Pressurized thermal shock events are system transients in a pressurized water reactor (PWR) in which severe overcooling occurs coincident with high pressure. The thermal stresses caused by rapid cooling of the reactor vessel inside surface combine with the stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a preexisting flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent to the core where neutron radiation gradually embrittles the material over the lifetime of the plant may be such a material.

The toughness of ferritic reactor vessel materials is characterized by a “reference temperature for nil ductility transition” (RT NDT ). RT NDT is referred to as a ductile-to-brittle transition temperature. At temperatures below RT NDT fracture occurs very rapidly, by cleavage, a behavior referred to as “brittle.” As temperatures increase above RT NDT , progressively larger amounts of deformation occur before rapid cleavage fracture occurs. Eventually, at temperatures above approximately RT NDT  + 60 °F, there is no longer adequate stress intensification to promote cleavage and fracture occurs by the slower mechanism of micro-void initiation, growth, and coalescence into the crack, a behavior referred to as “ductile.”

At normal operating temperature, ferritic reactor vessel materials are usually tough. However, neutron radiation embrittles the material over time, causing a shift in RT NDT to higher temperatures. Correlations based on test results for unirradiated and irradiated specimens have been developed to calculate the shift in RT NDT as a function of neutron fluence (the integrated neutron flux over a specified time of plant operation) for various material compositions. The value of RT NDT at a given time in a reactor vessel's life is used in fracture mechanics calculations to determine the probability that assumed pre-existing flaws would propagate when the reactor vessel is stressed.

The Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61, adopted on July 23, 1985 (50 FR 29937), establishes screening criteria below which the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The screening criteria effectively define a limiting level of embrittlement beyond which operation cannot continue without further plant-specific evaluation. Regulatory Guide (RG) 1.154, “Format and Content of Plant-Specific Pressurized Thermal Shock Analysis Reports for Pressurized Water Reactors,” indicates that reactor vessels that exceed the screening criteria in the rule may continue to operate provided they can demonstrate a mean through-wall crack frequency (TWCF) from PTS-related events of no greater than 5 × 10 −6 per reactor year.

Any reactor vessel with materials predicted to exceed the screening criteria in 10 CFR 50.61 may not continue to operate without implementation of compensatory actions or additional plant-specific analyses unless the licensee receives an exemption from the requirements of the rule. Acceptable compensatory actions are neutron flux reduction, other plant modifications to reduce PTS event probability or severity, and reactor vessel annealing, which are addressed in 10 CFR 50.61(b)(3), (b)(4), and (b)(7); and 10 CFR 50.66, respectively.

No currently operating PWR reactor vessel is projected to exceed the 10 CFR 50.61 screening criteria before the expiration of its 40 year operating license. However, several PWR reactor vessels are approaching the screening criteria, while others are likely to exceed the screening criteria during their first license renewal periods.

The NRC's Office of Nuclear Regulatory Research (RES) has completed a research program to update the PTS regulations. The results of this research program conclude that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. This finding indicates that the screening criteria in 10 CFR 50.61 are unnecessarily conservative and may impose an unnecessary burden on some licensees. Therefore, the NRC is proposing a new rule, 10 CFR 50.61a, which would provide alternative screening criteria and corresponding embrittlement correlations based on the updated technical basis. The updated embrittlement correlation is the projected increase in the Charpy V-notch 30 ft-lb transition temperature for reactor vessel materials resulting from neutron radiation and is calculated using equations 5 through 7 of the proposed rule. The proposed rule would be voluntary for all holders of a PWR operating license under 10 CFR part 50 or a combined license under 10 CFR part 52, although it is intended for licensees with reactor vessels that cannot demonstrate compliance with the more restrictive criteria in 10 CFR 50.61. The requirements of 10 CFR 50.61 would continue to apply to licensees who choose not to implement 10 CFR 50.61a.

The following two reports provide the technical basis for this rulemaking: (1) NUREG-1806, “Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report,” and (2) NUREG-1874, “Recommended Screening Limits for Pressurized Thermal Shock (PTS).” These reports summarize and reference several additional reports on the same topic. The updated technical basis indicates that, after 60 years of operation, the risk of reactor vessel failure due to a PTS event is much lower than previously estimated. The updated analyses were based on information from three currently operating PWRs. Because the severity of the risk-significant transient classes (i.e., primary side pipe breaks, stuck open valves on the primary side that may later re-close) is controlled by factors that are common to PWRs in general, the NRC concludes that the TWCF results and resultant RT-based screening criteria developed from their analysis of three plants can be applied with confidence to the entire fleet of operating PWRs. This conclusion is based on an understanding of characteristics of the dominant transients that drive their risk significance and on an evaluation of a larger population of high embrittlement PWRs. This evaluation revealed no design, operational, training, or procedural factors that could credibly increase either the severity of these transients or the frequency of their occurrence in the general PWR population above the severity/frequency characteristic of the three plants that were modeled in detail.

The current guidance provided by Regulatory Guide 1.174, Revision 1, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” for large early release frequency (LERF) was used to relate the PTS screening criteria in 10 CFR 50.61a to an acceptable yearly limit of 1 × 10 −6 per reactor year on reactor vessel TWCF. Although many post-through-wall cracking accident progressions are expected to lead only to core damage (which suggests a 1 × 10 −5 events per year limit on TWCF per Regulatory Guide 1.174), uncertainties in the accident progression analysis led to the recommendation of adopting the more conservative TWCF limit of 1 × 10 −6 per reactor year based on LERF.

The updated technical basis uses many different models and parameters to estimate the yearly probability that a PWR will develop a through-wall crack as a consequence of PTS loading. One of these models is a revised embrittlement correlation that uses information on the chemical composition and neutron exposure of low alloy steels in the reactor vessel's beltline region to estimate the resistance to fracture of these materials. Although the general trends of the embrittlement models in 10 CFR 50.61 and the proposed rule are similar, the form of the revised embrittlement correlation differs substantially from the correlation in the existing 10 CFR 50.61. The correlation in 10 CFR 50.61a has been updated to more accurately represent the substantial amount of reactor vessel surveillance data that has accumulated since the embrittlement correlation was last revised during the 1980s.

This proposed rule would differ from the current rule in that it would contain a requirement for licensees who choose to follow its requirements to analyze the results from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Section XI in service inspection volumetric examinations. This requirement would be provided in paragraph (e) of the proposed rule. The examinations and analyses would confirm that the flaw density and size in the licensee's reactor vessel beltline are bounded by the flaw density and size utilized in the technical basis. The technical basis was developed using a flaw density, spatial distribution, and size distribution determined from a small amount of experimental data, as well as from physical models and expert elicitation. The experimental data included 22,210 cubic inches of weld metal, 3845 cubic inches of plate, and 1650 cubic inches of clad. The experimental data were obtained from samples removed from reactor vessel materials from cancelled plants (Shoreham and the Pressure Vessel Research Users Facility (PVRUF) vessel). The NRC considers that the analysis of the ASME Code inservice inspection volumetric examination is needed to confirm that the flaw density and size distributions in the reactor vessel to which the proposed rule may be applied are consistent with those in the technical basis because the experimental data was obtained from a limited number of reactor vessels.

Paragraph (g)(6)(ii)(c) of 10 CFR 50.55a requires licensees to implement Supplements 4 and 6 in Appendix VIII to ASME BPV Code Section XI after November 22, 2000. Supplement 4 contains qualification requirements for the reactor vessel inservice inspection volume from the clad-to-base metal interface to the inner 1.0 inch or 10 percent of the vessel thickness, whichever is larger. Supplement 6 contains qualification requirements for reactor vessel weld volumes other than those near the clad-to-base metal interface.

The performance of inspectors who have gone through the Supplement 4 qualification process has been documented in a paper by Becker (Becker, L., “Reactor Pressure Vessel Inspection Reliability,” Proceeding of the Joint EC-IAEA Technical Meeting on the Improvement in In-Service Inspection Effectiveness, Petten, the Netherlands, November 2002). Analysis of the results reported in this paper indicates that an inspector using a Supplement 4 qualification procedure would have an 80 percent probability of detecting a flaw with a through-wall extent of 0.1 inch and would have an approximately 99 percent probability of detecting a flaw with a through-wall extent of 0.3 inch. Therefore, there is an 80 percent or greater probability of detecting a flaw that contributes to crack initiation from PTS events in reactor vessels with embrittlement conditions characteristic of 1 × 10 −6 per reactor-year TWCF when they are inspected using ASME BPV Code Section XI, Appendix VIII, Supplement 4 requirements.

The true flaw density for flaws with a through wall extent of between 0.1 and 0.3 inch can be inferred from the ASME Code examination results and the probability of detection. The proposed rule would require licensees to determine if:

(1) The indication density and size within the weld and base metal inservice inspection volume from the clad-to-base metal interface to the inner 1.0 inch or 10 percent of the vessel thickness are within the flaw density and size distributions that were used in the technical basis represented in Tables 2 and 3 in the proposed rule;

(2) Any indications within the weld and base metal inservice inspection volume from the clad-to-base metal interface to the inner 1.0 inch or 10 percent of the vessel thickness are larger than the sizes in Tables 2 and 3;

(3) Any indications between the clad-to-base metal interface and three-eights of the vessel thickness exceed the size allowable in ASME BPV Code Section XI, Table IWB-3510-1; or

(4) Any linear indications that penetrate through the clad into the welds or the adjacent base metal.

The technical basis for the proposed rule concludes that flaws as small as 0.1 inch deep contribute to TWCF and that nearly all of the contributions come from flaws in the range below 1 inch deep for reactor vessels with embrittlement characteristics of TWCF equal to 1 × 10 −6 per reactor year. The peak contribution comes from flaws between 0.1 and 0.2 inch deep, because that is the range that has the maximum combined effect from the number of flaws, which is decreasing with flaw size, and their susceptibility to brittle fracture, which is increasing with flaw size. For weld flaws that exceed the sizes in the table, the risk analysis indicates that a single flaw can be expected to contribute a significant fraction of the 1 × 10 −6 /reactor-year limit on TWCF. Therefore, if a flaw of that size is found in a reactor vessel, it is important to more accurately assess if its size and location with respect to the local level of embrittlement challenge the regulatory limit.

The technical basis for the proposed rule indicates that flaws buried deeper than 1 inch from the inner surface of the reactor vessel are not as susceptible to brittle fracture as similar size flaws located closer to the inner surface. Therefore, the proposed rule would not require the comparison of the density of such flaws, but still would require large flaws, if discovered, to be evaluated for contributions to TWCF if they are within the inner three-eights of the vessel thickness. This requirement would be provided in paragraph (e)(4)(iv) of the proposed rule. The limitation for flaw acceptance, specified in ASME Code Section XI Table IWB-3510-1, approximately corresponds to the threshold for flaw sizes that can make a significant contribution to TWCF if present in reactor vessel material at this depth. Therefore, this proposed rule would require these flaws to be evaluated for contribution to TWCF in addition to the other evaluations for such flaws that are prescribed in the ASME Code.

The numerical values in Tables 2 and 3 of the proposed rule would represent the number of flaws in each size range that were derived from the technical basis. Table 2 for the weld flaws is limited to flaw sizes that are frequent enough to be expected to occur in most plants. Similarly, Table 3 for the plate and forging flaws stops at the maximum flaw size that was modeled for these materials in the technical basis. If one or more larger flaws are found in a reactor vessel, they must be evaluated to ensure that they are not causing the TWCF for that reactor vessel to exceed the regulatory limit.

Surface cracks that penetrate through the stainless steel clad into the welds or the adjacent base metal were not included in the technical basis because these types of flaws have not been observed in the beltline of an operating PWR reactor vessel. However, flaws of this type were observed in the Quad Cities Unit 2 reactor vessel head in 1990 (NUREG-1796, “Safety Evaluation Report related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2”). The observed cracks had a maximum depth into the base metal of approximately 6 mm (0.24 inch) and penetrated through the stainless steel clad. Quad Cities Units 2 and 3 are boiling water reactors which are not susceptible to PTS events and hence are not subject to 10 CFR 50.61. The cracking at Quad Cities Unit 2 was attributed to intergranular stress corrosion cracking (IGSCC) of the stainless steel cladding, which has not been observed in PWR reactor vessels, and hot cracking of the low alloy steel metal base. If these cracks were in the beltline region of a PWR, they would be a significant contributor to TWCF because of their size and location. The proposed rule would require licensees to determine if cracks of this type exist in the beltline weld region at each ASME Code Section XI ultrasonic examination. This requirement would be provided in paragraph (e)(2) of the proposed rule.

The ASME Code specifies that the dimension of flaws detected by nondestructive examination be expressed to the nearest 0.05 inch for indications less than 1 inch. Hence, the examination results from the ASME Code volumetric examination will be reported in multiples of 0.05 inch with a range of ±0.025 inch. Therefore, Tables 2 and 3 in the proposed rule describe the flaw density in multiples of 0.05 inch with a size range of ±0.025 inch.

The ASME Code standard for reporting flaw sizes did not match the size increments in the technical basis. Therefore, the NRC staff developed a procedure to distribute the flaws used in the technical basis into ASME Code-sized ranges. This is explained in greater detail in the NRC staff document “Development of Flaw Size Distribution Tables for Draft Proposed Title 10 of the Code of Federal Regulations (10 CFR) 50.61a” (refer to ADAMS accession number ML070950392).

The values in Tables 2 and 3 of the proposed rule exceed the values for those size ranges that were developed from the laboratory analyses of the two reactor vessels. It was decided to allow licensees to use the Table 2 and 3 values instead of the values that would come from the laboratory results because it is still conservative to model all of the flaws as if they were the largest size for each of the ASME Code size ranges. In effect, some of the conservatism that was in the original risk modeling is being made available to licensees for demonstrating that the results of an individual plant's ASME Code examinations are consistent with the underlying technical basis.

In SECY-06-0124, dated May 26, 2006, the NRC staff presented a rulemaking plan to the Commission to amend fracture toughness requirements for PWRs. In this SECY paper, the NRC staff proposed four options for rulemaking. The NRC staff recommended Option 3, which would allow licensees to voluntarily implement the less restrictive screening limits based on the updated technical basis and insert the updated embrittlement correlation into 10 CFR 50.61 to maintain regulatory consistency and implement the best state-of-the-art embrittlement correlation in both 10 CFR 50.61 and 10 CFR 50.61a. This recommendation was based on providing the necessary relief to licensees that would otherwise expend considerable resources to justify continued plant operation beyond the screening criteria in 10 CFR 50.61 (via compensatory actions, plant-specific analyses, annealing or exemption), while also requiring all licensees to recalculate their embrittlement metric to ensure that all plants' analyses are consistent.

In a Staff Requirements Memorandum (SRM) dated June 30, 2006, the Commission approved the initiation of the rulemaking as specified in Option 2 of the rulemaking plan. This option would require licensees to continue to meet the requirements of 10 CFR 50.61, which provides adequate protection against PTS events, without implementing the updated embrittlement correlation. For licensees whose reactor vessels do not meet the requirements of 10 CFR 50.61, Option 2 would allow licensees to voluntarily implement 10 CFR 50.61a which utilizes the less restrictive screening limits based on the updated technical basis as well as the updated embrittlement correlation. Accordingly, the proposed rule provides for a voluntary alternative to the current set of PTS requirements for any PWR licensee. The NRC considered requiring new plants to use the best available embrittlement correlation (i.e., the embrittlement correlation developed for the new rule). The NRC believes that such a requirement was not necessary to provide adequate protection of public health and safety. The NRC believes that imposing the existing 10 CFR 50.61, without modification, on new reactors would ensure that adequate protection concerns would be met. The NRC believes that the proposed rule's requirements should be a voluntary alternative available to new plants, if needed.

In implementing the rulemaking plan, the proposed rule would provide a new section, 10 CFR 50.61a, for the new set of fracture toughness requirements. The NRC decided that providing a new section containing the updated screening criteria and updated embrittlement correlations would be appropriate because the Commission directed the NRC staff to prepare a rulemaking which would allow current PWR licensees to implement the new requirements of 10 CFR 50.61a or continue to comply with the current requirements of 10 CFR 50.61. Alternatively, the NRC could have revised 10 CFR 50.61 to include the new requirements, which could be implemented as an alternative to the current requirements. However, providing two sets of requirements within the same regulatory section was considered confusing and/or ambiguous as to which requirements apply to which licensees. The proposed rule would provide a voluntary alternative to the current rule, which further prompted the NRC to keep the current, mandatory requirements separate from the new, voluntarily-implemented requirements. As a result, the proposed new rule would retain the current requirements in 10 CFR 50.61 for PWR licensees choosing not to implement the less restrictive screening limits, and would present new requirements in 10 CFR 50.61a as a voluntary relaxation for any PWR licensee.

Section 50.61 contains the current requirements for pressurized thermal shock screening limits and embrittlement correlations. Paragraph (b) of this section would be modified to reference the proposed new section, § 50.61a, as a voluntary alternative to compliance with the requirements of § 50.61. No changes are made to the current pressurized thermal shock screening criteria, embrittlement correlations, or any other related requirements in this section.

Proposed new § 50.61a would contain pressurized thermal shock screening limits based on updated probabilistic fracture mechanics analyses. This new section would provide similar requirements to that of § 50.61, fracture toughness requirements for protection against pressurized thermal shock events for pressurized water nuclear power reactors. However, § 50.61a would differ extensively in how the licensee determines the resistance to fractures initiating from different flaws at different locations in the vessel beltline, as well as in the fracture toughness screening criteria. The proposed rule would require quantifying PTS reference temperatures (RT MAX-X ) for flaws along axial weld fusion lines, plates, forgings, and circumferential weld fusion lines, and comparing the quantified value against the RT MAX-X screening criteria. Although comparing quantified values to the screening criteria is also required by the current § 50.61, the proposed § 50.61a would provide screening criteria that vary depending on material product form and vessel wall thickness. Further, the embrittlement correlation and the method of calculation of RT MAX-X values in § 50.61a would differ significantly from that in § 50.61 as described in the technical basis for this rule. The new embrittlement correlation was developed using multivariable surface-fitting techniques based on pattern recognition, understanding of mechanisms, and engineering judgement. The embrittlement database used for this analysis was derived primarily from the Power Reactor Embrittlement Data Base (PR-EDB) developed at Oak Ridge National Laboratory. The updated RT MAX-X estimation procedures provide a more realistic (compared to the existing regulation) method for estimating the fracture toughness of reactor vessel materials over the lifetime of the plant.

Paragraph (a) would contain definitions for terms used in § 50.61a. It would also provide that terms defined in § 50.61 also have the same meaning in § 50.61a unless otherwise noted.

Paragraph (b) would describe the applicability of § 50.61a to PWRs as an alternative to the requirements of § 50.61. The requirements of this section would provide a voluntarily-implemented alternative to the current requirements of § 50.61 for any current PWR licensee or future holder of a PWR operating license or combined license.

Paragraph (c) would set forth the requirements governing NRC approval of a licensee's use of § 50.61a. The licensee would make the formal request to the NRC via a license amendment, and only upon approval of the license amendment by the NRC would a licensee be permitted to implement § 50.61a. In the licensee's amendment request, the required information would include (a) calculating the values of RT MAX-X values as required by paragraph (c)(1), (b) examining and assessing flaws discovered by ASME Code inspections as required by paragraph (c)(2), and (c) comparing the RT MAX-X values against the applicable screening criteria as required by paragraph (c)(3). In doing so, the licensee would also be required to utilize paragraphs (e)(1) through (e)(3), paragraph (f), and paragraph (g) in order to perform the necessary calculations, comparisons, examinations, assessments, and analyses.

Paragraph (d) would define the requirements for subsequent examinations and flaw assessments after initial approval to use § 50.61a has been obtained under the requirements of paragraph (c). It would also define the required compensatory measures or analyses to be taken if a licensee determines that the screening criteria will be exceeded. Paragraph (d)(1) would define the requirements for subsequent RT MAX-X assessments consistent with the requirements of paragraphs (c)(1) and (c)(3). Paragraph (d)(2) would define the requirements for subsequent examination and flaw assessments utilizing the requirements of paragraphs (e)(1), (e)(1)(i), (e)(1)(ii), (e)(2), and (e)(3). Paragraphs (d)(3) through (d)(7) would define the requirements for implementing compensatory measures or plant-specific analyses should the value of RT MAX-X be projected to exceed the PTS screening criteria in Table 1 of this section.

Paragraph (e) would define the requirements for verifying that the PTS screening criteria in § 50.61a are applicable to a particular reactor vessel. The proposed rule would require that verification be based on an analysis of test results from ultrasonic examination of the reactor vessel beltline materials required by Section XI of the ASME Code.

Paragraph (e)(1) would establish cumulative limits on flaw density and size within the ASME Code, Section XI, Appendix VIII, Supplement 4 inspection volume, which corresponds to a depth of approximately one inch from the clad-to-base metal interface. The allowable number of flaws provided in Tables 2 and 3 are cumulative values. If flaws exist in larger increments, the allowable number of flaws is the value in Table 2 or 3 for that increment minus the total number of flaws in all larger increments. Flaws in this inspection volume contribute approximately 97-99 percent to the TWCF at the screening limit.

Paragraph (e)(1)(i) would describe the flaw density limits for welds.

Paragraph (e)(1)(ii) would describe the flaw density limits for plates and forgings.

Paragraph (e)(1)(iii) would describe the specific ultrasonic examination and neutron fluence information to be submitted to the NRC. The NRC would utilize this information to evaluate whether plant-specific information gathered in accordance with this rule suggests that the NRC staff should generically re-examine the technical basis for the rule.

Paragraph (e)(2) would require that licensees verify that no clad-base metal interface flaws within the ASME Code, Section XI, Appendix VIII, Supplement 4 inspection volume open to the vessel inside surface. These types of flaws could have a substantial effect on the TWCF.

Paragraph (e)(3) would establish limits on flaw density and size beyond the ASME Code, Section XI, Appendix VIII, Supplement 4 inspection volume to three-eights of the reactor vessel thickness from the interior surface. Flaws in this inspection volume contribute approximately 1-3 percent to the TWCF at the screening criteria. Flaws exceeding this limit could affect the TWCF. Flaws greater than three-eights of the reactor vessel thickness from the interior surface do not contribute to the TWCF at the screening limit.

Paragraph (e)(4) would establish requirements to be met if flaws exceed the limits in (e)(1) and (e)(3) or open to the inside surface of the reactor vessel. This section requires an analysis to demonstrate the reactor vessel would have a TWCF of less than 1 × 10 −6 per reactor-year. The analysis could be a complete, plant-specific, probabilistic fracture mechanics analysis or could be a simplified analysis of flaw size, location and embrittlement to demonstrate that the actual flaws in the reactor vessel are not in locations that would cause the TWCF to be greater than 1 × 10 −6 per reactor-year. This paragraph would be required to be implemented if the requirements of (e)(1) through (e)(3) are not satisfied.

Paragraph (e)(5) would describe the critical parameters to be addressed if flaws exceed the limits in (e)(1) and (e)(3) or if the flaws would open to the inside surface of the reactor vessel. This paragraph would be required to be implemented if the requirements of (e)(1) through (e)(3) are not satisfied.

Paragraph (f) would define the process for calculating RT MAX-X values. These values would be based on the vessel's copper, manganese, phosphorus, and nickel weight percentages, reactor cold leg temperature, and neutron flux and fluence values, as well as the unirradiated RT NDT of the product form in question.

Paragraph (g) would provide the necessary equations and variables required by paragraph (f) of this section.

Table 1 would provide the PTS screening criteria for comparison with the licensee's calculated RT MAX-X values. Tables 2 and 3 would provide values to be used in paragraph (e) of this section. Tables 4 and 5 would provide values to be used in paragraph (f) of this section.

Under the “Policy Statement on Adequacy and Compatibility of Agreement States Programs,” approved by the Commission on June 20, 1997, and published in the Federal Register (62 FR 46517; September 3, 1997), this rule is classified as compatibility category “NRC.” Agreement State Compatibility is not required for Category “NRC” regulations. The NRC program elements in this category are those that relate directly to areas of regulation reserved to the NRC by the Atomic Energy Act or the provisions of Title 10 of the Code of Federal Regulations (10 CFR). Although an Agreement State may not adopt program elements reserved to NRC, it may wish to inform its licensees of certain requirements via a mechanism that is consistent with the particular State's administrative procedure laws, but does not confer regulatory authority on the State.

The following table lists documents relating to this rulemaking which are available to the public and how they may be obtained.

Public Document Room (PDR). The NRC's Public Document Room is located at the NRC's headquarters at 11555 Rockville Pike, Rockville, MD 20852.

NRC's Electronic Reading Room (ERR). The NRC's electronic reading room is located at http://www.nrc.gov/reading-rm.html.

The Presidential memorandum dated June 1, 1998, entitled “Plain Language in Government Writing” directed that the Government's writing be in plain language. This memorandum was published on June 10, 1998 (63 FR 31883). The NRC requests comments on the proposed rule specifically with respect to the clarity and effectiveness of the language used. Comments should be sent to the address listed under the ADDRESSES caption of the preamble of this document.

The National Technology Transfer and Advancement Act of 1995, Pub. L. 104-113, requires that Federal agencies use technical standards that are developed or adopted by voluntary consensus standards bodies unless using such a standard is inconsistent with applicable law or is otherwise impractical.

The NRC considered using American Society for Testing and Materials (ASTM) standard E-900, “Standard Guide for Predicting Radiation-Induced Temperature Transition Shift in Reactor Vessel Materials. This standard contains a different embrittlement correlation than that of this proposed rule. However, the correlation developed by RES has been more recently calibrated to available data. As a result, ASTM standard E-900 is not a practical candidate for application in the technical basis for the proposed rule because it does not represent the broad range of conditions necessary to justify a revision to the regulations.

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requirements are utilized as part of the volumetric examination analysis requirements of the proposed rule. ASTM Standard Practice E 185, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels” is incorporated by reference in 10 CFR 50 Appendix H and utilized to determine 30-foot-pound transition temperatures. These standards were selected for use in the proposed rule based on their use in other regulations within Part 50 and their applicability to the subject of the desired requirements.

The NRC will consider using other voluntary consensus standards if appropriate standards are identified.

The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and, therefore, an environmental impact statement is not required. The basis for this determination is as follows:

This environmental assessment focuses on those aspects of § 50.61a where there is a potential for an environmental impact. The NRC has concluded that there will be no significant radiological environmental impacts associated with implementation of the rule requirements for the following reasons:

(1) Section 50.61a would maintain the same functional requirements for the facility as the existing PTS rule in § 50.61 as a voluntary alternative to the existing rule. This proposed rule would establish screening criteria, limiting levels of embrittlement beyond which operation cannot continue without further plant-specific evaluation or modifications, as well as require calculation of the maximum embrittlement predicted at the end of the licensed period of operation. The screening criteria provide reasonable assurance that licensees operating below (predicted embrittlement less than) the screening criteria could endure a pressurized thermal shock event without fracture of vessel materials, thus assuring integrity of the reactor pressure vessel.

(2) The new rule is risk-informed and in accordance with the NRC's 1995 PRA policy statement and risk-informed regulation guidance. Sufficient safety margins are maintained to ensure that any potential increases in core damage frequency (CDF) and large early release frequency (LERF) resulting from implementation of § 50.61a are negligible.

The action will not significantly increase the probability or consequences of accidents, result in changes being made in the types of any effluents that may be released off site, or result in a significant increase in occupational or public radiation exposure. Therefore, there are no significant radiological environmental impacts associated with this action.

With regard to potential nonradiological impacts, implementation of the rule requirements has no impact on the facility other than to provide a more realistic method of calculating PWR vessel fracture toughness with associated limits. Nonradiological plant effluents are not affected and there are no other environmental impacts. Therefore, the NRC concludes that there are no significant environmental impacts associated with the action.

As an alternative to the rulemaking described above, the NRC considered not taking the action (i.e., the “no-action” alternative). Not adopting the more realistic and less conservative regulation would result in no change in environmental impacts for current PWRs or those that would be expected for future PWRs under 10 CFR 50.61.

The NRC staff developed the proposed rule and this environmental assessment. Under the NRC's stated policy, a copy of this environmental assessment will be provided to the state liaison officials as part of the publication of the proposed rule for public comment.

On the basis of this environmental assessment, the NRC concludes that the action would not have a significant effect on the quality of the human environment. Accordingly, the NRC has determined not to prepare an environmental impact statement for the action.

The determination of this environmental assessment is that no significant offsite impact to the public from this action would occur. However, the general public should note that the NRC is seeking public participation. Comments on any aspect of the environmental assessment may be submitted to the NRC as indicated under the ADDRESSES heading.

The NRC has sent a copy of this proposed rule to every State Liaison Officer and requested their comments on the environmental assessment.

This proposed rule would contain new or amended information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501, et seq. ). This proposed rule has been submitted to the Office of Management and Budget for review and approval of the information collection requirements.

Type of submission, new or revision: Revision.

The title of the information collection: 10 CFR part 50, “Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events (10 CFR 60.61 and 50.61a)” proposed rule.

The form number if applicable: Not applicable.

How often the collection is required: Collections would be initially required for PWR licensees utilizing the requirements of 10 CFR 50.61a as a voluntary alternative to the requirements of 10 CFR 50.61. Collections would also be required, after voluntary implementation of the new § 50.61a, when any change is made to the design or operation of the facility that affects the calculated RT MAX-X value. Collections would also be required during the scheduled periodic ultrasonic examination of beltline welds.

Who will be required or asked to report: Any PWR licensee voluntarily utilizing the requirements of 10 CFR 50.61a in lieu of the requirements of 10 CFR 50.61 would be subject to all of the proposed requirements in this rulemaking.

An estimate of the number of annual responses: 2.

The estimated number of annual respondents: 1.

An estimate of the total number of hours needed annually to complete the requirement or request: 264 hours (24 hours annually for recordkeeping plus 240 hours annually for reporting).

Abstract: The NRC is proposing to amend its regulations to provide updated fracture toughness requirements for protection against pressurized thermal shock (PTS) events for pressurized water reactor (PWR) pressure vessels. The proposed rule would provide new PTS requirements based on updated analysis methods. This action is necessary because the existing requirements are based on unnecessarily conservative probabilistic fracture mechanics analyses. This action would reduce regulatory burden for licensees, specifically those licensees that expect to exceed the existing requirements before the expiration of their licenses. These new requirements would be voluntarily utilized by any PWR licensee as an alternative to complying with the existing requirements.

The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collections contained in this proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility?

2. Estimate of burden?

3. Is there a way to enhance the quality, utility, and clarity of the information to be collected?

4. How can the burden of the information collection be minimized, including the use of automated collection techniques?

A copy of the OMB clearance package may be viewed free of charge at the NRC Public Document Room, One White Flint North, 11555 Rockville Pike, Room O-1 F21, Rockville, MD 20852. The OMB clearance package and rule are available at the NRC worldwide Web site: http://www.nrc.gov/public-involve/doc-comment/omb/index.html for 60 days after the signature date of this notice.

Send comments on any aspect of these proposed information collections, including suggestions for reducing the burden and on the above issues, by November 2, 2007 to the Records and FOIA/Privacy Services Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail to INFOCOLLECTS@NRC.GOV and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management and Budget, Washington, DC 20503. Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given to comments received after this date. You may also comment by telephone at (202) 395-3087.

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control number.

The Commission has prepared a draft regulatory analysis on this proposed regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The Commission requests public comments on this draft regulatory analysis. Availability of the regulatory analysis is provided in Section IV. Comments on the draft regulatory analysis may be submitted to the NRC as indicated under the ADDRESSES heading of this document.

In addition, the Commission also requests public comments on the cost and benefit of requiring PWR licensees to revise their vessel analyses if the updated embrittlement correlation were imposed in 10 CFR 50.61. This would differ from the proposed rule, which leaves the technical content of 10 CFR 50.61 unchanged.

Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the Commission certifies that this rule would not, if promulgated, have a significant economic impact on a substantial number of small entities. This proposed rule would affect only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of “small entities” set forth in the Regulatory Flexibility Act or the size standards established by the NRC (10 CFR 2.810).

The NRC has determined that the requirements in this proposed rule do not constitute backfitting as defined in 10 CFR 50.109(a)(1). Therefore, a backfit analysis has not been prepared for this proposed rule.

The requirements of the current PTS rule, 10 CFR 50.61, would continue to apply to all PWR licensees, and would not change as a result of this proposed rule. The requirements of the proposed PTS rule, 10 CFR 50.61a, would not be required, but could be voluntarily utilized, by any PWR licensee. Licensees choosing to implement the proposed PTS rule would be required to comply with its requirements as a voluntary alternative to complying with the requirements of the current PTS rule. Because the proposed PTS rule would not be mandatory for any PWR licensee, but rather could be voluntarily implemented by any PWR licensee, the NRC finds that this amendment would not constitute backfitting.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 553; the NRC is proposing to adopt the following amendments to 10 CFR part 50.

The Commission believes that it is prudent for nuclear power plant designers to take into account the potential effects of the impact of a large, commercial aircraft. The Commission has determined that the impact of a large, commercial aircraft is a beyond-design-basis event and has chosen an approach consistent with NRC's previous approach to such events. The overriding objective of this rule is to require nuclear power plant designers to perform a rigorous assessment of design and other features that could provide additional inherent protection to avoid or mitigate, to the extent practicable, the effects of an aircraft impact, with reduced reliance on operator actions. In this manner, this rule would result in newly designed power reactor facilities being more inherently robust with regard to a potential aircraft impact than if they were designed in the absence of this rule. This rule thus provides an enhanced level of protection beyond that which is provided by the existing adequate protection requirements, which all operating power reactors are required to meet, and which would be provided by the proposed adequate protection requirements that the facilities will be required to meet when finalized (see the proposed 10 CFR part 73, “Physical Protection of Plants and Materials,” power reactor security requirements (71 FR 62663; October 26, 2006)).

The proposed rule would require applicants for new standard design certifications that do not reference a standard design approval; new standard design approvals; combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and new manufacturing licenses that do not reference a standard design certification or standard design approval, and those applicants with applications pending on the effective date of this rule (relevant applicants), to perform an aircraft impact assessment of the effects on the designed facility of the impact of a large, commercial aircraft. Based on the insights derived from that assessment, the application would have to include a description and evaluation of the design features, functional capabilities, and strategies to avoid or mitigate the effects of an aircraft impact, addressing core cooling capability, containment integrity and spent fuel pool integrity. The applicant would be required to describe how such design and other features avoid or mitigate, to the extent practicable, the aircraft impact effects with reduced reliance on operator actions.

The Commission has determined that the impact of a large, commercial aircraft is a beyond-design-basis event. For this reason, the Commission-approved final design basis threat (DBT) does not include an aircraft attack. The NRC published its final DBT rule, Title 10, Section 73.1, “Purpose and Scope,” of the Code of Federal Regulations (10 CFR 73.1), in the Federal Register on March 19, 2007 (72 FR 12705). Two well-established bases support the exclusion of aircraft attacks from the DBT. First, it is not reasonable to expect a licensee with a private security force using weapons legally available to it to be able to defend against such an attack. Second, such an act is in the nature of an attack by an enemy of the United States. Power reactor licensees are not required to design their facilities or otherwise provide measures to defend against such an attack, as provided by 10 CFR 50.13, “Attacks and Destructive Acts by Enemies of the United States; and Defense Activities.”

The Commission has addressed aircraft attacks by regulatory means other than the DBT rule in 10 CFR 73.1. By Order dated February 25, 2002 (Interim Compensatory Measures (ICM) Order), the Commission required all operating power reactor licensees to develop and adopt mitigative strategies to cope with large fires and explosions from any cause, including beyond-design-basis aircraft impacts (67 FR 9792; March 4, 2002). The Commission has proposed incorporating the continuing requirement to provide for such mitigative measures in the NRC's regulations in the proposed 10 CFR part 73 power reactor security requirements, specifically the proposed revisions to 10 CFR 73.55, “Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage,” and Appendix C, “Licensee Safeguards Contingency Plans,” to 10 CFR part 73. If these requirements, which are promulgated on the basis of adequate protection of public health and safety and common defense and security, are finalized, all current and future power reactors must satisfy them.

The current requirements, in conjunction with the currently proposed revisions to the security regulations in 10 CFR 73.55 and Appendix C to 10 CFR part 73, will continue to provide adequate protection of the public health and safety and the common defense and security. Nevertheless, the Commission has decided to require relevant applicants to evaluate possible additional features to avoid or mitigate the effects of an aircraft impact beyond satisfying the current regulations and the proposed 10 CFR part 73 regulations (assuming they become final).

The Commission's DBT requirements (both orders and existing rules) are based on adequate protection of the public health and safety and common defense and security. As such, they are excepted from the cost-benefit analysis that otherwise would be required under 10 CFR 50.109, “Backfitting.” This new proposed rule to address the capability of newly designed power reactors relative to a potential aircraft impact is based both on enhanced public health and enhanced safety and common defense and security but is not necessary for adequate protection. Rather, it would be to enhance the facility's inherent robustness.

Requiring applicants for new reactor designs to perform a rigorous aircraft impact assessment and describe design features to address the effects of a beyond-design-basis aircraft impact is consistent with the NRC's historic approach to beyond-design-basis events and is consistent with the NRC's position in its “Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants” (50 FR 32138; August 8, 1985) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML003711521). The policy statement notes, “The Commission expects that vendors engaged in designing new standard [or custom] plants will achieve a higher standard of severe accident safety performance than their prior designs.” The NRC reiterated that regulatory approach in its “Policy Statement on the Regulation of Advanced Nuclear Power Plants,” dated July 8, 1986 (ADAMS Accession No. ML051660651), “The Commission expects that advanced reactors would provide more margin prior to exceeding safety limits and/or utilize simplified, inherent, passive, or other innovative means to reliably accomplish their safety functions.” This regulatory approach has demonstrated its success, as all designs subsequently submitted to and certified by the Commission represent substantial improvement in safety for operational events and accidents. Therefore, the NRC is proposing to require applicants for newly designed facilities to assess the effects of an aircraft impact on the designed facility.

The Commission considered the appropriate location for requirements on an aircraft impact assessment during its deliberations on the security assessment rulemaking (10 CFR 73.62) proposed by the NRC staff in SECY-06-0204, “Proposed Rulemaking—Security Assessment Requirements for New Nuclear Power Reactor Designs (RIN 3150-AH92),” dated September 26, 2006 (ADAMS Accession No. ML062300068). In its Staff Requirements Memorandum on SECY-06-0204, dated April 24, 2007 (ADAMS Accession No. ML071140119), the Commission disapproved the staff's recommended rulemaking as described in SECY-06-0204. The Commission directed the NRC staff to include the aircraft impact assessment requirements in 10 CFR part 52, “Licenses, Approvals, and Certifications for Nuclear Power Plants,” (72 FR 49352, August 28, 2007) to encourage reactor designers to incorporate practicable measures at an early stage in the design process. This proposed rule is the result of that effort.

This proposed rule would revise 10 CFR part 52 to require applicants for new standard design certifications that do not reference a standard design approval; new standard design approvals; combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and new manufacturing licenses that do not reference a standard design certification or standard design approval to assess aircraft impact assessments and describe the design features and other means to avoid or mitigate, to the extent practicable, the effects of an aircraft impact with reduced reliance on operator actions. This proposed rule renders as duplicative and therefore unnecessary the staff's proposed 10 CFR 73.62 rule to require security assessments. That rule would have required a security assessment which would include mitigation of large fires and explosions, a target set analysis, and design features to protect target sets against DBTs. The large fires and explosions provisions of that rule would be subsumed by this proposed 10 CFR part 52 aircraft impact rule. Sufficient target set provisions are included in the NRC's proposed changes to 10 CFR 73.55, which applicants for new facilities would have to satisfy if that rule is made final. Designers of new facilities are encouraged to account for the provisions of 10 CFR 73.55 in the facility design so as to minimize more costly, post-design features to meet those requirements. Accordingly, the proposed 10 CFR 73.62 is not necessary because the Commission is proposing a new 10 CFR part 52 aircraft impact assessment rule.

In contrast to its relation to a possible security assessment rule, however, the new 10 CFR part 52 aircraft impact assessment rule would complement the proposed revisions to 10 CFR 73.55 and Appendix C to 10 CFR part 73, to mitigate the effects of large fires and explosions. The proposed 10 CFR 73.55 and Appendix C to 10 CFR part 73 provisions on mitigating large fires and explosions codify the adequate protection requirement imposed on existing operating reactors by ICM Order, Item B.5.b. These provisions of the proposed 10 CFR part 73 rule, therefore, are necessary for adequate protection and must remain in regulations  1 that are applicable to all currently operating reactors and must be satisfied by all newly licensed reactors.

Regarding large fires and explosions, which are two likely effects of an aircraft impact, the proposed Appendix C to 10 CFR part 73 is limited to mitigative strategies “using existing or readily-available resources,” which effectively can be adapted to existing facilities. For example, certain facility features might be virtually cost-free if designed into the facility (e.g., spatially diverse containment penetrations) but effectively impossible to retrofit. Thus, strategies that were not required by ICM Order, Item B.5.b, and are not required by proposed 10 CFR 73.55 and Appendix C to 10 CFR part 73 because they do not use existing or readily available resources, might be implemented in new reactor designs because of the aircraft impact assessment rule. The proposed aircraft impact assessment rule does not deal with a design basis event but would result in facilities with additional features to avoid or mitigate the effects of an aircraft impact because they would be designed into the facility. Such features would reduce reliance on operator actions to cope with an aircraft impact event. Thus, this proposed rule is not necessary for adequate protection, but rather is an enhancement that will result in newly designed facilities being more inherently robust against aircraft impacts than the facilities not subject to this proposed rule.

In contrast to the adequate protection requirements of proposed 10 CFR 73.55 and Appendix C to 10 CFR part 73, a proposed rule that would enhance safety and security by requiring an evaluation of newly designed facilities to avoid or mitigate the effects of aircraft impacts is appropriate for inclusion in 10 CFR part 52. The NRC is therefore proposing to require applicants for newly designed reactors, which will have to satisfy the revised 10 CFR part 73 provisions if the rule is made final, to also perform an aircraft impact assessment under this proposed rule. However, the NRC expects that, compared to a licensee for a facility that was not designed to meet the requirements of the proposed rule, licensees for facilities that are designed to comply with the proposed rule would have much less of a need to develop specific procedures, guidance, or other strategies to cope with the loss of large areas of the plant due to explosions or fires in order to comply with the requirements in the proposed 10 CFR 73.55 and Appendix C to 10 CFR part 73. The Commission sees this as a significant benefit of this proposed rule; namely, that features to avoid or mitigate the effects of an aircraft impact are designed into the facility and will result in much less reliance on operator actions for such protection.

Consideration of a rule to require applicants for newly designed reactors to perform an aircraft impact assessment and describe design and other features addressing such impacts, which are beyond-design-basis scenarios, is similar to the Commission's consideration in the mid-1980's of new rules addressing accidents more severe than design basis accidents. The 1985 “Policy Statement on Severe Reactor Accidents” explained the Commission's conclusion that, although it was proposing criteria to show new reactor designs to be acceptable for severe accident concerns, then-existing plants posed no undue risk to public health and safety, and thus, there was no need for action on operating reactors based on severe accident risks. The Commission's reasoning in the severe accident context supports its conclusion that although new reactor designs should be assessed for aircraft impacts and designed to avoid or mitigate the effects of an aircraft impact, existing reactors and designs provide adequate protection of the public health and safety and common defense and security.

The Commission has determined that the existing designs of currently operating nuclear power plants, together with the security program actions mandated by the NRC's orders (some of which are codified in the NRC's final DBT rulemaking and others of which are being incorporated into other NRC regulations), as well as the protection provided by other Federal, State, and local entities, provide an adequate level of protection to public health and safety and common defense and security against aircraft impacts. As a result of the events of September 11, 2001, the NRC has undertaken a series of actions to provide continued reasonable assurance of adequate protection to public health and safety and common defense and security at the United States commercial nuclear power facilities. The NRC has assessed the potential vulnerabilities of operating nuclear power reactors to aircraft impact, and it has issued orders and provided associated guidance to licensees for implementing a range of mitigative strategies. The results of these aircraft impact assessments were derived from detailed calculations of plant damage mechanisms (e.g., structural failures, shock and vibration effects, and fire effects). The NRC ensured that implementation of the February 25, 2002, ICM Order included measures to mitigate such scenarios.

The Commission's ICM Order, Item B.5.b, first established the requirement for licensees to implement certain mitigation measures at existing power reactors for these beyond-design-basis events. This requirement was specifically intended to address “losses of large areas of a (reactor) plant due to fires and explosions.” The Commission has since incorporated this requirement into the proposed rulemaking for 10 CFR 73.55 and Appendix C to 10 CFR part 73. Under the proposed 10 CFR part 73 rulemaking, future license applicants must identify and implement mitigative measures similar to those required for currently operating plants.

Most recently, the Commission published a final rule on March 19, 2007 (72 FR 12705), amending the DBT in 10 CFR 73.1. The DBT rule describes general attributes that nuclear power plant licensees must defend against with high assurance. This rulemaking enhanced the DBT by codifying generically applicable security requirements similar to those previously imposed by the Commission's April 29, 2003, DBT Orders.

On the basis of the previous information, the NRC concludes that existing power reactors pose no undue risk to public health and safety or common defense and security from the effects of an aircraft impact based on the Commission's specified aircraft characteristics. Therefore, the NRC is not applying the aircraft impact assessment requirement in this rulemaking to existing operating nuclear power plants.

The Commission has concluded that the proposed rule need not be applied to the four currently approved standard design certifications in Appendices A through D to 10 CFR part 52. 2 Therefore, applicants would not need to recertify these standard designs to meet this proposed rule. This follows from the Commission's determination that the aircraft impact rule is an enhancement above and beyond what is necessary for adequate protection and that the aircraft impact scenario, as previously explained, is a beyond-design-basis event. Just as the currently operating power reactor facilities continue to meet adequate protection requirements and do not need to meet this new aircraft impact rule, so too are the already certified standard designs sufficient to meet adequate protection design requirements. Any reactor facility built to one of these already-certified designs will, of course, have to satisfy all adequate protection requirements applicable to operating power reactors.

The original applicant (or successor in interest of any of the four current standard design certifications) may voluntarily seek to amend the standard design certification to add design features, functional capabilities, or strategies in accordance with the requirements of proposed 10 CFR 52.500, “Aircraft Impact Assessment.” The NRC encourages voluntary enhancement by the applicants for the four current standard design certifications because it will increase the already high levels of safety and security provided by these reactor designs. Applicants may implement these design modifications in different ways, including:

• Applications to amend the existing design certifications.

• Application for a new design certification based on the existing certification, but containing the design features, functional capabilities, and strategies identified as a result of this rule.

• Requests submitted by combined license applicants for plant-specific departures from the standard design, where the departure implements the modifications developed by the original design certification applicant's voluntary implementation of the provisions of the proposed 10 CFR 52.500 (these requests may be submitted by each individual combined license applicant, or they may be submitted by a group of combined license applicants under the provisions of Appendix N, “Standardization of Nuclear Power Plant Designs: Combined Licenses to Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites,” to 10 CFR part 52 and subpart D, “Additional Procedures Applicable to Proceedings for the Issuance of Licenses to Construct and/or Operate Nuclear Power Plants of Identical Design at Multiple Sites,” to 10 CFR part 2, “Rules of Practice for Domestic Licensing Proceedings and Issuance of Orders.”

The NRC's proposed rulemaking does not require updating of the assessment of aircraft impacts required by proposed 10 CFR 52.500 as part of an application for either a renewed design certification under 10 CFR 52.57, “Application for Renewal,” a renewed combined license under 10 CFR 52.107, “Application for Renewal,” and 10 CFR part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” or a renewed manufacturing license under 10 CFR 52.177, “Application for Renewal.” The NRC's requirement for assessment of large, commercial aircraft impacts is not an aging-related matter, nor is it based on time-limited considerations. Hence, aircraft impacts under the proposed rule are outside the scope of any combined license renewal proceeding under 10 CFR part 54 and combined license holders do not need to update the assessment required by 10 CFR 52.500(b) at the license renewal stage. Similarly, aircraft impacts under the proposed rule are outside the scope of any manufacturing license renewal proceeding under 10 CFR 52.177.

Under this proposed rule, relevant applicants for newly designed power reactors would be required to undertake the following:

• Perform an assessment of the effects on the designed facility of a beyond-design-basis aircraft impact

• Evaluate potential design features, functional capabilities, and strategies for avoiding or mitigating the effects of a beyond-design-basis aircraft impact on the key safety functions of the facility

• Describe how such design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions

The proposed rule is based on the premise that it is desirable for future power reactors to avoid or mitigate the effects of the applicable aircraft impact through design features that reduce or eliminate the need for operator actions. Because this type of consideration needs to occur during the development of the design itself, the NRC directs the requirements at plant designers.

The NRC does not expect plant designers to demonstrate that design features alone, without any operator action or mitigative response activity, will practicably avoid or mitigate the effects of the beyond-design-basis aircraft impact. The NRC recognizes that the decision to rely on design features (as opposed to operator action or mitigative strategies) is complex, and often involves a set of trade-offs between competing considerations. The NRC's goal is that the designer implement a rigorous assessment process to ensure that the design process constitutes a reasoned approach for assessing the plant design to identify practicable design or other features that either minimize the effects of, or mitigate, a beyond-design-basis aircraft impact.

Since September 11, 2001, the Commission has used state-of-the art technology to assess the effects of aircraft impacts on nuclear power plants. As part of a comprehensive review of security for NRC-licensed facilities, the NRC conducted detailed, site-specific engineering studies of a limited number of nuclear power plants to assess potential vulnerabilities of deliberate attacks involving large, commercial aircraft. In conducting these studies, the NRC consulted national experts from several Department of Energy laboratories using state-of-the-art structural and fire analyses. The agency also used realistic predictions of accident progression and radiological consequences.

The proposed rule sets forth a general description of the aircraft characteristics that are required to be used to perform the beyond-design-basis aircraft impact assessment. The assessment must be based on the Commission's specified aircraft characteristics used to define the beyond-design-basis impact of a large, commercial aircraft used for long distance flights in the United States, with aviation fuel loading typically used in such flights, and an impact speed and angle of impact considering the ability of both experienced and inexperienced pilots to control large, commercial aircraft at the low altitude representative of a nuclear power plant's low profile.

Beyond these general characteristics, the Commission will specify for plant designers in a Safeguards Information (SGI) guidance document more detailed characteristics of the large, commercial aircraft that should be used in the required assessment. Although the detailed aircraft characteristics will be described in an SGI guidance document and will not be publicly available because of their potential value to terrorists, the description of some of the factors used in selecting the parameters is offered to foster a better understanding of this rulemaking:

1. The aircraft used by the terrorists on September 11, 2001. The staff has reviewed the results of the September 11, 2001, attacks on the World Trade Center and the Pentagon. The NRC has used these reviews in previous studies for operating reactors. The NRC also used these reviews to make its decisions with respect to this rulemaking.

2. Communications with other U.S. Government agencies. Since September 11, 2001, the NRC has worked closely with the Department of Homeland Security, the Department of Defense, and other agencies both to understand their information on terrorist threats and to communicate the NRC's study results.

3. Communications with foreign governments. A number of foreign governments are considering the construction of new nuclear power plants. The NRC is communicating with the regulatory authorities in these countries to understand their requirements and to convey its own results and plans.

4. Evaluations of commercial aircraft. The NRC has studied the types, numbers, and characteristics of commercial aircraft flown in U.S. airspace.

Because this proposed rule is intended to provide added features to avoid or mitigate the effects of a beyond-design-basis event, the choice of aircraft characteristics and the scenario used for this analysis will not be linked to threat assessments or to any evolution of aircraft design. The proposed rule would require that the design-specific impact assessment use the Commission-specified aircraft characteristics as described in proposed 10 CFR 52.500(b). As stated previously, more specific details about the aircraft characteristics specified by the Commission will be contained in a separate guidance document under SGI controls. Because the guidance containing the more detailed aircraft characteristics will be SGI, the document will only be made available to those individuals with a need-to-know and who are otherwise qualified to have access to SGI. Plant designers (including their employees and agents) who meet the Commission's requirements for access to SGI would have access to the guidance document containing these more detailed characteristics in order to perform the assessments required by the proposed rule.

The Commission has carefully balanced the public interest in knowing the characteristics of the specific aircraft to be used in the aircraft impact assessment and the need for meaningful comment on specific details of the aircraft impact assessment. The result is an aircraft impact assessment proposed rule that describes the general aircraft characteristics which applicants are required to use in their aircraft impact assessments. The text of this proposed rule and the associated supplementary information, provide ample information to enable meaningful comment on what the aircraft impact assessment should entail. No additional information is necessary to understand or to comment on the proposed aircraft impact assessment rule. Members of the public can provide the Commission their views on this rulemaking regarding the design areas to be addressed in the assessment, functions to be evaluated for possible enhancement, and criteria for assessing practicability without having access to the more detailed SGI aircraft characteristics. Therefore, access to the proposed SGI aircraft characteristics contained in the regulatory guidance is not necessary to enable meaningful public comment on the proposed aircraft impact assessment rule.

This regulatory approach is similar to that used by the NRC in describing the DBT against which security programs under 10 CFR part 73 must defend. The general characteristics of the DBT appear in 10 CFR 73.1. More detailed information of a sensitive nature is contained in adversary characteristics documents. As is the case with the Commission's aircraft characteristics, the technical bases for these documents derive largely from intelligence information and contain SGI that must be withheld from public disclosure. They are available only on a need-to-know basis to those who are approved for access. In the final DBT rule, the NRC was careful to set forth rule text that does not compromise licensee security, but it also acknowledges the need to keep the public informed of the types of attacks against which nuclear power plants and Category I fuel cycle facilities are required to defend. The NRC is taking a similar approach in this proposed aircraft impact rule. This approach strikes the appropriate balance between public disclosure of the regulatory requirements governing nuclear power plants, and protection of public health and safety and common defense and security.

Because the aircraft impact is a beyond-design-basis event, the methods and acceptance criteria used should be based on realistic assumptions. The aircraft impact assessment would include the items detailed in the following paragraphs:

1. Consideration of aircraft characteristics. The assessment must consider a large, commercial aircraft of the type currently in use for long distance flights in the United States as described previously in this document and in 10 CFR 52.500(b). More detailed characteristics of the large, commercial aircraft to be used in this assessment will be contained in a separate guidance document under SGI controls.

2. Plant functions, structures, systems, components, and locations to be assessed. The critical functions required to be evaluated in the aircraft impact assessment include core cooling, containment integrity, and spent fuel pool integrity. Evaluation of the survivability of these functions, should consider not only the key components, but also power supplies, cable runs, and other components that support these functions. The evaluation may take credit for the availability of both safety and non-safety equipment. The assessment should evaluate whether the structures containing equipment that provides needed functions are likely to be affected by the specified large, commercial aircraft impact. Factors to be considered in the evaluation include the size and location of the structures and the presence of external impediments to impact.

3. Damage mechanisms. The assessment should model the structural response, shock and vibration effects, and fire effects of the postulated aircraft impact.

a. Structural assessment. The structural assessment should be based on a detailed structural model of the plant taking into account the nonlinear materials and geometric behavior. The assessment should consider both local and global (plant-wide) behavior, as well as thermal effects resulting from fire.

b. Shock assessment. The assessment should evaluate the local and global (plant-wide) shock and vibration effects resulting from the postulated impact.

c. Fire assessment. The fire assessment should consider the extent of structural damage and aviation fuel deposition, if any, and spread within the impacted buildings. The assessment should consider both short- and long-term fire effects.

The impact assessment is subject to audit and review by the NRC and, therefore, must be maintained by the applicant along with the rest of the information that forms the basis for the relevant application, consistent with paragraph (b) of 10 CFR 52.0, “Scope; Applicability of 10 CFR Chapter I Provisions,” 10 CFR 50.70, “Inspections,” and 10 CFR 50.71, “Maintenance of Records, Making of Reports.” The applicant does not need to submit the impact assessment—as opposed to the “description and evaluation of the design features, functional capabilities, and strategies” required by proposed 10 CFR 52.500(c)—to the NRC in its application.

Under the proposed rule, the NRC will confirm that the impact assessment was performed consistent with the regulatory requirements and related guidance documents. The NRC may take appropriate enforcement action for any violations of applicable NRC requirements, including, but not limited to, proposed 10 CFR 52.500, 10 CFR 52.4, “Deliberate Misconduct,” and 10 CFR 52.6, “Completeness and Accuracy of Information.” A failure to perform the assessment would be a violation of the rule. The NRC expects the assessment to be rigorous. Any assessment that is inadequate to reasonably assess the aircraft impact; to identify practicable design features, functional capabilities, or strategies; or to justify non-adoption of potentially advantageous design features, functional capabilities, or strategies, could be considered a violation of the rule.

The NRC's decision on an application subject to proposed 10 CFR 52.500 would be separate from any NRC determination that may be made with respect to the adequacy of the impact assessment which the rule does not require be submitted to the NRC. Applicants would only be required to submit a description and evaluation of the design features, functional capabilities, and strategies to avoid or mitigate the effects of the applicable, beyond-design-basis aircraft impact in their final safety analysis report (FSAR) with the understanding that the complete aircraft impact assessment would be available for NRC audit and review at the applicant's offices, if needed. The NRC expects that, generally, the information that it needs to perform its review of the application to assess the applicant's compliance with 10 CFR 52.500 would be that information contained in the applicant's FSAR. Therefore, the adequacy of the impact assessment would not be a matter which may be the subject of a contention submitted as part of a petition to intervene under 10 CFR 2.309, “Hearing Requests, Petitions to Intervene, Requirements for Standing, and Contentions.” A person who seeks NRC rulemaking action with respect to a proposed standard design certification on the basis that the impact assessment is inadequate could submit comments in the notice and comment phase of that rulemaking. A person who seeks rulemaking action after the NRC has adopted a final design certification rule on the basis that the impact assessment performed for that design certification is inadequate could submit a petition for rulemaking under 10 CFR 2.802, “Petition for Rulemaking,” and 10 CFR 2.803, “Determination of Petition,” seeking to amend the standard design certification. A person who seeks agency enforcement-related action on a combined license or manufacturing license on the basis of an inadequate impact assessment could file a petition under 10 CFR 2.206, “Requests for Action Under This Subpart.”

Once the applicant completes the impact assessment, accomplishes the evaluation required by proposed 10 CFR 52.500(c) based on insights gained from the proposed 10 CFR 52.500(b) assessment, and includes the description and evaluation required by proposed 10 CFR 52.500(c) in the FSAR, the purpose of the impact assessment would be achieved. Accordingly, the proposed rule would not require the impact assessment to be updated, by either: (1) The design certification applicant whose application references a design approval, (2) the design certification applicant following the NRC's adoption of a final standard design certification rule, (3) a design approval holder, (4) a manufacturing license applicant or holder whose application references a design certification or design approval, (5) a combined license applicant or holder whose application references a design certification, design approval, or manufactured reactor, or (6) a combined license holder whose application does not reference a design certification, design approval, or manufactured reactor and is required to prepare its own assessment.

The provisions of 10 CFR 50.71(c) require that records that are required by the regulations in 10 CFR part 50, “Domestic Licensing of Production and Utilization Facilities,” or 10 CFR part 52 must be retained for the period specified by the appropriate regulation. If a retention period is not otherwise specified, the licensee must retain these records until the Commission terminates the facility license. Because proposed 10 CFR 52.500(b) would require the performance of the aircraft impact assessment, it falls under the category of “records that are required by the regulations” and therefore, the licensee would be required to retain the assessment until the Commission terminates the facility license. The NRC also expects to add specific provisions to each standard design certification rule for a design covered by proposed 10 CFR 52.500 governing retention of the aircraft impact assessment by both the applicant for the design certification (including an applicant after the Commission has adopted a final standard design certification rule) and a licensee who references that design certification. The NRC expects to require applicants and licensees to retain the assessment required by 10 CFR 52.500(b) throughout the pendency of the application and for the term of the certification or license (including any period of renewal). An example of such requirements can be found in any of the current design certification rules, Section X, “Records and Reporting,” of Appendices A through D of 10 CFR part 52.

As discussed in Section VIII, “Specific Request for Comments,” of this document, the NRC is requesting comments on whether, in lieu of the specific design certification rule provisions or reliance on 10 CFR 50.71(c), it should adopt as part of the final 10 CFR 52.500 rulemaking a specific provision that would explicitly mandate the retention of the assessment. Such a provision, to be included in an additional paragraph of proposed 10 CFR 52.500, would also set forth the proposed period of retention for the assessment as the term of the design certification, combined license, or manufacturing license.

The proposed rule would require designers of new facilities to describe how the design features, functional capabilities, and strategies adopted based on the insights of the aircraft impact assessment avoid or mitigate the effects of the aircraft impact. Plant structures critical to maintaining facility safety functions should be designed, if practicable, such that an impact does not result in structural failure, and aircraft parts and jet fuel do not enter the structures. In circumstances in which an impact results in aircraft parts and jet fuel entering structures or affecting equipment, plant structures and layouts should be evaluated with respect to maintaining key safety functions by addressing equipment survivability following the entry of aircraft parts and jet fuel and key safety functions are accomplished notwithstanding the resulting internal damage resulting from structural loads, shock and vibration, and fire.

As discussed previously, the Commission has issued orders to operating plants requiring mitigation of the effects of losing large areas of the plant from fires and explosions. These requirements include some reliance on operator actions, such as realigning systems to ensure continued core cooling following the loss of a large area. Because this proposed rule would apply to newly designed facilities before construction of the facility, the Commission expects that improvements can be made in the plant's design that have the same result as operator actions credited in operating plants. Thus, these designs should have reduced reliance, relative to current operating plants, on operator actions.

The proposed rule would require applicants to describe how the design features, functional capabilities, and strategies avoid or mitigate, “to the extent practicable,” the effects of the applicable aircraft impact with reduced reliance on operator actions. The NRC intends this standard to include those design features, functional capabilities, and strategies that are realistically and reasonably feasible from a technical engineering perspective. For example, the NRC believes it may be practicable to employ new technologies currently in use in the commercial nuclear power industry or in another industry. Alternatively, it would not be practicable to introduce a design feature that could have adverse safety or security consequences under a different operational or accident scenario. This consideration of practicability allows the designers to evaluate potential competing technical factors, such as the response to earthquakes, while at the same time addressing aircraft impacts.

Nuclear power plants are inherently very robust, secure structures designed to withstand tornadoes, hurricanes, earthquakes, floods, and other severe events. They have redundant and diverse safety equipment so that if an active component becomes unavailable, another component or system will satisfy its function. The results of the Commission's evaluation of postulated aircraft impacts on operating reactors reinforced the value of design features such as the following:

• Reinforced concrete walls

• Redundancy and spatial separation of key systems, structures and components

• Diversity of power supplies

• Compartmentalization of interior structures with pressure resisting concrete walls and doors

The NRC expects the required evaluation to consider the value of such design features and of possible improvements in these and other features. The applicant must base the evaluation on insights gained from the impact assessment performed under proposed 10 CFR 52.500(b).

The NRC will confirm that the evaluation required by 10 CFR 52.500(c) was performed and that the FSAR includes the necessary description and evaluation of the design and other features adopted to avoid or mitigate, to the extent practicable, the potential effects of the applicable, beyond-design-basis aircraft impact. The NRC will review the evaluation contained in the application and reach a conclusion as to whether the applicant has conducted an evaluation reasonably formulated to identify practicable design and other features to avoid or mitigate the potential effects of the applicable, beyond-design-basis aircraft impact. However, NRC's review of the adequacy of the evaluation, and the effectiveness and practicability of the applicant- selected features, capabilities, and strategies, are separate and distinct from the NRC's determination whether to issue a final standard design certification rule, a final design approval, a combined license, or a manufacturing license. Therefore, as is the case with the impact assessment, the NRC will use its established audit and review process to ensure the evaluation and determination of practicability was performed consistent with the regulatory requirements and related guidance documents. The NRC may take appropriate enforcement action for any violation of applicable NRC requirements. Inasmuch as the adequacy of the evaluation and the practicability of the applicant-selected features, capabilities, and strategies, are separate and distinct from the approval of the final design in the design certification, design approval, combined license, or manufacturing license, there would be no issue resolution associated with the assessment regarding the lack of effectiveness or practicability of potential design features, functional capabilities, and strategies not selected by the applicant for inclusion in the certified design.

The NRC is proposing that the design features, functional capabilities, or strategies credited for avoiding or mitigating the effects of an aircraft impact be described in Chapter 19 of the FSAR, which addresses severe accidents. The design features may include structures or features unchanged from the plant design as it existed before the aircraft impact assessment (e.g., an existing wall is found to be effective), structures or features included in the plant design but enhanced to improve the response to an aircraft impact (e.g., an existing wall is made stronger), or new structures or features added solely to address aircraft impacts (e.g., a new wall). The regulatory treatment of the design features (e.g., how changes to the features are controlled) depends on which of the above categories apply. For example, a design feature added specifically to avoid or mitigate the effects of an aircraft impact would be controlled only by requirements specifically for control of those design features added in this proposed rule or requirements that the NRC expects to add to future design certifications that would be subject to proposed 10 CFR 52.500. A safety-related structure credited in the aircraft impact assessment as a design feature would continue to be controlled by Appendix B to 10 CFR part 50, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants;” 10 CFR part 21, “Reporting of Defects and Noncompliance;” and other regulations establishing technical and administrative requirements on the non-aircraft impact functions, in addition to the proposed requirements for control of features to address aircraft impacts.

For combined licenses not referencing a certified design, the NRC is proposing to have change control governed by the requirements in a new 10 CFR 52.502, “Control of Changes to FSAR Information,” to address changes to any design features, functional capabilities, or strategies credited for avoiding or mitigating the effects of an aircraft impact for a combined license that does not reference a certified design. Specifically, the proposed 10 CFR 52.502(c) would require that, if the licensee changes the information required by 10 CFR 52.79(a)(47) to be included in the FSAR, the licensee re-perform that portion of the evaluation required by proposed 10 CFR 52.500(c) that addresses the changed feature, capability, or strategy. The licensee would also be required to describe, in the re-evaluation, how the modified design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions. Because this rule is being proposed to address a beyond-design-basis event, the NRC has determined that it is appropriate to apply the same standard to any licensee-proposed changes to features, capabilities, and strategies that would be applied during the original evaluation of those design features, functional capabilities, and strategies.

A combined license holder subject to proposed 10 CFR 52.500 (i.e., a licensee whose application does not reference a standard design certification, standard design approval, or manufactured reactor) may change the design features, functional capabilities, and strategies incorporated into the design, in accordance with proposed 10 CFR 52.502, without prior NRC review and approval, as long as the licensee re-performs that portion of the evaluation required by proposed 10 CFR 52.500(c) addressing the changed feature, capability, or strategy. The licensee must also describe, in the re-evaluation documented in a change to the FSAR, how the modified design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions. Licensees' submittal of this updated information to the NRC will be governed by the existing FSAR update requirements in 10 CFR 50.71(e).

A design feature, functional capability, or strategy described in a standard design certification may not be changed generically except by notice and comment rulemaking, see 10 CFR 52.63, “Finality of Standard Design Certifications,” paragraphs (a)(1) and (2), and such a change must meet one of the criteria in 10 CFR 52.63(a)(1). All referencing combined licenses must implement any generic change to a design certification rule, as required by 10 CFR 52.63(a)(3).

The NRC expects to add a new change control provision to future design certification rules subject to proposed 10 CFR 52.500 to govern combined license holders referencing the design certification that request a departure from the design features, functional capabilities, or strategies in the referenced design certification. The new change control provision would require that, if the licensee changes the information required by 10 CFR 52.47(a)(28) to be included in the FSAR for the standard design certification, then the licensee must re-perform that portion of the evaluation required by proposed 10 CFR 52.500(c) addressing the changed feature, capability, or strategy. The licensee must also describe, in the re-evaluation documented in a change to the FSAR ( i.e. , a plant-specific departure from the generic design control document), how the modified design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions. Licensees' submittal of this updated information to the NRC will be governed by the reporting requirements in the applicable design certification rule. The NRC expects to continue, in future standard design certification rulemakings, its practice of adopting reporting requirements analogous to Section X.B of the four existing standard design certification rules. Licensees making changes to design features, capabilities, or strategies included in the certified design or in the plant-specific FSAR may also need to develop alternate means to cope with the loss of large areas of the plant from explosions or fires to comply with the requirements in the proposed 10 CFR 73.55 and Appendix C to 10 CFR part 73.

A design feature, functional capability, or strategy described in a standard design approval may not be changed generically except under an application for a new design approval. There are no provisions in 10 CFR part 52 for making generic changes to a standard design approval. Paragraph (a) of Section 52.145, “Finality of Standard Design Approvals; Information Requests,” states that an approved design must be used by and relied upon by the NRC staff and the Advisory Committee on Reactor Safeguards in their review of any individual facility license application that incorporates by reference a standard design approval unless there exists significant new information that substantially affects the earlier determination or other good cause. Therefore, any changes to a design feature, functional capability, or strategy described in a standard design approval would be subject to review by the NRC in any application that references the design approval. Note that 10 CFR 52.131, “Scope of Subpart,” states that the an applicant may submit standard designs for a nuclear power reactor or major portions thereof. To the extent that a standard design approval is issued for only portion of a nuclear power reactor, any applicant referencing that design approval will have to separately comply with the requirements of 10 CFR 52.500 for any portion of the design not addressed in the design approval issued by the NRC.

Under the provisions of 10 CFR 52.171, “Finality of Manufacturing Licenses; Information Requests,” the holder of a manufacturing license may not make changes to the design features, functional capabilities, or strategies described in the FSAR without prior Commission approval. The request for a change to the design must be in the form of an application for a license amendment, and must meet the requirements of 10 CFR 50.90, “Application for Amendment of License, Construction Permit, or Early Site Permit,” and 10 CFR 50.92, “Issuance of Amendment.”

Under the provisions of 10 CFR 52.171(b)(2), a combined license applicant or licensee who references or uses a nuclear power reactor manufactured under a manufacturing license under this subpart may request a departure from the design features, functional capabilities, or strategies described in the FSAR for the manufactured reactor. The Commission will grant such a request only if it determines that the departure will comply with the requirements of 10 CFR 52.7, “Specific Exemptions,” and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure.

Once the evaluation required by proposed 10 CFR 52.500(c) is completed and the application includes descriptions of the design features, functional capabilities, and strategies, the purpose of the evaluation would be largely achieved. Thus, as with the assessment required by proposed 10 CFR 52.500(b), the applicant or licensee would not be required to update the paragraph (c) evaluation after the design certification, design approval, combined license, or manufacturing license is issued, or in an application for renewal under either 10 CFR 52.57, 10 CFR 52.107 and 10 CFR part 54, or 52.177. However, licensees would be required to maintain the paragraph (c) evaluation, inasmuch as proposed 10 CFR 52.79(a)(47) and 10 CFR 52.157(f)(32) require the proposed 10 CFR 52.500(c) evaluation and description to be included in the FSAR portion of the application.

Following issuance of a final design certification rule, the design certification applicant would not be required to update the evaluation so long as it does not request a significant change to any of the design features, functional capabilities, or strategies in the design certification. Similarly, the holder of a combined license or manufacturing license would not be required to update the evaluation so long as the licensee makes no significant change to the design features, functional capabilities, or strategies described in the FSAR.

As with the aircraft impact assessment required by proposed 10 CFR 52.500(b), in accordance with 10 CFR 50.71(c), each combined license holder and manufacturing license holder whose application was subject to proposed 10 CFR 52.500 would be required to retain the documentation supporting the proposed 10 CFR 52.500(c) evaluation for NRC review. With respect to a standard design certification, proposed 10 CFR 52.47(a)(28) would require the proposed 10 CFR 52.500(c) evaluation to be included in the FSAR submitted as part of the design certification application. The NRC acknowledges that the applicant for a standard design certification is not, per se, responsible for maintaining the FSAR information once a final design certification rule is adopted by the NRC. Nonetheless, the NRC continues to believe, for the reasons set forth in the statement of considerations for the first design certification rulemaking, see 62 FR 25800, May 19,1997, at 25813-25814, 25826, that the original standard design certification applicant should be required to maintain the accuracy of the design certification information. Therefore, in future standard design certification rulemakings, the NRC expects to continue its practice of adopting a records management requirement analogous to Section X.A of the four existing standard design certification rules. In addition to the information included in the FSAR for the design certification or combined license, the supporting documentation retained onsite should describe the methodology used in identifying and evaluating the practicability of potential features, capabilities, and strategies for inclusion in the design; and list the features, capabilities, and strategies that were considered but rejected, along with the basis for their rejection.

Section 52.11 identifies the information collection requirements contained in 10 CFR part 52 approved by the Office of Management and Budget (OMB) as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq. ). The NRC is proposing to modify paragraph (b) to include proposed 10 CFR 52.500 in the list of requirements with approved information collections.

Section 52.47 identifies the required technical information to be included in an application for a standard design certification. The proposed rule would revise this section by adding a new paragraph (a)(28) requiring that the FSAR contain the information required by proposed 10 CFR 52.500, “Aircraft Impact Assessment.” This information, as currently set forth in paragraph (c) of proposed 10 CFR 52.500, is limited to the following:

1. A description of the design features, functional capabilities, and strategies credited by the applicant to avoid or mitigate the effects of the applicable, beyond-design-basis aircraft impact; and

2. An evaluation of how such design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions.

The 10 CFR 52.47(a)(28) requirement applies only to those standard design certification applications which are subject to proposed 10 CFR 52.500, that is, those design certifications issued after the effective date of the final rule (see 10 CFR 52.500(a)) that do not reference a design approval. Thus, any standard design certification application not referencing a standard design approval that is docketed and under review by the NRC but has not yet been issued in final form as of the effective date of the final 10 CFR 52.500 must amend its application to include the information required by final 10 CFR 52.500.

Section 52.79 identifies the required technical information to be included in an FSAR submitted in a combined license application under 10 CFR part 52, subpart C, “Combined Licenses.” The proposed rule would revise this section by adding a new paragraph (a)(47) requiring that the FSAR contain the information required by proposed 10 CFR 52.500. This is the same type of information that an applicant for a standard design certification would need to submit, namely, the following:

1. A description of the design features, functional capabilities, and strategies credited by the applicant to avoid or mitigate the effects of the applicable, beyond-design-basis aircraft impact; and

2. An evaluation of how such design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions.

Only those combined licenses issued after the effective date of the final rule that do not reference a standard design certification, standard design approval, or manufactured reactor would be subject to 10 CFR 52.79(a)(47). Thus, a combined license application filed after the effective date of the final 10 CFR 52.500 and referencing a standard design certification, standard design approval, or manufactured reactor would not have to include the information required by 10 CFR 52.500. The NRC notes that this would be true even for a combined license application which references one of the four current standard design certifications (ABWR, 10 CFR part 52, Appendix A; System 80+, 10 CFR part 52, Appendix B; AP600, 10 CFR part 52, Appendix C; and AP1000, 10 CFR part 52, Appendix D). This is consistent with the requirements of 10 CFR 52.79(c), (d), and (e) which state that, if the combined license application references a standard design certification, standard design approval, or manufactured reactor, then the FSAR need not contain information or analyses submitted to the Commission in connection with the design certification, design approval, or manufacturing license, as applicable. By contrast, a combined license applicant not referencing a standard design certification, standard design approval, or manufactured reactor whose application is docketed and under review by the NRC but for which a license has not yet been issued as of the effective date of the final 10 CFR 52.500, must amend its application to include the information required by 10 CFR 52.500.

Section 52.137 identifies the required technical information to be included in an application for a standard design approval. The proposed rule would revise this section by adding a new paragraph (a)(26) requiring that the FSAR contain the information required by proposed 10 CFR 52.500. This information, as currently set forth in paragraph (c) of proposed 10 CFR 52.500, is limited to the following:

1. A description of the design features, functional capabilities, and strategies credited by the applicant to avoid or mitigate the effects of the applicable, beyond-design-basis aircraft impact; and

2. An evaluation of how such design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions.

The 10 CFR 52.137(a)(26) requirement applies only to those standard design approval applications which are subject to proposed 10 CFR 52.500, that is, those design approvals issued after the effective date of the final rule (see 10 CFR 52.500(a)). Thus, any standard design approval application that is docketed and under review by the NRC but has not yet been issued in final form as of the effective date of the final 10 CFR 52.500 must amend its application to include the information required by final 10 CFR 52.500.

Section 52.157 identifies the required technical information to be included in an application for a manufacturing license. The proposed rule would revise this section by adding a new paragraph (f)(32) requiring that the FSAR contain the information required by proposed 10 CFR 52.500. This information, as currently set forth in paragraph (c) of proposed 10 CFR 52.500, is limited to the following:

1. A description of the design features, functional capabilities, and strategies credited by the applicant to avoid or mitigate the effects of the applicable, beyond-design-basis aircraft impact; and

2. An evaluation of how such design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions.

The 10 CFR 52.157(f)(32) requirement applies only to those manufacturing license applications which are subject to proposed 10 CFR 52.500, that is, those manufacturing licenses issued after the effective date of the final rule that do not reference a design certification or design approval (see 10 CFR 52.500(a)). Thus, any manufacturing license application that is docketed and under review by the NRC but has not yet been issued in final form as of the effective date of the final 10 CFR 52.500 must amend its application to include the information required by final 10 CFR 52.500.

Section 52.303 identifies the regulations in 10 CFR part 52 that are not issued under Sections 161b, 161i, or 161o for the purposes of Section 223 of the Atomic Energy Act of 1954, as amended, which provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under Sections 161b, 161i, or 161o of the Act. The NRC is proposing to modify paragraph (b) to include proposed 10 CFR 52.500 and proposed 10 CFR 52.502 in the list of regulations not issued under Sections 161b, 161i, or 161o for the purposes of Section 223 of the Act.

The NRC proposes to add a new subpart K, “Additional Requirements,” to 10 CFR part 52. This subpart would be reserved for requirements applicable only to the licenses, certifications, and approvals under 10 CFR part 52 that have unique characteristics which mitigate against placing them in other parts of 10 CFR Chapter I, “Nuclear Regulatory Commission” of Title 10.

Proposed 10 CFR 52.500 would be a new requirement for assessing a large, commercial aircraft impact at nuclear power plants and incorporating design features, functional capabilities, and strategies to avoid or mitigate, to the extent practicable, the effects of such aircraft impacts.

Paragraph (a) would state that the requirements of this section would be applicable to all standard design certifications issued after the effective date of the final rule that do not reference a standard design approval; standard design approvals issued after the effective date of the final rule; combined licenses issued after the effective date of the final rule that do not reference a standard design certification, standard design approval, or manufactured reactor; and manufacturing licenses issued after the effective date of the final rule that do not reference a standard design certification or standard design approval. A design certification rule issued after the effective date of the final 10 CFR 52.500 rule that does not reference a design approval is subject to the requirements of the rule even if its application was filed before the effective date of the final 10 CFR 52.500 rule. Similarly, a design approval issued after the effective date of the final rule is subject to the requirements of the rule even if its application was filed before the effective date of the final rule. A combined license issued after the effective date of the final 10 CFR 52.500 rule that does not reference a design certification, design approval, or manufactured reactor would be subject to the requirements of the rule, even if its application was filed before the effective date of the final 10 CFR 52.500 rule. As noted earlier in the section-by-section discussion of proposed 10 CFR 52.79(a)(47) of this document, a combined license issued after the effective date of the final 10 CFR 52.500 rule referencing one of the four current standard design certifications, would not be subject to the requirements of proposed 10 CFR 52.500.

Paragraph (b) would require those applicants subject to proposed 10 CFR 52.500 to perform a design-specific assessment of the effects on the designed facility of the impact of a large, commercial aircraft (impact assessment). By “design-specific,” the NRC means that the impact assessment must address the specific design which is either the subject of the standard design certification, standard design approval, combined license, or manufacturing license application. The proposed rule would require that the design-specific impact assessment be based on Commission-specified general aircraft characteristics used to define the beyond-design-basis impact of a large, commercial aircraft used for long distance flights in the United States, with aviation fuel loading typically used in such flights, and an impact speed and angle of impact considering the ability of both experienced and inexperienced pilots to control large, commercial aircraft at the low altitude representative of a nuclear power plant's low profile. Beyond these general characteristics, the Commission will specify for plant designers in an SGI guidance document more detailed characteristics of the large, commercial aircraft to be used in the required assessment. This approach is discussed in more detail in Section V.B of the Supplementary Information of this document. Because the assessment of an aircraft impact is a beyond-design-basis event, the methods and acceptance criteria used in the assessment should be based on realistic assumptions.

Paragraph (c) would require the relevant applications to include a description and evaluation of the design features, functional capabilities, and strategies (features, capabilities, and strategies) to avoid or mitigate the effects of the aircraft impact that applicants must assess under paragraph (b). Design features, functional capabilities, and strategies could include such things as reinforced concrete walls (in the original design, modified, or added); redundancy and spatial separation of key systems, structures and components; diversity of power supplies; and compartmentalization of interior structures. The NRC expects the required assessment to include an evaluation of such features, capabilities, and strategies and of possible improvements in them. The evaluation of such design features, functional capabilities, and strategies must include core cooling capability, containment integrity, and spent fuel pool integrity.

The description of the features, capabilities, and strategies should be equivalent in detail to descriptions of other design features and functional capabilities addressing beyond-design-basis events or severe accidents which are required to be described in the design certification, design approval, combined license, or manufacturing license FSAR. However, the NRC reiterates that the aircraft impact at which these features, capabilities, and strategies are directed is not a design basis event. Therefore, these features, capabilities, and strategies need not meet the “special treatment” requirements  3 applicable to safety-related or important to safety structures, systems, and components.

The paragraph (c) evaluation should be a structured process which would require consideration of the insights gained by the impact assessment performed under proposed 10 CFR 52.500(b) and identify features, capabilities, and strategies which are practicable to include in the facility design. The evaluation should summarize the bases for the applicant's determination that the selected features, capabilities, and strategies incorporated into the facility design avoid or mitigate, to the extent practicable, the effects of the applicable, beyond-design-basis aircraft impact, with reduced reliance on operator actions. As with the impact assessment, the evaluation would address a beyond-design-basis event, and therefore, need not be performed in accordance with the NRC's “special treatment” requirements, such as the quality assurance/quality control requirements in Appendix B of 10 CFR part 50. The proposed 10 CFR 52.500(c) evaluation must be included in the FSAR in accordance with 10 CFR 52.47(a)(28), 10 CFR 52.79(a)(47), 10 CFR 52.137(a)(26), or 10 CFR 52.157(f)(32) and should address only those features, capabilities, and strategies selected by the applicant for inclusion in the plant design. In addition to describing and evaluating the incorporated design and other features, the application must describe how such design and other features, avoid or mitigate, to the extent practicable, the effects of the impact with reduced reliance on operator actions. The NRC intends this standard to include those design features, functional capabilities, and strategies which are realistically and reasonably feasible from a technical engineering perspective. For example, the NRC believes that it may be practicable to employ existing technologies currently in use in the commercial nuclear power industry or in another industry. However, it would not be practicable to introduce a design feature that could have adverse safety or security consequences under a different operational or accident scenario.

Inclusion of any SGI in the evaluation submitted in the FSAR as part of a relevant application must be in accordance with applicable requirements in 10 CFR part 73. The NRC will process and address requests for access to this information from the general public in accordance with the NRC's existing procedures.

Paragraph (a) would state that, for standard design certifications which are subject to proposed 10 CFR 52.500, generic changes to the information required by 10 CFR 52.47(a)(28) to be included in the FSAR are governed by the applicable requirements of 10 CFR 52.63. A design feature, functional capability, or strategy described in a standard design certification may not be changed in the design certification except by notice and comment rulemaking, see 10 CFR 52.63(a)(1) and (2), and such a change must meet one of the criteria in 10 CFR 52.63(a)(1). Any generic change to a design certification rule must be implemented by all referencing combined licenses, as required by 10 CFR 52.63(a)(3).

Paragraph (b) would state that, for combined license applicants or holders which are not subject to proposed 10 CFR 52.500 but reference a standard design certification which is subject to proposed 10 CFR 52.500, proposed departures from the information required by 10 CFR 52.47(a)(28) to be included in the FSAR for the standard design certification are governed by the change control requirements in the applicable design certification rule. A combined license holder referencing a standard design certification subject to proposed 10 CFR 52.500 (i.e., a design certification issued after the effective date of the final 10 CFR 52.500), who seeks to depart from the design features, functional capabilities, or strategies in the referenced design certification, would be governed by a new change control provision that the NRC expects to add to future design certification rules to which proposed 10 CFR 52.500 would apply. The new change control provision would require that, if the licensee changes the information required by proposed 10 CFR 52.47(a)(28) to be included in the FSAR for the standard design certification, then the licensee shall re-perform that portion of the evaluation required by proposed 10 CFR 52.500(c) addressing the changed feature, capability, or strategy, and describe, in the re-evaluation, how the remaining design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions.

Paragraph (c) would state that, for combined licenses which are subject to proposed 10 CFR 52.500, if the licensee changes the information required by 10 CFR 52.79(a)(47) to be included in the FSAR, then the licensee shall re-perform that portion of the evaluation required by 10 CFR 52.500(c) addressing the changed feature, capability, or strategy, and describe, in the re-evaluation, how the modified design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions. The NRC believes that, because this rule is being proposed to address a beyond-design-basis event, it is appropriate to apply the same standard that was applied during the original evaluation of design features, functional capabilities, and strategies to any licensee-proposed changes to such features, capabilities, and strategies.

Paragraph (d) would state that, for manufacturing licenses which are subject to 10 CFR 52.500, generic changes to the information required by 10 CFR 52.157(f)(32) to be included in the final safety analysis report are governed by the applicable requirements of 10 CFR 52.171. Paragraph (b)(1) of 10 CFR 52.171 does not allow the holder of a manufacturing license to make changes to the design of the nuclear power reactor authorized to be manufactured without prior Commission approval. Any request for a change to the design must be in the form of an application for a license amendment, and must meet the requirements of 10 CFR 50.90 and 10 CFR 50.92.

Paragraph (e) would state that, for combined license applicants or holders which are not subject to 10 CFR 52.500 but reference a manufactured reactor which is subject to 10 CFR 52.500, proposed departures from the information required by 10 CFR 52.157(f)(32) to be included in the FSAR for the manufacturing license are governed by the applicable requirements in 10 CFR 52.171(b)(2). Paragraph (b)(2) of 10 CFR 52.171 allows an applicant or licensee who references or uses a nuclear power reactor manufactured under a manufacturing license under this subpart to request a departure from the design characteristics, site parameters, terms and conditions, or approved design of the manufactured reactor. The Commission may grant a request only if it determines that the departure will comply with the requirements of 10 CFR 52.7, and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure.

The NRC staff is preparing new regulatory guidance on the requirements in proposed 10 CFR 52.500. This guidance is intended to provide an acceptable method by which relevant applicants can perform the assessment of aircraft impacts to meet the requirements of proposed 10 CFR 52.500. The proposed rule would require that the design-specific impact assessment use the Commission-specified general aircraft characteristics. A more detailed description of the aircraft characteristics that should be used in the assessment will be set forth in the regulatory guidance, which will be issued in draft form following publication of this proposed rule. Because the portion of this regulatory guidance describing the detailed aircraft characteristics is likely to contain SGI, that portion of the document will only be made available to those individuals with a need-to-know, and who are otherwise qualified to have access to SGI. A version of the document without the SGI would be made publicly available. Final publication of the regulatory guidance is planned to coincide with publication of the final rule. The Commission reiterates, however, that the commenters on this proposed rule need not await the publication of the draft guidance in order to be able to comment on the proposed rule, in as much as the rule and the supplementary information in this document are sufficient to comment on the substance of the proposed rule.

In addition to the general invitation to submit comments on the proposed rule, the NRC also requests comments on the following questions:

1. Inclusion of impact assessment in application. The proposed rule does not require that the assessment of aircraft impacts that would be mandated by proposed 10 CFR 52.500(b) be included in the FSAR or otherwise submitted as part of the application for a standard design certification, standard design approval, combined license, or manufacturing license. However, the NRC is proposing that a description of the design features, functional capabilities, and strategies credited by the applicant to avoid or mitigate the effects of the applicable, beyond-design-basis aircraft impact be included in the FSAR submitted with the relevant application. In addition, the FSAR must contain an evaluation of how such design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions. The NRC is seeking specific comments on the desirability, or lack thereof, of requiring, in the final rule, that applicants include the aircraft impact assessment required by proposed 10 CFR 52.500(b) in the FSAR or another part of the application.

2. Acceptance criteria. The acceptance criterion contained in proposed 10 CFR 52.500 by which the NRC may judge the required assessment and evaluation is the practicability criterion addressed in paragraph (c), that is, that the applicant must describe how the “design features, functional capabilities, and strategies avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions.” The NRC is considering adding an additional acceptance criterion to proposed 10 CFR 52.500 for judging the acceptability of the applicant's aircraft impact assessment and evaluation. The NRC is seeking specific comments on the desirability, or lack thereof, of adding an additional acceptance criterion in the final rule beyond the proposed rule's practicability criterion. Such an additional acceptance criterion could read, for example:

3. Records retention. The proposed rule relies on the general record retention requirements in 10 CFR 50.71(c) for retention of the assessment required by proposed 10 CFR 52.500 for combined license and manufacturing license applicants subject to proposed 10 CFR 52.500. The NRC intends to similarly rely on a general design certification rule provisions for retention of the assessment required by proposed 10 CFR 52.500 for design certification applicants and combined license and manufacturing license holders that reference a design certification. The NRC is requesting specific comments on whether, in lieu of the specific design certification rule provisions or reliance on 10 CFR 50.71(c), the NRC should adopt as part of the final 10 CFR 52.500 rulemaking a specific provision that would explicitly mandate the retention of the assessment. Such a provision, would be included in an additional paragraph of final 10 CFR 52.500, and would set forth the proposed period of retention. Inclusion of a generic records retention requirement in final 10 CFR 52.500 would preclude the need for the NRC to include a specific records retention provision in each standard design certification subject to final 10 CFR 52.500. The NRC requests comments on whether such a provision should be included in final 10 CFR 52.500, together with specific reasons in support of the commenter's position.

The NRC also requests comments on the appropriate period for retention of the assessment, evaluation, and supporting documentation. The NRC is considering the following alternatives:

• For a standard design certification, combined license, and manufacturing license the period of NRC review prior to NRC final action on the application.

• For a standard design certification and manufacturing license, the duration of the design certification or manufacturing license (i.e., the period during which the design certification or manufactured reactor may be referenced, including any renewal).

• For a standard design certification or manufacturing license, until the licensee of the final referencing license has submitted a certification under 10 CFR 50.82(a), or the final referencing license has been terminated.

• For a combined license, when the licensee has submitted a certification under 10 CFR 50.82(a), or the combined license has been terminated.

4. Requests to amend existing standard design certifications to address aircraft impacts. The NRC has concluded that it does not need to apply the proposed rule to the four currently approved standard design certifications, as discussed in detail in Section III of the Supplementary Information of this document. Nonetheless, the original applicant (or successor in interest of any of the four current standard design certifications) may request an amendment to the standard design certification to add design features, functional capabilities, or strategies in accordance with the requirements of 10 CFR 52.500. The NRC encourages such requests for amendment by the applicants for the four current standard design certifications because it will further enhance the already high levels of safety and security provided by these reactor designs. These design modifications may be implemented in different ways as described in Section III of the Supplementary Information of this document. However, under the proposed rule, there are no standards, other than those contained in 10 CFR 52.63(a), for judging changes to the design to address the effects of an aircraft impact. The NRC requests specific comments on whether it should use the same criterion to judge amendments to an existing design certification as it would use on a new design certification applicant under the proposed 10 CFR 52.500.

5. Applicability to future 10 CFR part 50 license applicants. The NRC is proposing to apply the requirements in proposed 10 CFR 52.500 to 10 CFR part 52 applicants only, specifically, to applicants for standard design certifications issued after the effective date of the final rule that do not reference a standard design approval; standard design approvals issued after the effective date of the final rule; combined licenses issued after the effective date of the final rule that do not reference a standard design certification, standard design approval, or manufactured reactor; and manufacturing licenses issued after the effective date of the final rule that do not reference a standard design certification or standard design approval. However, the NRC is considering extending the applicability of the proposed 10 CFR 52.500 requirements to future applicants for construction permits under 10 CFR part 50. The NRC requests specific comments on the desirability, or lack thereof, of extending, to future 10 CFR part 50 construction permit applicants, the applicability of the proposed requirements to perform an aircraft impact assessment and to evaluate the design features, functional capabilities, and strategies to avoid or mitigate, to the extent practicable, the effects of the applicable, beyond-design-basis aircraft impact.

6. Addition of technical requirements to 10 CFR part 52. In the recent revision to 10 CFR part 52, the NRC took a comprehensive approach to reorganizing 10 CFR part 52 and making conforming changes throughout 10 CFR Chapter I, “Nuclear Regulatory Commission,” to reflect the licensing and approval processes in 10 CFR part 52. In that rulemaking, the NRC reviewed the existing regulations in 10 CFR Chapter I to determine if the existing regulations needed to be modified to reflect the licensing and approval processes in 10 CFR part 52. In making conforming changes involving 10 CFR part 50 provisions, the NRC adopted the general principle of keeping the technical requirements in 10 CFR part 50 and maintaining all applicable procedural requirements in 10 CFR part 52. This proposed aircraft impact rule represents a departure from that general principle in that it proposes to include specific technical requirements in 10 CFR part 52 and would create a separate subpart for inclusion of future, similar, technical requirements. The NRC is considering relocating the proposed aircraft impact requirements from 10 CFR 52.500 to a new section in 10 CFR part 50 in order to maintain the general principle it established in the comprehensive 10 CFR part 52 rulemaking. The NRC requests specific comments on the desirability, or lack thereof, of relocating the proposed aircraft impact requirements from 10 CFR 52.500 to a new section in 10 CFR part 50.

7. Applicability to design approvals and manufacturing licenses. The proposed rule would apply to future design approvals or manufacturing licenses. In the recent comprehensive rulemaking on 10 CFR part 52, the NRC strived for a high level of consistency in the requirements for design certifications, design approvals, and manufacturing licenses, given the similarity in the regulatory functions of these three processes. However, it is not clear that there will be future design approval applications, in light of the NRC's recent determination to remove the design approval as a prerequisite for obtaining a design certification. Similarly, there does not appear to be any near-term interest in obtaining a manufacturing license for the manufacture of a nuclear power plant. Therefore, the NRC is considering eliminating the applicability of the proposed 10 CFR 52.500 requirements to future applicants for design approvals and manufacturing licenses. The NRC requests specific comments on the desirability, or lack thereof, of eliminating the applicability of the proposed 10 CFR 52.500 requirements to future applicants for design approvals and manufacturing licenses.

8. Scope of design evaluated. The proposed 10 CFR 52.500 would be applicable to all standard design certifications, standard design approvals, and manufacturing licenses issued after the effective date of the final rule and to all combined licenses issued after the effective date of the final rule that do not reference a standard design certification, standard design approval, or manufacturing license. However, the proposed rule does not address the difference in the scope of the facility design that would be considered by an applicant for a standard design certification, standard design approval, or manufacturing license and the scope of the design that would be considered by a combined license applicant. For a standard design certification, standard design approval, or manufacturing license, the applicant is required to address only a subset of the facility design that a combined license applicant is required to address. In general, a design certification, design approval, or manufacturing license applicant is required to address such items as the reactor core, reactor coolant system, instrumentation and control systems, electrical systems, containment system, other engineered safety features, auxiliary and emergency systems, power conversion systems, radioactive waste handling systems, and fuel handling systems. In contrast, a combined license applicant also must address site-specific design features, such as the ultimate heat sink. Combined license applicants that do not reference a design certification, design approval, or manufactured reactor could address such site-specific design features in their evaluation of design features, functional capabilities, and strategies to avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions. However, the proposed rule does not impose any requirements on a combined license applicant that references a design certification, design approval, or manufactured reactor with regard to addressing the potential effects of an aircraft impact on such site-specific portions of the design. The proposed rule could, therefore, introduce an inconsistency in the treatment of combined license applicants that reference a design certification, design approval, or manufactured reactor and combined license applicants that submit a custom design. Therefore, to ensure consistent treatment of all combined license applicants, the NRC is considering an alternative approach in the final rule. One approach that the NRC is considering is to adopt additional requirements for combined license applicants that reference a design certification, design approval, or manufactured reactor that would require such applicants to evaluate that portion of the design excluded from the design certification, design approval, or manufactured reactor for additional design features, functional capabilities, or strategies to avoid or mitigate, to the extent practicable, the effects of the applicable aircraft impact with reduced reliance on operator actions. Alternatively, the NRC is considering limiting the scope of the evaluation for combined license applicants not referencing a design certification, design approval, or manufactured reactor to that portion of the design that would otherwise be covered in a design certification, design approval, or manufacturing license application, which would include the majority of the facility considered most vulnerable to an aircraft impact. The NRC requests specific comments on the desirability, or lack thereof, of adopting one of these alternative approaches in the final rule.

The NRC is making the documents identified below available to interested persons through one or more of the following methods as indicated.

Public Document Room (PDR). The NRC Public Document Room is located at 11555 Rockville Pike, Rockville, Maryland 20852.

NRC's Electronic Reading Room (ERR). The NRC's public electronic reading room is located at http://www.nrc.gov/reading-rm.html.

The Presidential memorandum “Plain Language in Government Writing” published June 10, 1998 (63 FR 31883) directed that the Government's documents be in clear and accessible language. The NRC requests comments on the proposed rule specifically with respect to the clarity and effectiveness of the language used. Comments should be sent to the NRC as explained in the ADDRESSES caption of this notice.

Under the “Policy Statement on Adequacy and Compatibility of Agreement States Programs,” approved by the Commission on June 20, 1997, and published in the Federal Register (62 FR 46517; September 3, 1997), this rule is classified as compatibility “NRC.” Compatibility is not required for Category “NRC” regulations. The NRC program elements in this category are those that relate directly to areas of regulation reserved to the NRC by the Atomic Energy Act or the provisions of 10 CFR. Although an Agreement State may not adopt program elements reserved to the NRC, it may wish to inform its licensees of certain requirements via a mechanism that is consistent with the particular State's administrative procedure laws. Category “NRC” regulations do not confer regulatory authority on the State.

The National Technology Transfer and Advancement Act of 1995, Pub. L. 104-113, requires that Federal agencies use technical standards that are developed or adopted by voluntary consensus standards bodies unless using such a standard is inconsistent with applicable law or is otherwise impractical. The NRC is not aware of any voluntary consensus standard that could be used instead of the proposed Government-unique standards. The NRC will consider using a voluntary consensus standard if an appropriate standard is identified.

The NRC has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A to 10 CFR part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and, therefore, an environmental impact statement is not required.

The determination of this environmental assessment is that there will be no significant offsite impact to the public from this action. However, the general public should note that the NRC is seeking public participation; availability of the environmental assessment is provided in Section IX of this document. Comments on any aspect of the environmental assessment may be submitted to the NRC as indicated under the ADDRESSES heading.

The NRC has sent a copy of the environmental assessment and this proposed rule to every State Liaison Officer and requested their comments on the environmental assessment.

This proposed rule contains new or amended information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501, et seq. ). This rule has been submitted to the Office of Management and Budget for review and approval of the information collection requirements.

Type of submission, new or revision: Revision.

The title of the information collection: 10 CFR part 52; “Consideration of Aircraft Impacts for New Nuclear Power Reactor Designs,” proposed rule.

The form number if applicable: N/A.

How often the collection is required: One time; to be submitted with each application for a design certification not referencing a design approval, a design approval, a combined license not referencing a design certification, design approval, or manufactured reactor; or a manufacturing license not referencing a design certification or design approval.

Who will be required or asked to report: Designers and any person eligible under the Atomic Energy Act to apply for a design certification, design approval, combined license, or manufacturing license for a nuclear power plant.

An estimate of the number of annual responses: 1.

The estimated number of annual respondents: 1.

An estimate of the total number of hours needed annually to complete the requirement or request: 3,960 hours.

Abstract: The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to require applicants for new standard design certifications that do not reference a standard design approval; new standard design approvals; combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and new manufacturing licenses that do not reference a standard design certification or standard design approval to assess the effects of the impact of a large, commercial aircraft on the nuclear power plant. Based on the insights gained from this assessment, the applicant shall include in its application a description and evaluation of design features, functional capabilities, and strategies to avoid or mitigate the effects of the aircraft impact. The evaluation of such design features, functional capabilities, and strategies must include core cooling capability, containment integrity, and spent fuel pool integrity. The impact of a large, commercial aircraft is a beyond-design-basis event, and the NRC's requirements applicable to the design, construction, testing, operation, and maintenance of design features, functional capabilities, and strategies for design basis events would not be applicable to design features, functional capabilities, or strategies selected by the applicant solely to meet the requirements of this rule. The objective of this rule is to require nuclear power plant designers to perform a rigorous assessment of design features that could provide additional inherent protection to avoid or mitigate, to the extent practicable, the effects of an aircraft impact, with reduced reliance on operator actions.

The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collections contained in the proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility?

2. Is the estimate of burden accurate?

3. Is there a way to enhance the quality, utility, and clarity of the information to be collected?

4. How can the burden of the information collection be minimized, including the use of automated collection techniques?

A copy of the OMB clearance package may be viewed free of charge at the NRC Public Document Room, One White Flint North, 11555 Rockville Pike, Room O-1 F21, Rockville, MD 20852. The OMB clearance package and rule are available at the NRC worldwide Web site: http://www.nrc.gov/public-involve/doc-comment/omb/index.html for 60 days after the signature date of this document.

Send comments on any aspect of these proposed information collections, including suggestions for reducing the burden and on the above issues, by November 2, 2007 to the Records and FOIA/Privacy Services Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail to INFOCOLLECTS@NRC.GOV and to the Desk Officer, Nathan Frey, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0151), Office of Management and Budget, Washington, DC 20503. Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given to comments received after this date. You may also e-mail comments to Nathan_J._Frey@omb.eop.gov or comment by telephone at 202-395-4650.

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control number.

The NRC has prepared a regulatory analysis on this proposed rule and has included it in this Federal Register document. The analysis examines the costs and benefits of the alternatives considered by the NRC.

This proposed rule would amend 10 CFR part 52 to require applicants for new standard design certifications that do not reference a standard design approval; new standard design approvals; combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and new manufacturing licenses that do not reference a standard design certification or standard design approval to assess the effects of the impact of a large, commercial aircraft on the nuclear power plant. Based on the insights gained from this assessment, the applicant would need to include in its application a description and evaluation of design features, functional capabilities, and strategies to avoid or mitigate the effects of the aircraft impact. The objective of this rule is to require nuclear power plant designers to perform a rigorous assessment of design features that could provide additional inherent protection to avoid or mitigate, to the extent practicable, the effects of an aircraft impact, with reduced reliance on operator actions.

The only alternative considered was to conduct a rulemaking to require applicants to perform an aircraft impact assessment on newly designed facilities because the Commission directed the NRC staff in a staff requirements memorandum dated April 24, 2007, to so revise the regulations. However, the NRC staff considers the no-action alternative as the baseline from which to measure the costs and benefits of the proposed rule.

The regulations in 10 CFR part 52 would be amended to require applicants for new standard design certifications that do not reference a standard design approval; new standard design approvals; combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and new manufacturing licenses that do not reference a standard design certification or standard design approval, and those applicants with applications pending on the effective date of this rule, (relevant applicants) to perform an aircraft impact assessment of the effects on the designed facility of the impact of a large, commercial aircraft. Based on the insights derived from this assessment, the application would need to include a description and evaluation of the design features, functional capabilities, and strategies to avoid or mitigate the effects of an aircraft impact, addressing core cooling capability, containment integrity, and spent fuel pool integrity. The applicant would need to describe how such design and other features avoid or mitigate, to the extent practicable, the aircraft impact effects with reduced reliance on operator actions. The proposed rule would result in newly designed power reactor facilities being more inherently robust with regard to a potential aircraft impact than if they were designed in the absence of this rule.

The NRC identified the attributes that the proposed regulatory action could affect by using the list of potential attributes provided in Chapter 5 of NUREG/BR-0184, “Regulatory Analysis Technical Evaluation Handbook,” issued January 1997. Affected attributes include the following:

Public Health (Accident). The proposed regulatory action would reduce the risk that public health will be affected by the release of radioactive materials to the environment from the impact of a large, commercial aircraft on a nuclear power plant.

Occupational Health (Accident). The proposed regulatory action would reduce the risk that occupational health will be affected by the release of radioactive materials to the environment from the impact of a large, commercial aircraft on a nuclear power plant.

Offsite Property. The proposed regulatory action would reduce the risk that offsite property will be affected by the release of radioactive materials to the environment from the impact of a large, commercial aircraft on a nuclear power plant.

Onsite Property. The proposed regulatory action would reduce the risk that onsite property will be affected by the release of radioactive materials to the environment from the impact of a large, commercial aircraft on a nuclear power plant.

Industry Implementation. The proposed regulatory action would require applicants for new standard design certifications that do not reference a standard design approval; new standard design approvals; combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and new manufacturing licenses that do not reference a standard design certification or standard design approval to assess the effects of the impact of a large, commercial aircraft on the nuclear power plant. Based on the insights gained from this assessment, the applicant would need to include in its application a description and evaluation of design features, functional capabilities, and strategies to avoid or mitigate, to the extent practicable, the effects of an aircraft impact, with reduced reliance on operator actions.

NRC Implementation. Under the proposed regulatory action, the NRC would incur costs to develop guidance on performing an aircraft impact assessment, to review the actions taken by the applicant based on the insights gained from the assessment, and to review the assessment if the NRC needs additional information to verify compliance with this rule. The NRC would also incur the costs of completing this rulemaking.

Improvements in Knowledge. The proposed regulatory action would improve knowledge by ensuring that applicants for newly designed nuclear power facilities perform a rigorous assessment of the effects of the impact of a large, commercial aircraft on the designed facility. Based on the insights gained from this assessment, the applicant would need to include in its application a description and evaluation of the design features, functional capabilities, and strategies to avoid or mitigate the effects of the aircraft impact, addressing core cooling capability, containment integrity, and spent fuel pool integrity. The applicant would need to describe how such design and other features avoid or mitigate, to the extent practicable, the aircraft impact effects with reduced reliance on operator actions.

Safeguards and Security Considerations. The proposed regulatory action to address the capability of newly designed power reactors relative to a potential aircraft impact is based both on enhanced public health and enhanced safety and common defense and security but is not necessary for adequate protection. Rather, it would be to enhance the facility's inherent robustness.

This section describes the process used to evaluate benefits and costs associated with the proposed regulatory action. The benefits (values) come from any desirable changes in the affected attributes which are solely qualitative for the proposed regulatory action; the costs (impacts or burdens) come from any undesirable changes in the affected attributes (e.g., monetary costs, increased exposures). As described in Section 3.1 of this regulatory analysis, the attributes expected to be affected include public health (accident), occupational health (accident), offsite property, onsite property, industry implementation, NRC implementation, improvements in knowledge, and safeguards and security considerations.

Ideally, a cost-benefit analysis quantifies the overall costs and benefits of the regulatory options relative to each of these attributes. This analysis relies on a qualitative evaluation of several of the affected attributes (public health, occupational health, offsite property, onsite property, improvements in knowledge, and safeguards and security considerations) because of the difficulty in quantifying the impact of the current rulemaking. The proposed regulatory action would affect these attributes through the associated reduction in the risks of aircraft impact damage to core cooling, containment integrity, spent fuel pool integrity, and other structures, systems, and components.

The remaining attributes (industry implementation and NRC implementation) are evaluated quantitatively. Quantitative analysis requires a characterization of the universe, including factors such as the number of applicants and the scope of the aircraft impact assessment being performed. The NRC analyzed incremental costs and benefits of the proposed regulatory action relative to the baseline (i.e., the no-action alternative described in Section 2 of this regulatory analysis).

Under OMB guidance and NUREG/BR-0058, “Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission,” Revision 4, issued September 2004, the results of the cost analysis are presented as discounted flows of funds using 3- and 7-percent real discount rates.

The NRC derived information on the estimated number of applications submitted for a new standard design certification that does not reference a standard design approval from industry announcements. Given the uncertainty in the number of (1) new standard design approvals; (2) combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and (3) new manufacturing licenses that do not reference a standard design certification or standard design approval, the NRC staff applied its professional judgment in this analysis.

The proposed regulatory action would apply only to applications for new standard design certifications that do not reference a standard design approval; new standard design approvals; combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and new manufacturing licenses that do not reference a standard design certification or standard design approval, and those applicants with applications pending on the effective date of this rule. It would not apply to (1) a standard design certification or combined license issued before the effective date of the final rule, (2) the design certification rule, (3) Appendices A through D to 10 CFR part 52, or (4) a nuclear power reactor with a current operating license.

For Sections 3.5.1 through 3.5.4, the cost-benefit analysis of the proposed regulatory action is based on the assumed number of applicants in each category. In each case, industry would incur both implementation and operation costs. Furthermore, because all of the benefits are measured qualitatively in this analysis, only costs are included in these subsections.

This analysis uses $105 per hour for both NRC and industry staff rates. The annual results are derived as present values using the 3- and 7-percent discount rates as described in Appendix B to NUREG/BR-0184.

In implementing the proposed regulatory action, standard design certification applicants would incur one-time costs to develop an SGI program, purchase an appropriate SGI storage container, perform the aircraft impact assessment, and incorporate the design features, functional capabilities, or strategies into the design based on the insights gained from the assessment. The NRC estimates that each applicant would spend 120 hours to develop the SGI program. Using the assumed staff rate of $105 per hour, the one-time cost of developing the SGI program would be $13,000 per applicant (120 hours × $105/hour). The NRC also estimates it would cost $2,500 to purchase an appropriate SGI storage container. Finally, the NRC estimates it would take an applicant 24 staff-months for a one-time cost of $400,000 (24 staff-months 4 weeks/month × 40 hours × $105/hour) per application to complete the assessment and incorporate the results into the design. Thus, the one-time cost for an applicant to implement the proposed regulatory action is estimated to be $420,000.

For the standard design certification process, this analysis assumes that three applications would be affected by the proposed rule in the first year following promulgation of this rule, and thereafter, one application would be submitted every 4 years at years 4, 8, 12, 16, and 20. Table 1 shows the discounted flow of funds (using 3- and 7-percent discount rates) of the total industry implementation costs for standard design certification applications over a 20-year period.

With respect to industry operational costs, there would be recordkeeping costs for retention of the assessment, evaluation, and supporting documentation. In addition, it is assumed that an applicant spends 3 hours per year to maintain the records. The estimated annual cost for recordkeeping is $315 per applicant (3 hours × $105/hour). However, for this analysis, it is assumed that it takes 4 years for the Commission to adopt the application as a final standard design certification rule, after which the records are retained by the applicant for 15 years as required by the standard design certification rule. Table 2 shows the discounted flow of funds of the recordkeeping costs (using 3- and 7-percent discount rates) for applications submitted over a 20-year period, using the schedule discussed previously.

Under the proposed regulatory action, an applicant for a standard design approval would need to comply with the requirements for an aircraft impact assessment in 10 CFR 52.500. However, the NRC staff concludes that it is unlikely that a request for a standard design would be submitted to the NRC for approval during the next 20 years. Therefore, no cost-benefit analysis was done for a standard design approval.

Although the NRC concludes that there is a low probability of a combined license applicant not referencing a standard design certification, standard design approval, or manufactured reactor, this analysis assumes that one application would be submitted to the NRC in year 10 following promulgation of the rule.

In implementing the proposed regulatory action, combined license applicants would incur one-time costs to develop an SGI program, purchase an appropriate SGI storage container, perform the aircraft impact assessment, and incorporate the design features, functional capabilities, and strategies into the design based on the insights gained from the assessment. The NRC estimates that each applicant would spend 120 hours to develop the SGI program. Assuming a staff rate of $105 per hour, the one-time cost of developing the SGI program would be $13,000 per applicant (120 hours × $105/hour). The NRC also estimates it would cost $2,500 to purchase an appropriate SGI storage container. Finally, the NRC estimates it would take an applicant 24 staff-months for a one-time cost of $400,000 (24 staff-months × 4 weeks/month × 40 hours × $105/hour) per application to complete the assessment and incorporate the results into the design. Thus, the one-time cost for an applicant to implement the proposed regulatory action is estimated to be $420,000. For one application submitted in year 10, following promulgation of the rule, the discounted flow of funds of the implementation costs are $310,000 and $210,000 using 3- and 7-percent discount rates, respectively.

With respect to industry operational costs, there would be recordkeeping costs for retention of the assessment, evaluation, and supporting documentation. In addition, it is assumed that an applicant spends 3 hours per year to maintain the records. The estimated annual cost for recordkeeping is $315 per applicant (3 hours × $105/hour). For this analysis, it is assumed that it takes 4 years for the Commission to approve the application, after which the records are retained by the licensee for 60 years, at which time the Commission terminates the facility license. The discounted flow of funds of the recordkeeping costs for one application are $5,400 and $1,700 using 3- and 7-percent discount rates, respectively.

The total industry cost is the sum of the implementation and operation costs. The implementation cost is the present value of the assumed one application ($420,000) which when discounted is $310,000 (using a 3-percent discount rate) and $210,000 (using a 7-percent discount rate). The operating costs are $5,400 and $1,700 using the 3- and 7-percent discount rates as shown above. Therefore, the total discounted industry costs are $315,400 and $211,700 using 3- and 7-percent discount rates, respectively.

Although the NRC concludes that there is a low probability of a manufacturing license application not referencing a standard design certification or standard design approval, this analysis assumes that one application would be submitted to the NRC in year 10 following promulgation of the rule.

In implementing the proposed regulatory action, manufacturing license applicants would incur one-time costs to develop an SGI program, purchase an appropriate SGI storage container, perform the aircraft impact assessment, and incorporate the design features, functional capabilities, and strategies into the design based on the insights gained from the assessment. The NRC estimates that each applicant would spend 120 hours to develop the SGI program. Assuming a staff rate of $105 per hour, the one-time cost of developing the SGI program would be $13,000 per applicant (120 hours × $105/hour). The NRC also estimates it would cost $2,500 to purchase an appropriate SGI storage container. Finally, the NRC estimates it would take an applicant 24 staff-months for a one-time cost of $400,000 (24 staff-months × 4 weeks/month × 40 hours × $105/hour) per application to complete the assessment and incorporate the results into the design. Thus, the one-time cost for an applicant to implement the proposed regulatory action is estimated to be $420,000. For one application submitted in year 10, following promulgation of the rule, the discounted flow of funds of the implementation costs are $310,000 and $210,000 using 3- and 7-percent discount rates, respectively.

With respect to industry operational costs, there would be recordkeeping costs for retention of the assessment, evaluation, and supporting documentation. In addition, it is assumed that an applicant spends 3 hours per year to maintain the records. The estimated annual cost for recordkeeping is $315 per applicant (3 hours × $105/hour). For this analysis, it is assumed that it takes 4 years for the Commission to approve the application, after which the records are retained by the licensee for 15 years, at which time the Commission terminates the license. The discounted flow of funds of the recordkeeping costs for one application are $3,300 and $2,200 using 3- and 7-percent discount rates, respectively.

The total industry cost is the sum of the implementation and operation costs. The implementation cost is the present value of the assumed one application ($420,000) which when discounted is $310,000 (using a 3-percent discount rate) and $210,000 (using a 7-percent discount rate). The operating costs are $3,300 and $2,200 using the 3- and 7-percent discount rates as shown above. Therefore, the total discounted industry costs are $313,300 and $212,200 using 3- and 7-percent discount rates, respectively.

Cost to Review the Applicant's Results. The NRC would incur costs to review the applicant's actions in response to the requirements of the proposed rule. The one-time cost to (1) review the actions taken by each applicant for a new standard design certification that does not reference a standard design approval; a combined license that does not reference a standard design certification, standard design approval, or manufactured reactor; and a new manufacturing license that does not reference a standard design certification or standard design approval, and (2) review the assessment if the NRC requires additional information to verify compliance with this rule is estimated to be $17,000 (1 staff-month × 4 weeks/month × 40 hours × $105/hour).

As an example, the total NRC cost for the first year of implementing the proposed rule is the present value of the costs of reviewing three applications for standard design certifications and reviewing the assessments at $51,000 for all three applications. This corresponds to a net present value of $50,000 (using a 3-percent discount rate) and $48,000 (using a 7-percent discount rate).

Cost to Develop Guidance. The NRC would incur 0.5 full-time equivalent (FTE) of staff time to develop guidance to support implementation of the proposed regulatory action. The cost for this action is estimated to be $76,000 (0.5 FTE at $152,000/FTE).

Cost to Provide Training. The NRC would incur costs to develop a training course to instruct NRC staff on the proposed changes to 10 CFR part 52. Assuming that it would take 20 staff-hours to develop the training course, the cost is estimated to be $2,100 (20 staff-hours x $105/hour). The cost to train 20 people for 2 hours, plus the instructor's time of 2 hours is estimated to be $4,400 (21 people × 2 hours × $105/hour). The total cost to the NRC to provide training for the proposed regulatory action is estimated to be $7,000.

Cost of the Regulatory Action. The NRC would incur 2.2 FTE of staff time and $150,000 in contractor support to complete this rulemaking after publishing the proposed rule. The cost of this action is estimated to be $484,000 (2.2 FTE at $152,000/FTE + $150,000).

Table 3 shows the discounted flow of funds of the total NRC implementation costs for the proposed regulatory action over 20 years.

The NRC staff has not identified any impacts to other stakeholders or the Agreement States. However, the proposed action would lead to an increase in public confidence because nuclear power plant designers would perform a rigorous assessment of design and other features that could provide additional inherent protection to avoid or mitigate, to the extent practicable, the effects of an aircraft impact, with reduced reliance on operator actions.

The benefits of the proposed rule can be evaluated only on a qualitative basis. The analysis estimates that the proposed action would result in qualitative benefits in public health (accidental), occupational health (accidental), offsite property, onsite property, improvements in knowledge, and safeguards and security considerations.

Specifically, the benefits would include improvements in knowledge because applicants for new power reactor designs would need to evaluate the design features, functional capabilities, and strategies to avoid or mitigate the effects of the aircraft impact, addressing core cooling capability, containment integrity, and spent fuel pool integrity. If the effects are not assessed by facility designers, it would be more difficult to enhance the inherent robustness of the facility to avoid or mitigate the effects of the aircraft impact. Furthermore, designers of new facilities would need to describe how the design features, functional capabilities, and strategies adopted based on the insights of the assessment avoid or mitigate, to the extent practicable, the effects of an aircraft impact, with reduced reliance on operator actions. In this manner, this proposed rule would result in newly designed nuclear power reactor facilities being more inherently robust with regard to a potential aircraft impact than if they were designed in the absence of this rule.

In addition, because the impact of a large, commercial aircraft is a beyond-design-basis event, this rule would provide an enhanced level of protection beyond that which is provided by the existing adequate protection requirements, which all operating facilities are required to meet, and the proposed adequate protection requirements that facilities will be required to meet if they are made final (see Section I of this document).

Table 4 summarizes the results of the NRC's cost-benefit analysis for industry.

Table 5 shows the total costs of the proposed regulatory action.

The total present-valued costs of this proposed action are $4.0 million and $3.3 million for 3- and 7-percent discount rates, respectively. The benefits are expressed only qualitatively and are discussed in Section 3.5.7 of this regulatory analysis. As noted previously, the key benefit is improvements in knowledge because applicants for new standard design certifications that do not reference a standard design approval; new standard design approvals; combined licenses that do not reference a standard design certification, standard design approval, or manufactured reactor; and new manufacturing licenses that do not reference a standard design certification or standard design approval would need to assess the effects of the impact of a large, commercial aircraft on the nuclear power plant. Based on the insights gained from this assessment, nuclear power plant designers could provide additional inherent protection to avoid or mitigate, to the extent practicable, the effects of an aircraft impact, with reduced reliance on operator actions.

After publication of the proposed rule in the Federal Register and consideration and resolution of the public comments, the NRC would publish a final rule which would become effective 30 days after publication.

The Commission requests public comments on the draft regulatory analysis. Interested persons may submit comments on the draft analysis to the NRC as indicated under the ADDRESSES heading.

In accordance with the Regulatory Flexibility Act (5 U.S.C. 605(b)), the Commission certifies that this rule will not, if promulgated, have a significant economic impact on a substantial number of small entities. This proposed rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of “small entities” set forth in the Regulatory Flexibility Act or the size standards established by the NRC (10 CFR 2.810).

The NRC has determined that neither the backfit rule, 10 CFR 50.109, nor any of the finality provisions in 10 CFR part 52, apply to this proposed rule and, therefore, a backfit analysis is not required, because the proposed rule does not contain any provisions that would impose backfitting as defined in the backfit rule, nor does it contain provisions that would require a finding of compliance or adequate protection under the finality provisions in 10 CFR part 52.

The proposed rule applies only to applicants for a standard design certification not referencing a standard design approval; a standard design approval; a combined license not referencing a standard design certification, standard design approval, manufacture reactor; or a manufacturing license not referencing a standard design certification or standard design approval that are pending at the time of or submitted after the effective date of the final rule. There are no existing combined licenses or manufacturing licenses protected by the backfitting restrictions in 10 CFR 50.109 or the finality provisions in 10 CFR part 52. To the extent that the proposed rule would revise the requirements for future design certifications, combined licenses, or manufacturing licenses the requirements would not constitute backfitting or otherwise be inconsistent with the finality provisions in 10 CFR part 52, because the requirements are prospective in nature and effect. Neither the backfit rule nor the finality provisions in 10 CFR part 52 were intended to apply to every NRC action which substantially changes the expectations of future applicants under 10 CFR part 52.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 553, the NRC is proposing to adopt the following amendments to 10 CFR part 52.

We encourage you to participate in this rulemaking by submitting comments and related material. If you do so, please include the docket number for this rulemaking (USCG-2007-29354), indicate the specific section of this document to which each comment applies, and give the reason for each comment. We recommend that you include your name, mailing address, and an e-mail address or other contact information in the body of your document to ensure that you can be identified as the submitter. This also allows us to contact you in the event further information is needed or if there are questions. For example, if we cannot read your submission due to technical difficulties and you cannot be contacted, your submission may not be considered.

All comments received will be posted, without change, to http://www.regulations.gov and will include any personal information you have provided. We have an agreement with the Department of Transportation (DOT) to use the Docket Management Facility. You may review the Department of Transportation's Privacy Act Statement in the Federal Register published on April 11, 2000 (65 FR 19477), or you may visit http://DocketsInfo.dot.gov.

Please submit all comments and related material in an unbound format, no larger than 8 1/2 by 11 inches, suitable for copying. If you would like to know they reached us, please enclose a stamped, self-addressed postcard or envelope. We will consider all comments and material received during the comment period. We may change this proposed rule in view of them. If, as we anticipate, we make this temporary final rule effective less than 30 days after publication in the Federal Register , we will explain in that publication, as required by 5 U.S.C. 553(d)(3), our good cause for doing so.

We do not now plan to hold a public meeting. But you may submit a request for one to Lieutenant (Junior Grade) Quincy Adams at U.S. Coast Guard Sector Honolulu, Sand Island Parkway, Honolulu, Hawaii 96819-4398, explaining why one would be beneficial. If we determine that one would aid this rulemaking, we will hold one at a time and place announced by a later notice in the Federal Register .

The Hawaii Superferry (HSF) is a 349-foot large passenger vessel documented by the U.S. Coast Guard with an endorsement for coastwise trade, and certificated for large passenger vessel service in the United States. The HSF, operating Hawaii's first inter-island vehicle-passenger service, is intended to provide service among the islands of Oahu, Maui and Kauai.

The HSF enters Kauai at Nawiliwili Harbor, a federally maintained waterway. During the HSF's inaugural commercial trip to Kauai on August 26, 2007, nearly 40 swimmers and obstructers on kayaks and surfboards blocked Nawiliwili Harbor's navigable channel entrance to prevent the lawful entry of the HSF into Kauai. Many of the obstructers entered the water from the jetty that is south of Nawiliwili Park, which is adjacent to the Matson shipping facility in Nawiliwili Harbor. Other demonstrators ashore on the jetty threw rocks and bottles at Coast Guard personnel who were conveying detained obstructers to shore. Coast Guard Station Kauai resources were eventually able to clear the channel for the HSF's arrival while also ensuring the personal safety of the waterborne obstructers. The HSF was able to dock on August 26, 2007.

On the following day, August 27, 2007, approximately 70 persons entered the water again to block the channel entrance, thereby preventing the HSF from docking in Nawiliwili Harbor. Due to the difficulty of maneuvering in the small area of Nawiliwili, and in the interest of ensuring the safety of the protesters, the HSF's master chose not to enter the channel until the Coast Guard cleared the channel of obstructers. However, because the vessel remained outside the harbor, and because the obstructers did not approach within 100 yards of the vessel, the existing security zone for large passenger vessels (33 CFR 165.1410) did not provide the Coast Guard with the authority to control obstructer entry into Nawiliwili Harbor or clear the channel of obstructers before the HSF commenced its transit into the harbor.

After waiting 3 hours, and with nearly 20 obstructers still in the water actively blocking the HSF, the HSF was forced to return to Oahu without mooring in Kauai. This decision was made by the Superferry's master, in consultation with company officials.

As a result of the events of August 26 through 27, 2007, the HSF voluntarily suspended operations between Oahu and Kauai on August 28, 2007. HSF's goal, however, was and is to resume operations between Oahu and Kauai as soon as possible. As of September 26, 2007, there are no, nor have there been, state court injunctions or other legal prohibitions on the HSF resuming operations between Oahu and Kauai.

Responding to these unexpected events, the Coast Guard's Fourteenth District Commander established a temporary fixed security zone in Nawiliwili Harbor. That emergency rulemaking established a temporary security zone in order to prevent persons and vessels from endangering themselves and HSF passengers and crew by attempting to impede the vessel's passage after it commences the difficult transit into the harbor. That rule, which became effective September 1, 2007, was issued by the Coast Guard's Fourteenth District Commander on August 31, 2007 (72 FR 50877, September 5, 2007).

Several events have occurred since August 31, 2007, that have precluded the HSF from resuming commercial service between Oahu and Kauai, notwithstanding the existence of an effective temporary rule that permits activation of a fixed security zone to clear Nawiliwili Harbor for its arrivals and departures. The HSF's parent company is involved in litigation in Maui that has resulted in a temporary restraining order prohibiting HSF commercial service between Oahu and Maui. That litigation is still unresolved. Also, several groups have initiated a lawsuit in Kauai to enjoin the Superferry from operating between Oahu and Kauai; that lawsuit is still ongoing as well. The HSF was scheduled to resume commercial service between Oahu and Kauai on September 26, 2006, but voluntarily decided not to resume service on that date due to considerations of safety and the public interest. As before, however, the HSF could resume operations to Kauai at any time, since there are no federal or state legal impediments to such operations.

The purpose of this proposed temporary rule is several-fold. First, by designating significant portions of the waters of Nawiliwili Harbor as a security zone, activated for enforcement 60 minutes before the HSF's arrival into the zone through 10 minutes after its departure from the zone, the temporary rule would provide the Coast Guard and its law enforcement partners the authority to prevent persons and vessels from endangering themselves and the HSF passengers and crew by attempting to impede the vessel's passage after it commences the difficult transit into the harbor. Extending the security zone to Nawiliwili Jetty and its access road would provide law enforcement personnel with the authority necessary to control access into the water so the HSF may enter and depart the harbor safely and unimpeded by obstructers. Furthermore, closing off the jetty and its access road would prevent violent protesters from continuing to impede law enforcement operations and endanger law enforcement personnel by throwing rocks, bottles, and other dangerous objects. Finally, the security zone would make land adjacent to the harbor available for law enforcement purposes, and in fact would be used by the Patrol Commander (the person in overall command of all waterborne law enforcement assets present in Nawiliwili Harbor enforcing the security zone) as the command post during any Superferry protests.

This NPRM proposes a rule that would be the successor to the original temporary final rule that is set to expire on October 31, 2007. There is continued uncertainty regarding when, if ever, the HSF might resume service into Nawiliwili Harbor. The resolve of obstructers to continue attempting to impede the Superferry's passage into and through Nawiliwili Harbor, should it indeed resume service there, has been vocally manifested. Therefore, the Coast Guard has determined there is a need to ensure that law enforcement personnel will still have a fixed security zone available to them beyond the expiration date of the original temporary final rule to facilitate the safe arrival of the HSF, should it again return to Nawiliwili Harbor. This is the reason for this proposed rulemaking.

The proposed rule would be in effect from November 1, 2007, until December 31, 2007. It is reasonably anticipated that the need for a fixed security zone of this nature will no longer be needed on Kauai after December 31, 2007, though the Coast Guard may, if necessary, draft further rules as necessary and appropriate to ensure safe operation of the HSF in and around Nawiliwili Harbor.

This proposed rule would create a security zone in most of the waters of Nawiliwili Harbor, and on Nawiliwili Jetty in Nawiliwili Harbor. The security zone would be activated for enforcement 60 minutes before the Hawaii Superferry's arrival into the zone, and would remain activated for 10 minutes after the Hawaii Superferry's departure from the zone. The activation of the zone for enforcement would be announced by marine information broadcast and by a red flag, illuminated after sunset, displayed from Pier One and the Harbor Facility Entrance on Jetty Road. During its period of activation and enforcement, entry into the land and water areas of the security zone would be prohibited without the permission of the Captain of the Port, Honolulu, or his or her designated representative.

In preparing this proposed temporary rulemaking, the Coast Guard made sure to consider the rights of lawful protestors. To that end, the Coast Guard excluded from the security zone two regions which create a sizeable area of water in which demonstrators may lawfully assemble and convey their message in a safe manner to their intended audience. These areas include the waters west of a line running from the southeastern-most point of the breakwater of Nawiliwili Small Boat Harbor due south to the south shore of the harbor, and the waters from Kalapaki Beach south to a line extending from the western most point of Kukii Point due west to the Harbor Jetty. These areas of the harbor not included in the security zone are completely accessible to anyone who desires to enter the water, and are fully visible to observers ashore, at the HSF mooring facility, aboard the HSF when transiting the harbor, and from the air.

The Coast Guard also took into account the lawful users of Nawiliwili Harbor in its creation of this rule. As previously noted, the rule will only be activated 1 hour before the HSF's arrival into port, and will be deactivated 10 minutes after the HSF departs the port. The harbor is fully available to all users during the period when the zone is not activated. Furthermore, the rule affords persons desirous of using the harbor, even during a period when the zone is activated, with the opportunity to request permission of the Captain of the Port to do so.

This proposed rule is not a “significant regulatory action” under section 3(f) of Executive Order 12866, Regulatory Planning and Review, and does not require an assessment of potential costs and benefits under section 6(a)(3) of that Order. The Office of Management and Budget has not reviewed it under that Order.

We expect the economic impact of this proposed rule to be so minimal that a full Regulatory Evaluation is unnecessary. This expectation is based on the short activation and enforcement duration of the security zone created by this temporary rule, as well as the limited geographic area affected by the security zone.

Under the Regulatory Flexibility Act (5 U.S.C. 601-612), we have considered whether this proposed rule would have a significant economic impact on a substantial number of small entities. The term “small entities” comprises small businesses, not-for-profit organizations that are independently owned and operated and are not dominant in their fields, and governmental jurisdictions with populations of less than 50,000.

The Coast Guard certifies under 5 U.S.C. 605(b) that this proposed rule would not have a significant economic impact on a substantial number of small entities. While we are aware that the affected area has small entities, including canoe and boating clubs and small commercial businesses that provide recreational services, we anticipate that there will be little or no impact to these small entities due to the narrowly tailored scope of the temporary rule, and to the fact that such entities can request permission from the Captain of the Port to enter the security zone when it is activated.

If you think that your business, organization, or governmental jurisdiction qualifies as a small entity and that this rule would have a significant economic impact on it, please submit a comment (see ADDRESSES ) explaining why you think it qualifies and how and to what degree this rule would economically affect it.

Under section 213(a) of the Small Business Regulatory Enforcement Fairness Act of 1996 (Pub. L. 104-121), we want to assist small entities in understanding this proposed rule so that they can better evaluate its effects on them and participate in the rulemaking. If the rule would affect your small business, organization, or governmental jurisdiction and you have questions concerning its provisions or options for compliance, please contact Lieutenant (Junior Grade) Quincy Adams, U.S. Coast Guard Sector Honolulu, (808) 522-8264. The Coast Guard will not retaliate against small entities that question or complain about this rule or any policy or action of the Coast Guard.

This proposed rule would call for no new collection of information under the Paperwork Reduction Act of 1995 (44 U.S.C. 3501-3520).

A rule has implications for federalism under Executive Order 13132, Federalism, if it has a substantial direct effect on State or local governments and would either preempt State law or impose a substantial direct cost of compliance on them. We have analyzed this proposed rule under that Order and have determined that it does not have implications for federalism.

The Unfunded Mandates Reform Act of 1995 (2 U.S.C. 1531-1538) requires Federal agencies to assess the effects of their discretionary regulatory actions. In particular, the Act addresses actions that may result in the expenditure by a State, local, or tribal government, in the aggregate, or by the private sector of $100,000,000 or more in any one year. Though this proposed rule would not result in such an expenditure, we do discuss the effects of this rule elsewhere in this preamble.

This proposed rule would not effect a taking of private property or otherwise have taking implications under Executive Order 12630, Governmental Actions and Interference with Constitutionally Protected Property Rights.

This proposed rule meets applicable standards in sections 3(a) and 3(b)(2) of Executive Order 12988, Civil Justice Reform, to minimize litigation, eliminate ambiguity, and reduce burden.

We have analyzed this proposed rule under Executive Order 13045, Protection of Children from Environmental Health Risks and Safety Risks. This rule is not an economically significant rule and would not create an environmental risk to health or risk to safety that might disproportionately affect children. While some obstructers used small children in obstruction tactics, both on land and on shore, during the August 26 and 27 Superferry arrivals into Kauai, and while online forums and other sources indicate that organizers are actively recruiting adolescents and small children with the intent of putting them into harm's way as obstructers of the Superferry's passage should it ever again approach and enter Nawiliwili Harbor, any heightened harm faced by children as a result of these tactics has no relation to the creation of this rule. Instead, those heightened risks are entirely the product of persons who recruit and employ adolescents and children to put themselves at risk of death or serious physical injury by attempting to physically obstruct the passage of a large passenger vessel in a small harbor.

This proposed rule does not have tribal implications under Executive Order 13175, Consultation and Coordination with Indian Tribal Governments, because it would not have a substantial direct effect on one or more Indian tribes, on the relationship between the Federal Government and Indian tribes, or on the distribution of power and responsibilities between the Federal Government and Indian tribes.

We have analyzed this proposed rule under Executive Order 13211, Actions Concerning Regulations That Significantly Affect Energy Supply, Distribution, or Use. We have determined that it is not a “significant energy action” under that order because it is not a “significant regulatory action” under Executive Order 12866 and is not likely to have a significant adverse effect on the supply, distribution, or use of energy. The Administrator of the Office of Information and Regulatory Affairs has not designated it as a significant energy action. Therefore, it does not require a Statement of Energy Effects under Executive Order 13211.

The National Technology Transfer and Advancement Act (NTTAA) (15 U.S.C. 272 note) directs agencies to use voluntary consensus standards in their regulatory activities unless the agency provides Congress, through the Office of Management and Budget, with an explanation of why using these standards would be inconsistent with applicable law or otherwise impractical. Voluntary consensus standards are technical standards (e.g., specifications of materials, performance, design, or operation; test methods; sampling procedures; and related management systems practices) that are developed or adopted by voluntary consensus standards bodies.

This proposed rule does not use technical standards. Therefore, we did not consider the use of voluntary consensus standards.

We have analyzed this proposed rule under Commandant Instruction M16475.lD and Department of Homeland Security Management Directive 5100.1, which guide the Coast Guard in complying with the National Environmental Policy Act of 1969 (NEPA) (42 U.S.C. 4321-4370f), and have made a preliminary determination that this action is not likely to have a significant effect on the human environment. Draft documentation supporting this preliminary determination is available in the docket where indicated under ADDRESSES. We seek any comments or information that may lead to the discovery of a significant environmental impact from this proposed rule.

For the reasons discussed in the preamble, the Coast Guard proposes to amend 33 CFR part 165 as follows:

Table of Contents I. What Proposed Actions Are EPA Taking? II. What Is the Background for EPA's Proposed Actions? III. What Are the Criteria for Redesignation? IV. Why Is EPA Proposing These Actions? V. What Is the Effect of EPA's Proposed Actions? VI. What Is EPA's Analysis of the Request? VII. What Is EPA's Analysis of North Carolina's Proposed VOC Insignificance Finding and the Proposed Subarea NO X MVEBs for the Triangle Area? VIII. What Is an Adequacy Determination? IX. What Is the Status of EPA's Adequacy Determination for the Proposed Subarea NO X MVEBs for the Years 2008 and 2017, and the VOC Insignificance Determination? X. Proposed Action on the Redesignation Request and Maintenance Plan SIP Revision Including Proposed Approval of the 2008 and 2017 Subarea NO X MVEBs, and the Proposed VOC Insignificance Determination for the Triangle Area

XI. Statutory and Executive Order Reviews

EPA is proposing to take two related actions, which are summarized below and described in greater detail throughout this notice of proposed rulemaking: (1) To redesignate the Triangle Area to attainment for the 8-hour ozone NAAQS; and (2) to approve North Carolina's 8-hour ozone maintenance plan into the North Carolina SIP, including the associated MVEBs for NO X and the VOC insignificance determination. In addition, and related to today's proposed actions, EPA is also notifying the public of the status of EPA's adequacy determination for the Triangle Area subarea  1 NO X MVEBs and the insignificance determination for VOC emission contribution from motor vehicles to 8-hour ozone pollution in the Triangle Area.

First, EPA is proposing to determine that the Triangle Area has attained the 8-hour ozone standard, and that the Triangle Area has met the other requirements for redesignation under section 107(d)(3)(E) of the CAA. EPA is now proposing to approve a request to change the legal designation of the Triangle Area from nonattainment to attainment for the 8-hour ozone NAAQS.

Second, EPA is proposing to approve North Carolina's 8-hour ozone maintenance plan for the Triangle Area (such approval being one of the CAA criteria for redesignation to attainment status). The maintenance plan is designed to help keep the Triangle Area in attainment of the 8-hour ozone NAAQS through 2017. Consistent with the CAA, the maintenance plan that EPA is proposing to approve today also includes 2008 and 2017 subarea MVEBs for NO X , and an insignificance determination regarding the contribution of VOC emissions from motor vehicles to the ozone pollution in the Triangle Area. Today, EPA is proposing to approve (into the North Carolina SIP) the 2008 and 2017 subarea NO X MVEBs and the VOC insignificance determination, that are included as part of North Carolina's maintenance plan for the Triangle Area for the 8-hour ozone NAAQS. The VOC insignificance determination applies to the entire Triangle Area, whereas the NO X MVEBs are subarea MVEBs that apply to individual counties within the Triangle Area. Please see Section V of this rulemaking for a listing of the MVEBs for these individual counties.

Third, EPA is also notifying the public of the status of EPA's adequacy process for the newly-established 2008 and 2017 subarea NO X MVEBs, and its insignificance determination for VOC for the Triangle Area. The adequacy comment period for the Triangle Area's 2008 and 2017 subarea NO X MVEBs, and the VOC insignificance determination began on March 21, 2007, with EPA's posting of the availability of North Carolina's maintenance plan submittal on EPA's Adequacy Web site ( http://www.epa.gov/otaq/stateresources/transconf/currsips.htm ). The adequacy comment period for these subarea MVEBs, and the VOC insignificance determination closed on April 20, 2007. No adverse comments were received during the adequacy public comment period. Please see section VIII of this proposed rulemaking for further explanation of this process, and for more details on the MVEBs and the VOC insignificance determination.

Today's notice of proposed rulemaking is in response to North Carolina's June 7, 2007, SIP submittal, which supersedes North Carolina's March 12, 2007, submittal that included a request for parallel processing. The June 7, 2007, submittal requests the redesignation of the Triangle Area, and includes a SIP revision addressing the specific issues summarized above and the necessary elements for redesignation described in section 107(d)(3)(E) of the CAA.

Ground-level ozone is not emitted directly by sources. Rather, emissions of NO X and VOC react in the presence of sunlight to form ground-level ozone. NO X and VOC are referred to as precursors of ozone. The CAA establishes a process for air quality management through the NAAQS.

On July 18, 1997, EPA promulgated a revised 8-hour ozone standard of 0.08 parts per million (ppm). This new standard is more stringent than the previous 1-hour ozone standard. Under EPA regulations at 40 CFR part 50, the 8-hour ozone standard is attained when the 3-year average of the annual fourth highest daily maximum 8-hour average ambient air quality ozone concentrations is less than or equal to 0.08 ppm ( i.e. , 0.084 ppm when rounding is considered). (See, 69 FR 23857 (April 30, 2004) for further information.) Ambient air quality monitoring data for the 3-year period must meet a data completeness requirement. The ambient air quality monitoring data completeness requirement is met when the average percent of days with valid ambient monitoring data is greater than 90 percent, and no single year has less than 75 percent data completeness as determined in Appendix I of part 50. Specifically, section 2.3 of 40 CFR part 50, Appendix I, “ Comparisons with the Primary and Secondary Ozone Standards ” states:

The CAA required EPA to designate as nonattainment any area that was violating the 8-hour ozone NAAQS based on the three most recent years of ambient air quality data. The Triangle 8-hour ozone nonattainment area was designated using 2001-2003 ambient air quality data. The Federal Register document making these designations was signed on April 15, 2004, and published on April 30, 2004 (69 FR 23857).

The CAA contains two sets of provisions—subpart 1 and subpart 2—that address planning and control requirements for ozone nonattainment areas. (Both are found in title I, part D.) Subpart 1 (which EPA refers to as “basic” nonattainment) contains general, less prescriptive, requirements for nonattainment areas for any pollutant—including ozone—governed by a NAAQS. Subpart 2 (which EPA refers to as “classified” nonattainment) provides more specific requirements for certain ozone nonattainment areas. Some 8-hour ozone nonattainment areas are subject only to the provisions of subpart 1. Other 8-hour ozone nonattainment areas are also subject to the provisions of subpart 2. Under EPA's Phase 1 8-hour ozone implementation rule (69 FR 23857) (Phase 1 Rule), signed on April 15, 2004, and published April 30, 2004, an area was classified under subpart 2 based on its 8-hour ozone design value (i.e., the 3-year average of the annual fourth-highest daily maximum 8-hour average ozone concentrations), if it had a 1-hour design value at or above 0.121 ppm (the lowest 1-hour design value in Table 1 of subpart 2). All other areas are covered under subpart 1, based upon their 8-hour ambient air quality design values.

Durham and Wake Counties, and the Dutchville Township portion of Granville County were originally designated as a moderate nonattainment area for the 1-hour ozone standard on November 6, 1991 (56 FR 56694). Durham and Wake Counties, and the Dutchville Township portion of Granville County were redesignated as attainment for the 1-hour ozone standard on April 18, 1994 (59 FR 18300). On April 30, 2004, EPA designated the Triangle Area (of which Durham and Wake Counties, and the Dutchville Township portion of Granville County are a part) as a “basic” 8-hour ozone nonattainment area (see, 69 FR 23857, April 30, 2004). Thus, on June 7, 2007, when North Carolina submitted its final redesignation request, the Triangle Area was classified under subpart 1 of the CAA, and was obligated to meet only the subpart 1 requirements.

Various aspects of EPA's Phase 1 Rule were challenged in court. On December 22, 2006, the U.S. Court of Appeals for the District of Columbia Circuit (D.C. Circuit Court) vacated EPA's Phase 1 Rule (69 FR 23951, April 30, 2004). South Coast Air Quality Management Dist. (SCAQMD) v. EPA, 472 F.3d 882 (D.C. Cir. 2006). On June 8, 2007, in response to several petitions for rehearing, the D.C. Circuit Court clarified that the Phase 1 Rule was vacated only with regard to those parts of the Rule that had been successfully challenged. Therefore, the Phase 1 Rule provisions related to classifications for areas currently classified under subpart 2 of title I, part D of the CAA as 8-hour nonattainment areas, the 8-hour attainment dates and the timing for emissions reductions needed for attainment of the 8-hour ozone NAAQS remain effective. The June 8th decision left intact the Court's rejection of EPA's reasons for implementing the 8-hour standard in certain nonattainment areas under subpart 1 in lieu of subpart 2. By limiting the vacatur, the Court let stand EPA's revocation of the 1-hour standard and those anti-backsliding provisions of the Phase 1 Rule that had not been successfully challenged. The June 8th decision reaffirmed the December 22, 2006, decision that EPA had improperly failed to retain measures required for 1-hour nonattainment areas under the anti-backsliding provisions of the regulations: (1) Nonattainment area New Source Review (NSR) requirements based on an area's 1-hour nonattainment classification; (2) Section 185 penalty fees for 1-hour severe or extreme nonattainment areas; and (3) measures to be implemented pursuant to section 172(c)(9) or 182(c)(9) of the CAA, on the contingency of an area not making reasonable further progress toward attainment of the 1-hour NAAQS, or for failure to attain that NAAQS. The June 8th decision clarified that the Court's reference to conformity requirements for anti-backsliding purposes was limited to requiring the continued use of 1-hour motor vehicle emissions budgets until 8-hour budgets were available for 8-hour conformity determinations, which is already required under EPA's conformity regulations. The Court thus clarified that 1-hour conformity determinations are not required for anti-backsliding purposes.

This section sets forth EPA's views on the potential effect of the Court's rulings on this proposed redesignation action. For the reasons set forth below, EPA does not believe that the Court's rulings alter any requirements relevant to this redesignation action so as to preclude redesignation, and do not prevent EPA from proposing or ultimately finalizing this redesignation. EPA believes that the Court's December 22, 2006, and June 8, 2007, decisions impose no impediment to moving forward with redesignation of the Triangle Area to attainment, because even in light of the Court's decisions, redesignation is appropriate under the relevant redesignation provisions of the CAA and longstanding policies regarding redesignation requests.

With respect to the 8-hour standard, the Court's ruling rejected EPA's reasons for classifying areas under subpart 1 for the 8-hour standard, and remanded that matter to the Agency. Consequently, it is possible that this Area could, during a remand to EPA, be reclassified under subpart 2. Although any future decision by EPA to classify this area under subpart 2 might trigger additional future requirements for the area, EPA believes that this does not mean that redesignation cannot now go forward. This belief is based upon (1) EPA's longstanding policy of evaluating requirements in accordance with the requirements due at the time the request is submitted and (2) consideration of the inequity of applying retroactively any requirements that might in the future be applied.

First, at the time the redesignation request was submitted, the Triangle Area was classified under subpart 1 and was obligated to meet only subpart 1 requirements. Under EPA's longstanding interpretation of section 107(d)(3)(E) of the CAA, to qualify for redesignation, states requesting redesignation to attainment must meet only the relevant SIP requirements that came due prior to the submittal of a complete redesignation request. September 4, 1992, Calcagni Memorandum (“Procedures for Processing Requests to Redesignate Areas to Attainment,” Memorandum from John Calcagni, Director, Air Quality Management Division). See also, Michael Shapiro Memorandum, September 17, 1993, and 60 FR 12459, 12465-66 (March 7, 1995) (Redesignation of Detroit-Ann Arbor, Michigan). See, Sierra Club v. EPA, 375 F.3d 537 (7th Cir. 2004) (upholding this interpretation). See, e.g. also, 68 FR 25418, 25424, 25427 (May 12, 2003) (redesignation of St. Louis, Missouri).

Moreover, it would be inequitable to retroactively apply any new SIP requirements that were not applicable at the time the request was submitted. The D.C. Circuit Court has recognized the inequity in such retroactive rulemaking, ( Sierra Club v. Whitman, 285 F. 3d 63 (D.C. Cir. 2002)), in which the Court upheld a district court's ruling refusing to make retroactive an EPA determination of nonattainment that was past the statutory due date. Such a determination would have resulted in the imposition of additional requirements on the area. The Court stated: “Although EPA failed to make the nonattainment determination within the statutory time frame, Sierra Club's proposed solution only makes the situation worse. Retroactive relief would likely impose large costs on the states, which would face fines and suits for not implementing air pollution prevention plans in 1997, even though they were not on notice at the time.” Id. at 68. Similarly here, it would be unfair to penalize the area by applying to it for purposes of redesignation, additional SIP requirements under subpart 2 that were not in effect at the time it submitted its redesignation request.

With respect to the requirements under the 1-hour standard ozone standard, only the Durham and Wake Counties, and the Dutchville Township portion of Granville County of the Triangle Area were originally designated as a moderate nonattainment area for the 1-hour ozone standard on November 6, 1991 (56 FR 56694); the remainder of the Triangle Area was designated as attainment. Durham and Wake Counties, and the Dutchville Township portion of Granville County were redesignated as attainment for the 1-hour ozone standard on April 18, 2004 (59 FR 18300). Therefore, the entire Triangle Area was redesignated to attainment of the 1-hour ozone standard prior to its nonattainment designation for the 8-hour ozone standard. As a result, the Triangle Area is considered to be a 1-hour attainment area subject to a CAA section 175A maintenance plan for the 1-hour standard. The D.C. Circuit Court's decisions do not impact redesignation requests for these types of areas, except to the extent that the Court, in its June 8th decision, clarified that for those areas with 1-hour MVEBs in their maintenance plans, anti-backsliding requires that those 1-hour budgets must be used for 8-hour conformity determinations until they are replaced by 8-hour budgets. To meet this requirement, conformity determinations in such areas must comply with the applicable requirements of EPA's conformity regulations at 40 CFR part 93.

First, there are no conformity requirements relevant for the Triangle Area redesignation request, such as a transportation conformity SIP. It is EPA's longstanding policy that it is reasonable to interpret the conformity SIP requirements as not applying for purposes of evaluating a redesignation request under section 107(d) because state conformity rules are still required after redesignation, and Federal conformity rules apply where state rules have not been approved. See, 40 CFR 51.390; see also, Wall v. EPA, 265 F.3d 426 (6th Cir. 2001) (upholding EPA's interpretation). See also, 60 FR 62748 (Dec. 7, 1995) (redesignation of Tampa, Florida). Durham and Wake Counties, and the Dutchville Township portion of Granville County, currently have a fully approved 1-hour ozone transportation conformity SIP, which was approved on December 27, 2002 (67 FR 78983).

Second, with regard to the three other anti-backsliding provisions for the 1-hour standard that the D.C. Circuit Court found were not properly retained, Durham and Wake Counties, and the Dutchville Township portion of Granville County comprise an attainment area subject to a maintenance plan for the 1-hour standard, and the NSR, contingency measure (pursuant to section 172(c)(9) or 182(c)(9)), and fee provision requirements no longer apply to this area because it was redesignated to attainment of the 1-hour standard. As a result, the decisions in SCAQMD should not alter any requirements that would preclude EPA from finalizing the redesignation of the Triangle Area to attainment for the 8-hour ozone standard.

As noted earlier, in 2006, the ambient ozone data for the Triangle Area indicated no further violations of the 8-hour ozone NAAQS, using data from the 3-year period of 2004-2006 to demonstrate attainment. As a result, on June 7, 2007, North Carolina requested redesignation of the Triangle Area to attainment for the 8-hour ozone NAAQS. The redesignation request included three years of complete, quality-assured ambient air quality data for the ozone seasons (April 1st until October 31st) of 2004-2006, indicating that the 8-hour ozone NAAQS has been achieved for the entire Triangle Area. Under the CAA, nonattainment areas may be redesignated to attainment if sufficient, complete, quality-assured data is available for the Administrator to determine that the area has attained the standard and the area meets the other CAA redesignation requirements in section 107(d)(3)(E).

The CAA provides the requirements for redesignating a nonattainment area to attainment. Specifically, section 107(d)(3)(E) of the CAA allows for redesignation providing that: (1) The Administrator determines that the area has attained the applicable NAAQS; (2) the Administrator has fully approved the applicable implementation plan for the area under section 110(k); (3) the Administrator determines that the improvement in air quality is due to permanent and enforceable reductions in emissions resulting from implementation of the applicable SIP and applicable Federal air pollutant control regulations and other permanent and enforceable reductions; (4) the Administrator has fully approved a maintenance plan for the area as meeting the requirements of section 175A; and, (5) the state containing such area has met all requirements applicable to the area under section 110 and part D of the CAA.

EPA provided guidance on redesignation in the General Preamble for the Implementation of Title I of the CAA Amendments of 1990, on April 16, 1992 (57 FR 13498), and supplemented this guidance on April 28, 1992 (57 FR 18070). EPA has provided further guidance on processing redesignation requests in the following documents:

1. “Ozone and Carbon Monoxide Design Value Calculations,” Memorandum from Bill Laxton, Director, Technical Support Division, June 18, 1990;

2. “Maintenance Plans for Redesignation of Ozone and Carbon Monoxide Nonattainment Areas,” Memorandum from G. T. Helms, Chief, Ozone/Carbon Monoxide Programs Branch, April 30, 1992;

3. “Contingency Measures for Ozone and Carbon Monoxide (CO) Redesignations,” Memorandum from G. T. Helms, Chief, Ozone/Carbon Monoxide Programs Branch, June 1, 1992;

4. “Procedures for Processing Requests to Redesignate Areas to Attainment,” Memorandum from John Calcagni, Director, Air Quality Management Division, September 4, 1992 (hereafter referred to as the “Calcagni Memorandum”);

5. “State Implementation Plan (SIP) Actions Submitted in Response to Clean Air Act (ACT) Deadlines,” Memorandum from John Calcagni, Director, Air Quality Management Division, October 28, 1992;

6. “Technical Support Documents (TSD's) for Redesignation of Ozone and Carbon Monoxide (CO) Nonattainment Areas,” Memorandum from G. T. Helms, Chief, Ozone/Carbon Monoxide Programs Branch, August 17, 1993;

7. “State Implementation Plan (SIP) Requirements for Areas Submitting Requests for Redesignation to Attainment of the Ozone and Carbon Monoxide (CO) National Ambient Air Quality Standards (NAAQS) On or After November 15, 1992,” Memorandum from Michael H. Shapiro, Acting Assistant Administrator for Air and Radiation, September 17, 1993;

8. “Use of Actual Emissions in Maintenance Demonstrations for Ozone and CO Nonattainment Areas,” Memorandum from D. Kent Berry, Acting Director, Air Quality Management Division, November 30, 1993;

9. “Part D New Source Review (Part D NSR) Requirements for Areas Requesting Redesignation to Attainment,” Memorandum from Mary D. Nichols, Assistant Administrator for Air and Radiation, October 14, 1994; and

10. “Reasonable Further Progress, Attainment Demonstration, and Related Requirements for Ozone Nonattainment Areas Meeting the Ozone National Ambient Air Quality Standard,” Memorandum from John S. Seitz, Director, Office of Air Quality Planning and Standards, May 10, 1995.

On June 7, 2007, North Carolina requested redesignation of the Triangle 8-hour ozone nonattainment area to attainment for the 8-hour ozone standard. EPA's evaluation indicates that North Carolina has demonstrated that the Triangle Area has attained the standard and has met the requirements for redesignation set forth in section 107(d)(3)(E) of the CAA. EPA is also announcing the status of its adequacy determination for the 2008 and 2017 subarea NO X MVEBs, and the VOC insignificance determination, which are relevant to the requested redesignation.

EPA's proposed actions establish the bases upon which EPA may take final action on the issues being proposed for approval today. Approval of North Carolina's redesignation request would change the legal designation of the Durham, Franklin, Granville, Johnston, Orange, Person and Wake Counties in their entireties, and Baldwin, Center, New Hope and Williams Townships in Chatham County for the 8-hour ozone NAAQS found at 40 CFR part 81. Approval of North Carolina's request would also incorporate into the North Carolina SIP, a plan for the Triangle Area for maintaining the 8-hour ozone NAAQS in the Area through 2017. This maintenance plan includes contingency measures to remedy future violations of the 8-hour ozone NAAQS. The maintenance plan also establishes subarea NO X MVEBs and provides a VOC insignificance determination for the Triangle Area. The following Table identifies the subarea NO X MVEBs for the year 2008 and 2017 for this Area.

Approval of North Carolina's maintenance plan would also result in approval of the subarea NO X MVEBs, and the VOC insignificance determination. Additionally, EPA is notifying the public of the status of its adequacy determination for the 2008 and 2017 subarea NO X MVEBs, and its VOC insignificance determination, pursuant to 40 CFR 93.118(f)(1).

EPA is proposing to make the determination that the Triangle Area has attained the 8-hour ozone standard, and that all other redesignation criteria have been met for the Triangle Area. The basis for EPA's determination for the area is discussed in greater detail below.

EPA is proposing to determine that the Triangle Area has attained the 8-hour ozone NAAQS. For ozone, an area may be considered to be attaining the 8-hour ozone NAAQS if there are no violations, as determined in accordance with 40 CFR 50.10 and Appendix I of part 50, based on three complete, consecutive calendar years of quality-assured air quality monitoring data. To attain this standard, the 3-year average of the fourth-highest daily maximum 8-hour average ozone concentrations measured at each monitor within an area over each year must not exceed 0.08 ppm. Based on the rounding convention described in 40 CFR part 50, Appendix I, the standard is attained if the design value is 0.084 ppm or below. The data must be collected and quality-assured in accordance with 40 CFR part 58, and recorded in the EPA Air Quality System (AQS). The monitors generally should have remained at the same location for the duration of the monitoring period required for demonstrating attainment.

EPA reviewed ozone monitoring data from ambient ozone monitoring stations in the Triangle Area for the ozone season from 2004-2006. This data has been quality assured and is recorded in AQS. The fourth high averages for 2004, 2005 and 2006, and the 3-year average of these values (i.e., design values), are summarized in the following Table:

As discussed above, the design value for an area is the highest design value recorded at any monitor in the area. Therefore, the design value for the Triangle Area is 0.080 ppm, which meets the standard as described above. As discussed in more detail below, North Carolina has committed to continue monitoring in this area in accordance with 40 CFR part 58. The data submitted by North Carolina provides an adequate demonstration that the Triangle Area has attained the 8-hour ozone NAAQS.

Below is a summary of how these two criteria were met.

EPA has determined that North Carolina has met all applicable SIP requirements for the Triangle Area under section 110 of the CAA (general SIP requirements). EPA has also determined that the North Carolina SIP satisfies the criterion that it meet applicable SIP requirements under part D of title I of the CAA (requirements specific to subpart 1 basic 8-hour ozone nonattainment areas) in accordance with section 107(d)(3)(E)(v). In addition, EPA has determined that the SIP is fully approved with respect to all applicable requirements in accordance with section 107(d)(3)(E)(ii). In making these determinations, EPA ascertained which requirements are applicable to the area and that if applicable, they are fully approved under section 110(k). SIPs must be fully approved only with respect to applicable requirements.

The September 4, 1992, Calcagni Memorandum (see “Procedures for Processing Requests to Redesignate Areas to Attainment,” Memorandum from John Calcagni, Director, Air Quality Management Division, September 4, 1992) describes EPA's interpretation of section 107(d)(3)(E). Under this interpretation, to qualify for redesignation, states requesting redesignation to attainment must meet only the relevant CAA requirements that come due prior to the submittal of a complete redesignation request. See also, Michael Shapiro Memorandum, (“SIP Requirements for Areas Submitting Requests for Redesignation to Attainment of the Ozone and Carbon Monoxide NAAQS On or After November 15, 1992,” September 17, 1993), and 60 FR 12459, 12465-66 (March 7, 1995) (redesignation of Detroit-Ann Arbor, Michigan). Applicable requirements of the CAA that come due subsequent to the area's submittal of a complete redesignation request remain applicable until a redesignation is approved, but are not required as a prerequisite to redesignation. See, section 175A(c) of the CAA; Sierra Club , 375 F.3d 537; see also, 68 FR 25424, 25427 (May 12, 2003) (redesignation of St. Louis, Missouri).

General SIP requirements. Section 110(a)(2) of title I of the CAA delineates the general requirements for a SIP, which include enforceable emissions limitations and other control measures, means, or techniques, provisions for the establishment and operation of appropriate devices necessary to collect data on ambient air quality, and programs to enforce the limitations. General SIP elements and requirements are delineated in section 110(a)(2) of title I, part A of the CAA. These requirements include, but are not limited to, the following: submittal of a SIP that has been adopted by the state after reasonable public notice and hearing; provisions for establishment and operation of appropriate procedures needed to monitor ambient air quality; implementation of a source permit program; provisions for the implementation of part C requirements (Prevention of Significant Deterioration (PSD)) and provisions for the implementation of part D requirements (NSR permit programs); provisions for air pollution modeling; and provisions for public and local agency participation in planning and emission control rule development.

Section 110(a)(2)(D) requires that SIPs contain certain measures to prevent sources in a state from significantly contributing to air quality problems in another state. To implement this provision, EPA has required certain states to establish programs to address the transport of air pollutants (NO X SIP Call, Clean Air Interstate Rule (CAIR)). EPA has also found, generally, that states have not submitted timely SIPs under section 110(a)(1) to meet the interstate transport requirements of section 110(a)(2)(D)(i). However, the section 110(a)(2)(D) requirements for a state are not linked with a particular nonattainment area's designation and classification in that state. EPA believes that the requirements linked with a particular nonattainment area's designation and classifications are the relevant measures to evaluate in reviewing a redesignation request. The transport SIP submittal requirements, where applicable, continue to apply to a state regardless of the designation of any one particular area in the state. Thus, we do not believe that the CAA's interstate transport requirements should be construed to be applicable requirements for purposes of redesignation.

In addition, EPA believes that the other section 110 elements not connected with nonattainment plan submissions and not linked with an area's attainment status are not applicable requirements for purposes of redesignation. The area will still be subject to these requirements after the area is redesignated. The section 110 and part D requirements, which are linked with a particular area's designation and classification, are the relevant measures to evaluate in reviewing a redesignation request. This approach is consistent with EPA's existing policy on applicability (i.e., for redesignations) of conformity and oxygenated fuels requirements, as well as with section 184 ozone transport requirements. See, Reading, Pennsylvania, proposed and final rulemakings (61 FR 53174-53176, October 10, 1996), (62 FR 24826, May 7, 1997); Cleveland-Akron-Loraine, Ohio, final rulemaking (61 FR 20458, May 7, 1996); and Tampa, Florida, final rulemaking at (60 FR 62748, December 7, 1995). See also, the discussion on this issue in the Cincinnati, Ohio redesignation (65 FR 37890, June 19, 2000), and in the Pittsburgh, Pennsylvania redesignation (66 FR 50399, October 19, 2001).

EPA believes that section 110 elements not linked to the area's nonattainment status are not applicable for purposes of redesignation. Any section 110 requirements that are linked to the Part D requirements for 8-hour ozone nonattainment areas are not yet due, since, as explained below, no part D requirements for 8-hour standard became due prior to submission of the redesignation request. Therefore, as discussed above, for purposes of redesignation, they are not considered applicable requirements. Nonetheless, EPA notes it has previously approved provisions in the North Carolina SIP addressing section 110 elements under the 1-hour ozone NAAQS (See, 51 FR 19834, June 3, 1986). EPA believes that the section 110 SIP approved for the 1-hour ozone NAAQS is also sufficient to meet the requirements under the 8-hour ozone NAAQS (as well as satisfying the issues raised by the D.C. Circuit Court in the SCAQMD case).

Part D requirements. EPA has also determined that the North Carolina SIP meets applicable SIP requirements under part D of the CAA since no requirements became due prior to the submission of the Area's redesignation request. Sections 172-176 of the CAA, found in subpart 1 of part D, set forth the basic nonattainment requirements applicable to all nonattainment areas. Section 182 of the CAA, found in subpart 2 of part D, establishes additional specific requirements depending on the area's nonattainment classification. Subpart 2 is not applicable to the Triangle Area.

Part D, subpart 1 applicable SIP requirements. For purposes of evaluating this redesignation request, the applicable part D, subpart 1 SIP requirements for all nonattainment areas are contained in sections 172(c)(1)-(9). A thorough discussion of the requirements contained in section 172 can be found in the General Preamble for Implementation of title I (57 FR 13498). No requirements applicable for purposes of redesignation under part D became due prior to the submission of the redesignation request, and therefore none are applicable to the Area for purposes of redesignation. For example, the requirements for an attainment demonstration that meets the requirements of section 172(c)(1) are not yet applicable, nor are the requirements for Reasonably Achievable Control Technology (RACT) and Reasonably Available Control Measures (RACM) (section 172(c)(1)), reasonable further progress (RFP) (section 172(c)(2)), and contingency measures (section 172(c)(9)).

In addition to the fact that no part D requirements applicable for purposes of redesignation became due prior to submission of the redesignation request and therefore are not applicable, EPA believes it is reasonable to interpret the conformity and NSR requirements as not requiring approval prior to redesignation.

Section 176 Conformity Requirements. Section 176(c) of the CAA requires states to establish criteria and procedures to ensure that Federally supported or funded projects conform to the air quality planning goals in the applicable SIP. The requirement to determine conformity applies to transportation plans, programs and projects developed, funded or approved under title 23 of the United States Code (U.S.C.) and the Federal Transit Act (transportation conformity) as well as to all other Federally supported or funded projects (general conformity). State conformity revisions must be consistent with Federal conformity regulations relating to consultation, enforcement and enforceability that the CAA required the EPA to promulgate.

EPA believes it is reasonable to interpret the conformity SIP requirements as not applying for purposes of evaluating the redesignation request under section 107(d) because state conformity rules are still required after redesignation and Federal conformity rules apply where state rules have not been approved. See, Wall , 265 F.3d 426 (upholding this interpretation). See also, 60 FR 62748 (December 7, 1995, Tampa, Florida).

NSR Requirements. EPA has also determined that areas being redesignated need not comply with the requirement that a NSR program be approved prior to redesignation, provided that the area demonstrates maintenance of the standard without a part D NSR program in effect since PSD requirements will apply after redesignation. The rationale for this view is described in a memorandum from Mary Nichols, Assistant Administrator for Air and Radiation, dated October 14, 1994, entitled “Part D New Source Review (Part D NSR) Requirements for Areas Requesting Redesignation to Attainment.” North Carolina has demonstrated that the Triangle Area will be able to maintain the standard without a part D NSR program in effect, and therefore, North Carolina need not have a fully approved part D NSR program prior to approval of the redesignation request. North Carolina's PSD program will become effective in the Triangle Area upon redesignation to attainment. See, rulemakings for Detroit, Michigan (60 FR 12467-12468, March 7, 1995); Cleveland-Akron-Lorraine, Ohio (61 FR 20458, 20469-70, May 7, 1996); Louisville, Kentucky (66 FR 53665, October 23, 2001); Grand Rapids, Michigan (61 FR 31834-31837, June 21, 1996). Thus, the Triangle Area has satisfied all applicable requirements for purposes of redesignation under section 110 and part D of the CAA.

EPA has fully approved the applicable North Carolina SIP for the Triangle Area, under section 110(k) of the CAA for all requirements applicable for purposes of redesignation. EPA may rely on prior SIP approvals in approving a redesignation request, see Calcagni Memorandum at p. 3; Southwestern Pennsylvania Growth Alliance v. Browner , 144 F.3d 984, 989-90 (6th Cir. 1998); Wall , 265 F.3d 426, plus any additional measures it may approve in conjunction with a redesignation action. See, 68 FR 25426 (May 12, 2003) and citations therein. Following passage of the CAA of 1970, North Carolina has adopted and submitted, and EPA has fully approved at various times, provisions addressing the various 1-hour ozone standard SIP elements applicable in the Triangle Area (59 FR 18300, April 18, 1994; and 69 FR 56163, September 20, 2004).

As indicated above, EPA believes that the section 110 elements not connected with nonattainment plan submissions and not linked to the area's nonattainment status are not applicable requirements for purposes of redesignation. EPA also believes that since the part D requirements applicable for purposes of redesignation did not become due prior to submission of the redesignation request, they also are therefore not applicable requirements for purposes of redesignation.

EPA believes that North Carolina has demonstrated that the observed air quality improvement in the Triangle Area is due to permanent and enforceable reductions in emissions resulting from implementation of the SIP, Federal measures, and other state adopted measures. Additionally, new emissions control programs for fuels and motor vehicles will help ensure a continued decrease in emissions throughout the region.

In its request to redesignate the Triangle Area to attainment, North Carolina submitted a SIP revision to provide for the maintenance of the 8-hour ozone NAAQS for at least 10 years after the effective date of redesignation to attainment.

Section 175A of the CAA sets forth the elements of a maintenance plan for areas seeking redesignation from nonattainment to attainment. Under section 175A, the plan must demonstrate continued attainment of the applicable NAAQS for at least 10 years after the Administrator approves a redesignation to attainment. Eight years after the redesignation, the State of North Carolina must submit a revised maintenance plan, which demonstrates that attainment will continue to be maintained for the 10 years following the initial 10-year period. To address the possibility of future NAAQS violations, the maintenance plan must contain such contingency measures, with a schedule for implementation as EPA deems necessary to assure prompt correction of any future 8-hour ozone violations. Section 175A of the CAA sets forth the elements of a maintenance plan for areas seeking redesignation from nonattainment to attainment. The Calcagni Memorandum provides additional guidance on the content of a maintenance plan. The Calcagni Memorandum explains that an ozone maintenance plan should address five requirements: the attainment emissions inventory, maintenance demonstration, monitoring, verification of continued attainment, and a contingency plan. As is discussed more fully below, North Carolina's maintenance plan includes all the necessary components and is approvable as part of the redesignation request.

North Carolina selected 2005 as “the attainment year” for the Triangle Area for the purposes of demonstrating attainment of the 8-hour ozone NAAQS. This attainment inventory identifies the level of emissions in the area, which is sufficient to attain the 8-hour ozone standard. North Carolina began development of this attainment inventory by first developing a baseline emissions inventory for the Triangle Area. The year 2005 was chosen as the base year for developing a comprehensive ozone precursor emissions inventory for which projected emissions could be developed for 2008, 2011, 2014, and 2017. Nonroad mobile emissions estimates were based on the EPA's NONROAD2005c model. On-road mobile source emissions were calculated using EPA's MOBILE6.2 emission factors model. The 2005 VOC and NO X emissions, as well as the emissions for other years, for the Triangle Area were developed consistent with EPA guidance, and are summarized in Tables 4 and 5 in the following subsection.

The June 7, 2007, final submittal includes a maintenance plan for the Triangle Area. This demonstration:

(i) Shows compliance and maintenance of the 8-hour ozone standard by providing information to support the demonstration that current and future emissions of VOC and NO X remain at or below attainment year 2005 emissions levels. The year 2005 was chosen as the attainment year because it is one of the most recent three years (i.e., 2004, 2005, and 2006) for which the Triangle Area has clean air quality data for the 8-hour ozone standard.

(ii) Uses 2005 as the attainment year and includes future emission inventory projections for 2008, 2011, 2014, and 2017.

(iii) Identifies an “out year,” at least 10 years after the time necessary for EPA to review and approve the maintenance plan. Per 40 CFR part 93, subarea NO X MVEBs were established for the last year (2017) of the maintenance plan. Additionally, North Carolina chose, through interagency consultation, to establish subarea MVEBs for the year 2008 for NO X , and to determine insignificance for VOC for the Triangle Area. See, section VII below.

(iv) Provides the following actual and projected emissions inventories, in tons per day (tpd) for the Triangle Area. See, Tables 4 and 5.

A safety margin is the difference between the attainment level of emissions (from all sources) and the projected level of emissions (from all sources) in the maintenance plan. The attainment level of emissions is the level of emissions during one of the years in which the area met the NAAQS. North Carolina has decided to allocate a portion of the available safety margin to the subarea NO X MVEBs for the years 2008 and 2017 for the Triangle Area, and has calculated the safety margin in its submittal. See, Tables 4 and 5, above. This allocation and the resulting available safety margin for the Triangle Area are discussed further in section VII of this proposed rulemaking.

There are currently eight monitors measuring ozone in the Triangle Area. North Carolina has committed in the maintenance plan to continue operation of these monitors in compliance with 40 CFR part 58, and has addressed the requirement for monitoring.

North Carolina has the legal authority to enforce and implement the requirements of the ozone maintenance plan for the Triangle Area. This includes the authority to adopt, implement and enforce any subsequent emissions control contingency measures determined to be necessary to correct future ozone attainment problems.

North Carolina will track the progress of the maintenance plan by performing future reviews of actual emissions for the Area using the latest emissions factors, models and methodologies. For these periodic inventories, North Carolina will review the assumptions made for the purpose of the maintenance demonstration concerning projected growth of activity levels. If any of these assumptions appear to have changed substantially, North Carolina will re-project emissions.

The contingency plan provisions are designed to promptly correct a violation of the NAAQS that occurs after redesignation. Section 175A of the CAA requires that a maintenance plan include such contingency measures as EPA deems necessary to assure that the state will promptly correct a violation of the NAAQS that occurs after redesignation. The maintenance plan should identify the contingency measures to be adopted, a schedule and procedure for adoption and implementation, and a time limit for action by the state. A state should also identify specific indicators to be used to determine when the contingency measures need to be implemented. The maintenance plan must include a requirement that a state will implement all measures with respect to control of the pollutant that were contained in the SIP before redesignation of the area to attainment in accordance with section 175A(d).

In the June 7, 2007, submittal, North Carolina affirms that all programs instituted by the State and EPA will remain enforceable, and that sources are prohibited from reducing emissions controls following the redesignation of the area. The contingency plan included in the submittal provides tracking and triggering mechanisms to determine when contingency measures are needed and a process of developing and adopting appropriate control measures. The primary trigger of the contingency plan will be a violation of the 8-hour ozone NAAQS, or when the three-year average of the fourth-highest value is equal to or greater than 0.085 ppm at any of the Triangle Area monitors. The trigger date will be 60 days from the date that the State observes a fourth-highest value that, when averaged with the two previous ozone seasons' fourth highest values, would result in a three-year average equal to or greater than 0.085 ppm. The secondary trigger will apply where no actual violation of the 8-hour ozone standard has occurred, but where the State finds monitored ozone levels indicating that an ozone NAAQS violation may be imminent. An imminent violation exists where there is a pattern. A pattern will be deemed to exist when there are two consecutive ozone seasons in which the fourth-highest values are 0.085 ppm or greater at a single monitor within the Triangle Area. The trigger date will be 60 days from the date that the State observes a fourth-highest value of 0.085 ppm or greater at a monitor for which the previous season had a fourth-highest value of 0.085 ppm or greater. Similarly, the tertiary trigger is a first alert to a potential air quality problem in the future and will not be an actual violation of the 8-hour ozone standard. The trigger will be activated when a monitor in the Triangle Area has a fourth-highest value of 0.085 ppm or greater, starting the first year after the maintenance plan has been approved. The trigger date will be 60 days from the date that the State observes a fourth-highest value of 0.085 ppm or greater at any monitor.

In the submittal, if there is a measured violation of the 8-hour ozone NAAQS in the Triangle Area, contingency measures would be adopted and implemented as expeditiously as possible, but no later than eighteen to twenty four months after the triggering event. Once the primary or secondary trigger is activated, the proposed schedule for these actions would be as follows:

• NCDENR will begin analyses, including trajectory analyses of high ozone days, and emissions inventory assessment to determine required emission control measures for attaining or maintaining the 8-hour ozone standard;

• By May 1st of the year following the ozone season in which the primary (a violation of the 8-hour ozone NAAQS occurs) or secondary trigger has been activated, NCDENR will complete sufficient analyses to begin adoption of necessary rules for ensuring attainment and maintenance of the 8-hour ozone NAAQS; and

• Rules would become State-effective by the following January 1st, unless legislative review is required.

North Carolina will consider one or more of the following contingency measures to re-attain the standard:

• NO X RACT on stationary sources in the Triangle Area;

• Diesel inspection and maintenance program  2 ;

• Implementation of diesel retrofits programs, including incentives for performing retrofits; and

• Additional controls in upwind areas.

Once the tertiary trigger is activated, NCDENR will commence analyses including meteorological evaluation, trajectory analyses of high ozone days, and an emissions inventory assessment to understand why a fourth highest exceedance of the standard has occurred. NCDENR will then work with the local awareness program and develop an outreach plan to identify any additional voluntary measures that can be implemented. If the fourth highest exceedance occurs early in the season, NCDENR will work with entities identified in the outreach plan to determine if the measures can be implemented during the ozone season. Otherwise, NCDENR will work with the local air awareness coordinator to implement the plan for the following ozone season.

EPA has concluded that the maintenance plan adequately addresses the five basic components of a maintenance plan: Attainment inventory, maintenance demonstration, monitoring network, verification of continued attainment, and a contingency plan. The maintenance plan SIP revision submitted by North Carolina for the Triangle Area meets the requirements of section 175A of the CAA and is approvable.

Today's actions address two related elements regarding on-road motor vehicle emissions and the requirement to establish MVEBs. First, EPA is proposing to find that the VOC emission contribution from motor vehicles to 8-hour ozone pollution in the Triangle Area is insignificant. The result of this finding, if finalized, is that North Carolina need not develop an MVEB for VOC for the Triangle Area. See below for further information on the insignificance determination. Second, EPA is proposing to approve the subarea NO X MVEBs for the Triangle Area.

In certain instances, the Transportation Conformity Rule allows areas not to establish an MVEB where it is demonstrated that the regional motor vehicle emissions for a particular pollutant/precursor is an insignificant contributor to the air quality problem in an area. The general criteria for insignificance findings can be found in 40 CFR 93.109(k). Insignificance determinations are based on a number of factors, including (1) The percentage of motor vehicle emissions in context of the total SIP inventory; (2) the current state of air quality as determined by monitoring data for that NAAQS; (3) the absence of SIP motor vehicle control measures; and (4) historical trends and future projections of the growth of motor vehicle emissions. EPA's rationale for the providing for insignificance determinations is described in the July 1, 2004, revision to the Transportation Conformity Rule at 69 FR 40004. Specifically, the rationale is explained on page 40061 under the subsection entitled “XXIII. B. Areas With Insignificant Motor Vehicle Emissions.” Any insignificance determination under review of EPA is subject to the adequacy and approval process for EPA's action on the SIP.

Through the adequacy and SIP approval process, EPA may find that a SIP demonstrates that regional motor vehicle emissions are an insignificant contributor to the air quality problem for the pollutant/precursor at issue. In the case of the Triangle Area, EPA intends to make its finding as part of EPA's final action on the redesignation request of North Carolina for the Triangle Area. Upon the effective date of EPA's adequacy finding or the publication date of the final rule for this SIP revision (i.e., which includes the VOC insignificance determination), federal regulations waive the regional emissions analysis requirements (for the purpose of transportation conformity implementation) for the relevant pollutant or precursor. Areas with insignificant regional motor vehicle emissions for a pollutant or precursor are still required to make a conformity determination that satisfies other relevant requirements. Additionally, such areas are required to satisfy the regional emissions analysis requirements for pollutants or precursors for which EPA has not made a finding of insignificance.

The maintenance plan for the Triangle Area, included as part of the SIP revision, contains MVEBs for NO X and an insignificance determination for VOC contribution from motor vehicles to the 8-hour ozone pollution in the Triangle Area. As part of the preparation for its redesignation request, North Carolina consulted with the interagency consultation group for the Triangle Area regarding the insignificance determination for VOC. For the purposes of regional emissions analysis, the information provided by North Carolina supports EPA's proposal to determine VOC contribution to 8-hour ozone pollution from motor vehicles in the Triangle Area as insignificant. The information provided by North Carolina to EPA as part of the SIP revision addresses each of the factors listed in 40 CFR 93.109(k), and is summarized below.

The future on-road VOC emissions are projected to be less than 10 percent in the Triangle Area, in the context of the total SIP inventory. According to information provided by North Carolina, biogenic emissions account for approximately 90 percent of the VOC emissions in future years in the Triangle Area. Support for these percentages is found in Figure 4.1.6-3, located in Appendix C.3—Mobile Source Inventory Documentation on pages 4-36 of North Carolina's submittal (available in the Docket for this proposed rulemaking) which also indicates on-road VOC emissions declining by about 50 percent by 2017 and vehicle miles traveled (VMT) going up by about 25 to 30 percent by 2017. In addition, North Carolina conducted a sensitivity analysis (a photochemical model) that indicated that 8-hour ozone levels in the Triangle Area were not impacted by reductions in man-made VOC emissions (i.e., reductions from motor vehicles). Specifically, the photochemical model was run for a 39-day scenario with a modeled 30 percent reduction in man-made VOC emissions. According to the photochemical model, in the year 2009, even with anticipated increases in VMT, the mobile source inventory is still projected to be less than 6 percent of the total inventory for VOC emissions. In comparison, biogenic emissions are expected to account for at least 84 percent of the total inventory for VOC emissions. As discussed in North Carolina's submittal, the biogenic sector is the most abundant source of VOC in North Carolina and accounts for approximately 90 percent of the total VOC emissions statewide. As a result, the information provided by North Carolina indicates that VOC contribution to 8-hour ozone pollution from motor vehicle emissions is insignificant.

With regard to the factor relating to the absence of motor vehicle control measures in the SIP, EPA considered the existence of an inspection and maintenance (I/M) program in the North Carolina SIP, and its implementation in the individual counties comprising the Triangle Area. The I/M program was not added to the North Carolina SIP as a VOC control measure, but rather, a NO X control measure. The I/M program is currently being implemented in all but one of the counties (Person County) in the Triangle Area. Implementation of the I/M program in the Triangle Area began from July 2002 through July 2004, and continues to be ongoing in the Area. In North Carolina's SIP submittal, the State explains that the I/M program was established to achieve additional reductions in NO X emissions, and that while there are incidental VOC emission reductions (approximately 2 tons per day in 2005) as a result of implementing this program in the Triangle Area, the program was not implemented to reduce VOC emissions from motor vehicles. As a result, the existence of this program in the SIP for the purpose of NO X reductions does not prohibit EPA from finding the VOC contribution to 8-hour ozone pollution from motor vehicles insignificant.

After evaluating the information provided by North Carolina and weighing the factors for the insignificance determination outlined in 40 CFR 93.109(k), particularly the biogenic contribution to the overall VOC inventory, EPA is now proposing to approve North Carolina's determination that the VOC contribution from motor vehicle emissions to the 8-hour ozone pollution for the Triangle Area is insignificant. If this finding is completed through the adequacy process (see Section VIII below) or approved through the final rulemaking on this SIP submission, the insignificance determination should be considered and specifically noted in the transportation conformity document that is prepared for this Area.

Under the CAA, states are required to submit, at various times, control strategy SIPs and maintenance plans in ozone areas. These control strategy SIPs (reasonable further progress and attainment demonstration, etc.) and maintenance plans create MVEBs for criteria pollutants and/or their precursors to address pollution from cars and trucks. Per 40 CFR part 93, an MVEB is established for the last year of the maintenance plan. A state may adopt MVEBs for other years as well. The MVEB is the portion of the total allowable emissions in the maintenance demonstration that is allocated to highway and transit vehicle use and emissions. See, 40 CFR 93.101. The MVEB serves as a ceiling on emissions from an area's planned transportation system. The MVEB concept is further explained in the preamble to the November 24, 1993, transportation conformity rule (58 FR 62188). The preamble also describes how to establish the MVEB in the SIP and how to revise the MVEB.

North Carolina, after interagency consultation with the transportation partners for the Triangle Area, has elected to develop county-level subarea MVEBs for NO X . North Carolina is developing these MVEBs, as required, for the last year of its maintenance plan, 2017, and for an additional year, 2008. The MVEBs reflect the total on-road emissions for 2008 and 2017, plus an allocation from the available NO X safety margin for each year. Under 40 CFR 93.101, the term safety margin is the difference between the attainment level (from all sources) and the projected level of emissions (from all sources) in the maintenance plan. The safety margin can be allocated to the transportation sector; however, the total emissions must remain below the attainment level. These MVEBs and allocation from the safety margin were developed in consultation with the transportation partners and were added to account for uncertainties in population growth, changes in model VMT and new emission factor models. The subarea NO X MVEBs for the Triangle Area are defined in Table 6 below.

As mentioned above, North Carolina has chosen to allocate a portion of the available safety margin to the 2008 and 2017 subarea NO X MVEBs. The following table identifies the amount of the NO X safety margin that was allotted to the 2008 and 2017 subarea NO X MVEBs.

The total allocation is 7,741 kg/day (8.53 tpd) in 2008 and 6,049 kg/day (6.67 tpd) in 2017 for NO X . The remaining NO X safety margin after allocation of some of the safety margin to the MVEBs for the Triangle Area is 19.20 tpd in 2008 and 79.39 tpd in 2017.

Through this rulemaking, EPA is proposing to approve the 2008 and 2017 subarea MVEBs for NO X for the Triangle Area because EPA has determined that the Area maintains the 8-hour ozone standard with the emissions at the levels of the budgets. As mentioned above, these MVEBs are subarea MVEBs for each individual county in the Triangle Area. Once the new subarea MVEBs for the Triangle Area (the subject of this rulemaking) are approved or found adequate (whichever is done first), they must be used for future conformity determinations.

As discussed above, the MVEB is the portion of the total allowable emissions in the maintenance demonstration that is allocated to highway and transit vehicle use and emissions. The MVEB concept is further explained in the preamble to the November 24, 1993, transportation conformity rule (58 FR 62188). The preamble also describes how to establish the MVEB in the SIP and how to revise the MVEB. Additionally, the transportation conformity rule (see 93.109(k)) allows for areas not to establish a MVEB for a particular pollutant or precursor if it can be demonstrated that motor vehicle emissions contributions do not significantly contribute to an area's pollution. North Carolina's submittal for this area establishes MVEBs for NO X and provides an insignificance determination for VOC contribution.

Under section 176(c) of the CAA, new transportation projects, such as the construction of new highways, must “conform” to (i.e., be consistent with) the part of the state's air quality plan that addresses pollution from cars and trucks. “Conformity” to the SIP means that transportation activities will not cause new air quality violations, worsen existing violations, or delay timely attainment of the NAAQS. If a transportation plan does not “conform,” most new projects that would expand the capacity of roadways cannot go forward. Regulations at 40 CFR part 93 set forth EPA policy, criteria, and procedures for demonstrating and assuring conformity of such transportation activities to a SIP. The regional emissions analysis is one, but not the only, requirement for implementing transportation conformity. Transportation conformity is a requirement for nonattainment and maintenance areas. Maintenance areas are areas that were previously nonattainment for a particular NAAQS but have since been redesignated to attainment with a maintenance plan for that NAAQS.

When reviewing submitted “control strategy” SIPs or maintenance plans containing MVEBs, EPA must affirmatively find the MVEB contained therein “adequate” for use in determining transportation conformity. Once EPA affirmatively finds the submitted MVEB is adequate for transportation conformity purposes, that MVEB can be used by state and Federal agencies in determining whether proposed transportation projects “conform” to the SIP as required by section 176(c) of the CAA.

EPA's substantive criteria for determining “adequacy” of an MVEB, including EPA's determination that an MVEB need not be established because of an insignificance determination, are set out in 40 CFR 93.118(e)(4). The process for determining “adequacy” consists of three basic steps: public notification of a SIP submission, a public comment period, and EPA's adequacy finding. This process for determining the adequacy of submitted SIP MVEBs was initially outlined in EPA's May 14, 1999, guidance, “Conformity Guidance on Implementation of March 2, 1999, Conformity Court Decision.” This guidance was finalized in the Transportation Conformity Rule Amendments for the “New 8-Hour Ozone and PM2.5 National Ambient Air Quality Standards and Miscellaneous Revisions for Existing Areas; transportation conformity rule amendments—Response to Court Decision and Additional Rule Change,” on July 1, 2004 (69 FR 40004). EPA follows this guidance and rulemaking in making its adequacy determinations. EPA must also use a similar process to determine the adequacy of an insignificance determination that is submitted by a state as a part of a control strategy SIP or maintenance plan. Additional information on the adequacy process for both MVEBs and insignificance determinations is available in the proposed rule entitled, “Transportation Conformity Rule Amendments: Response to Court Decision and Additional Rule Changes,” 68 FR 38974, 38984 (June 30, 2003).

As discussed earlier, North Carolina's maintenance plan submission includes new county-level subarea NO X MVEBs for the Triangle Area for the years 2008 and 2017. Additionally, the maintenance plan included a VOC insignificance determination for the entire Triangle Area, and therefore, no MVEB for VOC is included as part of the SIP revision. EPA reviewed both the NO X MVEBs and the VOC insignificance determination through the adequacy process. The North Carolina SIP submission, including the Triangle subarea NO X MVEBs and the VOC insignificance determination, was open for public comment on EPA's adequacy Web site on March 21, 2007, found at: http://www.epa.gov/otaq/stateresources/transconf/currsips.htm . The EPA public comment period on adequacy of the 2008 and 2017 subarea NO X MVEBs, and VOC insignificance determination closed on April 20, 2007. EPA did not receive any comments on the adequacy of the MVEBs or the VOC insignificance determination, nor did EPA receive any requests for the SIP submittal.

EPA intends to make its determination on the adequacy of the 2008 and 2017 subarea NO X MVEBs, and the VOC insignificance determination for the Triangle Area for transportation conformity purposes in the final rulemaking on the redesignation of the Triangle Area. If EPA finds the 2008 and 2017 subarea NO X MVEBs, and the VOC insignificance determination adequate or approves these MVEBs and the VOC insignificance determination in the final rulemaking action, the new MVEBs for NO X must be used, and the VOC insignificance determination should be noted, for future transportation conformity determinations. If the new 2008 and 2017 subarea NO X MVEBs are found adequate, and both the NO X MVEBs and the related VOC insignificance determination are approved in the final rulemaking, the NO X MVEBs and the VOC insignificance determination will be effective on the date of publication of EPA's final rulemaking in the Federal Register . For required regional emissions analysis years that involve the year 2016 or before, the applicable budget for the purposes of conducting transportation conformity will be the new 2008 subarea NO X MVEBs for the Triangle Area. For required regional emissions analysis years that involve 2017 or beyond, the applicable budgets will be the new 2017 subarea NO X MVEBs. Both the 2008 and 2017 subarea NO X MVEBs are defined in section VII of this proposed rulemaking. More detail on the VOC insignificance determination can be found in section VII of this proposed rulemaking as well.

EPA is proposing to make the determination that the Triangle Area has met the criteria for redesignation from nonattainment to attainment for the 8-hour ozone NAAQS. Further, EPA is proposing to approve North Carolina's June 7, 2006, SIP submittal including the redesignation request for the Triangle Area. EPA believes that the redesignation request and monitoring data demonstrate that the Triangle Area has attained, and will continue to maintain the 8-hour ozone standard.

EPA is also proposing to approve the maintenance plan for the Triangle Area included as part of the June 7, 2006, SIP revision. The maintenance plan includes subarea NO X MVEBs for 2008 and 2017, and a VOC insignificance determination for motor vehicles' contribution to the ozone pollution in this Area, among other requirements. EPA is proposing to approve the 2008 and 2017 subarea NO X MVEBs for the Triangle Area because the maintenance plan demonstrates that in light of expected emissions for all other source categories, the Triangle Area will continue to maintain the 8-hour ozone standard. EPA is also proposing to approve the insignificance determination for the VOC contribution from motor vehicle emissions to the 8-hour ozone pollution for the Triangle Area.

Further as part of today's action, EPA is describing the status of its adequacy determination for the 2008 and 2017 subarea NO X MVEBs, and VOC insignificance determination, in accordance with 40 CFR 93.118(f)(1). Within 24 months from the effective date of EPA's adequacy finding for the MVEBs, or the publication date for the final rule for this action, the transportation partners will need to demonstrate conformity to the new subarea NO X MVEBs pursuant to 40 CFR 93.104(e) as effectively amended by section 172(c)(2)(E) of the CAA as added by the Safe, Accountable, Flexible, Efficient Transportation Equity Act—A Legacy for Users (SAFETEA-LU), which was signed into law on August 10, 2005. Additionally, the transportation partners should note EPA's finding of adequacy and approval for the VOC insignificance determination for future conformity determinations.

Under Executive Order 12866 (58 FR 51735, October 4, 1993), this proposed action is not a “significant regulatory action” and therefore is not subject to review by the Office of Management and Budget. For this reason, this action is also not subject to Executive Order 13211, “Actions Concerning Regulations That Significantly Affect Energy Supply, Distribution, or Use” (66 FR 28355, May 22, 2001). This proposed action merely proposes to approve state law as meeting Federal requirements and imposes no additional requirements beyond those imposed by state law. Redesignation of an area to attainment under section 107(d)(3)(e) of the CAA does not impose any new requirements on small entities. Redesignation is an action that affects the status of a geographical area and does not impose any new regulatory requirements on sources. Accordingly, the Administrator certifies that this proposed rule will not have a significant economic impact on a substantial number of small entities under the Regulatory Flexibility Act (5 U.S.C. 601 et seq. ). Because this rule proposes to approve pre-existing requirements under state law and does not impose any additional enforceable duty beyond that required by state law, it does not contain any unfunded mandate or significantly or uniquely affect small governments, as described in the Unfunded Mandates Reform Act of 1995 (Public Law 104-4).

This proposed rule also does not have tribal implications because it will not have a substantial direct effect on one or more Indian tribes, on the relationship between the Federal Government and Indian tribes, or on the distribution of power and responsibilities between the Federal Government and Indian tribes, as specified by Executive Order 13175 (65 FR 67249, November 9, 2000). This action also does not have Federalism implications because it does not have substantial direct effects on the states, on the relationship between the national government and the states, or on the distribution of power and responsibilities among the various levels of government, as specified in Executive Order 13132 (64 FR 43255, August 10, 1999). This action merely affects the status of a geographical area, does not impose any new requirements on sources, or allow a state to avoid adopting or implementing other requirements and does not alter the relationship or the distribution of power and responsibilities established in the CAA. This proposed rule also is not subject to Executive Order 13045 “Protection of Children from Environmental Health Risks and Safety Risks” (62 FR 19885, April 23, 1997); because it is not economically significant and because the Agency does not have reason to believe that the rule concerns an environmental health risk or safety risk that may disproportionately affect children.

In reviewing SIP submissions, EPA's role is to approve state choices, provided that they meet the criteria of the CAA. In this context, in the absence of a prior existing requirement for the State to use voluntary consensus standards (VCS), EPA has no authority to disapprove a SIP submission for failure to use VCS. It would thus be inconsistent with applicable law for EPA, when it reviews a SIP submission; to use VCS in place of a SIP submission that otherwise satisfies the provisions of the CAA. Redesignation is an action that affects the status of a geographical area but does not impose any new requirements on sources. Thus, the requirements of section 12(d) of the National Technology Transfer and Advancement Act of 1995 (15 U.S.C. 272 note) do not apply. This proposed rule does not impose an information collection burden under the provisions of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq. ).

You may be potentially affected by this action if you are an agricultural producer, food manufacturer, or pesticide manufacturer. Potentially affected entities may include, but are not limited to:

• Crop production (NAICS code 111), e.g., agricultural workers; greenhouse, nursery, and floriculture workers; farmers.

• Animal production (NAICS code 112), e.g., cattle ranchers and farmers, dairy cattle farmers, livestock farmers.

• Food manufacturing (NAICS code 311), e.g., agricultural workers; farmers; greenhouse, nursery, and floriculture workers; ranchers; pesticide applicators.

• Pesticide manufacturing (NAICS code 32532), e.g., agricultural workers; commercial applicators; farmers; greenhouse, nursery, and floriculture workers; residential users.

This listing is not intended to be exhaustive, but rather provides a guide for readers regarding entities likely to be affected by this action. Other types of entities not listed in this unit could also be affected. The North American Industrial Classification System (NAICS) codes have been provided to assist you and others in determining whether this action might apply to certain entities. If you have any questions regarding the applicability of this action to a particular entity, consult the person listed under FOR FURTHER INFORMATION CONTACT .

1. Docket . EPA has established a docket for this action under docket identification (ID) number EPA-HQ-OPP-2007-0546. Publicly available docket materials are available either in the electronic docket at http://www.regulations.gov , or, if only available in hard copy, at the Office of Pesticide Programs (OPP) Regulatory Public Docket in Rm. S-4400, One Potomac Yard (South Bldg.), 2777 S. Crystal Dr., Arlington, VA. The hours of operation of this Docket Facility are from 8:30 a.m. to 4 p.m., Monday through Friday, excluding legal holidays. The Docket Facility telephone number is (703) 305-5805.

The Atlantic HMS fisheries are managed under the dual authority of the Magnuson-Stevens Fishery Conservation and Management Act (Magnuson-Stevens Act) and the Atlantic Tunas Convention Act (ATCA). The Consolidated HMS FMP, finalized in 2006, and amendments to that FMP are implemented by regulations at 50 CFR part 635.

On July 27, 2007 (72 FR 41392), NMFS published a proposed rule that requested comments on the draft Amendment 2 to the Consolidated HMS FMP, and scheduled 10 public hearings throughout August and September 2007 to receive comments from fishery participants and other members of the public regarding the proposed rule and draft Amendment 2 to the Consolidated HMS FMP. NMFS has since received many requests to extend the comment period in order to allow for more adequate comment submissions. In order to accommodate these requests and to provide additional opportunities for public comment by constituents, NMFS is extending the public comment period on the proposed rule and draft Amendment 2 to the Consolidated HMS FMP to 5 p.m., November 2, 2007.