ML20133H638
| ML20133H638 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 10/15/1985 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20133H641 | List: |
| References | |
| ULNRC-1192, NUDOCS 8510180051 | |
| Download: ML20133H638 (14) | |
Text
aus UNON RECHNC COMMW 1901 Grotiot street. St. Louis Doncid F. Schnell Vice President October 15, 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Denton:
ULNRC-1192 DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 REVISION TO TECHNICAL SPECIFICATION FIGURES 3.9-1, 5.6-1 AND SECTIONS 3/4 9.12, 5.3, 5.6 CONCERNING SPENT FUEL POOL STORAGE Union Electric herewith transmits three (3) original and forty (40) conformed copies of an application for amendment to Facility Operating License No. NPF-30 for Callaway Plant, Unit 1.
This amendment request replaces Technical Specification Figure 3.9-1 with curves that represent criteria for storing either Westinghouse optimized fuel or standard fuel in Region 2 of the spent fuel pool. This is necessary to accommodate the Callaway reload design change from Westinghouse standard to optimized fuel. In addition, the maximum initial enrichment limit for storage in the pool is increased from 3.5 w/o U235 to 4.2 w/o U235. This change is reflected in Technical Specification Sections 5.3.1 and 5.6.1.1.a. Due to the substantial margins to criticality limits based on the original criticality analysis of standard fuel at a maximum initial enrichment of 3.5 w/o U235; and due to improved analysis techniques, as built spent fuel pool rack dimensions, and the optimized fuel design, a criticality re-analysis justifies extending the pool storage limit to a maximum initial enrichment of 4.2 w/o U235 fuel. Technical Specification Section 5.6.1.1.a is also revised to reference Figure 3.9-1 instead of Figure 5.6-1. The same proposed curves would replace Figure 5.6-1 as well as Figure 3.9-1. To avoid unnecessary duplication the reference in Section 5.6.1.1.a will be made to Figure 3.9-1.
Technical Specification Section 5.6.1.1.b is revised to give the center-to-center distance between fuel assemblies placed in storage racks as 9.24 inches nominal. This dimension change is based on an approved design drawing that was issued subsequent to performance of the original criticality analysis. The current value is verified to be a typical measured as-built dimension.
0510180051 851015 PDR ADOCK 05000403 PDH P
gl Momng Address: Ro sox u9. st. Louis Mo 63166 ' w) ebA Y lWh_ _ Y ___ . _-- _WA _ _
In addition a criticality re-analysis was performed on the new fuel storage racks. The analysis verified that reload fuel with the Westinghouse optimized design and with enrichments up to 4.2 w/o U235 can be safely stored in the new fuel storage racks
- without exceeding criticality safety limits. The new fuel storage racks however are not included in this amendment request since new fuel storage rack revisions are not required to the Technical Specifications.
Union Electric's date of February 1, 1986 for reload fuel on site and the effective date for implementation of the proposed Technical Specification changes are subject to NRC approval.
Enclosed is a check for the $150.00 application fee required by 10 CFR 170.21.
Very truly yours, J
\- -
s Dona ^d ri' chnell DJW/bjk 6
Enclosures:
1-Safety Evaluation 2-Significant flazards Consideration i
3-Marked Technical Specification Pages i
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STATE OF MISSOURI )
Robert J. Schukai, of lawful age, being first duly sworn upon oath says that he is General Manager-Engineering (Nuclear) for 1 Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information ar.d belief.
O
[
By Robe-t J rSdhuRai Gene 11 Mancler-Engineering g
Nuclear SUBSCRIBED and sworn to before me this /5 day of , 198 f tW hf/
DAR3 ARA J. PfAkf 8 8 fi0f AV PJDUC, STAf[ Of MISSMt MY C0"M5IC1 EXP!E5 AFRIL 22,106)
ST. LOUIS Cout4Tf
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l cc: Gerald Charnoff, Esq. '
, Shaw, Pittman, Potts & Trowbridge 1800 M. Street, N.W.
j Washington, D.C. 20036 Nicholas A. Petrick .
Executive Director ,
4 SNUPPS
! S Choke Cherry Road i Rockville, Maryland 20850 G. C. Wright Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission 2 Region III i
! 799 Roosevelt Road
- Glen Ellyn, Illinois 60137 Bruce Little l; l Callaway Resident Office
, U.S. Nuclear Regulatory Commission ,
i RRil '
Steedman, Missouri 65077 ,
Tom Alexion ;
- of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
,! Mail Stop P-316 4
7920 Norfolk Avenue
- Bethesda, MD 20014 Ron Kucera, Deputy Director Department of Natural Resources i P.O. Box 176 Jefferson City, Missouri 65102 !
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J Enclosure 1 l ULNRC-1192 l October 15, 1985 Page 1 of 7 l
SAFETY EVALUATION Callaway's initial core contains Westinghouse Standard Fuel Assembliec (SPA) . The Callaway roload will contain Westinghouse Optimized Fuel Assemblics (OFA) . In addition, the fuel management plan is to normally utilize 18 month fuel cycles which nominally requires about 80 new feed assemblies per reload, instead of a third of the 193 assembly core. Baced on these j facts, a reevaluation of the spent fuel pool design bases wan l performed to assess the impact of changen for the OFA design and
! for the 18 month cycle fuel management plan.
i l The evaluation included calculations that verified (1) the spent fuel criticality limits are maintained and that within the
- existing safety margins storage of higher initial enrichment fuel I to 4.2 w/o is permiccible; (2) the thermal-hydraulic, structural, l and seismic responso design bacon of the fuel rack / fuel j pool / cooling system are met; and (3) any environmental or j radiological aspects are within the bounds of the licensing bacia j for the plant.
l As a result of these ovaluations, this amendment requests j that Technical Specification Figuros 3.9-1 and 5.G-1 be replaced with curves reprocentative of both OFA and SFA fuel; that the 1
maximum initial enrichment limit for storage in the pool ic
, increased from 3.5 w/o to 4.2 w/or and the contor-to-conter i distance between fuel accomblica in storage racks be revised to 9.24 inches in accordance with the approved design drawing.
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! Descripti__on of _the _ Cal _laway Spent Fuel Pool "
The Callaway spent fuel pool utilizes the maximum density i rack (MDR) design concept. Under this concept, the opent fuel l pool is divided into two neparate and distinct regions which for .
j the purpoce of criticality considerationo may be considered no l ceparato pools. Suitability of this design accumption regarding ,
pool separability is accured through appropriato design i restrictions at the boundarios between Region 1 and Region 2.
The smaller region, Region 1, of the pool is designed on the basis of conservative criteria which allow for the cafo storago of a number of fresh unirradiated fuel acaemblien (including a full core unloading if that should prove necoccary). The larger i region of the pool, Region 2, la designed to cafely otore i
! irradiated fuel annemblion in large numboro. The only change in
+ critoria betwoon Region 1 and Region 2 is the recognition of i actual fuel and fission product inventory accompanied by a nyatem i for verifying fuel burnup prior to moving any fuel acaembly from i Rogion 1 to Rogion 2. In both Region 1 and 2, ouberiticality ,
(Koff < .95) in maintained during all normal, abnormal, or accident conditionc.
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Unclocure 1 1
ULNRC-1192 October 15, 1985 i Page 2 of 7 The spent fuel pool ic a reinforced concrete ctructure with 4 a stain 1ccc steci liner. Fuel storage rack modules are constructed with square boxcc which form a honeycomb ctructure.
The rack modulec are frccatanding on the floor liner plate of the pool. The pool in filled with borated water with a boron i
concentration of 2000 ppm. The fuel pool cooling and cicanup cystem concictc of two 100 percent capacity cooling trainc. Thic cystem functions to limit the pool temperature to 135*F with one ,
train operating during normal plant conditions; removec impuritice for vicual clarity; and limits the radiation doce to operating personnel during normal and refueling operationc.
Criticality Analycic The phycical characterictica of OPA and SPA fuel acccmblica are cimilar. Both designa employ 17x17 fuel rod arraya and the fuel rods are zircaloy clad. The ora decign, however, utilizcc a cmaller fuel rod diameter with chamfered pelletc and employc i zircaloy rather than inconel mixing vano spacer gridc. For given initial enrichments, theco decign changcc result in different recctivity characterictics both for unirradiated fuel acccmblics and ac a function of burnup. In addition, the ora reload decignc will utilize initial enrichmente greater than the 3.5 w/o accumed for the original criticality analycic. For thece reaconc, a new criticality cafety analycic was performed to demonstrate that the
, OPA design may alco be cafely ctored in the Callaway opent fuel 4
racks, and to determine the maximum initial enrichment allowabic for storage within the margins to criticality cafety limito. The new criticality analycic wac performed ucing the came calculational methodology, computer coacc, and crocc acction libratica ac uced in the original analyclc. The calculational J method used the Lt0 PARD transport theory code for crocc-acction l generation, the CINDER code for a detailed analycic of ficcion j product buildup and contribution to the abcorption croco coction, and the PDQ7 dif fucion theory code for criticality determination.
As diccucced in the original analycic these codes, their crocc j
coction librarica, and the methodology of their uno han been verified by comparicon to mencured criticalc and measured juct critical accomblica cciccted oc cppropriate for benchmarking for spent fuel rack criticality calculationc. In addition, the benchmarking approach included ceparately comparing both the transport theory and diffucion theory calculationa againct applicabic critical accomblicc. Thuc the calculational differencoc betwoon transport and diffucion theory are impilcitly included in the derived calculational uncertainty factor.
- Calculational blaccc and uncertaintica were determined ac in the 4
original calculationc.
on geactivity of variationeCalculational blocco include in the pool temperature theG0 from cr{cct P to 160 r; calculational modeling accumptions, including calculational mcch clze; and the effectc of not axial leakage.
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l Enclosure 1
- I ULMRC-1192 l October 15, 1985 i Page 3 of 7 l These blaccc are summed and added to the basic result gf the !
- calculated multiplication factor of the problem. ToleranceA and i uncertaintics are also concidered. The tolerances and i uncertaintics include manufacturing uncertaintica cuch as fuel i
, rack storage cell dimoncion and wall thickness, uncertainty of ,
j fuel position within the box, variation in fuel doncity, !
calculational methods and model uncertainticc, and depicted fuel reactivity uncertaintics. Thecc tolerancoc and uncertaintics are combined statictically and then also added to the result of the basic problem multiplication factor. Thus the maximum t
- multiplication factor calculated for the variouc probicm ,
configurations using the most concervative inputs (for exampics a !
! Cuc1 pocition yiciding the minimum pitch betwoon accomblics, and !
l' the maximum fuel pc11ct density) is adjusted for the addition of 7 the total biaccc and total uncertaintica, and is then compared -
againct the critical limit of Kef f < .95. !
Region 1 calculations determined the reactivity for an
[ infinite array of unirradiated 4.2 w/o accomblics arranged in the f j Region 1 rack configuration. Region 1 maximum Koff including :
biaccc and uncertaintics for the reanalysis was .9307. No credit l i wac given for the 2000 ppm borated water in the pool, however, i I including thic reactivity effect would reduce the maximum Koff to i
.6875 for Ecgion 1. Region 2 calculations were baced on '
! reactivitico determined at 36,000 ftWD/MTU for 4.2 w/o acccmbile in the Region 2 rack configuration. The maximum offective j j multiplication factor, including total blocos and uncertaintica, !
I wac determined to be a value of .0720. In addition to mcoting !
the criticality limit, curvcc were developed for both ora and SFA i
, to provido criteria for celecting which bund 1cc in Region 1 could l be trancforred for storage in Region 2. In thic cace, the SPA ;
i curve was redeveloped to account for reviced rock dimencionc and i 1
to uce improved ctatistical methods for combining the !
! uncertainticc. The original criticality analycic wac performed prior to the design change of incroacing the storage rack cell ,
'j incido dimoncion. The reculting curven were developed by an i iterativo prococs of calcu..' ting the reactivitics for variouc i initial enrichmento at varicus expocuro points for both ora and I j SFA fuel. Ac determined in the analycic, the total uncertainty [
j on the reculta ic compriced largely of the calculational
$ uncertainty neccccary for a 95/95 confidence ctatement, the ;
i depicted fuc1 calculational uncertainty, and the analytic model !
blac. Thece componente comprico more than 00 percent of the j total uncertainty. Since the came calculational methode were i i employed for both fuci decignc, the magnitude of thoce threc l
- componento are the came for both fuel decigna and the came value !
4 of the multiplication factor can be uced for both fuel typoc an a l
! criterion to accure compliance with the 0.95 regulatory limit. !
i The total uncertainty, along with the calculated multiplication !
factor, was uced to determine the value of the multiplication l
- factor that would coticfy the 0.95 regulatory limit at the 95% j
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Enclosure 1 ULNRC-1192 October 15, 1985 Page 4 of 7 1
l confidence icvel. This computed value was .9203. However, to 1
account for interpolation errors in developing the curves, a i maximum target value of 0.9150 was used. Using this maximum j sciccted limit of .9150 (below the limit of .9a), the various j cnrichment/burnep curvca were used to interpolate or extract j points having multiplication factors below .9150. These f amiliec
, of curves were used to generate the proposed curvec for uce in
! the Technical Specificationc.
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! The new analycic, therefore, verified that fuel, up to 4.2
( w/o initial enrichmentc, can be cafely stored in the spent fuel
{ pool without violating the criticality cafety limit.
i Thermal-nydraulic and Structural Analysis A review of the thermal hydraulic and structural design l bacco of the fuel rack / fuel pool / cooling system was performed to verify that the change to OFA fuel and an 18 month cycle did not advercely affect thoco systema. The OFA atructural chcngcc as diccucced above did not impact the thermal hydraulic 3 calculations; however, the change to an 18 month cycic with higher accombly burn-upc and greater numbers of acccmblica por
) reload required reanalycic be performed for the design basic thermal / hydraulic and structural cacca. Thic reanalycia utilized
! the original methodology.
i l Two thermal / hydraulic caccc were reanalyzed: Cace 1 i concidered the responce of the opent fuci pool af ter a refueling j off-load, and Caco 2 analyzed a refueling of f-load followed by a
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full coro dicchargo. Thece caccc and their acceptanco criteria are identical to the original cace doccribed in Section 9.1 of the FSAR. The calculations included the following accumptionat I 1. OFA Fuel operated continuously for 1288 days at 100 j percent power followed by 0 day refueling period.
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- 2. A nominal 80 accombly diccharge por refueling (84 acccmblics were analyzed for diccharge f rom Cycle 1) .
- 3. 51,000 MD/MTU burn-up for all diccharged fuel l 4. Only one train of fuci pool cooling availabic with heat i exchanger performance based upon ASME Code Section 2I, j Clace 3; TEMA, paragraph n (Tubular Exchangero j Manuf acturcrc Accociation) ; and limiting component j cooling water conditionc.
l i 5. Decay heat calculated in accordance with nucicar i Regulatory Commincion Standard noview Plan (NUREG-800) j Branch Technical Pocition ASD 9-2.
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Enclosure 1 ULNRC-1192 October 15, 1985
- Page 5 of 7 A computer program was developed to implement this methodology. This computer code was validated to the Standard Revicw Plan decay heat correlation.
l Soncitivity ctudica were performed to determine the moct conservative refueling times. The most concorvative models and parameters were used in thg final cafculation. The original i design requiremento of 135 F and 160 F for both caccc were met.
It chould be noted that these conditions are more stringent
, acceptance criteria than 140*F for Case 1 and no boiling for Cace
- 2 in the Standard Review Plan. The accumptions used in both cases are very concervative and the probability of all accumptions applying at the came time ic very low. Therefore, l Union Electric cccc no probicmc with the calculated difference between the calculationc and their acceptance criteria.
! To verify the structural decign of the spent fuel pool, a third cace was analy:cd using the conservative methodology developed for the thermal / hydraulic calculations. Thic cace accumed non-operation of the fuel pool cooling system for two
! hourc after a discharge identical to Cace 2. The acceptance critoria uged for Cace 3 ic the fuci pool liner plate chall not exceed 240 F. These calculgtlons confirmed a fuel pool liner plate temperature below 240 F (186*F calculated). In addition, analyccc demonctrate that for the hotect fuel acccmbly, no bulk boiling occurred within the coolant channel.
i Scicmic Analycic An evaluation of the exicting colcmic analycic wac l performed. Concidering both the vertical and horizontal loada
'l for the Callaway Fuc1 Pool Decign, the vertical load ic the limiting load. Because OFA fuel weighc 15% Iccc than SFA fuel the vertical load in reduced. Concequently lower coicmic ctrecccc are realized and the original analycic ic conservative.
The frequency of the acismic analycin was alco evaluated. The lower weight of the OFA accomblice 1cada to an inctcaco in the fundamental rack frequency and cince the floor cpectrum decreaccc with frequency the original calculations are conservative. !
Environmental Evaluation ]
A change to the OFA fuci decign and to nominal 80 acccmbly I reloads on an 18 month cycle docc not alter the normal
! performance of the fuci pool cicanup cystem or the fuel building j ventilation and radiological control cyctemc. Calculationc were performed to verify that the propocod changoc do not alter the
, concorvatiamo procent in the original analycon. Decign limits on pool temperaturoc, pool structural integrity, and criticality cafety limite are not altered. A review of the poctulated j accident scenario of a dropped bundic (maximum reactive OFA
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. Enclosure 1 i ULWRC-1192 i
! October 15, 1985 [
j Page 6 of 7 [
} I assembly) and total inventory rolcace in the fuel building was performed. Calculations of the doses to the public were l, performed and compared to regulatory 11mitc. These calculationc [
j verify that the dose calculations to the public are well within [
! the required limits of 10CFR100 (1.3 Rem Thyroid and 0.05 Rom i j Whole Body). In performing this calculation, Fuel Building !
{ Emergency Exhaust Filters with an iodinc officiency of 90% werc .
j accounted for. Original calculations for standard fuel found in !
i the FSAR, Table 15.7-0 did not include thic accumptions :
j therefore, the doces for the standard fuel are greater than the j doccc calculated for the optimized fuel. '
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l Mixed Storage of OFA and SFA Fuel i i i The original analyses provided the accurance of compliance !
j to criticality limits of the SFA fuci design. Since the ;
i reactivity of the unirradiated OFA design is always greater than i the unirradiated SFA design, and cince the criticality analycoc
( assume infinite arrays of unieradiated 4.2 w/o OFA fuel in the ;
j radial planc for Region 1, the criticality analysis bounds any ;
i combination of unirradiated SPA and OFA designs with initial !
enrichments up to 4.2 w/o. :
i Compliance with the new burnup/ initial enrichT.cnt limit i i curvcc for cach fuci design accurca that the placement of a !
I bundic in Region 2 will not exceed the criticality limit. !
! Compliance with the curvoc would bound any combination of OFA or !
l SFA fuel in Region 2 of the pool.
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} The ef fect of a mix of OFA and SPA accomblice on the cpont !
- fuci pool thermal-hydraulic analycic ic negligibic. Standard i l
fuc1 accomblica will be discharged to the opent fuci pool carly i in plant life. The cpont fuci pool limite are only reached much i later in plant life when the pool in nearly filled to capacity. !
Near end of plant life the diccharged ctandard accomblice will l have decayed to cuch a Icyc1 ao to make the expected contribution I to the total heat load negligibly different from that of a [
comparabic optimized fuel accombly. In addition, becauce of the !
low 1cyc1 of decay heat in a ctandard accombly near end of plant !
life, the cooling of a standard accombly will not be compromiced l below that of an optimized fuel accombly even during locc of cpont fuel pool cooling. !
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} Conclucions The above cummarice of the evaluations of the propocod ;
j Technical Specification changcc indicato that the OFA fuel may bc ,
cafely stored in the cpent fuci pool without exceeding deofgn limite or climinating safety marginc. Phycically, the two fuci ,
decignc are cimilar and utilize similar materialc. OFA fuel in l i
geometrien11y compatibic with SPA fuel and the fuel ancombly f
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Enclosure 1
- ULNRC-1192 l October 15, 1985 '
- Page 7 of 7 i
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l j dimensional envelope, skeletal structure, and internal grid i locations are essentially the same. The major structural i l differences are a smaller fuel rod outer diameter and zircaloy
- spacer grids in the OFT design. Dased on those discussions, the !
proposed Technical Specification changes do not adversely affect or endanger the health or the safety of the general public, and ,
j does not involve an unreviewed safety question. !
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October 15, 1985 Page 1 of 3 t
)i SIGNIFICANT !!AZARDS CONSIDERATION ,
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! This amendment request replaces Technical Specification i Figuroc 3.9-1 and 5.6-1 with curves that reprocent criteria for j storing either Westinghouse Optimized Fuel Assemblica (OFA) or -
- Standard Fuel Assemblics (SPA) in Region 2 of the spent fuel ;
l pool. The maximum initial enrichment limit for storage in the i pool is incroaced from 3.5 w/o to 4.2 w/o. This change is ;
I reficcted in Technical Specification Sectionc 5.3.1 and j 5.6.1.1.a. Section 5.G.1.1.b is revised to give the conter-to- L i center distance between fuel accomblics placed in storage racks l l as 9.24 inches nominal. This dimension change is based on an
- approved design drawing.
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The Safety Evaluation cupporting this amendment requent ;
! provides the bacco for concluding that the proposed changes are j l concictent with the licencing bacca of the opent fuel pool and l 1 verify that the changen do not alter safe operation of opent fuel
' l pool cyctems nor violato pool criticality cafety limits. The l l' reevaluationc further demonstrate that an incroaco in maximum ,
initial enrichment for storage can be up to 4.2 w/o and that a '
l reload cize can be extended to nominally include 80 acccmblico j j without a cignificant reduction in a margin of cafety. As I i discucced in the Safety Evaluation an 84 anacmbly discharge for j the thermal-hydraulic evaluation was analyzed for Cycle 1.
{ Further, cince the criticality cafety analycic and the thermal-j hydraulic and structural analycic confirm that original criteria i
are mot by both OFA and SPA fuci, the poccibility of a new or different kind of accident or condition over previous ovaluations ic not credibic. Physically the two fuci designs are cimilar.
OFA fuel is geometrically compatibic with SPA fuel. The fuel accombly dimoncional envelope, ckeletal otructure, and internal grid locations are cccentially the came. The structural differences for OFA fuel is a smaller fuci rod outer diamotor and zircaloy cpacer grida rather than inconel. Ucutronic differencoc 4 betwoon the two fuci designs have been analyzed and determined i not to alter opent fuci pool criticality cafety limits. In l cacence, the Technical Specification changoc result from a nucloor reactor core reloading where the reload fuel acccmblice are not cignificantly different from thoco found previously acceptabic to the NRC. In addition, the WCAP 9500A (May 1982)
I cots forth the Optimized Fuel Design and has boon reviewed and
! approved by the NRC.
In ovaluating the incroaco in probability or concoquencon of any previously analyzed accident the following cconarion wore considered.
Enclocure 2 ULURC-1192 October 15, 1985 Page 2 of 3 Accident Scenarion The original accident and hazard scenarios were reevaluated concidering the OPA fuel design and using the 4.2 w/o initial enrichment. The scenarios consider dropping a fuel accombly on top of the rackc; dropping a fuel acccmbly which penetrates the rackc and occupice a position other than a normal ctorage
, location; dropping the fuel cack or other heavy objects into the pool; and the effects of toronado or carthquake on the deformation and relative position of the fuel racks. The loca of cooling cyctemc under accident conditionc was not considered cince the Callaway one hundred percent capacity redundant cooling trains provide a single failure proof cyctem.
In the case of a dropped fuel bundic lying on top of the racks, the only negative impact on criticality limits would be a i
reduction in the axial neutron leakage from the rack. Ac demonstrated in the criticality aralycic, the contribution of the axial neutron leakage only clightly reduccc the value of the maximum effective multiplication factor. For exampic, calculatienc chow the total axial neutron leakage blac to be only
.0022AK. Without any axial neutron leakage, the multiplication factor of .9387 for Region 1 would be clightly increaced to .9409 which ic ctill within the cuberitical limit by an acceptabic margin. Since there are coveral inches between the top of the active fuel and the top of the storage racks the reduction in axial neutron 1cakage from the rack chould be negligibic.
The spent fuel rackc include ctainlecc ctcc1 standoffs which maintain a spacing of at icact 3.0 inchec between the racks and any fuel accombly which might be inadvertently located immediately adjacent to a rack. The ctandoffs limit the reactivity increase in this event to a few tenthc of it A K. In addition, for thic abnormal condition credit may be taken for the colubic boron in the pool which hac a negative reactivity worth about .24746K. Since the pocitive reactivity would only be a few tenths of 1%AK and clnce the colubic boron contributen about
.2474LK negative reactivity, the criticality limit in not exceeded and pool cubcriticality in maintained.
The ef fectivenecc of neutron moderation can be charatterized by the moderator-to-fuct ratio as a function of water dencity, fuci doncity, and fuct lattice geometry. Plots of the moderator-to-fuel ratio againct multiplication factor yield a curve that give an optimum rctio for the effectivenenc of moderation. Moderator-to-fuci ratloc to the left of thic optimum are referred to an undermoderated. For an enrichment of 4.2 w/o, the lattice of fuel acccmb11cc recults in an undermoderated configuration and any cruching or compaction of the fuct accomblica would tend to l reduce the modcrator-to-fuct ratio and the multiplication factor i i
of the cpent fuci pool. Since the SPA fuct han a larger fuci rod i
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i Enclocure 2 i ULMRC-1192 i October 15, 1985 >
} , Page 3 of 3 I I i
) diameter, the lattice of SFA acccmblics is even more '
undermoderated than OFA fuel. Plots of Region 1 and Region 2 i
, multiplication factora versus water doncity further demonstrate !
i that a reduction in the water / fuel ratio chows a decreasing t I multiplication factor. Therefore, the dropping of heavy objects
- into the pool or deformationc from the effects of carthquakes or f
l 4 tornadoes could not produce a criticality accident. In addition, !
1 due to a lighter fuel weight for OFA fuel the sciamic analycio !
confirmed the bounding nature of the original analycic for an -
carthquake event. i l i I Finally in addition, as done in the original analysis, an i j accident cconario was analyzed in which a fuel acccmbly from ;
j Region 1 of the apont fuel pool wan incorrectly transferred to i
- Region 2. The new analycic accumed the extreme case of l j completely loading Region 2 with unirradiated 4.2 w/o U-235 fuel l I acccmblics. For thic abnormal Region 2 condition, credit was i i taken for 2000 ppm colubic boron. The resulting multiplication [
j factor for this case was 0.8905 and below the 0.95 criticality- .
j limit. !
i j Decign Conservatisms !
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{ As discucced in the Safety Evaluation, decian conservatisme !
were used in the criticality and thermal hydraulic analysec. In 1 brief cummary, the calculations uced the conservative accumptionc *
] accumed in the original analysec. The criticality cell i 4
calculations accumed that the fuci pool water contained no boric !
- acid. Scncitivity ctudies were performed to determine the mont :
concervative refueling timec. The moct conservative models and '
paramotors were uced in all final calculationc. i j Conclucion 4
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- hic amendment requect does not involve a cignificant f incroaco in the probability or conacquence of an accident or '
other adverce condition over previouc evaluations; or create the poccibility of a new or different kind of accident or condition over previouc evaluationc; or involve a significant reduction in a margin of cafety. Daced on thic information, the requested ;
licence ar.cndment doce not precent a cignificant hazard. ,
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