L-PI-06-070, Supplement to License Amendment Request to Incorporate Large Break Loss of Coolant Accident Analyses Using Automated Statistical Treatment of Uncertainty Method

Supplement to License Amendment Request to Incorporate Large Break Loss of Coolant Accident Analyses Using Automated Statistical Treatment of Uncertainty Method

ML062610088
Person / Time
Site:	Prairie Island
Issue date:	09/15/2006
From:	Thomas J. Palmisano Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-06-070
Download: ML062610088 (4)

Text

"9 Committed to Nuclear Excellence Prairie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC SEP 1 5 2006 L-PI-06-070 10 CFR 50.90 U S Nuclear Regulatory Commission 10 CFR 50.46 ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Supplement to License Amendment Request (LAR) to Incorporate Large Break Loss Of Coolant Accident (LOCA) Analyses Using the Automated Statistical Treatment of Uncertaintv Method By letter dated July 6,2006, Nuclear Management Company (NMC) submitted an LAR to incorporate new Large Break LOCA (LBLOCA) analyses using the realistic LBLOCA methodology in the NRC approved WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)" and revise Technical Specification (TS) 5.6.5.b to include reference to WCAP-16009-P-A. This letter supplements the LAR to address NRC requests for additional information (RAls) regarding this M R . NMC submits this supplement in accordance with the provisions of 10 CFR 50.90. provides the NRC RAls and NMC responses.

The supplemental information provided in this letter and enclosure does not impact the conclusions of the Determination of No Significant Hazards Consideration and Environmental Assessment presented in the July 6, 2006 submittal.

In accordance with 10 CFR 50.91, NMC is notifying the State of Minnesota of this LAR by transmitting a copy of this letter and enclosure to the designated State Official.

Summarv of Commitments This letter contains no new commitments and no revisions to existing commitments.

1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1 121

Document Control Desk Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on SEP 1 5 2006 Thomas J. Palmisano Site Vice President, Prairie Island Nuclear Generating Plant Units 1 and 2 Nuclear Management Company, LLC Enclosures (1) cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota

Enclosure 1 NRC Question:

Q1. To show that the referenced generically approved LBLOCA analysis methodology applies specifically to each of the Prairie lsland plants, provide a statement that NMC and its vendor have ongoing processes that assure that the ranges and values of the input parameters for the Prairie lsland Unit 1 LBLOCA analyses, and for the Prairie lsland Unit 2 LBLOCA analyses, conservatively bound the ranges and values of the as-operated plant parameters.

Nuclear Management Company, LLC (NMC) response:

NMC and its vendor (Westinghouse) have ongoing processes that assure that the ranges and values of the input parameters for the Prairie lsland Nuclear Generating Plant (PINGP), Unit 1 Large Break Loss of Coolant Accident (LBLOCA) analyses conservatively bound the ranges and values of the as-operated PINGP Unit 1 parameters. Likewise, NMC and its vendor (Westinghouse) have ongoing processes that assure that the ranges and values of the input parameters for the PINGP, Unit 2 LBLOCA analyses conservatively bound the ranges and values of the as-operated PINGP Unit 2 parameters.

Q2. The LBLOCA analyses results for Prairie lsland Unit 1 (reported in Table 1-2) and Prairie lsland Unit 2 (reported in Table 2-2), each report "0" core wide oxidation. Even at the relatively low peak cladding temperatures calculated, this is not consistent with staff expectations.

The amount of total core-wide oxidation should be estimated and reported, regardless of how small that amount is.

NMC response:

The amount of core-wide oxidation was conservatively estimated for PINGP Units 1 and

2. The total amount of core-wide oxidation for both PINGP Unit 1 and Unit 2 is less than 0.01%.

Page 1 of 2

L-PI-06-070 Q3. Please describe how pre-existing oxidation is factored into the results for maximum local oxidation (LMO). If pre-existing oxidation is not factored into the reported results, please provide an estimate of the contribution.

NMC response:

The pre-existing oxidation was not factored into the maximum local oxidation results presented for PlNGP Unit 1 and Unit 2. The maximum expected total of the normal operation (pre-transient) and loss of coolant accident (LOCA) transient oxidation, for any time in life, was considered for the PlNGP optimized fuel assembly (OFA) fuel. The pre-transient oxidation increases with burnup, from zero at beginning of life (BOL) to a maximum value at the discharge of the fuel (end of life, or EOL). The transient oxidation decreases from the near BOL value of 0.2% for PlNGP Unit 1 and 0.5% for PlNGP Unit 2 to a negligible value at EOL. The sum of the pre-transient plus transient oxidation remains below 17% at all times in life for the OFA fuel.

Page 2 of 2