ML20151F106

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Forwards Response to 880615 Addl Questions Re Seismic Design,Impact of Dropped Fuel Assemblies & Safety Procedures.Calculations of Effects of Dropped Fuel Assemblies Will Be Submitted by 880812
ML20151F106
Person / Time
Site: Vogtle Southern Nuclear icon.png
Issue date: 07/21/1988
From: Bailey J
GEORGIA POWER CO.
To:
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
GN-1475, NUDOCS 8807260348
Download: ML20151F106 (49)
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{{#Wiki_filter:.(t a Gior0ia Powtr Company Route 2, Box 299A Waynesboro, Georgia 30830 Telephone 404 554 9961 404 724 8114 Southern Company Services,Inc. Post Office Box 2625 Birmingham, Alabama 35202 Vogtle PioJect Telephone 205 8704011 July 21, 1988 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk File: X7BC35 Washington, D. C. 20555 Log: GN-1475' PLANT V0GTLE - UNIT 2 NRC DOCKET NUMBER 50-425 CONSTRUCTION PERMIT NUMBER CPPR-109 SPENT FUEL RACKS Gentlemen: NRC letter dated June 15, 1988, transmitted additional questions on the Unit 2 Spent Fuel Racks. These questions requested information concerning seismic design, the impact of dropped fuel assemblies and safety procedures. Georgia Power Company's (GPC) responses to your questions are enclosed. Your question number IV requested the calculations for the effects of dropped fuel assemblies that were reported in Section 12 of the Seismic Design Summary Report, contained in our letter of December 23, 1987. Those analyses were performed using the combined weight of a fuel assembly and its associated control rod assembly. However, the FSAR has cumitted to an analysis that also includes the additional weight of the spent fuel handling tool. Therefore, these analyses are being redone for the higher weight and will be submitted by August 12, 1988. If your staff requires further information, please do not hesitate to contact In order to expedite the continuing review of the Unit 2 Spent Fuel me. Racks, GPC will, if necessary, meet with the NRC staff to explain our responses or to provide additional information so that your review can be completed on schedule. Sincerely, t J. A. Bailey Project Licensing Manager ve/ 8807260348 880721 PDR ADOCK 05000425 A PDC y

'h G. .U.-S. Nuclear Regulatory Commission - Files X7DC35 July 21, 1988. Log: ' GN-1475 .Page 2 3 JAB /1g Enclosure 'xc: NP.C Regional Administrator NRC Resident Inspector P. D. Rice J. P. Kane R. A. Thomas-B.,W. Churchill, Esquire -J. B. Hopkins (2) G. Bockhold, Jr. R. J. Goddard', Esquire R. W. McManus-Vogtle Project File-1576V-1 4 f

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-eJ L s s I.1 SPENT FUEL POOL ANALYSIS Proside sketches and/or drawings of the pool ~ showing elevations,. base mat and pool wall thicknesses, water levels, and safety-related compo'ents (such as piping) in the pool, and their clearances from the racks.

RESPONSE

The normal water level in the spent fuel pool is 218'-6" as shown in FSAR Figure 1. 2. 2-11. .The pool basemat, north, south, east, and west concrete. walls are 6, 6-1/2, 6, 5, and 6 feet-thick, respectively. These are shown as FSAR Figures ~1.2.2-7, 1.2.2-8, and 1.2.2-12. These FSAR figures are attached for convenience. The bottom of the gates to the fuel transfer' canal and cask loading pit are at elevation 192'-7-1/2". All other obstructions are approximately 6 feet or more above the top of the spent fuel. racks (approximate elevation 193'-10"). The location of the gates and other obstructions is shown on Figure I-1, Pool Layout, and Figure I-2,-Fuel Rack Interference Study. 1 I.1-1

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5 g r n 3 q % 4 m, <-,- e 18t4 Ortrl 4 g 4 r twetort. s*.s* ooc aeassany.n, D'auor-17ttysAnarr-w f g,M ri ~;," "o SPENT FUEL POOL ' " " * "* *'~' ** 709 of LlaER Ptart Ell.W.oMr* <am-..a.o -,... .-? 2 [c/Norn warra ax.nrs t rvercat on esacas) l 7* l ODC AG116Aaw-f -l R-E*. 4- % * -l SOL EL E13'-o* ? EMY6 tort - Jo~n IS' U'I ~O*' W 4 *- G ~ ~. ~~ .}' MEFERENtr M*:=*aans b o, e%y ~~ 5 FL/D

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  • n.ru'-s-t ruel nanoss A

A . -.~.u< Scagt I-Z l l FUEL RACK INTERFERENCE STUDY vm?voo, nuct:4R rumv 9510 SK-RFE-001 ~ ""M

DATE, OR.

(Met gessy y C$' Df TR CORP. LOS ANE3 m i j

I.2 Provide information on how the additional weight.of high-density racks (HDRs) and impacts on floor and walls under the postulated seismic events are incorporated in the design of the pool-structure. Provide information related.to pool structure seismic responses due to the proposed reracking, controlling load combinations and stresses at critical' structural sections.

RESPONSE

The current licensed configuration is for two HDRs in the Unit 1 -pool. .The fuel racks'for the Unit 2 pool have also been designed as high-density free-standing racks and are located in the Unit 2 pool to preclude impact on the walls or other racks during a seismic event. Pool Structure Seismic Response due to the Proposed a. Reracking The existing calculations for generation of floor response spectra are based on a weight of.7,860 kips for racks and fuel elements, considering.both the Unit 1 and the Unit 2 spent fuel pools. The weight of two racks (together with the fuel elements) previously installed in Unit 1 is 506 kips. The weight of HDRs (including fuel elements) provided for Unit 2 is 4,146 kips. The total weight of racks and' fuel elements in the Unit 1 pool and the Unit 2 pool is 506 + 4,146 = 4,652 kips. This is less than the 7,860 kips' weight considered in the seismic response evaluation. Therefore, the proposed reracking has no impact on the design floor response spectra. b. Effects on Structural Design The calculations that address the design of the fuel handling building are based on an analysis which utilizes a three-dimensional finite-element computer model. The computer model includes the basemat, fuel-pool walls, and other major structural elements. his T analysis assumed high-density fuel racks with a total weight of 3,930 kips for each pool, including fuel. The calculation identifies the controlling loading combination to be U = 1.4D + 1.7L + 1.9E (where D is dead load, L is live load, and E is OBE seismic load) and demonstrates that the flexural capacity of the reinforced concrete basemat and fuel pool walls exceeds the worst-case resultant section forces. This is accomplished by the use of interaction diagrams which represent the capacity of the basemat and fuel pool wall I.2-1

sections. The worst-casu section forces of the basemat and walls fall within the interaction diagrams. The calculation also determined that the. worst-case out-of-plane shear force in the basemat 1s well below the ~ allowable shear' capacity. The weight of the Unit 2 HDRs including spent fuel is 4,146 kips. This represents only a 5.5% increase over the weight of the fuel and racks considered for the Unit 2 pool in the existing analysis, which would cause an increase of no greater than 2% in the resultant section forces due to the total load from the structure (weight of basemat, rack, fuel, water, etc.). .This increase of no greater _than 2% is more than compensated by the existing design margin in the flexural and shear capacity of the: Unit fuel-pool basemat and walls. For example, the flexural utilization factor in the fuel pool basemat is only 83%, indicating that sufficient design margin. exists to accommodate the increase in loading. The allowable concrete bearing capacity under the fuel rack support feet was determined based on the requirements of ACI 318-71. The bearing pads under the rack support feet were sized such that the allowable concrete stress would not be exceeded. This assures not only that localized crushing of the' concrete under the support feet bearing pads will not occur but also that the reactions from the impact of the rack loads will be transferred through the liner plate to the concrete structure below without damage to the pool liner. i ) I.2-2 i

I.3 Provide information on the effects of HDR weight on the soil bearing capacity and liquefaction potential. -RESPONSE: a. Bearing Pressure The' gross static loading pressure applied by-the fuel handling building is 8.1 KSF as shown in Table 2.5.4-12 of the FSAR. Because the weights of the high-density fuel racks and fuel. elements are enveloped by the rack 'and fuel element loads assumed for the bearing capacity analysis, tha factors of safety shown in the table remain applicable. Assumed loads are discussed in the response to part a of question I.2. FSAR Table-2.5.4-12 is attached for convenience, b. Liquefaction Potential. As noted above, the loading pressures in FSAR Table 2.5.4-12 remainthe loading valid with the addition of the HDRs and-fuel elements. Additionally, even if there were a slight change in these pressure loadings, the liquefaction analysis would remain unaffected since the structure bearing pressure is not a controlling parameter in the VEGP liquefaction analysis. Therefore, the existing liquefaction analysis, which demonstrates that liquefaction will not occur under'SSE conditions, remains applicable. Furthermore, as indicated in letter GN-673, dated August 5, 1985, Dr. H. Bolton Seed has_ concluded, based on his evaluation of standard penetration test results, that there is no possibility of liquefaction occurring in this soil for any level of ground acceleration that-may develop at the Vogtle site and that liquefaction is simply not a credible mode of f ailure for the Vogtle - fill. I.3-1

TABLE 2.5.4-12 (SHEET 1 OF 2)

SUMMARY

OF RESULTS OF BEARING CAPACITY ANALYSIS I Factor or Sa rety. rI or Urtemate Net Beareng lop or Capacity i Net fout<lation/ Sta t e c or Dynamic El or 8 t oad i ng Pressiere ustamate*8 Allowabic"I Computed Computed Bottom Approw. _ J oadgq_Preanu re Not Net Beareng StateO Dynamic or Mat Cross Net (4 Gross Net Bea r s ng Capagegy_ _ factor factor supporting Foundatsor S t.*e stato statec S t ruc tie re luaina f r M_ _..Lrt}_ (k/ rt_g] Dyn.e m Dynam c C.spacety Statoc Dyn.emec or or [kf rt ] [kf r_tb' c Lkyg2J (k/r3 j i k/ r t# 1 Lk/ft') Saregy par ty 2 r Dieses generator Backrill 220/211 11Sx93 3.8 2.7 13.6 12.5 60.9 20.3 30.5 22.6 4.9 e buiIding Turbine Duelding Backre s e 195/186 184=604 3.6 - 0. '> 8.7 s.6 56.5 18.8 23.3 ve r y'88 12.3 e ro e gh "J Control buildeng Backrete' tou/1/3 ',75 = l t,9 4.3 -1.1 13. ta 7.5

  • >l.8 19.3 20.9 ve r> ge:

7, c, M< he9h O Containment Backritt 169/158.5 0: 1%. S 8.4 1.0 20.9 13.S 61.7 20.6 30.9 61.7 86 s buisding 6 9 12] fuct building Sach ri l l 160/IM 198x15 8.1 o.1 ?3.84 15.8s 68. 0 21.3 32.0 640 4.2 4 Nuclear service Mars 13//128 D.100 M8 -2.4 34.? 23.1

61. /

20.6 30.9 ve ry'88 2.7 cool ng water b tower fe gh AuxeIeary Marl 119.2S/199.?S a40 130 119. 7 -3,1 PM./ I 'j., 63 / ?1.? 33.9 Ve e y (88 (s. ? s bas e id e n9 1 h 9fe Ntclear hervice (Sac k r a i l 204.'j/195.5 140=33 3.2 u.6 17.6 In.u 133.S h is. ", 66.6 ??/. *> 13. is cooling water vaive house Aunifiery reed-Bac k r e l l 2tS/212 86=40 1.6 es.6 3.? .'. ?

97. ',

30.M water pumphouse r 6. 3 ISs, ) is2,o e e Condensate fl.eck r e t s ??u/212 11',.63 ' 3.I ?.1 'f. 3 is. 3 I t *,. 3 3 n, r4 storage tank ' /. / S's. 9 26.8 Dieses ruce oit Itac k r o s s 2tt.5/?u9.', 1 1 a.. ?(, 1.1 p.sa ?. '> 1.? 81.9 2 f. 3 storage tank s t.o Pu's. 8 68.3 e poemphouse (4 Q Q 333 O. O. O. W *%8 m M p t/B (13 N\\\\ g co co m aa b m

TABLE 2.5.4-12 (SIIEET 2 OF 2) ractor or Sarety = Ultimate Net Bea ring Et or top or Capacity - Net foundation / Sta tic or Dynamic toadino Pressure EI or ustimate A Iowable Computed Computed Bottom Approx. Loading Pressure Net Net Bea ring Static Dynamic or Mat Cross Net Gross Net Bearing Capacity Factor Factor Supporting roundation size Static Static Dynamic Dynamic Capac ty StatigI Dynamic or. or E.t r_pe tu re _3_ra..t um__ frt) __( r tJ_ Lkirgzy tu,ftg 1 (k/rta1 (k/rtz) 2 (k/rt l-(k/ft (k/ftal Sarety Serety rector makeup BackriiI 220/212 51x51 2.3 1.3 5.3 4.3 95.7 31.9 47.9 73.6-22.3 water storage tank Raruct ing water Backrill 220/216 62x62 3.7 3.2 11.8 11.3 88.9 29.6 44.5 27.8 7.9 storage tank R*dwaste Backrill 220/216.5 93x45 3.6 3.2 7.7 7.3 86.5 28.8 43.2 27.0 12.0 tesnsrer buiiding 14 M O

  • v 5

m in>

  • U s

M l {> ! h m. The net static soad is the somd in excess or the overburden pressure at the base or the structure.

s b.

The ustimate net bearing capaci ty is the soad in excess or the overburden pressure at the roundation level at which shee r o. fairure wist occur in the roundation stratum. The allowable net static and dynamic bearing capacities are obtained by dividing the net ultimate bearing capacity by rectors c. or 3 and 2 respectively. [ d. The net static bearing pressure is negative, u \\ co tn 9

O II.1-INPUT SEISMIC l'OTION The results of the three-dimensjonal analysis of HDRs are quite sensitive to the"frequency content and duration of the time histories being used for the analysis. Provide information on how the power spectral densities synthetic time histories possess r(PSDs) of the proposed easonably distributed energy content over a frequency range of interest and generally match the target PSD criteria, in addition to the enveloping criteria of the standard Review Plan 3.7.1. Guidelines for. developing PSD can be found in NUREG/CR-3509, "Power Spectral Density Functions compatible with NRC Regulatory Guide 1.60."

RESPONSE

NUREG/CR-3509 (Reference 1) describes the generation of a power spectral density (PSD) function of a stationary process.in auch a way.that the sample functions satisfy the NRC Regulatory Guide 1.60 response spectrum. The PSD function criteria for the ground acceleration time histories (Target Ground PSD Function) are provided in equatica 2 of this report with So = 1100 in.8 /sec' (for 1 g acceleration) recommended in the Conclusions section of this report. The guidelines provided in this report are intended only for ground PSD. However, based on the target graund PSD function, a conservative assessment of the adequacy of tre flocr time history PSD can be made. Figure II.1-1 shows a comparison of the target ground PSD function for SSE (i.e., corresponding to 0.2 g) with the PSDs for the Bechtel ground horizontal acceleration time history and the floor horizontal (H1) acceleration time history. It can be seen that the PSD corresponding to Bechtel ground time history is,.in general, an order of magnitude higher than the target ground PSD function in the frequency range of interest, where the floor response spectra exhibit peak amplifications. This comparison is consistent with Figuras 3.1 through 3.14 of l NUREG/CR-3480 (Reference 2), where con.parisons of PSDs of 14 time histories, obtained from 11 firms in the nuclear industry, with the target ground PSD function are'provided. Figure II.1-1 hiso shows that the floor time history PSD is i significantly above the ground time history PSD. Both these f PSDs were plotted using a three point moving average. Because of the oscillatory nature of the PSDs, it is difficult to ascertain the relative amplifications. Therefore, in order to better assess the amplification in the floor time history PSD, j the PSDs for the ground and floor time histories were also plotted using 21 and 13 point moving averages, respectively. s II.1-1 i

The higher averaging number 'for the free fiel'd PSD was necessary oue to the high frequency content of the free field. This comparison (Figure II.1-2 in the 2 to 5.5 cps range)(the amplified region of the floorshows that the response spectra, Figure II.1-3) is about 10 times the ground. time history PSD. From 1 to 2 cps and 5.5 to 10 cps, the ratio tapers to about 5.0. Similar comparicons are provided for the floor SSE-H2, SSE-V, OBE-H1, OBE-H2, and CBE-V time histories in Figures II.1-4 through II.1-8. Figures IIs1-9 through II.1-11 provide SGE-V, OBE-H, and OBE-V floor response spectra. ~ Considering the order of magnitude conservatista of the ground time history PSD over the target ground PSD, and the amplification of the floor time history PSD over the ground time history PSD without any missing frequency windows in the frequency range of interest, adequate distributed energy content.the floor time history possesses Referencee i 1. Shinozuka, et al: Power Spectral Density Functions Compatible with NRC RG 1.60 Response Spectral. NUREG/CR-3509 March 1984. 2. David W. Coates, Jr. and David A. Lappa: Value/ Impact Assessment for Seismic Design Criteria. USI A-40, NUREG/CR-3489 August 1984. l t S F II.1-2

SSE -11 POWER SPECTRAL DENSITY COMPARISON FLOOR TH (AVE 3) VS GROUND TH (AVE 3) VS NRC TARGET GROUND 10000 ,e' ^ i r ~ ,et s, ni 1-4 e,% ' s a 't ,a s, s i f l ~ l t e

    • s'.

I 1000 - c. n i' i k3 i I

f. i, i

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a. TM,I (i [

i 3 I I3 l%I'9 fAfskt .L ' I,X SMJ t M 3 'Il N/$i ] l@l\\,': L' 4 1 1 - FREQUENCY (CPS)10. 100. l Figure II.1-1

e SSE-H1 POWER SPECTRAL DENSITY COMPARISON Fl.OOR TH (AVE 11) VS GROUND TH (AVE 21) VS NRC TARGET GROUND 10000 / .1 1000 m m s r \\_ ~ ~ y 'w % s, Q / l L.! ' "g 1 a1 N t, 7 100 )l 1, Z G I' ~f ' ... 3 s 1, _f \\ 1. O \\ f [ ', I ' I M N l '. Il 1 O_ \\ l I 10 \\ 'i '.

/.

4 n.,.Ib 'i N i '.IN i et, 'i \\ Win' T., '( Nefh,1) 1 1 1-FREQUENCY (CPS)10. 100. Figure II.1-2 4

VNP DESIGN MANUAL REV I DATE 4-4-83 DC-1005 i F R EQUE NCY (CYCLES PE R S ECOND) 100 50 25 to 5 2 1 .5 .2 -" i BE CHTE L POWE R T CORPOR M M oncN eAss$. P$AR $3 AND $4 mi- -t--+ hh W GEORGIA POWER COMPANY ~'&"--~~ h h 1-ALVIN W.V0GTLE NUCLEAR PLANT l l

4^

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tav.g p = 2 4 .q -q Y h h ^ ZN Nk! [Mf i h5 hU l g , 3, , v e 1 1 x ~ __..--.. -t3& g g__ v -lh M-m" 4: ? 1 i m i i rW - e y = w > &_p p a m k I I ^ 'hi e .01 .o2 .03.04 oe.oe.1 .2 .s .4 .s .s i 2 3 4 s m:-oo tsEcoNosi Figure I7. 1-3

SSE-H2 POWER SPECTRAL DENSITY COMPARISON FLOOR TH (AVE 11) VS GROUND TH (AVE 21) VS NRC TARGET GROUND 10000 1 ~~ i ,e s' i ^ ~ ~ -,s 1000 ^ 0 -l 's y ~ / V 1/gJ, W ' ' g' / i (/) i N ( /. N 100 z ~ X I '. f er. A 1 _I I,i, a 'N / " B9 i cn w 4 ~ \\ Il' Q_ l 10c \\ 'i ',' 'N r,3 U N i,; P t.e t, i r \\ M,8k T.,' I i %5,7 lll h3$I i 1 1-FREQUENCY (CF3)10. 100. 1 Figure II.1-4

SS E-V ? OWE R SP EC-RA J ENS - Y CON PAR.' sob FLOOR TH (AVE 11) VS GROUND TH (AVE 21) VS NRC TARGET GROUND 1oooo s' 1 s 1000 i n .i Q T,' g' / s W r ^ '/ (' g'

ig D

i,,, s' N ig-4,' ? 100 Z w in X i s' U N L a g O k s i m x y CL \\ gil li. 10 N s.

.. ac ;

i \\ ' f,,[,',i / \\ J.T6, , p \\ '# 3'I' 3 3,C XVI ' %' s i j i l j 1-FREQUENCY (CPS)10. 100. I'ig u re II.1-5 l l

I OBE-F1 POWER SPECTRAL DENSITY COMPARISON FLOOR TH (AVE 11) VS GROUND TH (AVE 21) VS NRC TARGET GROUND 10000 / \\' s 1000 s,s n V) i is i O La g -s. t 1 N ~ 3" '^ ? 100 r ,l/.i s' ,, e [ ![i C W i E N fii ' ' l 'l' s i 10 \\ ', 6 ! ' _ : N , ', )' 'O,. s A

  • r 11 1,

\\ '4' 1.(, v,,e '6 ' 's 5 ,-a, \\ f, I', ! ;'l' 1 l l-FREQUENCY (CPS)10. 100. Figure 11.1-6

03E-H2 POWER SPECTRAL DENSITY COMPARISON FLOOR TH (AVE 11) VS GROUND TH (AVE 21) VS NRC TARGET GROUND 10000 / t 2 1, ?' 1000 m V) i. e Q s'. w ~ - s. g ~ t / ) ), g s s ? 100 Z i 1 1 ' t, G 14 + n'l l,.. Q -l', i.) 8 I t l 10 \\ '; J \\ i i,,' i P2 A \\ '.. : 'r il, . b, e l _ = L m r, \\ 18 f t f f. fit e T 16 3'a' l 1 ,t j 1 l 1 - FREQUENCY (CPS)10. 100. Figure 11.1-7

I 1 03E-V 30WER S? EC-RA D ENS TY COV 3A R SON l FLOOR TH (AVE 11) VS GROUND TH (AVE 21) VS NRC TARGET GROUND 10000 ^ 1000 PQ s-s Q w s, I i',, y) N h' ~ ) N 100-s z i . i' ,V,i '1 e, a -f, Y.sy I {f 's ", 10 \\ ', h ! ' x i. ',.. i.3'I s \\ ' tf \\ ' '/ '. I* u _ T a fr, \\ e It IN iw I I I 'g kI[I t j 1., t F 1 1-FREQUENCY (CPS)10. 100. 5 Figure II.1-5 l

VNP DESIGN MANUAL DC-1005 REV I DATE 4~4*03 ( \\ FREQUENCY (CYCLES PCR SECCND) 100 60 25 10 6 2 1 .5 .2 8ECHTEL POWE R -otsu tasts. ma sa' Aho u" -- " ~ CORPORATION p ti+- ++&tt, GEORot A POWER COMPANY i ~ 1- - -T HM ALVIN W.VOOTLE NUCLE AM PLANT ""Q FUEL HANDLING BUILDING et SAFt 1HufDOWN EARTHOUAKE 5 --4 1 l-a di-VERilCAL ACCEL.RtsP0451 $hCTRA i ^t h* ; E L. t 79'4" T i 4

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I VNP DESIGN MANUAL DC-1005 1 I100 74w - t FREQUENCY (CfCLES PER SECOND) 50 25 to 5 2 1 .5 .2 BECHTEL POWTR S 1.. bMNibfN5 1. CORPORATION te, l % DuiGN e AstS. PiAR $3 AND k L-tttr *'t ; GEORGI A POWER CCMPANY f*f jr g ALVIN W. VOGTLE NUCLEAR PLANT i s pu i FUEL HANDLING BUILDING OM ?P(RAflhG Bast $ E ARTHQU AKE g } i HORil ACCEL RESP 0.hst SPECTR A g g.4 a-w i y gg q % er: memed er: Approved er: g j, i O h N ^ }{' g 4 I 'k Joe evo. X2CR447 M E V. O j setScot i g.p.y w.4 .... a ik!.- . i,. ~+4 < fi'- h { I -t' j: !i ' itt jh w q.--.,9 1 q 3, i !, ! ' ~~, . h .U .*-* *N* W ,1,., g ~ I ~- y~- 4> .. m + i p 1 1

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III.1 ANALYSIS AND DESIGN OF HDRL With regard to the modelling and model parameters, provide-the 'following information: Flexural rigidities of the fuel assemblies and a. justification for modelling.them as five separated rattling masses. b. How is the_ vertical mass of the fuel assemblies 4 accounted for in the model? Calculations supporting the assumption that fuel c. racks ~are rigid. d. Provide additional information on how the rotations of the support legs lare simulated. How is the limiting condition-of slab moment (P. 14, Seismic Analysis) calculated? Provide information on how the non-symmetrical e. arrangement of the fuel-cell lateral supports (Figure 9.1.2-2B) is accounted for in the rack analysis. j J f. Because of the unpredictable nature of the rack-movements under seismic event, it is not conceivable to assign one value (so called "conservative") to model parameters such as spring constants. Provide t the numerical values associated with Table 2 of the Seismic Analysis and an estimate of' potential variation from the best estimate values.

RESPONSE

Flexural rigidities of the fuel assemblies are not a. i accounted for in the model. The use of uncoupled i i lumped masses reflects the fact that the fuel assembly flexural stiffness is insignificant compared to the rack flexural rigidity.. The cantilever mode fundamental natural frequency of a Westinghouse PWR fuel assembly in air is approximately 1.2 Hz. It is intuitively apparent that the bending and shear springs connecting the fuel assembly nodal masses { j will be quite weak, and would therefore have an insignificant effect on the results. Fuel is modelled as five lumped masses at Z=0.0, 0.25H, 0.5H, 0.75H, and H, where H is the height of the rack above the baseplate. Actual fuel-to-rack i impacts, if they occur,'will occur at the locations of any "support plates" for the rod bundle, and i 1 III.1-1 i r ase

certainly at the top of the rack. The impact locations in the dynamic model were chosen to balance the modelling of the fuel as a uniform mass distribution with the actual location of potential impact points, b. The vertical mass of the fuel assemblies is assumed to be attached to the rack and to move vertically in accordance with the assumed location of the fuel assembly centroid with respect to the rack centroid. Justification of Rigid Rack Assumption c. Seismic forcing frequencies are on the order of 5 Hz. The use of a rigid rack assumption can be justified if we can show that the lowest natural frequency of rack vibration in water is 33 Hz or greater. To show this, we consider the rack as a cantilever beam with metal moment of inertia computed from the grid structure. To be conservative, we assume the mass of the vibrating rack to be the rack itself plus all of the water enclosed in the rack. We assume the rack is modelled as a beam of length L, area A, and cross-section inertia I around the weak axis. 1. Mass of Rack Per Inch mt = 0.293 A/g where A = rack cross-soction metal area Mass of Water in Rack Per Inch 64 (ab-A) m2 = 1728 g where a (sb), b are the length and width of the rack, respectively. The total mass per unit length is the sum of the abova two quantities. m = mi +ma III.1-2 k

s For, Young's Modulus, we use E so that the frequency is - 1.8758 'EI 2/2 ~- freq = __ 2 m L' For the 10x12 rack, a = 105 in., b = 124 in.,.A ' 362 in.8,. L = 169 in. A conservative' estimate of the weakest metal moment of inertia is I = 202,757 in.' For E = 27.9 x 10' psi i

  1. sec*
  2. sec 8 mi = 0.293 x 362/386.4 = 0.2744 1.213

_; m2 = EI 27.9 x 10' x 0.202757 x 10' m L' 1.4874 x 8.157 x 10' = 0.4662 x 10' i Therefore, using the previous frequency formula i i freq = 38.2 Hz This value is greater than 33 Hz; therefore, the j assumption of a rigid rack is justified, d. The effect of local floor elasticity on rocking motion (bending of the support) is represented by a rotational spring having spring rate KR = 2E I/ II~M ) c B i I III.1-3 l i a

where I = D'/12, and B is the characteristic dimension of the support foot. Ec and y are the Young's Modulus and the Poisson ratio of the foundation concrete, respectively. Therefore, for the given rack support pads KR* /6 (1-9 8)' c The spring constant X is used'in the following manner. With the support leg rotation in each direction obtained from the dynamic analysis program at each time step, the base moment on each leg due-to a rotation, d, is determined from the relation m=KR' applied in the x,y directions. Since the base moment cannot exceed the moment provided if edging occurs, these base rotational springs (two on each support) are modelled as rotational frictional elements, with maximum ellowable moment given by the current downward load times the edge distance B/2. The calculation of inertia properties for the rack e. cross-section assumes a grid construction with all edges completely connected by welding. Since assembly of the configuration does not permit connection at all corner locations, for conservatism, we neglect certain mass when computing rack inertia. The calculation of inertia properties Ix, I for the rack cross-section uses the "as analyzed" y configuration with each cell modelled as a square section of thickness t and x = y = c where c is the average cell-to-cell pitch (see Figure III.1-1). No credit is taken for the inner segments of side length E. The use of lower calculated values for Ix, I leads to conservative stresses computed at the rack y base. In the calculation of rack metal area, however, the actual available cross-section is used, since the axial loading does not depend on shear transfer. The calculation of torsional properties also uses the idealized square cross-section with the inner segments neglected. In all dynamic analyses, however, we use the idealized values. of Ix, Iy; thus, the values used for rack inertia properties are i III.1-4 i 1

E 's lower, which implies higher values-for rack stresses used to assess rack cell-to-baseplate attachment

welds, f.

The model parameters used for the Vogtle analysis are summarized in Table III.1-1. Also included, for reference only, is a two-dimensional idealization of e the model which describes the pacameters (Figure III.1-2). The spring. mass model in-this. figure is intentionally simplified for clarity. -The-actual. analysis used a three-dimensional model with five rattling masses. i 3 -t t 6 .i 'i l 'l l .i - l i 1 III.1-5 i

-s 7 ] Table III.1-1 Summary of Properties 10 x if; 9 x 11 10 x 11 AX (in.) 104.875 94.3 105 AY (in.) 124.063 114 114 H (in.) 169 169 169 4 h (in.) 5.25 5.25 5.25 K (#/in.) 0.164 x 108 (x) 0.162 x 108 0.180 x 10' y O.234 x 10' (y) 0.240 x los 0.215 x 10'~ K (#/in.) 953379 953379. 9T3379. 6 K (in-#) 5.668 x 10' 5.668 x 10' 5.568 x 10' p in. Kg (#/in.) 3.0247 x 10' 3.024 x 10' 3.0247 x 10' We would expect approximately 10% variation'in some of the spring rates except for Ky where we have deliberately chosen significantly larger values than would reasonably be expected. Some previous work (undocumented numerical experiments) has indicated that changes in K1 affect the peak impact force between cell and assembly, but do not significantly affect the overall rack resp';nse. This is primarily a result of large modifications to Kyto give the same over(by a factor of 10 downward) still tending all impulse to the rack. ? 4 e u III.'.-a

n, m, n, 13 7 12 16 [ [ l l 1 ~ T-m, m, 9 m, I \\ 6+e 10 i 15 -f I e--- re 1 i E [ ] l ] o n, r< r 9 5 8 14 [ [ ] ] ] r, r r, 3 1 2 4 ,a ia o E 6 t J a Figure III.1-1 Box Assembly Sequence _. = _

Z t Ki n )j 2 2 TYPICAL RATTLING FUEL MASS RACK CENTROID (ASSUMED at H/2) l H 4 RIGID RACK BASEPLATE d 6 Kb K6 h Kt I N" U* _U fg / 0 KR KR 3 UU hyy ~ I 1 1 Figure III.1-2 Spring Mass Simulation for Two-Dimensional Motion c, ,--+e-w.

- e.- t. n III.2 The staff had previously reviewed and accepted the results of rack analysis using DYNARACK computer _ code. However, it appears that the cede format has been changed to include the use of non-linear vibration theory and cross-coupling effects as indicated in items "n" and "o" of Section 3.1. In light-of the changes provide the following information: Demonstrate the validity of the non-linear vibration a. theory used to simulate gap variation in the ~ fuel-cell impact analysis and in the rack analysis. Assthe analysis results_will lead to less-conservative design, provide inclusion of all the associated factors such as realistic modelling of fuel rods comprising a fuel assembly,' fluid flow through-rods, differences in various fluid paths along the-length of the fuel assemblies. Provide verification of the theory with the statistically evaluated experimental results and the limits of-their applicability. b With regard to the cross-coupling effects due to the movement of fluid around the adjacent racks, the staff had accepted its use for two dimensional multi-rack analyses. However, in absence of-any experimental verification, th3 gap distance was conservatively recommended as three times the nominal space between the adjacent racks. Provide information on how the formulation will be used to analyze the rack to rack impact at Vogtle 2. Provide a comparative tabulation showing c. displacements and impact loads using the linear vibra' tion formulation and those using the proposed non-linear vibration theory.

RESPONSE

The Vogtle rack analysis did not use any non-linear a. vibration theory for calculating fluid coupling terms. The Licensing Seismic report (paragraph 3.1.n) is being revised as follows: "The rattling of the fuel assemblies inside the storage locations causes the ' gap' between the fuel assemblies and the cell wall to change from a maximum of twice the-nominal gap to a theoretical zero gap. Therefore,'the fluid coupling coefficients (Referenceo 8,9) utilized are based on the average effective gap during the seismic event Linear vibratior. theory is used to simulate the fluid coupling effects. " III.2-1 -w

e O References 1 and 2 below outline the development of the fluid coupling expressions: 1. R. J. Fritz, "The Effects of Liquids on the Dynamic Motion of Immersed Solids," Journal of Engineering for Industry, ASME, Feb. 1972, pp. 167-172. 2. "Dynamic Coupling in a Closely Spaced Two-Bady System Vibrating in a Liquid Medium: The Case of Fuel Racks," by K.P. Singh and A.I. Soler, Proc. of the Third Confgrence on Vibration in Nuclear Plants, 1982. Fritz consistently makes the small deflection assumption in his work, but the approach of determining fluid velocities to satisfy continuity and subsequently forming system kinetic energy is not limited to small deformations. Previous studies on simple models in Reference 2 indicated that inclusion of large deformation effects could lead to a lowering of the structural response.

n the detailed rack analysis, a complate large deflection non-linear analysis can lea.. to a significant increase in computational t.mo.

Therefore, in the interest of conservatism, in lieu of an instantaneous recalculation of fluid coupling terms, an equivalent linear approach was used. The fluid coupling effect between the fuel assembly and the storage cell is considered by treating the former as a blunt body. The standard PWR fuel assemblies used in the Vogtle reactor typically contain 264 fuel rods in a 17x17 array. The fuel rods are 0.374 inch in diameter arranged in a square lattice with a pitch of 0.496 inch. Therefore, the gap between the adjacent fuel rods is less than 1/8 inch (0.122 inch nominal) The cross-sectional dimension of the rod array is 8.424 inches square. Since the storage cell opening cross-sectional dimension is 8.75 inches, the net lateral spacing between the fuel assembly and the storage cell is 0.326 inch. The lateral movement of the fuel assembly in the storage cell causes the water to flow past the assembly. Since the flow between these narrow channels formed by the array of rods involves repeated changes in the flow cross-section of width from 0.122 inch to 0.496 inch--a fourfold change in transverse flow area--the hydraulic pressure losses III.2-2

s: o t through these channels are an order of magnitude greater than what the fluid s 1 counters flowing through the assembly / cell wall gap around the array periphery. The hydraulic pressure loss due.to flow through these narrow convergent / divergent channels is an important mechanism for energy loss from the vibrating rack system. However, in the conservative approach used to model fluid coupling, no such flow, and therefore no such loss, occurs; all the fluid is assumed to flow in the assembly / cell wall space around the array periphery, b. Fluid external to a rack acts in the gaps perpendicular to a given seismic excitation, and also in the gaps parallel to the excitation direction. To obtain the correct characterization of the rack module motion, it is necessary to include the effect of the racks,and/or walls around the subject rack in an appropriate manner. This is done by conservatively estimating the cross-coupling terms in the three-dimensional motion in the manner of the Diablo Canyon simulation.- The methodology for estimating the cross-coupling terms does not depend on whether two-dimensional or three-dimensional analysis is being carried out. Performing the cross-coupling term evaluation in an identical manner to Diablo Canyon leads to the nominal gap multiplier equal to 2.78 for the Vogtle Unit 2 pool. The potential for rack-to-rack impact is assessed by monitoring the maximum rack displacements. Since no rack-to-rack impact is to occur, we show that the maximum rack excursion is less than 50% of the spacing to an adjacent rack, or less than 100% of the spacing to an adjacent wall. Not applicable (see response paragraph a on c. page 111.2-1). III.2-3

4 r IV.1 DROPPED FUEL ASSEMBLY IMPACT Provide calculations demonstrating the assertion in the submittal that the high shear stresses due to the postulated impact load will be sbove the active fuel region.

RESPONSE

Section 12.2 of the seismic analysis report for the VEGP Unit 2 spent fuel racks (which was transmitted to the NRC by letter on December 23, 1987) summarized an analysis of a dropped fuel asse.-bly striking the top of.a spent fuel rack. This analysis was f or the pool in the dry condition. 1k) spent fuel will be stored or handled in the spent fuel pool in the dry conditions. The consequences of such an evaluation are of little interest from the standpoint of radiological releases. As stated in Section 9.1.2 of the FSAR, "Fuel assembly drop accidents involving non-irradiated-new fuel, which may be stored in the spent fuel racks in a dry pool do not effect criticality criteria; therefore, analysis considers only the core of a dropped spent, irradiated fuel assembly in a flooded pool and takes credit for dissolved boron in the water." Therefore, Section 12.2 of the report will be revised to show I the results of the analysis performed for the dropped irradiated fuel assembly. The analysis described in the FSAR also conservatively assumes the weight of a fuel assembly, control rod assembly, and spent fuel handling tool to be 2300 lb. This weight exceeds the weight of the fuel assembly with control rod assembly (1647 lb) which was used in the analysis described in Section 12.2 of the seismic analysis report. Since the analysis described in Section 12.2 of the seismic report is not consistent with the FSAR, even though it meets the Standard Review Plan guidance, a revised analysis will be provided. The revised analysis will be provided by August 12, 1988. i i l i i 2 Iv.1-1

.~ IV.2 Provide calculations showing the effects of the postulated impact load on the rack baseplate will not damage the pool liner.

RESPONSE

The response to this question will be provided as part of the response to question IV.1. l l IV.2-1 4

i. .V.1 MISCELLANEOUS ITEMS With regard to the planned partial loading of the pool, provide the following information: Calculation demonstrating the stability of the end a. racks (factors of safety against tilting and overturning) where the fluid coupling effects are negligible. b. Calculations in "a' with partial and eccentric loading of the racks.

RESPONSE

A number of runs were carried out in dry and wet pool conditions for a variety of loading conditions. A measure of stability against overturning is provided by referring to the attached Figure V.1-1. If OH2>d where 8 is the maximum angle of rotation in either direction, obtained from the dynamic analysis, then we may have reached a threshold for an instability. If we define Om = d/H, then Om/0 provides a measure of the safety margin available. 2 Table V.1-1 gives results for typical cases. The maximum values of 8 ara tracked in every dynamic analysis. In each case H2* 0.5 x 169 + 5.13 = 89.75 inches. The value used for d varies in accordance with the loading pattern. The attached table contains results for the A module in dry storage condition. It is recognized that module B, which has a more balanced planform (11x10 cells for B versus 11x9 cells for A) will have even larger margins of safety against kinematic instability. The factors of safety against overturning presented in Table V.1-1 are typical of the Vogtle Unit 2 spent fuel pool modules. Other scenarios of storage, such as a storage pattern resulting in the center of gravity of the stored array non-collinear with the rack centerline, produce even larger margins against overturning. This is because of the characteristic of the seismic excitation at the Vogtle site, the fact that a partially filled rack, by definition, has lessand rattling inertia than a fully loaded n.odule. 4 V.1-1 l

2 6 'I L I P' = Centroid of rack + fuel P' I o H i g i n ed H2 I d L U 1 4 W = Figure V,1-1 .I

? Table V.-1-1 Margins of Safety Against Incipient Overturning RUN d 0 0 O d8 m m Al dry 28.7 C.320 0.61x10 8 524.6 16 assemblies u = 0.8 Al dry 47.1-0.525 0.34x10 8 1544. 32 assemblies y = 0.2 Al dry 47.1 0.525 0.192x10 8 273.4 32 assemblies y = 0.8 Al dry 47.1 0.525 0.903x10 8 581.4 99 assemblies y = 0.2 Al dry 47.1 0.525 0.415x10 8 1265.1 99 assemblies u = 0.8 D2 wet 52.5 0.585 0.254x10 8 230.3 120 assemblies u = 0.8 Al wet 47.1 0.525 0.243x10 8 216 99 assemblies 2 racks in pool y = 0.8 Al wet 47.1 0.525 0.273x10 8 192 99 assemblies 4 racks in pool u = 0.8 B3 wet 52.5 0.585 0.262x10 8 223.3 110 assemblies partially isolate u = 0.8 D2 wet 26.25 0.2925 0.106x10 2 275.94 60 assemblies y = 0.8 4 r_.- y.. ,-p.- ,_ _ _--. -... _ -- ~.. -.. y

e s' e i V.2 Provide details of the proposed installation procedure indicating how the elevations of the racks and designated gaps between the racks will be maintained and monitored.

RESPONSE

The pool and racks are surveyed to predetermine rack locations maximizins rack-to-wall, rack-to-rack, and rack foot-to-leak chase distances (2-1/8-inch, 1-1/2-inch, 1-inch minimum, respectively). The racks are placed in these locations and leveled with an overhead crane, rollers, and jacks. Minimum clearances are confirmed by use of a feeler gage,.and rack levelness is verified using a plumb or appropriate (optical) sighting equipment. See the response to V 3 below for a description of procedures implemented after a seismic event. l 1 i 1 i V.2-1

'ko ? e V.3 Provide a description of plant safety procedures in case of: a. Fuel drop accident b. A seismic event Loss of water from the pool detected by leak chases. c. R_ESPONSE: In the event of a fuel handling accident in the fuel a. handling building, the building will be evacuated and the building ventilation will be verified to be operating in the post-accident mode. In addition, the conditions will be evaluated for implementatior. of the Site Emergency Plan. After the building has been approved for reentry, the Engineering Support Department will analyze the damaged fuel assembly and implement steps to recover the assembly. b. In the event of a seismic event, the Site Emergency Plan will be implemented. The Engineering Support Department will review the data from the seismic monitors and initiate inspection programs of various plant systems, including the fuel handling equipment. Specific actions, including plant shutdown, w 4.1 ) be based on the results of this evaluation. In the event a loss of water from the spent fuel pit c. is detected by the leak chases, all fuel movement will be suspended and immediate action will be taken to restore proper water levels in the pool. 1174n V.3-1}}

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