[Federal Register Volume 62, Number 249 (Tuesday, December 30, 1997)]
[Notices]
[Pages 67922-67923]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-33846]
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NUCLEAR REGULATORY COMMISSION
[Docket No. STN 50-457]
Commonwealth Edison Company (Braidwood Nuclear Station, Unit No.
2); Exemption
I.
Commonwealth Edison Company (ComEd, the licensee) is the holder of
Facility Operating License No. NPF-77, which authorizes operation of
the Braidwood Nuclear Station, Unit 2. The license provides, among
other things, that the licensee is subject to all rules, regulations,
and orders of the Commission now or hereafter in effect.
II.
In its letter dated November 30, 1994, as supplemented on May 11,
1995, 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
comply with the fracture toughness and material surveillance program
requirements for the reactor coolant pressure boundary as stated 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
[[Page 67923]]
service lifetime, and that are obtained by conforming to the methods of
analysis and the margins of safety in the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),
Section XI, Appendix G. It is required in 10 CFR 50.55a that any
reference to the ASME Code, Section XI, in 10 CFR Part 50 refers to the
addenda through the 1988 Addenda and editions through the 1989 Edition
of the Code unless otherwise noted. It is specified in 10 CFR 50.60(b)
that alternatives to the described requirements in Appendix G to 10 CFR
Part 50 may be used when an exemption is granted by the Commission
under 10 CFR 50.12.
To mitigate low-temperature overpressure transients that would
produce pressure excursions exceeding the required limits while the
reactor is operating at low temperatures, the licensee installed a low-
temperature overpressure protection (LTOP) system. The system contains
pressure-relieving devices called power-operated relief valves (PORVs).
The PORVs are set at a low enough pressure so that if an LTOP transient
occurred, the mitigation system would prevent the pressure in the
reactor vessel from exceeding the required limits. To prevent the PORVs
from lifting as a result of normal operating pressure surges, some
margin is needed between the PORV setpoint and the normal operating
pressure. In addition, normal operating pressure must be high enough to
prevent damage to reactor coolant pumps that may result from cavitation
or inadequate differential pressure across the pump seals. Hence, the
licensee must operate the plant within a pressure window that is
defined as the difference between the minimum pressure required for
reactor coolant pumps and the operating margin to keep the PORVs from
lifting. When instrument uncertainty is considered, the operating
window is small and presents difficulties for plant operation.
The licensee has requested the use of the ASME Code Case N-514,
``Low Temperature Overpressure Protection,'' for determining the LTOP
system setpoint. Code Case N-514 allows use of an LTOP system setpoint
so that system pressure does not exceed 110 percent of the P-T limits
during an LTOP event. Code Case N-514 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. The
content of this code case has been incorporated into the ASME Code,
Section XI, Appendix G, and was 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 (1) when 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 50.60 and 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 Appendix G to 10 CFR Part 50
requires that the reactor vessel be operated with P-T limits at least
as conservative as those obtained from following the methods of
analysis and the required margins of safety of Appendix G of Section XI
of the ASME Code.
Appendix G of the ASME Code requires that the P-T limits be
calculated (1) Using a safety factor of 2 on the principal membrane
(pressure) stresses, (2) assuming a flaw at the surface with a depth of
one-fourth of the vessel wall thickness and a length of 6 times its
depth, and (3) 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 Braidwood reactor vessel
material.
For determining the LTOP system setpoint, the licensee proposed to
use safety margins based on an alternate methodology consistent with
ASME Code Case N-514. The code case allows the setpoint for mitigating
LTOP events to be so determined that the maximum pressure in the vessel
would not exceed 110 percent of the Appendix G P-T limits. 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, the proposed criteria will produce
adequate margins of safety for 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
PORVs would be reduced, thereby making the plant safer.
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)(ii), in that
application of 10 CFR 50.60 is not necessary in order to achieve the
underlying purpose of this regulation which is to provide adequate
fracture toughness of the reactor pressure boundary.
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 the 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; in accordance with ASME Code
Case N-514, the LTOP system setpoint may be determined so that system
pressure does not exceed 110 percent of the Appendix G P-T limits 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
granting this exemption will not have a significant effect on the
quality of the human environment (62 FR 59008).
This exemption is effective upon issuance.
Dated at Rockville, Maryland, this 12th day of December 1997.
For the Nuclear Regulatory Commission.
Frank J. Miraglia,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 97-33846 Filed 12-29-97; 8:45 am]
BILLING CODE 7590-01-P
