Source: https://www.federalregister.gov/documents/2006/09/29/E6-16015/exelon-generation-company-llc-byron-station-unit-nos-1-and-2-braidwood-station-unit-nos-1-and-2
Timestamp: 2018-04-26 00:01:11
Document Index: 634144332

Matched Legal Cases: ['art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50']

Federal Register :: Exelon Generation Company, LLC, Byron Station, Unit Nos. 1 and 2; Braidwood Station, Unit Nos. 1 and 2; Environmental Assessment and Finding of No Significant Impact
https://www.federalregister.gov/d/E6-16015 https://www.federalregister.gov/d/E6-16015
The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an exemption from the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 50.60(a), for Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77, issued to Exelon Generation Company, LLC (the licensee), for operation of the Byron Station, Unit Nos. 1 and 2 (Byron), and Braidwood Station, Unit Nos. 1 and 2 (Braidwood), located in Ogle County, Illinois and Will County, Illinois, respectively. Therefore, as required by 10 CFR 51.21, the NRC is issuing this environmental assessment and finding of no significant impact.
The proposed action would allow the use of the methods described in Westinghouse Commercial Atomic Power Report (WCAP)-16143, “Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2,” dated November 2003, in calculating the reactor pressure vessel (RPV) pressure-temperature (P-T) limits for Byron and Braidwood, in lieu of 10 CFR Part 50, Appendix G, “Fracture Toughness Requirements,” paragraph IV.A.2.c as required by 10 CFR 50.60(a).
The proposed action is in accordance with the licensee's application for exemption dated October 3, 2005.
The proposed action is needed because utilization of WCAP-16143 will enhance overall plant safety by widening the P-T operating window, especially in the region of low temperature operations. The primary two safety benefits that would be realized are the following: (1) A reduction in the potential challenges to the low-temperature overpressure protection system and resultant inadvertent opening of a power operated relief valve, and (2) a reduction in the risk of damaging the reactor coolant pump seals due to pump operation under conditions in which it is difficult to maintain adequate seal differential pressure to ensure proper pump operation.
Appendix G to 10 CFR Part 50 contains requirements for P-T limits for the primary system and requirements for metal temperature of the closure head flange and vessel flange regions. The P-T limits are to be determined using the methodology of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Appendix G, but the flange temperature requirements are specified in 10 CFR Part 50, Appendix G. This regulation (Table 1 of 10 CFR Part 50, Appendix G) states that the metal temperature at the closure flange regions must exceed the material unirradiated nil-ductility transition reference temperature (RTNDT) by at least 120 °F for normal operation when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure.
This requirement was originally based on concerns about the fracture margin in the closure flange region. During the boltup process, outside surface stresses in this region typically reach over 70 percent of the steady state stress, without being at steady state temperature. The margin of 120 °F and the pressure limitation of 20 percent of hydrostatic pressure were developed in the mid-1970s using the ASME Code lower bound crack arrest/dynamic test fracture toughness (KIa) to ensure that appropriate margins would be maintained.
Improved knowledge of fracture toughness and other issues that affect the integrity of the reactor vessel have led to the recent change to allow the use of the ASME Code lower bound static crack initiation fracture toughness (KIc) in the development of P-T curves, as contained in ASME Code Case N-640, “Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1.” ASME Code Case N-640 has been approved for use without conditions by the NRC staff in Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,” published in August 2005.
However, P-T limit curves can still produce operational constraints by limiting the operational range available Start Printed Page 57578to the operator during heatup and cooldown of the plant, especially when considering requirements in the closure head flange and the vessel flange regions. Implementing the P-T curves that use KIc material fracture toughness without exempting the flange requirement of 10 CFR Part 50, Appendix G, would place a restricted operating window in the temperature range associated with the closure head flange and reactor vessel flange, without a commensurate increase in plant safety.
The NRC has completed its evaluation of the proposed action and concludes that the more conservative minimum temperature requirements related to footnote (2) to Table 1 of 10 CFR Part 50, Appendix G are not necessary to meet the underlying intent of 10 CFR Part 50 Appendix G, to protect the Byron and Braidwood RPVs from brittle fracture during normal operation under both core critical and core non-critical conditions and RPV hydrostatic and leak test conditions.
The action does not involve the use of any different resources than those previously considered in the Final Environmental Statement for the Byron and Braidwood stations, NUREG-0848 dated April 1982, and NUREG-1026 dated June 1984, respectively.
In accordance with its stated policy, on June 19, 2006, the NRC staff consulted with the Illinois State official, Mr. Frank Niziolek of the Illinois Emergency Management Agency, regarding the environmental impact of the proposed action. The State official had no comments.
For further details with respect to the proposed action, see the licensee's letter dated October 3, 2005. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room (PDR), located at One White Flint North, Public File Area O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible electronically from the Agencywide Documents Access and Management System (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS should contact the NRC PDR Reference staff by telephone at 1-800-397-4209 or 301-415-4737, or send an e-mail to pdr@nrc.gov.
Dated at Rockville, Maryland, this 22nd day of September 2006.
Project Manager Plant Licensing Branch III-2, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
[FR Doc. E6-16015 Filed 9-28-06; 8:45 am]