[Federal Register Volume 65, Number 173 (Wednesday, September 6, 2000)]
[Notices]
[Pages 54081-54082]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-22781]


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NUCLEAR REGULATORY COMMISSION

[DOCKET NO. 50-352]


Peco Energy Company; Limerick Generating Station, Unit 1; 
Environmental Assessment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (NRC) is considering 
issuance of an exemption from certain requirements of 10 CFR 50.60(a) 
for Facility Operating License No. NPF-39, issued to PECO Energy 
Company (PECO, or the licensee) for operation of the Limerick 
Generating Station, Unit 1 (Limerick Unit 1), located in Montgomery and 
Chester Counties in Pennsylvania.

Environmental Assessment

Identification of the Proposed Action

    Appendix G to Title 10 of the Code of Federal Regulations, Part 50 
(10 CFR Part 50, Appendix G), requires that pressure-temperature (P-T) 
limits be established for reactor pressure vessels (RPVs) during normal 
operating and hydrostatic or leak rate testing conditions. 
Specifically, 10 CFR Part 50, Appendix G, states, ``The appropriate 
requirements on both the pressure-temperature limits and the minimum 
permissible temperature must be met for all conditions.'' Appendix G of 
10 CFR Part 50 specifies that the requirements for these limits are the 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME Code), Section XI, Appendix G, limits.
    To address provisions of amendments to the technical 
specifications' P-T limits, the licensee requested in its submittal 
dated May 15, 2000, as supplemented May 19, 2000, that the staff exempt 
Limerick Unit 1 from application of specific requirements of 10 CFR 
Part 50, Section 50.60(a) and Appendix G, and substitute use of ASME 
Code Cases N-588 and N-640. Code Case N-588 permits the postulation of 
a circumferentially-oriented flaw (in lieu of an axially-oriented flaw) 
for the evaluation of the circumferential welds in RPV P-T limit 
curves. Code Case N-640 permits the use of an alternate reference 
fracture toughness (KIC fracture toughness curve instead of 
KIA fracture toughness curve) for reactor vessel materials 
in determining the P-T limits. Since the pressure stresses on a 
circumferentially-oriented flaw are lower than the pressure stresses on 
an axially-oriented flaw by a factor of two, using Code Case N-588 for 
establishing the P-T limits would be less conservative than the 
methodology currently endorsed by 10 CFR Part 50, Appendix G, and 
therefore, an exemption to apply the Code Case would be required by 10 
CFR 50.60. Likewise, since the KIC fracture toughness curve 
shown in ASME Code, Section XI, Appendix A, Figure A-2200-1 (the 
KIC fracture toughness curve) provides greater allowable 
fracture toughness than the corresponding KIA fracture 
toughness curve of ASME Code, Section XI, Appendix G, Figure G-2210-1 
(the KIA fracture toughness curve), using Code Case N-640 
for establishing the P-T limits would be less conservative than the 
methodology currently endorsed by 10 CFR Part 50, Appendix G, and 
therefore, an exemption to 10 CFR 50.60 to apply the Code Case would 
also be required.
    The proposed action is in accordance with the licensee's 
application for exemption dated May 15, 2000, as supplemented May 19, 
2000.

The Need for the Proposed Action

    ASME Code Case N-640 is needed to revise the method used to 
determine the reactor coolant system (RCS) P-T limits, since continued 
use of the present curves unnecessarily restricts the P-T operating 
window. Since the RCS P-T operating window is defined by the P-T 
operating and test limit curves developed in accordance with the ASME 
Code, Section XI, Appendix G, procedure, continued operation of 
Limerick Unit 1 with these P-T curves without the relief provided by 
ASME Code Case N-640 would unnecessarily require the RPV to maintain a 
temperature exceeding 212  deg.F in a limited operating window during 
the pressure test. Consequently, steam vapor hazards would continue to 
be one of the safety concerns for personnel conducting inspections in 
primary containment. Implementation of the proposed P-T curves, as 
allowed by ASME Code Case N-640, does not significantly reduce the 
margin of safety and would eliminate steam vapor hazards by allowing 
inspections in primary containment to be conducted at a lower coolant 
temperature.
    ASME Code Case N-588 allows a licensee to postulate a 
circumferential flaw in circumferential RPV welds in lieu of the axial 
flaw that is normally assumed to be present by the ASME Code, Section 
XI, Appendix G, analysis. The staff has determined that the assumption 
of an axial flaw in a circumferential RPV shell weld would provide an 
overly-conservative margin of safety on stress intensities resulting 
from the operating pressure, and that postulation of a circumferential 
flaw in the circumferential welds would continue to satisfy the margin 
of safety of two required by Appendix G to Section XI of the ASME Code.
    In the requests for exemptions to use Code Cases N-588 and N-640, 
the staff has determined that, pursuant to 10 CFR 50.12(a)(2)(ii), the 
underlying purpose of the regulation will continue to be served by the 
implementation of these Code Cases.

Environmental Impacts of the Proposed Action

    The NRC has completed its evaluation of the proposed action and 
concludes that the exemption described above would provide an adequate 
margin of safety against brittle failure of the Limerick Unit 1 RPV.
    The proposed action will not significantly increase the probability 
or consequences of accidents, no changes are being made in the types of 
any effluents that may be released offsite, and 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 nonradiological environmental impacts, the 
proposed action does not involve any historic sites. It does not affect 
nonradiological plant effluents and has no other environmental impacts. 
Therefore, there are no significant nonradiological impacts associated 
with the proposed action.
    Accordingly, the NRC concludes that there are no significant 
environmental impacts associated with the proposed action.

Alternatives to the Proposed Action

    As an alternative to the proposed action, the 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

    This action does not involve the use of any resources not 
previously considered in the Final Environmental

[[Page 54082]]

Statement for the Limerick Generating Station, Units 1and 2, dated 
April 1984.

Agencies and Persons Consulted

    In accordance with its stated policy, on August 7, 2000, the staff 
consulted with the Pennsylvania State official, David Ney of the 
Pennsylvania Department of Environmental Protection, 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 May 15, 2000, as supplemented by letter dated 
May 19, 2000, which are available for public inspection at the NRC 
Public Document Room, The Gelman Building, 2120 L Street, NW., 
Washington, DC. Publicly available records will be accessible 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room).

    Dated at Rockville, Maryland, this 29th day of August, 2000.

    For The Nuclear Regulatory Commission.
Bartholomew C. Buckley,
Sr. Project Manager, Section 2, Project Directorate I, Division of 
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 00-22781 Filed 9-5-00; 8:45 am]
BILLING CODE 7590-01-P