ML13044A312

Electronic Transmission, Draft Request for Additional Information Regarding 30-Day Report for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46

ML13044A312
Person / Time
Site:	Three Mile Island
Issue date:	02/13/2013
From:	Bamford P J Plant Licensing Branch 1
To:	Khanna M K Plant Licensing Branch 1
Bamford P J
References
TAC ME8237
Download: ML13044A312 (4)

Text

February 13, 2013 MEMORANDUM TO: Meena Khanna, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

FROM: Peter Bamford, Project Manager Plant Licensing Branch I-2

/ra/ Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

SUBJECT:

THREE MILE ISLAND, UNIT NO. 1 - ELECTRONIC TRANSMISSION, DRAFT REQUEST FOR ADDITIONAL INFORMATION REGARDING 30-DAY REPORT FOR EMERGENCY CORE COOLING SYSTEM MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 (TAC NO. ME8237)

The attached draft request for additional information (RAI) was transmitted by electronic transmission on February 12, 2012, to Mr. Thomas Loomis, at Exelon Generation Company, LLC (Exelon, the licensee). This draft RAI was transmitted to facilitate the technical review

being conducted by the Nuclear Regulatory Commission (NRC) staff and to support a conference call (if needed) with Exelon in order to clarify the licensee's submittal reporting a change or error discovered in an Emergency Core Cooling System evaluation model or in the application of such a model that affects the peak cladding temperature (PCT) calculation at Three Mile Island Nuclear Station, Unit 1 (TMI-1). The draft RAI is related to the licensee's submittal dated March 21, 2012. The draft questions were sent to ensure that they were understandable, the regulatory basis was clear, and to determine if the information was previously docketed. Additionally, review of the draft RAI would allow Exelon to evaluate and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not represent an NRC staff position. Docket Nos. 50-289

Enclosure:

As stated February 13, 2013 MEMORANDUM TO: Meena Khanna, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

FROM: Peter Bamford, Project Manager /ra/ Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

SUBJECT:

THREE MILE ISLAND, UNIT NO. 1 - ELECTRONIC TRANSMISSION, DRAFT REQUEST FOR ADDITIONAL INFORMATION REGARDING 30-DAY REPORT FOR EMERGENCY CORE COOLING SYSTEM MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 (TAC NO. ME8237)

The attached draft request for additional information (RAI) was transmitted by electronic transmission on February 12, 2012, to Mr. Thomas Loomis, at Exelon Generation Company, LLC (Exelon, the licensee). This draft RAI was transmitted to facilitate the technical review

being conducted by the Nuclear Regulatory Commission (NRC) staff and to support a conference call (if needed) with Exelon in order to clarify the licensee's submittal reporting a change or error discovered in an Emergency Core Cooling System evaluation model or in the application of such a model that affects the peak cladding temperature (PCT) calculation at Three Mile Island Nuclear Station, Unit 1 (TMI-1). The draft RAI is related to the licensee's submittal dated March 21, 2012. The draft questions were sent to ensure that they were understandable, the regulatory basis was clear, and to determine if the information was previously docketed. Additionally, review of the draft RAI would allow Exelon to evaluate and agree upon a schedule to respond to the RAI. This memorandum and the attachment do not represent an NRC staff position. Docket Nos. 50-289

Enclosure:

As stated DISTRIBUTION

Public RidsNrrPMPBamford LPL1-2 R/F Accession No.: ML13044A312

• via email OFFICE LPL1-2/PM SRXB/BC NAME PBamford CJackson*DATE 02/13/13 02/11/2013 OFFICIAL RECORD COPY DRAFT Enclosure REQUEST FOR ADDITIONAL INFORMATION REGARDING THREE MILE ISLAND NUCLEAR STATION, UNIT 1 30-DAY REPORT FOR EMERGENCY CORE COOLING SYSTEM MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10 CFR 50.46 DOCKET NO. 50-289

By letter dated March 21, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12081A083), Exelon Generation Company, LLC (Exelon, the licensee), sent a notice reporting a change or error discovered in an evaluation model or in the application of such a model that affects the peak cladding temperature (PCT) calculation for Three Mile Island Nuclear Station, Unit 1 (TMI-1). This report was submitted pursuant to the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46, which requires, in part, that licensees report a change in the evaluation model used resulting in a significant change in PCT (greater than 50 degrees Fahrenheit). As described in the statements of consideration published in the Federal Register (FR), the intent of this requirement is to enable the Nuclear Regulatory commission (NRC) staff to establish the safety significance of this change (53 FR 35996-36005).

The submittal dated March 21, 2012, was supplemented by letter dated December 12, 2012 (ADAMS Accession No. ML12349A175). The December 12, 2012, letter referenced an additional letter from AREVA NP Inc. (AREVA), which was submitted to the NRC on December 6, 2012 (ADAMS Accession No. ML12342A381). The following questions pertain to the AREVA submittal, insofar as it applies to the TMI-1 report.

1. For the analyses completed pertaining to the Emergency Core Cooling System (ECCS) bypass error for the lowered loop design, a 2.506-ft peak power location was used, and the analyses for the ECCS bypass error for the raised loop design used a 9.536-ft peak power location. In the December 6, 2012, supplemental letter, the effects of the end-of-bypass timing error are expressed in terms of liquid inventory available to reach the lower plenum and initiate a bottom-up core reflood. The effects of an adiabatic heatup, which is terminated by the core reflood, are also discussed. In consideration of these phenomena, it would appear that a higher elevation in the core would be a more limiting location to evaluate the effects of an error associated with end-of-bypass timing.

Provide information to demonstrate that the bottom-peaked power shape being used for the lowered loop design is conservative and/or appropriate.

2. After evaluating a 177 fuel assembly (FA) lowered loop plant with column weldments modeled for a 205 FA plant, details of the column weldments for a 177 FA plant were developed. The model for column weldments of a 177 FA plant were then used for the analyses of a raised loop plant. Two 177 FA raised loop cases showed that the newly developed column weldments increased PCT by 3 degrees Fahrenheit.

DRAFT It was also reported that the column weldments in a lowered loop plant increased PCT by 35.6 degrees Fahrenheit for an unruptured fuel segment. To determine the effect of column weldments on a ruptured fuel segment, the result is doubled. This resulted in a PCT increase of 71.2 degrees Fahrenheit for a ruptured fuel segment. This was bounded by generically estimating the effect of column weldments to be an increase in PCT of 80 degrees Fahrenheit.

Column weldments in a raised loop plant increased PCT by 8.9 degrees Fahrenheit for an unruptured segment, which is one fourth of the effect seen in the lowered loop design.

a. Provide justification to show that analyzing column weldments modeled for a 177 FA plant has an effect on PCT of the same magnitude in a lowered loop plant as in a raised loop plant.

b. Provide justification to show that the generic 80 degree increase in PCT is bounding for a ruptured fuel segment in a lowered loop plant using the 177 FA column weldment model.

c. Describe the nodalization for the column weldments used in RELAP5 analyses.

3. Provide drawings to compare the column weldment design for a 205 FA plant to the column weldments for the 177 FA plant.