[Federal Register Volume 66, Number 180 (Monday, September 17, 2001)]
[Notices]
[Pages 48069-48070]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-23149]


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

NUCLEAR REGULATORY COMMISSION

[50-458]


Entergy Operations, Inc.; River Bend Station; Environmental 
Assessment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (the NRC) is considering 
issuance of an exemption from 10 CFR part 50, Appendix G for Facility 
Operating License No. NPF-47, issued to Entergy Operations, Inc. (the 
licensee), for operation of the River Bend Station, Unit 1 (RBS) 
located in West Feliciana Parish, Louisiana. 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 exempt the licensee from certain 
provisions of 10 CFR part 50, Appendix G. Pursuant to 10 CFR part 50, 
Appendix G, pressure-temperature limits (P-T) are required to 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, ``***[t]he appropriate requirements on 
both the pressure-temperature limits and the minimum permissible 
temperature must be met for all conditions.'' Appendix G to 10 CFR part 
50 specifies that the requirements for these limits are the American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code 
(the Code), Section XI, Appendix G limits.
    The proposed action would substitute ASME Code Case N-640 for 
specific requirements in 10 CFR part 50, Appendix G. Code Case N-640, 
``Alternative Reference Fracture Toughness for Development of P-T Limit 
Curves Section XI, Division 1,'' permits the use of an alternative 
reference fracture toughness (KIc fracture toughness curve 
instead of the KIa fracture toughness curve) for RPV 
materials in determining the P-T limits. Since the KIc 
fracture toughness curve shown in ASME Code Section XI, Appendix A, 
Figure A-4200-1 provides greater allowable fracture toughness than the 
corresponding KIa fracture toughness curve of ASME Code 
Section XI, Appendix G, Figure G-2210-1, using the KIc 
fracture toughness, as permitted by Code Case N-640, in establishing 
the P-T limits would be less conservative than the methodology 
currently endorsed by 10 CFR part 50, Appendix G. Considering this, an 
exemption to apply the Code Case would be required by 10 CFR 50.60. 
Accordingly, the licensee requested an exemption from the requirements 
in 10 CFR part 50, Appendix G.
    Use of the KIc curve in determining the lower bound 
fracture toughness in the development of P-T operating limits is more 
technically correct than the KIa curve, since the rate of 
loading during a heatup or cooldown is slow and is more representative 
of a static condition than a dynamic condition. The KIc 
curve appropriately implements the use of static initiation fracture 
toughness behavior to evaluate the controlled heatup and cooldown 
process relative to an RPV. The ASME Code Section XI, Appendix G, 
procedure was conservatively developed based on the level of knowledge 
existing in 1974 concerning RPV materials and the estimated effects of 
operation. Since 1974, the level of knowledge about these topics has 
been greatly expanded. The NRC staff concludes that this increased 
knowledge permits relaxation of the ASME Code Section XI, Appendix G 
requirements by applying KIc fracture toughness, as 
permitted by Code Case N-640, while maintaining, pursuant to 10 CFR 
50.12(a)(2)(ii), the underlying purpose of the ASME Code and the NRC 
regulations to ensure an acceptable margin of safety.
    The proposed action is in accordance with the licensee's 
application for amendment and exemption dated January 24, 2001, as 
supplemented by letters dated July 2, and August 6 and 20, 2001, and is 
needed to support the technical specification (TS) amendment that is 
contained in the same submittal and is being processed separately. The 
proposed TS amendment will revise the P-T limits of TS 3.4.11, RCS 
[Reactor Coolant System] Pressure and Temperature Limits,'' related to 
the heatup, cooldown, and inservice test limitations for the RCS to a 
maximum of 16 Effective Full Power Years (EFPY). The proposed action 
replaces TS Figure 3.4-11, ``Minimum Temperature Required Vs. RCS 
Pressure,'' with recalculated RCS P-T limits based, in part, on the 
alternative methodology in Code Case N-640.

The Need for the Proposed Action

    The revised P-T limits are needed to allow required reactor vessel 
hydrostatic and leak tests to be performed at a significantly lower 
temperature. These tests are to be performed during the upcoming 
refueling outage scheduled to commence in September 2001. The lower 
temperature for the tests can reduce refueling outage critical path 
time by reducing or eliminating the heatup time to achieve required 
test conditions.

Environmental Impacts of the Proposed Action

    The NRC has completed its evaluation of the proposed action and 
concludes that the exemption and associated license amendment described 
above would provide an adequate margin of safety against brittle 
failure of the RBS reactor vessel. The lower temperature, is also safer 
for test inspectors due to lower ambient drywell temperatures and could 
result in lower radiological dose due to increased inspection 
effectiveness at the lower temperature.
    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 off site, 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 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 staff considered 
denial of the

[[Page 48070]]

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 different resource than 
those previously considered in the ``Final Environmental Statement,'' 
NUREG-1073, January 1985, for the RBS.

Agencies and Persons Consulted

    On August 13, 2001, the staff consulted with the Louisiana State 
official, Ms. Soumaya Ghosn of the Louisiana Department of 
Environmental Quality, Radiation Protection Division, 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 January 24, 2001, as supplemented by letters 
dated July 2, and August 6 and 20, 2001. Documents may be examined, 
and/or copied for a fee, at the NRC's Public Document Room (PDR), 
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 Library component on the NRC web site, http://www.nrc.gov (the Public Electronic Reading Room). If you do not have 
access to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-
4209, or 301-415-4737, or by e-mail to [email protected].

    Dated at Rockville, Maryland, this 7th day of September, 2001.

    For the Nuclear Regulatory Commission,
Robert E. Moody,
Project Manager, Section 1, Project Directorate IV, Division of 
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-23149 Filed 9-14-01; 8:45 am]
BILLING CODE 7590-01-P