[Federal Register Volume 66, Number 93 (Monday, May 14, 2001)]
[Notices]
[Pages 24410-24412]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-12036]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-272 and 50-311]


PSEG Nuclear LLC, Salem Nuclear Generating Station, Unit Nos. 1 
and 2; Environmental Assessment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (NRC) is considering

[[Page 24411]]

issuance of an exemption from certain requirements of Title 10 of the 
Code of Federal Regulations (10 CFR) Part 50, Appendix G, for Facility 
Operating License Nos. DPR-70 and DPR-75, issued to PSEG Nuclear LLC 
(the licensee), for operation of the Salem Nuclear Generating Station 
(Salem), Unit Nos. 1 and 2. The facility is located at the licensee's 
site on the southern end of Artificial Island in Lower Alloways Creek 
Township, Salem County, New Jersey. Salem, New Jersey is located 
approximately 7.5 miles northeast of the site.

Environmental Assessment

Identification of the Proposed Action

    Title 10 of the Code of Federal Regulations, 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.'' The purpose of 10 CFR Part 50, Appendix G, is to 
protect the integrity of the reactor coolant pressure boundary in 
nuclear power plants. This is accomplished through these regulations 
that, in part, specify fracture toughness requirements for ferritic 
materials of the reactor coolant pressure boundary. Appendix G of 10 
CFR Part 50 specifies that the requirements for these limits are the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code (Code), Section XI, Appendix G limits.
    The proposed action would exempt the licensee from implementing 
specific requirements of 10 CFR Part 50, Appendix G, for operation of 
Salem, Unit Nos. 1 and 2. In conjunction with the staff granting the 
proposed exemption to the requirements of 10 CFR Part 50, Appendix G, 
the licensee is proposing to substitute ASME Code Case N-640, 
``Alternative Reference Fracture Toughness for Development of P/T Limit 
Curves for ASME Section XI, Division I.''
    The proposed action is in accordance with the licensee's 
application for exemption dated November 10, 2000, as supplemented by 
letters dated March 28 and April 2, 2001.

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. The methodology currently used to determine the lower bound 
fracture toughness of RPV material for development of P-T limit curves 
is based on the KIa fracture toughness curve of ASME Section 
XI, Appendix G, Figure G-2210-1. The licensee has determined that the 
use of the KIa curve provided appropriate conservatism when 
it was codified in 1974 due to the limited knowledge of RPV materials. 
However, since that time, additional knowledge has been gained about 
RPV materials, that demonstrates that the lower bound on fracture 
toughness provided by the KIa curve is well beyond the 
margin of safety required to protect the public health and safety from 
potential RPV failure. Implementation of ASME Code Case N-640 would 
provide an alternative to the methodology used to develop P-T limit 
curves. The code case methodology uses the KIc fracture 
toughness curve shown in ASME Section XI, Appendix A, Figure A-2200-1, 
in lieu of the KIa fracture toughness curve of ASME Section 
XI, Appendix G. Other margins involved with the ASME Section XI, 
Appendix G process for establishing P-T limit curves would remain 
unchanged. P-T curves based on the KIc curve would enhance 
overall plant safety by opening the P-T operating window with the 
greatest safety benefit in the region of low temperature operations. 
The operating window through which the operator heats up and cools down 
the RCS is determined by the difference between the maximum allowable 
pressure determined by ASME Section XI, Appendix G, and the minimum 
required pressure for the reactor coolant pump (RCP) seals adjusted for 
instrument uncertainties.
    The staff has determined that, pursuant to 10 CFR 50.12(a)(2)(ii), 
the underlying purpose of the regulation to protect the integrity of 
the reactor coolant pressure boundary will continue to be served with 
the implementation of Code Case N-640.

Environmental Impacts of the Proposed Action

    The NRC has completed its evaluation of the proposed action and 
concludes that the exemption and implementation of the proposed 
alternative as described is consistent with the intent of the 
applicable regulations and would provide an acceptable margin of safety 
against brittle failure of the Salem, Unit Nos. 1 and 2 RPV material.
    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 impact. 
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 Statement for the 
Salem Nuclear Generating Station, dated April 1973.

Agencies and Persons Consulted

    In accordance with its stated policy, on May 1, 2001, the staff 
consulted with the New Jersey State official, Mr. R. Pinney of the New 
Jersey 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 November 10, 2000, as supplemented by letters 
dated March 28 and April 2, 2001. Documents may be examined, and/or 
copied for a fee, at the NRC's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible electronically 
from

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the ADAMS Public Library component on the NRC Web site, 
http:\\www.nrc.gov (the Electronic Reading Room).

    Dated at Rockville, Maryland, this 8th day of May 2001.

    For the Nuclear Regulatory Commission.
Robert J. Fretz,
Project Manager, Section 2, Project Directorate I, Division of 
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-12036 Filed 5-11-01; 8:45 am]
BILLING CODE 7590-01-P