ML20141M112

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Application for Amend to License DPR-36,consisting of TS Proposed Change 165,revising Fuel Enrichment Limit & Allowing Radial & Axial Zoning of Enrichment Up to 3.95 Weight Percent U-235
ML20141M112
Person / Time
Site: Maine Yankee
Issue date: 03/24/1992
From: Frizzle C
Maine Yankee
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20141M115 List:
References
CDF-92-31, MN-92-27, NUDOCS 9204010169
Download: ML20141M112 (6)
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MaineYankee HEEfMUM!OMMSOMOM cm wanue rmnom Preu1tra Awh Mwe 04336 March 24, 1992 3 Mfi 27 CDF-92-31 Proposed Change 165 UtilTED STATES fiUCLEAR REGULATORY COMMISS10fi Attenticn: Document Ccatrol Desk Washington, DC 20555

References:

(a) License No. DPR-36 (Docket flo. 50-309)

(b) Lotter MYAPCn to NRC dated february 12,1988 (MN-88-17) " fuel Storage Criticality Analysis Methodology" (c) Lctter NRC to MYAPCo dated June 23,1988 (HMY-88--085), Amendment No, 105 and Attached Safety Evaluation

Subject:

Technical Specification Proposed Change 165: fuel Enrichment Lit.,it Gentlemen:

Maine Yankee Atomic Power Cor.: party requests, with this submittal, to amend its Technical Specification (15) pertaining to the fuel enrichment limit of the reactor core. Specifically, we request to amend the fuel enrichment limit and allow radial and axial zoning of enrichment up to . raximum as-f abricated radially-averaged enrichment of any axial enrichment zone w. in a fuel assembly of 3.95 weight percent U-235. . We also request to reviso the specincation to specify a range for the active fuel length from "136 to 137 inches", as the current nominal value of "136.7 inches" changes slightly with advanced fuel types.

The proposed changes do not directly impact the results of the plant safety analysis. Many factors must be evaluated along with the specific fuel enrichment and

active fool height for each reload core design to demonstrate that the applicable l- acceptance criteria are met prior to core reloading. The proposed Technical Specification 1.3 places an upper limit on the maximum nominal fuel enrichment and an upper and lower limit on active fuel height allowed in the reactor core. These changes have resulted in no direct impact on the existing plant safety analysis, t

i The operation of Maine Yankee with a ...aximum radially-averaged enrichment, of l any axial enrichment zone, of 3.95 weight percent U-235 is anticipated for Cycle 14

( and beyond. As customary during the reload design process, each transient and l accident considered in the safety analyses is reviewed and reanalyzed where necessary to meet the appropriate safety criteria.

The storage of fuel assemblies with a maximum radially-averaged enrichment, of
any axial enrichment zone of 3.95 weight percent enriched fuel in the spent fuel pool I has been analyzed using the methodology described in Reference (b) and approved in Reference (c). The analysis used uniform enrichments only. Radial enrichment zonine

, was evaluated and shown to be conservative relative to uniform anrichments. Radial l zoning with lower enrichments nearer the guide tubes reduces assembly reactivity in both the core and spent fuel > ack geometries. Using the uniform enrichment limit in the specification means the specifics of the enrichtent zoning patterrs will not i require reevaluat,io,n Mi t.h e#.h new fuel type. Likewise, the wording of the U. b t 'l /'

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UNITED STATES flVCLEAR RECULAiORY COMMISSI0ff Mti-92-27 Attentivn: - Document Control Desk Page 2 I

sntcification addresses both non-blanketed and blanketed assemblies; since the ,

maximum enrichment of any axial zone is specified, with no credit taken for the  ;

reduced enrichment in the blanket regions. The analysis detailing this evaluation is presented as Attachment C. i A description of the proposed chmge and a summary of Maine Yankee's significant i hazards evaluation is presented in Mtachment A. As discussed in the attachment, this change does- not involve a significant increase in the 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 sigt,1ficant  !

reduction in a trargin of safety. Therefore, this proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92.

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

This proposed change has been reviewed t:y the Plant Operations Review Committee and the liuclear Safety Audit and Review Comittee.

  • A state of Maine representative is being notified of this proposed change by )

copy of this letter.  :

Upon your review and approval of this amendment request, we request that the amendment be effective immediately.  :

Very truly yours, c i ,

a

$f //

Charles D. Frizzle '

President BWS/sjj c: Mr. Thomas T. Martin Mr. E. H. Trottier ,

Mr. Charles S. Marschall l Mr. Patrick Dostie i Mr. W. Clough Toppan STATE OF MAINE Then personally appeared before me, Charles D. Frizzle, who being duly sworr. did state that he is President of Maine Yankee Atomic Power Company, that he is duly 1 i authorized to execute and file the foregoing request in the name and on behalf of

Maine Yankee Atomic Power Compeny, and that the statements therein are true to the best of his knowledge and belief.

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4 ATIACHMENT A page 1 of 4 EtKripilDILDLPlRMRd_ Chang This proposal would change the maxim m nominal enrichment of the fuel allowed to be used in the reactor core, it would also allow for slight changes in the active fuel length.

Maine Yankee proposes to change, in lechnical Specification 1.3, " Reactor", the ,

fuel enrichment specification from a maximum nominal 3.70 weight percent U-235 to a maximum, as-fabricated, radially-averaged enrichment, of any axial enrichment zone within a fuel assembly, of 3.95 weight percent U-235 and to change the nominal active fuel length of 136.7 inches to a range of 136 to 137 inches. The pre 9 sed change is aueptable since: _

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. Maine Yankee's fuel handling equipment and storage areas have been analyzed and demonstrated to meet appropriate acceptance criteria fnr a maximum, as-fabricated, radially-aver qed enrichment, of any axial enrichment zone within a fuel assembly, of 3.9b weight percent U-235.

Simlficant HRanlL[yAluadsn 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 dces 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 or malfunction of equipment important to safety previously evaluated. _

The proposed increase in fuel enrichment and change to a range of fuel height does not affect the probability of the accidents previously evaluated.

The proposed changes in enrichment and fuel height do not increase the consequences of accidents previously analyzed. The adequacy of a given core design must be demonstrated for each core prior to . ore reloading.

The fuel enrichment and active length are only some of the factors that must be considered in this determination. The fuel enrichment and minor changes in active length alone do not directly impact the results of the plant safety analysis.

The facility's f uel and storage areas have been analyzed for a maximum a; fabricated radially-averaged enrichment of any axial enrichment zone within a fuel assembly of 3.95 weight percent U-235. The criticality analysis for Maine Yankee's spent fuel pool is presented as Attachment C.

The results of these analyses indicate that handling and storage of such fuel assemblies do not involve an unreviewed safe.y question. The resd ts of these analyses are within the acceptance criterion defined in Technical L:\ISDM H %\K R s E l

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l ATTACH liENT A Page 2 of 4 j l

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Specification 1.1, " fuel Storage" of K.,, less than or equal to 0.95. The i new fuel vault is addressed in the original analysis of Reference (b).

The applicable codes, standards and regulations of criticality safety for  !

spent fuel and new fuel storage include the following: l

Storage and Section 9.1.1, New fuel Storage.

ANSl/ANS-57.2-1983, Design Requirements fer Spent fuel Storage l

facilities At Nuclear Power Plants, Section 6.4.2.  ;
  • ANSI /AN5-75.3-1983, Design Requirements for New fuel Storage f acilities at LWR Plants, Section 5.2.4.

These regulations and guides require that for spent fuel racks the maximum calculated K.,,, including margin for uncertainty in calculational method ,

and mechanical tolerances be, less than or equal to 0.95 with a 95% >

I probability at a 95% confidence level. ,

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

  • Pure, unborated water at 68af 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 K ,,, including uncertainties, is less than or equal to 0.95 when flooded and less than or equal to 0.98 under conditions of " optimum moderation".

Because the new fuel vault is normally dry, and low density moderation of

" optimum moderation" produces strong coupling between assemblies, the -

following conservative assumptions are used:

  • The vault is water tight,
  • - Unborated water is-introduced uniformly throughout the vault and the I space between fuel pins, i

Water density is varied uniformly from flooded to dry, L:\PROPChNG\Pc!6s

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ATTACHMENT A  !

Page 3 of 4 l

  • Neutron absorption from spacer grids is negiected, i.e., replaced by water, and
  • A 3-0 semi-infinite array is modelled with reflect ton from the floor,  !

wall and ceiling.

In new fuel vault criticality analysis, leakage is explicitly modelled, because the assumption of an infinite array with no radial or axial ,

leakage is unrealistic under conditions of low density moderation.  :

Leakage suppresses criticality at low moderator density. Without 3-D modelling of the array, erroneously high va ues of K,,, are calculated.

Thus, the assumption on array leakage is relaxed, but reflection from the walls, floor and ceiling is included. '

In addition to the above items, the following conservative assumptions are .

applied to both: analyses: i a No credit is taken for the presence of burnable poison shims. These pins displace fuel pin positions and are an integral part of selected  !

, fuel assemblics,

-* The upper statistica, bound of the fuel density tolerance, as based on the fuel fabrication spe';ification, is included in the statistical evaluation of uncertainties, and a The upper statistical bound of the fuel assembly enrichment, as based on the fuel fabrication specification, is included in the statistical evaluation of uncertainties. '

The criticality analysis of the Maine Yankee spent fuel racks show:; that the maximun, as-fabricated, radially-averaged f uel enrichment of any axial enrichment zone within a fuel assembly which meets the 0.95 NRC limit wit!

uncertainties is-3.95 weight percent U'". Criticality analysis of the Maine Yankee new fuel vault shows that fresh fuel with enrichment at least 5.b weight percent U'" is allowable in the vault even under condition of -

" optimum moderation".

Although . a higher enrichment fuel cycle may result in fuel burnup ,

consisting of a slightly different mixture of nuclides and inventory, the effect is-insignificant because the isotopic mixture and inventory 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 pessibility of a new or d;fferent kind of ,

accident or malfunction of equipment important-to safety from any accident previously evaluated. ,

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i ATTACHMENT A Page 4 of 4 The adequacy of a given core design shall be demonstrated for each core prior to s eloading. The fuel enrichment and active fuel length are only some of the factors that mtist be considered in this determination.

Operation _of Maine Yankce with the proposed enrichment limit change will not create any new or different kinds of accidents from those previously evaluated. Operation of Maine Yankee with a range specified for active fuel length rather than a tenths place of significant digit specified will ,

not create any new or different kinds of accidents from those previously evaluated.

Fuel handling and storage of fuel with as-fabricated radially-averaged enrichment of any axial enrichment Zone within a fuel assembly of 3.95 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, ircluding a consideration of a significant reduction in the margin of safety, See response provided in item A for information pertaining to the demonstratien of the adequacy of each core design.

The nargin to criticality for fuel assemblies with a maximum as-fabricated radially-averaged enrichment of any axial enrichment zone nf 3.95 weight percent in the Maine Yankee- fuel pool storage racks meets the NRC ccceptance criterion of 0.95 f or K.,,, Attachment (c), even with the many conservative assumptions used in the calculation of K,,,.

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

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