[Federal Register Volume 66, Number 85 (Wednesday, May 2, 2001)]
[Notices]
[Pages 22018-22019]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-10965]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-338 and 50-339]


Virginia Electric and Power Company, North Anna Power Station, 
Units 1 and 2; Environmental Assessment and Finding of No Significant 
Impact

    The U.S. Nuclear Regulatory Commission (NRC) is considering 
issuance of an exemption from the requirements of Title 10 of the Code 
of Federal Regulations (10 CFR) part 50, appendix G, for Facility 
Operating License Nos. NPF-4 and NPF-7, issued to Virginia Electric and 
Power Company (the licensee), for operation of the North Anna Power 
Station, Units 1 and 2, located in Louisa County, Virginia.

Environmental Assessment

Identification of the Proposed Action

    10 CFR Part 50, Appendix G, requires that the 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 P-T limits and low-temperature overpressure protection 
(LTOP) system setpoints, the licensee requested in its submittal dated 
June 22, 2000, as supplemented on January 4, February 14, March 13, and 
March 22, 2001, that the NRC staff exempt North Anna Power Station from 
the requirements of 10 CFR Part 50, Appendix G, to allow the use of 
ASME Code Case N-641.
    Code Case N-641 permits the use of an alternate reference fracture 
toughness (K<INF>IC</INF> fracture toughness curve instead of the 
K<INF>IA</INF> fracture toughness curve) for reactor vessel materials 
in determining the P-T limits, LTOP system setpoints and 
T<INF>enable</INF>, and provides for plant-specific evaluation of 
T<INF>enable</INF>. Since the K<INF>IC</INF> fracture toughness curve 
shown in ASME Section XI, Appendix A, Figure A-2200-1 (the 
K<INF>IC</INF> fracture toughness curve) provides greater allowable 
fracture toughness than the corresponding K<INF>Ia</INF> fracture 
toughness curve of ASME Section XI, Appendix G, Figure G-2210-1 (the 
K<INF>Ia</INF> fracture toughness curve), and a plant-specific 
evaluation of T<INF>enable</INF> would give lower values of 
T<INF>enable</INF> than use of a generic bounding evaluation for 
T<INF>enable</INF>, use of Code Case N-641 for establishing the P-T 
limits, LTOP system setpoints and T<INF>enable</INF> would be less 
conservative than the methodology currently endorsed by 10 CFR Part 50, 
Appendix G. Although the use of the K<INF>IC</INF> 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 June 22, 2000, as supplemented by letters dated January 
4, February 14, March 13, and March 22, 2001.

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 T<INF>enable</INF>.
    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 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. 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 32.3 effective full-power years (EFPYs) for Unit 1 and 
34.3 EFPYs for Unit 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,

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create an unnecessary burden to plant operations, and challenge control 
of plant evolutions required with OPPS enabled. Continued operation of 
North Anna Units 1 and 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 K<INF>Ic</INF> curve in determining the lower bound 
fracture toughness of RPV steels is more technically correct than use 
of the K<INF>Ia</INF> 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 K<INF>Ic</INF> 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 
K<INF>Ia</INF> 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 knowledge has been gained about RPV materials, 
which demonstrates that the lower bound on fracture toughness provided 
by the K<INF>Ia</INF> curve greatly exceeds the margin of safety 
required, and that the K<INF>IC</INF> 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 
reactor coolant pressure boundary 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 that the proposed action provides adequate margin of safety 
against brittle failure of the reactor coolant pressure boundary. 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.

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 
North Anna Power Station, Units 1 and 2, dated April 1973.

Agencies and Persons Consulted

    In accordance with its stated policy, on April 2, 2001, the staff 
consulted with the Virginia State official, Mr. J. Dekrafft of the 
Radiological Health Program of the Virginia Department of Health, 
regarding the environmental impact of the proposed action. The State 
official had no comments.

Finding of No Signficant 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 June 22, 2000, as supplemented by letters dated 
January 4, February 14, March 13, and March 22, 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 ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room).

    Dated at Rockville, Maryland, this 26th day of April 2001.

    For the Nuclear Regulatory Commission.
Gordon E. Edison,
Senior Project Manager, Section 1, Project Directorate II, Division of 
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-10965 Filed 5-1-01; 8:45 am]
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