ML080440184

Summary of Telephone Conference Call Held on February 6, 2008, Concerning Comments Related to the Proposed Rule on the Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (Rin 3150-AI01, RM-668)

ML080440184
Person / Time
Site:	Oconee, Mcguire, Catawba, McGuire
Issue date:	02/14/2008
From:	Rodriguez V M NRC/NRR/ADRA/DPR/PRAB
To:	Harrall T P Duke Energy Corp
Rodriguez V M NRR/DLR/RLRB 415-3703
References
RIN 3150-AI01, RM-668
Download: ML080440184 (5)

Text

February 14, 2008 Thomas P. Harrall, Jr., VP Plant Support Nuclear Generation Duke Energy Corporation P.O. Box 1006 Charlotte, NC 28201-1006

SUMMARY

OF TELEPHONE CONFERENCE CALL HELD ON FEBRUARY 6, 2008, CONCERNING COMMENTS RELATED TO THE PROPOSED RULE ON THE ALTERNATE FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL SHOCK EVENTS (RIN 3150-AI01, RM-668)

The U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of Duke Energy and Information Systems Laboratory held a telephone conference call on February 6, 2008. The purpose of the telephone conference call was to clarify comments provided by Duke Energy on the subject proposed rule.

provides a list of the participants and Enclosure 2 contains a list of the comments discussed, including a summary of the discussion.

/RA/ Veronica M. Rodriguez, Project Manager Regulatory Analysis, Policy and Rulemaking Division of Policy and Rulemaking Office of Nuclear Reactor Regulations

Enclosures:

1. List of Participants 2. List of Comments

ML080440184 OFFICE PM:PRAB:DPR BC:PRAB:DPR NAME VRodriguez JZimmerman DATE 02/14/2008 02/14/2008

Enclosure 1 PROPOSED RULE ON THE ALTERNATE FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL SHOCK EVENTS TELEPHONE CONFERENCE CALL LIST OF PARTICIPANTS February 6, 2008 PARTICIPANTS AFFILIATIONSVeronica M. Rodri guezNuclear Re gulator y Commission (NRC) Robert Hardies NRC Mark Kirk NRC Bill Arciere Information S ystems Laboratories (ISL) Don Fletcher ISL R. L. Gill Duke Ener gy (Duke)Gre gg Swindlehurst Duke

Enclosure 2 PROPOSED RULE ON THE ALTERNATE FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL SHOCK EVENTS TELEPHONE CONFERENCE CALL COMMENTS DISCUSSED February 6, 2008

The U.S. Nuclear Regulatory Commission (NRC or the staff) held a telephone conference call on February 6, 2008 to clarify comments provided on the subject proposed rule. The following comment was discussed.

Comment 1:

The summary report for NUREG-1806, along with the apparent assignment of event sequences to an insignificant bin in the probabilistic risk assessment (PRA) report, prompt the following concerns:

(1) Some main feedwater overfeed Babcock & Wilcox (B&W) design. The B&W design will overcool more rapidly than other pressurized water reactor (PWR) designs because of the once-through steam generators. The initial secondary water inventory is low, and the overfeed will immediately influence the rate of heat transfer. The event progresses to a counter flow water-solid heat exchange process, and the temperature of the primary side cold leg water returning from a steam generator will approach the main feedwater temperature. This low cold leg water temperature along with the cold safety injection water has the potential to severely overcool the reactor vessel. Insights based on overfeed analyses for PWR designs with u-tube steam generators are not applicable to the B&W design.

(2) The overfeed events that were analyzed are described as only filling to the top of the steam generator. Perhaps this assumption of a limited duration overfeed is supported by the plant design and/or by operator recovery actions credited by the PRA. A continued overfeed would be more severe relative to pressurized thermal shock (PTS).

(3) The PRA report considers a zero power (low decay heat) initial plant condition. That initial condition is much more severe for main feedwater overfeed events. Thermal hydraulic analyses of main feedwater overfeed events should consider this initial condition.

(4) The statement in the summary report "... the extent of the cooldown is limited because the ultimate heat sink temperature is the saturation temperature at atmospheric pressure" is not correct for a B&W design. The extent of the cooldown for a main feedwater overfeed is related to the main feedwater temperature, which will be low at zero power with no preheating, and the primary cooldown will be enhanced by the cold safety injection water.

Enclosure 2 The significance of the above comments in the overall integrated risk due to PTS for B&W design plants is not known, but additional consideration of the issues summarized above is warranted. There is a possibility that the conclusions drawn in the references may be incomplete.

Discussion: The staff informed Duke Energy (Duke) that this comments required clarification. The staff expressed concerns with the possibility that Duke could be using a different model to determine how overfeed scenarios progresses. The staff requested that Duke explain the comment.

Duke stated that their main concern is that it seems that the overfeed transients were improperly modeled for B&W plants, specifically for Oconee. Duke expressed concerns with the scenario when either the integrated control system or a human operator fails to terminate main feedwater. Duke stated that these scenarios were not addressed and that they could lead to sever overcooling that had not been properly accounted for in the PTS study. The staff explained that the scenario described had indeed been modeled in initial scoping studies, but had not been documented. The staff stated that when the sequence was modeled, as described, similar results were obtained. However, it was decided that the failure of both, the integrated control system and the operator action, was a low probability event; and therefore, was excluded from the base case. The staff informed Duke that the PRA reports on Oconee will be reviewed to clarify the details of how the overfeed sequences were modeled.

The NRC will publish the formal response to this comment in the Federal Register.