[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
