[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