[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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