[Federal Register Volume 75, Number 41 (Wednesday, March 3, 2010)]
[Notices]
[Pages 9623-9625]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-4388]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. STN 50-528, STN 50-529, and STN 50-530; NRC-2010-0058]


Arizona Public Service Company, et al.; Palo Verde Nuclear 
Generating Station, Units 1, 2, and 3; Exemption

 1.0 Background

    The Arizona Public Service Company (APS, the facility licensee) is 
the holder of Facility Operating License Nos. NPF-41, NPF-51, and NPF-
74, which authorize operation of the Palo Verde Nuclear Generating 
Station (PVNGS, the facility), Units 1, 2, and 3, respectively. The 
licenses provide, among other things, that the facility is subject to 
all rules, regulations, and orders of the Nuclear Regulatory Commission 
(NRC, or the Commission) now or hereafter in effect.
    The facility consists of three pressurized-water reactors located 
in Maricopa County, Arizona.

2.0 Request/Action

    Title 10 of the Code of Federal Regulations (10 CFR) Part 50, 
Appendix G, ``Fracture Toughness Requirements,'' which is invoked by 10 
CFR 50.60, requires that pressure-temperature (P-T) limits be 
established for the reactor coolant pressure boundary 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].'' 
The regulations in 10 CFR Part 50, Appendix G, also specify the 
applicable editions and addenda of the ASME Code, Section XI, which are 
incorporated by reference in 10 CFR 50.55a. In the most recent version 
of 10 CFR (2009 Edition), the 1977 Edition through the 2004 Edition 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.''
    By letter dated February 19, 2009, as supplemented by letter dated 
December 22, 2009 (Agencywide Documents Access and Management System 
(ADAMS) Accession Nos. ML090641014 and ML100040069, respectively), the 
licensee submitted a request for exemption from 10 CFR Part 50, 
Appendix G regarding the pressure-temperature (P-T) limits calculation, 
and a license amendment request to revise Technical Specification (TS) 
3.4, ``Reactor Coolant System (RCS),'' to relocate the P-T limits and 
the low temperature overpressure protection (LTOP) system enable 
temperatures from the TS to a licensee-controlled document; the 
Pressure and Temperature Limits Report (PTLR). In the license amendment 
request, the

[[Page 9624]]

licensee identified Combustion Engineering (CE) Owners Group Topical 
Report CE NPSD-683-A, Revision 6, ``Development of a RCS Pressure and 
Temperature Limits Report (PTLR) for the Removal of P-T Limits and LTOP 
Requirements from the Technical Specifications'' (ADAMS Accession No. 
ML011350387), as the PTLR methodology that would be cited in the 
administrative controls section of the PVNGS, Units 1, 2, and 3 
Technical Specifications governing PTLR content. 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 (ADAMS Accession No. ML010780017), 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 use the methodology in CE NPSD-683-A, Revision 6, 
for the calculation of flaw stress intensity factors due to membrane 
stress from pressure loading (K<INF>IM</INF>), an exemption was 
required, since the methodology for the calculation of K<INF>IM</INF> 
values in CE NPSD-683-A, Revision 6, could not be shown to be 
conservative with respect to the methodology for the determination of 
K<INF>IM</INF> provided in editions and addenda of the ASME Code, 
Section XI, Appendix G through the 2004 Edition. Therefore, in addition 
to the license amendment request, the licensee's February 19, 2009, 
submittal also contains an exemption request, consistent with the 
requirements of 10 CFR 50.12 and 50.60, to apply the K<INF>IM</INF> 
calculational methodology of CE NPSD-683-A, Revision 6, as part of the 
PVNGS, Units 1, 2, and 3 PTLR methodology.
    During the NRC staff's review of CE NPSD-683, Revision 6, the NRC 
staff evaluated the K<INF>IM</INF> calculational methodology of that 
report versus the methodologies for the calculation of K<INF>IM</INF> 
given in the ASME Code, Section XI, Appendix G. In the NRC'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 K<INF>IM</INF> factors, and 
instead applies FEM [finite element modeling] methods for estimating 
the K<INF>IM</INF> factors for the RPV [reactor pressure vessel] shell 
* * * the staff has determined that the K<INF>IM</INF> calculation 
methods apply FEM modeling that is similar to that used for the 
determination of the K<INF>IT</INF> 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 CE/ABB Evaluation 
063-PENG-ER-096, Revision 00, ``Technical Methodology Paper Comparing 
ABB/CE PT Curve to ASME Section III, Appendix G,'' dated January 22, 
1998 (ADAMS Accession No. ML100500514, non-proprietary version). 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.'' This conclusion regarding the comparison between the CE NSSS 
methodology and the 1995 Edition of the ASME Code, Section XI, Appendix 
G methodology also applies to the 2004 Edition of the ASME Code, 
Section XI, Appendix G methodology because the evolution of the ASME 
Code Section XI, Appendix G methodology does not affect the 
K<INF>IM</INF> calculation significantly.
    In summary, the staff concluded in its March 16, 2001, SE that the 
calculation of K<INF>IM</INF> 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 methods, 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 RPV due to membrane stress from pressure loadings in 
lieu of meeting the requirements in 10 CFR 50.60 and 10 CFR Part 50, 
Appendix G. 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 P-T limits and the minimum permissible 
temperature are established for the RPV under normal operating and 
hydrostatic or leak rate test conditions. The licensee's alternative 
methodology for establishing the P-T limits and the LTOP setpoints is 
described in CE NPSD-683-A, Revision 6, which 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 used 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 stress 
intensity factors in the reactor vessel material has no negative 
implications for 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 PVNGS, Units 1, 2, and 3 with P-
T limit curves developed in accordance with the ASME Code, Section XI, 
Appendix G is not necessary to achieve the underlying purpose of 10 CFR 
part 50, appendix G. Application of the K<INF>IM</INF> 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 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

[[Page 9625]]

the reactor coolant system 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 agrees with the licensee's determination that an 
exemption is required to approve the use of the K<INF>IM</INF> 
calculational methodology of CE NPSD-683-A, Revision 6. The staff 
concludes that the application of the K<INF>IM</INF> calculational 
methodology of CE NPSD-683-A, Revision 6, for PVNGS, Units 1, 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 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 
K<INF>IM</INF> 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 K<INF>IM</INF> 
calculational methodology of CE NPSD-683-A, Revision 6, is acceptable 
for use in the PVNGS, Units 1, 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 APS an exemption from the 
requirements of 10 CFR part 50, appendix G to allow application of the 
K<INF>IM</INF> calculational methodology of CE NPSD-683-A, Revision 6 
in establishing the PTLR methodology for PVNGS, Units 1, 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 (75 FR 8149; dated February 23, 2010).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 24th day of February 2010.

    For the Nuclear Regulatory Commission.
Allen G. Howe,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2010-4388 Filed 3-2-10; 8:45 am]
BILLING CODE 7590-01-P