ML20148P769

States That Staff Has Reviewed GE Proposal W/Regard to Burnup Allowances & Fuel Storage Limits Based on Certain Values.Recommends Issuance of Amend W/Listed Lic Conditions. W/Encl Safety Rev of GE Responses to Questions

ML20148P769
Person / Time
Site:	07001308
Issue date:	10/20/1975
From:	Stevenson R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
Shared Package
ML20148P758	List: ML20148P769 ML20148P783
References
NUDOCS 7811290039
Download: ML20148P769 (8)

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. NUCLEAll HEGULATORY COMMISstDN W ASHING TON, D. C. 20555 OCT 2 0 1975 Note to Files GENERAL ELECTRIC COMPANY (CE), MORRIS, ILLIN0IS, REVISED FUEL STORAGE AT MIDWEST FUEL RECOVERY PLANT (MFRP), DOCKET 70-1308, SEPTEMBER 12, 1975 The subject submittal provides responses to the nuclear criticality safety questions that were raised following receipt of the initial application from General Electric. This note summarizes the nuclear criticality safety review of the GE responses and should be considered a supplement to my note to files dated July 23, 1975. The earlier note gives a general description of the GE proposal, which will not be repeated here. A summary of our review of'the GE responses is given as an attachment,

, The major problems with the initial application concerned the proposed use of a burnup allowance and the use of fuel storage limits based on maximum k. values. The two proposals implied a prim,ary importance for administrative controls in assuring the nuclear criticality safety of the stored fuel. However, our subsequent computer calculations check the applicant's calculations which indicate that fuel types more reactive than would be allowed by the limits would be suberitical even in the absence of burnup. The importance of GE's administrative controls is thus diminished and the need for a burnup allowance put into question.

CE has presented graphs of k as a function of fuel enrichment and reactor type and correction factors for the main va'riables affecting the k . Our calculations have confirmed the validity of the limiting cor-rection factors and the nuclear criticality safety of the stored fuel when the limiting km values are met. The technical effort needed to con-firm the k of the incoming fuel to the storage basin has thus been reduced to the operation of determining fuel parameters, multiplying the appropriate factors, and comparing the km with one of the three limits.

Our confirmatien of the nuclear criticality safety of the stored fuel and the simplification of GE's review procedure for k. provide satis-factory resolution of the main nuclear criticality safety questions on the GE application. It is therefore recommended that a license amend-ment be issued in which the questions on burnup and other potential questions are obviated by license conditions:

1. to limit fuel stored Lo that which meets the k. limits without burnup, 7 8112 9 0 osq

2. to exclude changes in the design or ma.terial of fuel racks or c'anisters that could reduce the nuclear criticality safety,

3. to limit fuel stored to that covered by the range of the parameters defined by the graphs (Figures F.1-1A and F.1-1B) and the correction factor for stainless steel cladding, which shall be at least 0.0165.in. thick, and

4. to prevent the use of containers or devices that would invalidate the safety analysis through replacement of isolating water by less effective isolators such as air or aluminum.

In addition to the foregoing conditions, the recommendation of approval of the application is also subject to the concurrence of the structural reviewer: ,

5. that the device to prevent dropping of fuel from a basket into the cask unloading pit will perform its intended function,

6. that the air equivalent drop for the fuel basket is about 12 ft., and

7. that the basket drop in the cask unloaditrg pit will not permit assemblies to extend more than 42 inches (fueled section) beyond the pipes.

Based on a discusrion with F. M. Empson on October 2, 1975, items 6 and 7 have aircady been confirmed by the structural reviewer. If the applicant does not agree with proposed Condition 1, questions on burnup which should then be resolved are given in Item 2 of the attached Record of Nucicar Criticality Safety Review.

Attention is also directed to assumptions concerning the fulfillment of certain general quality assurance, structural, and inspection requirements identified in Item 20 of the attachment.

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, R. L. Stevenson ,

Fuel Cycle Licensing Branch 1 Division of Fuel Cycle and Material Safety

Attachment:

Record of Nuclear Criticality Safety Review Submittal Dated 9/12/75, Docket 70-1308 g/ cc : P. H. rnnenn

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RECORD OF ?!UCI,EAlt CRITICALI1Y SAFETY REVICW Of GM RESPONSES TO _

QUCSTT0!!S ON NtorOSED RRVISION OF MFRT F001.: Int:l'CT 70-1308 Snhmittal Dated September 12, 1975

1. Check of storage array nafety when storate criteria are met. (No allowance for buronp was made in any of the ilPC calculations reported in the following.)

a. Our calculated keff for an infinite, water-immersed single-tier artay of 15 x 15 type, 3.6% enriched, mirconium clad pWR assemblies in the 12 inch diameter, Schedule 5 SS pipes in groups of.four on 27 inch centers was 0.9508 at the 95% confidence limit.

Larger margins of suberiticality were found when calculations were made for 4.1% enriched, 15 x 15 type, SS clad PWR assemblies and for 14 x 14 type FWR's of the smaller width (7.78 inch or 8.12 inch) at 3.6% enrichment.

b. Our calculated keff for a water-immersed 200 x 200 x 1 array of BWR assemblico at 3.2% enrichment in the stainless steel pipes in contact won 0.7001 and the same array was subcritical (keff 0.846

+ 0.00615) when the SS pipes were replaced by water. It does not appear that any current type of BWR assembly would present a critical-Jty hazard whon stored in the 8 inch SS p3 pes, with or without burnup.

2. Adequacy _of burnno allowance and associated administrative requirements.

Calculations repotted by the applicant and.ont indopendent calculations summarized under Item 1, foregoing, indicate that usd of a burnup allow-ance is not necessary to establish the nucicar criticality safety of nearly any f uel type for which storage is now planned. It is therefore recommended that the 1Jcense ha conditioned to limit the stored fuel to assemblies meetinn Criterion 2 without a burnup allowance. This approach obviates thn n<#d to obtain nnd evaluate answere to the following quentinie -

a. '1he elfectn of burnup incJude credit for fisalon product buildup, althonnh the footnote on page H.13 of the nattelle report (" Criticality sar"ly un n .l t, l ot t h" tiritP P r oj"c t -- J roe) lunulle Storage Danlects")

and the concInnlonn on page 2L o f t hee t. ieporL recommend agajnnt this appioach unicun adeyonto sel f nhieldlon f act ord were entablinhed for Iho fInnion pi pince s i .

Furthur inntil1 ration should be provided for this calculational apptoach. Ihfa comment relates to entablishing the accuracy of the calculatlonal procedure assuming that burnnp levels are known. ,

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b. What oncertaint.lcs are introduced in'the analysis through.varia-tions in aanon bly axial exposure, vntidt lonn in core location, uti))I.y practicen and vendor practicon?

c. The responsen to d and b shonid demonstratin'the adequacy of the Cnc. tor of two proposed. '

3, Validity of correction' factors for pellet size and unter-to-oxide volume ratio (Pinures F . J.- 1 A and F .1.-1P.) -- e '

a. (Pellet diameter correction),, Our independent calculations confirmed that the correction f actors for PWR tv. values ucre conservative. l Confirmation of the values for the PWR nssemblies was deemed' un-necessary in view of the results of the kef f. calculations for the 1 arrnva or non annenblien.

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b. (Water-to-oxide volume ratio correction) Volutne ratio correction f actora derived f rom data in the U.Y. AllSB llandbook and Figure III-5 of UCAP- 2909 agreed with the valuen in Finnre F.1-1A. M' Confirmatl.co of the values for BWR n'semh11es una deemed unnecessary for rennono nn ntated ohnve.

4. Valldity of cortcction f actor for SS cladding, in_.pjace of zirconium.

'fbe validity of Ts proposed 0.951 f actor was conf 1'rmed by reference to Fly,uro 111-13 of WeAP-2999, which Indientos the factor to be conservative by nt lenst 3 or 4 percent. The factor was also confirmed by compating the k.,. value for SS clad rods with that for zirconinm clad rods as'calcula-ted wir.h our computer. code.

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5. Adequacy or ru"1 IdgntJJicatIon_procedura., '

ConsiderIny, thot most fuel typen, and all fuci now planned to be stored, wou.ld be subcritical regardlesn of buronp. the identification, procedure -

with cheel> arrangements + in aderlunto f or sa fety purposen. As noted in the lJtst round i nview, l.he .bani:et des igno p terj ude interchange of-PWR asnemhilen wilb HUR annemhjlen. (See itema 3e and 3i, on par,e 4 of the revfem i wm d at t en lu d t o Ihe . inly 23,- lo n note to rilen.)

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6. Adequacy of k, cheeky rocedure.

The procedure GE proponen to uan to check the k. calculationh is simple and straightforward no that the probability of errors In the GE review shout ri he acceptably'Juw. ,

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7. Check effect of banket spacing on keff.

The NRC cniculations for the'first'round review were based on, the. pipes in adjacent banketn being in contact. Since the subsequent NRC calcula-tions to check the kaff for unburned' fuel of maximum anticipated reactivity (l'UR's of 3.6% onriclonent , oxide at 100% of t heoretien1 density,15 x 15, zirconium clad) were made-for groups of four pipes at'27 inches center- ,.

to -renter, eniculations were also made at lesner spacings to determine the sennitivity of keff to basket spacing. The keff for the'27 inch spacing meets the 0.95 criterion at the 93% confl.dence icvel. The array in 9tfIl snhc.ritlent nt 26.25 inch spacing nin e koff is 0.95618 1 0.00469.

Aspreviouslynoted,theCEcontrolsonAnhon1dnotpermitinbroduction of fac1 on rencl.Is" on th.tt a n n.l y z ed .

8. A_de,riuae v o f P1 a n t Sa f c t y_ Comm 13,t;e_e, to rey I n'.Lc,h a ny.cn,., [

The itcense "ho"ld he condis f oned to exc.lude chnuges in the design or the matntlo) of the statogn rocks and cantutern that would reduce the nuclear critienlity safety of the ntored I"el. With this condition, the conpetence of, and procedures used by, tha Pinnt Sa'fety Committee become s of accondary importance, and, as defined, the Committee should have suf f ln tont uper t inn and authority to enf orce t he proposed requirement.

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9. Aderjuncy_.of a rranyment' to,,p.r,eci,nde dropy1:1g_fu_n1 from a loaded' basket into the fuel ontondine. ett..

It in ausumed thit the device being designed to precinde the drop accident ws.11 be reviewed and approved by the struct urn t consultants as adequate to '

f ul f.I ll t h" Intend"d function.

10. Check of concluoJon of subctiticality for four PWR nanemblics compacted by, impact of tornnelo loduce,d, minnile.

• The appl icant 'n conc lunion uns con f irmed 'n tel"tence to an accident analysis fot Lonr l' Wit asnemi'lics , SS clatt,15 x 3 5, 4.5% curiched , reported in Dochet 110. 7ft- W . (See Ifemorandum to Filen,.luly 18, 1973, " Westinghouse a

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by C. 'B throtto.) The reactivity of the 4,5% SST clad rods is about equivalent t.o zirconium clad rods of 3.1% enrichment, approximately the maximum reactivity for 15 x 15 type assemblien permitted under Criterion 2.

11. S_uberiticality of fuel after a drop accident.

The applicant's response establishes with reasonabic certainty that fuel damage would he minor af ter .a basket drop or drop of an individual fuel annembly equivalent to a 12 ft. dror in air. -

12. R,esponse to show the relationship between assemblyA and rod ' lattice km. )

i The' applicant's response has been confirmed by our independent. computer calculotiono, which demonstrate a large margin of saberiticality for.BWR assemblies, and adequate margins for the more reactive PWR assemblies.

(See Item 1, nhor".). ,

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13. Analysis _of_ fuel _ansemblies containing r ds_of various enrichment. ,

The t!P.C calculations summarized in Item 1 render the response to this question of secondary concern, since the question relates only to BWR assemhjles. Calculations reported by Dr. L. I, Kopp of the Core Performance Branch indicate that the use of the average enrichment does indeed provide Conservative k... VaiUes, ,

14. Assurance that most reactive assembly deniy,ns_ were. considered.I,.

Ourindependentcalculationsforavarietyoffuelassemblytypbs(PWR's of various rod diametern, large and small cross scetions, SS clad and ~

zirconium clad) confirm the correctness of the applicant's choice of rn[erence denJnno.

15. A 3 teqmcy, of npplJennt's validation calculations.

The arpileant's validation calculations, ns supplemented, bracket the range of the reactivity of the fuel to be nLoted and demonstrate the acenrney nnd connervatinm of the calculational method used in the analysis.

16.- Proceduro for "talled fuel" and possible ac,cl<lental drop of a basket or asseinbly on top of stored funiz

'The'11 cense should be condition $d to preclude invalidation'of the nuclear ,

criticality safety nnalysis such as by displacement of water by less ,

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effective isolators (such as aluminum or al,r). The fuel handling equipment and procedures preclude occurrance of the postulated drop.'

accident. '

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17. Subcriticality of acetdent situation involving placement of a fuel _

assembly next to the fuel baskets being loaded in the fuel unloading pit.

This sit.uation, not specifically n'ddressed by the applicant, is similar ]

to the fue) conpaction accident. summarized in item 10, although the conditions of spacing would be lens severe. An " extra" PWR assembly would be held nt least two inches from the..nentest assembly within the boskots by the pipe and guide plate. The low kerf of the'large ,

arrov of BUR assembiles without the steel pipes (Item 1 b) indicates that the Bull ntcengement would be suberitical. The subcriticality'of the PWR system nt 3.1% enrichment'was confirmed: (1) by a calculation based on pairn of P4tt'ansemblies separated'by water gapn, and (2) by a calcula-

, tion In Phich the situation was m'ocked up cractly f or TWF assemblics with the " extra" assembly tangent to on6 of the two center pipes. The kef f for t hin ,rrannement was 0.936 at the 95% confidence limit'. '

11:1. Suberitica11tr or accident nituation involving, assemblies projecting from bushet with 42 inches of Euci e> posed. _,-

C. R. throtta 'n calcula tiono for infinite nqu're arroys of 3.5% enriched.

15 x 15 type PUB's indicate that h values at this enrichment would range from about 1.01 to 1.1 as spacing between aspemblics is reduced from font inches to three inches in water. The top guide plates in'the baskets ensure a four inch gap between ansemblies where they emerge iron thn basket, the geometric buckling of the four assembly array assures it9 ouberiticality at 3.1% enrichmnnt and with as little as

.b r eu 1nches of war er between assemblies. (The conclusion based on llr. l'orot ta'n ca.lculntion wan conf irmed by a specific computer calcula-tion bancel on inut PWH assemblies in water. -A calculation was also-modo to con [11m t h" nnberittentity of a 3 x 3 array of BWR assemblies of 'l .47 co r t elnern t .in watnr.)

19. A_q rp p t a n c e o f t he ditinrencen In k... valo-t no calculated by the: applicant and hv flpr The nocle m etlitrolliy "afety review of the In3 tint amende'ent nepl ication-r e vt.) I re t i dIi1, e o se e nl ihout ' . 'G he t we. n ou r k,e vn t ur< vnd thnne of the appilcant tot onrictenents above d 2.5%. (See' attachment t o ncte of July 23, 1975,.ltem 2a, of Record of Initini Nuclear Cr:lticality Safety Review.) '

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'lhin difference in oot conoideied of concern because the limiting values for the rear.t iv i t y (l% limits) of the ct ro eil f uel .m based on the appilcant'n calculations. Our chiculat.J oun ( i t en. 1) confirm the nuclear l

criticality nafety of the stor"d fuel when the applicant's limits are met. l l

20. Comparison of the propoued__ storage conditions,.vith relevant available I guidos.

a. Draf t Guide 3.24.'1 dated Auf;ust 27,;)p 7 5_ i i l

Requirements assumed fulfilled for: (1) quality assurance in installa-tion and during useful lifetime by inspection, (2) for pool integrity,

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and (3) for J imited storage array deformation f rom tornado missiles.

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b. Regulatory _ Guide 3.24 - As above, the inspectJon of safety related I components (and inspectability) assurved. -

. c. AM">I H210, Draft of section on Fuel Storage Racks (January 1975 revision) - The pertinent requiremchta, for kef f and for physical l measuren to prevent positioning of fuel assemblics in locations other than npe"Ifled for storage, are met.

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d. . Standard Review Plan - SEent Fuel Stgrang, nJ'eb.ruary 1975 - The intent of Criterion 02, re[ctred to in Section 11.1.e, is met. It is assumed that the requiroment of III.2.c is mel ("Eailure of systems or structures not designed to Seismic Category I st andards and located in the vicinity of l ,

tl.e opent fuel storage f acility will not cause. a decrease in the degree l

of subcriticality provided..."). t-

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