ML091740255

Approval of Plant Specific Emergency Core Cooling System (ECCS) Evaluation Model Reanalysis

ML091740255
Person / Time
Site:	Fermi
Issue date:	06/30/2009
From:	Mahesh Chawla Plant Licensing Branch III
To:	Jennifer Davis Detroit Edison
Chawla M, NRR/DORL, 415-8371
References
TAC MD9169
Download: ML091740255 (8)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 30, 2009 Mr. Jack M. Davis Senior Vice President and Chief Nuclear Officer Detroit Edison Company Fermi 2 - 210 NOC 6400 North Dixie Highway Newport, MI 48166

SUBJECT:

FERMI 2 - APPROVAL OF PLANT SPECIFIC EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL REANALYSIS (TAC NO. MD9169)

Dear Mr. Davis:

By letter dated June 23, 2008, as supplemented by letter dated June 10, 2009, Detroit Edison Company, the licensee for Fermi 2, submitted a reanalysis of the Loss of Coolant Accident Analysis. The reanalysis was performed due to an error in the General Electric Plant Specific Emergency Core Cooling System evaluation for Fermi 2. The reanalysis was completed and a new Licensing Basis Peak Clad Temperature was established.

The Nuclear Regulatory Commission (NRC) staff reviewed the request and finds the proposed changes to the Licensing Basis Peak Clad Temperature acceptable. Enclosed is the NRC staff evaluation of the Fermi 2 Loss of Coolant Reanalysis. In case of any further questions, you can contact me at 301-415-8371 or Mahesh.chaw/a@nrc.gov.

Sincerely, Mahesh L. Chawla, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION CONCERNING FERMI 2 PLANT-SPECIFIC EMERGENCY CORE COOLING SYSTEM EVALUATION MODEL REANALYSIS

1.0 INTRODUCTION

By letter dated June 23, 2008 (Agencywide Documents Access and Management Systems (ADAMS) Accession No. ML081830408) and June 10,2009 (ADAMS Accession No. ML091680380), Detroit Edison Company (DECO), the licensee for Fermi 2, submitted a reanalysis of their Loss of Coolant Accident Analysis. The reanalysis was performed due to an error in the General Electric (GE) Plant Specific Emergency Core Cooling System (ECCS) evaluation for Fermi 2. The reanalysis was completed and a new Licensing Basis Peak Clad Temperature was established.

DECO reported an error in the GE Plant Specific ECCS evaluation for Fermi 2 (ADAMS Accession No. ML081130561). DECO identified plans to complete and completed reanalysis of the SAFER/GESTR-Loss-of-Coolant Accident Analysis due to the error. The reanalysis was performed for both the GE11 and GE14 fuel types.

The analysis showed that the small break Loss of Coolant Accident (LOCA) is the limiting Licensing Basis Peak Clad Temperature (LBPCT) for both fuel types. The LBPCT results were 1830 of and 1990 of for GE11 and GE14 fuels, respectively.

2.0 REGULATORY EVALUATION

The Nuclear Regulatory Commission (NRC) staff reviewed the reanalysis performed by DECO based on the SAFER/GESTR-LOCA methods, which were previously approved by the NRC.

The NRC staff reviewed the Fermi 2 application using the following regulatory criteria:

• Title 10 of the Code of Federal Regulations (10 CFR) 50.46, which requires that each boiling or pressurized light-water nuclear power reactor be provided with an ECCS that is designed so that its calculated cooling performance following postulated LOCAs maintains the Peak Cladding Temperature below 2200 of.

10 CFR 50, Appendix K, which outlines the required and acceptable features of evaluation models, such as sources of heat during LOCA, fission product decay heat, and bJowdown phenomena.

• NRC Information Notice 97-78, "Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times."

Enclosure

- 2 NUREG-1764, "Guidance for the Review of Changes to Human Actions."

NUREG-0711, "Human Factors Engineering Program Review Model," Revision 2.

NUREG-0800, "Standard Review Plan", Chapter 18.0, "Human Factors Engineering,"

Revision 1.

3.0 TECHNICAL EVALUATION

DECO performed the Plant Specific ECCS evaluation of both GE11 and GE14 fuel types. The analysis was performed using the SAFER/GESTR-LOCA models, which have been previously approved by the NRC. The limiting break with a single failure for large breaks is the design basis accident (DBA) recirculation suction line break with Division II Battery failure in the case of both fuel types. For both fuel types, the limiting failure for small breaks is the Division I Battery failure.

Fermi 2 is currently licensed to operate at 3430 megawatts thermal (MWt). The Extended Power Uprate (EPU) thermal power is 3952 MWt. The thermal power used for the Appendix K calculation is 4031 MWt. Calculations considered 100 percent core flow, a nominal Vessel Dome Pressure of 1045 psia, and an Appendix K Vessel Dome Pressure of 1060 psia. The Appendix K assumptions also include a requirement to use 1.2 times the fission product decay heat for an infinite operating time, and assumes the use of the Moody model to calculate the break flow for two phase flow. Appendix K also requires that the analysis assume that the reactor power level is at least 1.02 times the licensed power level at the beginning of the transient.

3.1 GE1'I Fuel For GE11 fuel, the SAFER/GESTR-LOCA analysis showed that the limiting large break LOCA event for Fermi 2 is the maximum recirculation suction line break. The event was analyzed with Nominal and Appendix K assumptions. This resulted in the above limiting break and single failure combination of the maximum recirculation suction line break with Division II Battery failure as described above for both Nominal and Appendix K assumptions.

During the maximum recirculation suction line break, the vessel will rapidly depressurize due to large inventory loss through the break. The core will uncover and heat up. When the pressure becomes low enough, the Low Pressure Core Spray (LPCS) system will begin to inject into the core. The Low Pressure Coolant Injection (LPCI) system will also inject into the core when the pressure gets low enough and the LPCI system permissive is reached. The level will eventually recover to the level of the jet pump suction. Appendix K assumes a higher bundle power and decay heat, which causes the Peak Clad Temperature (PCT) for the Appendix K model to be higher than that of the nominal model.

The small break LOCA analysis demonstrated that the limiting single failure for small line breaks is the Division I Battery failure. The event was analyzed with Nominal and Appendix K assumptions. During a small break LOCA event, inventory is lost to containment through the break. At the initiation of the event, high containment pressure is expected to initiate a scram and close the main steam isolation valves (MSIVs). With the MSIVs closed, the condenser will

- 3 remove no heat. The vessel will be losing inventory through the break but not at a rate that will reduce the vessel pressure rapidly. In addition, the high-pressure coolant injection (HPCI) system injection flow may be lost through the break and will therefore be unable to restore level.

The core will heat up due to decay heat and the lack of heat sink. The pressure will slowly drop until it becomes low enough for LPCS and LPCI to inject. More time will pass before LPCS and LPCI can inject compared to the large break because the pressure does not drop as quickly as in the case of the large break. Therefore, the core will heat up more than the large break LOCA analysis indicated and will result in a higher PCT than indicated in the large break evaluation.

Therefore, the small break LOCA with Division I Battery failure would be the limiting PCT.

Non-Recirculation line breaks were found to be non-limiting in previous analyses by DECO. The feedwater line break accident analysis assumed operator actions are required to depressurize the reactor during a Division I Battery failure. This was addressed to DECO to gain understanding of how credit for operator action was applied to this break and failure analysis.

The operators are routinely trained on the Fermi 2 simulator in reactor vessel water level control and use of the plant emergency operating procedures (EOPs). The EOPs instruct the operators to manually depressurize the reactor vessel during this scenario.

After evaluation of the breaks/failures, DECO concluded that the Appendix K overall limiting break is the limiting small recirculation suction line. DECO used Fermi 2 plant-specific variables and the NRC approved SAFER/GESTR-LOCA methodology to calculate a licensing basis PCT of 1830 of for the GE11 fuel.

3.2 GE14 Fuel For GE14 fuel, the SAFER/GESTR-LOCA analysis showed that the limiting large break LOCA event for Fermi 2 is the maximum recirculation suction line break. The event was analyzed with Nominal and Appendix K assumptions. This resulted in the above limiting break and single failure combination of the maximum recirculation suction line break with Division II Battery failure as described above for both Nominal and Appendix K assumptions.

During the maximum recirculation suction line break, the vessel will rapidly depressurize due to large inventory loss through the break. The core will uncover and heat up. When the pressure becomes low enough, LPCS will begin to inject into the core. LPCI will also inject into the core when the pressure gets low enough and the LPCI system permissive is reached. The level will eventually recover to the level of the jet pump suction. Appendix K assumes a higher bundle power and decay heat, which causes the PCT for the Appendix K model to be higher than that of the nominal model.

The small break LOCA analysis demonstrated that the limiting single failure for small line breaks is the Division I Battery failure. The event was analyzed with Nominal and Appendix K assumptions. During a small break LOCA event, inventory is lost to containment through the break. At the initiation of the event, high containment pressure is expected to initiate a scram and close the MSIVs. With the MSIVs closed, the condenser will remove no heat. The vessel will be losing inventory through the break but not at a rate that will reduce the vessel pressure rapidly. In addition, the HPCI injection flow may be lost through the break and will therefore be unable to restore level. The core will heat up due to decay heat and the lack of heat sink. The pressure will slowly drop until it becomes low enough for LPCS and LPCI to inject. More time

- 4 will pass before LPCS and LPCI can inject compared to the large break because the pressure does not drop as quickly as in the case of the large break. Therefore, the core will heat up more than the large break LOCA analysis indicated and will result in a higher PCT than indicated in the large break evaluation. Therefore, the small break LOCA with Division I Battery failure would be the limiting PCT.

Non-Recirculation line breaks were found to be non-limiting in previous analyses by DECO. The feedwater line break accident analysis assumed operator actions are required to depressurize the reactor during a Division I Battery failure. This was addressed by DECO to gain an understanding of how credit for operator action was applied to this break and failure analysis.

The operators are routinely trained on the Fermi 2 simulator in reactor vessel water level control and use of the plant EOPs. The EOPs instruct the operator to manually depressurize the reactor vessel during this scenario.

After evaluation of the breaks/failures, DECO concluded that the Appendix K overall limiting break is the limiting small recirculation suction line. DECO used Fermi 2 plant-specific variables and the NRC approved SAFERIGESTR-LOCA methodology to calculate a licensing basis PCT that was less than the Upper Bound PCT but since the licensing basis PCT cannot be less than the Upper Bound PCT the licensing basis PCT was set at 1990 OF for the GE14 fuel.

3.3 Manual Depressurization During its review, the NRC staff questioned an assumption in the model regarding the completion of depressurization of the reactor pressure vessel (RPV) within 10 minutes. The NRC staff requested additional information from the licensee to demonstrate that the operators have appropriate training, procedures, indications, and time to effect the manual depressurization of the RPV in response to the limiting small break scenario. The licensee submitted additional information on June 10, 2009, confirming the assumption that the operator actions involved in RPV depressurization could be completed in 10 minutes or less, as assumed in the reanalysis.

In its June 23, 2008, letter the licensee stated that"...the feedwater line break basis includes an assumption of operator action to depressurize the reactor during the Division I Battery failure scenario." The time assumed for this action was 10 minutes (600 seconds). The NRC staff questioned this assumption and asked the licensee to validate the estimated time for this action with plant data such as past simulator runs with trained operators.

In its response on June 10, 2009, the licensee clarified that during 2008 Licensed Operator Training (Cycle 2) shift crews were evaluated on a Fermi 2 evaluation scenario that includes Division I Battery failure, an event that requires emergency depressurization using manual actions. In accordance with the EOPs, after reactor water level reaches the top of active fuel, but before it reaches the Minimum Steam Cooling Reactor Water Level, manual emergency depressurization is performed. The ability to recognize the need for manual depressurization is dependent upon the operators' observations of reactor vessel pressure and core narrow-range reactor water level instrumentation.

These indications are designed as safety-related divisional Class 'IE loops. Therefore, the design of the plant ensures the operators are provided with reliable instrumentation necessary to

- 5 indicate the need for the action and provide the necessary feedback that the action is successful. The EOPs used in this training scenario are based on the currently approved Boiling Water Reactor Owners Group Emergency Procedure Guidelines (EPGs)/Severe Accident Guidelines (SAGs).

3.4 NRC Staff Evaluation of Reanalysis The purpose of the Fermi 2 reanalysis was to correct an error that was included in previous calculations. The NRC staff finds that the proposed ECCS LOCA reanalysis for Fermi 2 was performed in accordance with the methods and analytical practices approved by the NRC for the SAFER/GESTR ECCS-LOCA code for the evaluation of large and small break LOCAs.

The limiting break produced a PCT of 1830 of for GE11 fuel and 1990 of for GE14, which are both within the 10 CFR 50.46 limits. The analysis showed that the small break LOCA is the limiting Licensing Basis PCT for both fuel types. The NRC staff finds the analysis to be acceptable and the use of the SAFER/GESTR-LOCA methodology to be correct.

The assumption in the model regarding the completion of depressurization of the RPV within 10 minutes is considered a Critical Task during this simulator evaluation scenario and is required to be correctly performed in order to pass the evaluation. Four licensed operator requalification crews received and passed this evaluation scenario.

Based on the use of EOPs that were developed from approved EPGs and the successful execution of the depressurizations by four different crews in less than the estimated time, the NRC staff concludes that Fermi 2 operators are provided with the training, procedures, instrumentation, and the time necessary to ensure a high probability of success for manual depressurization.

4.0 CONCLUSION

The NRC staff finds that the proposed LOCA reanalysis for Fermi 2 was performed in accordance with the methods and analysis practices approved by the NRC for the SAFER/GESTER-LOCA code for evaluations of LOCA ECCS performance. The analysis produced PCT within the regulatory limit of 2200 of.

- 6

5.0 REFERENCES

1. Letter from Joseph H. Plona (Detroit Edison Company) to U.S. NRC, "Submittal of Plant Specific Emergency Core Cooling System Evaluation (ECCS) lVIodel Reanalysis," dated June 23,2008 (ADAMS Accession No. ML081830408).

2. NEDE-23785-1-P-A Rev. 1, 'The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 3, SAFER/GESTR Application Methodology," October 1984 (ADAMS Accession No. ML090780920).

Principal Contributors: J. Miller, NRR G. Lapinsky, NRR M. Chawla, NRR Date: June 30, 2009

June 30, 2009 Mr. Jack M. Davis Senior Vice President and Chief Nuclear Officer Detroit Edison Company Fermi 2 - 210 NOC 6400 North Dixie Highway Newport, MI 48166 SUB~IECT:

FERMI 2 - APPROVAL OF PLANT SPECIFIC EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL REANALYSIS (TAC NO. MD9169)

Dear Mr. Davis:

By letter dated June 23, 2008, as supplemented by letter dated June 10,2009, Detroit Edison Company, the licensee for Fermi 2, submitted a reanalysis of the Loss of Coolant Accident Analysis. The reanalysis was performed due to an error in the General Electric Plant Specific Emergency Core Cooling System evaluation for Fermi 2. The reanalysis was completed and a new Licensing Basis Peak Clad Temperature was established.

The Nuclear Regulatory Commission (NRC) staff reviewed the request and finds the proposed changes to the Licensing Basis Peak Clad Temperature acceptable. Enclosed is the NRC staff evaluation of the Fermi 2 Loss of Coolant Reanalysis. In case of any further questions, you can contact me at 301-415-8371 or Mahesh.chawla@nrc.gov.

Sincerely, IRA!

Mahesh L. Chawla, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:

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