[Federal Register Volume 61, Number 138 (Wednesday, July 17, 1996)]
[Notices]
[Pages 37294-37295]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-18137]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. STN 50-454 and STN 50-455]


Commonwealth Edison Company; Byron Station, Units 1 and 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 Nos. NPF-37 and NPF-66, 
issued to Commonwealth Edison Company (ComEd, the licensee), for 
operation of Byron Station, Units 1 and 2, located in Ogle County, 
Illinois.

Environmental Assessment

Identification of the Proposed Action

    The proposed action would allow the licensee to utilize the 
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 March 14, 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 Byron Station, Units 1 
and 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 Power 
Operated Relief Valves (PORVs) 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

[[Page 37295]]

pressure must be maintained below the PORV 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 PORVs 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 PORV setpoint would 
result in an operating window between the LTOP setpoint and the minimum 
pressure required for reactor coolant pump seals which is significantly 
restricted when physical conditions such as PORV overshoot, RCP pump 
Ps, and static head corrections are taken into account in 
setpoint determination. Operating with these limits could result in the 
lifting of the PORVs 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. The 
net effect of using Code Case N-514 is that the setpoint will not 
change significantly with the next setpoint analysis. Therefore, the 
licensee proposed that in determining the PORV setpoint for LTOP events 
for Byron, 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 March 14, 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 lower bound of static, dynamic, and crack arrest 
fracture toughness tests on material similar to the Byron reactor 
vessel material.
    In determining the PORV 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 the 
Byron Station, Units 1 and 2.

Agencies and Persons Consulted

    In accordance with its stated policy, on June 19, 1996, the staff 
consulted with the Illinois State official, Mr. Frank Niziolek; Head, 
Reactor Safety Section; Division of Engineering; Illinois Department of 
Nuclear Safety; 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 March 14, 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 Byron Public Library District 109 N. Franklin, P. O. Box 434, 
Byron, Illinois 61010.

    Dated at Rockville, Maryland, this 11th day of July 1996.

    For the Nuclear Regulatory Commission.
George F. Dick, Jr.,
Project Manager, Project Directorate III-2, Division of Reactor 
Project--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 96-18137 Filed 7-16-96; 8:45 am]
BILLING CODE 7590-01-P