L-2019-202, CFR 50.46 - Emergency Core Cooling System LOCA 30-Day Report

CFR 50.46 - Emergency Core Cooling System LOCA 30-Day Report

ML19325C588
Person / Time
Site:	Duane Arnold
Issue date:	11/19/2019
From:	Parks W Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2019-202
Download: ML19325C588 (2)

Text

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FPL, November 19, 2019 L-2019-202 10 CFRS0.46 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001-Re: N extEra Energy Duane Amold, I.LC Duane Arnold Energy Center, Docket No. 50-331 10 CFR 50.46 - Emergency Core CooliAg System WCA 30-Day Report Florida Power & Light (FPL), on behalf ofNextEra Energy Duane Amokl, I.LC, and pursuant to 10 CFR 50.46(a)(3)(11), is submitting this letter to provide a 30-day report for the Duane Amold Energy Center (DAEq for the emergency core cooling system analysis performed by General Electric Hitachi (GEH)._

One legacy error was identified by GEH that affects the GNF2 Loss-of-Coolant Acciden~ (LOCA) analysis.

As the reported error 1S of a*0°F inipact, the PCT continues to remain within the limits. However, as the cumulative PCT change already exceeds 50 °F for the WCA analysis, a 30 day 10CFRS0.46 report must be issued. Evaluation of the reported errpr has concluded that re-analysis was not required Tbis letter contains no new or revised regulatory commitments.

• Should you have any questions regarding this report, please contact Mr. Steve Catron, Fleet Licensing Manager, at (5,61) 304-6206.

Very truly yours, 11/,r~

William L Parks Safety Assurance and Learning, General Manager -

Florida Power & Light Company Attachment (1) cc: USNRC R~onal Administrator, Region ill USNRC Project Manager, Duane Amold Energy Center USNRC Senior Resident Inspector, Duane Arnold Energy Center Florida Power & Light Company 700 Unrverse Boulevard. Juno Beach, FL 33408

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' L-2019-202

' Attachment 1 Page I 1 of 1 Attachment 1 Duane Arnold GNF2 LOCA PCT 30-Day Report Evaluation Methodology: ,

General Electric, "The GESTR-WCA .and SAFER Models for the Evaluation of the Loss-of Coolant Accident Volume ill - SAFER/ GESTR Application Methodology," NEDE-23785-1-PA, 1

• February 1985. . . . .

Global Nuclear Fuel, Licensing Topical Report, The PRThffi Model for Analysis of Fuel Rod Therinal-Mechanical Performance," Technical Bases - NED<.;:-33256P-A, Qualification - NEDC;-33257P-A, and Application'Methodology- NEDC-33258P-A, September 2010.

General Electric-Hitachi; Duane A.mold Energy' Center GNF2 ECCS-LOCA Evaluation," Engineering Report #0000-0133--6901-RO, DRF 0000--0133--6885-RO,'August 2012.

Evfiluation Model PCT: 1730 °P Net PCT Absolute

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Ef;fect PCT,Effect Prior 10 CPR 50.46 Changes or Error Correc;tions -up to 12/31/2018 -10 op 70 °F (Reference 1)

Prior 10 CPR 50.46 Changes or Error Corrections - 2019 None New 10 CPR 50.46 Changes or Error Correctlons -2019 SAFER Lower Limit on Differential Pressure for )

0 op 0 op Bypass Leakage Sum of 10 CPR 50.46 Changes or ;Errors Corrections -10 op 70 °F The sum of the PCTfrom thr most t?ant anafysis using an accptab/e walllation model and thr. utimatrs ofPCT {mpact for changes and mars identified since this 1720 °F <: 2200 °P anafysis SAFER Lower T.irnit on Differential Pressure for Bypass Leakage

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• The driving differential pressure for forward and backward bypass leakage is limited with an upper and lower hmit in the SAFER code. It was discovered that the limits are implemented correctly on all nine leakage paths except for one, the lower limit for the control rod guide tube to control rod drive housing interface backward leakage path. A SAFER version was developed with the correct lower limit on differential pressure for the contr<;>l rod guide tube to control rod drive housing interface backward leakage path, and this version confirmed that the software code error had no impact on the plant ECCS LOCA event evaluations. The PCT impact of the change/ error 1S estimated to be O °F for Duane A.mold

References:

1) Letter from W. Path (Florida Power & Light Company) to USNRC, "10 CPR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications,"

L-2019-057, March 19, 2019.