[Federal Register Volume 71, Number 129 (Thursday, July 6, 2006)]
[Notices]
[Pages 38430-38432]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-10529]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-361 and 50-362]


Southern California Edison Company, San Diego Gas and Electric 
Company, the City of Riverside, CA, the City of Anaheim, CA; San Onofre 
Nuclear Generating Station, Units 2 and 3; Exemption

1.0 Background

    Southern California Edison Company (the licensee) is the holder of 
Facility Operating License Nos. NPF-10 and NPF-15, which authorize 
operation of the San Onofre Nuclear Generating Station, Unit 2 and Unit 
3 (SONGS 2 and 3), respectively. The licenses provide, among other 
things, that the facility is subject to all rules, regulations, and 
orders of the U.S. Nuclear Regulatory Commission (NRC, the Commission) 
now or hereafter in effect.
    The facility consists of two pressurized-water reactors located in 
San Diego County, California.

2.0 Request/action

    Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 
Appendix G, which is invoked by 10 CFR 50.60, requires that pressure-
temperature (P-T) limits be established for reactor pressure vessels 
(RPVs) during normal operating and hydrostatic or leak rate testing 
conditions. Specifically, 10 CFR Part 50, Appendix G, states that 
``[t]he appropriate requirements on both the pressure-temperature 
limits and the minimum permissible temperature must be met for all 
conditions,'' and ``[t]he pressure-temperature limits identified as 
`ASME [American Society for Mechanical Engineers] Appendix G limits' in 
Table 3 require that the limits must be at least as conservative as 
limits obtained by following the methods of analysis and the margins of 
safety of Appendix G of Section XI of the ASME Code [Boiler and 
Pressure Vessel Code].'' Part 50 of Title 10 of the Code of Federal 
Regulations, Appendix G, also specifies that the editions and addenda 
of the ASME Code, Section XI, which are incorporated by reference in 10 
CFR 50.55a, apply to the requirements in 10 CFR Part 50,

[[Page 38431]]

Appendix G. In the 2005 Edition of the Code of Federal Regulations, the 
1977 Edition through the 2003 Addenda of the ASME Code, Section XI are 
incorporated by reference in 10 CFR 50.55a. Finally, 10 CFR 50.60(b) 
states that, ``[p]roposed alternatives to the described requirements in 
Append[ix] G * * * of this part or portions thereof may be used when an 
exemption is granted by the Commission under [10 CFR 50.12].''
    In the licensee's January 28, 2005, license amendment request to 
implement a pressure-temperature limits report (PTLR) for SONGS 2 and 
3, the licensee identified Combustion Engineering (CE) Owners Group 
Topical Report NPSD-683-A, ``The Development of a RCS [Reactor Coolant 
System] Pressure and Temperature Limits Report for the Removal of P-T 
Limits and LTOP [low temperature overpressure protection] Setpoints 
from the Technical Specifications,'' as the PTLR methodology that would 
be cited in the administrative control section of the SONGS 2 and 3 
Technical Specifications governing PTLR content. CE NPSD-683-A refers 
to an NRC-approved version of Topical Report CE NPSD-683. The NRC staff 
evaluated the specific PTLR methodology in CE NPSD-683, Revision 6. 
This evaluation was documented in the NRC safety evaluation (SE) of 
March 16, 2001, which specified additional licensee actions that are 
necessary to support a licensee's adoption of CE NPSD-683, Revision 6. 
The final approved version of this report was reissued as CE NPSD-683-
A, Revision 6, which included the NRC SE and the required additional 
action items as an attachment to the report. One of the additional 
specified actions stated that if a licensee proposed to utilize the 
methodology in CE NPSD-683, Revision 6, for the calculation of flaw 
stress intensity factors due to membrane stress from pressure loading 
(KIM), an exemption was required since the methodology for 
the calculation of KIM values in CE NPSD-683, Revision 6, 
could not be shown to be conservative with respect to the methodology 
for the determination of KIM provided in editions and 
addenda of the ASME Code, Section XI, Appendix G, through the 2003 
Addenda. Therefore, in connection with the licensee's January 28, 2005, 
license amendment request, as supplemented by its letter dated January 
12, 2006, the licensee also submitted an exemption request, consistent 
with the requirements of 10 CFR 50.60, to apply the KIM 
calculational methodology of CE NPSD-683-A, Revision 6, as part of the 
SONGS 2 and 3 PTLR methodology.
    During the NRC staff's review of CE NPSD-683, Revision 6, the NRC 
staff evaluated the KIM calculational methodology of CE 
NPSD-683, Revision 6, versus the methodologies for KIM 
calculation given in the ASME Code, Section XI, Appendix G. In the 
staff's March 16, 2001 SE, the staff noted, ``[t]he CE NSSS [nuclear 
steam supply system] methodology does not invoke the methods in the 
1995 edition of Appendix G to the Code for calculating KIM 
factors, and instead applies FEM [finite element modeling] methods for 
estimating the KIM factors for the RPV shell * * * the staff 
has determined that the KIM calculation methods apply FEM 
modeling that is similar to that used for the determination of the 
KIT factors [as codified in the ASME Code, Section XI, 
Appendix G]. The staff has also determined that there is only a slight 
non-conservative difference between the P-T limits generated from the 
1989 edition of Appendix G to the Code and those generated from CE NSSS 
methodology as documented in Evaluation No. 063-PENG-ER-096, Revision 
00. The staff considers that this difference is reasonable and that it 
will be consistent with the expected improvements in P-T generation 
methods that have been incorporated into the 1995 edition of Appendix G 
to the Code.''
    In summary, the staff concluded in its March 16, 2001, SE that the 
calculation of KIM using the CE NPSD-683, Revision 6, 
methodology would lead to the development of P-T limit curves, which 
may be slightly non-conservative with respect to those which would be 
calculated using the ASME Code, Section XI, Appendix G, and that such a 
difference was to be expected with the development of more refined 
calculational techniques. Furthermore, the staff concluded in its March 
16, 2001, SE that P-T limit curves that would be developed using the 
methodology of CE NPSD-683, Revision 6, would be adequate for 
protecting the RPV from brittle fracture under all normal operating and 
hydrostatic/leak test conditions.

3.0 Discussion

    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.
    This exemption results in changes to the plant by allowing the use 
of an alternative methodology for calculating flaw stress intensity 
factors in the reactor pressure vessel due to membrane stress from 
pressure loadings in lieu of meeting the requirements in 10 CFR 50.60. 
As stated above, 10 CFR 50.12 allows NRC to grant exemptions from the 
requirements of 10 CFR Part 50. In addition, the granting of the 
exemption will not result in violation of the Atomic Energy Act of 
1954, as amended, or the Commission's regulations. Therefore, the 
exemption is authorized by law.
    The underlying purpose of 10 CFR 50.60 and 10 CFR Part 50, Appendix 
G, is to ensure that appropriate pressure-temperature limits and the 
minimum permissible temperature are established for the reactor 
pressure vessel under normal operating and hydrostatic or leak rate 
conditions. The licensee's alternative methodology for establishing the 
P-T limits and low-temperature overpressure protection setpoints are 
described in Combustion Engineering Owners' Topical Report NPSD-683-A, 
and has been approved by the NRC staff. Based on the above, no new 
accident precursors are created by using the alternative methodology, 
thus, the probability of postulated accidents is not increased. Also, 
based on the above, the consequences of postulated accidents are not 
increased. In addition, the licensee will use an NRC-approved 
methodology for establishing P-T limits and minimum permissible 
temperatures for the reactor vessel. Therefore, there is no undue risk 
to the public health and safety.
    The exemption results in changes to the plant by allowing an 
alternative methodology for calculating flaw stress intensity factors 
in the reactor vessel. This change to the calculation of stresses in 
the reactor vessel material has no relation to security issues. 
Therefore, the common defense and security is not impacted by this 
exemption.
    Special circumstances, pursuant to 10 CFR 50.12(a)(2)(ii), are 
present in that continued operation of SONGS 2 and 3 with P-T limit 
curves developed in accordance with the ASME Code, Section XI, Appendix 
G, without the authorization to utilize the alternative KIM 
calculational methodology of CE NPSD-683-A, Revision 6, is not 
necessary to achieve the underlying purpose of 10 CFR Part 50, Appendix 
G. Application of the KIM calculational methodology of CE 
NPSD-683-A, Revision 6, in lieu of the calculational methodology 
specified in the ASME Code, Section XI, Appendix G, provides an 
acceptable alternative evaluation

[[Page 38432]]

procedure, which will continue to meet the underlying purpose of 10 CFR 
Part 50, Appendix G. The underlying purpose of the regulations in 10 
CFR Part 50, Appendix G, is to provide an acceptable margin of safety 
against brittle failure of the RCS during any condition of normal 
operation to which the pressure boundary may be subjected over its 
service lifetime.
    Based on the staff's March 16, 2001, SE regarding CE NPSD-683, 
Revision 6, and the licensee's rationale to support the exemption 
request, the staff accepts the licensee's determination that an 
exemption would be required to approve the use of the KIM 
calculational methodology of CE NPSD-683-A, Revision 6. The staff 
concludes that the application of the technical provisions of the 
KIM calculational methodology of CE NPSD-683-A, Revision 6, 
by SONGS 2 and 3 provides sufficient margin in the development of RPV 
P-T limit curves such that the underlying purpose of the regulations 
(10 CFR Part 50, Appendix G) continues to be met. Therefore, the NRC 
staff concludes that the exemption requested by the licensee is 
justified based on the special circumstances of 10 CFR 50.12(a)(2)(ii), 
``[a]pplication 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.''
    Based upon a consideration of the conservatism that is explicitly 
incorporated into the methodologies of 10 CFR Part 50, Appendix G, and 
ASME Code, Section XI, Appendix G, the staff concludes that application 
of the KIM calculational methodology of CE NPSD-683-A, 
Revision 6, as described, would provide an adequate margin of safety 
against brittle failure of the RPV. Therefore, the staff concludes that 
the exemption is appropriate under the special circumstances of 10 CFR 
50.12(a)(2)(ii), and that the application of the technical provisions 
of the KIM calculational methodology of CE NPSD-683-A, 
Revision 6, should be approved for use in the SONGS 2 and 3 PTLR 
methodology.

4.0 Conclusion

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), the exemption is authorized by law, will not present an undue 
risk to the public health and safety, and is consistent with the common 
defense and security. Also, special circumstances are present. 
Therefore, the Commission hereby grants Southern California Edison 
Company an exemption from the requirements of 10 CFR Part 50, Appendix 
G, to allow application of the KIM calculational methodology 
of CE NPSD-683-A, Revision 6, in establishing the PTLR methodology for 
SONGS 2 and 3.
    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 (71 FR 19553; dated April 14, 2006).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 5th day of June 2006.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E6-10529 Filed 7-5-06; 8:45 am]
BILLING CODE 7590-01-P