ML20150E407

Application for Amend to License DPR-36,increasing Max Nominal Fuel Enrichment from 3.5% to 3.7% U-235.Fee Paid

ML20150E407
Person / Time
Site:	Maine Yankee
Issue date:	03/24/1988
From:	Randazza J Maine Yankee
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20150E410	List: ML20150E407 ML20150E411 ML20150G053
References
MN-88-30, NUDOCS 8803310253
Download: ML20150E407 (7)

Text

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MaineYankee REUARE EMCTACTY f OR MAAE SME 194 Edison DRIVE . AUGUSTA. MAINE 04330 .(207) 623 3521

. March 24, 1988 HN-88-30 Proposed Change 138 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission

Hashington, D. C. 20555 Attenti?n

Document Control Oesk

References:

(a) License No OPR-36 (Docket No. 50-309)

(b) Letter MYAPCo to NRC dated February 12, 1988 (MN-88-17)

"Fuel Storage Criticality Analysis Methodology"

Subject:

Technical Specification Proposed Change 138: Fuel Enrichment Limit Gentlemen:

Maine Yankee Atomic Power Company requests, with this submittal, to amend its Technical Specification (TS) pertaining to the fuel enrichment limit of the reactor core. Specifically, we request to amend Technical Specification 1.3 to increase the maximum nominal fuel enrichment from 3.5 to 3.7 weight percent U-235.

The core design for Cycle 11, which uses 3.7 weight percent U-235, is a transition core to extend the operating cycle from 14 months to 18 months.

Cycle 11 is currently scheduled to begin in December, 1988.

The proposed change in fuel enrichment does not significantly impact the results of the plant safety analysis. Many other factors relating to the actual core configuration have a greater influence on safety analysis results than the enrichment change proposed herein. These factors include:

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a. The number and placement of fresh fuel assemblies, l J

l b. the exposure distribution and placement of fuel assemblies remaining i i

from previous cycles, j l c. the number and placement of burnable poison rods, and i

d. core operational strategy. i i

These factors, among others, must be evaluated along with the specific fuel enrichment for each reload core design to demonstrate that the applicable I acceptance criteria are met prior to core reloading. Technical Specification l 1 1.3 simply places an upper limit on the maximum nominal fuel enrichment allowed in the reactor core with minimal impact on the plant safety analysis. f The operation of Maine Yankee with a maximum nominal fuel enrichment of 3.7 weight percent U-235 is being reviewed and evaluated for Cycle 11. Each 1 transient and accident considered in earlier safety analyses is being reviewed i and reanalyzed where necessary. Preliminary results indicate that the core  ;

design with 3.7 weight percent enriched fuel will meet the appropriate safety i criteria. i The storage of 3.7 weight percent enriched fuel in the spent fuel pool and l new fuel storage area has been analyzed using the methodology proposed in  :

Reference (b). Both locations have been found acceptable for the storage of l

{ fuel with this enrichment level. The analysis detailing this evaluation is i presented as Attachment C.  ;

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A description of the proposed change and a summary of Haine Yankee's l l significant hazards evaluation is presented in Attachment A. As discussed in  !

! the attachment, this change does not involve a significant increase in the  !'

4 probability or consequences of an accident previously evaluated, the

possibility of a new or different kind of accident from any accident

! previously evaluated, or a significant reduction in a margin of safety.

' Therefore, this proposed change does not involve a significant hazards ,

consideration as defined in 10 CFR 50.92.

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] A revised page 1.3-1 is included as Attachment B.

, This proposed change has been reviewed by the Plant Operations Review

Committee and the Nuclear Safety Audit and Review Committee.

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United States Nuclear Regulatory Commission Page Three '

) Attention: Docu.nent Control Desk MN-88-30 1 le

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A $150.00 application fee is enclosed pursuant to 10 CFR 170.12.

I A state of Haine representative is being notified of this proposed change i by copy of this letter.  !

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! Upon your review and approval of this amendment request, we request that  !

the amendment be effective immediately. ,

Very truly yours, .

MAINE YANKEE ATOMIC POWER COMPANY l l

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John B. Randazza i

i j Executive Vice President 1 JBR/bjp i .

Attachments i t

cc: Mr. Richard H. Nessm-a i Hr. Hilliam T. Russell i

Hr. Cornelius F. Holden .

Mr. Pat Sears  :

Mr. Clough Toppan f STATE OF MAINC l

Then personally appeared before me, John B. Randazza, who being duly sworn  !

did state that he is Executive Vice President of Maine Yankee Atomic Power i Company, that he is duly authorized to execute and file the foregoing request  ;

in the name and on behalf of Haine Yankee Atomic Power Company, and that the l

statements therein are true to the best of 51s knowledge and belief.

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4 h}nklIC bil!@C ATTACHMENT A Descriotion of Prgpm ed Change This proposal would change the maximum nominal enrichment of the fuel allowed to be used in the reactor core for operating Cycle 11 and beyond.

Maine Yankee is currently in Cycle 10 operation.

Maine Yankee proposes to change, in Technical Specification 1.3, "Reactor", the fuel enrichment specification from a m3ximum nominal weight percent of 3.5 U-235 to 3.7 weight percent U-235. Changing the technical specification enrichment from 3.5 to 3.7 weight percent U-235 is acceptable since:

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a) The adequacy of a given core design relative to the acceptance criteria must be demonstrated for each core prior to core reloading, and b) The facility's fuel handlirg equipment and storage areas have been analyzed and demonstrated to meet appropriate acceptance criteria for an enrichment of 3.7 weight percent U-235.

Significant Hazards Evaluation Operation of the Maine Yankee plant in accordance with this change to its operating license has been evaluated using the standards in 10 CFR 50.92 regarding no significant hazards consideration. This proposed change does not involve a significant hazards consideration for the following reasons:

A. This change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed increase in fuel enrichment does not affect the probability of the accidents previously evaluated.

The proposed change in enrichment does not increase the consequences of accidents previously analyzed. The adequacy of a given core design  ;

must be demonstrated for each core prior to core reloading. The fuel l enrichment is only one factor that must be considered in this {

determination. The fuel enrichment itself does not significantly I impact the results of the plant safety analysis, Factors like the number and placement of fresh fuel assemblies, the exposure distribution and placement of the fuel assemblies remaining from previous cycles, the number and placement of burnable poison rods, and the core operational strategy have a more significant impact.

Preliminary results from an evaluation of the core loading planned for Cycle 11 (including a fresh fuel enrichment of 3.7 weight percent U-235) indicates that all applicable acceptance criteria will be met.

Maine Yankee's determination for Cycle 11 will be documented in our Cycle 11 Core Performance Analysis Report which is under development and scheduled for submittal in July, 1988.

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MaineYankee The facility's fuel and storage areas have been analyzed for enrichments of 3.7 weight percent U-235. The criticality analysis for Haine Yankee's spent fuel pool and new fuel vault is presented as Attachment C. The results of these analyses indicate that handling and storage of 3.7 weight percent enriched fuel does not involve an unreviewed safety question. The results of these analyses are within the acceptance criterion defined in Technical Specification 1.1, "Fuel Storage" of Kert less than or equal to 0.95.

The applicable codes, standards and regulations of criticality safety j for spent fuel and new fuel storage include the following -

f General Design Criterion (2 - Prevention of Criticality in Fuel Storage and Handling.

NUREG-0800, USNRC Standard Review Plan, Section 9.1.2, Spent Fuel Storage and Section 9.1.1, New Fuel Storage.

ANSI /ANS-57.2-1983, Design Requirements for Spent Fuel Storage Facilities At Nuclear Power Plants, Section 6.4.2.

ANSI /ANS-75.3-1983. Design Requirements for New Fuel Storige Facilities at LWR Plants, Section 6.2.4. ,

These regulations and guides require that for spent fuel racks the  !

maximum calculatede K rr including margin for uncertainty in l calculational method and mechanical tolerances be less than or equal to 0.95 with a 95% probability at a 95% confidence level. l j

In order to assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made for spent fuel rack criticality analysis:

Pure, unborated water at 68'F is used in all calculations.

An infinite array with no radial or axial leakage is modelled, and Neutron absorption from spacer grids is neglected, i.e., replaced by water.

For new fuel vaults, a dual criteria applies in which the maximum calculated Ke rr including uncertainties is less than or equal to 0.95 when flooded ud less than or equal to 0.98 under conditions of "optimum moderation".

Because the new fuel vault is normally dry and low density moderation l of "optimum moderation" produces strong coupling between assemblies, j the following conservative assumptions are used:

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j The vault is water tight, j I

Unborated water is introduced uniformly throughout the vault and  :

the space between fuel pins,  !

Hater density is varied uniformly from flooded to dry, i  !

Neutron absorption from spacer grids is neglected, i.e., replaced  !

l by water, and

! A 3-D semi-infinite array is modelled with reflection from the  !

j floor, wall and ceiling.

] In new fuel vault criticality analysis, leakage is explicitly '

1 modelled, because the assumption of an infinite array with no radial  ;

j or axial leakage is unrealistic under ("ditions of low density I i moderation. Leakage suppresses criti:ality at low moderator density. .

1 Hithout 3-D modelling of the array, erroneously high values of Keff j are calculated. Thus, the assumption on array leakage is relaxed, but d

reflection from the walls, floor and ceiling is included. '

4 I In addition to the above items, the following conservative assumptions

are applied to both analyses

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No credit is taken for the presence of burn 3ble poison shims. I

, These pins displace fuel pir. positions and are an integral part of ,

a selected fuel assemblies,

! The upper bound of the fuel density tolerance as based on the fuel  !

fabrication specification is used, and i The upper bound of the fuel assembly enrichment as based on the  !

fuel fabrication specification is used. l i .

The criticality analysis of the Maine Yankee spent fuel racks shows i 1

that the maximum fresh fuel enrichment to meet the 0.95 NRC limit with t l uncertainties is 3.72 weight percent U235 Criticality analysis of  !

l the Maine Yankee new fuel vault shows that fresh fuel with enrichment  :

l at least 5.5 weight percent U235 is allowable in the vault even j

! under condition of "optimum moderation".

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~4-An evaluation has also been performed to determine the effect of higher fuel enrichment on the fuel handling accident. The evaluation has resulted in the determination that an increase in fuel enrichment will not by itself affect the mixture of fission product nuclides.

Although a higher enrichment fuel cycle may result in fuel burnup consisting of a slightly different mixture of nuclides, the effect is insignificant because the isotopic mixture of an irradiated assembly is relatively insensitive to the fuel assembly's initial enrichment and the doses from postulated accidents are not significantly affected and continue to be acceptable.

B. This change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The adequacy of a given core design shall be demonstrated for each core prior to reloading. The fuel enrichment is only one factor that must be considered in this determination.

Operation of Maine Yankee with a maximum nominal 3.7 weight percent enriched fuel will not create any new or different kinds of accidents from those previously evaluated.

Fuel handling and storage of fuel with enrichment of 3.7 weight percent U-235 does not create the possibility of a new or different kind of accident from any accident previously evaluated.

C. This change will not involve a significant reduction in a margin of safety.

The evaluation performed for each reload core assures that the core design meets appropriate safety limits, including a consideration of a significant reduction in the margin of safety. See response provided in Item A for information pertaining to the results of preliminary evaluations performed for Cycle ll, the first reload core introducing 3.7 weight percent U-235 fuel.

The margin to criticality for fuel assemblies of 3.7 weight percent in the Maine Yankee fuel pool storage racks meets the NRC acceptance criterion of 0.95 for Kerr, Reference (b), even with the many conservative assumptions used in the calculation of Kert assuming 3.7 weight percent fresh fuel. Similar conclusions have been reached for the new fuel storage area.

Based on the above evaluation, this proposed change does not constitute a significant hazards consideration.

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