ML20034C542

Application for Amends to Licenses DPR-53 & DPR-69,to Allow Movement of Spent Fuel Shipping Cask Into Spent Fuel Pool

ML20034C542
Person / Time
Site:	Calvert Cliffs
Issue date:	04/30/1990
From:	Creel G BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20034C543	List: ML20034C542 ML20034C545
References
NUDOCS 9005040104
Download: ML20034C542 (6)

Text

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I bALTI M O RE

( 4AS AND

/i ELECTRIC 1

CHARLES CENTER P. O ROX '475 BALTIMORE,WRYL AND 21203 v$c"#',',[.d"["'

April 30, FA Nsca n ENEROY aoo.seo 4 sn 7

U. S. Nuclear Regulatory Coihmissivn Washington, DC 20555 f

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ATTENTION:

Document Contron Dest.

St.tLJECT:

Calvert Cliffs Audeu, Power Plant Unit Nos.1 & 2: Docket No. 50-317 & 50-318 Tt.,anical SpeV fication Change Request:

Soent Friel Cap Handline Crane m

REFERENCE:

Letters from A. E. Lundvall, Jr. (BG&E) to D. G. Eisenhut (NRC), datea L

January.,1982 and March 1,1982, Control of Heavy Loa:Is Gentlemen:

The Baltimone CM and Lem ic O..apatt hereby requests an Amu amer.t to its Operathg

_.icense Nas. DPI't 53 and JPR-69 fo. Calvert Cliffs Uait % i & 2, respectively. 11 accordanes with t h ' ' R ).90 rg ' 50.91 to allow movement if s spent fuel shipping cask into Our spent fuel pool. The task will be used to ship selecud fuel rods in support of o. t-ce. I wt tk sponsoro ' by EPRI. The primary purpose of the EPRI program is :o pr,. Fde hhh burnnp fuel )eiformance data that wiii support operation at extended.uel bu m ps.

s ta will be /atan < d selc&g to com ion and hydrogen pick up in g--

Zirtloy-4. Jel rod Cladding and assr/ dbly structural Compef tSts, ' effects of extended operation (including corrosion atu firence) on mechanical ' 3 operties of Zircaloy-4 cladding.nd co7 ponents, and fission il s release and its rd stionship with UO, pellet microstructure and fuel rod in Wnal pres'ure.

In ordei te meet scheduIes foe i

av.a ht'nir e the spent fuel shipping cua.', and for comnirebn of the hot - (1

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e art h.a0 in, tne license avendraent would neeG.o be (. sued by ~. sly 1,1990.

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I Besidrs t ermittirq a shipment of fuel rods la n.pport of th EPRI program, the proi).s<H i

chan3e 4 ill also allov/

reacter vessel weld surveillance capsule to be removed irm t our spent fuel poet using a shipning cask.

DESCRIPTVV Ot) CHANTT l

A change to },.nri,:al Specificat. ion 3/4.9.1 x " Spent Fue. C.sk Pandling Crane" is

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proposed to illow if sveny tt d? spmt fiel shipping et th wVlf a a cask le,psth of fuel

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ceit). n the pool. The ms. c nk nt 6 to be adowed ont'; W

.M becon concentration of the

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spect fuM pool is greate.- than or equal to 1000 pt m 3ND the following critenu are met by all assemblict wi:hil one ask length radius of the pathway: 1) initial 0

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3 Documeet Contrst Desk

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April 30,1996 Page 2 enrichment less than or equal to 4.1 w/o U-235; 2) burnup greater than or equal to 28,000 MWD /MTU; and 3) greMor than 4@ days elapsed from the shutdown of the last operating cycle, in which the n$tembly was present in the core. The change is applicable only for a Ahipment of spent fuel rods supporting EPRI sponsored hot-cell work and for a sh!pnont of a reactor vessel weld surveillance capsule in support of our

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life cycle management program.

p The proposed change 'tattached) will add the following footnote to ' Technical Specification 3/4.9.13:

'These conditions are modified to permit shipping cask travel to and from the cask pit in the presence of fuel within one cask length radius of the pathway provided the boric acio concentration in the rpent fuel pool is greater than or equal to 1000 ppm AND thq following criteria are met by all assemblies within one cask length radius of the pathway: 1) initial enrichment less than or equal to,4.1 w/o I.b235, 2) Burnup greater than or equal to 28,000 MWD /MTU, and

3) eccater than 440 days elapsed from the shutdown of the last carating cycle in which the assembly was present in the core. Crane interlocks and physical stops which restrict a spent fuel shipping cask from passinE over any area within one sh;pping cask length of any j

fuel assembly not satisfying the above criteria shall be demonstrated OPERA DLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to using ti.e crane for moving a cask I

within one cask length of fuel assemblies raeeting the above criteria.

These modifications are applicable only for the shipment of fuel reeis supporting the EPRI sponsored hot-cell work and fer the shipment of a reactor vessel weld material surveillance capsule "

REASON FOR CilANGE Compliance with Technical Specification 3/4.9.13 would require movement of fuel out of rack locations located within one cask length of the spent fuel shipping cask pit vhenever a cask is brought into the pit. Figure 1 (attached) shows the area of the spent fuel pool affected by this requirement. Because of the amount of spent fuel stored in the pool, which has been increased because of the offload of the Unit 2 Cycle 8 core into the pool, there is presently insufficient space to comply with the requirements of Technical bpecification 3/4.9.13. Unit 2 core onload and modiNeations to fuel handling eq,uipment which would provide access to additional storage locations cannot occur until later this year. As described below, cask shipments must occur before those evolutions can take place. Consequently, we cannot comply with Technical Specification 3/4.9.13 and accommodate the cask shipments. Therefore, a change to the Technical Specification is required to allow a one-time shipment of fuel rods supporting EPRI sponsored hot-cell work and a one-time shipment of a reactor vessel weld material surveillance capsulS supporting our life cycle management program.

The surveillance capsule contains neutron irradiation dosimetry data and must be removed from the spent fuel pool and analyzed within the next year. Otherwise, important data needed to monitor reactor vessel fracture toughness properties will be lost through radioactive decay. We have considered other means of removing the capsule but all would result in increased radiation exposure to the personnel involved. Thus, the use of a shipping cask to make this shipment is preferred.

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t Docume:t Cc ttrl Desk April 30,1990

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ig With regard to the hot cell program, EPRI has stated that the fuel rod shipment to the hot cell must be scheduled for the summer of 1990. The other schedule constraint is a window for shipping cr.sk availability which is currently from June to mid July 1990.

The tight restriction on availability is due to DOE's heavy use of all available casks. In consideration of the schedule constraints, we are scheduling a cask shipment for early July 1990.

DETERMINATION OF NO SIGNIFICANT llAZARDS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:

(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or The referenced letters considered a heavy load drop in response to NUREG-0612,

• Control of Ileavy Loads at Nuclear Power Plants

• and while the proposed Technical Specification change will not increase the probability of the drop, it will affect the consequences. Our NUREG responses reported that no fuel would be impacted by the drop due to restrictions placed on locating fuel within the are shown in Figure 1. The Basis for Technical Specification 3.9.13 relates to the fact that when fuel storage within one cask length of the pathway is prohibited, a dropped cask will not cause fuel damage (the only safety consequences of fuel damage being i

offsite dose) or result in a critical array, Because the proposed change would allow a dropped cask to impact the fuel, we have evaluated the offsite dose and criticality cMeerns of a postulated cask drop.

Our evaluation of the dose consequences conservatively assumes that every fuel assembly within a cask length radius of the shipping cask's pathway would experience complete cladding failure of every rod in the assembly.

The methods for calculating gas gap activity and resulting off-site dose are the same as those used in performing calculations in Chapter 14 of the FSAR for the Fuel liandling incident (Fill), where total cladding failure is assumed for one fuel assembly damaged three days after shutdown. For this analysis we decayed the F111 gas gap inventory for 440 days as opposed i:

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to the 3 day decay assumed in the FIII analysis. This resulted in the gas gap activity being reduced by a factor of 267 from the Flil activity, Using acceptance criteria contained in NUREG-0612, which states that the heavy load drop is acceptable if the resulting offsite doses are limited to less than 25% of the 10 CFR 100 limits, over 5000 assemblies would have to be damaged to reach the NUREG limits based on the gas gap activity calculated for a 440-day decay. Because a maximum of approximately 500 assemblies could be contained within a cask length radius of the pathway, we would remain well within the NUREG limits even if all 500 assemblies could be damaged by the cask drop, ii i

i Doe:me:t Contr:1 Desk April 30,1990 Page 4

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A more realistic analysis would predict that every assembly within the are j

could not conceivabiy be damaged. The are contains fuel assemblies in both 4

the North and South pools, separated by a dam. Because the cask pit is tocated in the North pool and the placement of mechanical stops, controlled 1

by a cask handling procedure, will prohibit travel of the cask over the dam I

into the South pool, fuel damage from a cask :frop should occur only in the 4

North pool, Considering this approach, the predicted offsite doses for the I

cask drop can even be shown to be bounded by the doses predicted for the i

F}li, v here one assembly is damaged 3 days after shutdown. Bned on the calcula%n described above for a 440-day decay, up to 267 assemblies could be damaged and the Fill dose would _ still not be exceeded. Since the number of fuel assemblies in the affected area of the North pool can be limited to 260, the Fill dose would bound the cask drop dose even if gli of the fuel in the North pool lying within the arc was to be damaged.

Furthermore, damage to every assembly in the North pool within the arc is highly unlikely, since if the cask fell, even if it rotated on its side, 'it would not be likely to roll continuously in a trajectory which would result in damage to every assembly within the arc.

in summary, the offsite dose consequences associated wlth-the proposed modification remain well within limits specified in Criterion I of NUREG 0612, Section 5.1 (< 25% of 10 CFR 100 limits), and with a more realistic analysis, the results can even be shown to be bounded by doses previously calculated for the Fill. We thus conclude that the probability or offsite dose consequences of a hean load drop have not increased significantly.

Criticality concerns were considered for the proposed modification because a cask drop would cause a geometrical distortion of the fuel /r'ack system.

Because the distortion is difficult to predict, assumptions were macie to bound the most reactive configuration. For the calculations, we assumed that the geometry of individual fuel assemblies was not deformed (this maximized reactivity) and that the storage racks were deformed to remove the inter-storage cell gap (neutron flux trap).- Additionally, the poison j

material contained v'ithin the racks was ignored and replaced by pool water, f

While NUREG-0612 states that criticality analysis for a dropped load may 1

assume that poison material integral to the rack remains in place, we conservatively chose to ignore this benefit. We also assumed an initial fuel assembly enrichment limit of 4.1 w/o U-235, a minimum assembly burnup of 28,000 MWD /MTU, and a minimum boron concentration in the spent fuel pool of 1000 ppm. Consistent with NUREG requirements, these assumptions (enrichment, burnup and boron concentration) are incorporated directly into the proposed Technical Specification change. Our procedures will require that serial numbers of fuel assemblies located within one cask length radius of the pathway be checked to confirm that all fuel within the are conforms to the above restrictions. Also the boron concentration of the pool will be verified to be greater than or equal to 1000 ppm prior to cask movement into the pool and prior to cask movement out of the pool.

The method used to assess criticality concerns in the same method used to address criticality concerns relative to our spent-fuel pool enrichment upgrade, which has received NRC approval. A two-dimensional analysis l

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Doe:mert C=tr:I Desk April 30,1990 Page 5 l

using the DOT-.IV computer code was performed fo< an ~ infinite array of fuel 4

assembly storage modules distorted as described above. The resuitti yielded a K-eff of 0.698, Because uncertainties are less than 0.03, the Keeff value with uncertainties will be no greater than 0.928. Consequently, results are well within the criticality limit (i.e.,

K-eff 5

0.95) i i

specified in Criterion 11 of NUREO 0612, Section 5.1.

(ii) create the possibility of a new or different type of accident from 1

any accident previously evaluated; or The proposed change would allow fuel to be stored in an area previously prohibited when a shipping cask is being moved into the cask pit. Because fuel is normally stored around the cask pit when cask movement 'is not taking place, the condlitions within the cpent fuel. pool have not changed.

Since a heavy load drop was previously considered in our NUREG-0612 responses, we have not created a new accident scenario, but rather have' I

increased the consequences or such 'an accident and as discussed above, these consequences remain well within established limits.

(iii) involve a significant reduction in a margin of safety.

1 The potential a,dverse effects. on r,afety margins associated with the i

proposed change involve offsite dose and criticality concerns caused by a-I cask drop. A *. previously discussed, by complying with the restrictions contained within the proposed change, offsite dose and suberitical margin are not significantly affected.

Consequently, there is no significant decrease in a margin of safety.

In the March 6 1986, Federal Register Notice, the NRC listed examples of changes which are considered not likely to involve significant hazards considerations. Example tvi) from this list states:

"A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or.may reduce in some way a safety margin, but where the results of the change are clear'y within all acceptable criteria with respect to the system or component specified in the Standard Review I

Plan,...".

The proposed change is similar to Example (vi) in that the change conforms with NRC guidance for heavy loads and the resulting consequences are well within established limits. Accordingly, the proposed change does not involve a significant hazards I

consideration.

SAFETY COMMFITEE REVIEW l

The proposed change to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Off-Site Safety Review Committees, and they have concluded that implementation of this change will not 4.

result in an undue risk to the health and safety of the public.

Document Control Desk April 30,1990 Page 6 Very truly yours, i

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i STATE OF MARYLAND.

bouin #d Md

,19 fB, before me, Jhe I here rtify that on the _dd V day of U 3%

~ Mr. hoAa 6 subsp,riber, p Notary Public of the State of Maryfand in and for

('Mr *Rv

, p,ersonally appeared George C. Creel, being duly sworn, an'd states that he is V4co President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belief; and thet he was authorized to provide the response on behalf of said Corporation.

WITNESS my lland and Notarial Scal:

kA b I'

i o ary Yu'blic l

My Commission Expires:

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1>a GCC/DBO/bjd Attachment cc:

D. A. Brune Esquire J. E.

Silberg, Esquire R. A.Capra NRC I

D. G. Mcdonald, Jr., NRC

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T. T. Martin, NRC L. E. Nicholson, NRC T. Magette, DNR i

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