ML20086E207

Core Operating Limits Rept for Cycle 3

ML20086E207
Person / Time
Site:	Braidwood
Issue date:	11/21/1991
From:	Simpkin T COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
NUDOCS 9111270098
Download: ML20086E207 (8)

Text

..

.. N Comrnonw:alth Edison

• g O / 1400 Opus Place O

V} Downers Grove, Illinois 60515 November 21, 1991 Dr. Taomas E. Murley, Director Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn

Document Control Desk

Subject:

Braidwood Station Unit 2 Cycle 3 Reload NRC Docket No. 50-457

References:

See Attachment 3

Dear Dr. Murley:

Braidwood Unit 2 has completed its second cycle of operation and is conducting a refueling outage that began September 15, 1991. Braidwood Unit 2 Cycle 2 attained a final cycle burnup of approximately 17,100 MHD/MTV. Cycle 3 is expected to commence on November 25, 1991. This letter is to summarize Commonwealth Edison Company's (CECO) plans and evaluation regarding the Braidwood Unit 2 Cycle 3 reload core.

Attachment I describes the core reload including a summary of CECO's safety evaluation, performed in accordance with the provisions of 10CFR50.59 as there are no unreviewed safety issues or necessary Technical Specification changes.

Attachment 2 provides the Core. Operating Limits Report for Cycle 3 pursuant to Technical Specification 6.9.1.9. CECO and its vendor (Westinghouse) apply NRC apprnved reload design methodologies developed by Westinghouse as described in Reference 1. Commonwealth Erlison performed the neutronic portion of the reload design using the-methods and codes described in References 2 & 4 and the NRC approved these requests in References 3 & 5 respectively. -Specifically, the Braidwood Unit 2 Cycle 3 reload design, including the development of the core operating limits, was generated by Commonwealth-Edison using the NRC approved methodologies.

Please direct any questions regarding this notification to thit office.

Very truly yours, 9

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rweddlefW T.H. Simpkin Nuclear Licensing Administrator i

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l A.B. Davis - Regica III Administrator QL- S. DuPont - NRC' Resident Inspector - Braidwood /

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The Braidwood Unit 2, Cycle 3 reload ccre was designed to perform under L current nominal design parameters, Technical Specifications and related bases, and current Technical Specification setpoints such that:

1. Core characteristics will be equivalent or less limiting than those previously reviewed and accepted; or

2. For those postulated incidents analyzed and reported in the Updated Braidwood/ Byron Final Safety Analysis Report (UFSAR) which could potentially be affected by fuel reload, reanalyses or reevaluations have demonstrated that the results of the postulated events are l

within allowable limits.

The Braidwood Unit 2 Cycle 3 core is a " Low Leakage" design.

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

During the Cycle 2/3 refueling, eighty-four (84) VANTAGE 5 fuel assemblies will be inserted into the core. The Braidwood Unit 2 core will then contain a combination of fresh and previously irradiated 17x17 VANTAGE 5 Fuel Assemblies (84 new and 84 once-burned), and 25 twice-burned 17x17 Optimized fuel Assemblies (0FA's). The NRC approved the use of VANTAGE 5 at Braidwood Unit.2 for Cycle 2 and thereafter, under the provisions of 10CFR50.90 in Reference 6. Reference 8 and Braidwood/ Byron UFSAR addressed the compatibility of Westinghouse OFA and VANTAGE-5 assemblies in a reload

, core, and also verified compatibility with control rods and reactor internals L interfaces. A mixture of Integral fuel Burnable Absorber (IFBA) rods and Het Annblar Burnable _ Absorbers (WABAs) will be used as the burnable poison. The IFBA rods contain fuel pellets with enriched B-10 ccating. Both WABAs and IFBA fuel r'ods have been used previously by Commonwealth Edison.

The reload VANTAGE 5 fuel assemblies will incorporate Westinghouse standardized fuel pellets, reconstit" table top nozzles (RTN), extended burnup design features, modified Dcbris filter Bottom Nozzle (DFBN), and snag

_ resistant Intermediate Flow Mixers (IFM) and grids. Sinillar features have o been successfully utilized previously in Commonwealth Edison's Byron and Braidwcod Units and similar units elsewhere.

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The Braidwood Unit 2 Cycle 3 core has been designed and evaluated using NRC licensed and approved methods. Commonwealth Edison requested approval to perform the neutronic portion of the PHR reload design using the methods described in Reference 2, and the NRC approved this request (Reference 3).

Specifically, the Braidwood Unit 2 Cycle 3 reload design, including the development of the core operating limits, was generated and verifted by Commonwealth Edison using NRC :pproved methodology, The reload fuel's nuclear design is evaluated generically in Reference 8 and the UFSAR. As OfA and VANTAGE 5 fuel have the same pellet and fuel rod diameters, most reactivity parameters are insensitive to fuel type. Changes in nuclear characteristics due to the transition from OTA to VANTAGE 5 fuel are within the range normally seen form cycle to cycle due to fuel management effects. The loading pattern dependent parameters wer evaluated in detail in the CECO /Hestinghouse reload safety evaluation process.

Commonwealth Edison has determined that all neutronic reload parameters remain within the previously established and recently revised reload safety and transient Safety Parameter Interaction list (SPIL) limits. These include, but are not limited to SPil items for non-LOCA and LOCA considerations, and have considered the resolution of Westinghouse issue PI-91-020. Issue PI-91-020 addresses the Boron Dilution re-analysis for Byron /Braldwood for CVCS malfunctions in Modes 3, 4, and 5.

The thermal-hydraulic design for the Cycle 3 reload core has not significantly changed from ' oat of the previously reviewed and accepted initial cycle design. The FNDH limits of less than 1.55 for OfA assemblies and 1.65 for VANTAGE 5 assemolies etmoras that the DNB ratio of the limiting power rod during Condition I and Coi Htion 11 events is greater than or equal to the DNBR limit of the DNBR correlation being applied.

Commonwealth Edison's reload safety evaluation process (SPil/RSE review) is a verification to ensure that the previously reviewed and approved accident analyses are not adversely impacted by the cycle specific reload core design.

CECO's Braidwood Unit 2 Cycle 3 Reload Safety Evaluation relied on previously reviewed and accepted analyses reported in the UfSAR, fuel technology reports, the Reference 8 VANTAGE 5 Reload Transition Safety Report (RTSR), and previous reload safety evaluation reports. A detailed review of the core characteristics was performed to determine those parameters affecting the postulated accident analyses reported ir the Braidwood UFSAR. Commonwealth Edison verified that fer those accident analyses presented in the UFSAR, the conclusions were not ai 3cted by the reload core characteristics.  ;

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<- finally,: additional verification of the Braidwood Unit-2 Cycle 3 reload core-design will be performed during the standard reload startup physics tests. These tests include, but are not limited to:

1. A physical inventory of the fuel in the reactor by serial number and location prior to the replacement of the reactor head;

2. Control rod drive tests and drop times;

3. Critical boron concentration measurements;

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

5. Moderator temperature coefficient measurements;

6. Startup power distribution measurements using the incore flux mapping system.

In addition, per the requirements of the NRC SER for CECO's design methods benchmark topical (Reference 3), Braidwood will provide a report to i_ the NRC detailing the results of the startup physics tests.

In summary,. CECO's use of advanced neutronics methods and VANTAGE 5 fuel l (as described in References 2 and 7, respectively) have been previously approved by the NRC (References 3 and 6 respectfully).

Therefore, pending. completion of'the .in-Site and Off-Site Reviews, no additional-prior.NRC review and opproval of the reload core analyses or application for amendment to the traidwood Unit 2 operating license, is

. required as a result of the specific reload design for Cycle 3.

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Braldwood Unit 2 Cycle 3 Operating Limit Report - Fxy Portion This Radial Peating factor Limit Report is provided in accordance with '

Paragraph G.9.1.9 of the Draldwood Unit 2 Nuclear Plant Technical tp:d iitations, the Fxy limits for RATED THERMAL POWER within specified core plan for Cycle 3 shall be:

a. For the lower core region from greater than or equal to 0% to less than or equal to 50%:

RTP

1. f xy less than or equal to 2.140 for all core planes containing bank "0" control rods, and RTP

2. F x

less than or equal to 1.709 for all unfoddedcoreplanes,

b. for the upper core region from greater than 50% to less than or equal to 100%:

RTP

1. F x

less than or equal to 2.140 for all core planes cobtainingbank"0"controlrods,and RTP less than or equal to 1.825 for a'i unrodded

2. F xy cof e planes.

These fxy(z) limits were used to confirm that the heat flux hot channel factor FQ(z) WIll be limited to the Technical Specification values of:

Eg(z)1[Z 5D] [K(z)] for P > 0.5 and, P

fg(z)1[5.00] [K(z)] for P 1 0.5 assuming the most ilmiting axial power distributions expected to 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 Olstributton Control and Load following Procedures",

HCAP-8403, September, 1974. Therefore, these F x limits provide assurance thatthe-initialconditionsassumedintheLOCAknalysisaremet,alongwith the ECCS acceptance criteria of 10 CFR 50.46.

See figure 1 tor a plot of [fg.PRel] vs. Axlai Core Height.

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REFERENCES:

1. Hestinghouse HCAP-9272-P-A, dated October 1985;

" Westinghouse Reload Safety Evaluation Methodology",

(originally issued March 1978).

2. CECO submittal, J.A. S11ady to T.E. Murley dated July 13, 1990; titled " Commonwealth Edison Company Topical Report on Denchmark of PHR Nuclear Design

Methods Using The Phoenix-P and ANC Computer Codes, NRC Docket Nos. 50-295/304, 50-454/455, and 50-456/457".

3. HRC SER on Ceco's Neutronics Topical (Ref. 2) dated March 11, 1991.

4. CECO submittal, F.G. Lentine to H.R. Denton dated July 27, 1983; titled " Zion Stations Units I and 2, Byron Station Units I and 2, Braldwood Station Units 1 and 2, Commonwealth Edison Company Topical, Report on Benchmark of PHR Nuclear Design Methods, NRC Docket i Nos. 50-295/304, 50-454/455, and 50-456/457".

5; NRC SER on CECO's Neutronics Topical (Ref. 5) dated December 13, 1983.

6. NRC Letter from S.P. Sands to T.E. Kovach, -

" Amendment No.23 - Use of VANTAGE 5 fuel", dated April 19, 1990.

7. CECO submittal, S.C. Ilunsader to T.E. Hurley "Braldwood Stations Unit-1 and 2 Application to facility Operating License NPF-72 and NPF-77 dated October.14, 1989.

8. Hes','nghouse reports S. L. Davidson, July 1989;

" VANTAGE 5 Reload Transition Safety Report for the Byron /Draldwood Stations Unit 1 and 2".

Braidwood Unit 2 Cycle 3 Reload Description Att 3-1 l

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