modd Dmeriptian - Ucwnity Topk:a! Ryurt", NFDC-3217dP, Phar (GE PROPRJETARY)

}

32 "TRACG QuaFaion - Ucass Topic 4 Rgxzf, NEDE-321MP, Reviion 1, June 1993. (dEPROIRETARY) t i

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i J

I Roulslan a 31

nt'r.-13 vo IUa 14

• b /L Yhhu.h hiOfijU~A 6 62 fhX h0. 0406941537

~

P.15 mit im '95 39:23M GC iOCLEm ClOGY GE Nuclear Fnergy ogNg.ast Aot$-02ss

4.

SUMMARY

AND CONCLUS'ONS I

The key conclusions toai the'Genenc t.eiict 94 05 maiuation perRxmed for jhe BWR/4 shroud are sormwined hue

t. The minimum reghed ligament from a mudars2 nurg,in viewp utis 5 to 10% of the wall W' hile the initid depth cgaret be definitely established, the pgeneric evah:ation i

perfb ased for B%Ts indicates that the cunmit crack depth could be up to 30% of the wall thickness. Even if this conse: vat.ive inniting ahm is assumed, enhea* muotural margin is mairsained fbr the currow cycle sure the predicted crack smwth for the cauwd cydeis small.

i

2. The safety ansessmais for the ma'n suwe ths break combined 'with a seisado event thows that the liA of the siiroud wil not meeed the depth of the top guide ihr the postulated steam bne tnsk. Thus, the atwlity to scramis mamtaned. De halloads for the rocinmlation line bruak somWwd with a saide event wem aise detennised and fomed in cause at worw. timited dpping and crack opming followed by subsequent return tothe original geometry.

In wrnmary, the story analysis construs & abnic/ su opeiste saMy with)imiting crack depth samtretions e

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AMska 0 32

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