[Federal Register Volume 61, Number 140 (Friday, July 19, 1996)]
[Notices]
[Pages 37774-37775]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-18373]


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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 (the Commission) is 
considering issuance of an exemption from certain requirements of its 
regulations to Facility Operating License No. NPF-6, issued to Entergy 
Operations, Inc. (the licensee), for operation of Arkansas Nuclear One, 
Unit 2, located in Pope County, Arkansas.

Environmental Assessment

Identification of the Proposed Action

    The proposed action would allow the licensee to utilize American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code 
(Code) Case N-514, ``Low Temperature Overpressure Protection'' to 
determine its low temperature overpressure protection (LTOP) setpoints 
and is in accordance with the licensee's application for exemption 
dated April 11, 1996. The proposed action requests an exemption from 
certain requirements of 10 CFR 50.60, ``Acceptance Criteria for 
Fracture Prevention Measures for Lightwater Nuclear Power Reactors for 
Normal Operation,'' to allow application of an alternate methodology to 
determine the LTOP setpoints for ANO-2. The proposed alternate 
methodology is consistent with guidelines developed by the ASME Working 
Group on Operating Plant Criteria (WGOPC) to define pressure limits 
during LTOP events that avoid certain unnecessary operational 
restrictions, provide adequate margins against failure of the reactor 
pressure vessel, and reduce the potential for unnecessary activation of 
pressure relieving devices used for LTOP. These guidelines have been 
incorporated into Code Case N-514, ``Low Temperature Overpressure 
Protection,'' which has been approved by the ASME Code Committee. The 
content of this Code Case has been incorporated into Appendix G of 
Section XI of the ASME Code and published in the 1993 Addenda to 
Section XI. However, 10 CFR 50.55a, ``Codes and Standards,'' and 
Regulatory Guide 1.147, ``Inservice Inspection Code Case 
Acceptability'' have not been updated to reflect the acceptability of 
Code Case N-514.
    The philosophy used to develop Code Case N-514 guidelines is to 
ensure that the LTOP limits are still below the pressure/temperature 
(P/T) limits for normal operation, but allow the pressure that may 
occur with activation of pressure relieving devices to exceed the P/T 
limits, provided acceptable margins are maintained during these events. 
This philosophy protects the pressure vessel from LTOP events, and 
still maintains the Technical Specifications P/T limits applicable for 
normal heatup and cooldown in accordance with 10 CFR Part 50, Appendix 
G and Sections III and XI of the ASME Code.

The Need for the Proposed Action

    Pursuant to 10 CFR 50.60, all lightwater nuclear power reactors 
must meet the fracture toughness requirements for the reactor coolant 
pressure boundary as set forth in 10 CFR Part 50, Appendix G. 10 CFR 
Part 50, Appendix G, defines P/T limits during any condition of normal 
operation including anticipated operational occurrences and system 
hydrostatic tests, to which the pressure boundary may be subjected over 
its service lifetime. It is specified in 10 CFR 50.60(b) that 
alternatives to the described requirements in 10 CFR Part 50, Appendix 
G, may be used when an exemption is granted by the Commission under 10 
CFR 50.12.
    To prevent transients that would produce excursions exceeding the 
10 CFR Part 50, Appendix G, P/T limits while the reactor is operating 
at low temperatures, the licensee installed an LTOP system. The LTOP 
system includes pressure relieving devices in the form of relief valves 
that are set at a pressure below the LTOP enabling temperature that 
would prevent the pressure in the reactor vessel from exceeding the P/T 
limits of 10 CFR Part 50, Appendix G. To prevent these valves from 
lifting as a result of normal operating pressure surges (e.g., reactor 
coolant pump starting and shifting operating charging pumps) with the 
reactor coolant system in a solid water condition, the operating 
pressure must be maintained below the relief valve setpoint.
    In addition, to prevent damage to reactor coolant pump seals, the 
operator must maintain a minimum differential pressure across the 
reactor coolant pump seals. Hence, the licensee must operate the plant 
in a pressure window that is defined as the difference between the 
minimum required pressure to start a reactor coolant pump and the 
operating margin to prevent lifting of the relief valves due to normal 
operating pressure surges. The 10 CFR Part 50, Appendix G, safety 
margin adds instrument uncertainty into the LTOP setpoint. The 
licensee's current LTOP analysis indicates that using this 10 CFR Part 
50, Appendix G, safety margin to determine the relief valve setpoint 
would result in an operating window between the LTOP setpoint and the 
minimum pressure required for reactor coolant pump seals which is too 
small to permit continued operation. Operating with these limits could 
result in the lifting of relief valves or damage to the reactor coolant 
pump seals during normal operation. Using Code Case N-514 would allow 
the licensee to recapture most of the operating margin that is lost by 
factoring in the instrument uncertainties in the determination of the 
LTOP setpoint. Therefore, the licensee proposed that in determining the 
relief valve setpoint for LTOP events for ANO-2, the allowable pressure 
be determined using the safety margins developed in an alternate 
methodology in lieu of the safety margins required by 10 CFR Part 50, 
Appendix G. The alternate methodology is consistent with ASME Code Case 
N-514. The content of this Code Case has been incorporated into 
Appendix G of Section XI of the ASME Code and published in the 1993 
Addenda to Section XI.
    An exemption from 10 CFR 50.60 is required to use the alternate 
methodology for calculating the maximum allowable pressure for LTOP 
considerations. By application dated April 11, 1996, the licensee 
requested an exemption from 10 CFR 50.60 to allow it to utilize the 
alternate methodology of Code Case N-514 to compute its LTOP setpoints.

Environmental Impacts of the Proposed Action

    Appendix G of the ASME Code requires that the P/T limits be 
calculated: (a) using a safety factor of two on the principal membrane 
(pressure) stresses, (b) assuming a flaw at the surface with a depth of 
one quarter (1/4) of the vessel wall thickness and a length of six (6) 
times its depth, and (c) using a conservative fracture toughness curve 
that is based on the

[[Page 37775]]

lower bound of static, dynamic, and crack arrest fracture toughness 
tests on material similar to the ANO-2 reactor vessel material.
    In determining the relief valve setpoint for LTOP events, the 
licensee proposed the use of safety margins based on an alternate 
methodology consistent with the proposed ASME Code Case N-514 
guidelines. ASME Code Case N-514 allows determination of the setpoint 
for LTOP events such that the maximum pressure in the vessel will not 
exceed 110% of the P/T limits of the existing ASME Appendix G. This 
results in a safety factor of 1.8 on the principal membrane stresses. 
All other factors, including assumed flaw size and fracture toughness, 
remain the same. Although this methodology would reduce the safety 
factor on the principal membrane stresses, use of the proposed criteria 
will provide adequate margins of safety to the reactor vessel during 
LTOP transients.
    The change will not 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 the 
allowable individual or cumulative occupational radiation exposure. 
Accordingly, the Commission concludes that there are no significant 
radiological environmental impacts associated with the proposed action.
    With regard to potential nonradiological impacts, the proposed 
action does involve features located entirely within the restricted 
area as defined in 10 CFR Part 20. It does not affect nonradiological 
plant effluents and has no other environmental impact. Accordingly, the 
Commission concludes that there are no significant nonradiological 
environmental impacts associated with the proposed action.

Alternatives to the Proposed Action

    Since the Commission has concluded there is no measurable 
environmental impact associated with the proposed action, any 
alternatives with equal or greater environmental impact need not be 
evaluated. As an alternative to the proposed action, the staff 
considered denial of the proposed action. 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 ANO-2.

Agencies and Persons Consulted

    In accordance with its stated policy, on May 13, 1996, the staff 
consulted with the Arkansas State official, Mr. Bernard Bevill Director 
of Radiation Control and Emergency Management, regarding the 
environmental impact of the proposed action. The State official had no 
comments.

Finding of No Significant Impact

    Based upon the environmental assessment, the Commission concludes 
that the proposed action will not have a significant effect on the 
quality of the human environment. Accordingly, the Commission 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 April 11, 1996, which is available for public 
inspection at the Commission's Public Document Room, 2120 L Street, 
NW., Washington, DC, and at the local public document room located at 
the Tomlinson Library, Arkansas Tech University, Russellville, AR 
72801.

    Dated at Rockville, Maryland, this 15th day of July, 1996.

    For the Nuclear Regulatory Commission.
George Kalman,
Senior Project Manager, Project Directorate VI-1, Division of Reactor 
Projects III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 96-18373 Filed 7-18-96; 8:45 am]
BILLING CODE 7590-01-P