[Federal Register Volume 67, Number 67 (Monday, April 8, 2002)]
[Notices]
[Pages 16769-16770]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-8387]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-368]


Entergy Operations, Inc.; Arkansas Nuclear One, Unit 2 
Environmental Assessment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (NRC) is considering 
issuance of an exemption from Title 10 of the Code of Federal 
Regulations (10 CFR) part 50.60 and 10 CFR part 50, Appendix G, for 
Facility Operating License No. NPF-6, issued to Entergy Operations, 
Inc. (the licensee), for operation of the Arkansas Nuclear One, Unit 2 
(ANO-2), nuclear power plant, located in Pope County, Arkansas. 
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 allow a one-time exemption from 10 CFR 
part 50, Appendix G requirements that pressure-temperature (P-T) limits 
be established for reactor pressure vessels (RPVs) during normal 
operating and hydrostatic or leak testing conditions. Specifically, 10 
CFR part 50, Appendix G, states that ``[t]he 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 contained in 
the American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code (Code), Section XI, Appendix G.
    To address provisions of an amendment to the Technical 
Specifications (TSs) P-T limits and low-temperature overpressure 
protection (LTOP) system TS restrictions, the licensee requested in its 
submittal dated October 30, 2001, as supplemented by letters dated 
February 25 and March 13, 2002, that the NRC staff exempt the ANO-2 
nuclear power plant from the requirements of 10 CFR part 50, Appendix 
G. The exemption requested would allow the use of ASME Code Case N-641 
in establishing the reactor vessel pressure limits at low temperatures.
    Code Case N-641 permits the use of an alternate reference fracture 
toughness (KIC fracture toughness curve instead of the 
KIA fracture toughness curve) for reactor vessel materials 
in determining the P-T limits, LTOP system setpoints, and LTOP system 
effective temperature (also known as the LTOP system enable 
temperature, Tenable), and provides for plant-specific 
evaluation of Tenable. Since the KIC fracture 
toughness curve shown in ASME 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 Section XI, Appendix G, Figure G-2210-
1 (the KIA fracture toughness curve), and a plant-specific 
evaluation of Tenable would give lower values of 
Tenable than use of a generic bounding evaluation for 
Tenable, use of Code Case N-641 for establishing the P-T 
limits, LTOP system setpoints, and Tenable would be less 
conservative than the methodology currently endorsed by 10 CFR part 50, 
Appendix G. Although the use of the KIC fracture toughness 
curve in ASME Code Case N-641 was recently incorporated into Appendix G 
to Section XI of the ASME Code, an exemption is still needed because 10 
CFR part 50, Appendix G requires a licensee's analysis to use an 
edition and addenda of Section XI of the ASME Code incorporated by 
reference into 10 CFR part 50, Section 50.55a, i.e., the editions 
through 1995 and addenda through the 1996 addenda (which do not include 
the provisions of Code Case N-641). Therefore, an exemption to apply 
the Code case is required by 10 CFR part 50, Section 50.60.
    The proposed action is in accordance with the licensee's 
application for exemption dated October 30, 2001, as supplemented by 
letters dated February 25 and March 13, 2002.

The Need for the Proposed Action

    ASME Code Case N-641 is needed to revise the method used to 
determine the reactor coolant system (RCS) P-T limits, LTOP setpoints, 
and Tenable.
    The purpose of 10 CFR part 50, Section 50.60(a), and 10 CFR part 
50, Appendix G, is to protect the integrity of the reactor coolant 
pressure boundary (RCPB) in nuclear power plants. This is accomplished 
through these regulations that, in part, specify fracture toughness 
requirements for ferritic materials of the RCPB. Pursuant to 10 CFR 
part 50, Appendix G, it is required that P-T limits for the RCS be at 
least as conservative as those obtained by applying the methodology of 
the ASME Code, Section XI, Appendix G.
    Current overpressure protection system (OPPS) setpoints produce 
operational constraints by limiting the P-T range available to the 
operator to heat up or cool down the plant. The operating window 
through which the operator heats up and cools down the RCS becomes more 
restrictive with continued reactor vessel service. Reducing this 
operating window could potentially have an adverse safety impact by 
increasing the possibility of inadvertent OPPS actuation due to 
pressure surges associated with normal plant evolutions, such as 
reactor coolant pump start and swapping operating charging pumps with 
the RCS in a water-solid condition. The impact on the P-T limits and 
OPPS setpoints has been evaluated for an increased service period for 
operation to 32 effective full-power years for ANO-2, based on ASME 
Code, Section XI, Appendix G requirements. The results indicate that 
these OPPS setpoints would significantly restrict the ability to 
perform plant heatup and cooldown, create an unnecessary burden to 
plant operations, and challenge control of plant evolutions required 
with OPPS enabled. Continued operation of ANO-2 with P-T curves 
developed to satisfy ASME Code, Section XI, Appendix G, requirements 
without the relief provided by ASME Code Case N-641 would unnecessarily 
restrict the P-T operating window, especially at low temperature 
conditions.
    Use of the KIC curve in determining the lower bound 
fracture toughness of RPV steels is more technically correct than use 
of 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 of a reactor 
vessel. The staff has required use of the conservatism of the 
KIA curve since 1974, when the curve was adopted by the ASME 
Code. This conservatism was initially necessary due to the limited 
knowledge of the fracture toughness of RPV materials at that time. 
Since 1974, additional

[[Page 16770]]

knowledge has been gained about RPV materials, which demonstrates that 
the lower bound on fracture toughness provided by the KIA 
curve greatly exceeds the margin of safety required, and that the 
KIC curve is sufficiently conservative to protect the public 
health and safety from potential RPV failure. Application of ASME Code 
Case N-641 will provide results that are sufficiently conservative to 
ensure the integrity of the RCPB, while providing P-T curves that are 
not overly restrictive. Implementation of the proposed P-T curves, as 
allowed by ASME Code Case N-641, does not significantly reduce the 
margin of safety.
    In the associated exemption, the NRC staff has determined that, 
pursuant to 10 CFR part 50, section 50.12(a)(2)(ii), the underlying 
purpose of the regulation will continue to be served by the 
implementation of ASME Code Case N-641.

Environmental Impacts of the Proposed ACTION:

    The NRC has completed its evaluation of the proposed action and 
concludes, as set forth below, that there are no significant 
environmental impacts associated with the use of the alternative 
analysis method to support the revision of the RCS P-T limits, LTOP 
setpoints and proposed Tenable.
    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 nonradiological 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 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 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:

    The action does not involve the use of any different resource than 
those previously considered in the Final Environmental Statement for 
the ANO-2 nuclear power plant, dated June 1977 (NUREG-0254).

Agencies and Persons Consulted:

    On March 18, 2002, the staff consulted with the Arkansas State 
official, Mr. B. Bevill of the Division of Radiation Control and 
Emergency Management of the Arkansas Department of Health, 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 October 30, 2001, as supplemented by letters 
dated February 25 and March 13, 2002. 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 System (ADAMS) 
Public Electronic Reading Room on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff by 
telephone at 1-800-397-4209 or 301-415-4737, or by e-mail to 
[email protected].

    Dated at Rockville, Maryland, this 2nd day of April 2002.

    For the Nuclear Regulatory Commission.
Robert A. Gramm,
Chief, Section I, Project Directorate IV, Division of Licensing Project 
Management, Office of Nuclear Reactor Regulation.
[FR Doc. 02-8387 Filed 4-5-02; 8:45 am]
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