[Federal Register Volume 61, Number 238 (Tuesday, December 10, 1996)]
[Notices]
[Pages 65084-65085]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-31324]


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NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-454, STN 50-455]


Commonwealth Edison Company (Byron Station, Units 1 and 2); 
Exemption

I.

    Commonwealth Edison Company (ComEd, the licensee) is the holder of 
Facility Operating License Nos. NPF-37 and NPF-66, which authorize 
operation of Byron Station, Units 1 and 2, respectively. The licenses 
provide, among other things, that the licensee is subject to all rules, 
regulations, and orders of the Commission now or hereafter in effect.
    The facility consists of two pressurized-water reactors located at 
the licensee's site in Ogle County, Illinois.

II.

    In its letter dated March 14, 1996, the licensee requested an 
exemption from the Commission's regulations. Title 10 of the Code of 
Federal Regulations, Part 50, Section 60 (10 CFR 50.60), ``Acceptance 
Criteria for Fracture Prevention Measures for Lightwater Nuclear Power 
Reactors for Normal Operation,'' states that all lightwater nuclear 
power reactors must meet the fracture toughness and material 
surveillance program requirements for the reactor coolant pressure 
boundary as set forth in Appendices G and H to 10 CFR Part 50. Appendix 
G to 10 CFR Part 50 defines pressure/temperature (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 Appendices 
G and H to 10 CFR Part 50 may be used when an exemption is granted by 
the Commission under 10 CFR 50.12.
    To prevent low-temperature overpressure transients that would 
produce pressure excursions exceeding the P/T limits of Appendix G to 
10 CFR Part 50 while the reactor is operating at low temperatures, the 
licensee installed a low-temperature overpressure protection (LTOP) 
system. The system includes pressure-relieving devices called power-
operated relief valves (PORVs). The PORVs are set at a pressure low 
enough so that if an LTOP transient occurred, the mitigation system 
would prevent the pressure in the reactor vessel from exceeding the P/T 
limits of Appendix G to 10 CFR Part 50. To prevent the PORVs from 
lifting as a result of normal operating pressure surges (e.g., starting 
reactor coolant pumps, and shifting operating charging pumps) with the 
reactor coolant system in a solid water condition, the operating 
pressure must be maintained below the PORV setpoint. Applying LTOP 
instrument uncertainties as required by WCAP-14040, Revision 1, results 
in an LTOP setpoint that would have resulted 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.
    The licensee has requested the use of the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Case 
N-514, ``Low Temperature Overpressure Protection,'' which allows 
exceeding the safety limits of 10 CFR Part 50, Appendix G, by 10 
percent. ASME Code Case N-514, the proposed alternate methodology, is 
consistent with guidelines developed by the ASME Working Group on 
Operating Plant Criteria 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. ASME Code Case N-514 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.

III.

    Pursuant to 10 CFR 50.12, the Commission may, upon application by 
any interested person or upon its own initiative, grant exemptions from 
the requirements of 10 CFR Part 50 when (1) the exemptions are 
authorized by law, will not present an undue risk to public health or 
safety, and are consistent with the common defense and security, and 
(2) when special circumstances are present. Special circumstances are 
present whenever, according to 10 CFR 50.12(a)(2)(ii), ``Application of 
the regulation in the particular circumstances would not serve the 
underlying purpose of the rule or is not necessary to achieve the 
underlying purpose of the rule.''
    The underlying purpose of 10 CFR Part 50, Appendix G, is to 
establish fracture toughness requirements for ferritic materials of 
pressure-retaining components of the reactor coolant pressure boundary 
to provide adequate margins of safety during any condition of normal 
operation, including anticipated operational occurrences, to which the 
pressure boundary may be subjected over its service lifetime. Section 
IV.A.2 of this appendix requires that the reactor vessel be operated 
with P/T limits at least as conservative as those obtained by following 
the methods of analysis and the required margins of safety of Appendix 
G of the ASME Code.
    Appendix G of Section XI 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 setpoint for LTOP events, the licensee proposed 
to use safety margins based on an alternate methodology consistent with 
the ASME Code Case N-514 guidelines. The ASME Code Case N-514 allows 
determination of the setpoint for LTOP events such that the maximum 
pressure in the vessel would not exceed 110 percent of the P/T limits 
of the existing ASME Code, Section XI, Appendix G. This approach 
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, the proposed criteria will 
provide adequate margins of safety to the reactor vessel during LTOP 
transients and, thus, will satisfy the underlying purpose of 10 CFR 
50.60 for fracture toughness requirements. Further, by relieving the 
operational restrictions, the potential for undesirable lifting of the 
PORV would

[[Page 65085]]

be reduced, thereby improving plant safety.

IV.

    For the foregoing reasons, the NRC staff has concluded that the 
licensee's proposed use of the alternate methodology in determining the 
acceptable setpoint for LTOP events will not present an undue risk to 
public health and safety and is consistent with the common defense and 
security. The NRC staff has determined that there are special 
circumstances present, as specified in 10 CFR 50.12(a)(2), in that 
application of 10 CFR 50.60 is not necessary in order to achieve the 
underlying purpose of this regulation.
    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), an exemption is authorized by law, will not endanger life or 
property or common defense and security, and is otherwise in the public 
interest. Therefore, the Commission hereby grants an exemption from the 
requirements of 10 CFR 50.60 such that in determining the setpoint for 
LTOP events, the Appendix G curves for P/T limits are not exceeded by 
more than 10 percent in order to be in compliance with these 
regulations. This exemption is applicable only to LTOP conditions 
during normal operation.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of this exemption will not have a significant effect on the 
quality of the human environment (61 FR 37294).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 29th day of Nov. 1996.

    For the Nuclear Regulatory Commission.
Frank J. Miraglia,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 96-31324 Filed 12-9-96; 8:45 am]
BILLING CODE 7590-01-P