Federal Register / Vol. 75, No. 41 / Wednesday, March 3, 2010 / Notices 9623 

Nuclear Energy Institute). The licensee’s 

request for an exemption is therefore 

consistent with the approach set forth 

by the Commission as discussed in the 

June 4, 2009, letter. 

VEGP Schedule Exemption Request 

The licensee provided detailed 

information in its letter dated November 

6, 2009, as supplemented November 20, 

2009, requesting an exemption. It 

describes a comprehensive plan to 

install equipment related to a certain 

requirement in the new Part 73 rule and 

provides a timeline for achieving full 

compliance with the new regulation. 

The submittals contain proprietary 

information regarding the site security 

plan, details of the specific requirement 

of the regulation for which the site 

cannot be in compliance by the March 

31, 2010, deadline and why, the 

required changes to the site’s security 

configuration, and a timeline with 

critical path activities that will bring the 

licensee into full compliance by 

September 27, 2010. The timeline 

provides dates indicating (1) when 

various phases of the project begin and 

end (i.e., design, field construction), (2) 

outages scheduled for each unit, and (3) 

when critical equipment will be 

ordered, installed, tested and become 

operational. 

Notwithstanding the schedular 

exemption for this limited requirement, 

the licensee will continue to be in 

compliance with all other applicable 

physical security requirements as 

described in 10 CFR 73.55 and reflected 

in its current NRC approved physical 

security program. By September 27, 

2010, VEGP will be in full compliance 

with all the regulatory requirements of 

10 CFR 73.55, as issued on March 27, 

2009. 

4.0 Conclusion for Part 73 Schedule 

Exemption Request 

The NRC staff has reviewed the 

licensee’s submittals and concludes that 

the licensee has provided adequate 

justification for its request for an 

extension of the compliance date to 

September 27, 2010, with regard to a 

specific requirement of 10 CFR 73.55. 

Accordingly, the Commission has 

determined that pursuant to10 CFR 

73.5, ‘‘Specific exemptions,’’ an 

exemption from the March 31, 2010, 

compliance date is authorized by law 

and will not endanger life or property or 

the common defense and security, and 

is otherwise in the public interest. 

Therefore, the Commission hereby 

grants the requested exemption. 

The NRC staff has determined that the 

long-term benefits that will be realized 

when the VEGP equipment installation 

is complete justifies extending the full 

compliance date with regard to the 

specific requirement of 10 CFR 73.55. 

The security measure, that VEGP needs 

additional time to implement, is a new 

requirement imposed by the March 27, 

2009, amendments to 10 CFR 73.55, and 

is in addition to those required by the 

security orders issued in response to the 

events of September 11, 2001. 

Therefore, it is concluded that the 

licensee’s actions are in the best interest 

of protecting the public health and 

safety through the security changes that 

will result from granting this exemption. 

As per the licensee’s request and the 

NRC’s regulatory authority to grant an 

exemption from the March 31, 2010, 

implementation deadline for the 

requirement specified in the SNC letter 

dated November 6, 2009, as 

supplemented November 20, 2009, the 

licensee is required to be in full 

compliance by September 27, 2010. In 

achieving compliance, the licensee is 

reminded that it is responsible for 

determining the appropriate licensing 

mechanism (i.e., 10 CFR 50.54(p) or 10 

CFR 50.90) for incorporation of all 

necessary changes to its security plans. 

Pursuant to 10 CFR 51.32, ‘‘Finding of 

no significant impact,’’ the Commission 

has previously determined that the 

granting of this exemption will not have 

a significant effect on the quality of the 

human environment 75 FR 3943; dated 

January 25, 2010. 

This exemption is effective upon 

issuance. 

Dated at Rockville, Maryland, this 24 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. 2010–4381 Filed 3–2–10; 8:45 am] 

BILLING CODE 7590–01–P 

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 pressuretemperature (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 

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9624 Federal Register / Vol. 75, No. 41 / Wednesday, March 3, 2010 / Notices 

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 

(KIM), an exemption was required, since 

the methodology for the calculation of 

KIM values in CE NPSD–683–A, 

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 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 KIM

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 KIM calculational 

methodology of that report versus the 

methodologies for the calculation of KIM

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 

KIM factors, and instead applies FEM 

[finite element modeling] methods for 

estimating the KIM factors for the RPV 

[reactor pressure vessel] 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 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 KIM calculation significantly. 

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 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 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 

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 

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Federal Register / Vol. 75, No. 41 / Wednesday, March 3, 2010 / Notices 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 KIM calculational 

methodology of CE NPSD–683–A, 

Revision 6. The staff concludes that the 

application of the KIM 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 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 KIM

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 KIM 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 

NUCLEAR REGULATORY 

COMMISSION 

[Docket Nos. 50–282 and 50–306; NRC– 

2010–0046] 

Northern States Power Company— 

Minnesota Prairie Island Nuclear 

Generating Plant, Units 1 and 2; 

Exemption 

1.0 Background 

Northern States Power Company, a 

Minnesota corporation (NSPM, the 

licensee) is the holder of Facility 

Operating License Nos. DPR–42 and 

DPR–60, which authorize operation of 

the Prairie Island Nuclear Generating 

Plant, Units 1 and 2 (PINGP). 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 

Goodhue County, Minnesota. 

2.0 Request/Action 

Title 10 of the Code of Federal 

Regulations (10 CFR) Part 73, ‘‘Physical 

protection of plants and materials,’’ 

Section 73.55, ‘‘Requirements for 

physical protection of licensed activities 

in nuclear power reactors against 

radiological sabotage,’’ published March 

27, 2009, effective May 26, 2009, with 

a full implementation date of March 31, 

2010, requires licensees to protect, with 

high assurance, against radiological 

sabotage by designing and 

implementing comprehensive site 

security programs. The amendments to 

10 CFR 73.55 published on March 27, 

2009, establish and update generically 

applicable security requirements similar 

to those previously imposed by 

Commission orders issued after the 

terrorist attacks of September 11, 2001 

and implemented by licensees. In 

addition, the amendments to 10 CFR 

73.55 include additional requirements 

to further enhance site security based 

upon insights gained from 

implementation of the post September 

11, 2001 security orders. It is from five 

of these new requirements that PINGP 

now seeks an exemption from the March 

31, 2010 implementation date. All other 

physical security requirements 

established by this recent rulemaking 

have already been or will be 

implemented by the licensee by March 

31, 2010. 

By letter dated November 5, 2009, as 

supplemented by letters dated 

November 30 and December 17, 2009, 

the licensee requested an exemption in 

accordance with 10 CFR 73.5, ‘‘Specific 

exemptions.’’ The licensee’s November 5 

and December 17, 2009, letters, and 

certain portions of its November 30, 

2009 (Agencywide Documents Access 

and Management System (ADAMS) 

Accession No. ML100050096), letter, 

contain security-related information 

and, accordingly, are not available to the 

public. The licensee has requested an 

exemption from the March 31, 2010, 

compliance date identified in 10 CFR 

73.55(a)(1), stating that specific parts of 

the new requirements will require more 

time to implement before all 

requirements can be met. Specifically, 

the request is to extend the compliance 

date for five specific requirements from 

the current March 31, 2010, deadline to 

June 30, 2011. Being granted this 

exemption for the five requirements 

would allow the licensee to complete 

the modifications designed to provide 

significant upgrades to the security 

system to meet the noted regulatory 

requirements. 

3.0 Discussion of Part 73 Schedule 

Exemptions From the March 31, 2010, 

Full Implementation Date 

Pursuant to 10 CFR 73.55(a)(1), ‘‘By 

March 31, 2010, each nuclear power 

reactor licensee, licensed under 10 CFR 

part 50, shall implement the 

requirements of this section through its 

Commission-approved Physical Security 

Plan, Training and Qualification Plan, 

Safeguards Contingency Plan, and Cyber 

Security Plan, referred to collectively 

hereafter as ‘security plans.’ ’’ Pursuant 

to 10 CFR 73.5, the Commission may, 

upon application by any interested 

person or upon its own initiative, grant 

exemptions from the requirements of 10 

CFR part 73 when the exemptions are 

authorized by law, and will not 

endanger life or property or the common 

defense and security, and are otherwise 

in the public interest. 

NRC approval of this exemption 

would, as noted above, extend the 

required compliance date for the 

requirements specified in the licensee’s 

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