[Federal Register Volume 66, Number 122 (Monday, June 25, 2001)]
[Notices]
[Pages 33717-33718]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-15816]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[Docket No. 50-354]


PSEG Nuclear LLC; Hope Creek Generating Station Environmental 
Assesment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (NRC) is considering 
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 No. NPF-57, issued to PSEG Nuclear LLC, (the 
licensee) for operation of the Hope Creek Generating Station (HCGS), 
located in Salem County, New Jersey.

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 HCGS from application of specific 
requirements of 10 CFR Part 50, Appendix G, and would substitute use of 
ASME Code Cases N-588 and N-640 as alternatives pursuant to 10 CFR 
50.60(b).
    The proposed action is in accordance with the licensee's 
application for exemption dated December 1, 2000, as supplemented by 
letters dated February 12, May 7, and May 14, 2001.

The Need for the Proposed Action

    The proposed action is needed to allow the licensee to implement 
ASME Code Cases N-588 and N-640 in order to revise the method used to 
determine the P-T limits.
    Code Case N-588, ``Alternative to Reference Flaw Orientation of 
Appendix G for Circumferential Welds in Reactor Vessels, Section XI, 
Division 1,'' amends the provisions of the 1989 Edition of ASME Section 
XI, Appendix G, by permitting the postulation of a circumferentially 
oriented reference flaw as the limiting flaw in a RPV circumferential 
weld for the purpose of establishing RPV P-T limits. The 1989 Edition 
of ASME Section XI, Appendix G, would require that such a reference 
flaw be postulated as an axially oriented flaw in the circumferential 
weld. The licensee addressed the technical justification for this 
exemption by citing industry experience and aspects of RPV fabrication 
which support the postulation of circumferentially oriented flaws for 
these welds. The reference flaw is a postulated flaw that accounts for 
the possibility of a prior existing defect that may have gone 
undetected during the fabrication process. Postulating the Appendix G 
reference flaw in a circumferential weld

[[Page 33718]]

is physically unrealistic and overly conservative, because the length 
of the flaw is 1.5 times the vessel wall, which is much longer than the 
width of the circumferential weld. Industry experience with the repair 
of weld indications found during preservice inspection, inservice 
nondestructive examinations, and data taken from destructive 
examination of actual vessel welds confirms that any remaining defects 
are small, laminar in nature, and do not cross transverse to the weld 
bead. Therefore, any postulated defects introduced during the 
fabrication process, and not detected during subsequent nondestructive 
examinations, would only be expected to be oriented in the direction of 
weld fabrication. ASME Code Case N-588 also provides appropriate 
procedures for determining the stress intensity factors for use in 
developing RPV P-T limits per ASME Code, Section XI, Appendix G, 
procedures. The procedures allowed by ASME Code Case N-588 are 
conservative and provide a margin of safety in the development of RPV 
P-T operating and pressure test limits that will prevent nonductile 
fracture of the vessel.
    Code Case N-640, ``Alternative Reference Fracture Toughness for 
Development of P-T Limit Curves for ASME Section XI, Division 1,'' 
amends the provisions of ASME Section XI, Appendix G, by permitting the 
use of the K<INF>lc</INF> equation as found in Appendix A in ASME 
Section XI, in lieu of the K<INF>la</INF> equation as found in Appendix 
G in ASME Section XI. Use of the K<INF>lc</INF> equation in determining 
the lower bound fracture toughness in the development of the P-T 
operating limits curve is more technically correct than the use of the 
K<INF>la</INF> equation since the rate of loading during a heatup or 
cooldown is slow and is more representative of a static condition than 
a dynamic condition. Use of K<INF>la</INF> was justified by the initial 
conservatism of the K<INF>la</INF> equation since 1974 when the 
equation was codified. This initial conservatism was necessary due to 
the limited knowledge of RPV materials. Since 1974, additional 
knowledge has been gained about RPV materials, which demonstrates that 
the lower bound on fracture toughness provided by the K<INF>la</INF> 
equation is well beyond the margin of safety required to protect the 
public health and safety from potential RPV failure. The lower bound 
K<INF>lc</INF> fracture toughness provides an adequate margin of safety 
to protect the public health and safety from potential RPV failure.
    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 Cases N-588 and 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 
alternatives as described above are consistent with the intent of the 
applicable regulations and would provide an acceptable margin of safety 
against brittle failure of the HCGS RPV. Therefore, the proposed action 
will not have a significant impact on the environment.
    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 Statement for the 
HCGS.

Agencies and Persons Consulted

    In accordance with its stated policy, on June 7, 2001, the staff 
consulted with the New Jersey State official, Mr. Dennis Zannoni, 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 December 1, 2000, as supplemented by letters 
dated February 12, May 7, and May 14, 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 the Agencywide Documents Access and Management Systems (ADAMS) 
Public Electronic Reading Room on the Internet at the NRC web site, 
http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to 
ADAMS or if there are problems in accessing the documents located in 
ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-
800-397-4209, 301-415-4737 or by e-mail to <a href="/cdn-cgi/l/email-protection#7a0a1e083a140819541d150c"><span class="__cf_email__" data-cfemail="b0c0d4c2f0dec2d39ed7dfc6">[email protected]</span></a>.

    Dated at Rockville, Maryland, this 19th day of June 2001.

    For the Nuclear Regulatory Commission.
Richard B. Ennis,
Project Manager, Section 2, Project Directorate I, Division of 
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-15816 Filed 6-22-01; 8:45 am]
BILLING CODE 7590-01-P


</pre><script data-cfasync="false" src="/cdn-cgi/scripts/5c5dd728/cloudflare-static/email-decode.min.js"></script></body>
</html>