NRC-12-0025, Submittal of 2011 Safety Relief Valve Challenge Report, Main Steam Bypass Line Report, and ECCS Cooling Performance Evaluation Model Changes or Errors Report

Submittal of 2011 Safety Relief Valve Challenge Report, Main Steam Bypass Line Report, and ECCS Cooling Performance Evaluation Model Changes or Errors Report

ML12118A147
Person / Time
Site:	Fermi
Issue date:	04/26/2012
From:	Rachel Johnson Detroit Edison, Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NRC-12-0025
Download: ML12118A147 (9)

Text

Fermi 2 6400 North Dixie Hwy., Newport, MI 48166 Detroit Edison TS 5.6.6 10CFR50.46 April 26, 2012 NRC-12-0025 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001

Reference:

Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

Subject:

Submittal of 2011 Safety Relief Valve Challenge Report, Main Steam Bypass Line Report, and ECCS Cooling Performance Evaluation Model Changes or Errors Report The Fermi 2 Technical Specifications (TS) contains a requirement for submitting an annual report for safety relief valve challenges (TS 5.6.6). Enclosure 1 provides the Safety Relief Valve Challenge Report for 2011. provides the annual Service Life of the Main Steam Bypass Lines Report for 2011. This satisfies the commitment stated in Detroit Edison's letter to the NRC dated November 7, 1986 (VP-86-0154). provides the annual Emergency Core Cooling System (ECCS) Cooling Performance Evaluation Model Changes or Errors Report for 2011. This report is provided in accordance with 10 CFR 50.46(a)(3)(ii).

There are no new commitments included in this document.

Should you have any questions or require additional information, please contact me at (734) 586-5076.

Sincerely, Rodney W. Johnson Manager, Nuclear Licensing A DTE Energy Company

USNRC NRC-12-0025 Page 2

Enclosures:

1. Safety Relief Valve Challenge Report

2. Service Life of Main Steam Bypass Lines Report

3. ECCS Cooling Performance Evaluation Model Changes or Errors Report cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 4, Region III Regional Administrator, Region III Supervisor, Electric Operators, Michigan Public Service Commission

Enclosure 1 To NRC-12-0025 Fermi 2 Safety Relief Valve Challenge Report January 1 to December 31, 2011 Detroit Edison Company NRC Docket No. 50-341 Facility Operating License No. NPF-43 to NRC-12-0025 Page 1 Safety Relief Valve Challenges There were no instances in 2011 where reactor pressure was high enough to require Safety Relief Valve (SRV) actuation. There were also no instances in 2011 where an SRV actuation was demanded by an automatic logic system.

Enclosure 2 To NRC-12-0025 Fermi 2 Service Life of Main Steam Bypass Lines Report January 1 to December 31, 2011 Detroit Edison Company NRC Docket No. 50-341 Facility Operating License No. NPF-43 to NRC-12-0025 Page 1 Service Life of Main Steam Bypass Lines In accordance with Detroit Edison's letter to the NRC dated November 7, 1986 (VP-86-0154),

the cumulative time the main steam bypass lines are operated with the bypass valves between 30 and 45 percent open will be reported annually. A cumulative value of 100 days is not to be exceeded without prior NRC notification.

As discussed in Detroit Edison's letter number VP-86-0154, the bypass lines are acceptable for safe operation when operated within the 100 day constraint. The main steam bypass piping that was installed in 1985 has a service life that will allow it to function for the life of the plant under anticipated operating conditions.

As of December 31, 2011, the main steam bypass lines cumulative usage was 42.91 days.

Enclosure 3 To NRC-12-0025 Fermi 2 ECCS Cooling Performance Evaluation Model Changes or Errors Report January 1 to December 31, 2011 Detroit Edison Company NRC Docket No. 50-341 Facility Operating License No. NPF-43

Enclosure 3 to NRC-12-0025 Page 1 ECCS Cooling Performance Evaluation Model - Analysis of Record On June 23, 2008, Detroit Edison submitted a re-analysis of the SAFER/GESTR Loss of Coolant Accident (LOCA) (Reference 1). This re-analysis established a new licensing basis Peak Clad Temperature (PCT) of 1990°F which was associated with the Upper Bound analysis of the limiting small break LOCA.

Emergency Core Cooling System (ECCS) Cooling Performance Evaluation Model Changes or Errors Since the time of the submittal of the analysis of record identified above, General Electric -

Hitachi (GEH) and Global Nuclear Fuel (GNF) have issued the following notifications which indicated that changes had been made in the ECCS- LOCA analyses inputs that affect Fermi 2:

The below notification letters affect the Licensing Basis PCT for GNF 10x1O fuel type used at Fermi 2.

Notification Letter 2008-01 November 25, 2008 Reference 2 2011-02 July 20, 2011 Reference 3 2011-03 July 20, 2011 Reference 4 Additionally, Detroit Edison determined that the Low Pressure, Coolant Injection (LPCI) pump performance curve used in reference 1 to predict LPCI injection flow as a function of Reactor Pressure Vessel pressure was potentially non-conservative in that it did not allow for normal component degradation near pump shutoff consistent with that allowed under the design at rated flow conditions. Therefore, a design provision for degradation near shutoff was made and the impact on the limiting analyses was performed and documented in Reference 5.

This condition and its net estimated impact on the analysis of record were reported in accordance with 10CFR50.46 on August 19, 2011 (Reference 6). A tabulated summary of the impacts of all errors is provided below.

to NRC-12-0025 Page 2 Current LOCA Model Assessment for GE14 Fuel Description GE14 PCT 10CFR 50.46 Baseline Licensing Basis PCT (Reference 5) PCT = 1990°F 10 CFR 50.46 Notification Letter 2008-001 dated November APCT = 5°F 25, 2008, Impact of Steam Flow Induced Error on Level 3 Setpoint for Small Break LOCA Analysis.

10 CFR 50.46 Notification Letter 2011-02 dated July 20, APCT = 40°F 2011, Impact of Database Error for Heat Deposition on the Peak Cladding Temperature (PCT) for 10x10 fuel bundles.

10 CFR 50.46 Notification Letter 2011-03 dated July 20, APCT = -15°F 2011, Impact of Updated Formulation for Gamma Heat Deposition to Channel Wall for 9x9 and 10x10 Fuel Bundles.

Self-identified non-conservative assumption regarding LPCI APCT = 57°F pump degradation at shutoff pressure.

Net PCT PCT = 2077°F The net PCT of 2077°F provides a 123°F margin to the 2200°F PCT limit in 10 CFR 50.46.

References:

1. Detroit Edison's Letter to USNRC, "Submittal of Plant Specific Emergency Core Cooling System (ECCS)," NRC-08-0046, dated June 23, 2008. (ML081830408).

2. General Electric- Hitachi "10 CFR 50.46 Notification Letter 2008-01," dated November 25, 2008.

3. General Electric- Hitachi "10 CFR 50.46 Notification Letter 2011-02," dated July 20, 2011

4. General Electric- Hitachi "10 CFR 50.46 Notification Letter 2011-03," dated July 20, 2011.

5. Evaluation Report - GEH Report No. 0000-0121-0144-R1, "DTE Energy Enrico Fermi 2 -

Reduced LPCI Flow GE11 and GE14 ECCS-LOCA Evaluation," dated July 2010 (GEH Proprietary).

6. Detroit Edison's Letter to USNRC, "30-Day 10 CFR 50.46 Report - Plant Specific ECCS Evaluation Change," NRC-11-0042, dated August 19, 2011. (ML112340598)