ML20033E149

Forwards Unit 1 Cycle 4 Reload Core Rept.Vendor Elected Not to Exercise Option of Performing Cycle 4 Neutronic Design in-house Because of Transition to Westinghouse VANTAGE5 Fuel

ML20033E149
Person / Time
Site:	Byron
Issue date:	02/26/1990
From:	Schuster T COMMONWEALTH EDISON CO.
To:	Murley T Office of Nuclear Reactor Regulation
References
0708T, 708T, NUDOCS 9003090150
Download: ML20033E149 (7)

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4 February 26, 1990 3l Mr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation

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U.S. Nuclear' Regulatory Comunission n_

Washing ton, ' DC' 20555

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Subject:

Byron. Station Unit l' Cycle 4 Reload Core 1

NRC Docket No.__50-454

References:

1.

Westinghouse WCAP-9272-P-A, dated October 1985;

" Westinghouse Reload Safety Evaluation Methodology",

(originally issued March 1978).

2.

' CECO submittal, F. G. Lentine to H. P. Denton dated

~ July 27, 1983; titled " Zion Stations Units 1 and 2,.

Byron Station Units 1 and 2, Braidwood Station Units 1 and 2,' Commonwealth Edison Company Topical, Report on Benchmark of PWR Nuclear Design Methods, NRC Docket Nos. 50-295/304 50-454/455, and.50-456/457.

3..

NRC SER on Ceco's Neutronics Topical (Ref. 2) dated December 13, 1983.

-- 4.

Westinghouse WCAP-11596-P -A, dated June 1988; y

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" Qualification of the *PHONEIX/ANC Nuclear' Design-System for Pressurized Water Reactor Cores

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Ceco submittal, R.A. Chrzanowski to T.E. Murley,

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" Byron Station Units 1 and 2 Application for Amendment to. Facility' Operating Licenses NPF-37 and NPF-66,"

dated July'31, 1989.

Dear Mr. Murley:

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Byron Unit'I has completed its third cycle of operation and is

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conducting a refueling outye that began January 7,1990.

Byron Unit 1 Cycle 3 attained a final cycle burnup of approximately 16,200 MWD /MTU. - Cycle 4 is t

expected to conunence early in March 1990. This letter provides you with two

reports pertaining the Byron Unit 1 Cycle 4 reload core. Attachment 1

-describes.the reload core,and summarizes Edison's review performed in accordance with 100FR50.59. Attachment.2.provides the Core Operating Limits Report for' Cycle 4 pursuant to Technical Specification 6.9.1.9.

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. i Commonwealth Edison and its vendor (Westinghouse) apply NRC approved reload design methodology developed by Westinghouse as described ~in

Reference:

1

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Commonwealth Edison ~ requested approv31 to perform the neutronic portion of the reload design in Reference 2, and the NRC approved this: request _ (Reference -

3).

However, Commonwealth Edison elected to not. exercise the option of

- performirg th'. ! 'rnn Unit 1 Cycle 4 neutronic design in-house because of the transition-tt taa Westinghouse VANTAGE 5 fuel.

The Byron Unit.1 Cycle 4 neutronic design, including development of the core operating limits, was generated.by Westinghouse using the NRC approved methodology of Reference 1 and the NRC approved neutronic code package as described in Reference 4.

i Please address any questions regarding this-submittal'to this office.

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Very truly yours, P

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T. K. Schuster Nuclear Licensing Administrator Byron Station cc:

.P. C. Shemanski A. B. Davis B. Clayton Resident Inspector-Byron i,

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ATTACHMENT 1 l

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ByInn 1 Cycle _4 Agload Descriptien The Byron Unit 1, Cycle 4 reload core was designed to perform under current nominal design parameters. Technical Specifications and related bases, and current Technical Specification setpoints such that:

1.-

Core characteristics will be less limiting than those previously reviewed and_ accepted; or

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2.

For those postulated incidents analyzed and reported in the Updated Byron Final Safety Analysis Report (UFSAR) which could potentially be affected L

by fuel reload, reanalyses or reevaluations have demonstrated that the results of the postulated events are within allowable limits. The

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reanalysis, described in Reference 5, was reviewed and approved by the NRC.. Commonwealth Edison Company performed a detailed review with L

Westinghouse on the bases, including all the postulated. incidents H

considered in the UFSAR, of the Reload Safety Evaluation (RSE).

L The Byron Unit 1 Cycle 4 core will be a " Low Leakage" design.

Commonwealth Edison has successfully developed and used similar " Low Leakage" designs at its Byron and Zion Units.

During the Cycle 3/4 refueling, eighty-eight (88) VANTAGE 5. fuel assemblies will be inserted into the core.

The Byron Unit I core will then contain a combination of fresh Westinghouse VANTAGE 5 fue1~ assemblies and-Westinghouse's 17x17 Optimized Fuel Assemblies (0FA's), as described in the Reference 5 Amendment submittal. Reference 5 requested approval for the transition to VANTAGE 5 fuel and associated proposed changes to the Byron Technical Specifications. The submittal fully justified

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the-compatibility of. Westinghouse OEA and VANTAGE 5' assemblies in a reload l

core, and verifies. compatibility with control rods and reactor internals interfaces. A mixture of Integral Fuel Burnable Absorber'(IFBA) rods and Wet Annular Burnable Absorbers (WABAs) will be used as the burnable poison. WABAs have been used extensively by Commonwealth Edison. A description and evaluation of IFBA rods is presented in Reference 5.

The reload fuel assemblies will incorporate Westinghouse standardized fuel' pellets, reconstitutable top nozr.les (RTN), extended burnup design

.featurco, and snag resistant grids.

Similar features have been successfully utilized in Commonwealth Edison's Byron and Braidwood Units. Additionally, tho reload fuel assemblies will incorporate the Debris Filter Bottom Nozzle (DFBN).

The DFBN, hydraulically and structurally equivalent to the nozzle

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used on the existing fuel assemblies, is expected to improve fuel performance by reducing the size of any debris that enters the active fuel region. This feature is-currently in cperation in Braidwood Unit 1 Cycle 2.

The significant new mechanical features of the VANTAGE 5 design are the Intermediate Flow Mixer (IFM) grids and the Axial Blankets.

Structural evaluations of these fuel features provided in-Reference 5 verify that the VANTAGE 5. assembly design is strurourally acceptoble.

In addition, Westinghouse's Enhanced Performance Rod Cluster Control Assemblies (EP-RCCAs),

which'contain a Silver-Indium-Cadmium (Ag-In-Cd) absorber instead of hafnium, will be utilized in Cycle 4.

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_4-The reloc luel'a nuclear design has been evaluated in Reference 5.

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As lDFA and VANTAGE 5 fuel have the same pellet and fuel rod diameters, most i

reactivity parameters are insensitive to fuel type. Changes in nuc1 car H

characteristics due to the transition from 0FA to VANTAGE 5 fuel'are-within the range normally seen from cycle to cycle due to fuel management effects. The s

1oading pattern dependent parameters were evaluated'in-detail in the Ceco / Westinghouse-reload safety evaluation described below.

In addition,

. based upon the performance of an eighteen case FAC analysis, a total peaking factor (Fq) of less than 2.50 is'the maximum which could occur for the full rai.ge of power distributions, including load follow maneuvers, allowable under

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Constant Axial Offset Control (CA00). Therefore, additional surveillance of Fq(z) is not required in Cycle 4.

The Cycle 4 radial peaking factor (Fxy) limits are described in the attached " Core Operating Limits" report.

The thermal-hydraulic design for the Cycle 4 reload core has not p

significantly changed from that of the previously reviewed and accepted initial-cycle design. Tests and analysis have confirmed that the VANTAGE 5 assemblies are hydraulically compatible with the OFA assemblies reloaded as Regions 4 and 5.. The-approved VANTAGE 5 amendment FNDH limits of less than 1.55 fer OFA assemblies and 1.65 for VANTAGE 5 assemblies ensures that the DNB ratio of the limiting power rod during Condition I and Condition II events is greater than or equal to the DNBR limit of the DNBR correlation being applied.

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Commonwealth Edison's reload safety evaluation process is a l

sverification that previously reviewed and approved accident analyses are not 1:

adversely in.pacted by the cycle specific reload core design. Commonwealth i

Edison's Byron. Unit 1; Cycle.4 reload safety evaluation applied both the LOCA l'

and non-LOCA safety analyses presented in' Reference 5 and relied on previously l

. reviewed and accepted analysis reported in UFSAR, fuel technology reports, and previous reload safety evaluation reports. A detailed review of the core l

characteristics was performed to determine those parameters affecting.the 1

l postulated accident analyses reported in the Byron UFSAR, and in Reference 5.

Commonwealth Edison verifies that accident analyses presented in the UFSAR, as

. modified by the analyses described in Reference 5, were not affected by the reload core characteristics. Commonwealth Edison har previously verified that the results of the Reference 5 reanalyses were within previously reviewed and accepted limits.

l, LThe reload safety evaluation demonstrated that no additional Technical Specification changes, beyond those previously submitted for NRC approval in Reference 5, are required for operation of Byron Unit i during Cycle 4.

Commonwealth Edison has reviewed the cycle-specific licensing analysis results and concludes that no unreviewed safety questions exist for

'this reload, as defined by 10CFR50,59.

Therefore, no additional prior NRC n

review and approval'of the reload core analyses or application for amendment to the Byron Unit 1 operating license beyond that requested in Reference 5, is required as a result of the cycle-specific reload design for Cycle 4.

Finally, verification of the reload core design will be performed per the standard. reload startup physics tests. These tests include, but are not

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A physical inventory of the fuel in the reactor by serial number and (1

loc.atiotr prior; to the replacement of the reactor head; j'

2..

Control rod' drive tests and drop times; 1

3.. Critical' boron concentration measurements;.

4.

Control bank worth measurements using the rod swap technique; 5.

Moderator temperature coefficient measurements; t

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Stt tup power distribution measurement using the incore flux mapping l

system.

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<.i n Byron Unit 1 Cycle 4 Limit Report Fay Portion This Radiel Peaking Factor Limit is provided in accordance with Paragraph 6,9,1.9 of the Byron Unit 1 Nuclear Plant Technical Specifications.-

The 7 limits for RATED THIWIAL POWER within tho'specified core-planaIYfor Cycle 4 shall be:

1.' N iess'than or equal to 1.98 for all core planes;

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cEtainingbank'D'controlrods,and i

AU 2'P 1ess than or equal to 1.75 for all unrodded core pNoos.

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~: t These Fxy(s);1imits vert used to confira-that the heat flux hot l

channel factor:M2(z) will be limited to the Technical Specification values'of:

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F (z) $[L,121 [K(z)) for P > 0.5 and,-

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F (s) $[5.00) [K(s)) for P $ 0.5 q

assuming the most limiting axial power distributions expected to i

, result from the. insertion and removal of Control. Banks C and D during_

operation.. including the accompanying variations in the axial xenon

-and power distributions as described in the " Power Distribution m

-Control and Load Following Procedures". WCAP 8403, September, 1974.

limits provide assurance that the initial

Therefore-- these F*In the LOCA analysis are met,'along with the ECCS conditions nasumed acceptance criteria'of 10 CFR 50.46.

See Figure 'l for a plot of [F.Pg ] versus Axial Core Height.

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