ML20079P395

Forwards Final Licensing Rept on New Fuel Storage Racks & marked-up Tech Specs Re High Density Spent Fuel Storage Racks.Nshc Appropriate Based on Rept & 10CFR50.92 Evaluation

ML20079P395
Person / Time
Site:	Summer
Issue date:	01/23/1984
From:	Dixon O SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20079P400	List: ML20079P395 ML20079P405
References
NUDOCS 8401310206
Download: ML20079P395 (41)

Text

{{#Wiki_filter:i o SOUTH CAROLINA ELECTRIC & GAS COMPANY post o* Fica 7s4 CotuMelA. south CAROLINA 29218 O. W. OlXON, JR. January 23, 1984 vice Paes,osNr NUCLEAR OPERArioNS Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 High Density Spent Fuel Storage Racks

Dear Mr. Denton:

South Carolina Electric and Gas Company (SCE&G), in a December 28, 1983 letter from Mr. O. W. Dixon, Jr., to Mr. H. R. Denton, sub-mitted a Draft Licensing Report on High Density Spent Fuel Storage Racks and therein stated that Technical Specification changes would be forwarded upon successful review and approval,. A copy of the; Technical Specification changes, 10 copies of the/ Final l l Licensing Report on High Density Spent Fuel Storage Racks and 10 copies of the New Fuel Rack Criticality Report Reanalysis are attached. Also attached for your review are the Virgil C. Summer Nuclear Station Final Safety Analysis Report changes. SCE&G has determined that a finding of no significant hazards con-sideration is appropriate based on the attached Licensing Report and 10CFR50.92 evaluation. The Technical Specification changes have been reviewed and approved by both the Plant Safety Review Committee (PSRC) and Nuclear Safety Review Committee (NSRC). A check in the amount of Four Thousand Dollars ($4000.00) is enclosed for processing this change. If there should be any questions, please contact us at your con-venience. Very truly yours Jt O. W.

Dixon, WRM:OWD/fjc

Attachment:

cc: (See Page #2) q4 dk 8401310206 840123 \\ PDR ADOCK 05000393 OQ P PDR ci 1 v

'O Mr. Harold R. Denton High Density Spent Fuel Storage Racks January 23, 1984 Page #2 cc: V. C. Summer T. C. Nichols, Jr./O. W. Dixon, Jr. E. H. Crews, Jr. E. C. Roberts W. A. Williams, Jr. D A. Nauman J. P. O'Reilly Group Managers O. S. Bradham C. A. Price C. L. Ligon (NSRC) G. J. Braddick D. J. Richards NRC Resident Inspector J. B. Knotts, Jr. NPCF File 1 J

10CFR 50.92 SIGNIFICANT HAZARDS CONSIDERATION 1. Would the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated: ANSWER: No - With the appropriate Administrative controls in place, expanding the storage capacity of the spent fuel pool and ircreasing the allowable enrichment will not significantly increase the onsite or offsite radiological impact above that currently authorized. This amendment does not involve a significant increase in the probability or consequences of any accident previously analyzed. 2. Would the proposed amendment create the possibility of a new kind of accident from any accident previously evaluated? ANSWEP: No - The accidents previously evaluated are still pertinent to the High Density Spent Fuel Storage Rerack. The potential for a new accident has not been created by this proposed design change. 3. Would the proposed amendment involve a significant reduction in the margin of safety? ANSWER: No - The proposed amendment is consistent with the intent of Regulatory Guide 1.13, " Spent Fuel Storage Facility Design Basis," and does not involve a significant reduction in the margin of safety.

REFUELING OPERATIONS 3/4.9.12 SPENT FUEL ASSEMBLY STORAGE LIMITING CONDITION FOR OPERATION 3.9.12 The combination of initial enrichment and cumulative exposure for spent fuel assemblies stored in Regions 2 and 3 shall be within the acceptable domain of Figure 3.9-1 for Region 2 and Figure 3.9-2 for Region 3. APPLICABILITY: Whenever irradiated fuel assemblics are in the spent fuel pool. ACTION: a. With the requirements of the above specification not satisfied, suspend all other movement of fuel assemblies and crane operations with loads in the fuel storage areas and move the non-complying fuel assemblies to Region 1. Until these requirements of the above specification are satisfied, boron concentration of the spent fuel pool shall be verified to be greater than or equal to 2000 ppm at least once per 8 hours. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.12 The burnup of each spent fuel assembly stored in Regions 2 and 3 shall be ascertained by careful analysis of its burnup history prior to storage in Region 2 or 3. A complete record of such analysis shall be kept for the time period that the spent fuel assembly remains in Region 2 or 3 of the spent fuel pool. i SUMMER - UNIT 1 3/4 9-14

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REFUELING OPERATIONS BASES 3/4.9.12 SPENT FUEL ASSEMBLY STORAGE The restrictions placed on spent fuel assemblies stored in Regions 2 and 3 of the spent fuel pool ensure inadvertent criticality will. not occur. l l SUMMER - U14IT 1 B 3/4 9-3 i t .s

.y ~ D EESIGN FEATURES r ~ 7-f' 5= T': i!"'E ~ ~ ,c g, ~ l; [513 ~ REACTOR ' CORE;gg-.---{r.g_:E-2::d?pg; i;q{ip'_:5:y. a-.==----- -~ _._..._.=--:.---t--< .~=~.:'==---.=;; - - ---.=== =: :

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shall.have~s7'hoininar~ active ~ fuel'lengthi of 144 inches and contain'a maximum.

l total ~ weight [of.:1766 grans: uranium.-DThe_ initial core'. loading shall have a ~~ maximum' enrichment..of. 3.2.weightTpercent'U-235L:-Reload fuel shall be similar f_ig:yhysical'. design;to the.ini_tial core _ loading.and shall have a maximum ~ 7 enrichment of M weight, percent U-235.: r : = 2 cr.. 2 7: -

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= '. - 5.3.2rThe reactor core shall contain 48 full length.' control rod assemblies. ..The1 full-length control _ rod assemblies shall contain a nominal 142 inches of cabsorber material..The nominal values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall be iclad with stainles's steel tubing. i m 5.4 REACTOR COOLANT SYSTEM' ~ DESIGN PRESSURE AND TEMPERATURE 5.4.1.The reactor coolant system is designed and shall be maintained: a. _In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowin'ce for normal degradation pursuant to the applicable Surveillance Requirements, ~ b.' For a pressure of 2485 psig, and c. For a temperature of 650 F, except for the pressurizer which is 680 F. VOLUME 5.4.5 The total water and steam volume of the reactor coolant system is 9407 i 100 cubic feet at a nominal T,yg of 586.8*F. j \\ 5.5 11ETEOR0 LOGICAL TOWER LOCATION l' 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1. i SUMMER - UNIT 1 5-6

DESIGN. FEATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks consist of 1276 individual cells, each of which accommodates a single fuel assembly. The cells are grouped into 3 regions. Region 1 is designated for storage of freshly discharged fuel assemblies with enrichments up to 4.3 weight percent U-235. The cells in Region 2 are reserved for accommodating fuel assemblies with initial enrichments of 4.3 weight percent U-235 and a minimum burnup of 20,000 MWD /MTU. Both regions 1 and 2'are poisoned. Region 3 cells are capable of accommodating fuel assemblies with initial enrichments of 4.3 weight percent U-235 and a minimum burnup of 4 2,000 MWD /MTU. The spent fuel storage racks are designed and shall be maintained with: a. AKeff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance for uncertainties as described in Section 4.3 of the FSAR. b. Nominal center-to-center distance between fuel assemblies of 10.4025" in Region 1, 10.4025" x 10.1875" in Region 2, and 10.116" in Region 3. 5.6.1.2 The new fuel storage racks are designed and shall be maintained with a nominal 21 inch center-to-center distance between new fuel assemblies such that k gg will not exceed e 0.98 when fuel having a maximum enrichment of 4.3 weight percent U-235 is in place and various densities of unborated water are assumed including aqueous foam moderation. The k gg of e <0.98 includes the conservative allowance for uncertainties described in Section 4.3 of the FSAR. DRAINAGE 5.6.2 The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pocl below elevation 460'3". SUMMER - UNIT 1 5-7

4 DESIGN FEATURES CAPACITY 5.6.3 The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1276 fuel assemblies, 242 in Region 1, 99 in Region 2, and 935 in Region 3. 5.7 COMPCNENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. SUMMER - UNIT 1 5.7(a) ~

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'Ihis section identifies those criteria important$ to criticality sbf$ty analyses. et. 3.1. l. ) Nw Ful.stgo.S e, New fuel is normally stored in.b4' inch center to center racks in the 20 .~ ~ new fuel storage facility with no water present but which are designed so as to prevent accidental criticality even if unborated water is prescat. For the flooded condition assuming new fuel of the highest anticipated enrichment w/o U-235) 'in place, the effective multipli-cation factor does not exceed 0.95. For the normally dry condition the ef fective multiplication factor does not exceed 0.98 with fuel of the highest anticipated enrichment in place assuming optimum moderationMoce.e Mer m.d), Credit is taken for the inherent neutron absorbing effect of materials gf construction of the racks. () u Gd In the analysis for thef storage f acilities, the fuel assemblies are assumed to be in their most reactive condition, namely fresh or unde-pleted and with no control rods or removable neutron absorbers present. Assemblics can not be closer together than the design separation pro-l vided by the storage facility except in special cases such as in fuel shipping containers where analyses are carried out to establish the acceptability of the de. sign. The mechanical integrity of the fuel l assembly is assumed. J The fDllowios describes net fue{ stora e in t e spen fuel poolJ Unbo ated w ' ter of 1.0 g 'cc is assumec. in the analy is. er le range of w ter d tsities of ir terest (corre ipondin to 60 F thr ugh 12*F, ful densi y wate is a, conser ative ssumpt on si e a d cre ei I wa er den ty wil. caus the ffecti,e mult plicat on fa tor K f eff Itspouldb emph-ized at be lin is t t e syste to de rease.. rmitte to occhr und r anyleircums tances Re3 o n 1J iL c spui Ecl Au bc $40 red.., ) p Lt m ay dhtw u ed in 5J o, 't. 3. 2. 7. 2_ d o r mp. b,l.47, e 4.3-49 /GE R "T 20 AUOOCT, L940 i l l I

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.. ', egu. w.r.. con de ng possible variations, there is a 95 percent c'onfidence le 1r-. .g. egynv g as.-;3c.1; .? that t effective multiplication factor (Kef f) of' thk fuel storag ' f.< ,,.m 'y* te u ;- 3.g.g. g,ossible u.o i- .4. p i..- array wil be' l. ess than 0.9 5 as per A.,NSI Standard N18.2, ;;.; 4,. a, The .w~ p :x..: ,: u 9.. variations.1, the criticality analyses are in three categories: 1)_ cal-3... - m ~- culational unc reainties, 2) fuel rack fabrication uncertain es, and ' ~ 3). transport eff ts. The results of compa ng standard calculations with 10 critical experi-ments as su-arized in able 4. 3-4 indicate tha. j 1. The average difference between the calculat ons and experimental results or bias in the c mputations, was .1 parcent ok which is denoted as the calculation 1 bias, and 2. The standard deviation in the iff ence between the calculations and experimental results was 0. percent e.k. Multiplying the te one sided upper tolerance standard deviation by the appr r factor results in a calculat onal u certainty valid at the 95 percent confidence level. l Reactivity uncertainties co responding to fue rack fabrication uncer- [ tainties are defined as f 11ows: 1. The clearance sp cing to permit insertion of th fuel assemblies into the rack s equivalent to a clearance reacti ity. 3 .m ity associated with the tolerance on the p sition of 2. The react 11 is de-full le gth structural' support members within a given note as cell reactivity. f 3. e reactivity associated with the tolerance on center to ce ter ( l spacing between cells is denoted as center to center (c-c) rea - tivity. ~ k I 4.3-50 1

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.~ ..w ~- r 4 diffusion # lts at'several..' center .'"--n, T x Co aring correspon transport e.n resu . j l iN9. *Gt EE Mi i O!;diu k..-. .,,->.i. -..- gr. Flows a transport bias,,to be,.. lculated by the fol. T,U to?c ter spac gs al ca i SI h ""'W' lowi$d k i 3 0,: ' jf.$M,5 [' e ?.' -G..Keff (Transport) - Kett.(Diffusion) . Transport B x 100 = C. K,gg (Transport) s' The three types f uncertainties described above are statis cally com-bined in the foll ing equation: ' ~ Total Uncertainty = T nsport Bias + Calculational aa + [(Ca culational Uncertainty) (c-c Reactivity) + (Ce Reactivity) + (Cle rance Reactivity) ) ! i In evaluating the above expre ion for an finite array of 17 x 17 assem-blies enriched to 3.5 w/o U-235, the tot uncertainty is calculated to be 4.8 percent AK,f f. Subtractin O. 8 from 0.95 gives an effective ^ multiplication factor of 0.902 whic results in a center to center spacing 5 of 14.0 inches. Accordingly, a 1.8 ch center to center rack spacing corresponds to at least 95'per at of t time K,ff will not exceed 0.95 at a 95 percent confidence 1 el. Verification that the d ign criteria for wet uel storage is not ex-ceeded is achieved th ough the use of standard k tinghouse design methods such as th E0 PARD and PDQ codes It hould be noted that on i the basis of a r ent (July,1975) evaluation of WRE coupled-core experi-ments it has b n concluded that there is no need to a ly a transport bias to diff ion theory results obtained using the LEOP /PDQ com-putational ystem However, for additional conservati Westing-house vi continue to include a transport bias in the calcu tion of total certainty. In he design of dry new fuel storage racks (with fresh fuel prese t), c iculations are performed including sources of optimum moderation quch 4.3-51 A N.T 5.. 6'

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-w,qh. ^ .ww. -: n.y. ~~ " "- - - v rm n f.'culationsare, appropriate y conformance.D ' c./f,g,'4'{,fiy ,. i ' /.', - % 4 4.. 3 ~. = ~f. r. s An infinite er of dry fuel assembliesof design would have a K .80. ' ' ~ ' e.m.,.. t : e ea Fwel d b a y T h e d A). 4.3.2.8 Stability 4.3.2.8.1 Introduction ) The stability of the PWR cores against xenon-induced spatial oscilla-tions and the control of such transients are discussed extensively in Faferences [9], [17], [18), and [19]. A summary of these reports is given in the following discussion and the design bases are given in Section 4. 3.1. 6. ~ In a large reactor core, xenon-induced oscillations can take place with no corresponding change in the total power of,the core. The oscilla-tion may be caused by a power shif t in the core which occurs rapidly in comparison with the xenon-iodine time constants. Such a power shift occurs in the axial direction when a plant load change is made by con-trol rod motion and results in a change in the moderator density and fuel temperature distributions. Such a power shift could occur in the diametral plane of the core as a result of abnormal control action. Due to the negative power coefficieng of reactivity, PWR cores are in-herently stable to oscillations in total power. Protection against total power instabilities is provided by the control and protection system as described in Section 7. 7. Hence, the discussion on the. core stability will be limited here to xenon-induced spatial oscillations. ( 4.3-52 i

[nde(Y b The high density spent fuel storage racks fcr the c. E u..... - S t =

• i ~1 are designed to assure that a keff equal to or less than 0.95 is maintained with the racks fully loaded with fuel of the highest anticipated reactivity in each of three regions and flooded with unborated water at a temperature corre-sponding to the highest reactivity.

The maximum calculated re-activity includes a margin for uncertainty in reactivity cal-culations and in mechanical tolerances, statistically combined, such that the true keff will be equal to or less than 0.95 with a 95% probability at a 95% confidence level. l I i

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t }oM(k b kLodi. a i 2 assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made. o Moderator is pure, unborated water at a temperature cor-responding to the highest reactivity. e Lattice of storage racks is assumed infinite in all di-rections; i.e., no credit is taken for axial or radial neutron leakage (except in the assessment of certain abnormal / accident conditions where leakage is inherent). e Neutron absorption in minor structural members is ne.glected; i.e., spacer grids are replaced by water. The design basis fuel assembly is a 17 x 17 array of fuel at a maximum initial rods (1 2::ir' containing UO2 enrichment of 4.3% U-235 by weight, corresponding to 54.30 grams U-235 per axial centimeter of fuel assembly. Three independent regions are provided in the spent fuel storage

pool, with separate criteria defining the highest anticipated reactivity in each of the three regions as follows.

Region 1 is designed to accommodate new unirradiated fuel e with a maximum enrichment of 4.3 wt.% U-235, or spent fuel regardless of the discharge fuel burnup. e Region 2 is designed to accommodate spent fuel of 4.3 wt.% U-235 initial enrichment, which has accumulated a minimum burnup of 20,000 Mwd /mtU. Region 2 will also safely accept fuel of lower discharge fuel burnup pro-vided the initial enrichment is correspondingly lower. o Region 3 is designed to accommodate fuel of 4.3 wt.% U-235 initial enrichment which has accumulated a minimum burnup of 42,000 Mwd /mtU. Region 3 will also safely accept fuel of lower discharge fuel burnup provided the initial enrichment is correspondingly lower.

y,,,,,4 ALtoad. The reference nuclear criticality analyses of the high den-s> ae>u sh q,, u <<..u. ts e ~\\s t m, sity spent fuel storage rackAwere performed with the AMPX -KENO 2 I t.oA e. computer 4 package, using the 123-group GAM-THERMOS cross-section set and the NITAWL subroutine for U-238 resor.ance shielding effects (Nordheim integral treatment). AMPX-KENO has been ex-tensively benchmarked against a number of critical experiments ( (e.g., Refs. 3, 4, 5, and 6), including those4,6 most representa-l tive of spent fuel storage racks, Results of the benchmark calculations 6 i on a series of criti-1 cal experiments indicate a calculational bias of 0, with an un-1 l certainty of 0.003 (95% probability at a 95% confidence I level). In addition, a small correction in the calculational &, m. c a u. ., + L e-bias might be necessary to account for the A nternal water-gap i thickness 'C.110 ..M between rack walls and the fuel assemblies and 4kat ,in the V. C. Summer spent fuel rack w M t2-the-corrmpending ud mr u. in the benchmark critical experiments. s Based upon the correlation developed in Ref. 6, the correction for water-gap thickness in the V. C. Summer spent fuel storage rack indicates a small overprediction of ~ 0. 002 ak. For conser-vatism, the overprediction is neglected and the net calculational bias is taken as 0.000

• 0.003, including the effect of the water-gap thickness.

4-15

i For investigation of small reactivity effects due to manu-facturing tolerances, the CASMO computer code 7 m+- was used to calcu-late small incremental reactivity changes that would otherwise be e od lost in the normal statistical variation associated with Monte T Carlo techniques (i.e., KENO). Lei b#< n op tw\\ u\\a4 on3 I. i wcrt cd s e per Gem e d m,, ') gg L A,$ 3 0 ccde, + re ad a. k.s % A+.she d -u Aa.a B 69-e cd d W "4 " D y c..d.dease \\.~.+),c.ed ~'v <k"~nA~t> d - t. Ao * <~ ^ dc~ 6 "' # 9 b

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dhe calculated maximum reactivity in Regions 2 and 3, d ice includes a 0.015 Ak extra additive allowance for uncer-tainty in burnup calculations, p^"3 t: n addi icn:1 ar;ia ^t E _r m x= ~ 1 ^ ^E As cooling time ^# increases in long-term storage, decay of Pu-241 results in sig-nificant decrease in reac;ivity, which will provide an increasing suberiticality margin and tend to further compensate for any uncertainty in depletien calculations. Spacing between the three different rack modules is sufficient to preclude adverse inter-action between modules. Regions 2 and 3 can accommodate fuel of lower discharge fuel burnup provided the initial enrichment is correspondingly 't.3-33a 4.5 396 lower. Figures 4.3 a n d /es illustrate, as a function of the initial fuel enrichment, the minimum acceptable burnup which yields the maximum reactivity given in Table 4.1 for Regions 2 and 3, respectively. These curves are i'- i.u s incorpo-unct rated in the Technical Specifications supplemented with appro-priate adr;.inis tra tive procedures to assure verified burnup as specified in draft Regulatory Guide 1.13, Revision 2. l Although credit for soluble poison normally present in the pool water is permitted under abnormal / accident conditions, most abnornal or accident conditions will riot result in exceeding the limiting reactivity of 0.95 even in the absence of soluble poi-son. One accident condition that could potentially exceed the 1

y., u h blu^i-) limiting reactivity is the inadvertent loading of a new fuel assembly (4.3% enrichment) into Region 2 or Region 3 storage cells, with the simultaneous occurrence of a loss of all soluble poison. Administrative procedures =ill za-u preclude y the possibility of simultaneous occurrence of these two inde-pendent accident conditions. Effects on reactivity of other abnormal and accident condi-tions evaluated are summarized in Table c 4,3 -4, k positive reactivity effect results only from an increase in temperature above the nominal maximum pool temperature of 150'F assumed for criticality evaluation of the Region 3 storage rack. Tempe r-atures above 150 F are considered accident conditions, in which case the soluble poison would maintain reactivity at an accepta-bly low value. Nevertheless, in the absence of soluble poison, the increase in reactivity is calculated to be only 0.01 a k at 248* F (approximate boiling temperature of the bulk coolant at the submerged depth of the fuel racks). Thus, even in the simul-taneous occurrence of two independent accident conditions, the maximum reactivity does not exceed the limiting value of 0.95. At 248'F, voids resulting from boiling have a negative reactivity effect.

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T -1 r,x:;f y %r.: ~ "' ~ 4.:; ^ 7 *%.Lt :%.%.-.;"J',Qi~"l?. c..J .:M:~.n - I " s.'. 2, G.. ~ n. ;- ? w p:.it1 . 4~ 4.; S4 't, 7 L." ::.p:. ':.W~ ',.,a.t.c u + R g ' Ri; jg..p u - n iQ'[l[ ihhh* $Y Y ,g'. v;g. .p'pc. z:.' 3,*rn..r. d.g,,.g,g y,o e,.g J._.,.q.F ~ 13.~ Altomare,. S. and Barry, R'. F.~, "The TURTLE 24.0 Diffusion Deple 4-aw y -)phtM ': eW e r-M e 4-o.. tion Code," WCAP-7213-P-A (Proprietary) and WCAP -7758-A (Non ~ -~-,.: a-s- Proprietary), January, 1975.

14. ' Ce:mak',~ J. 0., et al'., Pressurized Water Reactor pH - Reactivity

~ Effect Final Report," M' CAP-3696-8 (EURAEC-2074), October, 1968.. 15. Ouzts, J. E., " Plant Startup Test Report, H. B. Robinson Unit No. r-2," WCAP-7844, January, 1972. h 16. Skrawbr dge, . E. nd B ry, R F., Criti litj Cale atio for g 2 (1 5). ," Nup1. Sc. and Eng. J-3, Wate -Moderbted tric B '7nifo 2.W M. Poncelet, C. G. and Christie, A. M., " Xenon-Induced Spatial In-stabilities in Large PWRs," WCAP-3680-20, (EURAEC-1974) March, 1968. "l6, % Skogen, F. B. and McFarlane, A. F., " Control Procedures for Xenon-Induced X-Y Instabilities in Large PWRs," WCAP-3680-21, (EURAEC-2111), February, 1969. 2.b. M. Skogen, F. B. and McFarlane, A. F., " Xenon-Induced Spatial Insta-bilities in Three-Dimensions," WCAP-3680-22 (EURA'IC-2116), September, 1969. (~ "' '""' "" "'" "^**"' **" ' """**"' '""'" "' """ " """" Gas and Electric Reactor," WCAP-7964, June, 1971. 2,9. 2b Barry, R. F., et al., "The PANDA Code," WCAP-7048-P-A (Proprietary) and WCAP-7757-A (Non-Proprietary), January, 1975.

• 11, 1 England, T.

R., " CINDER - A One-Point Depletion and Fission Product t l Program," WAPD-TM-334, August, 1962. I l ( 4.3-67 ( l

R-@ U tao + y

• p ed F 4 2 n Raub ser j

l /,

16. L.c en,, e 3 Re g o c +

co j hfgi\\ C. 5 %m m e r h \\ ear M " b' 0 '~n, N R L Dockd Mc. 50 -3% D esembe r, 19 g3 l7 Green, Lucious, Petrie, Ford, White, Wright, PSR-63/AMPX-1 (code package), AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Librariec from ENDF/B, ORNL-TM-3706, Oak Ridge National Laboratory, March 1976. 18, 4x L. M. Petrie and N. F. Cross, KENO-IV, An Improved Monte Carlo Criticality Program, ORNL-4938, Oak Ridge National Laboratory, November 1975. I o 3, S. R. Bierman et al., Criticgl Separation Between l g} Subcritical Clusters of 4.29 wt% U 5 Enriched UO2 Rods in Water with Fixed Neutron Poisons, NUREG/ CR-0 0 73, Battelle Pacific Northwest Laboratories, May 1978, with errata sheet p issued by the USNRC August 14, 1979. w E 6 ZO 4 r M. N. Baldwin et al., Critical Experiments Supporting Close F4 Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, The Babcock & Wilcox Company, July 1979. l LI.5-- R. M. Westfall and J. R. Knight, Scale System Cross-Section l Validation with Shipping-Cask Critical Experiments, ANS Transactions, Vol. 33, p. 368, November 1979. 11 6 S. E. Turner and M. K.

Gurley, Evaluation of AMPX-KENO Benchmark Calcula tions for Figh Density Spent Fuel Storage

Racks, Nuclear Science and Engineering, 80(2):

230-237, February 1982. 13 & A. Ahlin and M.

Edenius, CASMO - A Fact Transport Theory Depletion Code for LWR Analysis, ANS Transactions, Vol. 26, p.

604, 1977. CASMO-2E Nuclear Fuel Assembly Analysis, Application Users Manual, Rev. A, Control Data Corporation, 1982. l

t.. ...t;$p.f.l, ; ' :.,-,,-. n - r a.... r.:.e ?.

z..

it ,. n ' ~; b.i..r,,.c %,'.,5';;. f..,? .:- u u.

7.,.

p.. - :.

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3. !!

h.

1. 9.6l' m %. i &$ * ' '

,y%

sty. -

/- w. s.:. 6 .s. a - .,f:?71. {.f f,., N.W f A  :.; : p.r.y.~ f f ?*.g. lll? *.'.f *:,. l ~ T- ? l- ..' :' M &. Eggleston,'F. T.pn"S.afety-Related Resear,ch and Development for.CJ 9 ' ' w - x.s. =,c..ucs:A .ti ' c , :.grcm:2. - . G u. Wes.;tingho,use.y.Pr..essurized Water, Reactors.,-u.ogram Summaries, Spring Pr 7- ~.. .n 1976," WCAP-8768, June, 1976. at,'24...Poncelet, C. G., " LASER - A Depletion Program,for Lattice Calcula-tions Based on MUFT and THERMOS," WCAP-6073, April, 1966.

3) '1Mi, Olhoef t, J. E., "The Doppler Effect for a Non-Uniform Temperature 3

Distribution in Reactor Fuel Elements," WCAP-2048, July, 1962. 33, & Nodvik, R. J., et al., " Supplementary Report on Evaluation of Mass Spectrometric cad Radiochemical Analyses of Yankee Core I Spent . Fuel, Including Isotopes of Elements Thorium Through Curium," WCAP-6086, August, 1969. 33,2A Drake, M. K. (Ed), " Data Formats and Procedure for the ENDF/B Neutron Cross Section Library," BNL-50274, ENDF-102, Vol.1,1970.

35. % Suich, J. E. and Honeck, H.

C., "The HAMMER System, Hetergeneous Analysis by Multigroup Methods of Exponentials and Reactors," DP-1064, January, 1967. h.2A Flatt, H. P. and Buller, D. C., " AIM-5, A Multigroup, One Dimen-sional Diffusion Equation Code," NAA-SR-4694, March, 1960. 3'7. '3% Moore, J. S., " Nuclear Design of We pinghouse Pressurized Water Reactors with Burnable Poison Rods," WCAP-7806 (Non-Proprietary), '~ December, 1971. 3S, % Nodvik, R. J., "Saxton Core II Fuel Performance Evaluation," WCAP-3385-56, Fart II, " Evaluation of Mass Spectrometric and Radio-( chemical Analyses of Irradiated Saxton Plutonium Fuel," July, 1970. Fueled Critical Experiments," 33 3%, Leamer, R. D., et al., "PUO -UO2 2 WCAP-3726-1, July, 1967. ( i 4.3-68

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n -%mg. s..Aftj 342 L'hMIMS - ~ Me ' e r,*- R 7...~;- O s t*MC.,,Ef Mt@d. ? "Th e ~ Analy s i s o f,'Fu e l ' De n s if ,3f D f

• M

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TABLg '4. 3-4 a' 7 %[ ' ,I ~ ? ~;j ' . ; c:.,., BENCHMARK CRITICAL EXPERIMENTS Description f Number of LEOP K,ff Using C Experiments (1 Experiments Experi ntal Bucklings UO2 (s... Al clad 14 1.0012 SS clad 19 0.9963 Borated H O 7 0.9989 2 Subtotal 40 0.9985 U-Metal Al clad 41 0.9995

• Unclad 20 0.9990 Subtotal 61 0.9993 Total 101

.9990 Rep ted in Reference 116]. C 4.3-75 1 .,...-,.-.v

4 Ta bi e. 4.3 - 4 i Reactivity Effects of Abnormal _ and Accident Conditions Accident / Abnormal Conditions Reactivity Effect Temperature increase Negative in Regions 1 and 2; positive in Region 3 Void (boiling e 248* F) Negative in all regions Assembly outside rack Negligible Assembly lying on top of rack Negligible Lateral rack module raovement Negligible l l l l t .n.

nlil,o:Il

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• 2 y "1

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fI

,ll

=_ [ _.~-l S_'~'... _ _.. _ _ _ _... _. _ ~ l-L= 40- _....:~.._ _:- r ? = _. -4 S = 35- =- A c t. rK-GKi-'- a = 30-3 2 u.._. m m &T:A-st. E. a. 3g m 25- =_ -- _ _r ~. _ ~.__T. :_ _ - . r_Tr __=_- o ~_=~~~ ~__~~iT- = =~~ ___._._ W - - - ~ _ _ ~ - ~ ~ m I 20-o m. o

== _. 5 15-1 =-[--[.j_ W m = = _ _ = _ _ _ - - 4 =. _ _ - ~ ~ - 10-5- ;- a-- .o_h --=__-==-t-- + - - , r -... _ -g

_= -

r__ ..___.E-_" = t_3- _;_/ - ===- ; 2-

== 0 i a l 2 3 4 INITI AL ENRIC H MENT, wt. % U-23 5 Fig.4.3 *Hb Limiting discharge fuel burnup for Region 3 storage rack for fuel of various initial enrichments. T

9,0 . AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 9.1.1 NEW FUEL STORAGE i 9.1.1.1 Design Bases New fuel is stored in racks (see Figure 9.1-1). Each rack is compcsed of individual vertical cells which are factened together in any number to form a module that is firmly bolted to anchors in the floor of the new fuel storage area.f The new fuel storage racks are designed to inci d storage for 1/3 core at a center to center spacing of 21 inches. This spacing provides a minimum separation between adjacent fuel assemblies of 12 inches which is sufficient to maintain a subcritical array even in the event the builoing is flooded with unborated water. Surfaces that coue into contact with the fuel assemblies are made of annealed austenitic stainless steel, whereas the supporting structure s may be painted carbon steel. C The racks are designed to withstand nominal operating loads as well as safe shutdown earthquake (SSE) and operating basis earthquake (OBE) seismic loads. The racks are Safety Class 2b and meet the allowable stresses in the ASME Code, Section III, Appendix 17. The new fuel racks are designed to withstand a maximum uplift force of 5000 pounds. l t i j T l [ he racks con 3 3t d b map d d r y ce b, enck 1^ i5 pcM ern. to.d es. n. 30 cd3 in g ,ue:==T 1B ( 9.1-1 gt, 337 l

9.1.1.2 Freilitie D:scriptien The new fuel storage facility is located in the eastern end of the fuel handling building (see Figure 1.2-6). The facility consists of a con-crete pit 27 feet deep by 15 feet wide by 28 feet long, covered by a removable grating. The new fuel storage racks are anchored to the floor of the facility and are designed to store fuel assemblies for 1/3 ^ of a core. 9.1.1.3 Safety Evaluation The design of normally dry new fuel storage racks is such that the effective multiplication factor does not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming optimum moderation (under dry, fogged, or flooded conditions). Credit is taken for the inherent neutron-absorbing effect of the materials of construction. The new fuel assemblies are stored dry and the 21 inch center to center g h, spacing ensures an eversafe geometric array. Space between storage U positions is blocked to prevent insertion of fuel. Under these condi-tions, a criticality accident during refueling and storage is precluded. 9.1.2 SPENT FUEL STORAGE 9.1.2.1 Design Bases Spent fuel is stored in racks (see Figure 9.1-2). Ea h ra ' is com l posed of in vidua vert cal ells ich e fa tene tog ther in y nu er to orm a odu tha is f ly b ted oa hor in e f oo of e spe fuel ool ] y [%gpent 8 fuel storage racks hr : cep-sli.y of 12/2 ;;;c ;;. e c;;.ter te .y-l / =nt:r sposius vi 14 lu sc The racks maintain a separation between spent fuel assemblies sufficient to maintain a suberitical array. (' l Space between storage positions is blocked to prevent insertion of fuel. Surfaces ~that come into contact with fuel assemblies are made of ( 1*""*"t 10 9.1-2 OMARFR. l}78 s' N i I k e-m w, _m

Insut A T 4---- The high density spent fuel racks consist of individual cells with 8.85" x 8.85" ( nomina'l) square cross section, each of which accommodates a single "ce L W~as:4 fuel assembly. The cells are arranged in modules of varying size. A total of 1276 cells are arranged in 11 distinct modules in 3 regions. Region 1 is n e.- c.- A reshly discharged fuel assemblies with designated for storage of f enrichments up to 4.3 weight percent U-235. The cells in Region 2 are reserved for accommodating fuel assemblies with initial enrichments of 4.3 weight percent U-235 and a minimum burnup of 20,000 MWD /MTU. The remaining cells, i.e. Region 3 cells, are capable of accommodating fuel assemblies with initial enrichments I of 4.3 weight percent U-235 and a minimum burnup of 42,000 MWD /MTU. shows the arrangement of the rack modules in tne e ru-r peci i- ~ ? r+ ~ e e ::it;d ;tr 2. >pe d b l N ol. scifons cf 47pkc.l F,p re 't.1 - 1 b

• hce h o r i tc o +cd c rcs,

33 po,3c.3ed ( R e y e n, t = ed 2-)

o. a d we p o s o o < d ( R e,y.. 3) cc\\\\

ac ro ys. TcM e. Ct. l - 1. pec e des l l l ~ " relevant design data on each region. The modules in the

i. i -3 three regions are of 4 dif ferent types.

Table SG2 summarizes the physical data for each module type. l The modules are not anchored to the pool floor, to each other, } or to the pool walls. A minimum gap of' 1-7/8" is provided between the modules to ensure that kinematic movements of the modules during the Plant Design Basis Earthquake will not cause inter-module impact. Adequate clearance with other pool hardware, eg. light fixtures, etc. is also provided. m = -,en

9.1.2.3 Safety Evalustfon ) Design of this storage facility ensures a safe condition under normal and postulated accident conditions. Consideration of criticality safety analysis is discussed in Section 4.3.2.7. Compliance with Regulatory Guide 1.13 is discussed in Appendix 3A. j i The center to center distance between the adjacent spent fuel assemblies is sufficient to ensure a k,gf <0.95 even if unborated water is used to fill the spent fuel pool. The design of the spent fuel storage rack assembly is such that it is impossible to insert the spent fuel assemblies in other than prescribed locations, thereby preventing any possibility of accidental criticality. The spent fuel cooling system is discussed in Section 9.1.3. 9.1.3 SPENT FUEL COOIING SYSTEM The spent fuel cooling system cools spent fuel pool water to remove decay heat from the spent fuel elements. It also provides for purifi-cation of the water in tha spent fuel pool and refueling water storage tank (RWST), for transfer of water between the RWST and spent fuel pool, maintains boron concentrations, monitors radiation levels and monitors and maintains spent fuel pool water level to ensure adequate cooling and shielding. 9.1.3.1 Design Bases The spent fuel cooling system is designed to perform the following functions: ( S4h cme, hecd 4.x< Mc.nc)er e pe.c aM t.) 0* k .16# 1. (faintain the spent fuel pool water temperature less than H? P-(, i l with a heat load based upon decay heat generation from 1/3 ef e 76 bd cw.4l.o hewc. esse that ions been irradiated for 24,000 effective full power hours p b"n irnLdad b (EFPH) and cooled for six days plus.,po 0..\\ c.we bt e ik.d ba** 2.. _ a : cooled for mese 2.tece EpN, l T cad, -tinee one yearf ee m e,t (, (, 9.1-4 rrurwr 14_

~' ) W;4h 4wo head ca. chem 3c() oper cM t. ) o g g p. 2. pintain spent fuel pool water temperature less than MO*F. with a heat load based upon decay heat generation from the f:'-- _r E'--t' - n - - - - mi 2 _..J.m. 16'1 L e( Esse nblie3 d d have, becq 2. 1l2 : _... = =unec = 18,vu - a 8 uw. c a: :

, :+

i/5. ; :::: ur2n 2t d r in Ann r m r3 :: 2 _- 7; m -1 --- 1/2 of w e irradiated for 2,400 EFPH and cooled for 6 days o ncd Ill I A_ m -i. .;: ci = cars n.= 2ted f or l':, vuu t r en 2nc===1--d iniW {,,) ass e,bIsr3 lb cd bun ben "/2 of a core irradiat.ed for 24,000 EFPH and cooled for = re then-3 e. one year, o r m o r e,, 3. Transfer water between the RWST and the refueling cavity. 4. Maintain the purity and clarity of water in the spent fuel pool and/or the refueling cavity at acceptable levels by means of skim-ming, filtering and/or demineralizing the water to remove corrosion and fission pr: ducts. 5. Provide means for adding boric acid to the spent fuel pool to main-tain the boron concentration at a nominal 2000 ppm. 6. Provide means to add decineralized water to the spent fuel pool to make up for evapotation losses. ) 7. Monitor spent fuel coolant for excessive radioactivity due to defective fuel elements. 8. Provide for filtering and/or demineralization to clean the water in the RWST. 9.1-5 2'C ='I f 14 (%. ' io79 dBNr

Also, there are no drain lines connected to the spent full p cl. Th3 required water shielding height of approximately 9.5 feet above the ) stored spent fuel assemblies is maintained even.in the unlikely event of pool syphoning from the discharge end of the return cooling lines. The suction and return lines are located on opposite sides of the pool to prevent channeling and to obtsin maximum circulation of the pool water. During fuel handling operations with the pool at the normal water level, a minimum shielding of 11 to 13 feet of water is available during the transfer of spent fuel elements. i Normal makeup water to the pool is available from the demineralized water storage tank. Emergency makeup water to the pool is available from the RWST, which is Safety Class 2a, or from the reactor makeup water storage tank (RMWST), which is Safety Class 2b. Pumps from three separate systems - spent fuel cooling and purification pumps; RMWST pumps; and demineralized water transfer pumps - are available for the transfer of water. The spent fuel pool water is cooled by either one of two redundant spent fuel cooling water loops. Each loop contains a pump, heat exchanger, piping, valves and instrumentation. The pumps can be powered from sepa-rate emergency (diesel) power sources. l 76 Cd asse bi;,3 %t M been The normal design basis cooling situation is 1^!? cf :- n.;_; -aft u-12 = fu11w l 1,-3 a irradiated for 24,000 EFPH and cooled for 6 days pk 12.0 0 Ge.l.* a. aw[nbf.9 'iked hcus been ' ---- irradiated for 24,000 EFFH and cooled for one year e r m o rt., 1/ core irr iated/for 24,00 EFPH and c oled or tw) yests D 4. 1/ co e ir adiat d for 24,0 EFPi and ooled for t ree ears. 5 3e e i radia d fo 24,0,0 EFP' and coole for four ears forffve r me re - /3 re radia ed f 24,0p0E i an cool ) year. 9.1-9 .mMr. 1979

0 This combination results in design basis heat load of x 10 ~ BTU /hr. The calculated operating temperature of the pool, considering only heat removal through one heat exchanger, is no greater than the (h design criterion of N. No credit is taken for the evaporative y lyo F cooling effect. ( is7 Cd a>u,bG e> A t L h b<m (14 The off-normal design basis cooling situation ie 13,3 mi ii r.R- .mm as== =. = : : m c..a - G e.( M x: irradiated for 24,000 EFPH and cooled for 106 days ple ll L *1 blie3 4 b e d he. n b e m

^ * irradiated for 24,000 EFPH and cooled for MI Q o.1c y < - be.i,3n b m4n p Cell Pitch Min. B o Flux Trap i Region (nominal) Loading Gap (nominal) 2 1 10.4025" .022 gm/cm 1.1605 2 2 10.4025" x 10.1875" .0015 gm/cm 1.2605 x 1.0455 3 10.116" unpoisoned 1.086 l w am t l l l l l 9 0,1 - 3 Tab \\ c. 1 S eni Fuel 5 4c ra3c. has - M ockle. Dcd4 p Approximate Region Module Cells per Array Weight No. Type Quantity module Size (lb/ module) 1 A 2 121 lixil 36300 2 B 1 99 lix9 28300 3 C 5 121 lixil 25500 3 D 3 110 .lix10 23100 l J --...n. J n u9 5 - se e e.< l' eg l l d V i i t! e J a , i a-e- e i9 h . j = .a =- 3* j ** w ) J_____

a w

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r=t=)Rs

l i e...,. 1 s _.... A ) g\\ e_ t Figure 9.1-2

1 $khx o4) 'N' N,/ / / N N k/ / %/ / 'N' N N/ / N / 'N/ K x // s g x / ,/ N s s SOUTH CAROLIN A ELECTRIC & gas CO. vsRGIL C. SUMMER NU CLEAR STATION. FIGURE 91-2 HIGH DENSITY SPENT FUEL STORAGE RACKS,

ur = u_- u-120 w -tm w us 32 mgm e ja c}w REF. h h i a '9 / L* REGION 3 REGION I e REGION 3 RESERVED = E N p 9 ll X li ll X 11 Z ll X ll / AREA [ i p / o a . i u 9 he d REGION 3 REGIO N I e REGION 3 REGIO N 3 i-o o O ll Xll 11 X ll Z ll X ll 11 X 11 I ' co n E Lm n# =p, i', c c. o 4 mico 8 REGIO N 3 REGION Z i REGION 3 RE6IO N 3 g O 11 X l O 11 X 9 Il X 10 11 X 10 9 =m u I _3 I U p et h t

)

il4f" *g 11 0 [ 110$ 110 a 27 27" 7T = = = " ^7 3 REF. 3 " =.. 7 7.. 7 g g-1 '8 1'8 39-0 = = S ed L\\ S h hk - ModA Layod F,pe c 9.1 -la p y

I MI i ~ ~ a e f n / \\ p s l \\ p + + p g y I / f l / / p d d k k. .;d c k .~ i + l l o l p l / / / / A c-;, O .+ i e a' l /i ' I / s j i s t t s Poison e d [. q L., POISON INNER BOX OUTER BOX MATERIAL 3 I 7 1 1 4 s 1 l I k i e i i S 1 e l l e I i unpo.soned 4-l I I I O L -L _L e g 4 1 F.p re. 'i. l ' 2. b 5 e.d Fud sioc3 Rmd Celi A e c y., Typ'.c.d 3O p w.}}