ML14175B390
| ML14175B390 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/20/2014 |
| From: | Polson K Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML14175B390 (7) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 June 20, 2014 10 CFR 50.4 10 CFR 50.46 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 1 Renewed Facility Operating License No. DPR-33 NRC Docket No. 50-259
Subject:
10 CFR 50.46 30-Day Report for Browns Ferry Nuclear Plant, Unit I
Reference:
Letter from TVA to NRC, "10 CFR 50.46 Annual Report for Browns Ferry Nuclear Plant, Units 1 and 2,and 10 CFR 50.46 30-Day and Annual Report for Browns Ferry Nuclear Plant, Unit 3," dated April 15, 2014 (ML14108A327)
The purpose of this letter is to provide a 30-day report of changes and errors to the calculated peak cladding temperature (PCT) for Browns Ferry Nuclear Plant (BFN), Unit 1, Emergency Core Cooling System (ECCS) evaluation model. This report is required in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 50.46, "Acceptance Criteria for ECCS for Light-Water Nuclear Power Reactors," paragraph (a)(3)(ii).
The previous 10 CFR 50.46 report for BFN, Unit 1, was submitted on April 15, 2014 (Reference).
On May 23, 2014, the Tennessee Valley Authority (TVA) received reports of four errors/changes affecting the GE14 Loss of Coolant Accident (LOCA) analyses for BFN, Unit 1. The PCT changes and errors identified for BFN, Unit 1, described in the enclosed report, when expressed as the cumulative sums of the absolute magnitudes, exceed 50 degrees Fahrenheit (°F). In accordance with 10 CFR 50.46(a)(3)(ii), a holder of an operating license or construction permit is required to report changes and errors affecting an ECCS evaluation model to the NRC within 30 days when the cumulative sum of the absolute magnitudes of resulting PCT changes exceeds 50'F. The licensee is also required to include with the report, a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with the 10 CFR 50.46 requirements.
As presented in this report, compliance with 10 CFR 50.46 requirements is demonstrated by the calculated PCT for BFN, Unit 1, remaining below the 2200°F limit. Therefore, TVA has concluded no proposed schedule for providing a reanalysis or other action is required.
U.S. Nuclear Regulatory Commission Page 2 June 20, 2014 There are no new regulatory commitments in this letter. Please direct questions concerning this issue to Jamie L. Paul at (256) 729-2636.
Respectfully, K. J. Poison Site Vice President
Enclosures:
10 CFR 50.46 30-Day Report for Browns Ferry Nuclear Plant, Unit 1 cc (w/Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant
ENCLOSURE 1 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT I The Browns Ferry Nuclear Plant, Unit 1, core contains both the ATRIUM-10 and GEl4 fuel designs. This report documents errors and changes in the GEl4 Loss of Coolant Accident (LOCA) analyses.
ATRIUM-10 Fuel Evaluation The previous 10 CFR 50.46 report for BFN, Unit 1, was submitted on April 15, 2014 (Reference 1). This report cites References 2 and 3 as the Analyses of Record (AOR) for ATRIUM-10 fuel, with a baseline Peak Clad Temperature (PCT) for ATRIUM-10 fuel of 19440F.
No new changes or errors have been identified in the ATRIUM-1 0 LOCA analyses since the issuance of Reference 1.
Table 1 details the accumulated PCT impact due to errors and changes in the LOCA analyses since the AOR in Reference 3 of this enclosure.
Table 1: Cumulative Effect of PCT Changes - BFN, Unit I (ATRIUM-I0)
Baseline PCT (Reference 3) 19440F Thermal Conductivity Degradation (previously reported)
+ 0°F Accumulated changes since baseline analysis
+ 0°F New licensing PCT 19440F Sum of absolute value of individual changes
+ 0°F El -1 of 5
ENCLOSUREI 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT I GE14 Fuel Evaluation The previous 10 CFR 50.46 report for BFN, Unit 1, was submitted on April 15, 2014 (Reference 1). This report cites Reference 4 as the AOR for GE14 fuel. The applicability of this analysis to the current plant configuration was confirmed by GE Hitachi Nuclear Energy (GEH) in Reference 5. Reference 4 provides PCT results for both Extended Power Uprate (EPU) and Current Licensed Thermal Power (CLTP) conditions. The Tennessee Valley Authority (TVA) has elected to use the CLTP results for 10 CFR 50.46 reporting, because EPU has not been approved for BFN, Unit 1, and all GE14 fuel is scheduled to be discharged prior to the planned EPU implementation date. The baseline PCT for GE14 fuel at CLTP conditions is 1760°F.
On May 23, 2014, TVA received reports from GEH (References 6 through 9) documenting four errors/changes affecting the GE14 LOCA analyses for BFN, Unit 1.
SAFERO4A E4-Maintenance Update Changes A new version of SAFER04A has been released (E4). This version resolves several accumulated observations docketed on software control tracking tools, which are reported as Evaluation Model changes for an ECCS-LOCA calculation utilizing the updated SAFER code compared to earlier versions. The following updates are reported (Reference 6):
- 1. Fix for a non-physical, low prediction of channel wall temperature for quenched nodes due to alternate heat transfer mechanisms while the quench front traverses the node;
- 2. Fix of a logic flaw which slowed progression of the re-wetting front (calculated on the basis of average rods, applied to hot rods) as a result of quench front reaching two-phase level;
- 3. Fix of the boundary treatment between downcomer steam and liquid nodes - for time after feedwater is terminated until the liquid node reaches saturation - before the nodes are combined into one region;
- 4. Fix of input manipulation in core wide metal water reaction calculation, which inadvertently replaced the smeared local peaking factor constant with the input unsmeared local peaking factor in the subsequent group calculation, causing the fraction of rods assigned to each group to be slightly different;
- 5. Fix of a rare occurrence wherein the code aborts when attempting to divide by zero in the pressure rate of change calculation, checking mass of the node when the core is partially empty; and
- 6. Removal of a discontinuity when using the dynamic gap conductance model, if the fuel rod gap is initially closed and the calculation predicts gap opening.
El - 2 of 5
ENCLOSUREI 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT 1 SAFERO4A E4-Mass Non-Conservatism A new version of SAFER04A has been released (E4). This version resolves several accumulated observations docketed on software control tracking tools, the current item being reported as an Evaluation Model error for an ECCS-LOCA calculation utilizing earlier versions and corrected in the updated SAFER code revision. A logic error has been isolated occurring with an indication that the expected system mass diverges from the calculated actual system mass. This occurs when upper plenum liquid mass and core spray flow rate is low; system mass is gradually lost due to core spray being discarded, resulting in marginally less ECCS flow credited as reaching the core (Reference 7).
SAFERO4A E4-Minimum Core DP Model A new version of SAFER04A has been released (E4). This version resolves several accumulated observations docketed on software control tracking tools, the current item being reported as an Evaluation Model error for an ECCS-LOCA calculation utilizing earlier versions and corrected in the updated SAFER code revision. Due to calculation of a non-physically low delta pressure (Ap) for droplet flow above a two-phase level in the core, an earlier version of the model imposed a minimum core Ap. It has been observed that for cores with greater voiding (more steam flow), this minimum Ap could be non-conservative, actually driving the steam flow slightly, and offering inappropriate steam cooling benefit above the core two-phase level (Reference 8).
SAFER04A E4-Bundle/Lower Plenum CCFL Head A new version of SAFER04A has been released (E4). This version resolves several accumulated observations docketed on software control tracking tools, the current item being reported as an Evaluation Model error for an ECCS-LOCA calculation utilizing earlier versions and corrected in the updated SAFER code revision. A counter current flow limitation (CCFL) is applied on the interface between the hot bundle and the lower plenum. The pressure head applied at that location is based on the liquid water level in the bundle. It was found, upon exercising the routine to define CCFL, the output would replace the pressure head with a value revised by that calculation, resulting in a representation of pressure head slightly different from that of the calculated water level in the bundle. The iteration scheme for CCFL has been fixed in the SAFER04A E4 model so that, consistently, the level head is applied whenever CCFL is calculated in that location (Reference 9).
El - 3 of 5
ENCLOSURE I 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT 1 Table 2 details the accumulated PCT impact due to errors and changes in the GE14 LOCA analyses since the AOR in Reference 4 of this enclosure.
Table 2: Cumulative Effect of PCT Changes - Unit I (GEI4)
Baseline PCT 1760°F Input coefficient database error (previously reported)
+ 250F Revised gamma heat deposition formulation (previously reported)
+ 150F Pellet thermal conductivity degradation (previously reported)
+ 0°F SAFER04A E4-Maintenance Update Changes
+ 0°F SAFER04A E4-Mass Non-conservatism
+ 10°F SAFER04A E4-Minimum Core DP model
+ 20°F SAFER04A E4-Bundle/Lower Plenum CCFL Head
- 20°F Accumulated changes since baseline analysis 50°F New licensing PCT 1810°F Sum of absolute values of individual changes 90°F El - 4 of 5
ENCLOSUREI 10 CFR 50.46 30-DAY REPORT FOR BROWNS FERRY NUCLEAR PLANT, UNIT I References
- 1. TVA Letter to NRC, "10 CFR 50.46 Annual Report for Browns Ferry Nuclear Plant, Units 1 and 2, and 10 CFR 50.46 30-Day and Annual Report for Browns Ferry Nuclear Plant, Unit 3," dated April 15, 2014.
- 2. AREVA NP Inc., "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis,"
ANP-3015(P), Revision 0, September 2011.
- 3. AREVA NP Inc., "Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM TM-10 Fuel," ANP-3016(P) Revision 1, November 2013.
- 4. GE Nuclear Energy, "Browns Ferry Nuclear Plant Units 1, 2, and 3: SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-32484P, Revision 6, February 2005.
- 5. GE-Hitachi Nuclear Energy, "Browns Ferry Nuclear Plant Unit 1: Supplementary Report Regarding ECCS-LOCA Evaluation Additional Single Failure Evaluation at Current Licensed Thermal Power," NEDC-32484P Revision 6, Supplement 2, Revision 0, September 2012.
- 6. 10 CFR 50.46 Notification Letter from GEH to TVA, "Browns Ferry Nuclear Plant (Unit 1)
SAFER04A E4-Maintenance Update Changes," 2014-01, May 21, 2014.
- 7. 10 CFR 50.46 Notification Letter from GEH to TVA, "Browns Ferry Nuclear Plant (Unit 1)
SAFER04A E4-Mass Non-Conservatism," 2014-02, May 21, 2014.
- 8. 10 CFR 50.46 Notification Letter from GEH to TVA, "Browns Ferry Nuclear Plant (Unit 1)
SAFER04A E4-Minimum Core DP Model," 2014-03, May 21, 2014.
- 9. 10 CFR 50.46 Notification Letter from GEH to TVA, "Browns Ferry Nuclear Plant (Unit 1)
SAFER04A E4-Bundle/Lower Plenum CCFL Head," 2014-04, May 21, 2014.
El -5of5