[Federal Register Volume 63, Number 17 (Tuesday, January 27, 1998)]
[Notices]
[Pages 3927-3928]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-1902]
[[Page 3927]]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-454, STN 50-455 and STN 50-456, STN 50-457]
Commonwealth Edison Company; (Byron Station, Units 1 and 2);
(Braidwood Station, Units 1 and 2); Exemption
I
Commonwealth Edison Company (ComEd, the licensee) is the holder of
Facility Operating License Nos. NPF-37, NPF-66, NPF-72, and NPF-77,
which authorize operation of Byron Station, Units 1 and 2, and
Braidwood 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 Byron facility consists of two pressurized-water reactors
located at the licensee's site in Ogle County, Illinois. The Braidwood
facility consists of two pressurized-water reactors located at the
licensee's site in Will County, Illinois.
II
In its letter dated April 3, 1997, as supplemented on June 19,
1997, ComEd 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 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 service lifetime, and
specifies that these P-T limits must be at least as conservative as the
limits obtained by conforming to the methods of analysis and the
margins of safety of the American Society of Mechanical Engineers
(ASME) Boiler and Pressure Vessel Code (Code), Section XI, Appendix G.
10 CFR 50.55a requires that any reference to ASME Code Section XI in 10
CFR Part 50 refers to 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 requirements
described 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 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 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 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, when instrument uncertainty is considered, the
operating window between the PORV setpoint and the minimum pressure
required for reactor coolant pump seals is small and presents
difficulties for plant operation.
The licensee has requested the use of the 1996 Addenda to the ASME
Code, Section XI, Appendix G, which allows the use of lower stress
intensity factors for determining the applied stress intensity from
pressure and thermal stresses, and allows use of an LTOP system
setpoint so that system pressure does not exceed 110 percent of the P-T
limits. The 1996 Addenda to the ASME Code, Section XI, Appendix G, 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, Section XI, Appendix G, 1996 Addenda,
has been approved by the ASME Code Committee.
III
Pursuant to 10 CFR 50.12, the Commission may, upon application by
any interested entity 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 and
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 by following the methods of analysis
and the required margins of safety of Appendix G of Section XI of the
ASME Code. 10 CFR 50.55a requires that any reference to ASME Code
Section XI in 10 CFR Part 50, Appendix G, refers to addenda through the
1988 Addenda and editions through the 1989 Edition of the ASME Code,
unless otherwise noted.
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 of the vessel wall thickness (\1/4\ T) 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 reactor
vessel material.
For determining the P-T limits, the licensee proposed to use the
safety margins based on the 1996 Addenda to the ASME Code in lieu of
the 1989 Edition. When compared to the 1989 Edition of the ASME Code,
the 1996 Addenda permits the use of a lower stress intensity factor for
determining the applied stress intensity from pressure and thermal
stresses. This results in a slight reduction in the applied stress
intensity and a corresponding shift in the allowable pressure at a
given temperature in the non-conservative direction; however, this
difference is minor when compared to the explicit conservatisms
incorporated into Appendix G, and the changes in the stress intensity
factor are supported by the work performed for NRC and for others by
J.A. Keeney and T.L. Dickson at Oak Ridge National Laboratory (ORNL).
For determining the LTOP system setpoint, the licensee proposed to
use safety margins based on the 1996 Addenda to the ASME Code. The 1996
[[Page 3928]]
Addenda allows determination of the setpoint for mitigating LTOP events
so that the maximum pressure in the vessel would not exceed 110 percent
of the
P-T limits that are determined using the 1996 methodology. 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 provide
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
PORV would be reduced, thereby improving plant safety.
It should be noted that the provision to set the PORV setpoint so
that system pressure remains below 110 percent of the P-T limits has
already been incorporated into the Byron and Braidwood licensing basis.
This provision was approved by an exemption to 10 CFR 50.60 granted to
Byron, Units 1 and 2, on November 29, 1996, to Braidwood, Unit 1 on
July 13, 1995, and to Braidwood, Unit 2 on December 12, 1997, to allow
the use of ASME Code Case N-514. Therefore, although it represents a
change from the 1989 Edition of the ASME Code, it is not a change to
the current licensing basis for the facilities.
IV
For the foregoing reasons, the NRC staff has concluded that ComEd'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 10
CFR 50.60 need not be applied 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 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 so that the P-T limits may be
determined using the 1996 Addenda to the ASME Code, Section XI,
Appendix G, and the LTOP system setpoint may be determined so that
system pressure does not exceed 110 percent of the P-T limits.
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 (63 FR 2268).
This exemption is effective upon issuance.
Dated at Rockville, Maryland, this 16th day of January, 1998.
For the Nuclear Regulatory Commission.
Frank J. Miraglia,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 98-1902 Filed 1-26-98; 8:45 am]
BILLING CODE 7590-01-P