[Federal Register Volume 71, Number 189 (Friday, September 29, 2006)]
[Notices]
[Pages 57577-57578]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-16015]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. STN 50-454, STN 50-455, STN 50-456 and STN 50-457]


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

    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.

Environmental Assessment

Identification of the Proposed Action

    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 Need for the Proposed Action

    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 [deg]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 [deg]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

[[Page 57578]]

to 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.

Environmental Impacts of the Proposed Action

    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 details of the NRC staff's safety evaluation will be provided 
in the exemption that will be issued as part of the letter to the 
licensee approving the exemption to the regulation.
    The proposed action will not significantly increase the probability 
or consequences of accidents. No changes are being made in the types of 
effluents that may be released off site. There is no significant 
increase in the amount of any effluent released off site. There is no 
significant increase in occupational or public radiation exposure. 
Therefore, there are no significant radiological environmental impacts 
associated with the proposed action.
    With regard to potential non-radiological impacts, the proposed 
action does not have a potential to affect any historic sites. It does 
not affect non-radiological plant effluents and has no other 
environmental impact. Therefore, there are no significant non-
radiological environmental impacts associated with the proposed action.
    Accordingly, the NRC concludes that there are no significant 
environmental impacts associated with the proposed action.

Environmental Impacts of the Alternatives to the Proposed Action

    As an alternative to the proposed action, the NRC staff considered 
denial of the proposed action (i.e., the ``no-action'' alternative). 
Denial of the application would result in no change in current 
environmental impacts. The environmental impacts of the proposed action 
and the alternative action are similar.

Alternative Use of Resources

    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.

Agencies and Persons Consulted

    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.

Finding of No Significant Impact

    On the basis of the environmental assessment, the NRC concludes 
that the proposed action will not have a significant effect on the 
quality of the human environment. Accordingly, the NRC has determined 
not to prepare an environmental impact statement for the proposed 
action.
    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 [email protected].

    Dated at Rockville, Maryland, this 22nd day of September 2006.

    For the Nuclear Regulatory Commission.
Robert F. Kuntz,
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]
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